ML20235C010
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{{#Wiki_filter:* ~ fest'fr-So 36.2. c - 43 1 + JDcdKJT -50 353 ~ ~ A 0 f November 29, 1971 . 'ji% '$9l U,'U;f;j ,SATETY EVALUA1 t0N ,e;, -.Y' BY,THE rr. ppt s -", vir - h.., g DJVISION OF REACTOR LICENSING 1 1 f U.S. ATOMIC ENERGY 00XMI5SION { p-IN THE MATTER OF U' PHILADELPHIA ELECTRIC G)MPANY LIMERICK CENERATING STATION UNIT NOS.1 AND 2 l DOCKET HOS. 50-352 AND 50-353 ' I s e' h i I i i i i 8709240319 870921 PDR FOIA D MENZ87-111 PDR ) pSTRi3UT:CW Gi TH'S DOCUMENT is UMLP.MTID E t(
4 s i. i4 i fd 1 !$kg TABLE OF CONTENTS $k gg ABBREVIATIONS i l l
1.0 INTRODUCTION
1 j 2.0 SITE 4 l i c J 2.1 General 4 2.2 Meteorology 6 2.3 Eydrology 8 2.4 Geology. Seismolog, and Soil Mechanics 11 { 2.5 Environmental V ution Monitorias 14 i 2.6 Air Traffic 13 i' ! 2.7 Railroad and Rive e 1.-affic 16 '! d, 2.8 Conclusions 18 jQ 3.0 REACTOR DESIGN 18 0 i j 3.1 General 18 I t 3.2 Nuclear Design 19 i 3.3 thermal and Hydraulic Design 23 i [ 3.4 Reacto-Internals 25 1 1 3.4.1 Design 25 ( 3.4.2 Dynamic System Analysis for Seismic, Operating l' g and IDCA Imadings 26 3.4.3 Vibration Control 28 4.0 REAcr0R COOLANT SYSTDI 29 {'* 4.1 Ceneral 29 4.2 Reactor Coolant Pressure Loundary (RCPB) 30 4.2.1 Definition 30 j 4.2.2 Design 31 4.2.3 Pipe Wip Criteria 32 4.2.4 Main Steam Line Isolation valve (HSLIV) Leakage 33 4.2.5 Seismic Design of Main Steam Line Piping 33 L 4.2.6 Primary System Pressure Relief System 34 t' 4.3 App 1f cation of AEC Classification Croups 35 4.4 Reactor Vessel Material Surveillasca Program 37 4.5 Fracture Toughness Criteria 38 4.6 Inservice Inspection 39 l
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D9 gg );*q "4*J., TARLE OF COSIEN15 (cont.) t ?i%=4 L f SgI Zn'i l I 4.7 Reactor Coolant System Sensitized Stainless i
- l Steel 40 l
'f [.. 4.8 Electroslag Welding 41 4.9 Foreign Procurement 41 r ! : 4.10 Leak Detection 41 4.11 Corrosion Cracking of Pipe Metal 43 5.0 CONTAINMENT AND STRUCTURAL DESIW 44 5.1 containment Design and Comparison 44 5.1.1 Primary containment 44 5.1.2 Secondary containment 51 5.1.2.1 Reactor Building Recirculation' B. Systen 52 5.1.2.2 Stundby Gas Treatment System 55 5.2 Structural Design 56 Il i 5.2.1 Class 1 (Seismic) Structures 56 5.2.2 Environmental Effects 57 5.2.3 containment Structursi Design Analysis' SS 5.2.4 Testing and Surveillance 59 5.3 Seismic Design 60 5.3.1 Seismic Input 61 5.3.2 Seismic System Dynmic Analyses 61 ?$N 6.0 ENGINEERED SAFETT FEATURES 62 j 6.1 Emergency Core Cooling Systems (ECCS) 62 6.1.1 General 62 6.1.2 Emergency Core Cooling System Performance Evaluation 66 6.1.2.1 General 66 6.1.2.2 Discussion of ECCS Review 67 6.1.3 Conclusion 73 6.2 Hydrogen Generation in Primary Containment Following A LOCA 74
b<'. L 1 o., } i 1 'i M TABLE OF CONTINTS (cont.) W o. ^ 6.2.1 Rydrog. o Control System 74 6.2.2 Containment inerting 75 6.3 Other Engineered Safety Features 75 7.0 INSTRUMENTATION, C0dTROL AND ELECTRIC F0WER SYSTEM 75 7.1 Instrumentation and Control Systems 75 f 7.1.1 Post Accident Monitorin8 Instrumentation 76 h 7.1.2 E:virc :.sntel T tlu
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76 7.1.3 Enginested Safety Feature Testing 77 7.1.4 Design Criteria for Cable Installation 1 77 7.2 Electrical Power Systems 78 l 7.2.1 Offsite Power 78 7.2.2 Onsite Power 79 I8.0 AUXILIART SYSTEMB 81 L 8.1 General 81 8.2 Radioactive Waste Systems 4 81 8.2.1 Cenersi 81 8.2.2 Liquid Radwasta System 82 8.2.3 Caseous Radvasta System 86 8.2.4 Solid Radweste Systen. 89 8.3 Spent Puel Storage 90 8.4 ESW and RERSW Systems 91 8.5 Main control Roon Ventilation System 92 1 9.0 ACCIDENT ANALYSIS 92 9.1 General 92 9.2 Loss-of-coolant Accident (LOCA) 94 i N 9.2.1 Analytical Model for Pressure Suppression 94 9.2.2 LOCA Exposure Doses 95 9.3 Fuel Handling Accident (Rafueling Accident) 96 9.4 Control Rod Drop Accident 97 9.5 Main Steam (mL) Line Break i 98 9.6 Control Room 1xposure Doses During Accidents 99 9.7 Main Stems Line Isolation Valve Leakage 101 9.8 Instrument *.t.: or Process Line Break 102 9.9 Cryogenic Treatment Subsystem Accident 102
l l s. u ) 1. p 1 .m.j l-TABLE OF C.WTENTS teont.) L 'r.%U ?Ai& 10.0 CD:GUCI 0F OPERATIONS 103 l 10,1 Technical Qualifiestions 103 g 10,2 Organization of Plane Management 104 j {p[j f 10.3 Operating Procedures 106 j k 10.4 Startup, Pteoperation, and Power Tests 106 j L.h i 10.4.1' construction and startup tests 106 .a jl 10.4.2 Preoperational tests 107 l 10.4.3 Startup and power tsst program 107 i
- 10.4.4 Conclusions 108 10.5 Emergency Planning and Plant 3ecurity 108 l.
10.5.1 E:scrgency plans 108 a 10.5.2 Plant security 109 10.6 Independent Safety Review of Operator Actions 109 11.0 QUALITY ASSL' NANCE 110 1 12.0 TEmMCAL SPECIFICATIONS 1 12 { 13.0 REPORT OF TEE ADVIS0EY COM(ITTEE ON react 0R SAFECUAES 11 3 f 13.1 Process Line 3reaks 114 -iW.. 13.2 class I (Seismic) Main Steamlina (MSL) Design 114 I "- 3 13.3 MSL Isolation Valve (MIL-IV) Leakage 1 14 13.4 Reactor Shield lit i 13.5 Emergency Core Cooling System lh i 13.6 Radioactive Waste Disposal 1? 13.7 Anticipated Transients Withouc der.an (A'1WS) lit .l 13.8 Hydrogen Contrcl and Containment Inerting 1 16 13.9 Maximum Ground Acceleration 116 i 14.0 CONFORMANCE TO CENERAL DESIN CRITERIA 116 i j 15.0 COP 90N DEFENSE AND SECURITY 117 a~ { 16.0 FINANCIAL QUALIFICATIONS 117 17.0 (DNCLUSIONS 120 i i i i 5 i s 4,
I MI y s D ^ t L TABLE OF 00N1DiTS -(cont.) j' a w-P_ age b Liat of Tchles u < .3.1 Comparison of BWR Design Parameters 28a 4.3 AEC code classifications 37a l 5.1 Comparison of Containment Design Parameters 44a { 6.1 Comparison of ECC5 Capabilities 63a ,j I A?PENDICES f . Appendia A - Gironology of Events 123 Appendix 3 - Esport ei LLs AJvia.ury Cosmaittee on Reactor Safeguards (ACRS) 127 I g Appsadiz C - Reports of the Air Resources Environmental (1 & 2) Laboratory, National Oceanic and Atmospheric B a Administration 131 I I Appendia D - Report of the United States Geological Survey 135 Ii e y Appendia E - Report of the Seismology Division of the National Ocean Survey, National Oceanic and Atnaspheric Administration (formerly U.S. Coast. and Geodetic Survey) 141 i 1 Appendiz F . P (1, 2, 3) - Reports of the Fish and Wildlife Service 145 ~ ! l (4) - Applicant Response to Reporta 153 I Appendia C - Report of the Advisory Council on Historic l Pres ervation 167 3 Appendiz H - Report of Nathan M. Newmark Consulting Engineering l, Services 169 .i i il Appendia I - Financial Analysis 171 j i i J l, n
n W!i ABBkEVI/ CIONS h %.^ n f}M;9 ' ACI /nerican Concrete Institute
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ACitS Advisoty Connittee on Reactor Safeguards AD6 Automatic Depressurization System. y ABC United Ffates Atomic Energy Cossaission l AMS American Nuclaar Society { ANSI. American National Standard Institute i fred Atomic Power Equipment Department (GE) R API American Petroleum Institute l ASME American Society of Mechanical Engineert ASTM American Society for Testing and Materials -I BTU /hr-ft AWS American Welding Society } British Thermal Units per hour per square foot WR Boiling Water Reactor CIM Cubic feet per minute Ci/sec Curies per second r) CSCS Core Standby Cooling System W il CSS Core Spray System 1 DBA Design Basis AccHent DBE Design Basis Earthquaka y d-c direct current DEL Division of Reactor Licensing EOCS Emergency Core Cooling System H i o r i h ? l e \\ s 1 &n I
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l f 3 11 i 4 ft feet FSAR Final Safety Analysis Report i g acceleration, 32.2 feet per second per secots GDC AEC General Design Criteria Fbr Nuclear Power Plant Construction Fermits ( GE General Electric Company sps gallons per minute NEPA High Efficiency, Particulate, Air HPCI High Pressure Coolant Injection System 1EEE Institute of Electrical and Electronics Engf cours in inch k,gg effective multiplication factor (for the nuclear r fission process) Ak/k/'F reactivity (Ak/k) change per degree Fahrenheit 1 ERS Gundremmingen Nuclear Tcver Station (Germanyt I kV kilovolts kW/ft kilowatts per foot i Ib pmed LOCA Loes-of%1 ant Accident LPCI Low Freasure Coolant Injection System MQiD Minimum Critical Heat Flux Ratio i 2dified Mercalli (earthquake intensity) WC maximum permissible concentration aph miles per hour WD/ ton megawatt days per ton we megawatts electrical
.tC .. iii. ' 6:F h Nt megwatts thermal 1 -{ ]' NOAA National Oceanic and Atmospheric Administration NPSH not positive suction head i NSSS Nuclear Steam Supply System OBE Operating Basis Earthquake PEco Philadelphia Electric Company I 4 ppa Parts per aillion PSAR Preliminary Safety Analysis Report ',i-poi pounds per square inch 1,
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pois pounds per square inch gauge f QA Quality Assurwice l 1 } QC Quality Control l n i R&D Research and Development l RCPB Reactor Coolant Pressure Boundary RER Residual Beat Removal SBCTS Standby Ces Treatment System TEMA Tubular Exchanger Manufacturers Association j USGS United States Geological Survey w/o weight percent l uc/mi microcurie per milliliter lk - pc/see mic.ocurie per second i 10 CFR AEC, Title 10, Code of Federal Regulations Part 2 AEC Rules of Practice j, Part 20 AEC Standards for Protection Against Radiation Part 50 AEC Licensing of Production and Utilization Facilities Part 100 AEC Reactor Site Criteria J;' (
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1.0 INTRODUCTION
t ~ This report is the Atomic Ener8y Cornission's safety evaluation ,L.e-lyg of the application by the Philadelphia El+ ctric Cotpany for a l }h}:,qll construction permit and operating license for a dual-unit nuclear ljmj pcuer plant facility. The site for the Limerick Generating l-Station Unit Nos.1 and 2 is on the Schuylkill River,1.7 miles g. l i from Pottstown, Pennsylvania, and is about midway between ia a Reading and Philadelphia, Pennsylvania. 1he distance from the site to Philadelphia's city limits is about 21 miles. 4 1he Limerick Cenerating Station will use two identical boiling water reactors each with a rated thermal utput of 3293 W and a gross electrical output of 1152 m at the generators. The ulti-y mate thermal power of each reactor is anticipated to be 3440 W. The safety review and evaluation of the preposed Limerick facility 5' were based on the slightly higher thermal power level of 3440 W. 0 A The General Electric Company ~will supply the nuclest steam supply l a, ,) systems (NSSS), nuclear fuel, and' the turbine generators. Bechtel Corporation will furnish engineering and construction a services for the design and construction of the plant, integra-W.,, ting the itecs furnished by GE with couplete balance-of-plant i j. items; Bechtel is also responsible for procurement of all equip-5 j ment except the NSSS and turbine generators. Fabrication and f subsequent onsite asseubly of the reactor vessels will be accom-f pJished by the 011cago Bridge nd Iron Company. D 1 ( t i g _. _..
