ML19321A570

SER for Facility,Suppl 1 to NUREG-75/082

ML19321A570
Person / Time
Site:	Sterling
Issue date:	04/30/1976
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-76-0052, NUREG-76-52, NUDOCS 8007230637
Download: ML19321A570 (55)

Text

{{#Wiki_filter:, Y (Suppl.1 to NUREG 75/082) Evaluation Itepori

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related to construction of oveic g,eNyggor Stzrling Power Project Nuclear Unit No.1 Docket No. STN 50-485 Rochester Gas and Electric April 1976 Corporation, et al Supplement No.1 e 80 7 c637 l 1

I Available from National Technical Information Service Springfield, Virginia 22161 Price: Printed Copy $4.50 ; Microfiche $2.25 P 5-

NUREG 76/0052 (Supplement 1 to NUREG 75/082) ' April 14, 1976 i ' SUPPLEMENT N0. 1- -TO THE. SAFETY EVALUATION REPORT BY THE- .0FFICE OF NUCLEAR REACTOR REGULATION U.S. NUCLEAR REGULATORY COMMISSION 1 IN THE MATTER OF i ROCHESTER GAS AND ELECTRIC CORPORATION, ORANGE AND ROCKLAND' UTILITIES, INCORPORATED, CENTRAL HUDSON GAS AND ELECTRIC CORPORATION l l . AND NIAGARA M0 HAWK POWER CORPORATION STERLING POWER PROJECT' NUCLEAR UNIT NO. 1 i DOCKET NO. STN 50-485 1 r A T

TABLE OF CONTENTS Page

1.0 INTRODUCTION

AND. GENE RAL DISCUSSION...................................... 1-1 1.1 Introduction.......................... .............................. 1-1 2.0 SITE CHARACTERISTICS........................................ ............. 2-1 2.4.3 Cooling Water................................................. 2-1 2.6 Foundation Engineering............................................... 2-3 3.0 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS AND COMP 0NENTS.................... 3-1 3.7 Seismic Design....................................................... 3-1 3.7.1 Seismic Input................................................. 3-1 5.0 REACTOR COOLANT SYSTEM.......................................... ......... 5-1 5.6 Steam Generator Tubes.............................................. . 5-1 6.0 ENGINEERED SAFETY FEATURES................................................ 6-1 6.2 Containment Systems.................................................. 6-1 6.2.1 Containment Functional Design......................... 6-1 6.3 Emergency Core Cooling System........................................ 6-1 6.3.3 Performance Evaluation........................................ 6-1 6.3.5 Conclusions................................................... 6-3 7.0 INSTRUMENTATION AND CONTR0LS.............................................. 7-1 7.3 Engineered Safety Features initiation and Actuation Systems......... 7-1 8.0 ELECTRIC POWER SYSTEMS.................................................... 8-1 8.5 Fire Stops and Sea1s................................................. 8-1 i )

TABLE OF CONTENTS (Continued) Page 9.0 AUXILIARY SYSTEMS......................................................... 9-1 9.3 Wa t e r Sy s t ems........................................................ 9-1 9.3.1 Es senti a l Se rvi ce Wa ter System................................ 9-1 9.3.3 U l ti ma t e He a t S i n k................... c........................ 9-2 11.0 RADI0 ACTIVE WASTE MANAGEMENT............................................. 11 -1 11.1 Su mma ry De s c r i p t i o n............................................... 11 - 1 14.0 INITI AL TESTS AND OPERAT IONS........................................... 14-1 15.0 ACCIDENT ANALYSIS....................................................... 15-1 15.3 A c c i d e n t s......................................................... 15 - 1 15.5 Los s-o f-Coola nt Accident Dose Mode 1................................ 15-1 17.0 QUALITY ASSURANCE 17-l 17.1 Ge n e ra 1............................................................ 17 - 1 18.0 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS................... 18-1 - 19.0 COMMON DEFENSE AND SECURITY.............................................. 19-1 20 0 F I N ANC I AL QUAL I FI CAT IONS................................................. 20-1 20.1 Introduction................................................. .... 20-1 20.2 Rochester Gas and Electric Corporation........................... 20-2 20.3 Orange and Rockland Utili ties. Incorporated....................... 20-3 20.4 Central Hudson Gas and El ectric Corporation........................ 20-3 20.5 Ni a ga ra Mohawk Power Corporation.................................. 20-9

21.0 CONCLUSION

S........................................................ ..... 21-1 11

APPENDICES Page Appendix A CHRONOLOGY (CONTINUATION) 0F RADIOLOGICAL REVIEW............... A-1 Appendix B REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS.......... B-1 Appendix C EVALUATION OF LIQUID AND GASEOUS EFFLUENTS FROM STERLING FACILITY WITH RESPECT TO APPENDIX I TO 10 CFR PART 50......... C-1 Appendix D CHANGES AND CORRECTIONS TO THE SAFETY EVALUATION REPORT......... D-1 111

e LIST OF FIGURES Page Figure 2.1 SERVICE WATER SYSTEM............................................ 2-2 LIST OF TABLES Table 20.1 SOURCES OF FUNDS, ROCHESTER GAS AND ELECTRIC................... 20-4 Table 20.2 ASSUMPTIONS FOR SOURCES OF FUNDS.., ROCHESTER GAS AND ELECTRIC... 20-S Table 20.3 SOURCES OF FUNDS, ORANGE AND ROCKLAND......................... 20-6 'R SOURCES OF FUNDS, ORANGE AND ROCKLAND.......... 20-7 Table 20.4 ASSUMPTIONS Table 20.S SOURCES OF FU5!DS, CENTRAL HUDSON GAS AND ELECTRIC.............. 20-10 Table 20.6 ASSUMPTIONS FOR SOURCES OF FUNDS, CENTRAL HUCSON GAS AND ELECTRIC..................................................... 20-11 Table 20.7 SOURCES OF FUNDS, NIAGARA M0 HAWK.................... 20-13 Table 20.8 ASSUMPTIONS FOR SOURCES OF FUNDS, NIAGARA M0 HAWK...... ....... 20-14 iv

1.0 INTRODUCTION

AND GENERAL DISCUSSION .l.1 . Introduction - The Nuclear Regu!atory Comission's (Comission) Safety Evaluation Report in the matter of the application by the Rochester Gas and Electric Corporation to construct land operate'the proposed Sterling Power Project Nuclear Unit No.1 (facility) was issued on September 5, 1975. In the Safety Evaluation Report, we indicated certain s matters.(1) where additional information was required of the applicant, and (2) where our review was not yet complete. The purpose of this supplement is to update the Safety Evaluation Report by addressing the outstanding issues identified in Section 1.8 of the Safety Evaluation Report and by providing our evaluation of the additional information submitted for the applica- ~ tion since the issuance of the Safety Evaluation Report, and to address the comments made by the Advisory Comittee on Reactor Safegu ards in its report of October 16, 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 tht same as the section of the Safety Evaluation Report that is being updated and erept where specifically noted, is supplementary to and not in lieu of the discussion of the Safety Evaluation Report. Since the issuance of the Safety Evaluation Report, Rochester Gas and Electric has documented, in Amendr:ent 23 ?.- (ts application, the plan to share ownership of the proposed St:rling fccility with Orange and Rockland Utilities Incorporated Niagara Mohawk Power Corporction and Central Hudson Gas and Electric Corporation. In Amendment 3' to the 1pplication, all utilities sharing ownership of the Sterling facility wer designated as applicants. Rochester Gas and Electric has specified that. It retains f6 'l responsibility for the construction, operation and licensing of the facility. Appendix A to this supplement is a continuation of the chronology of our principal actions related _to the processing of the application. The report of the Advisory Comittee on Reactor Safeguards is attached as Appendix B. Appendix C is a detailed surnr.ary of our Appendix I evaluation, including source terms and doses. Appendix D - is a listing of corrections and changes to the Safety Evaluation Report. 'l-), A- -.3 _J..,-

2. 0 SITE CHARACTERISTICS 2.4.3 Cooling Water In the original cooling water intake design, as described in our Safety Evaluation Report, normal cooling water was to be supplied through a submerged intake structure located about 4200 feet from the shoreline. The water was to be discharged through a seismic Category I discharge canal located at the shoreline. Within the plunge basin of the discharge canal were to be two 30-inch diameter intakes through which emergency cooling water could be drawn.

In' the Safety Evaluation Report, we noted that Rochester Gas and Electric had not properly considered the potential of' blockage of the emergency cooling water intakes by littoral drift or ice buildups. We approved this original design with the condi-tion that Rochester Gas and Electric must justify to our satisfaction that littori.1 drift or ice buildup would not preclude an adequate supply of emergency cooling water to the plant at all times. If it could not be so justified, an alternate method of emergency cooling would be submitted for our approval. Rochester Gas and Electric has conducted field and analytical studies of littoral drift and ice buildup and is not certain that littoral drift or ice buildup will not cause blockage at the intakes. Therefore, a revised cooling water intake design was submitted for our review. In the revised design (see Figure 2-1), the applicants now propose to design the circulating water intake structure and tunnel to seismic Category I requirements and to provide a flow path to the seismic Category I service water pcmp structure via two 30-inch dianeter seismic Category I pipes connecting the tunnel to the forebay of the pump structure. In addition, an alternate flow path to the service water pump house will be provided by a 42-inch diameter pipe originating in the plunge basin of the circulating water discharge structure, thus meeting the single fallere criterion of Regulatory Guide 1.27. " Ultimate Heat Sink for Nuclear Power Plants." The applicants have committed to monitor the discharge canal for material buildup and maintain the canal to preclude blockage of the 42-inch diameter pipe. The details of the monitor-ing and maintenance program will be provided at the operating license stage of our review. We have reviewed the proposed cooling water intake design and conclude that an adequate emergency cooling water supply can be provided in the event of natural and postulated accidental phenomena, including floods, surges, seismic events, shipping accidents, littoral drift buildup, ice buildup and reasonable combinations thereof, as recommended in Regulatory Guide 1.27 (see Sections 2.6, 3.7.1, 9.3.1, 9.3.3 and 15.3 of this supplement). We, therefore, find the proposed design acceptable. 2-1

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, (located south cf the plant) should be prot:cted from Erosign. The applicants com- ' mitted to provide and have now specified the erosion protection of the structure to l prevent undennining. This protection will be provided by the use of dumped riprap, which will be placed where the flow enters the site drainage system. Based on our independent analysis, we conclude that the proposed protection is t.ceptable. ~ In the Safety Evaluation Report we stated that we would require Rochester Gas and Electric to commit to providing the final design of the shore protection for our review before the start of construction of the protection. In Revision 16 to the Site Addendum Report, the applicants have now committed to provide the required design details for our review and approval before the start of construction of the shore . protection. We find this comitment acceptable. 2.6 ' Foundation Engineering In the Safety Evaluation Report we stated that we would report the results of our evaluation of the slope stability of the seismic Category I discharge channel (which then served as the emergency service water intake). However, as described in Section 2.4.3 of this report, the applicants have removed the Category I classification of the discharge channel. As a result of this change we find that there are no natural or man-made slopes whose failure could adversely affect the safe shutdown of the facility. The revised cooling water intake design specified that the circulating water intake tunnel be designed to seismic Category I. The applicants have provided confirmatory data to demonstrate that the proposed seismically Category I designed intake tunnel is founded in and on tne glacial till and sandstone bedrock which support the seismi-cally Category I designed structures in the plant area. As reported in the Safety t Evaluation Report, we conclude that these deposits provide satisfactory support for plant structures and components designed to seismic Category I requirements. 2-3~ ~

r -- 3.0 DESIGN CRITERIA' FOR STRUCTURES. SYSTEMS ' AND COMPONENTS 3.7 Seismic Design 3.7.1 Seismic Input We have reviewed the applicants' methods for the seismic response analysis of the useismic Category.I offshore circulating water intake structure and tunnel, as pre-sented.in Revision 15 to the Site Addendum Report. We have evaluated the applicants' method for determining soil structure interaction of both rock and soil supported s tructures. On the bases of our evaluation, we conclude that the applicants'. pro-- visions '6r designing the circulating water. intake structure and tunnel to seismic - Category, are ecceptable. 3-1 E:

