ML19269F007

Suppl 1 SER Re Const of Facility

ML19269F007
Person / Time
Site:	South Texas
Issue date:	10/29/1975
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-75-075, NUREG-75-075-S01, NUREG-75-75, NUREG-75-75-S1, NUDOCS 7911140436
Download: ML19269F007 (75)

Text

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• 8'" 5 -488 and STN 50-499 Houston Lighting and Power Company, et al October 1975 Supplement No.1 2183 232 e

G 3911140

Available from liational Technical Information Service Springfield, Virginia 22161 Price: Printed Copy $4.75 ; Microfiche $2.25 2183 233

OCTOBER 29, 1975 NUREG-75/075 SUPPLEMENT NO. 1 TO THE SAFETY EVALUATION REPORT BY THE DIVISION OF REACTOR LICENSING U. S. NUCLEAR REGULATORY COMMISSION IN THE MATTER OF HOUSTON LIGHTING & POWER COMPANY, ET AL SOUTH TEXAS PROJECT UNITS 1 AND 2 DOCKET NOS. STN 50-498 AND STN 50-499 2183 234

TABLE OF CONTENTS Page

1.0 INTRODUCTION

AND GENERAL DISCUSSION 1-1 1.1 Introduction............... 1-1 1.9 Treatment of Interfaces with the RESAR-41 Nuclear Steam Supply System Standard Design..... 1-2 2.0 SITE CHARACTERISTICS. 2-1 2.2 Nearby Industrial, Transportation, and Military Facilities.... 2-1 2.5 Geology and Seismology 2-1 2.5.2 Site Geology 2-1 2.5.5 Foundation Engii.:ering... 2-1 3.0 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS AND C0t'PONENTS. 3-1 3.5 Missile Protection. 3-1 3.5.1 Missile Protection Criteria 3-1 3.11 Environmental Design of Electrical Equipment. 3-1 6.0 ENGINEERED SAFETY FEAldRES. 6-1 6.3 Emergency Core Cooling Systems 6-1 6.5 Control Room Habitability. 6-5 7.0 INSTRUMENTATION AND CONTROLS. 7-1 7.1 Introduction.. 7-1 7.1.2 Desgin Criteria 7-1 7.2 Reactor Trip System. 7-1 7.3 Engineered Safety Features Actuation Systerrs 7-2 7.3.1 Control Room Heating, Ventilation and Air Conditioning System. 7-2 7.3.4 Motor Operated Valves in the Emergency Core Cooling System. 7-2 7.3.5 Emergency Boration System.. 7-3 iO SystemsRequired'forSafeShutdown 7.4 7-4

Page 7-5 7.5 Safety Related Display Instrumentation 7.5.1 Monitoring of Heating, Ventilation and Air Conditioning 7-5 Systems.. 7-5 7.8 Anticipated Transients Without Scram... 8-1 8.0 ELECTRIC POWER.. 8.3 On Site Power System... 8-1 8.3.3 Thermal Overload Protection for Engineered Safety Features Motor Operated Valves. 8-1 9-1 9.0 AUXILIARY SYSTEMS.... 9.4 Heating, Ventilation, and Air Conditioning Systems. 9-1 9.4.1 Contr.1 Room and Eletrical Auxiliary tuilding Heating, Ventilation, and Air Conditioning System. 9-1 9.4.3 Fuel Handling Building Heating, Ventilation, and Air Conditioning 91 System 9.5 Other Auxiliary Systems. 9-2 9.5.1 Fire Protection System.. 9-2 11.0 PADI0 ACTIVE WASTE 'W4AbtMENT. 11-1 11.0 Surrinary Description.. 11-1 14.0 INITIAL TESTS AND OPERATIONS.... 14-1 17-1 17.0 QUALITY ASSURANCE..... 17.3 Brown & Root, Inc.. 17-1 18.0 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS................... 18-1 20-1 20.0 FINANCIAL QUALIFICATIONS.......... 21 -1

21.0 CONCLUSION

S..,.8....................2183236 i'

APPENDICES PAGE APPENDIX A UPDATED SUPPLEMENTARY INFORMATION FOR APPENDIX n.o TPE A-1 SAFETY EVALUATION REP 0kT " REPORT TO THE ADVISORY COMMITTEE ON REACTfR SAFEGUARDS BY THE OFFICE OF NUCLEAR Rete, TOR REGULATION - U. S. NUCLEAR REGULATORY COMMISSION IN THE MATTER OF WESTINGHOUSE ELECTRIC CORPORATION REFERENCE SAFETY ANALYSIS REPORT - RESAR DOCKET NO. STN 50-480" APPENDIX B CONTINUATION OF CHRON0 LOGY OF RADIOLOGICAL REVIEW OF SOUTH 8-1 TEXAS PROJECT UNITS 1 AND 2 APPENDIX C REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS, C-1 DATED SEPTEMBER 19, 1975 APPENDIX D MINIMUM CONTAINMENT PRESSURE MODEL FOR PWR ECCS PERFORMANCE D-1 EVALUATION APPENDIX E ANALYSIS OF FINANCIAL QUALIFICATIONS E-1 APPENDIX F ERRATA - SOUTH TEXAS PROJECT F-1 APPENDIX G NRC STAFF REVIEW OF THE WESTINGHOUSE EPERGENCY CORE G-1 COOLING SYSTEM EVALUATION t0 DEL E

LIST OF FIGURES Page FIGURE 17.2 EROWN & ROOT, INC. ORGANIZATION.. 17-4 FIGURE 17.3 BROWN & ROOT, INC. QUALITY ASSURANCE / QUALITY COTITROL (QA/QC) ORGANIZATION 17-5 2183 238

1.0 INTRODUCTION

AND GENERAL DISCUSSION 1.1 Introduction The Nuclear Regulatory Comission's (Comission) Safety Evaluation Report in the matter of "the application by the Houston Lighting & Power Company, the City Public Service of San Antonio, the Central Power and Light Company, and the City of Austin (hereinafter referred to as the applicants) to construct and operate the proposed South Texas Project Units 1 and 2 was issued on August 1, 1975. In this Safety Evaluation Report the staff indicated (1) certain matters requiring additional infomation from the applicants, (2) certain matters where our review is not yet complete and (3) certain comitments made by the applicants for which additional documentation would be required to pemit the staff to confim that these comitments meet our requirements. The purpose of this Supplement is to update the Safety Evaluation Report by providing the staff's evaluation of additional infome'; ion submitted by the applicants since the issuance of the Safety Evaluation Report, and to address the comments made by the Advisory Comittee on Reactor Safeguards in its report of September 19, 1975. In addition, a review of the Safety Evaluation Report has revealed areas where corrections or further explanations are in order. Each of the following sections in this Supplement is numbered the same as the section of the Safety Evaluation Report that is being updated and is supplementary to and not in lieu of the discussion in the Safety Evaluation Report. Also included in this Supplement as Appendix A is additional information which will update Appendix A of the Safety Evaluation Report " Report to the Advisory Comittee on Reactor Safeguards by the Office of Nuclear Reactor Regulation. U. S. Nuclear Regulatory Commission - In the Matter of Westinghouse Electric Corporation Reference Safety Analysis Report, RESAR-41 Docket No. STN 50-480". Appendix B to the Supplement is a continuation of the chronology of the staff's principal actions related to the processing of the application. The South Texas Project Report of the Advisory Committee on Reactor Safeguards is attached as Appendix C. The staff's technical paper " Minimum Containment Pressure Model for PWR ECCS Performance Evaluation" is attached as Appendix D and the " Analysis of Financial Qualifications" is attached as Appendix E. Appendix F is a listing of errata to the Safety Evaluc+. ion Report and the staff's technical paper "NRC Staff Review of the Westiphrase Emergency Core Cooling System Evaluation Model" ' s attached'as Appendix G. i ~ ' 2183 239 1-1

1.9 Treatment of Interfaces with the RESAR-41 Nuclear Steam Supply System Standard Design for standard nuclear steam supply system design applications, the Commission staff has been working to identify and define the type of interface infomation that should be provided to adequately assess the compatibility of a balance-of-plant design that references the standard nuclear steam supply system design. We recently requested additional inforCion regarding interfaces from Westinghouse. We will review this infomath prior to issuance of the Preliminary Design Approval for RESAR-41 and will report the results of our review in the RESAR-41 Safety Evaluation Report. Considerable progress has been made in this regard and we fully expect to completely resolve this matter so that future applications referencing RESAR-41 will be able to address the interface matter in an acceptable manner. However, we have recognized that for the near tem our objective of resolving this interface matter on the standard nuclear steam supply system designs could not be completed satisfactorily to preclude a schedule slippage of utility applications that currently reference the standard nuclear steam supply system designs, and are in the latter stage of the staff's construction pemit application review. Therefore, for the South Texas Project, which references the RESAR-41 nuclear steam supply system design, we decided to coreplete our review of areat of concern related to interface matters on the South Texas Project indepenoently of the final resolution of the RESAR-41 interfaces for issuance of a Preliminiry Design Approval. That is, we rev ewed South Texas Project and RESAR-41 as a custom plant application to confirm that the South Texas Project balance-of plant portion is compatible with the standard RESAR-41 design. We believe that ttis approach is practical since the South Texas Project application presents plant design infomation that is currently being reviewed with the review of RESAR-4i. It is our view that this approach is not very different than that we have used in the past in our reviews of custom plant applications. Using the above cited approach, we have been able to identify those areas of concern relating to the compatibility of the South Texas Froject balance-of-plant design to RESAR-41. The applicants were infomed of thcse areas. In amendments to the Preliminary Safety Analysis Repert, the applicants have provided infomation to confirm that the South Texas Project design is compatible with the standard RESAR-41 design. We have reviewed this information and the South Texas Project and RESAR-41 applications, and conclude that the Duth Texas Project balance-of-plant is compatible with the RESAR-41 nuclear steam supply system. 2183 240 m ttili 1-2

2.0 SITE CHARACTERISTICS 2.2 Nearby Industrial, Transportation, and Military Facilities We stated in the Safety Evaluation Report that a low level military airway (08-19) passes over the site area and that the U. S. Air Force had indicated in a letter to the applicants that the airway would be modified to assure a minimum clearance distance of five miles from the site. In Amendment 22 of the Preliminary Safety Analysis Report, the applicants state that flight route 09-19 was cancelled as of January 30, 1975. Therefore, on the basis of the information provided by the applicants, we have concluded that militt.ry flight on Flight 08-19 will not affect the safe operation of the facility. 2.5 Geoloqv and Seismology 2.5.2 Site Geology We stated in the Safety Evaluation Report (Sections 2.5.2 and 2.5.5) that we had not completed our review of the subsidence monitoring program. Our review of the applicants' subsidence monitoring program has now been completed. This program includes an instrumentation description, layout of the horizontal and vertical movement detectors as well as ground water instrumentation installation and frequency of observations. The results of the subsidence monitoring program will be submitted for our evaluation in the Final Safety Analysis Report. We have determined that this monitoring program will provide a satisfactory method of detecting subsidence at the proposed South Texas Project site and is, therefore, acceptable. 2.5.5 Foundation Engineerir,q We stated in the Safety Evaluation Report that we would report the results of our review of the applicants' design criteria for long term settlement, differential settlement and tilt of safety related structures. The applicants have presented these criteria in Section 2.5.4.13.1 of the Preliminary Safety Analysis Report (Amendment 24). We have completed our review of these criteria and find thcm acceptable for the foundation materials at the proposed South Texas Project site. On the basis of the design criteria and the settlement monitoring program that will be performed, there is reasonable assurance that the long term settlement of safety-related plant facilities will not cause a hazard to these facilities. 2183 241 2-1

3.0 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS AND COMPONENTS 3.5 Missile Protection 3.5.1 Missile Protection Criteria We stated in the Safety Evaluation Report that we would require that the South Texas Project be designed to withstand the impacts of missiles and impact velocities listed in Table 3.1 or Table 3.2 of the Safety Evaluation Report. The applicants have confirTned in Amendment 28 of the Preliminary Safety Analysis Report their intention to design the South Texas Project to withstand the tornado missiles and impact velocities that we set forth in Table 3.1 of the Safety Evaluation Report. We have reviewed this information and it satisfies our requirement for an acceptable spectrum of tornado missiles and velocities to be used as a design basis for the South Texas Project. Therefore, we have concluded that the proposed design is acceptable. 3.11 Environental Design of Electrical Equipment We stated in the Safety Evaluation Report that we required that all Class IE equipment be qualified to satisfy the requirements of the Institute of Electrical and Electronics Engineers (IEEE) Standard 323, 1974. In Amendment 28 of the PSAR, the applicants have agreed to conform with this standard. We have concluded that this cortnitment is acceptable. 2183 242 3-1