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m. (yg4). MN \\ n 2- ) y,$ 7M g ) g Our technical safety review and evaluation of the facility hsve j \\ ~~ l been based on the applicant's Prel'minary S2fety Analysis Report l d( '. (PSAR) and thirteen subsequant amendments thereto. During our L \\ ] review of the application we bald erny asetings with representa-3 b tives of the 'appitcant and its consultants to discuss the facility and the technical material subsitted. A chronological listing of the netting and other sig.ific::t evest: h give: 1: 4pcndix A to this evaluation. As a result of our review, we required a nue-bar. of changes to be made in the facility design; these changes i are described la the applicant's thirteen amendments to the PSAR l and are discussed in appropriate sections in this report. All of these documents are available for public inspection at the U.S. Atomic snergy Commission's Public Document Room,1717 H Street, i l N.W., Washington, D.C. and at the Pottstown Public Library, I [ Fot stown, Pennsylvania. i Our technical evaluation of the preliminary design of the facility i l t was accomplished with the assistance of consultants. The reports j of our consultants on meteorology, hydrology and geology, seis- ? mology, fish at 1 wildlife, and structural design are included as ( l appendices to this report. We have received a letter from the 1 Advisory Council on Historic Preservation which is attached to this evaluation. j
F ci ,[ ll' e M The Atcade 1*nergy Conair.cion's Advisory { aittce on Nactor Safeguards (AC2S) also reviewed the application. An ACRS 'sub-l ccanittee with representatives of our staff made a site visit on IIc vember 10, 1970. Subsequent ACES secommittee and iall connit-m toe meetings culminated in the ACRS report to the Coeurission -hat is included as Appendiz 5. This safety evaluation summarises the results of the tschaical 4 evaluation of the application for a construction permit. Many + features of this facility are aimilar to those we have already ,y evalasted and approved for other reactors new under construction I The review and evaluation 'of the proposed facility at the con-or la operation. To the extent feasible and appropriate in our revise, we have made use of these earlier evaluations. R,.. struction permit stage is only the first stage of the continuing review of the design, construction, and operation of the facility. Should the applicant receive a construction permit, prior to the issuance of an operating license, we will review the final design to determine if all the Counission's safety requirements were met. The facility would then be operated only in accordance with the terus of the operatins license, the Commission's reguistions, and i under the continued surveillance of the Commission's regulatory staff. ) l N
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- SJTE, 2.1 Gener'al l'
The 587-acre site for the Limerick Generating Sts' ion (LOS) is t located on the Schtylkill River 1.7 miles southeast of Pottstown, l Pennsylvania and about.e.idway between Philadelphia and Raeding. ] Philadelphia is about 21 miles away. lt he site is divided into three segments. The principal portion, where the major operating equipment and structures would_be located, is en the east bank of the river.- This portion is ] separated free the second segment, where cooling water intakes auld be located, by the triple tracks of the Reading Railroad's l c M n line. The third portion lies on the west bank of the j Schuylkill River adjacent to the tracks of the' Penn Central -e Railroad. 1 I The topography of the area is hilly, with elevations within five i milee of the site ranging from 110 feet above mean sea level (E) at the Schuylkill River to 560 feet Mst. The plant grade vill be at 218 feet MSL. The track of the Reading Railroad is at e 143 feet MSL, about five hundred feet from the plant. 'neo small l y l parallel streams run scuthwest through the site to the river. ) The minimum exclusion distance available is about 2500 fee * (760 meters) from the center of each reactor. There will be no h y,
w on '- ' sf f4 private reef dences located inside tha exclusion area. A sus 11 silt-fGreed island in the Schuylkill River is within the uclusion f r i area; the island is uninhabited and is owned by the emmanwealth 3:U of Pennsylvania. On some of the land inside the exclusion area, farning any be permitted and public recreation areas may be pro-l vided under the control of Philadelphia Electric Company. An information center for the public will also be built on the property. The city of Pottstown, the nearest desely populated center, had Hi a 1970 census population of 25,355 and a predicted population of 50,000 for the year 2000. The Population Center Distece (PCD) is 1.7 miles, the distace to the city limits of Potestavn. The Lev Population Zone distance (LPE) has been selected at 1.27 miles within which the 1970 resident population was found to be about 1071. 4 There are two major industrial facilities with a total of about 850 employees within the PCD (1.7 miles). The applicant indicated that nearby industrial activities, the two railroad lines (one through the exclusion area), and a stone quarry will operate without detrimental effect on the nuclear facility. Our study l 4 and that of the applicant have confirmed that no military installations are ' located within five miles of the site. A pri-vate swinuming club located about 1.3 miles from the site will be g
l' l 6- /d k.e operated during the wamer months of the year. n e maxiuum i attendance at this pon1 la estimated at approximately 80J ;eople, but on the average will be less. he pool usage will be reces-nised in preparation of emergency plans. Based on our evaluation of the population data, we conclude that' ) 4 the distances established for the exclusion sone, the low popula-j tion sone boundary, and the nearest ' population center me the guidelinas given in 10 CFR Part 100. il' h e applicant haa presented meteorological data fr= Vesther Bureau records ~ for Philadelphia and Reading airports (each about 25 miles from the site) and from the applicant's asteorological i j studies at its Peach Bottom site (about 50 miles away). Do data R3 f on atmospheric diffusion at the site has been provided for our review. De applicant has relied on extrapolation of Peach Bottom hkj data, converting from Brookhaven stability categories, to a form i i usable in classifying the wind data into appropriate Pasquill stability categories. He converted data were submitted in Appendix N (Amendment 8) to the Final Safety Analysis Report for 4 j Feach Botton Nuclear Generating Station, Unita 2 and 3, and have been found to give, in the most stable period, a Pasquill stabil-ity of Type F with a wind velocity of 0.7 meters per second. We meteorological situation at Limerick should be similar to that at t ,s V : w-
k i i ?;;. I Jw i Peach Eat tom since the general cli satcle,y and topography are j aint14r and they are about 50 miles apart. Our consultants at l. National Oceanic and Atmospheric Administration (NOAA) have re-viewed the Peach Bottom data and found it to be acceptable for use at Limerick. A copy of the consultant's rtport is attached as Appendiz C. For calculation of potential two hour exposure de:e: frc= airbe::e ;sdistien, we have used conservative meteorological assungtions of Type F and 0.5 meter per second . wind speed. The final onsite determination of actual conditions will await the collection and analysis of data following the applicant's planned installation of asteorological tavers and instruments. tro instrumented meteorological towers and one instrumented pole (for low-level monitoring) will measure and continuously record ., 9 speed, wind direction, and temperature from varied elevations. ' F.g:., Other instruments will acasure and record humidity, barometric pressure, and precipitation. Onsite locations of the weather towers were selected on the high ground northwest of the reactor and across the Schuylkill River in an open field; the pole is j located near the confluence of Possum Hollow ha and the Schuylkill River. l We have concluded that the meteorological program is acceptable. l
1 I I -e A l j 2.3 Hydrology I Me Schuylkill River bein extends about 80 miles above its con-fluence with the Delaware River. %s tota 1> drainage area is relatively small, being only 1900 square siles above the Limerick Station. Se applicant has esiculated a Probable Maximum Flood (PIf) discharge which corresponds to a discharge stage elevation T. _ _ < of 159 feet MSL at the site. The===i== flood level for which this nuclear facility will be designed is the PNF as defined by 1 the U. 8. Corps of Engineers. Se plant grade is at 218 feet MSL; bence, all safety-related plant facilities, except Schuyrstill I River water supply structures, are well above the PMF stage ele-vation. ne applicant has indicated _ that cooling water intake i structures, lines, and the pump house which are located below the i } PMF stage will be protected spinst flooding. We emelude that I l there will be adequate flood protection for safety related struc-I h tures and equipment. I 1 j Two natural draft, hyperbolic cooling towers will be employed to dissipata vasta heat to the atmosphere. Cooling water for makeup to the main circulating water system will come from either the Schuylkill River or the Delaware River, depending on the water 5 y usage permit obtained by the applicant. Delaware River water, if l used, will be routed via pipelinea and open channel to the site. %e total requirement for normal cooling water for the facility f operation is stated as 74 cfs, including 54 cfs for consumptive i
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25 y I use (cooling tower windage and evaporation). Dconstresa conste.p-
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tive requirements for Schuylkill River water are 467 cfs. The minimum instantaneous flow of record for the Schuylkill River at -j Pott <. town was 87 cfs in 1930. To provide an additieaal source of ~ l cooling water, Philadelphia Electric Company is completing agree-- ments with the Delaware River Basin Commission (DRBC) and other agencies to use Delimiste River' water. The Delamrare River wetar ~M will be piped eight miles from a proposed Point Pleasant pumping I. station (ownership and operation under consideration by DRBC) to the headwaters of the East Branch of the Perkiomen Creek where the water will he routed by open channel (gravity) flow through the East Branch for shout twenty miles to a pumping station near Craterford to be built and operated by the applicant. Consumptive cooling water requirements for the Limerick Station may then bc j extracted and pumped through a pipeline along the existing electrical transmission line right-of way for about eight miles to the site. About.20 cfs "nonconsumptive" water requirements will be obtained from the Schuylkill River for normal cooling. Nonconsumptive water is returned to the Schuylkill River as " blowdown" or " dilution water" from the cooling tower basin; 4 hence, diminishing of Schuylkill River flow would be minimized. i Shutdwn of one unit because of a design basis accident 'in one l tmit and normal shutdown of the other unit requires a nonconsumptive
rg.. l [ _ to _ j ,i-supply of ecolleg vater. %e rmrefncy shutdovn just described l ) }l would require a r_inimum cooling water flow of 38 cfa. We have concluded that the supply of,the water for this purpose is i adequa te. l There are seventeca major water intakes located along the ) gehuylkill River between the site and the Delaware River con-fluence in Philadelphia. The es;ccitic: cf==Mipal ussrs ranga I up to that of 198 millson saltons per day for the city of ~ 1 Phtisdelphia. To determine the mixing and dilution capacity of the gehuylkill River as it flows downstream, we have requested the applicant to evaluate dispersion and dilution factors for various ficw rates at the eleven downstream intakes. This infor-1 mation will be provided in the Final Safety Analysis Report. All the gromd water aquifers in the area occur in joints, frac-tures and other secondary openings in the consolidated rock be- ? neath the thin cofor of residual soils. The applicant has performed pumping and ground water percolation tests at the site, and has concluded that the residual soils are very impervious to watne movement and that only water wells generally in the direc-l tion of the strike of the strata may be affected by conditions
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st the site. After study of regional geology and the location of public ground water wells, the applicant has concluded that all such walls cannot be affected by conditions at the site. The l ) ) 5 k 't
-u-w, applicant has found approximately 77 wells for donestic water supply within one mile of the site. We agree with the applicant's conclusions that none of these wella can be affected by conditions i at the site because of their relative locations and depths,. the location of the river, and the local groistd water gradiants.