Gp 5.0 REACTOR COOLANT SYSTEM 5.6 Steam Generator Tubes We have evaluated.the measures that will be taken to assure that the steam generator tubes in the Sterling plant will not be subjected to conditions that will cause degradation of tube wall integrity. We have also evaluated the provisions made by the applicants to detect such degradation, should it occur, before it has progressed far enough to affect the safety of the plant. The facilities, steam generators, and operating procedures described in the construc-tion permit application for the Sterling plant are of more recent design than those plants that have experienced steam generator tube degradation. Nuclear steam supply system vendors of pressurized water reactors that have experienced significant steam generator tube corrosion have redesigned steam generators and have made signifi-cant changes in the secondary system water chemistry so that the new pr?ssurized water reactor plants, including the Sterling plant, should not experience degradation. The affected nuclear steam supply system vendors are obtaining experimental data on compatibility of the tube material with simulated secondary coolant conditions. In -Revision 14 to the SNUPPS Preliminary Safety Analysis Report, the applicants have submitted a design shange, consisting of the addition of a condensate cleanup system to the secondary coolant system to maintain the chemical purity of feedwater to the ' steam generators. With this design change approximately 68 percent of the feedwater will be processed by six 20-percent capacity demineralizer units (one of which will be - on standby) containing deep-bed regenerable mixed strong acid cation / strong base anion resins. The demineralizer units will remove condenser impurities by ion ex-change. ' Filtration will also be provided to remove corrosion products. With this ~ design change, the anticipated concentrations of the various corrosion products and fission products in the feedwater to the steam generators will be reduced by about 75 percent or more from the concentrations that would be present without this system. The condensate demineralizer regeneration wastes will be processed (whenever the radioactivity is above a pre-determined level) by the secondary liquid waste system, which is discussed in Section 11.1 of this supplement. For the Sterling plant steam generators, current regulatory requirements in combina-tion with the applicants' planned provisions for detection of degradation are con-sidered sufficient at the construction permit stage of review to assure plant safety. If future Commission action on this issue or future inspections of operating Westing-house steam generators develop significant safety issues concerning design features of systems or components for which preliminary designs are proposed in this applica- ' tion, post-construction permit design changes may be required of the applicants. 5-1 1

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. We conclude'that plant safaty will not be compromised by steam gInerat:r. tube degradation. ~In summary, our conclusions are based on the following considerations:

1. LThe steam generators will be of advanced design with improved secondary water

~ flow character' *ics, providing more tolerance for occasional lack of water -chemistry cr 2. 'The applicants will use an all volatile type of water chemistry that has been shown by service experience to minimize the probability of tube degradation. 3.' The, applicants will use a' condensate cleanup system. 4. Provisions for monitoring the secondary water chemistry will be included. These will be used to detect the presence of deleterious impurities'before significant = tube degradation can occur. 5. Provisions.for monitoring reactor coolant leakage to the secondary side are included in the design' and the limits on such leakage that will be imposed will assure that tube degradation, should it occur, will be detected before loss of integrity results in tube failure. ' 6. The design of the steam generators permits inservice inspection of_the tubes by methods that will detect tube degradation. Tubes.that could degrade to marginal conditions can be taken out of service by plugging. 9 5-2 a

6.0 ENGINEERED SAFETY FEATURES 6.2 Containment Systems 6.2.1 Containment Functional Design In the Safety Evaluation Report we stated that we had been reviewing recent infonnation sutznitted by Rochester Gas and Electric on the containment response to postulated main steam line breaks inside containment. Suisequently, the applicants submitted addi-tional infonnation which provided the results of the analysis of this postulated accident using a conservative mass and energy release model. We have now completed our evaluation of all the infonnation submitted by the applicants and our findings are presented below. The applicants have analyzed the containment pressure response for postulated main steam line failures inside the containment by considering several categories of breaks to provide an envelope of breaks. For the first break category, the most severe case was a 1.4 square-foot steam line break which resulted in the maximum containment temperature. This break corresponds to the size of t.'e flow restr ctor located in the steam generator. In this first category, reliance is placed on a signal generated by the primary protection system for main feedwater isolation and steam line isolation. Maximum containment pressure for the break categories considered was calculated for the second break category. For this break category, a break size of one square foot was considered and reliance is placed on the high containment pressure signal for main feedwater isolation and the mid-high containment pressure signal for steam line isolation. We have reviewed the calculational methods used by the applicants and have found them acceptable. In addition, we have perfonned confinnatory containment pressure and temperature calculations for these accidents and find that our results are in good agreement with those of the applicants. The applicants calculated a peak containment pressure of 48 pounds per square inch gauge for the limiting steam line failure. This is lower than the peak calculated containment pressure for the primary coolant system cold leg, pump suction double-ended rupture and is, therefore, acceptable since the calculated margin is greater than what we have accepted for the postulated loss-of-coolant accident case. 6.3 Emergency Core Cooling System 6.3.3 Perfonnance Evaluation In the Safety Evaluation Report we stated that we would report on our evaluation of the emergency core cooling system analysis to demonstrate compliance with the requirements of Appendix K to 10 CFR Part 50 (Final Acceptance Criteria), after Rochester Gas and Electric had submitted the analysis. This analysis has now been submitted. 6-1

- The applictts submitted cnalysis f4r liss-sf-coolant accidents that addristed small f breaks and majIr reactor cool 6nt system piru breaks. Fcr thilargn breaks, a spectrum ^ ' of three break sizes was submitted that'was specifically analyzed for the SNUPPS plants; To supplement this analysis, the applicants incorporated by reference Westinghouse Topical Report WCAP-8566, "bestinghouse ECCS Four-Loop Plant (17 x 17) . Sensitivity Studies", which covered other treak sizes, types and locations in confor- - mance with the break spectrum requirements in Section 50.46(a) of 10 CFR Part 50. The ' analyses submitted were perfcnned with an acceptable Westinghouse evaluation model which is in confonnance with Appendix K to 10 CFR Part 50, as-documented in the Consnission's letter to the Westinghouse Electric Corporation, dated May 30, 1975. The analyses submitted by the applicants identified the worst break' size as the double- = ended cold leg guillotine break with a Moody multiplier of 0.6. - The calculated peak clad temperature was 2178 degrees Fahrenheit which is below the acceptable limit of 2200 degrees Fahrenheit as specified in Section 50.46(b) of 10 CFR Part 50. In addi-tion, the calculated maximum local metal / water reaction of 7.6 percent and total core wide metal / water reaction of less than 0.3 percent are well below the allowable limits of 17 percent and one percent,-respectively. The analyses were performed based on an assumed total peaking factor of. 2.32, 102 percent of the rated nuclear steam supply system power level of 3411 megawatts thermal, and 102 percent of the peak linear power density of 12.6 kilowatts per foot. These analyses also assumed that there was a coin-cident loss of offsite power at the initiation of the loss-of-coolant accident, which t would result in pump coastdown and, therefore, increase the peak clad temperature. The assumptions made by the applicants for the containment net free volume, passive heat sinks, and operation of the containment heat removal systems, result in a canser-vative containment backpressure for the emergency core cooling system analysis is discussed in Section 6.2.1 of the Safety Evaluation Report. Appendix K to 10 CFR Part 50 of the Commission's regulations also requires that the combination of emergency core cooling subsystems to be assumed operative shall be those available after the most. severe single failure of emergency core cooling system equip-ment has occurred. The worst single failure was identified by the applicants as the loss of a low pressure emergency core cooling system pump, which provided within a consistent set of assumptions (1) the maximum containment cooling and a reduction in emergency core cooling flow and (2) the maximum calculated peak clad temperature. The applicants proposed to use a single valve in the line between the refueling water storage tank and the suction side of the safety injection pumps. This valve did not meet our single failure criterion. We informed the applicants of the unacceptability of the single valve and stated that the provision of a parallel valve would resolve our concern. The applicants have committed to such a provision but have indicated their 6-2 I

intent to continue Ciscussi:ns t3 convinct us cf the acceptability of altirnativt pr; visions. In the event that-the applicants provice informatirn t3 cenvince us that an alternative design provides an equivalent degree of protection, we will permit the alternative design to be used. With the applicants' comitmert we consider this matter

resolied.

. With t;,. above recommended condition for val.e redundancy, we can conclude that the emergency core cooling system perfomance will be adequate in the event of any postu-lated failure of a single component. We have also reviewed the proposed procedures and the system design for preventing . excessive boric acid buildup in the reactor vessel during the long-tem cooling phase ~ following a loss-of-coolent accident. One of the following two procedures would be implemented af ter the; initial cold leg injection: (1) simultaneous injection of borated water into the hot and cold legs, or (2) alternate injection of borated water into the hot and cold legs with injection times short enough to prevent excessive buildup of boric acid. We find that the procedures and the systems available for implementing the procedures will prevent the concentration of the boric acid in the core region from

exceeding solubility limits. At the operating license stage of review, the applicants will justify the injection times to be used in the long-tem cooling procedures follow-ing a postulated loss-of-coolant accident, and will demonstrate that all of the per-tinent valves are qualified to operate in their post loss-of-coolant accident environment.

4 The analyses presented by the applicants were based on the plant operating with all four loops. At the operating stage of review, the applicants will submit analyses to address plant operation with less than four loops. At that time, the Technical Specifications will incorporate the appropriate limits on plant operation to be consistent with the required perfomance of the emergency core cooling system, including any required limitations on plant operation with idle loops. 6.3.5 Conclusions On the basis of our review of the information submitted by the applicants, we have concluded that (1) the loss-of-coolant accident analyses that were performed are repre-sentative of the SNUPPS plants design and ere wholly in conformance with the require-ments of Appendix K to 10 CFR Part 50, (2) the emergency core cooling system performance confoms to the peak clad temperature and maximum oxidation and hydrogen generation criteria of Section 50.46 to 10 CFR Part 50, (3) with the' addition of the redundant - valve discussed in Section 6.3.3 of this supplement, the emergency core cooling system performance will be adequate despite any postulated failure of a single component, and (4)adequatesystemsareavailabletoprovidelong-termcorecooling. Therefore, we conclude that with the addition of a redundant valve, the design of the Sterling i plant emergency core cooling system is acceptable. Further, we conclude that the applicants will comply with the Final Acceptance Criteria. 6-3 w

w .7.0 INSTRUMENTATION AND CONTROLS -7.3' Engineered Safety' Features Initiation and Actuation Systems I'n the Safety Evaluation Report we stated that.we had requested the applicants to address for their application, their resolution of the deficiencies of Westinghouse Report WCAP-7705,." Engineered Safeguards Final Device Actuator Testing", identified in our letter to Westinghouse, dated July 14,1975. The applicants have now provided their resolutten, for the SNUPPS plants, of the deficiencies identified for WCAP-7705. The information provided includes the proposed testing of valves within the engineered - safety features system. The applicants also state that the main steam isolation valves and the main feedwater isolation valves and associated actuators and controls are of a design which permits partial stroking (closing) of the valves, thus providing a - capability for testing during reactor operation. We have reviewed the 'information provided and conclude that the applicants' proposed ~ resolution of the WCAP-7705 deficiencies is acceptable for the SNUPPS plants since the resolution proviaes for a capability of periodically testing containment isolation valves. In the Safety Evaluation Report we stated that we were reviewing the applicants proposed resolution for meeting the single failure criterion for manually-controlled, electrically-operated valves. We find that the ap,,licants' proposal, which is to lock out power to the controller for these valves, is in anfonnance with our position, attached as Appendix C to the Safety Evaluation Report. Therefore, we conclude that the proposed resolution is acceptable. 7-1

d 1 -- 8.0 ELECTRIC POWER SYSTEMS' l '8,5 Fire Stops and' Seals ~ f In'the Safety Evaluation Report, we ' tated that we had requested the applicar to s provide information regarding design criteria and procedures for the fire stop, and seals to be used in the electrical design _ of the facility. The applicants heve pro- ~ vided the requested infonnation and we have now completed our review. _ Based on our review of the information provided, we have determined that the design of the facility will provide for physical separation of redundant control. and protec- . tion system cables 'use of nonpropagating cable insulation materials, fire stops be- ' tween rooms and floors'and includes fire protection and detection systems as well as inspection capabilities for use both during construction and after the start of plant operation. We conclude that the electrical design will incorporate features that w'ill minimize the spreading of fires which could damage safety-related electrical equipment and associat-ed circuitry and, therefore, is acceptable. 8-1