6.0 ENGINEERED SAFETY FEATURES 6.3 Emergency Core Cooling Systems We stated in Section 6.3.5 of Appendix A to the Safety Evaluation Report that Westinghouse has comitted to provide an analysis which satisfies the require-ments of 10 CFR Part 50.46, the emergency core cooling system Final Acceptance Criteria. Westinghouse has submitted an evaluation of the emergency core cooling system performance in Amendments 18 and 19 to the RESAR-41 Preliminary Safety Analysis Report, pursuant to the requirements of the Commission's regulations,10 CFR Part 50.46. Westinghos,e's submittal addressed our concerns related to the applicability of the approved Westinghouse Evaluation Model to the analysis of the RESAR-41 emergency core cooling system design. Our review of the Westinghouse evaluation model is attached as Appendix G. Westinghouse provided justification for the evaluation model representation of the inherent asymetric geometry (three trains of the emergency core cooling system will be injected at different flow rates into the four cold legs of the reactor coolant system) of the RESAR-41 emergency core cooling system design, the use of the steam cc sling model, and the identification of the most limiting power distribution. Westinghouse submitted small and large break loss-of-coolant accident analyses in Amendments 18 and 19 to AESAR-41. The large break loss-of-coolant accident analysis was limited to a spectrum of four double-ended guillotine breaks with discharge coefficients of 1.0, 0.8, 0.6 and 0.4, and a double-ended split break with a discharge coefficient of 1.0. To supplement the analysis, Westinghouse submitted Topical Report WCAP-8565-P " Westinghouse ECCS Four Loop Plant (17 x 17) Sensitivity Studies" which covered other break sizes, types and locations and demonstrated that the guillotine breaks stre the worst cases for this type plant. The analyses submitted by Westinghouse identified the worst break as the double-ended cold leg guillotine break with a Moody maltiplier of 0.6. The calculated peak clad temperature was 1912 degrees Fahrenheit which is within the acceptable limit of 2200 degrees Fahrenheit as specified in 10 CFR Part 50.46(b). In addition, the maximum local metal / water reaction of 3.02 percent, whic5 was recorded for the 0.8 double-ended cold leg guillotine, and total core wide metal / water reaction of less than 0.3 percent were well below the allowable limits of 17 percent and one percent, respectively. The analyses were perfonned ^ based 'on an{as.sumed total peaking factor of 2.45, 102 percent of the rated nuclear steam supply system power level of 3800 megawatts thermal, and peak linear power density of 12.65 kilowatts per foot. 2183 243 6-1

The small break loss-of-coolant accident analysis included a five break spectrum specific to RESAR-41 and referenced Westinghouse Topical Report WCAP-8356 " Westinghouse Emergency Core Cooling System - Plant Sensitivity Studies". The six inch diameter pipe break was identified as the limiting small break with a calculated peak clad temperature of 1432 degrees Fahrenheit. This clearly indicated that the small break loss-of-coolant accident was not the limiting case. Westinghouse submitted in Amendment 19 of RESAR-41, responses to our concerns regarding the applicability of the approved Westinghouse evaluation model to the analysis of RESAR-41. The areas addressed included the geometric effects of asymetry of the emergency core cooling system design, the longer core length, the identification of the most limiting power distribution, and the core heat transfer during steam cooling. The inherent asymetry of the RESAR-41 emergency core cooling system design provides a nonunifonn distribution of emergency core cooling to the intact cold legs. We revicwed the evaluation model representation of the asymetric RESAR-41 emergency core cooling system design and found it acceptable based on: (1) the Westinghouse evalvation model representation conservatively assumed symetric emergency core cooling injection which, during blowdown, extended the time to end of bypass and caused an additional quantity of emergency core cooling to be spilled out the break, and (2) during reflood, a penalty for steam water mixing was taken in all three intact loops instead of only two (this is equivalent to about a two percent penalty in the flooding rate). Westinghouse provided a sensitivity study to justify that the assumed cosine axial power distribution represented the most limiting power shape for RESAR 41. Considering the longer core, the cosine shape when compared to skewed power shapes peaked at the top of the core, remained the most limiting case. The cosine power shape for the 14-foot core is assumed to be equivalent to a power shape skewed to seven feet on a 12-foot core. Since it was more appropriate to model the RESAR-41 core using a skewed power methodology (previously approved as part of the Westinghouse emergency core cooling system evaluation model) this method was used,to define an equivalent 12-foot power shape for purposes of interpreting the Full Length Emergency Cooling Heat Transfer (FLECHT) test results. This approach results in less favorable heat transfer and a degradation in heat flux during the steam cooling period. A complete rear.alysis of the guillotine breaks was submitted in Amendment 19 of RESAR-41 which showed the expected peak node clad temperature rise, when the reflood rate dropped below 1.0 inch per second. 2183 244 1 6-2

We stated in Section 6.3 of the Safety Evaluation Report, that we required an analysis to determine the minimum containment pressure in accordance with the requirements of Appendix K to 10 CFR Part 50. Appendix K to 10 CFR Part 50 of the Comission's regulations requires that the effect of operation of all the inAlled pressure reducing systems and processes be included in the emergency core cooling system evaluation. For this evaluation, it is conservative to minimize the containment nressure since, this will increase the resistance to steam flow in the reactor coolant loops and reduce the reflood rate in the core. Following a postulated loss-of-coolant accident, the pressure in the containment building will be increased by the addition of steam and water from the primary reactor system into the containment atmosphere. After initial blowdown, heat transfer from the core, primary metal structures and steam generators to the emergency core cooling system water will produce additional steam. This steam together with any emergency core cooling system water becomes energy that is released from the primary system postulated b"=k to the contain-ment during both the blowdown and later emergency core cooling vstem operational phases; i.e., reflood and post-reflood phases. Energy removal within the containment occurs by several means. Steam condensa-tion on tne containment walls and internal structures serves as a passive energy heat sink that becomes effective early in the blowdown transient. Subsequently, the operation of the containment heat removal systems such as containment sprays and fan coolers will roove energy from the containment atmosphere. When the energy removal rate exceeds the rate of eners addition from the primary system, the containment pressure will decrease from its peak value. The emergency core cooling system containment pressure calcula-tions for the South Texas Project Units 1 and 2 were made with the Westinghouse emergency core cooling system evaluation model. We reviewed the Westinghouse model and issued a Status Report on October 15, 1974. which was anended November 13, 1974. We had concluded that Westinghouse's containment pressure model was acceptable for the emergency core cooling system evaluation. We required, however, that justification of.he plant-dependent contain1ent input parameters used in the analysis be submitted for our review of each plant. The containment input data relating to containment net-free volume, passive heat sinks and containment heat removal systems were submitted for South Texas Project Units 1 and 2 in Amendment 30 of the Preliminary Safety Analysis Report. The containment input data were also submicted in Amendments 18 and 19 of RESAR-41, which is the reference plant for the South Texas Project Units 1 and 2. We have compared these two sets of containment input data and found that the data used by the South Texas Project, Units 1 and 2 are conservative for the emergency core cooling system containment pressure analysis. We find that the data for the p'assive, heat sinks are conservative in comparison with our recomendations contained in the staff technical paper. 'Pinimum Containunt Pressure tiodel for PWR ECCS Performance Evaluation", included as Appendix D to this supplement. The passive heat sink data are based on measurements within the. containment of similar nuclear plants. 2183 245 6-3

Based upon the data submitted by the applicants, we calculated the containment pressure by using the CONTEMPT computer code and found that the minimum containment pressure for the Sauth Texas Project Units 1 and 2 is conservative for the emergency core cooli9g system evaluation. We have concluded that the plant-dependent infomation used for the analysis to determine the minimum emergency core cooling system containment pressure following a postulated loss-of-coolant accident for the South Texas Project Units 1 and 2 is conservative. Therefore, the analysis is acceptable for use in the evaluation of emergency core cooling system performance. Appendix K to 10 CFR Part 50 of the Commission's regulations also requires that the combination of emergency core cooling subsystems to t,e assumed operative shall be those available after the most severe single failure of emergency core cooling systems equipment has occurred. The worst single failure was identified by Westinghouse as the loss of a low head safety injection pump, which provided within a consistent set of assumptions (1) the maximum contain-ment cooling and a reduction in emergen y core cooling flow, and (2) the maximum calculated peak clad temperattre. A review oC the RESAR-41 piping and instrumentation diagrams has indicated that spurious actuation of specific motor operated valves were considered in the selection of the worst single failure. We have concluded that the emergency core cooling system perfornance will be adecuate in the event of any postulated failure of a single active component. We have also reviewed the proposed procedures and the system design for preventing excessive boric acid buildup in the reactor vessel during the post loss-of-coolant accident, long-term coeling period, and have concluded that switchover time from cold to simultaneous hot and cold leg injection must be changed from 24 hours to 17.5 hours after a loss-of-coolant accident. This change is required to assure that, in the event of a cold leg break, the concentration of the boric acid in the core region does not exceed the solubility limits. We also require that in the case where only two subsystems are available, they should be aligned in such a manner that one subsystem injects into the bot leg and the other into the cold leg. This arrangement would assure that even in the case of a hot leg break, sufficient flow through the core is provided. We will require during the operating license stage of review that the South Texas Project Units 1 and 2 meet the above stated requirements. c h um 2183 246 6-4

We have reviewed the instrumentation and piping diagrams and found that the proposed emergency core cooling system can be operated in a manner complying with the single failure criterion. We will require that in the South Texas Project Units 1 and 2 Final Safety Analysis Report, the applicants use actual values for the emergency core cooling system piping flow resistances. emergency core cooling system and reactor coolant system volumes, and residual heat removal system piping flow resistance. In adition, the effects of rod-to-rod bowing will be considered in the development of the Technical Specifications for the nuclear peaking factors. To prevent water hamer, we will also require that venting provisions be described in the Final Safety Analysis Report for the emergency core cooling fill system. The emergency operating procedures will also be reviewed during the operating license stage of review. On the basis of our review of the information submitted by Westinghouse and the applicants, we have concluded that (1) the loss-of-coolant analyses that were performed conservatively represent the South Texas Project Units 1 and 2 design gRESAR-41) and are wholly in conformance with the reouirements of Appendix K to 10 CFR Part 50, (2) the emergency core cooling system performance conforms to the peak clad temperature and maximum oxidation and hydrogen generation criteria of 10 CFR Part 50.46, (3) the emergency core cooling system performance will be 3dequate in the event of any postulated failure of a single component and (4) adequate systems are available to provide long term core cooling. Therefore, we have concluded that the design of the South Texas Project Units 1 and 2 emergency core cooling system is acceptable. 5.5 Control Room Habitability We stated in the Safety Evaluation Report (Sections 6.5 and 7.3.1) that wr would require retention of the provision in the South Texas Project desic i for initiation of the auto,aatic isolation of the control room ventilation system by a signal from radiation detectors located within the outside air intakes for the control room. The applicants have documented their coroliance with cur requirement in Amendnent 28 of the Preliminary Safety Analysis Report. Therefore, we have concluded that with this comitment the design of the control room habitability system is acceptable. 2183 247 6-5

7.0 INSTRUMENTATION AND CONTROLS 7.1 Introduction 7.1.2 Design Criteria We stated in the Safety Evaluation Report that the applicants had agreed to revise the South Texas Project Preliminary Safety Analysis Report to be consistent with the criteria specified in RESAR-41. The applicants have presented Table 7.1-1 in the Preliminary Safety Analysis Report which agrees with the comparable table in Amendment 17 of RESAR-41. This table specifies the criteria for all instrumentation and control systems, including those which interface tatween the nuclear steam supply system furnished equipment and the balance-of-plant. We have determined after a review of Table 7.1-1 that this portion of the design satisfies the Comission's require-ments for a construction pemit, and is therefore acceptable. With respect to the design of electrical penetrations the applicants have comitted to comply with the recomendations of Regulatory Guide 1.63. We find this comitment to be acceptable. However, to assure proper implementa-tion of this requirement we recomend that prior to completion of the final design the applicants discuss the design of electrical penetrations with the s taf f. 7.2 Reactor Trip System We stated in the Safety Evaluation Report that the applicants will provide the reactor coolant pump undervoltage and underfrequency trips as inputs to the reactor trip system. The applicants comitted to meet the requirements of RESAR-41. However, Westinghouse has not specifically stipulated that these reactor trip inputs should confom to all criteria applicable to the pmtection system. Therefore, in Amendment 30 of the Preliminary Safety Analysis Report, the applicants comitted to meet all the requirements of the Institute of Electrical and Electronics Engineers (IEEE) Standard 279-1971 in the design of these reactor trip inputs and has also provided a preliminary design for our review. As a result of our review, we have detemined that this preliminary design presents an acceptable way to satisfy this standard and is therefore acceptable. 2183 248 1 7-1

We also stated in the Safety Evaluation Report that the applicants' implementa-tion of current RESAR-41 underfrequency trip design features was inadequate and unacceptable. RESAR-41 has since modified the desig1 and now provides a curve which defines the interface requirements regarding tre frequency decay rate. The applicants' design satisfies these 'nodified interface requirements of RES AR-41. We have determined that this design is now acceptable for the construction pemit stage of review. Mc.vever, we have not completed our generic review of the basis for the RESAR-41 interface regarding this matter. Upon completion of our review, if it is detemined that design changes are required, we will require the applicants to implement these changes in the South Texas Project design. 7.3 Engineered Safety Features Actuation Systems

7. 3.1 Control Room Heatina, Ventilation and Air Coaditioning System We stated in the Safety Evaluation Report that the applicants have not included the capability of automatically initiating the control room heating, ventilation and air conditioning system by the control room radiation monitors.