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fe_olgty. Seismoloar. and 3o11 Nechanics eo The site is in the Triassic Lowland section of the Piedmont Phymf a-- f graphic Province, a section characterized by gently rolling sur-face on en eroded low pistasu. A shallow residual soil overlies Triassic sediasatary rock of the Newark group which is about 8000 feet thick above the Palaczcic ar.d Prcestrian basamant roda. I Appalachian Orogenic belt, marked by a northeast-southwest orfan-The dominant structural feature of the region is the Regional tation (roughly perpendicular to the $daylkill at the Limerick Station) of the axis and linaation of the rock structure. Se nearest faults _ to the site are 9 miles to the southwest, a Paleo-s, [ soic thrust fault inactive for 200 million years; and 9 miles to the northeast, a Triassic strika slip. fault, inactive for 140 7 million years. Measurements at the site and laboratory analyses of site borings indicate that plant structures, such as the Reactor Building or Radwaste Building, will be founded on competent bedrock. Se 1
i{ Q .i - 12'- bhY i. N M;O y j i,M 1 lM applicant's criterion of founding :111 Class. I aciswie structurar / 1 i on coupe. tent bedrock is acceptabla. b Descriptive data on rock-fracture tones which are uncovered d excavation at the site were r m ested. One rock-fracture zone, discovered during excavation, will be stabilised sadsfactorily. t 0ther fractures, if encostored, will be stabilised against dif-fcrestici cettic:. cat er ccacclidctics by concrctc cpr.ning the fracture sono. A rock fracture known as the Sanatoga fault passes 1300 feet we of the alte and has not been active for 140 at111on years. i Otrar similar rock fractures are found in the lisvark basin fur the site but are also inactive. The area of the site has not experienced major earthquake activity.as far back as the early 18th century. .d Most such activity in southeastern Pennsylvania occurs is the Piedmont west of the fall sone, the physiogr : hic botadary between the Piedmont and the Coastal Plain occurrin about thirty miles southeast. The closest Intensity VII earth-quake was about thirty-five miles from the site near Wilmington Delaware, one hundred years ago. } I! , A definition of Class I and II seismic structures is given in section 5.2.1. A d i i ,p
. sp [ p.j Subterranean caverns, natural and nanmade, can be found in the StL-g, gene:al tegion; for Ir. stance, there are coal mines in the area aro and Wfik a-Barre, Pennsylvania, which is sixty-five miles to i the north of the site. Collapse or cave-in of aderground caverns - 1 have occurred; records exist for such occurrence at Wilkes-Barre. However, the shoch or earth tremors associcted with these cave-I ins are local and not transmitted over great distances. Following Y engineering seismology studies, the appiteent's consultant. Dames and Moore, concluded that the bedrock in the site area is compe-test siltstone, sandstone, and shals not subject to cavitation as in the Allentown area, and la not mined as la the Wilkes-Barre ares; hence, caverns are not in evidence around the site. Our scismic consultants at the United States Geological Survey and at the 8eismology Division of the National Ocean Survey of the National Oceanic and Atmospheric Administration (formerly U.S. g Coast and Geodetic Survey of the Envi' nmental Science Services Administration) concurred in the applicant's selection of the maximum ground horisontal acceleration of 0.063 resulting from an intensity VI (m) earthquake, to be used as the Operational Basis s Earthquake (08E). The consultants also concurred in selection by ,"l,$ the applicant of a value of 0.12h for the acceleration associated i with an intensity VII earthquake, to be used as the Design Basis 4 Earthquake (ISE). The Advisory Committee for Reactor Safeguards ~ ..~.- _- ~ %{
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l I [: i suggasted, in their August 10, 1971 letter to the Cenaissica } ) (Appendix 3), a more conurvative value of 0.153 for the mainum f 3round horisontal acceleration for the DBE. A vale.e of 0.153 for ) the DBE and a value of' 0.0753 for the 03E seismic design will be used. Of special interest in foundation design is the,use of engineered g. fill as a slope of 2:1 that will be provided for support or the f. Class I piping fran the schuylkill River pumphouse to the plant, N The applicant has given specific attention to the design and { placement of this subankment to assure satisfactory soil stability under reatic or dynamie leading conditions. 2.5 Environmental gadiation Monitch The applicant has presanted, in Amendment 11 to the PSAR, a description of the Preoperational surveillance Program. Regun in i February 1971 with analysis of water samples, the prograr has ex-panded and is bains performed by the Radiation Eanagcaent Corporation. A definition of the expanded sampling program, to include the number of samples, types, locations, frequency, and specific analysis, is presented in Amendment 11. Our review of the Preoperational Program revaala that a thorough prngram of data collection and analysis is planned for use as a basis for program modification and eventual preparation of the follow-on phases of the*overall program, namely, the Initial Operational Phase and the Continuing Operational Phase. ,A
v - 15 f Ou: consultant, the Fish and Wildlirr. ?ervice, has pmvided reports on its review of the app 11c,a's Preliminary Safety a,na y Analysis Report. The censultut reports were sent to the appli- ~ cent for considerate <.u and use in the preparation of the Faviron-asetal Radiation Monitoring Progras. Copies of the reports and- ./ the applicant's response are attached as Appendices F-1, F-2, ?$ P-3 and F-4.~ 9 Coordination by the applicant and its consultant with numerous L federal and state agencias was accomplished during the preparation of t2. program. I ve111ance of the envirens of the timerick site. In summary, we conclude that the applicant has described a workable and complete program for environzaontal radiological sur-We will continus our review of this progran at the aperating license stage o" the i review to assure action has been taken for its continuation.
- 2. 6 Air _ Traffic 1
r l Air traffic in the vicinity of the site has been reviewed; two ji ana11 airfields for small aircraft are nearby. v One field, owned b .g by the applicant, is about one mile to the northeast of the plant site. A f w small aircraft use this field, but the runway is oriented so that its extension does not pass over the site. About 15 flights per day is the average for the Pottstown u .s. ....... ~.,- e ruc h +* * * * " " ~ ~'"~ ~ ~ ~~ .j ~ ~~~ ~" 1
'r .I& l/ j"8 l. Municipal Airport which lies 4.3 miles northwest of the plant site. This airport has no scheduled airline service, but servas j *. a charter service, flying school, and some privately-ownd air-f] craft. The plant does not lic in the approach pattern for this t: f,j g airport and the rumtay's axis is oriented away from the plaat. i I: .k Connercial. aircraft to and from the Philadelphia area fly pat-terns such that most flights are above four thousand feet la the area of Pottstorm. Some local commercial traffic controlled by l.c Philadelphia Departure Control will be flying below four thousand feets a hip traffic day (July 2,1970) produced 25 local flishts. There are no holding patterns within five miles of the site. i i Flight frequencies, fli$t patterns, and type of aircraf t flying in the vicinity of the site are not significantly different fron hl! that in the ganaral region around Philadelphia. We conclude that l y no special provision need be made in plant design due to aircraft q flights in the area around Limerick. l
- 2. 7 Railroad and River Traffic r'
9 The Reading Railroad main line tracks run along the north bank of the Schuylkill River and traverse the site about five hundred [ feet from and below the plant. About fif teen "through" freight d. { trains (mtly in the late evening and early morning hours), one 5. h
11..q q .f I h"li ' f local freight train, and fourteen two to three car passenger lj (e i trains (batveen 6 a.m. and 11 p.m.) pass through the exclusion
- l E.J area each day. The applicant has co;asunicated with and will ester into an agreement with the Reading Railroad to allow the applicant to restrict access to the railroad right-of-way with-j 5
in the exclusion area is the event of en emergency. I f .I The Schuylkill River, itself, is not navigable to large Loata { because of the'dass located downstream from the site. Use of esall pleasure craf t in varmer weather, does occur, but this ') activity is stated to be relatively minor la the region of the i site. !5 t !i The applicant will have emergency plans to assure control over .!) l all portfow of the exc1mion area including those portions of { I exclusion area. the Reading Asilroad and the Schuylkill River located within the Structures at or near the Raading RailrcaJ, e.g., i intake structures and pumphouse, which have safety functior.s will f be designed to withstand effects of an exp1nsion or udssiles resulting from any conceivable railroad accident occurring in the I i vicinity of the facility. We conclude that the applicant's l l! criteria for protection of the facility against the effects of accidents at the rai. road are acceptable. b i l l l .b 4
w. L 'y ks ~ 2.8 Conclusion _s We conclude that the exclusion distance and the low population sone distance are acceptable, that natural ph'enomena are being adequately considered, that the meteorological and environmental programs are acceptable, that the effects of nearby railroad traffic have been adequately considered, that nearby air traffic requires as special consideration, and that the site is accsptatic for the cocotruction of the Limerick Station. g 3.0 RD.gotDEs1CN 3.1 General Each nuclear steam supply system is a General Electric Company (CE) boillag-water reactor (BWR) which generates steam for direct I ~, i l lj use in the steam turbine-generator. The reactor core containing l4 1 l nuclear fusi elements and control ro a is supported in a domed, d cylindrical shroud inside the reac cor vessel. Steam separators 9 tsp are mounted on the, shroud done. Two external, motor driven pumps i c l inject high-velocity water into 20 jet pumps which are located in .l the annulus between the shroud-and the reactor vessel. The high velocity water from the jet nozzles entrains additional water from 1 the annular region. The combined liquid flow (about 3 times that l ,j l 1 i of the high-velocity water flow) becomes a steam vster mixture as l it passes through and cools the reactor core. The steam emerges I I _____d
.u.c kg i from the steam serorators and drysts (md eders ferr 26-inch-diameter pfpt.n let. ding to llc tuibine-generator. 4 f i Reactor. power is controlled either by movezant of control rodt j or by changing the speed of the two external recirculating-water f pugs. Reactor power operation is terminated (reactor shutdown) by inserting control' rods into the core. A streSy liquid control f synLsin in provided m e *vec*muy eymL = for auctus~ shutdowu sud operates by pumping a sodium postaborate solution into the reactor. De reactor design of the Linarick Generating 5.ation is similar l 1 to that of other facilities now under construction; a 114 tics of ,j
- j' eelected pertinent features for Limerick and Peach Botton Units ll 2/3 are shown in Table 3.1.
Se nucisar fuel', the reactor vessels, a and the thermal-hydraulic characteristics of these plants are R basically the same. Me few differences are discussed in the f I following stesections. 3.2 Nuclear Design _ I g he nuclear fuel and its arrangement in the Limerick core is essen-j N tially identical to previously approved p1sts. A comparison of j f. .f selected nuclear feel design parameters is included in Table 3.1.. he design of the fuel will provide " negative reactivity feedback" i 1. that is sufficient in combination with other plant systems, to prevent fuel damage caused by enormal operational transients. f n I i
q gfl).. bf ,y 4 ' (jf The nuclear fuel asse 61y contains a 7 x 7 erray of cylindrical 1 fb1 rods enclosed in a Zircalloy fuel channel. The assemblies 'I are grouped = four-assembly modules in the core. Within a module,' there are two average fuel assembly-enrichments (weight percent of U-235) for the initist core loading: 2.50 w/o in the ~ three higher enriched fuel a4xablies and 1.10 w/o in the low ~! enriched fuel msedly. The avere;;c fus.1 snrichrcats fcr the initial core are chosen to provide excess reactivity in the fuel assemblies sufficient to overcome the neutron losses from core neutron leakage, mderator heating, fuel temperature..'se, and equilibrium xenon and samarium absorptica of neutrons (poisoning); i l l also included in the fuel enrichments is an allowance for fuel l depletion. f Within the high enrichmnt fuel asembly, there are four individual l rod enrichments; within the lo. enrichment fuel assembly, there i i l sre two enrichments. The combination of rod enrichments is de-i 3 j signed to reduce the local power peaking (the local power peaking l factor is defined as the marinum to average rod power over the four I bundle array). I ( Selection of two discrete (average) fuel assenbly enrichments, i 2.5 w/o and 1.10 w/o, enhances the transition from the initial i [ fuel load cycle of about 1.85 years, at 19,000 megawatt-days per I I i l .f.