- 9.0 AUXILIARY SYSTEMS 9.3 Water Systems Sections 9.3.1 and 9.3.3 replace the corresponding sections in the Safety Evaluation Report. 9.3.1 Essential Service Water System The essential service water system will provide cooling water for plant components that require cooling for safe shutdown of the reactor following postulated accidents. -These components include component cooling water heat exchangers, containment air coolers, diesel generator coolers, safety injection pump room coolers, residual heat removal pump room coolers, containment spray pump room coolers, centrifugal charging pump room coolers, component cooling water pump room coolers, auxiliary feedwater pump room coolers, control room air conditioning system cor,densers, Class IE switchgear air conditioning system condensers, penetration roam coolers and the station air compressor. The essential service water system will also provide emergency makeup to the spent fuel pool and component cooling water systems and the backup water supply to the auxiliary feedwater system. As described in Section 2.4.3 of this supplement, the applicants have revised the circulating water system and essential service water system designs. The revision to the circulating water system changed the design requirements of the circulating water intake structure and tunnel from non-seismic Category I to seismic Category I. Essential service water in the present design will be supplied from the circulating water intake tunnel instead of being supplied from the circulating water discharge canal as previously designed. The essential service water system includes four 100-percent (17,500 gallon per minute) capacity pumps, any one of which is capable of supplying water needed for the safe shutdown of the reactor. Controls will be pro-vided to automatically start the pumps from the diesel generators in the event of a loss of offsite power. The essential service water.ystem pumps will take suction from the fo.coaf of the service water pump structure. Water will be normally supplied to the pump structure from the seismic Category I circulating water inlet tunnel through two 30-inch diameter seismic Category I pipes. A 42-inch non-seismic Category I pipe from the non-seismic Category I circulating water discharae structure and canal will be provided for warm water recirculation to the service water pump structure to prevent freezing of the ~ emergency service water intake. This line, which is provided with a remotely con- -trolled motor operated valve, will also serve as an alternate intake. 9-1

During normal operation,'the cssentia'l s:rvic) wat:r will'be discharged to th) l circulating wat r return pipes which util discharga to the~ laka th.ough th2 dischargt ' canal.1During emergency operation, the essential service water wil) be returned,

through two.30-inch seismic Category I lines, to a ditch which dischaiqes to the

~ lake, or to the forebay of the-service water pump structure to prevent he accumula-tion on the screens during severe winter weather. The motor-operated valves located in the essential service water diversion structure can be actuated from the control room or locally.- The essential service water system will serve two ident; cal trains of engineered safety feature equipment, of which only one is requi'ed for the safe shutdown of the plant. Either essential service water header will be capable of supplying the required cooling water flow to either train by means of a cross-connection at the essential service water pump discharge headers. The essential seismic water pumps will be located in a seismic Category I pumphouse, which is designed to protect the pumps against flooding, tornadcas and tornaco missiles. All other parts of the essential service water system in the standardized plant design will also be designed as seismic Category I. All structures and components of the system will be located such that the failure of any non-Category I structure would not constitute a hazard to the essential service water system. Based on our review, we conclude that the design criteria and bases for the service water syftem conform to the applicable recommendations of Regulatory Guide 1.29, " Seismic Design Classification." The design criteria and bases also meet the re-quirements of Criterion 44 of the General Design Criteria regarding ability to transfer heat from safety related components to the ultimate heat sink and, Criteria 45 and 46 of the General Design Criteria regarding tests and inspections. We, therefore, conclude that the design of the service water system is acceptable. 9.3.3 Ultimate Heat Sink Lake Ontario will be the ultimate heat sink for the proposed Sterling facility and is connected to the facility by an intake structure, an inlet tunnel a screenwell and pump structure, a discharge canal and associated piping. The ultimate heat sink will provide cooling water for use in the service and essential service water systems during normal operation and shutdown, and for use in the essential service water system for emergency shutdown in the event of an accident or loss of offsite power. Lake Ontario provides an unlimited cooling wi.ter supply at temperatures consistent with the design requirements of the essential service water system. The maximum anticipated lake water temperature is 75 degrees Fahrenheit which is considerably below the essential service water system design temperature of 95 degrees Fahrenheit. The - ultimate heat sink will be capable of providing sufficient cooling water to permit safe shutdown ~ and cooldown of the plant under postulated natural phenomena such as tornado, flood,' draught and ice blockage. 9-2 a:

- As d; scribed in Sections 2.4.3 and 9.3.1 cf this supplement, heated watzr will nonnally .be discharged through the discharge canal. During emergency operati:n the watsr will be discharged through two redundant 30-inch diameter seismic Category I pipes to a drainage ditch located to the east.of the proposed plant. On the basis of our review f of the proposed plant design. we conclude that potential recirculation of heated - discharges is effectively precluded. . e have reviewed the applicants' design criteria and L ses, and the single-failure W analysis-and determined that they follow the heat-sink guidelines of Regulatory ' Guide 1.??, We, therefore, conclude that the system design is acceptable. = I m 9-3 Y

p 11.0 RADIOACTIVE WASTE MANAGEMENT 11.1 Summary Description We stated in the Safety Evaluation Report that we had not completed our review of the radioactive waste management systems to meet the requirements of Appendix I to 10 CFR Part 50. Prior to completing our review, the applicants revised their radioactive - waste management systems as described below. ~ The purpose of this supplement is to present (1) our review of the revision to the radioactive waste management systems, and (2) our evaluation of the radioactive waste management systems, as revised, to meet the requirements of Appendix I-to 10 CFR Part 50. The applicants have revised their radioactive waste nunagement systems, as described in Revision 14 to the SNUPPS Preliminary Safety Analysis Report. This revision added a secondary liquid waste system to process wastes in the regeneration of the demineral-izers in the condensate cleanup system. The condensate cleanup system is discussed in Section 5.6 of this supplement. The secondary liquid waste system will consist of tanks and pumps, an evaporator, demin-eralizer, charcoal bed, oil interceptor and filter. There will be no releases to the environment from this system since the effluent will be recycled back to the secondary system and evaporator bottoms will be solidified. The secondary liquid waste system will be housed in the radwaste building. Additional storage capacity will be provided in the solid radwaste system for holdup of the evaporator bottoms prior to transfer to the solidification holdup tank. The expected volume of wet solid waste will be increased by approximately 5,500 cubic feet per year, with a total activity increase of less than one curie. This increase will have a negligible effect on the solid radwaste system. The seismic and quality group designations of the equipment in the secondary liquid waste system, as delineated in Table 3.2-1 of the SNUPPS Preliminary Safety Evaluation . Report, are consistent with our guidelines as presented in staff Technical Position ESTB NO. 11-1, " Design Guidance for Radioactive Waste Management Systems Installed in Light-Water-Cooled Nuclear Power Reactor Plants," included with Standard Review Plan (NUREG-75/087, dated November 24,1975) Section 11.2, " Liquid Waste Management Systems". Therefore, we conclude that the proposed design of the secondary liquid waste system is acceptable. We have evaluated the radioactive waste management systems proposed for the Sterling plant, including-the above addition of a secondary liquid waste system, to reduce the quantities of radioactive materials expected to be released to the environment in 3 - 11-1

a . liquid and gaseous affluents in accardance with' Pction 50.34a of 10 CFR Part 50.

Thes2 systems,have_been previ;usly described in SIctions 11.2 and 11.3 of the Safety Evaluation Report and in Section 3.5 of 'the Draft Environmental Statement. The second-

.'ary 11guld waste system is described in' this supplement. ' Based on more recent infoma-tion applicable to the Sterling plant and changes in our calculational model, we have reviewed'the 11guld and gaseous source terms given in Tables 3.6 and 3.7 of the Draft Environmental Statement. These changes occurred subsequent to issuing the Draft Environmental Statement for the Sterling plant. The revised source tems were cal-culated using the model and methodology described in Draft Regulatory Guide 1.BB, " Calculation of' Releases of Radioactive Materials in Liquid and Gaseous Effluents from Pressurized Water Reactors (PWRs)," dated September 9,1975. On September 4,1975, the Comission amended,pendix 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 1. This option permits an applicant to design its radio- ~ . active waste management systems to satisfy the Guides on Design Objectives for Light- - Water-Cooled Nuclear Power Reactors 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 that it is unlikely that further reductions to radioactive material releases would be warranted on a cost-benefit basis for light-water-cooled nuclear power 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 Comission dated October 17, 1975, the applicants chose to comply with the September 4,1975 amendment to Appendix I rather than submit a cost-benefit - analysis as discussed in Paragraph 11.0 of Appendix I. Based on our reassessment of the liquid radioactive waste management systems we estimate that the quantity of radioactive materials expected to be released in liquid effluents, excluding tritium and dissolved noble gases, will be less than five Curies per year.and that the total calculated quantity of radioactive materials expected to be released in liquid effluents from the facility will not result in an annual dose or dose comitment to the total body or to any organ of an individual in an unrestricted area from all pathways of exposure in excess of five millirem. Based on our evalua-tion of the gaseous radioactive waste management systems, we estimate that the total quantity of radioactive materials expected to be' released in gaseous effluents from the plant will not result in a calculated annual gama air dose in excess of 10 milli-rad or a beta air dose in excess of 20 millirad at any location near ground level, at or.beyond the site boundary, which could be occupied by individuals. We estimate L 11-2

} 4 that the Ennual t tal quantity.of ' iodine-131' expected to be reltased Lin gassous - ~ ~ effluents will not exceed one curie per year and that the calculated annual total'

quantity of. radiciodine and radioactive particulates expected to be released in gaseous ef'1uents from the plant will not result in an annual dose or dose commitment to any organ of an individual. in an unrestricted area fran all pathways of exposure in excess -

of 15 millirem. Our detailed evaluation, which includes the calculated source terms and doses, is presented in Appendix C to this supplement. Based on our evaluation of the proposed liquid and gaseous radioactive waste management systems for the Sterling plant, we conclude that these systems are capable of meeting the criteria given in Appendix I of 10 CFR Part 50 for keeping releases of radioactive ~ materials to the environment "as low as is reasonably achievable," and therefore the - proposed systems are acceptable. 11-3 L_ .j

+ ^ 14.0 INITIAL TESTS AND OPERATIONS In the Safety Evaluation Report we concluded that an acceptable test and startup program can and will be implemented by the applicants. We have subsequently deter-mined there is one test we require that the applicants had not committed to performing. The' test, which is described in Section C 3b(2) of Regulatory Guide 1.79, "Preopera-

tional Testing of Emergency Core Cooling Systems for Pressurized Water Reactors",

Linvolves a test of the emergency core cooling system during the recirculation phase following a postulated loss-of-coolant accident. The test includes recirculation of water from the containment sump to verify vortex control and acceptable pressure drops across screening and suction lines and valves. We have infonned the applicants ~ of'our requirement for thi? test. As a result, the SNUPPS applicants have committed to provide the capability for performing this test, in conformance with Regulatory Guide 1.79, on the first SNUPPS unit. As the SNUPPS schedules presently stand, this ' capability will be provided in the first unit of the Callaway Plant, Units 1 and 2. - which _is the SNUPPS ' plant proposed by the Union Electric Company. The SNUPPS appli-cants. state that the sump design for all the SNUPPS plants are identical since the design is part of the standard portion of the plant. Therefore, the results of a sump test performed on the first SNUPPS unit will be applicable to all the SNUPPS plants. The. Sterling applicants state that a:lditional checkout tests will be por-formed on tne Sterling plant to assure that the sump system is operational. ~We concur with the SNUPPS applicants that the results of a sump test performed on the first SNUPPS unit will be applicable to all the PS plants. Therefore, on the basis of the SNUPPS applicants' commitment to pravide a sump test capability on the first SNUPPS unit in conformance with Regulatory Guide 1.79, we consider this matter resolved. 9 14-1

[1 ~ 15.0 ACCIDENT ANALYSIS -15.3 Accidents - We have' evaluated the possibility of the revised cooling water system as described in ' Section 2.4.3 of this' supplement for the proposed facility being disrupted by a ship

or barge accident. Based on a review of shipping patterns and traffic in the site vicinity and the design and submerged location of the seismic Category I intake struc-ture, we conclude that the likelihood of a ship or barge accident is sufficiently remote so that such an accident need not be considered as a design basis event for the Sterling facility.