In Amendment 28 of the Preliminary Safety Analysis Report, the applicants have modified the design of this system to include automatic initiation by the control room radiation monitors sensing high radioactivity in the outside air intakes. As discussed in Section 6.5 of this Supplement, the applicants have confomed with our requirements. Therefore, we have concluded that the design is acceptable. 7.3.4 Motor Operated Valves in the Emergency Core Cooling System We stated in Section 7.3.1 of Appendix A to the Safety Evaluation Report that we have concerns with respect to the interface criteria for motor operated valves in the emergency core cooling systems. Westinghouse has identified nine mctor-operated valves in the emergency core cooling system design which should not move from normal alignment durino certain phases of a postulated loss-of-coolant accident. We had detemined that the design would be unacceptable in those instances in which a single failure in the electric system would result in the loss-of-capability to perfom the specified function. As a result Westinghouse agreed to remove power to the valve operators in lieu of corrective design changes. tB!' 2183 249 7-2

We indicated to the applicants that the following criteria must be included in their design: (1) Electrical operated valves that are classified as " active" valves, i.e., are required to open or close in various safety system operational sequences, but are manually controlled, shall have the capability of electrical power restoration from the main control room withi-the time period required for these valves. (2) South Texas Project Units 1 and 2 Technical Specifications will require proper positioning of these nine valves and the locking out of power to them prior to the reactor being brought critical. (3) Redundant position indication will be provided for all nine valves and the position indication system will, itself, meet the single failure criterion. In Amendment 29 of the Preliminary Safety Analysis Report, the applicants have committed to the above criteria for six of the valves identified in the RESAR-41 review. The three accumulator valves are not considered to be active valves and do not require the capability of power restoration from the control room. Hcwever, the applicants stated that they will meet the remaining criteria identified above. We have reviewed the commitment presented by the applicants and have concluded that it will be an acceptable way of meeting the single failure criterion for the emergency core cooling system design. On this basis, we have concluded that the design of the emergency core cooling system valves is a cceptable. 7.3.5 Emergency Boration System In.Section 7.3.4 of Appendix A of the Safety Evaluation Report we discussed the fact that the power for the redundant heat tracing systems will be supplied from the redundant engineered safety feature busses. Therefore, we had concerns wiLh regard to terminating redundant engineered safety features power sources at single components like a common pipe or valve. Such a design might result in the compromise of the physical and electrical independence required between the plant redundant engineered safety features power sources. In addition, we also stated that it was not clear from the infomation provided how a single temperature monitoring system will provide reliable intelligence to the operator on emergency boration fluid temperature and how such a design will meet the single failure criterion requirements for safety systems. 2183 250 [ \\..n LOI3 ' ~. 7-3

In Amendment 31 of the Preliminary Safety Analysis Report, the applicants stated that electrical independence of redundant engineered safety features power sources at comon equipment will be preserved by providing sufficient physical separation between heat tracing elements energized by redundant engineered saf,ety features power sources to prevent damage to ora circuit by a short circuit or ground fault in a redundant circuit. Further, the applicants stated if it is not possible to provide sufficient physical separa-tion in congested spaces, one or each of the redundant heat tracing elements will be sheathed in a metallic sheath or braided metallic shield. The applicants stated that the design of the heat tracing will comply with the requirements of Regulatory Guide 1.75. The applicants also stated that as a further precaution, only one of the two redundant heat tracing systems will normally be energized at a time. During an accident, both heat tracing systems will be shed from the supply buses by the accident signal. We requested that the applicants provide a design feature or perfom a test that would assure that during switching from one system to the other, a fault which may exist on one system would not be transferred over to the redundant system. In a letter dated October 22, 1975, the applicants have comitted to the above requirements. We find this comitment acceptable. With respect to the temperature monitors, the applicants will provide redundant, physically separated temperature monitors powered from independent and physically separated power sources. On the basis of our review of these systems, we required the applicants to provide a comitment to the requirements of IEEE Std. 279-1971. The applicants have presented the design criteria for these monitors. We have reviewed this infomation and have detemined that the applicable sections of IEEE Std. 279-1971 will be satisfied. Therefore, we have concluded that the temperature monitoring system is acceptable. Based on the above, we have concluded that the proposo design of the emergency boration system for South Texas Project Units 1 and 2 is acceptable. 7.4 Systems Required for Safe Shutdown We stated in the Safety Evaluation Report that with the resolution of interface items, the systems required for safe shutdown conform to the Comission's requirements and are acceptable. We have reviewed the applicants' design of the instrumentation and conti system presented in Section 7.4.3.1 of the Preliminary Safety Analysis Report and identified 3 2183 251 7-4

in Table 7.1-1 with interface reconnendations of RESAR-41. On the basis of our review, we have concluded that the systems required for safe shutdown meet the requirements identified in Section 7.1 of ti.e Safety Evaluation Report and, therefore, are acceptable. 7.5 Safety Related Display Instrumentation 7.5.1 Monitoring of Heating, Ventilation and Air Conditioning Systems We stated in the Safety Evaluation Report that the applicants had originally provided radiation monitors in the control room and fuel handling building ventilation systems. However, they were later removed from the design. We required that the applicants prwide justification for the design change or include the radiation monitors in the design. In Amendment 23 of tne Preliminary Safety Analysis Report, the applicants have modified their design to include the radiation monitors indicated above. We have reviewed this design modification and have concluded that the design of the safety-related display instrumentation meets the requirements identified in Section 7.1 of the Safety Evaluation Report and, therefore, is acceptable. 7.8 Anticipated Transients Without Scram We stated in Section 15.5.7 of Appendix A to the Safety Evaluation Report that our evaluation in regard to Westinghouse's analysis related to anticipated transients without scram is not complete. Our requirements with respect to anticipated transients without scram are pro-vided in WASH-1270, " Anticipated Transients Without Scram for Water-Cooled Power Reactors." Westinghouse has stated that they believe the RESAR-41 design satisfies the requimments of WASH-1270 and that no hardware modifica-tions are required to mitigate the consequences of anticipated transients without scram. West,inghouse references Westinghouse Topical Reports, WCAP-8330, " Westinghouse Anticipated Transient Without Trip Analysis," and WCAP-8440 " Anticipated Transients Without Trip for a Four-Loop (3817 MWt) Westinghouse PWR", as the basis for this conclusion. These reports are currently under generic review by the staff. We will require that any design changes that are required as a result of our review, when it is completed, be implemented in the design of the South Texas Project. 2183 252 7-5

8.0 ELECTRIC POWER 8.3 On Site Power Sys+pm 8.3.3 Thermal Overload protection for Enqineered Safety Features Motor Operated Valves An item that was not addressed in the Safety Evaluation Report concerns the loss of function of safety related motor operated valves due to the malfunction of the therml overload protection circuits which are utilized to protect the motor windings against excessive heat. In nuclear power plants the criterion should be to drive the valve to its proper position to mitigate the effects of an accident, rather than be concerned with degradation or failure of the motor due to excessive heat. We informed the applicants of our reconmendations for implementing the design overload protection circuits for safety related motor operated valves as follows: (1) Thermal overload protection provided for safety related system motor operated valves shall have the trip setpoint 3t a value high enough to prevent spurious trips due to desica inaccuracies, trip setpoint drif t, or variations k the ambient temperature at the installed location. The trip setpoint chosen shall be consistent with that of any branch circuit protective device used. Periodic tests are reouired and shall be performed on each of the thermal overload devices to verify the accuracy and reliability of the overload trip setpoint, or: (2) Thermal overload protection may ba bypassed under accident conditions and the by;, ass circuitry shall be designed to IEEE Std 279-1971 criteria as appropriate for the rest of the safety related svstems. The applicants chose not to follow our recommendations and propose to remove the therral overload protecti.m function from all safety related motor enerated valves during all operating conditions, normal cr accident, and utilite the thermal overload devices for alarm purposes oniv. We do not recorrend this design approach in lieu of the choices identified above. Mcwever, '.e indicated to the applicants that thic is an acceptable eay to design aqairst the loss of function of a safety related motor operated valve due to malfunction or failure of equipront protective devices during accident conditions and therefore conclude that this design is acceptable. 2183 253 8-1

9.0 AUXILIARY SYSTEMS 9.4 Heating, Ventilation and Air Conditionin1 Systems 9.4.1 Control Room and Electrical Auxiliary Building Heating, Ventilation and Air Conditioning System We stated in the Safety Evaluation Report our concern regarding the adequacy of the smoke detector location at the control room air intake. A likely source of smoke would be a diesel fuel oil fire (see Section 9.5.1 of the Safety Evaluation Report). Subsequently, in Amendment 26 of the Preliminary Safety Analysis Report, the applicants stated that two redundant smoke detectors will be located near the junction between the two air intakes and the common duct to reduce transient time of smoke to the detectors. The smoke detectors will detect combustion products at a concentration of less than one percent of the toxicity level with instantaneous indication and alarm in the control room. The makeup air would travel through over 200 feet of duct work from the coninon intake before it reaches the control room. This would allow adequate warning for the operaters to isolate the control room and if necessary, utilize respirators. The applicants have provided an analysis that includes the control room carbon monoxide concentration in the event of a diesel fuel oil fire, (see Section 9.5.1 of this Supplement) and have demonstrated that carbon monoxide concentrations will be limited to acceptable levels. Based on our review of Amendment 26 of the Preliminary Safety Analysis Report, we have concluded that the control room heating, ventilation and air conditioning system design criteria and bases are in confonnance with Criterion 19 of the General Design Criteria and, therefore, are acceptable. 9.4.3 Fuel Handling Building Heating, Ventilation,and Air Conditioning System We stated in the Safety Evaluation Report (Sections 6.6, 9.4.3 and 15.7.3) that we would require the fuel handling building exhaust subsystem to be desianed as a seismic Category I system in accordance with Item 2(c) of Regulatory Guide 1.52. The applicants agreed to meet this position and have documented their compliance 4 Amendment 29 of the Preliminary Safety Analysis Report. We have concluded that the proposed design of the fuel handling building exhaust subsystem is acceptable. 2183 254 1 - t t 9-1

9.5 Other Auxiliary Systems 9.5.1 Fire Protection System We stated in ttje Safety Evaluation Report (Section 8.3.1 and 9.5.1) our concern regarding the fire potential of the diesel generator buildings. Each buildirrg will contain three 60,000 gallon fuel oil storage tt.nks in the upper level. A long lasting fuel oil tank fire of this magnitude cou'd incapacitate equipment required for safe shutdc n of the reactor. Subsequently, in Amendment 26 of the Preliminary Safety Analysis Report, the appeicants submitted an analysis, which assumed complete isolation of the diesel fuel tank room in the event of a fire. The fire would be rapidly extinguisheri because of lack of oxygen. No credit was taken for fire protection system actuation. The analysis predicted temperature effects on the structure, the amount of inerts, carbon dioxide, and soot that could be drawn into the unaffected engines and control room air intake, and concluded that the facility design was satisfactory. The analysis did not address itself to the potential effects of a major fire on the adjacent control room and cable spreading rooms other than an estimate of the carbon monoxide concentration in the control room air intake. The facility design included connecting doors between the upper level of the diesel generator building and the control room area. We were concerned about the predictability of the amount of air inleakage and the possibility of reignition upon sudden introduction of air following a fire. We expressed our concerns of the proposed system and our position regarding the need for continuous venting in the event of a fire, based on fire protection consultant experience, and the need for continuous air circulation, even in the event of loss of offsite power. The applicants committed in Amendments 31 and 32 of the Preliminary Safety Analysis Report that the following design changes and procedures would be implemented. (1) The applicants have comitted to move the diesel penerator building five feet away from the electrical auxiliary building containment the control room and the cable spreading room. gi (2)' Th'e first line of defense against a fire will be the building foam-water fire protection system. The system will be designed, installed, tested and maintained in accordarr.e with National t' ire Protection Association codes and stanJards. 2183 255 9-2

(3) The fuel oil tank room heating, ventilation and air conditioning system will be designed to assure continuous exhausting in the event of a fire. Exhaust; sents will exit directly from the tank rooms and will be routed such as to direct smoke and soot away from the diesel generators and control room air intakes. Exhaust fans will be utilized to maintain a slight negative pressure in the rooms. These fans will be powered from engineered safety feature electrical buses. Their power supply would be shed automatically in the evert of postulated design basis accidents and restarted manually wht.n compatible with emergency load requirements. (4) The tank rooms will be provided with drain lines of sufficient capacity to remove the fire protection system water and foam at the same rate that it is released, and will further have addit,1onal margin for simultaneous removal of spilled oil. (5) The heating, ventilation and air conditioning system will be designed to maintain the room ambient temperature safely below the minimum flash point of diesel fuel oil. (6) The diesel fuel tank room doors will be locked closed except during periodic inspections and maintenance. (7) An analysis has been submitted including prediction of effluent rates and the concentration of inerts, carbon monoxide and soot at the unaffected engines and c.ontrol rt i air intakes in the event of the postulated fire. On the basis of our review we have concluded that the fire protection system design criteria and bases are in conformance with Criterion 3 of the General Design Criteria and that the facility and system design is acceptable. 2183 256 9-3

11.0 RADI0 ACTIVE WASTE MANAGEMENT 11.1 Sumary Description We stated in the Safety Evaluation Report that we had not completed our review of the radicactive waste systems to meet the dose design objectives of Appendix ! to 10 *,FR Part E0 (effective June 4.1975) and the required cost-benefit analysis. We have evaluated the radioactive waste management systems proposed for South Texas Project, Units 1 and 2. to reduce the Quantities of radioactive materials released to the environment in liquid and gaseous effluents in accordance with 10 CFR Part 50.34a. These systems have been previously described in Sections 11.2 and 11.3 of the Safety Evaluation Report and in Section 3.5 of the Final Environmental Statement. Based on more recent information applicable to the South Texas Project and changes in our calculational model, we have revised the liquid and gaseous source terms given in the Final Environmental Statement. These changes occurred subsequent to issuing the Final Environment Statement for the South Texas Project. The revised source terms were calculated using the models and methodology described in Draft Regulatory Guide 1.08. " Calculation of Peleases of Radioactive Materials in Liquid and Gaseous Effluents from Pressurized Water Reactors (PWR's)." September 9,1975. On September 4,1975 the Comission amended Appendix I to 10 CFR Part 50 to provide persons who have filed applications for construction permits for light-water-cooled nuclear power reactors which were docketed on or after January 2.1971 and prior to June 4,1976 the option of dispensing with the cost-benefit analysis required by Paragraph II.D of Appendix I. This option permits an applicant to design their radwaste management systems to satisfy the Guides on Design Objectives for Light-Water-Cooled Nuclear Power Peactors proposed in the Concluding Statement of Position of the Regulatory Staff in the rule making hearing on as low as practicable (RM 50-2), dated February 20, 1974 As indicated in the Statement of Consideration included with this amendment the Comission noted it is unlikely that further reductions to radioactive material releases would be warranted on a cost-benefit basis for light-water-cooled nuclear powe. reactors having radwaste systems and equipment determined to be acceptable under the proposed staff design objectives set forth in RM 50-2. In a letter to the Commission dated October 1,1975, the applicants chose to comply with the September 4.1975 amendment to Appendix I rather than submitting a cost-benefit analysis as required by Paragraph II.D. 2183 257 11-1