10 55
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, *l K: + ten (WD/T) uranium, to the near equilibrium core situation. 1he ^ planned refueling strategy at tne end of the first reactor cycle is to remove the low enriched fuel and replace it with reload fuel. The '.eplacement fuel is designed to attain 25,000 MWD /T average exposure. The former practice of storing the removed fuel assemblies for leter reinsertion will not be necessary since their low reactivity lifetime is designed for only one fuel cycle. 1 1 Orifices have been provided in each' fuel assembly support piece to distribute the flow of core coolant. There are two'orificed ficw sones in the core: an outer sene that la a narrow, reduced I power region around the periphery of the core, and an inner some consisting of the core center region. Because of a relationship that makes the fuel assembly flow and power interdependent, the ics enriched assemblies would ordinarily tend to have more flow than the high enriched f uel. In order to provide better flow utilization in both fuel assembly types, each low enriched fuel i i assembly has an additional orifice Integral with the fuel assen-I bly inlet nose piece. The combined effect of the orifices pro-vide a flow distribution that results in a critical heat flux ratio (see section 3.3) that is approximately equal in each flow j, sone at design conditions. 1 Reactivity control will be achieved by using movable cruciform-shaped control rods and a variable recirculation flow rate. The 4 4 h' A y IA Ii Li
I ^ . l' i { recirculation flow rate will sutentically accorr:nodate a demand - for increase or decrease in power (for specified control rod pat-l terns) to permit " electrical load following." During the initial 7 l core lifetime prior to the first refueling, additional reactivity L i. f-centrol will be obtained from the neutron absorbing characteristic of gadoliniusmide that will be mixed with the uranium oxide ia L' four rods of the high enrichment fuel assemblies. Gadolinis-I ur :1= fu:1 red: will al:e be used fer reactivity centrol in tha Decwns Ferry, The Quad-Cities, and the Peach Botton Units 2/3 t. [ facilities. This use of a gadolinia-uranium mixture for resetiv-ity control was reviewed and found suitable during the Quad-cities review. The addition of gadolinia to the fuel will mai it un-l ascessary to install the temporary fixed poison (neutron absorber) ) l control eurtains that were used la earlier designs to control I reactivity in the initial period of the core lifetime. L l Mechanical and electrical methods are available to limit the in-sertion of positive reactivity into the core. h oe methods I include control rod velocity limiters, specified control rod l 1 operating patterns, limited control rod drive speed capability, a control rod worth minimiser program and its control' rod blocks l' (to inhibit selection, withdrawal or insertion of out-of-sequence l roda during startup, shutdown, or low power operation), and con-trol rod drive housing supports. The rod worth minimiser or the j i { ai
1 23 - 4 ' 66 i 2 centrol rod patterns administratively permitted will limit rod reactivity wrths to less than 0.01 Ak/k wMn operoting below 10f, of rated reactet power. The applicant has fadicated that the General Electric Co.ga.ny will obtain faservice data en use of the feel. An analysis of this data vill provide the inservice operating limits for reactor operation. We comelode the inforastion now available provides a mitable basis to expect safe performance of the reactivity control B =acha=4=== and the nuclear fuel. We intend to follow the in, I 3.3 service performance of General Eleccric Company's nuclear fuel to evaluate the expected standards of performance. h rmal and Hydriulie Desian h thermal and hydraulic characteristics of the Limerick reactor core are identical to those for Browns F(a and Peach Botton f' Units 2 and 3. h same rated and design power levels are planned l h for nach of these high power density cores. Core cooling systems 'y>$$ for the tacilities are identical: two recirculation loops and j J jf, twnty jet pumps for each reactor. The core thermal ar.d hydraulic i design basis is to control the local power density within the core to levels that assure the fuel rods do not overheat during normal plant operation including operational transients. I WD804 eh*MM **e eWMM hNaMM ^_ 7: g, d4 gg aan de^~- m M'
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I-( j (.{ hk%i )j gs By definition, the condition for which severe overheating of the A j l/j fuel rod causing clad damage is assumed to occur when the fuel i
- l l
rod heat flux equals that for the caset of tha transitico from 1j i i i the nucleate boiling regime to film boiling on the fuel rod ,'i g i cladding surface. This heat flux is defined as the critical heat flux. The ratio of the critical heat flux to the maximus
- j calculated heat flux is defined as the minimum critical heat fluz ratio (HCHFR). A EHFR paater than unity assures that cooling g
of the fuel is maintained through heat transfer by a nucleate f bolling process. EHFR values equal to or greater than unity I provide adequate mergia between expected conditions and those required to cause fuel clad damage since the critical hast flux correlation presented in General tiectric Company's Topical ) Report APED-5286, is based on a limit line drawn below all of the available relevant experimental data points. The Regulatory a ic l Staff has reviewed and found acceptable the methoda used to k calculate the EHFR, the experimental basis for the calculation, i and its validity as a dxmage limit. { During the normal, steady state operation of the reactor, the I thermal-hydraulic design of the core will provide a HCBFR of 4
diu bh ag about 1.9. The applicant's analysis of a series of postulated {N abnormal operational transienteE esulted in HCHFR values all t d; greater than unity. Cther limits have been established to prevent fuel rod clad damage. Cladding strain because of thermal expansion or interns 1 fuel rod pressurization could contribute to fuel damage. A value-of II plastic strt.in is applied as the limit below which no clad damage occurs. A II plastic strain any occur whom the linear - B (fuel rod) best generation rate exceeds 28 kilowatts per foot (kW/ft) in unirradiated fuel and 22 kW/ft. in fuel with a local exposure of 40,000 MW/T. The normal steady-state operation of the plant is established to assure that the maximum linear heat j generation rate is maintained below 18.5 kW/ft. We conclude that.the thermal and hydraulic design, including the applicant's analysas of normal operation and anticipated transiosts, O provides adequate safety margia to protect assinst feel failure. 3.4 Reactor Internals 3.4.1 Desian The reactor internal structures are classified as Class 1 (seismic) i components and will be designed to withstand normal loads, antici-pated transients and the Operational Basis Earthquake within the M Abnormal operational transients result from single equipment failures or single operator errors that can be reasonably expected during any mode of plant operaticas. l
7 l d i ~ 26 ~ p:[3 l 3 ":t )i$N j stresa limit criteria of Articia 4, Section III cf the ASME Boiler and Pressure Vressel Code. Under the loads calculated to result froma the Design Basis l Accident, the Design Basis Eatthquake and the combination of these postulated ovests Limerick reactor interns 1 structures will be designed to the requ'iressats of the G. E. Nuclear Steen System (!!!?) feedfag criteria, Appendis C of ths PSUt. The RSS ) loading criteria as oristaally proposed cents.ined a nusier of design stress, deformation, fatigue and buckling limits which B exceeded applicable limits as specified in the nuclear component codes. By amendaant, the applicant has stated that cely these limits of Appendiz C which are consistent with tbs limits specified in the nuclear component codes will be employed; en this basis we find the design criteria for the Liasrick reactor internals acceptable. Dynamic systm Analysis for Seismie, Operatina, and thCA Imadinas 3.4.2 3 Seismic loading on the reactor internals will be determined by means of a normal mode-time history analysis. We find this procedure acceptable. Dynamic loading due to normal and upset operating conditions will be computed by means of quasi-dynamic methods based on the actually measured vibration response of similar reactor designs. We are ,? 1 ) presently reviesing a euz,ny ef-CJ2 vibration test histeries and correlations of vibration test data with de.ais predictions l submitted as Amendment 19 (Proprietary) to the Quad-Cities application and referenced in Supplement 5 of the Limerick PSA1. 4 This amendment presents procedures which may be acceptable for 1 design of the reactor internals for Limerick. To assist us in l o \\ determining the acceptability of these procedures, we have asked the General Electric Company to supply additional iaiornatiuu. Finni acceptance of the procedures proposed for determining the g E dynamic loading due to aosasi and upset operating conditions for t tF.a Limerick reactor will be contigent upon a satisfactory review I of the requested Laformation. The cosplation of this review is expected to be completed prior to the component fabrication for Limerick. l Design loadings for the postulated Loss-of Coolent Accident (LOCA) will be determined by computing the respouca of each structural master to the calculated peak pre e t*.M. '.cr.etial applied as an equivalent static load. In response to our concerns retarding 1~ the validity of this static analysis the applicant has stated (PSAR Supplement 5) that the natural frequency of the BWR internal l I structures is more than ten times the calculated frequency of the { l IDCA loads thus assuring no significant dynamic anlification. On the basis of the information submitted by the applicant we find this analytical method acceptable. -3 .+ .~.~ -. .x, se n..-.. ~.. ~... -. . -...~.. -
.I A. isj,, f'u) s 5; 3.4.3 Vibration Control Vibration monitoring of repetor vessel internals daring preopera-L ~ tional tests and during normal operatire has been completed on i 0 a few boiling water reactors (Snas). General Electric Company topical report " APED-5433. Vibration Analysis and Testing of Reactor Internals" April 1967 presents the results of one tast t' l program. We have concluded chat a confirmato.ytype vibration test should be performed ac the Limerick plant. The applicant g has stated that confirmatory testing or monitoring would be 1 acceeplished. l We will require documentation of the test program -l ta the Final Safety Analysis Report. 4 B j 1a
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sf . fp-kh 4.0 REACTOR COOLANT SY51124
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4.1 Ceneral The principal design parameters of the reactor coolant ' ystem are s shown in Table 4-1 for Limerick and Peach gotton Units 2/3. m l Table 4-1 i Limerick Fasch Rotton Units 2/3 I ' Design Vessel Temp., 'F Design Thermal Power, led 3440 3440 Design Yassel Freasure (psig) 1265 1265 575 575 Total Core Coolant i F1wpte (full power) x 10,1b/br 106.5 102.5 StaanFlowR:q(1b/hr full I power) x 10 14.25 14.033 l Normal Operating Pressure (in dese), poig 1020 1020 Feedwater Flow Rate, l z 10 6, Ib/br 14.25 13.30 Feedvater Temperature. 'F 420 376.1 Reactor Coolant Racircula- ~g tion Line Design Pressure, peig 1 Inlet leg 1250 1148 Outlet les 1500 1326 Design Temp., 'F 562 562 Recirculation Flow Bate, gym 45,200 45,200 I As may be observed in Table 4.1, the differences among the listed l parameters for the three facilities are few in number. Limerick has a higher feedwater flow rate and* temperature than that for t a ._.....,_....s,,,,__,,,...
'h : thyr Lityp - C"f.'j ' I'E Peach Bottom. h higher feedwater temperature at Limerick is - l t,f ' 8: li} cLtained by using an additional (high pressure) feedwater heeter. 5 -N higbar flow race of feedwater compensates for the higher ~ steam flow rate of Limerick. N re are no. safety implications i not previously discussed and resolved in this increase of feed-water tagerature. t 4.2 Reactor Coolant ?ressure Boundary OtCPB) t 4.2.1 Definition g As defined in the Commissico's regulations,10 F. Part 50, 11 l-paragraph 50.2, the " reactor coolant pressure boundary" means all those pressure-containing compaments of boiling and pressurized I j piping, pumps, and valves, which are: water-cooled nuclear power reactors, such as pressure vessels, i t '^ i I s. Part of the reactor coolant system, 1 b. Connected to the reactor coolant system, up to and including any and all of the following: l 1 (1) The outermost containment isolation valve in system i piping which penetrates primary reactor contataant. (2) The second of two valves normally closed during normal p' reactor operation in system piping which does not pene-trate primary reactor containment, j (3) The reactor coolant system safety and relief valves, 1 l } ~. --____-_______________-_______1
1.i ef:.: (4) The outcruost contairmat isolation valve in the main steam and feedwater piping for reactors of the direct E cycle bolling water type. I 4.2.2 Design The reactor coolant pressure boundary (RCPB) will be designed as ,M a Class 1 (seismic) system. Components of the RCPB will be de-signed to withstand normal loads, saticipated transients and the 4: Operational Basis Earthquake within the normal and upset stress V limits'of the applicable codes cited below. Application of AEC code classifications is discussed in section 4.3. The reactor pressure vessel vill be designed, fabricated, and I inspected to Class A requirements of Section III of the ASME Boiler and Pressure Yessel Code,1%8 Edition, including the Summer 1969 Addenda. P.G'S piping will be designed, fabricated, B and inspected to Class I requirements of the ANSI B31.7 Code for l Nuchar Powr Piping,1969 Edition, including all applicable code l interpretations. The recirculation system pump casings anc valves j which are part of the RCPB will be designed, fabricated, and inspected to Class I requirements of the ASME Code for Pumps and Valves for Nuclear Power dated November 1968. The reactor pressure vessel and RCPB piping will be designed to i the emergency and f aulted stress limits of the applicable code for H ij N1 } l d' the loads calculated to result from the Design Ba.wis Accident, the Denip Basis Earthquake and the canbination of these postulated The remaining RCFB components, pimps and valves,. for which eveats. .eurgency or faulted stress limits are not included in the ASME Code for Pumps and Valves, will be designed to the requirements of the General Electric Nuclear Steam System (NSS) toading Criteria, Apptadix C of the Preliminary Safety Analysis Report (PSAR). The. NSS loading critaria contain a number of design. stress, deformation, fatigue and buckling limits that may exceed applicable limits as opecified.in the suelcar component codes; consequently, for those components within the reactor coolant pressure boundary, the appli-cast will use only those limits which are consistent with the limits specified in the nuclear component codes. He find that the design, fabrication, and inspection criteria discussed above are censistent with those accepted for all recently reviewed plants of this type and form an acceptable basis for the design of the reactor coolant pressure boundary. 4.2.3 Pipe IAsip Criteria The applicant states that a system of restraints and supports for recirculation and other piping will be so designed that reaction forces (pipe whip) associated with a circunferential break or 1 split will not prevent: (1) scram, (2) isolation of either i m< -}