15.5 Loss-of-Coolant Accident Dose Model In the Safety Evaluation Report we reported on the radiological consequences of a postulated loss-of-coolant accident based on the assumptions presented in Table 15.2 of the Safety Evaluation Report. We have reevaluated this postulated accident to reflect the following changes in the assumptions, which are included in the revised Table 15.2 presented in this supplement: 1. A change in the containment leakage rate (from 0.1 percent per day to 0.2 percent per day) to correctly reflect the SNUPPS design leakage rate, and 2. A change in the containment atmosphere mixing rate between sprayed and unsprayed volumes to give credit for the flow distributions established by the containment fan coolers and hydrogen mixing fans. The resultant radiological doses for this reevaluation are included in the revised Table 15.5 presented in this supplement. We find that the calculated low population i ' doses resulting'from containment purging following a postulated loss-of-coolant acci-dent, as discussed in Section.15.5 of the Safety Evaluation Report, when added to the revised loss-of-coolant accident doses, are well within the guidelines of 10 CFR ' Part 100. e 4 15 -

TABLE 15.2 ASSUMPTIONS USED FOR CALCULATION OF LOSS-0F-COOLANT ALTIlfENT 00SES Power Level (megawatts thennal): 3636 Operating Tirae (years): 3 Fraction of Core Inventory Available for Leakage (percent): Iodine!' 25 Noble Gases 100 Initial Iodine Composition in Containmp : (percent): Elemental 91 Organic 4 Particulate 5 Containment Leak Rate (percent per day): 0-24 hours 0.2 > 24 hours 0.1 Containment Volume: 6 Total Volume (cubic feet) 2.5 x 10 Unsprayed Fraction (percent) 11 Containment Mixing Rate between Sprayed and Unsprayed Volume: Forced Ventilation Rate (affecting 90 percent of unsprayed volume) (cubic feet per minute) 85,000 Natural Convection Rate (affecting 10 percent of unsprayedvolume)(volumectangesperhour) 2 Containment Spray System: Maximum Elemental Iodine 0econtamination Factor 100 Removal Coefficients (per hour): Elemental Iodine 10 . Particulate L,oine 0.45 Organic Iodine 0 Relative Concentration Values (seconds per cubic meter): 0-2 hours 1.1 x 10'4 0-8 hours 1.2 x 10-56 8-24 hours 8.2 x 10 6 1-4 days 3.5 x 10 6 4-30 days 1.0 x 10-15-2

TABLE 15.5 RADIOLOGICAL CONSEQUENCES OF ,w ! DENTS (PEM) ACCIDENT EXCLUSION AREA -LOW ?OFULATION ZONE ~(H00 meters) (4023 meters) Thyroid Whole Body Thyroid Whole Body Loss-of-Coolant 52 3 39 1 Fuel Handling 1.6 0.5 Hydrogen Purge Dose 4.5 0.6 i 1 1

17.0 QUALIVY ASSURANCE 17.1 General in Amendment 23,to the Application for Licenses, the Rochester Gas and Electric Corporation specified that it retains full responsibility for the construction, opera-tion and licensing of the proposed facility. Therefore. Rochester Gas and Electric remains responsible for the Quality Assurance program, which we have previously reviewed and and found acceptable, as stated in our Safety Evaluation Report. 17-1

.18.0 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS The Advisory Comittee on Reactor Safeguards completed its review of the application for a construction permit for the Sterling Power Project Nuclear Unit No I at its 186th meeting held on October 9-11, 1975. A copy of i e Comittee's report for the Sterling plant dated October 16, 1975, which contains.ertain comments and recomenda-tions, is attached as Appendix B. The actions we have taken or plan to take in re-sponse to these comments and recommendations are described in the following paragraphs: 1. The Comittee stated that it wishes to be kept informed on the resolution of p the following issues: a. The emergency core cooling system evaluation in compliance with the Final . Acceptance Criteria. b. The analyses of the effects of anticipated transients without scram. c. The evaluation of the plant design to meet the requirements of the new Appendix I of 10 CFR Part 50. Our evaluation of the emergency core cooling sy. em it, addressed in Section 6.3 of this supplement. This matter has been resa'ved in a manner satisfactory to -the staff. By means of this supplement the Comittee is being advised of the results of our evaluation. As stated in Section 7.2 of the Safety Evaluation "eport we are reviewing the analysis of the effects of anticipated transients without scram on a generic basis for Westinghouse plants. We will require that any design changes that are required as a result of our generic review, when it is completed, be implemented in the design of the Sterling plant. The results of our evaluation of the plant design to meet the requirements of the new Appendir I are addressed in Section 11.1 of this supplement and a more ~ detailed sumary of our evaluation is included as Appendix C. By means of this supplement, the Comittee is being advised of the results of our evaluation. 2. The Comittee recomended that the applicants and the Nuclear Regulatory Comis- - sion staff give particular attention to assuring proper coordination between the applicant and the Coast Guard to assure protection of the emergency service water -intake and discharge. 18-1

We will assure that the Final Safety Analysis R port includes the necessary written agreements between the applicants and the various Federal, State and local agencies whose services may be required to assure protection of the intake and discharge structures. 3. The Comittee stated they believe that the applicants and tb-Nuclear Regulatory Comission staff should continue to review the Sterling plant design for features that could reduce te possibility and consequences of sabotage. Tha Office of Nuclear Regulatory Research has funded studies concerning possible modes of sabotage at nuclear power plants. Any recommendations resulting from these studies, regarding additional design features to protect against acts of industrial sabotage, will be considered by the staff for incorporation in the Sterling plant design. 4. TI.6 Comittee recommended that the Nuclear Regulatory Commission staff and the applica9ts review the design features that are intended to prevent the occurrence of damaging fires and to minimize the consequences to safety-related equipment should a fire occur. The Comittee requested that it be kept infonned concerning this matter. The staff is considering a program to conduct a comprehensive review and evalua-tion 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. Any recomendations resulting from these studies will be considered by the staff to determine if any additional measures will be required to upgrade the fire protection systems for the Sterling plant. We will inform the Committee of the results of our review. 5. The Comittee stated that generic problems relating to large water reactors, should be dealt with appropriately by the applicants and the Nuclear Regulatory Comission 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 staff, various reactor vendors and other industrial organizations and will be the subject of continuing attention by the staff. 18-2

I 19.0 COMON DEFENSE AND SECURITY In the Safety Evalua' tion Report, we stated that all of the directors of the Rochester Gas and Electric Corporation were citizens of the United States and we concluded that - Rochester Gas and Electric was not owned, dominated or controlled by an alien, foreign corporation or foreign government. However, in Amendment 23 to the Application for Licenses, it was stated that one director of the Rochester Gas and Electric Corporation is a citizen of Great Britain. The applicants have since provided us (in a letter from R. Koprowski to B. Rusche, dated February 12,~ 1976) with the information that the British citizen, one of 15 members of the Rochester Gas and Electric Corporation Board of Directors, owns 991 shares of common stock of the 9,500,000 shares issued as of December 1974. . On the basis that there is only one foreign director of the 15 directors on the Rochester Gas and Electric Corporation Board of Directors and that that director owns less than one percent of the common stock issued by the corporation, we conclude that, as required by 10 CFR 50, the applicants are not owned, dominated or controlled by an alien, foreign corporation or foreign government. 19-1

20.0 FINANCIAL QUALIFICATIONS 20.1 Introduction. Section 50.33(f) and Appendix C of 10 CFR 50 are the Commission's regulations which relate to financial data and the information required to establish the financial qualifications of an applicant for a facility construction permit. Rochester Gas and Electric Corporation. Orange and Rockland Utilities Incorporated, Central Hudson Gas and Electric ' Corporation and Niagara Mohawk Power Corporation have applied for a construction permit for the Sterling Power Project Nuclear Unit No. 1. The applicants will share in the ownership, financial support and electrical output of the proposed facility as follows: Rochester Gas and Electric Corporation 28% Orange and Rockland Utilities, Incorporated 33% Central Hudson Gas and Electric Corporation 17% Niagara Mohawak Power Corporation 2_21 2 100% The most recent estimate of the total cost of the Sterling Power Project Nuclear Unit No.1, as provided by the applicants on October 30, 1975 in response to a staff request for additional financial information, may be summarized as follows: (dollars in millions) Nuclear production plant cost---------------------- $1083.7 Transmission and distribution costs---------------- 46.3 Nuclear fuel inventory cost for first core--------------------------------------------- 82.0 Total $1212.0 The estimated cost of the nuclear production plant has been compared with the cost estimated by the Energy Research and Development Administration's CONCEPT costing model. ~The CONCEPT costing model projected the cost of the nuclear productica plant I to be $1062.0 million. The applicants' estimate of $1083.7 million is 2.0 perant ) above CONCEPT. The Oak Ridge National Laboratory, which administers the CONCEPT ) costing model, states that " estimates produced by the CONCEPT costing ndel are not intended as substitutes for detailed engineering cost estimates, but were prepared - as a rough check on the applicant's estimate." Consequently, in the interest of ~ conservatism, we have used in our analysis the applicants' higher estimate as the . cost of the nuclear production plant. 20-1

The applicants till shar; th) cstimated costs as follows: (dollarsinmillions) Rochester Gas and Electric Corporation $ 339.4 Orange & Rockland Utilities. Incorporated 400.0 Central Hudson Gas and Electric Corporation 206.0 Niagara Mohawk Power Corporation 266.6 $1212.0 We have reviewed the financial information presented in the application, and amend-ments thereto, and concluded that there is reasonable assurance that the afore-mentioned applicants can obtain the necessary funds to design and construct the Sterling Power Project Nuclear Unit No.1. Accordingly, we find them financially qualified to carry out the activities for which this permit is sought. Our con-clusion is based upon the following analyses and the basic assumptions of tational regulatory policies and relatively stable capital market conditions. These as-sumptions are necessary because of the lengthy future period involved and the expected heavy dependence on external financing. 20.2 Rochester Gas and Electric Corporation Rochester Gas and Electric supplies electricity, gas and steam service in the north central part of New York State. In addition to the city of Rochester, which is the third largest city and a major industrial center in the state, the utility supplies a large and prosperous fanning area. Operating revenues increased from $215.0 million for the 12 months ended June 30,1974 to $262.1 million for the 12 months ended June 30, 1975 and net income increased from $17.9 million to $27.6 million. Invested capital on March 31. 1975 amounted to $546.5 million and consisted of 48.9 percent long-term debt 12.3 percent preferred stock and 38.8 peretnt common equity. The return on common equity for the 12 months ended June 30, 1975 was 11.6 percent, up from 8.44 percent for the 12 months ended December 31, 1974. Pretax coverages of long-term interest and total interest charges for the 12 months ended June 30, 1975 were 3.31 times and 2.88 times, respectively, versus 2.54 ti.nes and 2.05 times for the 12 months ended December 31,1974 Rochester Gas and Electric's first mortgage bonds are rated A by both Moody's and Standard and Poor's. Rochester Gas and Electric plans to finance its 28 percent share of the project costs by the use of internally generated funds, and the issuance of deist and equity securities. Available funds from these sources in 1974 totaled $75.1 million and were derived from $23.6 million of internally generated funds, $30.0 million of first mortgage bonds, $0.5 million of common stock, and a $21.0 million increase in notes payable. Internally generated cash in 1974 represented 42.0 percent of the 1974 construction budget expenditures (including nuclear fuel). 20-2