Based on our evaluation of the liquid radwaste manacement systems, we estimate that the quantity of radioactive materials released in liquid effluents, excluding tritium and dissolved noble gases, will be less than 5 Curies per year per reactor and the total calculated quantity of radio-active materials released in liquid effluents from Units 1 and 2 will not result in an annual dose or dose comittrent to the total body or to any organ of an individual in an unrestricted area from all pathways of exposure in excess of 5 millirem. Based on our evaluation of the gaseous radwaste management systems, we estimate that the total quantity of radioactive materials released in gaseous effluents from Units 1 and 2 will not result in calculated annual gamma air dose in excess of 10 millirad and a beta air dose in excess of 20 millf red at every location near ground level, at or beyond the exclusion boundary distance, which could be occupied by individuals. We estimate that the annual total quantity of iodine-131 released in gaseous effluents will not exceed 1 Curie per year per reactor and the calculated annual total quantity of radiolodine and radioactive particulates released in gaseous effluents from Units 1 and 2 will not result in an annual dose or dose comitment to any organ of an individual in an unrestricted area from all pathways of exposure in excess of 15 millirem. Our evaluation of the proposed liquid and gaseous radioactive management systems for South Texas Project Units 1 and 2 shows these systems to be capable of meeting the criteria given in Appendix ! to 10 CFR Part 50 for keeping releases of radioactive materials to the environment "as low as is reasonably achievable" and, therefore, we find the proposed systems to be acceptable. 2183 258 tein 11-2

14.0 INITIAL TESTS AND OPEPATIONS We stated in the Safety Evaluation Report that we believe that acceptable testing can be developed for the emergency core cooling system and instrument air system and that we will require the applicants to develop acceptable testing methods at tbe operating license stage of review. In amendment 30 of the Preliminary Safety Analysis Report the applicants provided additional information with regard to the testing of these systems. With respect to the emergency core cooling system. the applicants have proposed to perform tests on the "B" train sump of Unit 1. The two units are identical, and the three sumps in each unit are of similar geometry and length to diameter ratio. The applicants propose to construct temporary retaining walls around the sump and will install throttle valves and discharge segments in both the high and low injection lines. Thus, water can be maintained in the sump while the pumps are tested to run-out condition. The data to be taken will include pump suction and discharge pressures and the pump volumetric flow rate, and water le.el and temperature in the sump. Viewing ports will be installed to permit visual observation of vortex ccntrol. We have determined that the proposed test will provide the information necessary to meet the recommendations of Regulatory Guide 1.79, and therefore is acceptable. With respect to the instrument air system, the applicants have proposed to perform suitable testing of the air supply and purification equipment to demonstrate that they meet design requirerents and have proposed to conduct a simulation of loss of instrument air. We have determined that the proposed tests meet the applicable recommendations of Regulatory Guide 1.80 and, therefore, are acceptable. 2183 259 14.1

17.0 QUALITY ASSURANCE 17.3 Brown & Root. Inc. In Amendment 32 of the Preliminary Safety Analysis Report (PSAR), the applicants revised the organizational structure of the engineer-constructor, Brown & Root. Inc. As a result the following replaces Section 17.3 of the Safety Evaluation Report. Brown & Root, Inc. has been designated engineer-constructor responsible for the design, engineering, equipment and materials procurement, and construction of the South Texas Project. Units 1 and 2. This includes all plant structures, systems, and components except those provided by Westinghouse. Figure 17.2 shows the Brown & Root organization as it relates to enVneering, procurement, construction, and QA. The QA Manager of Brown & Root, respon;itle for the QA program, reports to the Power Divis on Senior Group Vice-President. The QA Manager is on a i comparable organization level and technically and administratively independent of the engineering, construction and project organizations whose work the QA organization oversees. Although the purchasing organization is ou'. side the construction and engineering organization, QA controls of procurements are well defined in the Preliminary Safety Analysis Report. Therefore, we find the organizational independence snown in Figure 17.2 acceptable. The Executive Vice-President of Brown & Root has issued a management statement of policy which requires mandatory implementation of the QA program. The Brown & Root QA Manager issues the quality assurance / quality control (QA/QC) procedures for the South Texas Project. Engineering procedures and purchasing procedures for the South Texas Project are reviewed and audited by Brown & Root QA. Quality verification activities such as inspection, audits, and surveillance are conducted by personnel in the QA organization (see Figure 17.3). The Preliminary Safety Analysis Report includes matrices of QVQC procedures (including inspection pmcedures for construction), engineering procedures, and purchasing procedures for the South Texas Project. These procedures are cross referenced to the criteria of Appendix B to 10 CFR Part 50. Based on our review of this information, we have concluded that each criterion of Appendix B to 10 CFR Part 50 has been adequately included in the QA program for the South Texas Project Unjts 1 and 2. i<' 2183 260 17-1

In response to our request, Brown & Root, Inc. has committed to follow the guidance provided by the, ' omission in " Guidance on Quality Assurance During Design and Procurement Phase o, "utiear Power Plants," (Revision 1) May 24,1974 (WASH 1283) and " Guidance on Quality Assurance Requirements During the Construc-tion Phase of Nuclear Power Plants," May 10,1974 (WASH 1309). Based on this, Bn)wn & Root, Inc. has comitted to the essential, requirements for a QA Program in compliance with Appendix B to 10 CFR Part 50. Brown & Root, Inc. has identified the safety related structures, systems, and components that are subject to the Brown & Root QA program in the PSAR. Thess safety related items, and those listed in RESAR-41 supplied by Westinghouse, will fall within the Brown & Root program upon receipt at the South Texas Project site. The Preliminary Safety Analysis Report describes a training and indoctrination program comitted to assuring that personnel performing quality affecting activities understand, implement, and enforce the Brown & Root QA policies and procedures. The program assures adequate training and qualification in the printiples and techniques of quality related activities. The Brown & Root QA organization is shown in Figure 17.3. Functions of Engineering QA include assurance that QA program requirements are implemented during the design and procurement phase of the project. This office also coordinates QA technical matters through the Brown & Root Houston QA office during the construction phase of the project. The Brown & Root site QA organization is supervised by the Site Project QA Manap who is directly responsible to the QA Manager. He coordinates QA project administration and policy with the Construction Project Manager. His staff performs the needed QA/QC functions at the South Texas Project site. In addition, testing laboratories perfoming QC functions will report to the Site Project QA Manager. QA inspection personnel report to QC supervision which, in turn, reports to the Site Project QA Manager. We have concluded that the site QC inspectors have sufficient authority and organizational freedom to perfom their functions effectively and without reservation. Brown & Root Inc. has described a system of planned and documented audits witn provision for corrective and followup actions. Audits will be performed in accordance with written checklists by appropriately trained personnel having no direct responsibility in the area audited. Audit schedules are based on the status and safety importance of the activities being performed. Audit report distribution includes management personnel of the audited area, Houston Lighting & Power Company and Brown and Root. The audit reports will be a part of the QA record files at the project. 2183 261 Us L8 n 17-2

Brown & Root, Inc. has established a QA Review Board, directed by the Executive Vice President, which includes the Senior Group Vire-President of the Power Division, the Senior Vice-Presidents of Power Con? truction and Power Engineering, the Purchasing Vice-President, and the QA Manager. This board meets at least semiannually to review and discuss the administrative activities of the QA Department to detemine and evaluate the effectiveness of the corporate QA program. Based on our review and evaluation of the QA program described in amended Section 17 of the Preliminary Safety Analysis Report, we have concluded that Brown & Root's QA program for the South Texas Project demonstrates an acceptable QA and QC organization with adeouate policies, procedures, and instructions to implement a program that will satisfy the reouirements of Appendix B to 10 CFR Part 50. 2183 262 < r, 2. 8 i S 37 3

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18.0 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS The Advis, y Comittee on Reactor Safeguards completed its review of the applicatis for a construction permit for the South Texas Project Units 1 and 2 at its 185th meeting held September 11-13, 1975. A copy of the Comittee's report for the South Texas Eroject dated September 19, 1975, which contains certain coments and recomendations, is attached as Appendix C. The actions we have taken or plan to take in response to these coments and recommendations are described in the following paragraphs. (1) The Comittee stated that they believe that the interface program, when completed in a manner satisfactory to the Nuclear Regulatory Comission staff, will provide an adequate design basis for the balance-of-plant. The interface program for the South Texas Project has been completed in a manner satisfactory to the staff and is addressed in Section 1.9 of this Supplement. (2) The Corr:ittee stated that the Nuclear Regulatory omission staff has identified a number of outstanding issues speci' to the South Texas Project application as well as to RESAR-41, somt of which will reauire resolution before the issuance of a construction permit. The Comittee recomended that these matters be resolved in a manner satisfactory to the staff. The outstanding issues specific. to the South Texas Project are identified in Section 1.8 of the Safety Evaluation Report. The outstanding issues related to RESAR-41 are identified in Section 1.7 of Appendix A to the Safety Evaluation Peport. The satisfactory resolution of these outstanding issues are discussed in the appropriate s?ctions of this Supplement. The Committee also requested that they be kept inforred on the resolution of the following items: (a) The emergency core cooling system evaluation. (b) Diesel generator building design and location of the storage tanks for the diesel fuel. 2183 265 x Lbn 18-1

Our evaluation of the emergency core cooling system is addressed in Section 6.3 of this Supplement. This matter has been resolved in a manner satisfactory to the staff. Our evaluation regarding the diesel generator building design and location of the storage tanks for the diesel fuel is addressed in Section 9.5.1 of this Supplement. This matter was also resolved in a manner satisfactory to the staff. (3) The Comittee recommended that the Nuclear Regulatory Commission staff and the applicants review the design features that are intended to prevent the occurrence of damaging fires and to minimize the consequenc,es to safety-related equipment should a fire occur. The Comittee also requested that it be kept infomed concerning this matter. The staff is considering a program to conduct a comprehensive review and evaluation of all nuclear power plants. The review will consider experience gained from the Browns Ferry Nuclear Generating Station fire, recommendations from the Nuclear Energy Liability-Property Insurance Association and from other qualified fire protection consulting agencies. The fire protection systems for the South Texas Project Units 1 and 2 will be upgraded if the results of our evaluation 50 dictates. Fe will inform the Co-Otee of the results of our review. (4) The Committee requested that it be kept informed with respect to our review of the radioactive waste systems to meet the dose Jesign objectives of Appendix I to 10 CFR 50 (effective June 4, 1975) and the optional cost-benefit analysis provided % the amendment to the rule effective September 4,1975. Our evaluation of the radioactive management systems proposed for South Texas Project, Units 1 and 2, to reduce the quantities of radioactive materials released to the environment in liquid and gaseous effluents in accordance with Appendix I to :D CFR Part 50 is addressed in Section 11.0 of this Supplement. As discussed in Section 11, we have determined that the South Texas Project ' Units 1 and 2 radioactive management systems meet the requirements of Appendix I to 10 CFR Part 50 and are therefore, acceptable. (5) The Comittee expressed its continuing concern regarding generic problems related to large water reactors, recomending that such problems be dealt with appropriately by the applicants and the Nuclear Regulatory Commission staff. These generic problems are discussed in a report by the Advisory Comittee on Reactor Safeguards dated March 12,1975. These problems are being worked on by the various reactor vendors and other industrial organizations and will be the subject of continuing attention by the staff. 2183 266 18-2

20.0 FINANCIAL QUALIFICATIONS Section 50.33(f) and Appendix C of 10 CFR Part 50 are the Comission's regulations which relate to financial data and information required to establish financial qualifications for an applicant for a facility construction permit. Houston Lighting & Power Company, Central Power and Light Company, City Public Service of San Antonio, and City of Austin have applied for construction permits for the South Texas Project. Units 1 and 2. The applicants shall own undivided interests as tenants in comon in the South Texas Project site, Units 1 and 2 and common station facilitics, as follows: Houston Lighting & Power Company 30.8% Central Power and Light Company 25.2% City Public Service of San Antonio 28.0% City of Austin 16.0% 100.0% The most recent estimate of the total cost of South Texas Project, Units 1 and 2 was provided on September 19, 1975 in response to our request for additional financial information. These costs may be sumarized as follows: Total Nuclear Plant Costs $1,352,700,000 TransmissionCostsU 110.770,000 Nuclear Fuel - Initial load 136,085,000 $1,599,555,000] 2 The estimated cost of the nuclear plant has been reviewed by comparing it to the cost projected by the Energy Research and Development Agency's CONCEPT calcula-tional model. This model currently uses construction inflation or escal'ation rates of 8 percent per year for site labor, materials, and purchased equipment. The CONCEPT model projected,the cost of the nuclear plant to be $1,339 million, compared with the applicants' estimate of $1,352,700,000. This is only a difference of about one percent. Consequently, it appears that the applicants' estimate is reasonable. The earliest construction completion dates for Unit 1 and 2 are estimated to be May,1980 and October,1981, respectively. The latest estimated completion dates are May,1982 and October,1983, respectively. Assuming adherence to the former schedule, comercial operation for Units 1 and 2 is scheduled for October,1980 and March,1982, respectively. O o be borne by each participant in varying amounts. T 2] Includes escalation and allowance for funds used during construction. 2183 267 20-1