,1 e - L 1 f 'f.{ j reactor vessel' or primary containment, (3) adequcte core cooling, e. D and (4) maintenance of primary contai:wt integrity. 1*o find .c these criteria _seceptable. 4.2.4 Main St.aam Line Isolation Valve (MSLIV) Leskaae leakage through the closed main steam line (N8L) isolation valves following a postulated loss-of-coolant accident is an uncertainty in calculating potential IAX:A doses. The applicant has co.anitted to providing a closed MsLIV sealing system and has in progress a study to selact a MSL isolation valve essi system. Under present BR E consideration are two concepts: (1) provide a third isolstica valve in the MSL with either compressed air or water to seal the I* volume between closed isolation valves er (2) provide a water-i essi system which introduces water as the saalant upstream of the M8 LIV inside the primary conemin= ant structure. Un and the appli-canc will continue the review and resolution of this design feature during the construction period. 4.2.5 Seismic Desian of Main 8 team Line Pisina h main steam line (N8L) piping from the reacter vessel out to and including the outermost isolation valve will be designed, fabricated and inspected to requirements of Class I (seismic) and .n-(ABC) Quality Group A. h main steam lines from the outermost .m isolation valve up to but not including the turbine stop valve j l ~ f G-. }
y \\ j:1 are classed.as (AEC) Quality Croup B. ll 4 { l To provide adequate protec-tion cgsinst failure of this latter repent of M51 piping + , the ]. applicant will provide c.ain steam line Class I (seismic) des j f from the outermost isolation valve up to and including the t ? stop valve and all branch lines 2-1/2 inches. La diameter and 1 'l j 1stger, up to and including their first valve and their associa 3 {> restraints. This design will be accomplished by a dynamic seismic analysis to withstand the Oss one bas aoses vitata the 11mits I T' RR the ANSI B31.1 Class Il piping code and Appendix A. PSAR ?!w dynamic input for MSL design will be derived from a taa i, modal analysis (or equivalent method) of the supporting stru { #$ nas conclude that this additional measure of conservative is acceptable. I i ' i 4.2.6 Primary System Pressure Reliaf q. The objectives of the pressure relief system are (a) to limit NS kkt overpressure of the reactor coolant pressure boundary (RCFB) that spy might occur from abnormal operational transients, and (b) to p vide a method for rapid depressurization of the primary coolant system in the event of carrsin loss-of-coolant accidents. t In the latter application, automatic depressurization for sus 11 breaks of the primary system enables low pressure coolant injection \\ or core spray system (CSS) operation. This automatic depressuriza-b tion system is a backup to the high pressure coolant injection 'O system described in section 6.0. i;
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E g' m hh. ew l There are eleven relief valves and two safety valves in the pressure relief systan. The valves are moteted on the e"ein staulicas inside primary containwat. 0}eratica of the relief valves will discharge sesam to the wetwell egyressies pool and ' will perform these fumetions: (a) limit overprec?ure and prevent safety valve opentag, (b) augment safety valve ergability, and (c) depressurier the primary system followiss small breaks to .) i allow LPCI and/or CSS operatise. The alsven relief valves are self-actuating ta their overpressure safety mode but can be operated todirectly to permit remote assual or automatic opera-tism at lower pressures. i The two safety valves will discharge to the dryuell interier and i functies to prevent overpressurtaaties of the primary coolant system. The safety valve capacity is based se the pressure rise ~ resulting from the following postulated events: mais steam flow B stops after a turbias trip at operating conditions, turbine f adatasion pwssure of 980 psig (103 percast rated pensure), so l 4 steam bypass of turbine, and a reactor scram due to high pressure. The analysis indicates that a design safety valve capacity of 10 percent rated flow ta conjunction with a design relief valve capacity of 60 percent rated flow, is capable of keeping an adequate pressure margin below the peak ASME Code allevable pressure of 1375 pois. 1 l 4
i 3 \\ II. %t?{ GR ,ikk We conclude that the primary sptna pressure relief system, den e \\ wppleunted by the active of the reactor prots.ction spres, i l 'l B d provides adequate protectios against overpressurisation of the i resetor coolast system. 4.3 Ape 11 cation of AEC Cleeetficatius Crr;q The appliaast has applied a system of code slaasification grou to those pressure cantainiad ce.,.wac. vote er. p.st of th. )) l N reactor eeolant pressure boundary (defined in section 4.2.1) and i B ether fluid systems importsat to safety. These classificacias 1 j groups generally cerroepend to the ABC seda classification Croups I i B, C'and D developed by the ragstatory staff. The cadas and standards applicable to the canyonents sa asch of the classifice-I tion groups are identified is tela 4.3. 1 f tre and the applicant are in agressant on the application of the 'e ' W] code classifiestico groups for the reactor coolant pressure boundary and other fluid systems lamportant to safety. yor those systems, portions of systems, er components where the appiteant's p classification grouping differ from ears, the applicant has OM , # 2 +g upgraded his classifications by provisions for a quality level substantially equivalent to that aerns11y required by the appli-y g cable staff classification code groups. Clarification of thsee . upgraded classifications is documented in Amendment 6, Supplemen to the Preliminary Safety Analysis Report. - x
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D2M For pressurs-retaining cast parts of pumps and valves in systems . which are classified as Group A or Croup B, we f and the non-destructive awa='a= tion requirsasats in the AC4I Puay and Yalve Code for lines over two inches up to and tactading four inches landequate. To be acceptable, we require volametric saamination ) (radiography or ultrasonic testing) of thsas pressure-retaintag i! cast parts la 11 ass ever two inches up to and including four taches. there size er casfiguration does not permit effective volumetric examination, eurface examinatsoa (magastic particle, E er liquid penetreat testing) any be substituted. Ruasimation pr.w eedures and acceptases standards should conform with tboes spect-find in the ARIB Code for pumps and Yalves. The applicant will apply theas requiremsats for ace-destructive examination of cast parts of pumps and valves in Group A er Group E afstems. 1 l lie cor. elude that the systes quality group classification as specified by the' applicant and supplemented by provistoca for i upgrading quality levels, and the additional nondestructive assaination requirement discussed above are acceptable for this I facility. i-l 4.4 _ Reactor Vessel Material surveillance Proarsa The applicant's proposed reactor vessel material surveillance program, described in Supplement 2 to the PSAR, is consistent with progress that have been acccpted on similar BWR plaats. j' l5 L f I ~ a -.- - n.
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j 33 i 4 h3 The program in acceptable with respect to the total nunber. of I p capsules, number of capsules to be withdtwa am y. r tasted, material '] chemistry documentation, and retention of archive material. 4.5 Fracture Touahness criteria-r + 4 Im *-P : t 2 (and Supplement 2) to the Preliminary Safety E { j Analysta Report, the applicant respooded' to our request for' addi-sienst inieraatios. en fracsure toushmou and twL require ate ~ = that have been specified for the ferritic meterials withis the reactor coolant pressure boundary. b applicset has indicated provisions for fracture control will be in accordcace with the { requirements of the ASME Boiler and Pressure Vessel Code, Sectiot 111. h current code rules permit a vessel to be pressurized i l enly above a temperature aquel to the Ni1 Ductility Transition 7 (NDT) temperature plus 60*F. M MDT tamparatura, according to g paragraph 5-311 of the Code, can be obtained by either the drop-9M weight test (DWr) or the Charpy V-sotch (Cv) impact test. s -l ][. la a meeting with the applicant we described the proposed ABC 4 y - .e fracture toughness criteria (recently published as proposed ?' d: Appendix C to 10 CFA Part 30) and advised the applicant that I adequate fracture toughness data vill be required to establish "I ~ l :; appropriate heatup and ceoldown limits for this plant at the m time of the operating license review. e h. r ~~~
M d 4 Aw h applicant has stated that it is possible for many of the Ii ferritic piping systems taside the dryaell to be at a tencrature as low as 40*F vben the syntaa pressure is in excess of 25 percent of the normal reacter operettag pressure. Tbs applicant has pro-posed to assure adequate fracture toughness for, these systems asistaining the lausst service estal temperature at laest 40'F aheve the WT temperature as asasored by tho' charry T-ootch I' specimens. In view of the possible uncertainties of Chattyy { Veotch speciosas to predict correctly the lef temperature, we have advised the applicant that the NFf toeperature abould ** verified by othat tests, seek as the Mrf test. ja He will require applicant decennatation ta the Final Safety Analysis Report of any deviatian from the (pkoposed) criteria la Appendia 4 10 CPR Part 50. 4.6 laservice Inasectica .{ /;j'y the applicent will provide access to the reactor coolant pressure gu l boundary la compliance with Section 11. " Rules for Inservice cyr ,i < } Inspection of Ranctor Coolant Systems." ASME Boiler and Pressure l Vessel Code. For engineered safety systems beyond the limits of
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the reactor coolant pressure boundary that are met presently covered in Section 11 of the A5ME Code, the applicant is providing i i access to pipe welds, and pump and valve bodies in these systems. i i-. I R__m,,m__._.- d I 3; i
'a n 10 Ey We conclude that the applicant is maklag adequate access provistens a for the inservice. inspectica of the Liuerick Statium. 1 4.7 Reactor Coolant Systen Sensitised Statalass Steel the applicant has indicated that me furance-eensitised stataloss steel components will be used within the pressure containing boundary i or load-beartag maahers vital to the structural integrity of the reactor vessel ens cors. no furnace-seseitised staininas si 1 B. components are to be used la piping. Cast meterial with less than n F 3 $1 sinism ferrite will be used for hard-surfaced austamitic stala- { f ees : teel discs, east rians, valve bodies with integral seats, and pano hsarina :.nd sosi composeats ukich any be subject to tempera-i tures that weald uitise wrought asterial. The applicant will I met use any stataloss steel for ukich deliberate ad,41 tion of mitrogee has been'allound to enhance tha straagth of the stccl. 3 j 0 1 l j Additional precautions used will include rapid coaldcue from solo- 'N tion heat treatnest temperature and the use of low carben stainlass steel electrodas during umiding. Ferritte measles and pipe ends 4 will be buttered with lor carbon stataloss steel applied by the veld deposition techniqueg this procedure a11minates use of separate l wrought austenitic stataloss steel sc!e-ends which could become furnace-sensitised. He eenelude th6t planning to avoid sensitisa-tion of sustenitic stainless steel during the fabrication period is f acceptable. l I j l
i 41 - %p <w. .l'k, 4.8 Jlectrowlsa P41 ding N-i;fl Q3 tW Tha eIcetroslag welding process will not be used within the resetor coolant pressure bourAary. The applicant indicated that if their i . m ) plans should be changed, the required data (process specifications, i j j variablas, quality control procedures, beat affected mone, and .., y i weld data) would be provided for evaluation. g; 4.9 Foreira Procurenest To date ao foreign manufacturer has been eagesed to design or fabricate any component within the reactor coolant pressure boundary. 4.10 Leak Detection Detection of leeks from the reactor coelant pressure boundary within the dryven will be possible using airborne particulate sampling equipment stallar to that new esed at Dresde:: 2. ftchie i games and iodine detectica equipment will capplement the partico-hg late detection capability. Other equipannt la the leak detectica system will permit the measurement of drywell pressure, tempera-I w h ture, and sump liquid level. The combination of these measure-meets win maahle the operator to reliably detect abnormal laakage ,j in the drywen. Sensitivity for laak detectica on the order of a bj-few cc/ min win result from the use of both continuous and periodic i 1 l L aampling of radioactive airborne particulate. Routine use of the I ( ,f . ~. .n~---
c !. } equipment will provide an operational tool for maintaining con-trollable leaks such as from valve stem packings. Determination of the exact location 'of a leak will require detailed it.spection of the containment internals. Leakage detection for vital fluid-carrying systems external to the primary containment has been adequately provided for. Tempera-ture, pressure, and flow sensoro with associated instrumentation and alarms are to be provided beyond the limits of the reactor coolant pressure boundary for the following systems (measured parameters in parentheses): i a. Main steam lines (high flow rate; low pressure at turbine inlet). b. Mais steam lines compartment (high radiation and temperature in f vicinity of MSL). l Reactor core isolation cooling system (RCIC equipment room c. i temperature; RCIC turbine high steam flow, steam line low pressure, turbine exhaust high pressure). d. Righ pressure coolant injection system (HPCI equipment room high temperature, turbine high steam flow, low steam flow, turbine exhaust high pressure). a. Reactor water cleanup system (high temperature in equipment 1 rooms, cleanup system high flow). 1 -[ m
l In addition to the ' systems mentioned, the Reactor Building ventila- ) l tion exhaust is monitored for high radiation which would signal a breach of the nuclear system process barrier. Airborne radiation l sampling and monitoring is provided in the compartments housing the Residual Heat Removal (RER) System and the Reactor "Jater Cleanup System. Finally, the 3CIC and HPCI turbines and core spray pump areas will be sampled for airborne particulate activity. We conclude that the leakage detection systems are acceptable for the Limerick Generating Station. 4.11 Corrosion Crackins of Pipe Metal Potential for corrosion cracking of certain pipe metals is incrassed wherever lines contain oxygen or other gases 'in a stag-l nant volume. To avoid this condition, the applicant will keep the oxygen content in feedwater to the reactor below 200 parts per billion (ppb) by deseration la the main steam condenser. The applicant's design personnel will perform a continuous review of piping layout, elevation, and plan drawings to eliminate high points where gas could collect. In appropriate piping runs, piping will be sloped towards the reactor to eliminate pockets of trapped fluid and/or gas. Remaining aress of concern such as thermal sleeves and sparger connection boxes will be designed I with a gas escape passage. Aufficient width and root radius will i be Frovided to avoid gas collection where crevices cannot be I'! avoided. We find this approach acceptable. l. I .,,.a-m.. 1 I
1 44 - l 5.0 CONTAINMENT AND STRUCTURAL DESIGN 5.1 Containment Design and Comparison The containment systems include the primar'y containment using t.e S pressure suppression concept and the secondary containment which includes the reactor building, its recirculating (atmospheric ventilation) system, and the standby gas treatment system (SBGTS). i The drywell is a frustum of a cone with steel-lined reinforced concrete walls that are six feet thick. 'Ihe vapor suppression chamber is a steel-lined right circular cylinder of six feet thick reinforced concrete located directly beneeth the drywell. The drywell and wetwell are separated by a 3-1/2 fc thick reinforced concrete floor penetrated by eighty-five vent pipes. A low-I leakage Reactor Building surrounds the primary containment to serve as a secondary barrier. A comparison of the containment I design parameters for '.imerick Generating Station and Paaah Sotton Units 2/3 is presented in Table 5.1. Both primary and l secondary containments will meet, among others, the criteria for Class I seismic design. 5.1.1 Prim ry Containment The vapor suppression concept for the reduction of pressure inside the primary containment following a loss-of-coolant accident (LOCA) has been used in the Limerick design as in other boiling wa ter s 0 .g R.