At our requ;st, Rochester Gas and E1;ctric supplied a projrct:d sources of funds statement for tha 1976 to 1984 period, with undirlying assumptions, demonstrating how the requisite funds might be raised. Its internally generated cash over this period is projected to be 49.7 percent of total construction expenditures and 213.5 percent of its expected outlays for the Sterling Power Project Nuclear Unit No.1. We have reviewed Rochester Gas and Electric projections, and underlying assumptions, and find them reasonable. A " Sources of Funds" statement for the Rochestec Gas and Electric Corporation and the assumptions applicable to the statement are presented as Tables 20.1 and 20.2 respectively. 20.3 Orange and Rockland Utilities, Incorporated Orange and Rockland's operating revenues increased from $154.4 million for the 12 months ended June 30,1974 to $202.6 million for the 12 months ended June 30, 1975 and net income increased from $15.6 million to $16.2 million. Invested capital at June 30,1975 amounted to $423.3 million and consisted of 54.1 percent long-term debt,12.7 percent preferred and preference stock and 33.2 percent coninon equity. The return on common equity for the 12 months ended June 30, 1975 was 9.4 percent, up from 8.5 percent for the 12 months ended December 31, 1974. Pretax coverage of total interest charges for the 12 months ended June 30,1975 was 2.11 times versus 1.83 times for the 12 months ended December 31, 1974. Orange and Rockland's first mortgage bonds are rated BAA by Moody's and BBB by Standard and Poor's. Orange And Rockland plans to finance its 33 percent share of the project costs by the use of internally generated funds, and the issuance of debt and equity securities. Available funds from these sources in 1974.atal $58.9 million and were derived from $10.4 million of internally generated funds, $38.0 million in first mortgage bonds, and $10.5 million in other funds, hternally generated cash in 1974 represented 25.5 percent of 1974 construction expenditures. At our request, Orange and Rockland supplied a projected sources of funds statement for the 1976 to 1984 period, with underlying assumptions, demonstrating how the requisite funds might be raised. Its internally generated cash over this period is projected to be 32.7 percent of total construccion expenditures and 87.0 percent of -its expected outlays for the Sterling Power Project Nuclear Unit No.1. We have reviewed Orange and Rockland's projections, and underlying assumptions, and find them reasonable. A " Sources of Funds" statement for Orange and Rockland Utilities. Incorporated and the assumptions applicable to the statement are presented as Tables 20.3 and 20.4, respectively. 20.4 Central Hudson Gas and Electric Corporation Central Hudson Gas and Electric supplies electricity and gas to a 2,600 square mile area in southeastern New York State. Operating revenues increased from $104.8 million for the 12 months ended June 30, 1974 to $153.2 million for the 12 months ended June 30, 1975 and net income increased from $11.3 million to $13.8 million. Invested capital at June 30, 1975 amounted to $309.3 million and consisted of 50.5 percent long-tem debt,14.9 percent preferred stock and 34.6 percent coninon equity. 20-3

1+ 1 l Table 20.1 .u SOURCES OF FUNDS FOR SYSTEM-WIDE CONSTRUCTION EXPENDITURES DURING PERIOD ' F CONSTRUCTION OF STERLING UNIT NO. 1 (MILLIONS OF D0LLARS) Applicant: ROCHESTER GAS and ELECTRIC CORPORATION Construction Years of Sub.iect Nuclear Power Plant Security' issues and other funds 1976 1977 1978 1979 1980 1981 1982' .1983 1984- . Conunon Stock. 0 $ 0 0 0- $ 0 ~ $763 $T33 0 $0 Preferred stock 0 15.0 15.0 .15.0 - 15.0 15.0' 15.0-15.0-0 Long-term debt-50.0 19.0 35.0 38.3 33.0 170.0 49.0 40.0" 20.0 Notes payable (3.6) 13.2 .(2.0)

4. 5 -

(1.7) '21.3 (0.9)- (18.0), '(28.9); Contributions from . parent-net 0 0 0 0 0 0 0 ~0 0-Other funds 0 1.5 0 0.4 12.0 2.4 0.2 2.8 2.9 -

Total.-

$T6% $ TEX $3CE $38-I $1 578 T $T93 $T3.6) ; -Internal funds ~ $81.7 $88.8 .. Net income. $39.3 $42.8 $47.4 $52.3 $58.3 $ 65.0 $73.8 m - Less: -97 Pref;rred dividents, (6.1) (6.8) ( 8.3) (9.8) (11.3) (12.8) (14.3) (15.8) '(16.6) Common dividends (13.8) (14.5) (15.2) (16.0) (16.8) (18.0)- (20.1) (21.9) (23.0)- Retained earnings 19.4 21.5 23.9 26.5 30.2 34.2 39.4-44.0 49.2 iDeferred taxes. 1.7 - 1.6 1.9 2.7 2.9 3.2 4.6 6.0 6.5 . Invest.: tax credit-deferred 0 0 0 0 0 0 0 0-0 Depreciation and_ amort. 23.7 27.2 28.8 31.7 33.4 34.3 42.2' 50.7 . 53.2 ta 3. TOTAL FUNDS $EC6 $E33 $353 $15CT $1577 $1'6V l 5137 3 ' $1T33 $3V3 Construction Expenditures

• Nuclear power plants

$21.7 $22.0 $28.6 $ 46.1 $ 63.8 $ 85.5 $ 47.5 $ 20.4 $ 4.6 Other 62.3 63.5 59.9 58.0 43.6 83.7 90.1 92.6 85.3 Total const, expenditures $84.0 TB53 588 5 1;104.1 lit 673 IT6TE 3T37% TTT3-6 13V3 Subject nuclear plant $ 2.2 TT7 T7X 1r1CI li 47 7 TT43 TTE'T TTF7 lITJ SEC coverage-3.19 2.97 2.95 2.74 2.76 2.64 2.60 2.68 3.12 Indenture coverage-3.59 3.27 3.14 2.92 2.83 2.51 2.63 2.64 3.31

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

I' Table 20.2 ASSUMPTIONS APPLICABLE TO SOURCES OF FUNDS STATEMENT ROCHESTER GA5 AND ELECTRIC CORPORATION 1. Rate of return on average cor i. equity - maintained between 13.5 and 13.7 percent. 2. Preferred stock dividend rate - IG.; percent 3. Short-term interest rate 8 percent 4. Bond interest rate - =.a percent 5. Market price of common stock - $17.00 (1976) increasing $1.00 per year to $21.00 (1930) remaining constant through 1984 6. Comon stock divident was increased 2 percent per annum from current level, provided the total annual cash dividend did not exceed a 50 percent payout ratio. 7. A 3 percent stock dividend was also assumed in all years. 8. Growth Rates: Retail sales (KWH) - 1976 (7.1 percent.1977 7.3 percent. 1978 6.8 percent). 1979 (6.6 percent. 1980 6.6 percent. 1981 6.4 percent). 1982(6.2 percent. 1983 6.0 percent. 1984 5.7 percent). Revenues - Sufficient to maintain approxitstely 13.5 percent return on average . equity 9. Target Capital Structure: Bonds 44-47 percent Preferred Stock 14-16 percent Short-Tenn Debt 0-3 percent Comon Equity 36-40 percent

10. Operation and maintenance expenses were projected either manually or by the best fit of historical data to a least squares curve. Interest charges were actually calculated based on known and assumed financings and assumed interest rates. Net income was a result of all the above assumptions ad projections.

20-5

o f Table 20.3 SOURCES OF FUNDS FOR SYSTEM-WIDE CONSTRUCTION EXPENDITURES DURING PERIOD OF CONSTRUCTION OF STERLING UNIT NO. 1 (MILLIONS OF D0LLARS) Applicant: ORANGE AND ROCKLAND UTILITIES, INCORPORATED AND SUBSIDIARIES Construction Years of Subject Nuclear Power Plant ' Security issues and other funds 1975 1976 1977 1978 1979 1980 1981 1982 1983 1984 Common stock 0 0 $24.9 0 $18.1 $15.6 $27.0 $22.0 $ 0 0 Preferred stock 0 0 0 0 5.0 6.2 11.7 10.5 5.0 0 Long-tenn debt 10.0 0 27.1 0 44.0 36.9 58.8 50.0 38.7 0 Notes Payable (3.1) 5.2 (32.3) 19.5 (7.1) 3.1 8.9 (22.2) 5.7 10.6 Crntributions from parent-net 0 0 0 0 0 0 0 0 0 0 Other funds (3.1) .5 (2.5) (1.4) 2.9 13.4 11.8 25.8 27.8 12.5 Total 3.8 5.7 17.2 18.1 62.9 75.2 118.2 86.' 77.2 23.1 Internal funds Retained earnings Net income 20.7 21.5 26.6 30.8 35.5 40.3 49.0 56.9 62.6 66.8 ? Less: preferred dividends (3.0) (3.0) (3.0) (3.0) (3.1) (3.7) (4.4) (5.6) (6.5) (6.9) y a comon dividends (11.5) (12.0) (15.4) (18.2) (21.1) 23.8 (29.0) (33.4) (?'.1. 55 39.0) 20.9 Ritained earnings 6.2 6.b 8.2 9.6 11.3 15 6 17.9 Deferred taxes and Investment tax credit 5.6 6.6 3.4 3.7 4.0 4.4 4.9 5.3 5.6 7.5 Depreciation 12.5 13.0 13.3 14.6 15.5 16.6 17.6 18.8 20.0 31.4 Less: AFDC (1.1) (1.5) (2.0) (2.6) (5.9) (12.9) (18.2) (26.9) (33.7) (11.3) Total 23.2 24.6 22.9 25.3 24.9 20.9 19.9 15.1 11.5 48.5 l TOTAL FUNDS $77~6 $30.3 $40.1 $43 4 $87 8 $96 1 $138.1 $ TOT f $88 7 $7T.T Construction expenditures

• Nuclear power plants

$ 3.5 $ 2.6 $ 2.0 $ 9.0 $33.2 $56.2 $ 87.8 $49.8 $23.3 $ 4.7 Other 23.5 27.7 38.1 34.4 54.6 39.9 50.3 51.4 65.4 66.9 Total $27.0 $30.3 $40.1 $43.4 $87.8 $96.1 $138.1 $101.2 $88.7 $71.6 Subject nuclear plant $ 3.5 $ 2.6 $ 2.0 $ 9.0 $33.2 $56.2 $ 87.8 $49.8 $23.3 $ 4.7

• Excluding AFDC (allowance for funds used during construction)

2 Table 20.4 t'~ ASSUMPTIONS APPLICA8LE TO SOURCES OF FUNDS STATEMENT ORANGE AND ROCKLAND UTILITIES, INCORPORATED 1.' Return'on average common e'uity: q 1975 and 1976-12 percent 1977 and 1978 14 percent 1979 and 1980 15 percent '1981.to 1984. 16 percent 2. Assumed the following interest rates' for new securities: a Long-term Debt 10 percent b Pollution Control Debt 8 percent c Preferred Stock 10 percent 'd -Short-Term Debt-10 percent 3. Assumed that any new common stock sold during the period would provide net proceeds to Orange and Rockland equal to approximately 75 percent of the book value at the end of the prior year. 4. Assumed an 81/2 percent rate in 1975 and 10 percent thereafter for computing the allowance for funds used during construction for both Orange and Rockland's construction . program and the Sterling plant. 5. Assumed a 65 percent payout or a minimum of $1.20 per share. 6. Federal Income Tax: a) ' Assumed that the federal corporate income tax rate would remain at 48 percent. b) Assumed that the investment tax credit would be 10 percent in 1975 and 1976 i and 4 percent thereafter. 7. Orange and Rockland's nornal construction budget has been included in these forecasts.

8. - Assumed the following costs for the Sterling Plant (000's) (total figures supplied by RochesterGasandElectric):

4 Orange and Rockland-Utilities, Incorporated Total 33 percent interest Cash Expenditures'for the plant total- $ 832,600 $ 274,824

AFDC, 297 200 98,076-Total 1.TM 372,900 Nuclear Fuel - Cash

. Expenditures-82,000 27,060 Total Expenditures 1.212,000 $ 399,960 Cost of facilities eligible for potiution control financing $ 150,000 $ 50,000 Investment tax credit-30,231 9,976' 9.'. Electric sales.' revenues and expenses were based on a-7.1 percent load growth. . 10. Gas sales, revenues and expenses wre assumed to have virtually no growth. y. 20-7'

Table 20.4 (Continued)

11. Operation and maintenance expenses were escalated at 8 percent per year through 1980, and at 6 percent per year thereaf ter, l

12. A desired capital structure of 50 percent debt,10 percent preferred and preference stock

- and 40 percent common stock equity was used.

13. Securities and Exchange Commission coverage ratios were as follows:

1975 1976 1977 1978 1979 1980 1981 1982 1083 1984 Security and 2.48 2.60 3.17 3.38 3.42 3.10 3.29 2.77 2.30 2.79 Exchange Commission ratio

14. Indenture coverage ratios are not available as the study was conducted on a consolidated basis and indenture coverage requirements are on an unconsolidated basis.