The total estimated costs to be bomo by each applicant are as follows: Houston Lighting & Power Company $498,724,780 Central Power & Light Company 402,942,820 City Public Service of San Antonio 450,695,800 City of Austin 247,191,600 $1,599,555,000 The electric operations of the applicants are not subject to the Federal Power Comission. In addition, at the present time there is no state-wide regulatory agency in Texas having jurisdiction over the rates and service of electric utilities. Most municipalities in Texas are empcwered by statute to regulate the investor-owned electric utilities within their jurisdiction. The San Antonin anc' Austin city councils establish the rates for their respective systems. We have reviewed the financial information presented in the application, and amendments thereto, and hava concluded that there is reasonable assurance that the aforementioned applicants can raise the necessary funds to design and construct the South Texas Project, Units 1 and 2. Accordingly, we find them financially qualified to carry out the activities for which this construction permit is sought. Our conclusion is based upon the analyses presented in Appendix E to this supplement and the basic assumptions of rational regulatory policies and relatively stable capital market conditions. These assumptions are necessary because of the lengthy future period involved and the expected heavy dependence on external financing. 2183 268 e c8K 20-2

21.0 C_0_NCLUSIONS Our conclusion that the issuance of permits for construction of the facilities will not be inimical to the corrmon defense and security or to the health and safety of the public, as stated in the Safety Evaluation Peport. Section 21.0 was conditioned on the favorable resolution of outstanding matters identifled in Section 1.8 of the Safety Evaluation Report. We have discussed each of these outstanding issues in this Sepplement and have indicated a favorable resolution of each matter. Therefore, we reaffirm our conclusions as set forth in Section 21.0 of the Safety Evaluation Report. 2183 269 21-1

APPENDIX A UPDATED SUPPLEMENTARY INFORMATION FOR APPENDIX A TO THE SAFETY EVALUATIOf REPORT 92 SOUTH TEXAS PROJECT UNITS 1 AND 2 " REPORT TO THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS BY THE OFFICE OF NUCLEAR REACTOR REGULATION U. S. NUCLEAR REGULATORY _C_NNISSION IN THE MATTER OF kTSTINGHOUSE ELECTRIC CORPORATION REFERENCE SAFETY ANALYSIS PEPORT RESAR-41 DOCKET NO. STN 50-480" 2183 270

• a.

A-1

3.0 DESIGN CRITERIA FOR SYSTEMS AND COMPONENTS 3.5 Seismic Design 3.5.2 Seismic System and Subsystem Analysis We stated in Appendix A to the Safety Evaluation Report that the proposed procedures of seismic analysis for the case where a component or system is supported from two or more locations, with relative displacements and different response spectra, were unacceptable. In Amendment 18 to RESAR-41. Westinghouse comitted to our requirements. Westinghouse agreed that a dynamic msponse spectrum analysis will first be made assuming no relative displacement between support points. The response spectra used in this analysis will be the worst floor response spectra. Secondly, the effect of differential seismic movement of components interconnected between f!oors will be considered statically in the integrated system analysis and in the detailed component analysis. The results of the building analysis will_ be reviewed on a mode-vy-mode basis to detemine the differential motion in each mode. The results of these two steps, the dynamic inertia analysis and the static differential motion analysis, are combined in an absolute manner, with due consideration for the American Society of Mechanical Engineers classification of the stresses. These results will be used in accordance with Section III of the Code, Paragraphs NB-3652 and NB-3753 for piping, and Paragraph NB-3227 for components. Based on our review of this infonnation, we have concluded that for the methods used to analyze components and systems supported from two or more locations with relative displacements and diffemnt response spectrum, the dynanic methods and procedures for seismic systems analyses proposed b.v Westin@oase provide an acceptable basis for seismic design.. 3.7 Seismic Qualification of Category I Instrumentation and Electrical Equipment We stated in Appendix A to the Safety Evaluation Report that we would require that Westinghouse comit to the seismic qualification af instrumentation and electrical equipment important to safety in accordance with the requirements of IEEE 344-1975. Westinghouse has provided in RESAR-41 a comitment to conduct the seismic qualiff-cation of instrumentation and electrical equipment important to safety in accordance with the criteria of IEEE 344-1975. We find this commitment to be acceptable for issuance of a Construction Pemit for the South Texas Project which references RESAR-41. 2183 271 A-2

4.0 REACTOR 4.2.4 Reactivity Control System 4,2.4.3 Part Length Control Pod Drive Mechanisms The design of the part length control rod drive mechanisms was not addressed in Appendix A to the Safety Evaluation Report. This design is currently under review by the staff. RESAR-41 incorporates a magnetic jack design for these mechanisms. Previous Westinghouse designs used the lead-screw concept. Because of potential departure from nucleate boiling (DNB) problems associated with the use of part length control rods identified by Westinghouse, the use of these rods is not now pemitted on Westinghouse pressurized water reactors. We will not authorize their use on RESAR-41 plants until the DNB problem has been resolved and we have satisfactorily completed our review of the part length drive mechanisms. Based on our experience with operating pressurized water reactors and our review of RESAR-41, we have determined that the reactor can be safely operated without the use of part length control rods. Therefore, we have concluded for the South Texas Project, that the reactor can be safely operated withcut the use of part length control rods and, therefore, with the restrictions on the use of the part length control rods the reactor system design is acceptable. 2183 272

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'U 3 A-3

5.0 REACTOR COOLANT SYSTEM 5.4 Component and Subsystem Design 5.4.3 Residual Heat Removal System We stated in Appendix A to the Safety Evaluation Report that the design of the residual heat removal system is an outstanding issue. We are currently discussing this issue with Westinghouse to resolve it on a generic basis. We have detemined that on the basis of our review of other approved residual heat removal systems, we can satisfactorily resolve the RESAR-41 residual heat removal system design.g The applicants have made a comitment to accept the generic resolution of this issue. On this basis, we find this comitment to be acceptable for the South Texas Project construction permit stage. S.4.8 Rapid Refuelir.1 Systec 5.4.8.2 Roto-Lok Closure System We stated in Appendix A of the Safety Evaluation Report that we would require an analysis of the consequences of a postulated failure of a single Roto-Lok head closure stud. Westinghouse submitted information on an analysis of this postulated event which showed that the consequences would be acceptable. Using conservative assumptions, Westinghouse has shown that the failure of one closure stud would result in stress levels in the adjacent studs with margin to that allowed by Section III of the ASME Boiler and Pressure Vessel Code and have concluded that failure of one stud will not caua the failure of any other studs. We have reviewed this infomation and concluded that Westinghouse has acceptably demonstrated that, for the design conditions for the plant, a postulated failure of a single Roto-Lok head closure stud will not result in the failure of any other studs and, therefore, the design is acceptable. 5.4.8.4 One Lift Concept We stated in Appendix A to the Safety Evaluation Report that we would require an analysis of the consequences from the postulated dropping of the reactor vessel head during refueling. The applicants intend to use the RESAR-41 proposed one-lift refueling concept. This entails lifting the reactor vessel head with the 2h A-4

upper internals, control rods and drive mechanisms as a unit. The mass of this assembly is approximately twice as much as previous designs. This has raised a concern about the integrity of the reactor coolant system and the ability to maintain core geometry and provide adequate core cooling if the reactor vessel head were dropped. Westinghouse submitted the results of an analysis of this postulated accident and Stated that there would be no consequential damage to the fuel and that core cooling capability would not be jeopardized. As a result of our review we detemined that additional information was required to confim Westinghouse's conclusions. Westinghouse has recently submitted this information which we are presently reviewing. However, we will require the applicants to provide an overhead reactor vessel head assedly handling system that is designed so that the connected load would not fall in the event of a single failure or malfunction, unless the analysis shows that dropping the reactor vessel head assedly will not effect the integrity of the safety related components. 5.4.9 Reactor Vessel Sunports On May 7,1975, we were infomed by a licensee of a pressurized water reactor, Virginia Electric and Power Company, that an asymetric loading resulting from a postulated pipe rupture at a particular location in the reactor coolant system had not been taken ir+o account in the original design of the reactor pressure vessel support system for the North Anna Units 1 and 2 (Docket Nos. 50-338 and 339). This loading results from the forces induced on the internals within the reactor vessel caused by differential pressure conditions within the vessel imediately following a postulated loss-of-coolant accident. In addition, the asymetric loading from transient differential pressures that would exist around tne exterior of the reactor vessel from the same postulated pipe rupture was not included in the original design analysis. However, the symmetric loadings from such a postulated pipe rupture were included in the original analysis of the reactor pressure vessel supports. It is our opinion that these factors related to the design of the reactor pressure vessel supports are, generic in nature and may apply to the RESAR-41 design. Accordingly, we are taking steps to review this problem on a generic basis to determine the extent of the problem. We have infomed Westinghouse of the nature of this problem and have requested Westinghouse to verify that the design procedures for the reactor pressure vessel support system will properly include the asymetric forces described above in the final design of the supports. In a letter dated October 24,1975 Westinghouse provided verification that the final design will include the asymetric forces. 2183 274 A-5

Based on our review of this generic problem to date, we have detemined that the methodology necessary to model the complete reactor coolant system in sufficient detail to detemine analytically the magnitudes and phase relationships of the vessel support system loads from the transient pressure differentials has been developed extensively by Westinghouse, and that the calculational techniques have been refined so that it is practical to evaluate the actual dynamic system response to all the known transient loads. Furthermore. Westinghouse has informed us that structural analyses based on the loads developed by the worst case loading (which is a rupture of the reactor coolant pipe at the cold leg nozzle) demonstrate that Westinghouse reactor coolant support systems now being designed can sustain these loads and remain within the conservative design basis stress limits comparable to those stress limits specified in Appendix F of Section III of the ASME Boiler and Pressure Vessel Code. On the basis of our review of this problem to date, we have concluded that Westinghouse can properly account for these forces during the final design of the reactor vessel support system. 2183 275 g A-6

7.0 INSTRUMENTATION AND CONTPOLS 7.6 Other Instrumentation Systems and Requirements for Safet_y 7.6.1 Environmental Qualification We stated in Appendix A to the Safety Evaluation Report that we would reouire that Westinghouse comit to the environmentel qualification of instrumentation and electrical equipment irportant to safety in accordance with the requirements of IEEE 323-1974. Westinghouse has provided in RESAR-41 a comitment to corsuct the environmental qualification of instrumentation and electrical equipment important to safety in accordance with the criteria of IEEE 323-1974 We find tMs comitment acceptable for issuance of a Construction Permit for the South Texas Project. We stated in Appendix A to the Safety Evaluation Report that we require that the nuclear instrumentation neutron detectors be qualified for the worst case environment in the containment in accordance with the requirements of IEEE 323-1974. Westinghcuse has amended RESAR-41 to show that the overpower reactor trips are not required functions to provide protection against steam line ruptures. In addition, we have determined that the neutron detectors are not required for protectior, from accidents where the containment atmosphere would exceed thequalificationlimitofthedetectors(175degreesFahrenheit). We have concluded that the environmental qualification of the nuclear instrumentation detectors is acceptable. 2183 276 A-7

15.0 ACCIDENT ANALYSES 15.3 Anticipated Transients We stated in Appendix A to the Safety Evaluation Report that the design of the chemical and volume control system must be modified to allow adequate time for operator action during a boron dilution accident while refueling or during startup. Westinghouse has amended the Technical Specifications on RESAR-41 to require that those valves that control fresh water to the charging pumps during refueling must be locked closed. In addition, they have cecified that makeup to the primary system will not be permitt.d during makeup to the refueling water storage tank or until after sampling of the tank following makeup to it. If dilution were to occur by some unknown method, indication of the approach to criticality will be available to the operators through audible alarms and visual and audible indica-tion of the nuclear instrumentation. We have reviewed the design provisions in RESAR-41 for preventing a baron dilution accident. We have concluded that the proposed methods of isolating all dilution flow paths and providing indication of the core reactivity provide adequate protection against a baron dilution accident and are acceptable. 2183 277 A-8

APPENDIX B CONTINUATION OF CHRONOLOGY OF RADIOLOGICAL REVIEW 0F SOUTH TEXAS PROJECT, UNITS 1 AND 2 July 24, 1975 Letter to applicants requesting updated financial infomation August 1, 1975 Letter to applicants requesting infomation concerning auxiliary and power conversion systems August 1,1975 !ssuance of Safety Evaluation Report August 15, 1975 Submittal of Amendment No. 29, consisting of infomation relative to comitments made at meeting held July 17, 1975, and other changes to the Preliminary Safety Analysis Report. August 20, 1975 Letter to applicants advising of staff position regarding design of diesel generator building and diesel generator building heating, ventilation and air conditioning system August 27, 1975 ACRS Subcomittee meeting with staff and applicants September 5,1975 Letter. rom applicants conceming monitoring of horizontal movement Septed er 8, 1975 Letter to, applicants requesting additional information concernind the emergency core cooling system containment pressure calculation September 8,1975 Letter to applicants requesting additional infomation concErning the component cooling water system Jesign September 10-11, 1975 tieeting with applicants to discuss design of diesel generator building September ll,1975 ACRS meeting with staff and applicants Septenter 17, 1975 Letter from applicants transmitting updated financial information September 19,1975 Report by the ACRS B -1