e i p a d i h 3 s )) s / e ss 2 b p uu d l a rr n s u h c c a t b c t t e i8 n (( 5 t r n7 ot s U2 ih i 621 0000 0 t a 0 / sg r 5 8 4000 e t 0 x r c 1 0, 0, 0, n7 si o 2 c u o7 e. t t2 995 n r t - p, 513 o t g o0 pl l 111 c s n B 5_ ue e r i ss s d e k h s s e p c c ee e c u e rv v l r s d o u g e o P sl l n s f l l se e i s n e e ee e ) ,d ee i e e rt t hh ud v e t t S PS S t tli R S S R ppcse E eennr T ddii u E s d M rrt3 s n d A e eeot e a n R t t t nf r a A e aa p s P r )) ww,0 ) l s c ll t0 rt l l N l n ll ll s4 or a l 3 G e e ee aae7t o w a I ne c ww mmr cp w. S ot yt rr eap e f E is re ooth eu t ko 6 D 3 s er dw nnat rs e co (( r ir 6 5 1 k4 s ,t e (( c h T c3 eeed 5 N i/ rnrn 5500 7000 0 n tk 0 a E r2 pocid 5 42 4050 o c 5 c i e3 pcnl e 32 7, 7, 4, '3 h 4 E 4 L D m3 u oyn B D i - sf cci 452 d t A A L0 o l 352 e d T T 5 e ,l 211 c e' N rmdal r t2 O uuece o a/ C st nie f am1 ssitt nfi - F eul rs i ot1 O rr e eos PF V R rE NO S I RAP ) M F O n C ))( w gd o ii. d ) ssp w y em (( e l ) d ru ppm o a eu t b)) / f c ee t l n n a rrn ti t tf o o o % v uug ssff( v i i ssi o p(( t t eo ) tg3 sss p( e e c c a t eee eem e u u rrd memmu r r r r f o e"4 n ( t pp uruul f t t n m mull o n s e nnu i soov I : u n i g/ e m ggm svv ( go o t a1 m c n iii ae r nc c ck u i ssx mreee e i uat l a eea peet t dl l rea o t dd m d rra a l e e tl V n elff w r i v v sny o lll tl ue e nia r ro C l l aaa aelll e Bl l o d i eN l l nnn l wlll g cn/ A t y e e rrr uyeee a rr r gl et rew w eee crwww k oe e fi o e t p esv e cm p a me a py t t t t ld yt t t oe r ak myr e nxn a ree A I a1 U RD T rc iTD W I EI C DWW ao r e PD P R t } i i. reactor (BWR) facilities. There is a departure from the tradi-tional "lightbulb-torus" structural design in that the drywell and wetwell configuration is of the "over-unde." type, i.e., the drywell is constructed above the wetwell and together eney form a continuous, single structure for the primary containment. The codes utilized in design of the primary containment are ACI 316-63, "Bu11 ding Code Requirements for Reinforced Concrete" and the ASME Boiler and Pressurs Vessel Code, Section III, Subsection B. The latt.cr code governs design of the drywell head, locks, penetrations, and other steel structures having pressure vessel functions. Eere appropriate, the'Istter code also applies to the steel liner; how-ever, the 1/4-inch thick steel liner plate is designed to function only as a leak tight membrane. Containment liner material is A-285 steel, Grade A, Firebox Quality, fot 1/4 inch plate and ASTM A-516, Grade 60, for sections of thickness greater than 1/4 inch. C,mnecting the drywell and wetwell are eighty-five straight pipe vents, 45-1/2 feet long and two feet in diameter. The vents pro-ject about eighteen inthes above the steel-lined, reinforced-con: rete drywell fixt and extend into the wetwell so as to be submerged by ebout eleven feet of wetwell water. Each vent has a i protective cap mounted about ten inches above its opening in the drywell to keep foreign objects from entering the vent and also to e improve the fluid flow properties of postulated accident atmosphere. l ) l l i g ) O i
4 i I i i i \\ BLANK PAGE O h t 4 I e I d
.f The vents are structurally tied together by beams to provide resistance against forces which would be developed during the poatulated 7 0CA. Installed on twelve of the vents c.re vacuum relief valves which ' re monitored in the control room for their a position (open or normally closed). These relief valves (nine f inches diameter) are check valves designed in accordance with i the ASME Boiler and Pressure Vessel Code, Section III B, and I are remotely testable. They function to relieve wetwell atmospheric pressure should it exceed the drywell atmospheric pressure more than 3 psi. I 1 A removable steel pressure head car
- he top of the truncated j
cone of the drywell. The attachment for the drywell head is I i f anchored in the drywell shell concrete and is welded to the e drywell steel liner for leaktight integrity. Leak test. of the drywell head seal gasket is possible. a The Regulatory Staff has concluded that criteria for design of the primary containment are acceptable. Particular consideration was given to the items outlined below. a. Post-LOCA Containment Pressure: An extensive review of the methods for calculating the peak drywell containment pressure following a postulated loss-of-coolant accident and for e e. u we o 6 e ene a e
' determining the margin to be applied for the design pressure has been made. A discussion of this review is presented in section 9.2, Loss of Coolant Accident. Using a refined vent flow model, the General Electric Company has recalculated the post-LOCA peak drywell pressure to be 47.6 psig. The applicaut has accepted our suggested margin of 15% to give a design pressure of 5" psig. The newly calculated peak drywell floor differential pressure is 23.3 psig. The new design value of 30 pois for the deck differential pressure results from the use of a 30% margin which we suggested 'to the applicant. We conclude that the esiculated containment pressures and design margins ure acceptable. b. Drywell Deck Design: The drywell dsck will be surfaced with a steel liner plate that will provide added protection against vant bypass. Bypassing of the vents during a LOCA results in increase of the. peak drywell pressure 4 psi for each square foot of bypass area. The applicant has indicated willingness to test for leakage through the drywell floor. The method of testing (whether at low or high drywell pressure for qualit-ative or quantitative results) will be resolved and documented as a technical specification at the operating license stage of the review. The steel liner on the drywell deck will have the 1 l a me-A r
e
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sami quality as that on the containment walls. The structural details of the floor to wall joints and the method of floor to wall liner details are acceptable. c. Turbine Missiles: In the event of a turbine disc breakup, ' missile protection for the primary containment, control room, and the fuel pool is provided by structural barriers (rein- { forced concrete walls and roofs) and geometric arrangement I l j of the equipment, components, and operating areas. i d. Hydronen Control. Post-LOCA: The applicant has accepted the l AEC Safety Guide criteria for designing a system to control i e and kaep the concentration of hydrogen below the lower flam-mable limit in the primary containment following a LOCA. See section 6.2 for further discussion. e. Inertina: The applicant has indicated that the primary containment will be inerted with nitrogen gas during plant operation. See section 6.2 for further discussion. f. Seismic Instrumentation: The applicant has indicated that strong motion seismographs and accelerometers will be installed on the primary containment. We have concluded that the appli-cant has an acceptance plan.
g. Reactor Shield: The reactor chield must have the capability to withstand the internal pressure and coincident jet impinge-ment loads resulting from failures of high pressure lines in the annular region between the reactor vessel and outside diameter of the shield. The design provides structurally rigid shield wall "blockouts" or " plugs" in the piping penetra-tions to prevent movement of a broken recirculation or feed-unter' pipe such that the effective area for water flow to the . shield wall-reactor vessel annulus is reduced. The applicant has shown that the shield wall annulus, penetrations, and 'blockouts are designed to prevent significant direct impinge-ment of jet forces on the wall following a nozzle to safe-end weld failure on a (25.67 inch ID) recirculation line outlet. Since a failure of the nozzle to safe-end welds on smaller lines might result in significant jet impingement, an analysis of the worst case, the 12-inch feedwater line break, was made; the jet force and static pressure combined gave a resultant prensure (60 paid) that is within the design pressure. For added conservatism, the reactor shield will be designed to take the internal pressure and jet forces of loads equiva-lent to those associated with a Design Basis Accident break area of 4.82 sq. f t. We conclude that the design of the reactor shield is acceptable. M .... 3
i 9 r : h. Primary Containment Penetrations: Penetrations of the primary containment are in accordance with current design criterin. The applicant has designed an instrument line isolation system in keeping with the intent of our Safety Guide No.11. Inside the containment a 1/4 inch orifice will be installed in instru-ment lines as close to the instrument repoff as possible. Outside, and near the containment wall, a combination manual bypass valve and improved automatically reopening excess flow - J check valve vill be installed. Position indication of the latter on one of two local panols and a remota alarm in the control room will si nal the operator of valve closure. 3 1. Main Steam Isolatior. Valves (MSLIV): Isolation valves for the f f main steamlines are presently two per line, one inside and one outside the primary containment. The applicant has committed to the concept of sealing age!nst MSLIV leakage during a post-LOCA situation. The applicant is investigating the acceptability i of uring a water sealed system or a third isolation valve outside the primary containment with a cenability for pressurizing the inner volumeL bet w n the valves. The layaut of piping has l been made with sufficient space for a third isolation valve. See section 4.2.4 for more details. j. Containment Vacuum or External Overpressure: The applicant has eliminated external vacoum relief valv'es for the primary I a
t l containment because study showed that the marianum cooldown I rate of the post-LOCA drywell atmosphere by containment spray produced no more than 2.5 psi pressure differential. The applicant will provide a design overpressure capability of 5 psig. 5.1.2 Secondary Containment The secondary containment structure or Reactor Building will be designed to Class I (seismic) standards and built to limit the release of airborne radioactive materials. The Reacter Building will provide for controlled release of building atmosphere so that offsite doses from the postulated design basis. accidents will be well below the 10 CFR Part 100 guideline values. The Reactor Building vill enclose both reactor units; however, each reactor unit with its own primary containment is physically separated from the other by a dividing wall. Basically, there will be three compartments: two enclosing the primary containment of each of I two reactor units and one for the volume above the level of the refueling floor. The Reactor Building, including the dividing wall below the level of the refueling floor, is constructed with reinforced concrete. With reinforced walls and roofs, 3 feet and 1.5 feet thick respectively, the structure will have greater pro-tective capability than the standard BWR secondary confinement design that uses structural steel members and steel panels above
1 l ' ~ the. level of the refueling floor. The Reactor Building will be designed to withstand internal pressures of up to 7 inches of water, forces from tornadoes or earthquakes, blast effects from. a railroad accident, and missiles from tornadoes or rotating equipment to the extent that safety feature equipment inside l would resnain operational. I The original design objective was to provide a leak-tightness that i I i would limit in-leakage to 100% of the building volume per day at j i 1/4-inch water (vacuum) while operating the Standby Gas n eatment l System. The applicant has further increased the capability of the I Reactor Building by changing the design objective in-leakage to 50% of the building voluane per day. l We conclude that the applicant's criteria for design of the secondary containment are acceptable. The following sections present information on ancillary features and equipment which contribute to the< 1eakage control capability of the Reactor Building. 5.1.2.1 _ Reactor Building Recirculation System A Class I seismic design Reactor Building Recirculation System will be provided to assure reduction in the activity of the contained atmosphere in the event of the postulated loss-of-coolant accident or the refueling accident. Within three to five seconds after
receipt of an appropriate signal (Iow reactor water level, dryvell high pressure, high radiation in the exhaust ventilation duct, or manual actuation) the Reactor Building is isolated from the out-side atmosphere. The Recirculation System and the Standby Cas Treatment System (SBCTS) both a' tomatiesily start on isolation of u the Reactor Building. The recirculation of Reactor Building air, following an isolation signal, utilizes the normal ventilation system ductwork which has j been sealed to prevent outleakage from the building. The arrange-ment of ventilation ductwork and isolation. valves will permit separate control (recirculation, cleanup, and discharge) of air from each of the three main compartments of the Reactor Building. The normal ventilation fans are shut down, isolation valves close within three to five seconds, and one of the two recirculation fans start. The fan suction draws air from the volumes above and below the refueling floor through the recirculation system filter assembly. The flow rate in the main duct to the recirculation fan is about 60,000 CFM. A small fraction, about 1000 CFM, of filtered air is exhausted downstream from the recirculation system filter assembly to the SBCTS in order to maintain a slight negative pressure (1/4" water vacuum) in the reactor building. The remaining flow of filtered air (about 59,000 CFM) is recirculated within the Reactor Building. The effect of the effluent radioactivity is dis-cussed in section 9.0 of this raport.
f i - 54 1 The Reactor Building recirculation equipment includes redundant full-capacity fans and filter assemblies within each reactor unit compa rtment. The filters are full-flow, and include pre-filters, REPA filters, and charcoal filters. Tests will be performed to demonstrate applicant stated removal efficiencies of 99.9% for inorganic todines, 85% for organic iodines, and 99.97% for 0.3 micron particulate. Our evaluation of the Reactor Building j recirculation filters used halogen removal efficiencies of 90% \\ l for elemental todine and 70% for organic fodines. In our eval- }; untion, we further assumed that 1% of the iodines cannot be i y. removed. The applicant has calculated the maxiaum temperatures f attained on the filcers during a IDCA and found that the system i l f1w rates will adequately cool the filters; however, the charcoal bed filters will have water sprays to prevent overheating should air flow be lost when a large radioactive load is present on the filters. Mixing of the Reactor Building atmosphere within each of the throe main compartments following isolation is assured since the Recir-culation System inteke and return louvers are located at many locations. Fresh air is admitted during isolation by in-leakage to the building; this flow is about 50% of the Reactor Building volume per day. Mixing in the ductwork assures that the S8GTS extracts only a as.all portion of any activity in the system; such a.