) 20-8

The return on corrion equity for the 12 months ended June 30,1975 was 10.4 percent, up from 9.2 percent for the 12 months ended December 31,1974. Pretax 7v; rages of long-term interest and total interest charges for the 12 months ended oune 30, 1975 were 2.92 times and 2.13 times, respectively, versus 2.43 times and 1.78 times for the 12 months ended December 31,1974 Central Hudson Gas and Electric's first mortgage bonds are rated A by both Moody's and Standard and Poor's. Central Hudson Gas and Electric plans to finance its 17 percent share of the project costs by the use of internally generated funds, and the issuance of debt and equity securities. Available funds from these sources in 1974 totaled $38.5 million and were derived from $14.3 million of internally generated funds, $15.0 million of first mortgage bonds, $8.2 million of common stock, and a $1.0 million increase in short-term debt. Internally generated cash in 1974 represented 56.5 percent of the 1974 construction budget expenditures. At our request Central Hudson Gas and Electric supplied a projected sources of funds statemes t for the 1976 to 1984 period, with underlying assumptions, demonstrating how the requisite funds might be raised. Its internally generated cash over this period is projected to be 26.3 percent of total construction expenditures and 116.4 percent of its expected outlays for the Sterling Power Project Nuclear Unit No.1. We have reviewed Central Husdon Gas and Electric's projections, and underlying assumptions, and find them reasonable. A " Sources of Funds" statement for the Central Hudson Gas and Electric Corporation and the assumptions applicable to the statement are presented as Tables 20.5 and 20.6, respectively. 20.5 Niagara Mohawk Power Corporation Niagara Mohawk Power Corporation is an investor-owned electric and gas utility supplying electricity to western New York State and southern Ontario. Operating revenues increased from $720.3 million for the 12 months ended April 30, 1974 to $880.0 million for the 12 months ended April 30, 1975, and net income increased from $75.5 million to $91.1 million. Invested capital at March 31, 1975 amounted to $2.3 billion and consisted of 53.8 percent long-term debt,13.0 percent preferred stock and 33.2 percent common equity. The return on common equity for 1974 was 11.64 percent, up from 8.16 percent in 1973. Pretax coverages of long-term interest and total interest changes in 1974 were 2.57 times and 2.18 times, respectively, versus 2.21 times and 2.00 times in 1973. Niagara Mohawk's first mortgage bonds are rated A by Moody's and BBB+ by Standard and Poor's. Niagara Mohawk plans to finance its 22 percent share of the project costs by the use of internally generated funds, and the issuance of debt and equity securities. Available funds from these sources in 1974 totaled $371.8 million and were derived from $74.7 million of internally generated funds, $125.0 million of first mortgage bonds, a $129.3 million increase in notes payable, $29.9 million of common stock, and $12.9 million of other funds. Internally generated cash in 1974 represented 26.3 percent of 1974 construction expenditures. 20-9 1

Table 20.5 S_003CES OF FUNDS FOR SYSTEM-WIDE CONSTRUCTION EXPENDITURES DURING PERIOD OF CONSTRUCTION OF STERLING UNIT NO. 1 (MILLIONS OF DOLLARS) Applicant: CENTRAL HUDSON GAS and ELECTRIC CORPORATION Construction Years of Subject Nuclear Power Plant Security Issues and Other funds 1976 1977 1978 1979 1980 1981 1982 1983 1984 12.0 $ 10.0 $ 14.0 $ 16.0 $22.0 $22.0 0 Common stock 0 $ 8.0 $' 14 0 0 0 16 0 22.0 0 24.0 Preferred stock 12 0 0 Long-tern debt 0 20.0 30 0 30.0 35.0 40.0 W 40 0 60.0 Notes payable 0 0 0 0 0 0 0 0 0 Refunding debt securities (8.0 0 (12.0) (8.0) (6.0) 0

11. S 31.9)

Other funds (short term debt) (1.5) ~J4 at maturity 0 (7.9) 8.3 8.1 4.1 23.4 Total 10.5 79.9 32.1 45.3 W 107.1 38 6 ~ Internal funds Net income 0 0 0 0 0 0 0 0 0 Less: preferred dividends 0 0 0 0 0 0 0 0 0 m? corrrnon dividends 0 0 0 6 0 0 0 0 0 5 Retained earnings 5.3 5.7 6.4 7.3 - T2~ 9.2 10.5 12.1 13.1 Deferred taxes 12 1.4 1.7 2.0 22 2.3 3.0 3.1 3.5 Investment. tax credit deferred 0 0 0 0 0 0 0 0 0 Depreciation and amort. 12.7 13.1 13.5 16.2 16.8 17.5 19 0 26 8 34 0 Less: AFDC (1.8) (3J2 (4 8) (6 8) (9.8) 20 8 (26 0) N Total 17 4 -~T7 0 16 8 18 7 1/ 4 15 0 32 8 TOTAL FUNDS 27 9 $ 31 3 $- E F-1 50 8 $'TET- $T6I $WRI $53 6 $ 83 9 Construction Expenditures

• Nuclear power plants 10.7

$ 13.4 $ 19.4 $ 30.9 $ 37.0 $ 56.2 $ 33.6 $13.2 $ 3.9 Other 17.2 $ 17.9 $ 77.3 $ 19 9 $W $ 30.2 $ 85 2 $40 4 $ B0 0 Total const. expen. 27.9 $ 31.? $ %.) $ 50.8 $TE7 $ 66 4 $118 8 $53 6 $ 83.9 Subject nuclear plant 1.3 $ l.0 4.7 $-TTT $ 28.9 3452 $'e5 7 $12.0 $ 2.4 l

• Erclusive of AFDC (allowance for funds used daring construction)

Tabla 20.6 ASSUMPTIONS APPLICABi.E TO SOURCES OF FUNDS STATEMENT CENTRAL' HUDSON GAS AND ELECTRIC CORPORATION 1. For purpos'es of de'veloping these figures, a capital structure of 50 percent debt,15 percent preferred stock and 35 percent common stock was'atsumed.' 2. A return on common equity of 13 percent (the amount allowed in the last rate case) and a l payout ratio of 65 percent were likewise assumed. It should be noted that the 13 percent return on commor, equity would be increasingly inadequate during this period and higher j return rates would have to be allowed in order to pennit the' sale cf bonds in the 10 percent to 11 percent ranges. 3. Total Interest Coverages are on the order of 3.50 to 2.20 times declining over the period at 13 percent return on equity. 4. Mortgage Interest Coverage is on the order of 4.25 to 2.75 times. 5. Short-term interest is assumed at 11 percent over the period. Central Hudson Gas and Elec'. ic has not prepared a detailed forecast income statement but has based its financing schedule on the maintenance of an assumed capital structure, return on equity, payout ratio, and assumed rates of interest and dividends on bonds and preferred stock and on short-term debt. 20-11

Y w At our requ:sti Niagara Mohawk supplied a projected sturcss of funds statement for the 1976 to 1984 period, with underlying assumptions, demonstrating how the requisite funds might be raised.. Its internally generated cash over this period is projected to be 33.7 percent of total construction expenditures and.616.9 percent of its expected outlays for the Sterling Power Project Nuclear Unit No.1.. We have reviewed Niagara Mohawk's projections, and underlying assumptions, and find them reasonable. A " Sources of Funds". statement for the Niagara Mohawk Power Corporation [ and the assumptions applicable to the statement are presented in Tables 20.7 and 20.8, respectively. l 6 _ 12 z

{.

i .L.,.-,, e ~

Table 20.7 SOURCES OF FUNDS FOR SYSTEM-WIDE CONSTRUCTION EXPENDITURES DURING PERIOO OF CONSTRUCTION OF STERLING UNIT NO. 1 (MILLIONS OF DOLLARS) Applicant: NIAGARA MdHAWK POWER CORPORATION Construction Years of Subject Nuclear Power Plant Security' issues 1976 1977-1978 '.479 1980 1981 1982 1983 1984 and other funds Common stock $23.3 $41.5 $44.1 $53.4 $15.7 $47.2- $110.8 $154.7 $ 98.4 Preferred stock 10.0 17.8 18.9 22.9 26.3 41.1 70.7 92.9 72.2 Long-term debt 13.4 59.2 62.9 76.2 87.5 137.2 235.7 309.7-240.5 Notes payable 0 0 0 0 0 0 0 0 0 Contributions from parent-net 0 0 0 0 0 0 0 0 0 Other funds 0 0 0 0 0 0 0 0 0 Total $66.7 $118.5 $125.9 $152.5 $129.5 $225.5 $417.2 $557.3 $411.1 Internal Funds g Net income $124.0 $138.0 $155.0 $174.0 $190.3 0 0 0 0 L Less: w Preferred dividends 24.0 27.0 31.0 35.0 38.4 0 0 0 0 Consnon dividends 70.0 78.0 87.0 97.0 106.3 113.9 126.4 144.8 163.4 Retained earnings 30.0 33.0 37.0 42.0 45.6 48.8 54.2 62.1 70.0 Deferred taxes 9.0 13.0 12.3 10.1 13.4 14.1 14.9 15.8 16.9 Invest, tax credit-deferred 0 0 0 0 0 0 0 0 0 Depreciation and amortization 79.3 84.0 87.0 93.0 96.0 100.0 105.0 109.0 113.0 Loss: AFDC 20.3 27.7 41.7 40.9 23.0 31.0 45.0 32.0 26.0 Total 98.0 102.3 94.6 104.2 132.0 131.9 129.1 154 9 T7JJ TOTAL FUNDS 5T54 7 $T25T $Fl53 $2WT $25TT $357J $5 W T $7T2'T $1,Q Construction Expenditures

• Nuclear power plants

$ 44.6 $ 52.6 $ 67.7 $ 65.1 $ 72.6 $ 80.1 $ 39.6 $ 15.5 $ 3.1 Other 120.1 168.2 152.8 191.6 188.9 277.3 506.7 696.7 581.9 Total construction expend. $T6TT $Fl5T $Fl53 $25E T $2ETT $357J $5'43'T $71 T T $51f5T Subject nuclear plants $ 4.5 $ 1.3 57 3 $ FlT $ 37T $T8T $ 3TT $T5T $ 3.1 'SEC coverage 2.01 2.49 2.52 2.53 2.52 2.50 2.49 2.45 2.42 Mortgage indenture 2.14 2.61 2.57 2.59 2.56 2.55 2.45 2.35 2.37

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

~ e Table 20.8 ASSUMPTIONS APPLICABLE TO SOURCES OF FUNDS STATEMENT l NIAGARA MOHAWK POWEP, CORPORAIION ,1. Return on average common equity; l. 1976 12.0 percent OL 1977 12.5 percent I' 1978 13.0 percent l I l - 1979 13.5 percent : 1980 to 1984 13.75 percent 2. Assumed cost of new securities: i i . a. ' Long-Term Debt -10.2 percent ' b. Short-Tem Debt - 8.0 percent c. Preferred Stock 10.5 percent . 3. Market Price of Common Stock: $12.40 (1976), increasing $1.00 per year to $16.40 ti980), and remaining constant through 1984. 4. Assumed Common Stock Divided Payout Ratio = 70 percent 5. Growth Rates: Annual Energy Sales (KWH) - 1976 3.9 percent 1977 3.8 percent. 1978 (3.8 percent) 1979 3.7 percent, 1980 3.7 percent, 1981 (3.6 percent) 1982 3.6 percent, 1983 3.5 percent, 1984 (3.5 percent) Revenues.- sufficient to maintain returns on average equity indicated above. f l 6. Target Capital Structure: Long-Term Debt 50 percent Preferred Stock 15 percent Common Equity -35 percent. 7. Operation and m'aintenanca expenses were projected maiually based on historical data and i-anticipated future developments. Interest, preferret dividend, and balance for common ' equity requiremente were based on the rates assumed a ove and the projected financing through 1984. (- l' l I i i 1 - 14-eMa = eg e

21.0 CONCLUSION

S Our conclusion that the issuance of a permit for construction of the facility will not be inimical to the common defense and security or to the health and safety of the public, as stated in the Safety Evaluation Report, Section 21.0 was conditioned on the favorable resolution of outstanding matters identified in Section 1.8 of the Safety Evaluation Report. We have discussed each of these outstanding issues in this supplement and indicated a favorable resolution of each matter. Therefore, we can reaffim our conclusions as set forth in Section 21.0 of the Safety Evaluation Report. 21-1