Septecter 26, 1975 Meeting with applicants to discuss outstanding issues October 1,1975 Subraittal of Amend:nnt No. 3a consisting cf infomation corcerning certain outstanding issues and otter changes to ths Preliainery Safety Analysis Report. October 9,1975 fubmittal of Amen'dment No. 31. consisting of additional information concerning outstanding issues and other changes to the Preliminary Safety Analysis Repo*t October 17, 1975 Submittal of Amendment No. 32, consisting of revised and updated information. October 20, 1975 Submittal of Amendment No. 2 to License Application, consisting of updated financial information. October 22, 1975 Letter from applicants concerning emergency boration system. 2183 279 i'bl% v.. B-2

APPENDIX C ADVISORY COMMITTEE ON REACTOR SAFEGUARDS NUCLEAR REGULATORY COMMISSION W ASHINGTON, D. C. 20555 September 19, 1975 lbnorable William A. Andern Gairman U. S. Nuclear Regulatory Comission Washington, D. C. 20555

Subject:

REPORf CN SolrI11 'IEXAS PRQJECT LNITS 1 AND 2 At its 185th meeting, September 11-13, 1975, the Advisory Comittee on teactor Safeguards reviewed the application of Ibuston Lighting and Tbwer Company, the City Public Service of San Antonio, the Central Power and Light Company and the City of Austin (Applicants) for a permit to construct the South ibxas Project thits 1 and 2. % e site was visited on August 26, 1975, and the project was considered at a Subcomittee meeting at Bay City, 7bxas on August 27, 1975. During its review, the Comittee had the benefit of discussions with representatives and consultants of the Applicants, Westinghouse Electric Corporation, Brown & Root, Incorporated, and the NRC Staff. We Comittee also had the benefit of the documents listed below. We Plant will be located on the Colorado River in Matagorda County, Texas, approximately 89 miles southwest of Ibuston and 12 miles south-southwest of Bay City, the designated population center (1970 population,11,733; projected 2020 population, 24,000). W e exclusion area has a minimum boundary distance of 1430 meters. %e radius of the low population zone (present population, 55) is three miles. fujor land use in the area of the plant site is for the production of rice and cattle. %e South Texas Project will be the first plant to reference the RESAR-41 Westinghouse Standard Design Nuclear Steam Supply System (NSSS). We South 'Ibxas Project will be in compliance with the RESAR-41 requirements. We Comittee reported on RESAR-41 in its letter of September 18, 1975. Each reactor unit will utilize a four-loop pressurized water nuclear steam supply system having a core power level of 3800 MW(t). Groendwater at the site area consists of a shallow, low quality aquifer occurring above depths of 90-150 feet and a high quality aquifer cormencing at depths in the vicinity of 300 feet. Groundwater usage is almost totally from the deep aquifer. Based upon observations at other areas of similar soil structure, suchlas the Ibuston area, continual pumping of ground 2183 280 C-1

Honorable William A. Anders September 19, 1975 water frorr. the high gaality acquifer is expected to cause subsidence in the vicinity of the plant site. % e Applicant has developed design criteria assuming long term settlements, and has cottmitted to the NRC Staff to nonitor ubsidence at the site over the life of the plant. We Comittee believes that the planned actions provide an adequate basis for the safety of the plant structures. %e ultimate heat sink for the plant will be an artificial pond eight feet deep covering over 40 acres. It will be capable of providing the cooling water required for shutdown and maintenance of both reactors in shutdown condition for a minimum of 30 days. %e Comittee has reviewed the plans of the Applicant and the NSSS designer to complete the identification and documentation of interface information required by the balance of plant contractor to neet the safety design requirements of the NSSS designer. %e Comittee believes that this program, when ecmpleted in a nanner satisfactory to the NRC Staff, will provide an adequate design basis for the balance of plant. We NRC Staff has identified a number of outstanding issues specific to this application as mil as to RCSAR-41, some of which will require resolution before the issuance of a construction permit. %e Comittee recomends that these matters be resolved in a manner satisfactory to the Staff. %e Comittee wishes to be kept informed on the resolution of the following items: 1. W e emergency core cooling system evaluation, 2. Diesel engine building design and location of the storage tanks for the diesel fuel. We Comittee recomiends that the NRC Staff and the Applicant review the design features that are intended to prevent the occurrence of damaging fires and to minimize the consequences to safety-related equignent should a fire occur. h is matter should be resolved to the satisfaction of the NRC Staff. We Comittee wishes to be kept informed. 2183 281 -r s C -2

Ibnorable William A. Anders September 19, 1975 he NRC Staff is currently reassessing the paraneters and mathematical models for calculating releases of radioactive materials in effluents from this plant. Although these calculations include the consideration of additional airborne releases such as carbon-14 and particulates, the Staff does not anticipate that the nodifications will result in any sub-stantial increase in the annual population doses previously estimated. W e Staff has offered the Applicant the option of including in the South Texas Plant waste management systems meeting the requirements of the earlier proposed Appendix I,10 CFR 50, or the revised guidance as outlined in the. Comission's issuance of April 30, 1975. W e revised guidance includes the requirement that cost-benefit analyses be taken into consideration in the determination of waste management needs. We Comittee wishes to be kept informed on this matter. Generic problems relating to large water reactors are discussed in the Comittee's report dated March 12, 1975. % ese problems should be dealt with appropriately by the NRC Staff and the Applicant. %e Comittee believes that the above items can be resolved during con-struction and that if due consideration is given to these items, the South Texas Project Units 1 and 2 can be. constructed with reasonable assurance that they can be operated without undue risk to the health and safety of the public. Sincerely yours, MM W. Kerr Chairman References attached. 2183 282 C -3

Honorable William A. Anders Septaber 19, 1975 REFERDK'ES 1. Houston Lighting and Power Company, et al, "Scuth Texas Project, Preliminary Safety Analysis Report," (PSAR), Vols 1-10. 2. Amendments 1-29 to PSAR. 3. U.S.N.R.C., Safety Evaluation Repcrt fo : the South Texas Project, August 1, 1975. 4. W2stinghouse Electric Corporation, " Reference Safety Analysis Peport-41" (RESAR-41), Vol 1-8. 5. Amendments 1-18 to RESAR-41. 2183 283 ,or. 3: LdI t N

APPENDIX D MINIMUM CONTAINMENT PRESSURE MODEL FOR PWR ECCS PERFORMANCE EVALUATION A. BACKGROUND Paragraph I.D.2 of Appendix K to 10 CPR Part 50 (Ref, li requires that the containment pressure used to evaluate the performan:e capability of a pressurized water reactor (PWR) emergency core coe ling ayctem (ECCS) not exceed a pressure calculated conservatively.or that pur-pose. It further requires that the calculation include the effects of operation of all installed pressure-reducing systems and processes. Therefore, the following branch technical position has )een developed to provide guidance in the performance of minimum containment pressure analysis. The approach described below applies only to the ECCS-related containment pressure evaluation and not to the :ontainment functiont.1 capability evaluation for postulated design basis accidents. B. BRANCH TECHNICAL POSITION 1. Input Information fo. Model a. Initial Containment Internal Conditions The minimum containment gas temperature, minimum containment pressure, and maximum humidity that may be encountered under limiting normal operating conditions should be used, b. Initial Outside Containment Ambient Conditions A reasonably low ambient temperature external to the containment should be used. c. Containment Volume The maximum net free containment volume should be used. This maximum free volume should be determined frcm the gross contain-ment volume minus the volumes of internal structures such as walls and floors, structural steel, major equipment and piping. The individual volume calculations should reflect the uncertainty in t:te component volumes. 2183 284 sci tua" D-1

2. Active Heat Sinks a. Spray and Fan Cooling Systems The operation of all engineered aafety feature containment heat removal systems operating at maximum heat removal capacity; i.e., with all containment spray trains operating at maximum flow con-ditions and all emergency fan cooler units operating, should be assumed. In addition, the minimum temperature of the stored water for the spray cooling system and the cooling water supplied to the f an coolers, based on technical specification limits, should be assumed. Deviatiens from the foregcing will be accepted if it can be shown that the worst conditions regarding a single active failure, stored water temperature, and cooling water temperature have been selected from the standpoint of the overall ECCS model. b. Containment Steam Mixing With Spilled ECCS Water The spillage of subcooled ECCS water into the containment pro-vides an additional heat sink as the subcooled ECCS water udxes with the steam in the containment. The effect of the steam-water mixing should be considered in the containment preseure calculations. Containment Steam Mixing With Water from Ice Melt c. The water resulting from ice melting in an ice condenser contain-ment provides an additional heat stak as the subcooled water mixes with the steam while draining from the ice condenser into the lower containment volume. The effect of the stesa-water mixing should be considered in the containment pressure calcu-lations. 3. Pcssive Heat Sinks a. Identification The passive heat sinks that should be included in the containment evaluation model should be established by identifying those structures and components within the containment that could -3 2183 285 o-2

influence the pressure response. The kinds of structures and components that should be included are listed in Table 1. Data on passive heat sinks have been compiled from previous reviews and have been used as a basis for the simplified model outlined below. This model is acceptable for minimum containment pressure analyses for construction permit applications, and until such time (i.e., at the operating license review) that a complete identification of available heat sinks can be made. This simplified approach has also been followed for operating plants by licensees complying with Section 50.46 (a)(2) of 10 CFR Part 50. For such cases, and for construction permit reviews, where a detailed listing of heat sinks within the containment of ten cannot be provided, the following procedure may be used to model the passive heat sinks within the containment: (1) Use the surface area.and thickness of the primary containment steel shell or steel liner and associated anchors and concrete, as appropriate. (2) Estimate the exposed surface area of other steel heat sinks in accordance with Figure 1 and assume an average thickness of 3/8 inch. (3) Jddel the internal concrete structures as a slab with a thickness of 1 foot and exposed surface of 160,000 ft. The heat sink thermophysical properties that would be acceptable are shown in Table 2. At the operating license stage, applicants should provide a detailed list of passive heat sinks, with appropriate dimensions and properties. b. Heat Transfer Coefficients ,.The following conservative condensing heat transfer coefficients for heat transfer to the exposed passive heat sinks during the 'r, 2183 286 o.3

a blowdown and post-blowdown phases of the loss-of-coolant accident should be used (See Figure 2): (1) During the blowdown phase, assume a linear increase in the condensing heat transfer coefficient from h d" 2 8 Btu /hr-ft _.F. at t = 0, to a peak value four times greater than the maximum calculated condensing heat j transfer coefficient at the end of blowdom, using the l Tagami correlation (Ref. 2), r0.62 h = 72.5 = max Vt lPJ 2 where h = maxi =um heat transfer coefficient, Btu /hr-ft

• F mu Q

= primary coolant energy, Btu .I 3 i V = net free containment volume, ft e t = time interval to end of bicwdown, sec. P l (2) During the long-term post-blowdown phase of the accident, characterized by low turbulence in the containment atmosphere, j assume condensing heat transfer coefficients 1.2 thes greater than those predicted by the Uchida data (Ref. 3) and given in Table 3. (3) During the transition phase of the accident, between the end j of blowdown and the long-term post-blowdown phase, a reasonably conservative exponential transition in the condensing heat transfer coefficient should be assumed (See Figure 2). The calculated condensing heat transfer coefficients based on the above method should be applied. to all exposed passive heat sinks, i both metal and concrete, and for both painted and unpainted surf aces. i i Heat transfer between adjoining materials in passive heat sinks should be based on the assumption of no resistance to heat flow An ava7 e of this is the containment 1 at the material interfaces. liner to concrete interface. I I 2183 787 ,c.e 1 ?1, tdl" D-4

C. REEERENCES 1. 10 CFR Section 50.46, "Acc.cotance Criteria for Emergency Core Cooling Systems for Light Waior Nuclear Power Reactors," and 10 CFR Part 50, Appendix K, "ECCS Evaluation Models." 2. T. Tagami, " Interim ~ Report on Safety Assessments and Facilities Es tablishment Project in Japan f or Period Ending June 1965 (No.1)," prepared for the National Reactor Testing Station, February 28, 1966 (unpublished work). 3. H. Uchida, A. Oyama, and Y. Toga, " Evaluation of Post-Incident Cooling Systems of Light Water Power Reactors," Proc. Third Inter-national Conference on the Peaceful Uses of Atomic Energy, Volume 13, Session 3.9, United Nations, Geneva (1964). 2183 288 ' cat D-5

TABLE 1 IDENTIFICATION OF CONTAINhENT HEAT SINKS 1. Containment Building (e.g., liner plate and esternal concrete walls, floor, and sump, and liner anchors). 2. Containment Internal Structures (e.g., internal separation walls and floors, refueling pool and fuel transfer pit walls, and shielding walls). 3. Supports (e.g., reactor vessel, steam generator, pumps, tanks, major components, pipe supports, and storage racks). 4. Uninsulated Systems and Components (e.g., cold water systems, heating ventilation, and air conditioning cptems, pumps, motors, fan coolers, recombiners,.and tanks). .G, 5. Miscellaneous Equipment (e.g., ladders, gratings, electrical cable trays, and cranes). 2183 289 D-6

LABLE 2 HEAT SINK THERM 0 PHYSICAL PROPERTIES Specific Thermal Density Heat Conductivity Material lb/ft3 Btu /lb *F Btu /hr-ft 'F Concrete 145 0.156 0.92 Steel 490 0.12 27.0 2183 290 4