t i mixing is accomplished by the suction ductwork design. We intend to follow this system closely to verify operation in conformity with design. I 5.1.2.2. Standby Gas Treatment System The Standby Gas Treatment System (SBGTS) provides controlled .l filtration and discharge of the secondary confinement air via the reactor building vent. This Class 1 seismic system uses EPA. and { deep-bed charcoal filters to assure maximum removal of todines and particulate. The charcoal bed depth is about 15 inches i (usual filter depth is about two inches). Efficiencies for re-moval will be greater than 99.9% for inorganic iodines, 90% for orgat.ic iodines, and 99.97% for 0.3 micron particulate, accord-l' ing to the applicant. For our accident calculations, we assume l I halogen removal efficiencies of only 90% for both elemental and i i organic iodines. We have further assumed that 1% of the iodines cannot be removed. The applicant's calculations show that the maximum temperatures attained by the charcoal filters during a LOCA are well below the ignition temperature of 640'F. Water I sprays are available for the charcoal filters to prevent over-heating if the cooling air flow should be totally lost while a f large radioactive iodine inventory is present on the filters. } ! i 6 L I l
1 I I ~ 5.2 Structural Design I s The applicant has provided defin!tions and identification of the structures, equipment, and systems that will be designed to Class I { or Class II: seismic criteria. Our consultant on seismic design was { j Machan M. Newmark Consulting Engineering Services whose rmort is 1 attached as.'.ppendix H. )[ 5.2.1 Class I (Seismic) Structures
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\\ i a structures and equipment whose failure could cause significant l . release of radioactivity or which are vital to a safe shutdown of 1 ( the station and the removal of decay and sensible heat are desig-f nated as Class I for the purposes of seismic design. Structures and equipment that may be essential to the operation of the aestion, but outside the definition of Class I structures and equipment are designated Class II. We have reviewed the applicant's listing to determine that the structures, equipment, and components are in the appropriate classification; we conclude that they are classified correctly. A discussion of the seismic design of Class I structures follows. For Class I (seismic) concrete structures, the factored load approach is used for design purposes. For Class I (seismic) steel structures, working stresses are used for normal operating loading combinations; however, under a combination of normal loads, design accident, and extreme environmer.tal conditions, increased stress 4: I i Y t
57 - 1 limits of 0.9Fy for bending, 0.85Fy for axial tension and 0.5Fy for shear are used (Fy is the yield stress). Thesa design approaches are acceptable. The applicant's design of Class II (seismic) structures will be such that failure of these structures will not degrade the integrity of any Class I (seismic) structure. According to the applicant, some less essential portions of Class I structures may be Class II if not related to loss of function and their failure would not render the class I structure inoperative. We find this criterion acceptable; we will review the specific portions involved to determine whether the criterion is fulfilled. The reactor vessel support is a typicci steel-skirt to steel-ring-girder to concrete pedestal design. This design of. t.he support is similar to previously reviewed supports and is acceptable. 5.2.2 Environmental Effects All Class I seismic structures, systems, equipment needed for safe shutdown, the primary containment, and essential heat removal systems will be protected from tornado effects. The design basis tornado with winds of 300 mph rotational velocity and 60 mph 1 transnational velocity and a pressure drop of 3 psi within 3 esconds will be used as design criterra. Tornado missiles assumed for design are similar to those previouely accepted. As indicated 1 -_w
l j ' t earlier, the Reactor Building will be built to withstand forces of earthquakes or tornadoes, including associated missiles to the extent that safety feature equipment will remain operational. We find these criteria to be f.cceptable. 5.2.3 containment Structural Desian Analysis To carry tangential shears due to earthquake, tangential, diagonal reinforcing will be placed in the shell of the ' primary containment. The two basic codes used in the containment design are the ACI 318-63 code and the ASME Boiler and Pressure Vessel Code, Section III, I I i Subsection B. The basic analytical method used for axisymmetric loada is,the finite element method. Nonaxisymmetric loads are ar. '.yzed using a computer program developed by S. Gosh and E. L. Ii Wilson.* Principal stresses and strains are calculated from the program, and then converted to rebar stresses using ACI 505. j As stated earlier in Section 5.1.1, the ASME code is used to establish a basis for designing the steel containment liner, wherever applicable, and also to design the drywell steel head., { l 0j The primary containment design analysio is acceptable. The secon-0{ dary containment will be designed by the same methods and to the same criteria as other Class I structures mentioned earlier in this report.
- mosh, S. and Wilson, E., Dynamic Stress Analysis of Axisymmetric l
Structures Under Arbitrary Leading, University of California, Berkeley, Earthquake Eng'.neering Research Center, Report EERC 69-10, September 1969. I l
- 5. 2. 4.
Tes ting and Surveillance During construction, rebar user tests will be performed on full section bars. It its not expected that are welding of rebar will i be necessary,' however if it does become necessary to arc weld, it will be done in accordance with the American Welding Society I i code AWS D12.1 requirements. Cadweld splice sampling for tensile tests is acceptable. Containment liner seams inaccessible after construction will be provided with a leak-chase system. A minimum of 4% will be radiographer, or a minimum of 10% will be tested by magnetic particle inspection where radiography is not possible. Initial structural integrity tests of the primary containment will be conducted at 115% of the design conditions. A full design pressure test of the primary containment can be performed at any time during plant life when not actually in operation. The reactor building (secondary containment) leak rate will be tested by isolating the building, operating the S5GTS, and missur-ing to assure an in-leakage rate no greater than 50% of the building free volume per day. ~ We conclude the capability for test of the primary and secondary containments as well as associated penetrations are acceptable. i J 6']
i I', t The details of the drywell deck test and the containment long-term surveillance r<squirements will be established during our review of the Final Safety Analysis Report and Technical Specifications.- 5.3 Seismic Design ,f We and our consultants have reviewed the seismic design criteria I and the results of analyses presented by the applicant for founda-t f tions, seismic response, damping f actors, and seismic analysis of i j structures, piping, reactor internals, equipment, and critical ,I items of control and instrumentation. Our consultants indicate l that the design and analysis procedures are in general in accord 6 l with the state of the art and incorporate an acceptable range of safety margin for the hazards considered. A copy of our consul-tant's report is attached as Appendix H. All Class I systems, components, and equipment outside of the reactor coolant pressure boundary will be designed to sustain normal loads, anticipated transients, and the Operational Basis Earthquake within the appropriate code allowable stress limits, I j and the Design Basis Earthquake within stress limits which are j f comparable to those associated with this emergency operating condition category (within the yield strength of the material for membrane stresses). nie consider that these stress criteria pro-vide an adequai:e margin of safety for Class I systems and compo-nents which may.be subjected to seismic loadings. .._._._.A
e i 5.3.1 Seismic Input The seismic design response spectra submitted by the applicant produce, with a 2% damping, an amplification factor of 3.5 within the vibration period range of 0.15 to 0.5 seconds. Proposed struc-ture and equipment damping factors are in accordance with those recommended by our consultant. The response spectra derived from the time histories to be used for design are adjusted in amplitude and frequency to envelope the response spectra specified for the site. We conclude that the seismic input criteria proposed by the applicant provide an eccer table basis for seisaic design. 5.3.2 Seismic System Dynamic Analyses Modal response spectrum, multi-degree-of-freedom, and normal mode-time history methods are used for all Category I structures, systems, and components. Governing response parameters are combined by the method of taking the square root of the sum of their squares to obtain the modal maximums when the modal response spectrum method is used. The absolute sum of responses is used for closely spaced frequencies. Floor response spectra inputs to be used for design and test verification of structure, systems, and components are generated from the normal mode-time history method. A vertical seismic-system dynamic analysis is being employed in lieu of a l constant vertical load factor for all structures, systems, and I components with natural frequencies greater than 30 Hz. We and I a
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our seismic consultants conclude that the seismic-system dynamic methods and procedures proposed by the applicant provide an acceptable basis for the seismic design. 6.0 ENGINEERED SAFETY FEATURES 6.1 Emergency Core Cooling Systems (ECCS) 6.1.1 Cene_rel In this section the emergency core cooling capability is discussed. / General features of the total design concept as well as specific fune:tions of individual systems are described. The phenomena associated with the loss-of-coolant accident and ECCS operation n] are presented for backgruend preliminary to a discussion of our application of acceptance criteria to the ECCS performance. The ECCS is designed to provide emergency core cooling in the event of postulated accidents involving mechanical failures in the primary pipir.g system that result in partial or complete loss of coolant from the reactor vessel at a rate greater than the available coolant makeup capacity using the normal operating equipment. The emergency core cooling systems 1/ consist of High Pressure Ceolant Injection System (HPCIS), the Automatic Depressurization System (ADS), the Core Spray System (CSS), and the low Pressure Coolant Injection System (LPCIS) which is one 1,The applicant uses " core standby cooling systems" for reference to ECCS; we shall use the latter designation. 4 b
~ - \\ mode of operation of the Residual Heat Removal System (RHRS). De HPCIS and ADS are high pressure systems while the CSS and LPCIS operate et relatively low pressures. De various systems are initiated by a high drywell pressure signal or a low reactor vessel water level signal except for the ADS which requires coincidence of the two signals and availability of low pressure I core coo'Ing equipment. Each of these systems is designed to cover a specific range of accident conditions. Collectively, the systems provide a redundsney in protection for the spectrtas of break sizes. Table 6.1 presents a comparison of the designed ECCS capabilities f or Limerick and Peach Botton Units 2 4. 3. Sufficient redundancy and reliability have been provided for sensors and associated controls and instrumentation of the ECCS such that no failure of a single initiating sensor or control device will prevent starting any one of the cooling system. He controls and instrumentation can be tested *and calibrated ( conditions representative of an accident. De ECCS equipment and piping will meet the requirements of a C1sas I seismic design. A loss of offsite power will not prevent ECCS operation. Provisions i have been made to provids capability for testing the sequential operability and functional performance of each system. he ECCS pumps which will provide cooling water will be designed and installed to eliminate dependence on primary containment l[ t t E'
- 63a - TABLE 6.1 .COWARISON OF ECCS CAPABILITIES Limerick Peach Bottom Units Q f' HPCI No. of pumps 1 l
- Capacity, 5,000 (100%)2/
5,000 Head, poidgm 1,120 to 150 1,120 to 150 Backup ADS + CS and sante LPCI LPCI No. of Pumps 4 4 Capacity (sa.) gpu 10,000 (22.3 cfs)37 10,000 % of Required Total, ( 33-1/3 33-1/3 Head, paidg.) 20 20 Backup CS CS Co_rg_ Spray (CS) No. of Loops 2 2 No. d Pumps 4 4 1 Capacity (ea.) spa 3,125 3,125 l 2 of Required Total, (ea.) 50 50 1 Head, poid 122 122 Backup LPCI LPCI Automatic Depressurisation System (ADS) 1 J No. of Relief Valves 5 5 % of Required Total (ea.) 25 25 I Capacity (pp.) lb/hr 800,000 800,000 Head, psidA 1,100 1.100 Backup Remote - Manual same l Relief Valves l M paid a posada per square inch differential between reactor vessel and primary containment. O Normal run of HPCI line moved from reactor vessel feedwater line to core i spray line to provide additional margin for cooling cladding early in the j heatup transient associated with LOCA. O Notaal run of,LPCI lines moved from the recirculation lines to four separate j reactor vessel penetrations for 61scharge inside the core shroud to provide [ additional margin for cooling cladding durinE IDCA. ? [ l
fll v .d ,I + -fl I 3. p j BLANK PAGE [ j 'i a I i -3
j l -u-c 4, pressure to achieve adequate net pump suction head. Core cooling water supply for the pumps is provided initially from the condensate . makeup tank and then the suppression pool when the former source is depleted. A break in any one of the suction lines leading from the suppression pool tc, the ECG pump will not lead to drain-ing the suppression pool since the ICCS pump rooms are to be water tight enclosures of limited dimensions. f the rate of reactor primary coolant loss and hence the size of 1 the primary piping system break determines the mode of ECG [ functioning sines each system is designed to function over a I specific range of break sizes. For small breaks in a liquid line (not steam line), the NPCI system can supply sufficient coolant to depressurize the vessel and also cool the core, depecding on the core spray system and LPCI only for long-term cooling following depressurization of the reactor vessel. For intermediate size breaks in liquid lines, the depressurising i function of the HPCI system and the large volume coolant makeup capability of either the core spray system or the LPCI system would act in combination to provide effective core cooling. In the event of a loss-of-coolant accident without the HPCI capability (i.e., the normal feedwater and HPCI are unavailable), the ADS would provide for reactor vessel steam depressurisation to permit operation of the core spray and/or the LPCI systes g%
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i. before excessive fuel clad heating occurs. For large breaks in liquid lines such that loss 'of coolant exceeds the makeup capacity of the HPCI system, the reactor vessel pressure drops at a rate fast enough for the' core spray system and/or LPCI to comunence adding coolant in time to cool the core. For steamline breaks, the most limiting for core cooling design would be a break in the main steamline inside the drywell and upstream of the ficw restrictor. The gCCS can provide sufficient core cooling for this accident. Se gCCS for Limerick is similar to those for other EWR facilities for which construction permits have recently been issued except for improvements in the MFCIS and LPCIS which enhance the cooling of the fuel cladding during postulated loss-of-coolant accidents (IDCA). These improvements are summarized belows a. HPCI System Improvement: Injection of RPCI fluid is modified so that the fluid is injected through the core spray sparser i rather than into a feedwater line. This alteration allows the cooling water to reach the core much faster and for longer periods early in the. IDCA transient. The result of the diange is the 1mtericE c! Mak fuel cladning temperature. The W CI flow rate swy also be increased by redesign of the pump impel-ler, thereby providing further temperature reduction. l .w. se _ w.g - h* 4