F APPENDIX A CONTINUATION OF CHRONOLOGY OF RADIOLOGICAL REVIEW OF . STERLING POWER PROJECT NUCLEAR UNIT NO. 1 September 5,~1975 Issuance of Safety Evaluation Report September 5,1975 Letter from SNUPPS transmitting Revision 10 to SNUPPS PSAR, incor-porating infomation previously submitted, infomation concerning fire stops and seals. ECCS, site data, and design changes September 8, 1975 Letter from applicant incorporating Revision 10 of SNUPPS PSAR' as Amendment No. 22 September 8, 1975 Letter from SNUPPS concerning changes in seismic design of the radwaste equipment and building September 9,1975 Letter from applicant incorporating SNUPPS letter dated September 8, 1975 September 12, 1975 Letter from SNUPPS transmitting infomation concerning main steam line break analysis, in response to request of August 22, 1975 September 15, 1975 Letter from applicarit incorporating SNUPPS letter of September 12, 1975 September 15, 1975 Submittal of Amendment No. 23, consisting of information reflecting .the addition Gf three co-participants -September 16, 1975 . Letter to applicants requesting updated financial information September 19, 1975 Meeting with applicants to discuss possible change in design of service water intake Stptember 23, 1975 Letter to appitcants concerning containment design leakage rate SIptember 24,1975 Letter from SNUPPS transmitting Revision 11 to SNUPPS PSAR, incorporating infomation previously sutnitted and revisions concerning changes in the seismic design of the radwaste building and equipment S:ptember 30, 1975, Letter from applicants incorporating Revision 11 to SNUPPS PSAR as Amendment No. 24 October 2, 1975 Meeting with SNUPPS Utilities to discuss main steam line break analysis October 6, 1975 Letter from SNUPPS transmitting information concerning containment design leakage rate, in response to request of September 23, 1975 October 6, 1975 . Meeting with SNUPPS Utilities to discuss outstanding issues October 9, 1975. . ACRS meeting with staff and applicant October 10, 1975 Letter from SNUPPS transmitting additional infomation concerning containment design leakage rate, in response to request of-September 23, 1975 October 13, 1975-Letter from SNUPPS transmitting Revision 12 to SNUPPS PSAR, con-sisting of clarification and minor changes to the SNUPPS PSAR . October 15, 1975 Letter from applicants incorporating Revision 12 to Sf 7 PS PSAR as Amendment No. 25 October 16. 1975 Report by the ACRS A-1

- ~. Octobe'r 17, 1975-Submittalof'AmendmentNo.26(R;visitn12_tsSiteAdd:ndum), .' consisting cf inftmation relativ2 to the sssIntial sirvice water system and Appendix I to 10 CFR Part 50 October 21,1975 , Letter to applicants requesting response to staff positions October 28,.1975 Letter from SNUPPS providing partial response to request of October 21 - 1975, concerning ECCS' ' October 31,1975 Letter from SNUPPS transmitting information concerning main steam line break analysis-November 3,~1975 Submittal of Amendment No.'27, consisting of updated financial infomation, and Amendment No. 28, consisting of revised general information November 7, 1975 Letter to applicants requesting response to staff position regarding WCAP-7705 November 10,'1975 Letter from' SNUPPS providing information in response to request of -November 7, 1975 November 11,1975 Letter from SNUPPS providing additional information in response to request of October 21, 1975, concerning ECCS , November 17,11975 Submittal of Amendment No. 29 (Revision 13 to Site Addendum), consisting of information relative to seismic design r . November 18. 1975 Meeting with SNUPPS Utilities to discuss outstanding issues ' November '21,1975 Letter from SNUPPS providing information regarding verbal comitments made at November 18, 1975 meeting November 26,19751 Letter from SNUPPS transmitting Revision 13 to SNUPPS PSAR, incorporating infomation previously submitted by letters and providing information concerning research and development plans ' December 4, 1975 Letter from applicants incorporating Revision 13 to SNUPPS PSAR as Amendment No. 30 I December 17,'1975 SubmittalofAmendmentNo.31(Revision 14toSiteAddendum), consisting of revised site infonnation.

January 12'1976 SubmittalofAmendmentNo.32(Revision 15toSiteAddendvi),

consisting of revised design and quality assurance infomation January 14,1976 Letter from SNUPPS transmitting Revision 14 to SNUPPS PSAR, consisting of updated design infomation January 16,1976 Letter from applicants incorporating Revision 14 to SNUPPS PSAR as Amendment No. 33 February 20,'1976 Submittal of..nendment No. 34 (Revision 16 to Site Addendum), ~ consisting of information relative to Appendix I to 10 CFR Part 50 March _17,1976 Letter to applicants' advising that application should be amended to include co-participants as applicants LApril 9,.1976. Submittal of Amendment No. 36*,'which revised the application to include Orange and Rockland Utilities. Inc., Central Hudson Gas and Electric Corporation, and Niagara Mohawk Power Corporation as- 'co-applicants a . *Amendnent No. 35 reserved for future submittal ~A-2' f 'T -h M'7 y

APPENDIX B ADVISORY COMMITTEE ON REACTOR SAFEGUARDS NU' CLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 October 16, 1975 2 Honorable William A. Anders 01 airman U. S. Nuclear Regulatory Comission Washington, D.C. 20555

Subject:

REPORT ON TIE STERLING PGER PROJECT NCELEAR INIT 1

Dear Mr. Anders:

During its 186th meeting, October 9-11, 1975, the Advisory Comittee on Reactor Safeguards reviewed the application of Rochester Gas and Electric Corporation for a permit to construct the Sterling Power Project, Unit No. 1. On Septanber 24, 1975, the site was visited and a Subcomittee meeting was held in Sterlire, New York to review site-related matters. %e " Standardized Nuclear Unit Power Plant System" (SNUPPS) to be utilized at the Sterling site, and at three other plant sites, was reviewed at Subcomittee meetings held at Washington, D. C. on August 19, 1975, and at Emporia, Kansas on September 26, 1975, and at the IS5th and 186th meetings of the Comittee. During its reviews, the Comittee had the benefit of dis-cussiens with the Nuclear Regulatory Comission (NRC) Staff and repre-sentatives of the applicant, the Westinghouse Electric Corporation and the Bechtel Corporation. We Comittee also had the benefit of the documents listed below. We Sterling unit will be located on a 2800-acre site of partially wooded rural land located on the southeastern shore of Iake Ontario, approximately 7 miles southwest of Oswego, New brk the nearest popu-lation center (1970 population: 23,844). W e minimum exclusion area boundary distance from the center of the reactor building is 1190 meters. Part of the exclusion area extends into Lake Ontario. In the event the applicant is unable to gain control over those three acres of shore land within the exclusion zone which he does not now own, the minimum exclu-sion area borndary distance will be reduced to 945 meters. NRC Staff calculations indicate that the applicant can meet the siting dose guide-lines at this reduced distance without additional engineered safety features. B-1

Honorabl3 William A Anders Octobnr 16, 1975 %e SNUPPS will utilize the RESAR-3 Consolidated Version, four-loop pressurized water nuclear reactor with a core power output of 3411 MW(t). his design is similar to that utilized at the Comanche Peak Steam Elec-tric Station, Units 1 and 2, reported on by the Comittee in its letter of October 18,1974. %e Comittee's continuing review of the SNUPPS was reported on in its Callaway letter of Septerber 17, 1975, and is further reported on in this letter. It is anticipated that the Com-mittee's report on the renainder of its review of SNUPPS will be included in its reporti on the Tyrone application. he NRC Staff has. identified several items in its review of the Sterliry application which are not yet cmpleted. %e Comittee recomends that any outstanding issues which may dev'elop in the course of empleting these reviews be dealt with in a manner satis-factory to the NRC Staff. '1he Comittee wishes to be kept informed on the resolution of the followirg iters: 1. he emergency core cooling system evaluatida in cmpliance with the Final Acceptance Criteria. 2. We analyses of the effects of anticipated transients without scram. 3. We evaluation of the plant design to meet the requirenents of the new Appendix I of 10 CFR Part 50. % e RESAR-3 Consolidated Version nuclear design utilizes the Westing-house 17xl7 fuel asserbly. Westinghouse has identified an integrated test program to confirm the safety margins associated with this design, nich it plans to cmplete late this year. W e RESAR-3 reactor core design has been calculated by Westinghouse to.be stable against radial xenon oscillations. Westinghouse has agreed to verliy this stability in a startup physics test for a 193 fuel asserbly core similar to SNUPPS. We Comittee will continue to review these matters as appropriate docu-mentation is sulaitted. he Ccumittee recomended in its report of Septenber 10,1973, on acceptance criteria for EOCS, that significantly improved ECCS capability should be provided for reactors for which construction permit requests are filed after January 7, 1972. W e SNUPPS design is in this category. %ese units will use the 17x17 fuel assemblies similar to those'to be used in Cananche Ibak Steam Electric Station, Units 1 and 2. Although calculated peak clad tenperatures in the event of a postulated IDCA are B-2 l l l

Ibnorabla Willim A. Anderg October 16, 1975 less for 17x17 assablies than for a 15x15 array, the Comittee believes that the applicant should continue studies that are responsive to the Comittee's September 10, 1973 report. If studies establish' that signifi-cant further ECCS improvements can be achieved, consideration should be given to incorporating them into this unit. %e part of the exclusion zone which extends into take Ontario, including the points of intake and discharge of mergency service moling water, will be under control of the thited States Coast Guard. We Comittee recomends that the NIf Staff and the applicant give particular attention i to assure proper coordination between the applicant'and the Coast Guard rppropriate to protection of the emergency equiprent. i te Comittee believes that the applicant and the NRC Staff should continue to review the Sterling plant design for features that could reduce the possibility and consequences of sabotage. We Comittee recomends 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. %is matter should be resolved to the satisfaction of the NBC Staff. We Comittee wishes to be kept informed. Generic problems relating to large water reactors are dimed in the - Comittee's report dated March 12, 1975. % ese problems should be dealt with appropriately by the NBC Staff and the applicant. %e Advisory Comittee on Reactor Safeguards believes that the items mentioned above and the itms mentioned in its Callaway letter, which are relevant to the Sterliry application, can be resolved during construc-tion and that if due consideration is given to the foregoing, the Sterling Power Project Nuclear thit No.1 can be constructed with reasonable assur-ance that it can be operated without undue risk to the health and safety of the public. Sincerely yours, W. Kerr 01 airman B-3

i- ] Ibnor<alo ifillica A. Anders October 16, 1975 REFERENCES 1. SNUPPE Peclimira.f Safety Analysis Reprt with R2 Visions 1 through 10 and the Sterling Site Addendum Report with Revisions 1 throgh 11. 2. RESAR-3 Consolidated Version, Wstinghouse Reference Safety Analysi; Repet with Amerdients 1 tirough 6.. 3. Safety Evaluation Report, NUREG 75/082 related to the Constructica of the Sterling Ibwer Project, Nuclear thit No.1, Docket Ib. Sni 50-425, fleptember,1975. 4. Besolution by the Town of Sterling Town Board, dated Fay 12, 1975. 5. Letter dated September 17, 1975, from Ms. Sue Reinert, Ecology Actics of Oswego. B-4

APPENDIX C EVALUATION OF LIQUID AND GASEOUS EFFLUENTS FROM STERLING POWER PROJECT NUCLEAR UNIT NO. 1 WITH RESPECT TO APPENDIX I 0F 10 CFR PART 50 Introduction The purpose of this appendix is to present the results of the detailed assessment performed to de.;ennine if the proposed Sterling Power Project Nuclear Unit No.1 meets the nu;erical design objectives specified in Sections IIA B, C and D of Appendix I of 10 CFR Part 50. On September 4,1975, the Commission amended Appendix ! of 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 nrior 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 its radwaste management systems to satisfy the Guides on Design Objectives for Light-Water-Cooled Nuclear Power Reactors 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 nf Considerations included with the Appendix I amendment, the Conunission noted that it is unlikely that further reductions to radioactive material releases would be warranted on a cost-benefit basis for light-water-cooled nuclear power reactors having radwaste systems and equirent determined to be acceptable under the proposed Commission's objectives set fortt in RM-50-2. In a letter to the Commission dated October 17, 1975, the Sterling applicants chose to comply with the Commission's September 4,1975 amendment to Appendix I, eliminating the necessity to perform a cost-benefit analysis as required by Paragraph II.D of Appendix 1. Evaluation We have evaluated the radioactive waste management systems proposed for the Sterling facility, to reduce the quantitles of radioactive materials released to the environ-ment in liquid and gaseous effluents. These systems have been previously described in Sections 11.2 and 11.3 of the Safety Evaluation Report, in Section 11.1 of this - supplement, and in Section 3.5 of the Draft Environmental Statement. Based on information provided by the applicants, on more recet operating data applicable to -the Sterling facility, and on changes in our cciculat:cnal models, we have generated i r a liquid and gaseous source terms to determine conformance with Appendix I. These values are different from those given in Tables 3.6 and 3.7 of the Draft Environmental L Statement. C-1