• h D-7

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APPENDIX E ANALYSIS OF FINANCIAL QUALIFICATIONS The following information provides the details of the financial analyses for the applicants for the South Texas Project, Units 1 and 2. E-1 Houston Lighting & Power Company (HL&P) HL&P is engaged in the generation, transmission, distribution and sale of electric energy in a 5600 square mile area in the Texas Gulf Coast Region. Operating revenues increased from $409.1 million in 1973 to $486.8 million in 1974, although net income declined somewhat to $69.4 million from $71.9 million. Invested capital at December 31, 1974 amounted to $1,380 million and consisted of 53.4% long-term debt, 6.17% preferreu stock, and 40.36% common equity. The return on common equity for 1974 was 11.74%, down from 13.9% in 1973. Pre-tax coverages of long-term interest and total interest charges in 1974 were 4.25 times and 3.52 times, respectively, versus 5.49 times and 4.64 times in 1973. HL&P's first mortgage bonds are rated double-A by both Moody's and Standard and Poor's. Het income for the twelve months ended June 30, 1975 was $62.5 iaillion, compared with $71.6 million for the comparable prior period. HP&L plans to finance its share in South Texas 1 and 2 by the use of internally generated funds and external financing. Available funds from these sources in 1974 totaled $255.7 million and were derived from $164.9 million of long and short-term debt financing and $90.9 million of internally generated cash. The latter represented 35.5% of 1974 construction expenditures. At our request, HL&P supplied a projected sources of funds statement for the 1975-82 period, with underlying assumptions, demonstratinc how the requisite funds might be raised. Its internally generated cash over this period is projected to be 44.51' of total construction expenditures and 395.7% of its expected outlays for auth Texas i and 2. We have reviewed HL&P's pro-jections and find them within the zone of reasonableness. 2183 294 E-1 L:

Applicant: llouston Lichtiric Efower Co. helear Plant: STP I Sources of Funds for Syntem-Wide Conctruction Exnenditures Durinn Period of Construction of Stbket ' ; clear Power Plant k, (thousanus of dollars) n p Conntniction Years of Subject Nuclear Power Plant ~ Security issues and N other funds 197h 1975 1976 1977 1978 1970 1980 1981 1982 A o=on stock $ $ 39,150 $ L7,128 $ 6L,313 $ $ 61,040 $ 28,553 $ 31.380 $ C Preferred stock 40,000 45,000 20,000 35,000 30,000 30,000 35,000 35,000 Long-term debt 188,000 125,000 100,000 125,000 125,000 200,000 175,000 200,000 100,000 Notes payable 13,000 29,789 (3,077) (36,7ho) 6,917 (7,551) (48,307) 13,983 (2h,822) Contributions from parent-net other funds (36,1L8) (702) 23,022 11,h24 16,51h 18,32h 11,527 (28,145) (8,521) Total 164,852 233,237 212,073 183,997 183,431 301,813 196,773 252,218 101,657 Internal funds Net income 69,h06 6h,459 119,067 1hh,867 159,053 170,668 213,520 222,528 2h3,979 n LLess: preferred dividends 5,83C 6,496 13,782 15,930 19,216 21,975 24,884 18,252 31,660 comon dividents 32,628 3L,517 39,0bh 57,304 64,h68 Th,L88 76,hSO 96,768 97,136 Retained earnings 30,9L8 23,Lh6 66,241 71,633 75,369 Th,205 112,156 97,508 115,183 Deferred taxes 15,522 18,509 20,256 21,372 25,176 29,952 37,692 46,176 60,600 Invest. tax cred. defer. 7,L86 7,973 h,609 L,120 11,054 8,88k 10,595 16,212 22,457 Depreciation and anort. 45,149 51,295 55 979 60,533 68,118 75,h65 80,213 97,148 113,547 Less: AFPC 8,228 9,695 16,398 27,750 31,163 37,1L7 51,293 bl,971 30,782 N Total 90,677 91,528 130,187 129,908 146,554 151,359 169,363 215,073 281,005 TOTAL FUNDS $255,729 $324,765 $3L2,260 $313,905 $331,965 $h53,172 $336,136 $h67,291 $382,662 CD Construction Expenditures

• Nuclear power plants

$ 5,625 $ 11,436 $ 39,820 $ 91,269 $ 83,288- $ 75,018 $ 23,677 $ 10,8ho $ 2,625 other 250,104 313,329 302,LLO 222,636 248,697 378,15h 362,h59 h56,h51 380,037 m Total const. exp's. $255,72v $32L,765 $3h2,060 $313,905 $331,985 $453,172 $386,13( $h67,291 $382,662 Subject nuclear plant 5,625 $ 11,h36 $ 39,620 $ 91,269 $ 83,268 $ 75,018 $ 23,6(7 $ 10,8ho $ 2,625

• Exclusive of AFDC (allowance for funds 1. sed during construction)

HOUSTON LIGHTING & POWER COMPANY SOUTH TEXAS PROJECT UNITS 1 & 2 Assunctions Applicable to Source of Funds Statement (a) Rate of return on average common stock equity - approximately 15% (b) Preferred stock dividend rate 3/4% (c) Growth rates: (1) KWH sales (compound growth rate): Residential......... 6.7% Commercial 8.5% Industrial 8.2% Municipal 5.0% Public Utilities.... 6.3% System total 7.8% (2) Revenues, expenses, interest chargos and net income are based on various dependent factors by functional classifications and/or rate of return requirements. Salaries were escalated at rates ranging from 5% to 10% while materials and supplies were escalated at 8%.- (d) Market / book ratio with respect to projected common stock offerings - 1 (e) Common stock dividend payout ratio - 50% (f) Target capitalization ratios: Common stock....... 40% Preferred stock..... 10% Debt............... 50% (g) Fixed charge coverage - minimum 2.5 times (h) long-term and short-term debt interest rates 3/4% 2183 296 1t t8!T E-3

E-2 Cent'al Power and Light Company (CP&L) CPEt., a subsidiary of Central and South West Corporation, is engaged in the generation, purchase, transmission, distribution, and sale of electric energy in 44 counties in south Texas comprising an area of about 44,000 square miles. Operating revenues increased from $161.3 million in 1973 to $223.6 million in 1974, while net income increased from $27.5 million to approximately $29 million. Invested capital at December 31, 1974 amounted to $509 million and consisted of 54.5% long-tenn debt, 8.6% preferred stock, and 36.9% common equity. The return on common equity for 1974 was 14.4%, compared with 15.2% for 1973. Pretax coverages of long-term interest and total interest charges for 1974 were 3.96 times and 3.78 times, respectively, versus 5.66 times and 5.47 times in 1973. CP&L's iirst mortgage bonds are rated double-A by both Moody's and Standard & Poor's. Net income for the twelve months ended June 30, 1975 was $29.5 million, compred with 27.8 million for the comparable prior period after the cumulative effect of an accounting change. CP&L plans to finance its share in South Texas 1 & 2 by the use of inter-nally generated funds and external financing. Available funds from these sourcesin1974 totaled $91.8millionandwerederivedfrom$35.3millionof internally generated cash and $56.5 million of long and short-term debt financing. The inter'nally generated cash in 1974 represented 38.5% of construc-tion expenditures. y ]J E-4

At our request, CP&L supplied a projected sources of funds statement for the 1975-82 period, with underlying assumptions, demonstrating how the requisite funds might be raised. CP&L's internally generated cash over this period is projected to be 44.9% of total construction expenditures and 129.6% of its expected outlays for South Texas 1 & 2. We have reviewed CP&L's projections and find them within the zone of reasonableness. 2183 298 o,; t-s

Applicant: Central Power 8 Licht Co. Nuclear Plant: STP I Sources of Funds for System 'Jide Construction Expenditures During Period of Constructicn of Subiect Nuclear Power Plant (thousands of dollars) Ccnstruction Years of Subiect Nuclear Power Plant Security issues and other funds 197h 1975 lo76 1977 1978 1979 1080 1981 1982 Cormon stock $25,000 $ $ 65,000 $ 54,000 $ Preferred stock 24,000 60,000 Long-I,ern debt h9,825 55,000 60,000 75,000 75,000 50,000 Notes payable (Net) 12,700 (1,100) (10,000) 62,7C0 28,200 (9.200) (55,300) 5,200 (17,200) Contributions fron parent-net Other funds (6,020) 3,309 (12,562) (11,502) (5,582) (9,988). (3,395) 3,h57 (h,879) Total 56,505 27,209 56,438 136,198 142,618 109,812 16,305 8,657 27,921 Internal funds Net income 28,985 31,291 38,415 48,153 59,088 69,079 76,649 67,950 82,020 Less: -Y preferred dividends 2,566 2,566 L,966 h,966 7,966 10,966 10,966 10,966 10,966 common dividends" 16,850 18,205 22,h11 28,935 34,251 38,936 hh,008 33,tT9 k7,606 Retained earnings 9,569 10,520 11,038 lb,252 16,871 19,177 21,675 18,805 23,hk8 g Deferred taxes 6,63h 6,967 8,097 8,996 8,h90 10,576 15,315 16,596 18,664 Invest. tax cred.-deferred 2,910 1,7h0 8,274 3,295 2,883 9,315 5,706 Sho 6,328 m Depreciation and amort. 19,852 22,h10 25,076 27,101 28,147 32,835 ho,880 45,329 49,795 h Less: AFDC 3,656 h,076 h,202 10,5h2 19,816 27,631 20,265 12,202 13,520 Total 35,309 37,561 43,263 h3,102 36,575 hh,272 63,311 69,068 8h,715 N TOTAL FUNDS $91,614 $6h,770 $104,721 $179,300 $179,143 $154,004 $79,616 $77,725 $112.636 Construction exper.ditures* Nuclear power plants $ 5,h87 $10,088 $ 33,310 $ 7h,675 $ 68,1h5 $ Th,h92 $33,182 $22,348 $ 24,411 Other 86,327 Sk,682 71,h11 10h,625 111,0h8 79,592 h6.h3h 55,377 88,225 Total const. exp's. $91,81h $64,770 $10h,721 $179,300 $179,3 93 $15h,08h $79,616 $77,725 $112,636 Subject nuclear plant $ 5,h87 $10,088 $ 33,310 $ Th,675 $ 68,1h5 $ Th,h92 $33,182 $29,697 $ 15,806

• Exclusive of AFDC (allowance for funds used during construction)

CENTRAL POWER AND LIGHT COMPANY Rate of Rucurn on Average Common nui'y 15% e Preferred Stock Dividend Rate 10% forecast on all new issues during construction period Growth Rate in KWH Sales 7.4% Growth Rate-Revenue (Base Revenue = 6.9%) 15.6% Growth Rate-Expenses 16.3% Growth Rate-Interest Chgs. 13.9% Growth Rate-Met Income 13 7% Common Dividend Payment Ratio 67% Tar 3at Canttal Structuro Debt 50% Preferred 15% Common Equity _353_ 1001 Resultint Interest Coverages SEC Formula Hi 4.39 Lo 3.66 Indenture Requirement Hi 3.78 to 2.38 I_nterest Rates 2183 300 Long-tern Debt, b..['a 10% Short-term Debt 9% t E-7

E-3 CITY PUBLIC SERVICE OF SAN ANTONIO The City Public Service of San Antonio is a municipal corporation and political subdivision of the State of Texas which owns, and through the City Public Service Board, operates the San Antonio public utility electric system. The City Public Service Board has full authority for the management and operation of the electric system which serves the City of San Antonio and the surrounding area. As of January 31, 1975, San Antonio's net utility plant, including $104.4 million attributable to its gas operations and $87.6 million to construction work in progress, was $542.9 million. Electric revenues for the year ended January 31,1975 were $137.9 million. San Antonio plans to finance its share of South Texas 1 and 2 by the issuance of revenue bonds and the use of internally generated funds. San Antonio's revenue bonds are rated triple-A by both Moody's and Standard and Poor's. Available funds from these sources totaled $77.8 million for the fiscal year ended January 31, 1975, 42.4% of which were derived from internally generated funds. San Antonio has provided a projected sources of funds state-ment demonstrating how the requisite funds might be raised. Internally gener-ated funds are projected to be 44.3% of total construction expenditures and 154.4% of expected outlays for South Texas 1 and 2. We have reviewed these projections and find them within the zone of reasonableness. 2183 301 E-8

Applicant: City Public fMr/ sce Icard ! uelear Plant: STP I Sources of Funds for Synten-Wide Con struction Expen:litures During Period of Construction cf Subject Huclear Power Plant (thousands of dollars ) Construction Years of Subject Nuclear Power Plant Bond issues and Fiscal Years Ended January 31, other funds 1975 3976 1977 1978 1979 1980 1981 1982 1983 Revenue Bond Issues $ 85,000 $ 50,000 $130,000 $110,000 $ 70,000 $125,000 $140,000 $130,000 $120,000 Other Funds (h&,182 ) Sh,315 (3h,3h9) (21,306) (1h,58h) (25,138) (49,330) (33,387) (37,7h6) Total hh,616 104,315 95,651 63,694 55,416 99,862 90,670 96,613 6.,-,4 Internal Funds Net Inec=e 17,556 ~ 28,68h 27,580 27,736 45,002 38,948 h9,902 59,082 *** 76,h27*** Depreciation and amort. 15,h18 18,185 19,0h6 20 542 27,9hh 28,929 30,563 32,266 39,867 Total

• 32,974 h6,669 h6,626

~E,i,27If 72,946 67,877 80,h65 91.3h8 116,294 TOTAL FUNDS $ 77,792 $151,184 $142,277 $1I6,972 $128,362 $167,739 $17h,135 $187,961 $198,5h8 m Const ruction E>Extenditures* Huclear Power Plants $ 6,726 $ 11,624 $ L1,839 $ 91,620 $ 87,069 $ 98,063 $ 57,834 $ 78,112 $142,3 73 Other** 71,066 139,560 100,L38 h5,352 h1,293 69,676 116,301 1c9,849 56,375 Total const. exp's. 77,792 151,184 IL2.277 136,972 128,362 167.739 174,135 187,961 198,548 Subject nuclear plant $ 6,08h $ 11,321 $ 37.3b2 $ 63,801 $ 76,h73 $ 83,S'7, $ 37,23F $ 22,10h $ 17,739

• Exclusive of AFD0 (allowance for funds used during construction)
  • Includes Transmis-sior Costs Assoc.

vith STP I $ 450 h62 490 519 $ 18,928 $ 12,987 N

• • • Includes Nuclear

$ 11,780 $ 24,02h Fuel Expense C33 U U CD N

CITY PUBLIC. SERVICE BOARD Growth Rates Cover Period 1974-75 -- 1982-83 Growth Rate in KWH Sales 10.5% Growth Rate in Elcetric Customers 2.1% Growth Rate Electric Peak Demand 10.1% (Long Term 1975-76 -- 1989-90=7.18%) Growth Rate in Total Operating Revenue 11.9% (Base Rate =8.7%) Growth Rate in Operating Expenses 12.6% (Based on Individual Trends for Oper. Maint & Common Expenses) Long Term Interest Rate 7.0% Debt 50% Bonding - Net Additions Resultant Interest Coverages - Bond Ordi:ance Requirements - 1.5 Construction Period 2.0 High - 1.6 Low Rate Increases: 1976 1979 1981 10% 10% 15% 2183 403 \\k' k,. 0, ct ?> E-10

E-4 CITY OF AUSTIN The City ti Austin is a municipal corporation and political subdivision of the State of Texas which owns and operates the public utility electric system. Austin has full authority for the management and operation of the public utility electric system in the city of Austin and the surrounding The net utility plant, including construction work in progress of area. $69.3 million, was $169.2 million at September 30, 1974. Electric revenues for the year ending September 30,1974 were $58.9 million. Austin plans to finance its share of South Texas 1 and 2 by the issuance of revenue bonds and the use of internally generated funds. Austin's revenue bonds are rated double-A by both Moody's and Standard and Poor's. In November of 1973, Austin's citizens approved a $397.6 million electric utility bond issue that provided a long term generation expansion program, including participation in the South Texas Project. Austin has provided a projected sources of funds statement demonstrating how the requisite funds might be raised. Internally generated funds are projected to be 56% of total construction expenditures and 166% of expected outlays for South Texas 1 and 2, although additional authority will be needed to issue revenue bonds beyond 1979. We have reviewed these projections and find-them within the zone of reasonableness. 2183 304 w (8Is E-ll

Applicant: City Of A% cin Nuclear Plant: STP I Saurces of Funds for Systen-Wide Construction Menditu es Durins Feriod of Construction of Subject nuclear Pcuer Plant (thousands of dollars) Construction Years of Subject Huclear Power Plant Security issues and other funds 1975 1976 1977 1978 1979 1980 1981 1982 Long-term debt $h5,000 $113,000 $125,000 $ 92,000 $52,000 Total h5,000 113,000 125,000 92,000 52,000 Internal funds Retained earnings 17,700 23,800 29,350 35,725 37,210 hl,150 hh,410 h5,290 Depreciation and amort. 6.200 6,800 7,550 8,380 9,300 10,300 11,h30 12.680 Total 23,900 30,600 36,900 bh,105 h6,510 51,450 55.cho 57,970 TOTAL FUNDS $66,900 $1h3,600 $161,900 $136,105 $98,510 $51,h50 $55,bho $57,970 rU Construction expenditures

• Nuclear power plant

$ 6,h06 $ 21,150 $ 47,k13 $ h3,266 $h7,298 $21,068 $12,506 $10,(35 other h7,797 94,h17 72,861 47,821 33,1ho 36,822 37,354 38,849 Total const exp's. Sh,203 115,567 120,27h 91,0B7 80,h38 57,890 49,860 h9,884 Subject nuclear plant $ 6,h06 $ 21,150 $ 47,h13 $ h3,266 $47,298 $21,068 $12,506 $10,035

• Exclusive of AFDC (allowance for funds used during construction)

N Le4 u CD Ln

City of Austin 1. Long-term debt carried throughout authorized amounts; additional authority i<, needed beyond 1979. Assumed to originate from vocer authority and subsequent sale of revenue bonds. 2. Depreciation and amortization are at current applicable rates. 3. Retained earnings maintained at current ratios. Approximately 50% of net revenue after debt available for retained earnings. 2183 T06 E-13

APPENDIX F ERRATA SOUTH TEXAS PROJECT Page Line 1-5 17 Delete " motor" and substitute " air" 1-7 21 Delete "offsite" and sutstitute "onsite" 2-4 Figure ?.3 Substitute the attached updated Figure 2.3 2-14 5 Delete " Colorado River" and substitute " main cooling reservoir" 2-29 8 After " earthquake." add " Seismic Category I" 3-3 2 After "on" add "the containment" 3-3 7 After "50-446)." add "For other seismic Category I structures, American National Standard Institute document A 58.1 ' Building Code Requirements for Maximum fI5 Design Load in Buildings and other i '< Structures ' is used." 5-3 12 Delete sentence beginning with " Examples" and substitute the following "An Example of ar ASME code Class 3 system is the component cooling water system." 7-1 35 Delete "underpower" and substitute "undervoltage" 11-3 Table 11.1 Under " Boron Recovery System-EVAPORATOR" change Quality Group "C" to Quality Group "D (Augmented)." 2183 307 F-1

Page Line 11-3 Table 11.1 Under " Liquid Waste Processing System" delete line beginning with " Chemical Drain Tank" 11-7 37 Delete sentence beginning with "This gas may be recycled" and substitute the following " This gas will be stored in the tanks for 60 days, sampled and analy1 zed for radioactivity prior to release." 11-6 Table 11.2 Under " Quality Group" add "***" after "D (Augmented)" 11-11 16 Delete " turbine building floor drain discharge line" 12.5-38 Delete sentence beginning with 12.6 8 "In response to" through "a nuclear power plant" 13-2 14 Delete " Plant" and substitute " Assistant" 2183 08 F-2

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APPENDIX G NRC_ Staff Review of the Westinghouse Emergency Core Cooling System Evaluation Model Background In forma *. ion Westinghouse submit ted a description of the Wes tinghouse Emergency Core Cooling System (ECCS) evaluation model on August 5, 1974, in several report s (Ref erences 1 through 18 of Enclosure 2). The NRC staf f reviewed the August submittal f or conformance with the requirements of Appendix K to 10 CFR Part 50 entitled "ECCS Evaluation !!odels," and published it s evaluation in " Status Report by the Directorate of Licensing in the Mat ter of the Westinghouse ECCS Evaluation Model Conformance to 10 CFR Part 50, Appendin K" (Reference 21) dated October 15, 1974. This repo rt addressed each requirement of Appendix K, discussed conformance by Westinghouse, indicated the acceptability of the analytical methods euployed in the West inghouse model, and assessed the impact of specific open items which were either unresolved or unacceptable. Additional documentation was subsequently submitted by Westinghouse addressing these open items and the staff review entitled " Supplement to the Status Report by the Directorate of Licensing in the Matter of Westinghause Electric Company ECCS Evaluation Model Conformance to 10 CFR 50, Appendix K," was published on November 13, 1974 (Reference 22). We concluded that certain modifications, which were described in the above mentioned documents, were required to achieve conformance with Appendix K to 10 CFR Part 50 On October 26 and Novenber 14, 1974, the staff presented its assessment of the Westinghouse evaluation model to the Advisory Committee on Reactor Safeguards (ACRS). In its report to the Chairman of the AEC, dated November 20, 1974, the ACRS concluded that "the four light-water reactor vendors have developed evaluation models which, with additional modifications required by the staf f, will conform to Appendix K to Part 50." The required model changes which were subsequently implemented by Westinghouse into their evaluation model included the following: 1 - Deletion of rod-to-rod radiation during blowdown. 2 - Revision of the injection section pressure drop. 3 - Deletica of :he 10% clad strain burst criterion. 4 - Inclusion of a hot wall time delay. 5 - Modifications to the steam cooling model for reflood rates less than 1 inch /second. I i'. 8 K 2183 310 G-1

On December 27, 1974, the Commission, in response to licensee submittals of additional information from operating Westinghouse plants, pursuant to Section 50.46 and Appendix K to 10 CFR Part 50, issued a Safety Evaluation Report and Orders for Modification of Licenses pertaining to the latest proposed Technical Specifications. In addition, the Commission requested that the above modifications to the Westinghouse evaluation model be made and that a reanalysis be submitted within six months. On April 17, 1974, Westinghouse formally submit ted proprietary and non-proprietary versions of a topical report (References 19 and 20) which documented all of the modifications required by the staf f in October and November 1974 (References 21 and 22). ' In add it ion to the modifi-cations required by the staf f, Westinghouse submitted supplemental information to complete the documentation requirements of Appendix K, including respoases to questions raised by the staf f in the course of reviewing the Westinghouse ECCS evaluation model, refinements to the steam cooling mooel, and documentation of minor modifications which had no significant effect on computational results. _ Conclusions The NfiC staff has completed it s review of the Westinghouse ECCS eval-uation model which is comprised of References 1 through 20 We closely followed the development of the Westinghouse ECCS evaluation model and ut ilized t he referenced reports to determine the compliance of the Westinghouse evaluation model with Appendix K to 10 CFR Part 'O. The details of our review are summarized in References 21 and 22. We conclude: 1. That the Westinghouse evaluation model is an acceptable model.to be used for ECCS performance evaluation for plants which satis fy the following plant classifications: Typical current Westinghouse two, three, and four loop a. plants. b. Dry, subatmospheric or ice containments. c. Power ratings up to 3800 MWt. d. Plants utilizing only bottom flooding emergency core cooling systems. 2. That References 1 through 20, which constitute the consolidated description of the Westinghouse ECCS evaluation model, may be incorporated by reference in licensing applications as an accepted ECCS evaluation model. This acceptence applies to the Westinghouse ECCS model and does not const it ute acceptance of the individual reports for any purpose other than for ECCS analyses. 2183

11 G-2

3. The app.lication of the LOTIC code for purposes of ECCS performance evaluation, in determining the minimum containment pressure response for all ice containments, will be evaluated on a plant-by plant basis. Until such time that LOTIC is modified to resolve the staff concerns, noted in references 21 and 22, a conservativ; minimum containment pressure of zero psig must be assumed in ECCS analyses of plants using an ice condenser containment 2183 312 .. a t G-3

REFERENCES 1. WCAP-8 200, Rev. 2, WFLASH - A Fort ran IV Computer Program for Simulation of Transients in a Multi-Loop PWR," (Proprietary), June 1974. 2. WCAP-8261, Rev. 1, "WFLASil - A Fort ran IV Computer Program for Simulation of Transients in a Multi-Loop PWR," (Non-proprietary), July 1974. 3. WCAP-8359 "Ef fects of Fuel Densification Power Spikes en Clad "Ihermal Transients," (Non-proprietary), July 1974. 4. WCAP-8354 "Long-Term Ice Condenser Containment Code - LOTIC Code," (Proprietary), July 1974. 5. WCAP-8355 "Long-Term Ice Condenser Containment Code - LOTIC Code," (Non-proprietary),, July 1974. LOi-6. WCAP-8340 "Wes t inghouse ECCS - Plant Sensitivity Studies," (Proprietary), Jaly 1974. 7. WCAP-8356 " West inghcuse ECCS - Plant Sensitivity Studies," (Non-proprietary), July 1974. 8. WCAP-8327 " Cont ainment Pressure Analysis Code (C0CO)," (Proprietary), July 1974. 9. WCAP-8326 " Containment Pressure Analysis Code (C0CO)," (Non-proprietary), July 1974. 10 WCAP-8302 " SATAN-VI Program: Comprehensive Space-Time Dependent. Analysis o f Loss-of-Coolant," (Proprietary), June 1974. 11. WCAP-8306 "SATAH-VI Program: Comprehensive Space-Time Dependent Analysis o f Loss-of-Coolant," (Non-proprietary), June 1974. 12. WCAP-8170 "Calculat ional Model for Core Reflooding Af ter a LOCA (WREFLOOD Code)," ( Proprietary), June 1974. 13. WCAP-8171 "Calculat ional Model for Core Reflooding Af ter a LOCA (WREFLOOD Code)," (Non-proprietary), June 1974. 14. WCAP-8301 "LOCTA-IV Program: Loss-of-Coolant Transient Analysis," (Proprietary), June 1974. 15. WCAP-8305 "LOCTA-IV Program: Loss-of-Coolant Transient Analysis," (Non-proprietary), June 1974. 2183 H 3 G-4

16. WCAP-8339 " West inghouse ECCS Evaluation Model - Summary," (Non-proprietary), June 1974. 17. WCAP-8341 West inghouse ECCS Evaluat ion Model Sensitivity Studies," (Proprietary), July 1974. 18. WCAP-8 342 West inghouse ECCS Evalua t ion Model Sensit ivit y Studies," (Non proprietary), July 1974. 19. WCAP-8471 " West inghouse ECCS Evaluation Model - Supplementary Infor-mation," (Proprietary), January 1975. 20. WCAP-8472 West inghouse ECCS Evaluat ion Model - Supplementary Infor-mat ion," (Non-proprietary), January 1975. 21. " Status Report by the Directorate of Licensing in the Matter of Westinghouse Electric Company ECCS Evaluation Model Conformance to 10 CFR 50, Appendix K,' October 15, 1974. 22. " Supplement to the Status Report by the Directorate of Licensing in the Mat ter of Westinghouse Electric Company ECCS Evaluation Model Conformance to 10 CFR 50, Appendix K," November 13, 1974. 218,3 $14 G-5}}