The new source terms shown in TablCs C-1 and C- ? were calculated using the models and methodology described in Draft Regulatory Guide 1.BB, " Calculation of Releases I of Radioactive Materials in Liquid and Gaseous Effluents from pressurized Water l Reactors (PWRs),"datedSeptember9,1975. These source terms were used to calculate the doses as described below. The dispersion of radionuclides in and the deposition of radionuclides from the atmosphere were based on analyses performed by the staff for this evaluation. The mathematical models used to perform the dose calculations are contained in Draft Regulatory Guide 1.AA, " Calculation of Annual Average Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Implementing Appen(ix I," dated September 23,1975. Dose evaluations were made for three effluent categories: 1)pathwaysassociatedwith liquid effluent releases to Lake Ontario, 2) noble gases released to the atmosphere, and 3) pathways associated with radioiodines, particulates, carbon-14 and tritium released to the atmosphere. Estimates of the doses were made for the adult individual (19+ years), the teen (12-18 years), the child (1-11 years) and the infant (1 year), with appropriate values of consumption as given in Draft Regulatory Guide 1.AA. The dose evaluation of pathways associated with liquid effluents was based on the assumed m ximum exposed individual. The dietary and living habits for an adult individual assumes 1) the consumption of 21 kilograms per year of fish harvested in the imediate vicinity of the discharge to Lake Ontario, 2) the consumption of 730 liters per year of water from the Oswego, New York municipal water supply, and 3) recreational use of the shoreline in the immediate vicinity of the discharge of 200 hours per year. We used the applicants' value of shoreline recreation usage, which was greater than the value used in Draft Regulatory Guide 1.AA. For an infant, the diet assumes consumption of 510 liters per year of water from the same municipal water supply. The dose evaluation of roble bases released to the atmosphere included a calculation of beta and gama air doses at the site boundary and total body and skin doses at the site boundary. The maximum doses at the site boundary were determined to occur 0.74 miles north-northeast of the facility. The dose evaluation of pathways associated with radioiodine, prticulates, carbor.-14 and tritium released to the atmosphere was also based on the assumed maximum exposed individual. Such an individual is an infant whose assumed diet includes the consump-tion of 330 liters per year of milk produced at the location of the dairy farm having the highest calculated dose from this pathway. This location is 1.1 miles east-northeast of the site. For the pathways associated with liquid effluents, the adult individual would receive the highest total body dose. The doses from noble gases released to the atmosphere constitute external exposure, and are therefore not age-dependent. For the pathways C-2

TABLE C-1 z CALCULATED RELEASES OF RADI0 ACTIVE MATERIALS IN GASEOUS EFFLUENTS FROM STERLING POWER PROJECT NUCLEAR UNIT NO. 1 (Cf/yr) Unit Vent-Turbine 24 Containment Radw..ste Building 8uilding Nuclides purges per year Vent Exhaust Kr-83m : a a a Kr-85m 5 a' a 'Kr 5 250 a Kr-87 1 a a Kr-88 9 .a a Kr-89 3 a a Xe-131m 9 3-a Xe-133m 22 a a Xe-133 2000 1 a Xe-135m a a a Xe-135 21 ? a Xe-137 a a Xe-138 1 a a I-131 1.1(-2). a 5.2(-4) I-133 1.3(-2) a 6.9(-4) Co-60 7.5(-5)b 7.0(-5) c Co-58 1.6(-4) 1.5(-4) c Fe-59

1. 6(-5) 1.5(-5) c Mn-54 4.8(-5) 4.5(-5) c Cs-137 8.1(-5) 7.5(-5) c Cs-134 4.8(-5) 4.5(-5) c Sr'-90 6.5(-7) 6.0(-7) c Sr.3.6(-6).

-3.3(-6) c C-14 1 8 H-3. 1000 Ar-41~- 25 a = less than 1.0 Ci/yr. noble gases, less than 10-4 C1/yr for iodine. .b = exponential notation: 7.5(-5) = 7.5 x 10-5 ~ c = less than 1% of total for nuclide.

a :-

C-3

-TABLE C-2 4 CALCULATED RELEASES OF RADIOACTIVE MATERIALS IN LIQUID EFFLUENTS -rROM STERLING POWER PROJECT NUCLEA1 UNIT NO. 1

NUCLIDE

' CI/YR NUCLIDE CI/YR . Corrosion & Activation Products Cr-51 1.2(-4)a Cs-134 1(-2) Mn-54' 6(-5) I-135 3(-3) Fe-55 1.2(-4) Cs-136 3(-3) .Fe-59 7(-5) Cs-137 7.8(-3) -Co-58 1.2(-3)- Ba-137m 6.6(-3) Co-60. 4.3(-4) Ba-140 1(-5) -Np-239 '4(-5) La-140 .1(-5) Ai1 others 6(-5) Fission Products Total Except Tritium 1.9(-1) Rb-86 2(-5) Sr-89' 2(-5) Tritium 410 Mo-99 2.7(-3) Tc-99m 2.5(-3) Te-127m 2(-5) -Te-127 '2(-5) Te-129m 9(-5) Te-129 6(-5)~ I-130 1(-4) Te-131m 4(-5) l I-131 1.1(-1) Te-132 9.3(-3) I-132 1.6(-3)~ l I-133 4(-2) -4 a = exponential notation: 1.2(-4)=1.2x10 I. ' l-L l1 C associated with radiciodine and the cther radi:nuclidts. reltased to the atmosphara, an infant drinking milk produced at the dairy locatid 1.1 miles east-northeast of .the site would receive the highest dose. It is necessary to compare the calculated doses from the plant with ti.e Design Objectives contained in the' Concluding Statement of Position of the Regulatory Staff (RM50-2). Tables C-3 and C-4 provide a comparison of the calculated doses, with the design objectives of Sections IIA, B and C of Appendix I and the proposed Commission's design objectives set forth in RM 50-2. As shown in Table C-2 the expected quantity of radioactive materials released in 1iquid effluents from the plant, excluding tritium and dissolved gases, will be 0.19 curies per year, or less than five curies per year, which is in conformance with the amendment to Section II.D. The liquid effluents released from the plant will not result in an annual dose or dose commitment to the total body or to any organ of an individual, in an unrestricted area from all pathways of exposure, in excess of five millirem (Table C-4). -Based on our evaluation of the gaseous radwaste management systems, the expected total quantity of radioactive materials released in gaseous effluents from the plant will not result in an annual gamma air dose in excess of 10 millirad and a beta air dose in excess of 20 millirad at every location near ground level, at or beyond the site boundary, which could be occupied by individuals (Table C-4). As shown in Table C-1 the annual total quantity of iodine-131 released in gaseous effluents will be 0.011 curies per year, or less than one curie per year, which is in confirmance with the amendment to Section II.D. The annual total quantity of radiciodine and radioactive particulates released in gaseous effluents from the plant 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 (Table C-4). Conclusion Our evaluation demonstrates that the expected doses associated with the nonnal operation of the Sterling facility meet the design objecthes of Sections II.A. II.B and II.C of Appendix I of 10 CFR Part 50, and that the expected quantity of radioactive materials released in liquid and gaseous effluents and the aggregate doses meet the design objectives set forth in RM-50-2. Our evaluation shows that the applic nts' proposed design of the plant satisfies the criteria specified in the option provided by the Commission's September 4,1975 amend-ment to Appendix I and, therefore, meets the requirements of Section II.D of Appendix I of 10 CFR Part 50. . Based on our evaluation, the proposed liquid and gaseous radwaste management systems for the Sterling Power Project Nuclear Unit No. I meet the criteria given in Appendix I and are..therefore, acceptable. C-5

,-I,-, ~ ~ TABLE C-3 c.,.. -COMPARISON OF CALCULATED DOSES FROM STERLING OPERATION WITH a SECTIONS II. A, II.B AND II.C OF APPENDIX I.10 CFR 50 Appendix I Calculated .. Criterion Design Objectives Doses Liquid Effluents Dose t'o total body from all pathways (adult) 3 mrem /yr .046 mrem /yr-Dose to any organ from. -allpathways(adult) -10 mrem /yr '.058 mrem /yr Noble Gas Effluents. Gama dose in air.~ 10 mrad /yr 0.23 mrad /yr ~ Beta dose in air 20 mrad /yr 0.54 mrad /yr ~ Dose to total body of an individual (child)' 5 mrem /yr 0.14 mrem /yr ' Dose to skin of an individual (child) 15 mrem /yr 0.41 mrem /yr b Radiolodines and Particulates Dose to any organ from all pathways (infant) 15 mrem /yr 6.9 mrem /yr 'a.As presented in the Federal Register V. 40, p. 19442, May 5, 1975. bCarbon-14 and tritium have been added to this category. ? t C-6

a TABLE C-4 COMPARISON OF CALCULATED DOSES FROM STERLING OPERATION WITH a GUIDE ON DESIGN OBJECTIVES PROPOSED BY THE STAFF ON FEBRUARY 20. 1974 (DOSES TO MAXIMUM INDIVIDUAL FROM THE ONE UNIT ON SITE) RM-50-2 Calculated Criterion Design Objectives Doses Liquid Effluents Doses to total body or any organ from all pathways (infant) 5 mrem /yr .058 mrem /yr Gasecus Effluents Gama Dose in air 10 mrad /yr 0.23 mrad /yr Deta dose in air 20 mrad /yr 0.54 mrad /yr Dose to total body of an individual 5 mrem /yr 0.14 mrem /yr Dose to skin of an individual 15 mrem /yr 0.41 mrem /yr b Radiciodine and Particulates Dose to any organ from all pathways (infant) 15 mrem /yr 6.9 mrem /yr afrom " Concluding Statement of Position of the Regulatory Staff," Docket No. RM-50-2 Feb. 20, 1974, pp. 25-30, U.S. Atomic Energy Commission, Washington, D.C. bCarbon-14 and tritium have been added to this category. C-7 . _. =

APPENDIX D CHANGES AND CORRECTIONS TO THE SAFETY EVALUATION REPORT Page 2-5, Table 2.1 Change 2,788 to 2.778 Page 2-9, First Paragraph, First Sentence Insert "majer" between " closest" and." highway" Page 2-9, Fourth Paragraph, Third Sentence Change "Hancock Air Force Base" to "Hancock Field" Page 2-10, First Paragraph, Second Sentence Change " portions" to " climate" Page 2-10, Second Paragraph, First Sentence, First Line Insert "about" between "on" and "five" Page 2-10, Fourth Paragraph, Sixth Sentence Change "one to "two" Page 2-14, Second Paragraph, Sixth Sentence Change the first " sea" to lake" Page 2-14, Fourth Paragraph, First Sentence Change " site" to " plant" Page 2-17, First Paragraph, Fourth Sentence Change "233.0" to "237.0" Page 2-17, Fourth Paragraph, Second Sentence Insert " Site" before " Elevations" Page 2-19 Third Paragraph, First Sentence Change "Clarenden-Linden" to "Clarendon-Linden." Page 2-19. Third Paragraph, First Sentence Change "and" to."that" Page 2-20, First Paragraph, First Sentence Change "across" to."along" Page 2-23, Second Paragraph, First Sentence Insert ", exclusive of the marsh areas," between " area" and "can" Page 2-23, Second Paragraph, Second Sentence Change " medium" to " fine" Page 3-5, First Paragraph First Sentence Change " comprehensive" to " compressive" D-1

'Pege 3-11. Third Paragraph, First Sentence Change " shipping" to " whipping" Page 4-6. Third Paragraph. Eighth Sentence Change " demonstrates" to " indicates" Page 6-7, First Line Change "2,000,000" to "2,300,000" Page 6-9, Seventh Paragraph, Second Sentence Change " leak" to line" Page 8-2, First Paragraph, Fourth Sentence Change "will" to "may" Page 8-2, Third Paragraph. First Sentence Change " Number 2" to " Number 1" Page 8-3, Sixth Paragraph, Eighth Sentence Change "any" to "the" Page 13-2, Third Paragraph, Third Sentence Delete "research" Page E-4, Reference 33 Change " Quad Angle" to " Quadrangle" D-2 .}}