I r; b. LPCI System Improvement: Rerouting of LPCI lines for dis-charge through four separate reactor vessel penetrations to l the inside of the core shroud eliminates the loop selection logic and operation of the LPCI injection valves thereby 1sr proving system reliability, and giws improved core cooling. The LPCI injection valves are replaced by faster opening valves thereby permitting earlier flooding of the core. a 6.1.2 Easraency Core Coolins System Performance Evaluation 6.1.2.1 General The AEC Regulatory Staff has conducted a general reevaluation of i the emergency core cooling systems (ECCS) for light water rasctors. q l Although analytical methods and models were the principal areas considered in the review, experbnental results accumulated over the past several years and their applicability to analytical methods and models were also examined. This section presents the results of our review and evaluation of the effectiveness of ECG performance. The General Electric Company's calculations mod our evaluation do not include the recently improved capabili-ties in ECCS performance that result from modifications in the HPCI and LPCI systems. These modifications, discussed in the preceding paragraph (6.1.la and b), will increase the capability and improve the performance of the ECCS beyond that considered 1 in our conservative calculation. We will require that analyses, i I e,e. +
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67 - using the edified and improved ECCS capabilities, be included in the Final Safety Analysis Report. On June 19, 1971, the AEC issued 'an Interim Policy Statessat containing interim acceptance criteria for the performance of emergency core cooling systems in light-water nuclear power plants. Following publication of the Interim Policy Statement, the applicant and General Electric Cogany performed additional loss-of-coolant calenlations using the criteria and assumptions set forth in the Interim Policy Statement, particularly Appendix A, Part 2 which describes a General Electric evaluation model acceptable to the Commission. The following section presents further details of our review of the emergency core cooling systems in accordance with the Interim Policy Statement. Tlopical Report NEDO-10329 (dated April,1971) with supplements thereto and (amended) Section 6 of the PSAR pro-vided the bases for the review. 6.1.2.2 Discussion of ECCS Review The design basis accident for purposes of analyzing the effective-ness of the ECCS has been postulated as an instantaneous doele-ended recirculation line break such that the area available for coolant discharge from the reactor vessel would be the sum of ten jet pump nozzle areas and the cross sectional areas of the ,,,w esso i
= ! main recirculation line and the reactor water cleanup suction line (4.28 f t ). This is the worst case for ECCS analysis. For the purpose of analyses, the changing thermal and hydraulic phenomena that would occur during a design basis loss-of-coolant accident (thCA) may be described in five phases: (a) temperature changes and heat removal during reactor blowdown with associated l flow coastdown, (b) achievement of a critical heat flux (OlF) { st any point on the fuel rod cladding and associated temperature rise of the fuel and cladding, (c) lower plenum (fluid) flashing causing a temporary resurgence of core flow, (d) temperature t rise of fuel and cladding with diminished cooling and complete [ depressurization, and (e) temperature changes and heat removal lj during the ECCS operation. In the following paragraphs, we - 1 discuss each of these periods. The applicant's evaluation as well as curs considered a broad spectrum of break sizes but the following discussion relates to the largest (and " worst case") pipe break. Since the HPCI and LPCI design improvements occurred af ter the evaluation, the additional conservatism of these changes provide added safety margin to the final results. \\ Flow Coastdown The first phase of the 1DCA is the short-term blowdown period during which energy is remaved from the core by coolant passing i i through the core and exiting through the postulsted break, causing e h e
. the reactor coolant system pressure to decrease rapidly. Init ially. conditions, are nearly the same as during normal operation and nucleate boiling continues undisturbed. A short time later (about seven seconds) the core flow and system pressure decrease sufficiently that nucleate boiling cannot be sustained and the fuel rod heat transfer rate decreases markedly. Flow Stagnation Period he flow coastdown or short-term blowdown phase ends when the coolant flow through the core is assumed to stop when the water level in the reactor vessel downconer region reaches the inlet of the jet pump nozzle. Even though flow actually would continue at a reduced rate, for conservatism in the analysis, the flow in the core is assumed to stop at this time. During this period of flow stagnation, the heat transfer coefficient uced in the analy-sis is assumed to be equal to zero. His implies that no heat transfer by conduction or convection from the fuel rods to the coolant occurs. He fuel heats up adiabatically except for the heat loss due to thermal radiation from the fuel surfaces which is assumed to take place during this phase, which exists for about 2 seconds, he pressure in the primary system during this second phase of the postulated IDCA is continuing to decrease because of fluid mass loss through the bresk. De depressuriza-tion rate during the short-term blowdown regime is of the order of 15 psi /sec. ey gg.a "****8*** Weep u mde*-
l L l l ~I Lower Plenum Flashing Its the third phase of the IDCA, lower plenum flashing occurs. This is a flow phenomenon during the blevdown wherein a sudden increase in the core flow begins a few seconds af ter the core flow has decayed to near zero. The increase in core flow results when the liquid level in the vessel drops below the recirculation line suction nozzles causing the finv out the break to change from a liquid phase to a steam phase thereby increasing the rate of j depraesurization of the system (to about 30-40 psi /sec). l Mis rapid depressurization results in a rapidly changing thermodynamic state of the fluid in the primary system. Because the fluid in the lorer plenum beneath the core was initially in a subcooled state, it does not change thermodynamic' state during early blowdown as does the rest of the fluid system; however, when the system pressure decreases to its saturation pressure (or lower), this fluid flashes to steam to give a large increase in flow through the core. his period of the LOCA is called " lower plenum flashing." Calculation of flows, temper-atures and pressures during this phase depends on the thermo-dynamics of the flashing process, the effect of flow maldistribution, the resistance to flow of a two-phase (stese-water) mixture through the core and jet pump diffusers, and the rate of blowdown through the break. t
.71 - e 4 During this period of increased core flow, GE originally assumed that eucleate boiling is reestablished and that relatively large heat transfer coefficients result. Although nucleate boiling may 1 be reestablished, we required GE to perform the analysis of tb the more conservative assumption that only stable film boiling occurs. his results in greatly reduced values of the heat transfer t coefficient, as determined by the Groeneveld correlation during { the period of lower plenum flashing. This assumption is in accord with Appendix A, Part 2 of the Coasaission's June 19, 1971, Interia l Policy Statement on emergency core cooling systems. ) l Core Heatup Following the period of lower plena flashing it is conservatively assumed that no additional coolant is available and no convective cooling occurs. The net effect of heat generation by radioactive f decay of fission products and thermal radiation among the fuel i rods causes the core to heat up. l ECCS Operation Although the loss of water level or the increase in drywell pressure resulting from a pipe break is sensed immediately and the EC2 is signaled to start, the actual injection of water by the low pressure systems does not occur for about 30 to 40 seconds i ) af ter the large break occurs. This time is needed for the diesel generators to start and accept load, the reactor pressure to fall 1 .WM**** s ee e. e a, -ww ww. m one, gg g,,,,, g. 8", go, p o m, L
l e 4
- l I
below the (low pressure) ECG pump discharge pressure, and the ) i ECG pumps to achieve full flev. Water is injected into the ) reactor thro agS both the LPCI system and the' core spray system. In accordance with the AEC Interim Policy Statement, the applicant has considered two limiting cases. These cases result from a sin-gle failure of an active compoesnt. Failure of the most critical emergency diesel-;,seerstor results in the operation of one core spray system in conjunction with 3 out of 4 LPCI pumps. Failure of the IPCI injection valve results in the operation of both core spray subsysteam only; this valve will be eliminated with the LPCI modification but has been used for the evaluation because of the conservative nature of this asstamed failure. j 1he appilcant and CE have pe formed these analyses for the entire t pipe break spectrum, up to and including a double-ended severance of the largest pipe of the reactor coolant pressure boundary. In the limiting case of a postulsted double-ended break of a primary coolant systa recirculation loop pipe with the simultaneous failure of the LPCI system, the calculated maximum fuel clad temperature is below the acceptability limit of 2300*F. The cladding-water chemical reaction is calculated to be well below the limit of 1% of the active fuel cladding in the reactor for all break sizes, using th.* AEC evaluation andel.
e I ) - 73 '- \\ For long-term cooling of the core following a IDCA, a continuous 1 heat removal capability is provided. 1he ECCS will remove the decey and residual heat from the core under DBA conditions using i suppression pool water. If the latter is unavailable for some I ~ reason, RRR service water is available through cross connections. 1 Under low river flow conditions, as discussed in section 2.3, the applicant will provide assurance of an adequate supply of water for the Residual Heat Removal (RRR) Service Water ti,vstem and the Emergency Service Water System (ESWS), that provide the cooling water for the RHR System and emergency equipment, respectively. The RER Service Water System, the ESWS, and their structural enclosures, are to be designed to Class I seismic standards with adequate redundancy of equipment and pipelines to assure adherence to the " single failure criterioc, " 6.1.3 Conclusion We conclude that the design of the Limerick emergency core cooling system is acceptable based on the analysis which shows that the consequences of the loss-of-coolant accident are such that (a) the calculated maximum fuel rod cladding temperature does not exceed 2300*F, (b) the amount of fuel rod cladding that reacts chemically with water or steam does not exceed 1% of the total amount of cladding in the reactor, (c) the clad temperature transient is terminated at a time when the core geometry is still amenable to
l I l l I l l 1 1 For long-term cooling of the core following a 14CA, a continuous heat removal capability is provided. The ECCS will remove the decay and residual heat from the core under IISA conditions using ) suppression _ pool water. If the latter is unavailable for some b reason, RER service water is available through cross connections. Under low river flow conditions, as discussed in section 2.3, the applicant will provide assurance of an adequate supply of water for the Residual Heat Removal (IER) Service Water 5,vstem and the Emergency Service Water Systes (ESWS), that provide the cooling water for the RRR System and emergency equipment, respectively. The RER Service Water Systes, the ESWS, and their structural enclosures, are to ibe designed to Class I seismic standards with adequate redundancy of equipment and papalines to assure adherence to the " single failure criterion." l 6.1.3 Conclusion l We conclude that the design of the Limerick emergency core cooling system is acceptable based on the analysia which shows that the consequences of the lose-f.% wlant accident are such that (a) the calculated maximum fuel rod cladding temperature does not exceed 2300*F, (b) the amount of fuel rod cladding that reacts chemically with water or steam does not exceed 11 of the total amount of l cladding in the reactor, (c) the clad temperature transient is terminated at a time when the core geometry is still amenable to 0-;
l 3 l I i ) _ 74 cooling, and before the cladding is so embrittled as to fail during or after quenching, and (d) the core temperature is reduced and decay heat is removed for an extended period of time. We conclude that the design of the emergency core cooling system mesta the requirements of criterion 35 Emergency Core Cooling, of the General Design Criteria published in 10 CFR Part 50, Appendix A. 6.2 Hydrogen Generation in Primary Containment Following a LOCA 6.2.1 Hydromen Control System To eliminate the possibility of hydrogan buildup during the 1 period following a loss-of-coolant accident, the containment atmosphere will be processed through a catalytic recombination l system to remove hydrogen and oxygen. The reccabiner trains, l i* l located outside the primary containment, will pass containment I atmosphere through a halogen profilter (to remove iodine and its compounds), a catalytic recombiner, and a cooler-condenser to remove water vapor and to cool the remaining gases for return to the containment. The AEC design guidance, Safety Guide No. 7, will be used for the design of this system. We find auch cri-teria to be acceptable and the description of this system ade-l quate for this stage of the review, but will require further l information on design details and operation to assure its capa-bility for keeping hydrogen concentrations in the post-LOCA 'l 1 1 O +
i i l l containment below the flammable limit. This information can be reviewed during the construction of the facility. 6.2.2 Cont ainmen t Inertina As an operational technique to reduce llammable gas concentrations, the containment will be inertad with nitrogen. Information to be presented in the PSAR will further amplify the description of this technique. 6.3 Other Engineered Safety Features Other engineered safety features for the Limerick facility includes (a) a containment cooling sesystem that is one asode of operation of the Residual Heat Removal (RER) System, (b) seein steamline flow restrictors, (c) control rod velocity limiters, (d) control rod housing supports and (e) a standby liquid control system. These safety features are common to other WR facilities recently authorised a construction permit. 7.0 INSTRUMENTATION. CONTROL AND ELECTRIC POWER SYSTEMS 7.1 Instrumentation and Control Systems The instrumentation and control syntess have been evaluated against the Coensission's General Design Criteria (GDC) dated February 20, 1971 and the Proposed IEEE Criteria for Nuclear Power Plant Protection Systems (IEEE-279) dated August 28, 1968. A ceuparative review of the Limerick plant with previous WR 6
l l1 l i l r i L l
- nuclear plants was undertaken.
The reactor proteccion system and the instrumentation which initiates and controls the engineered rt fety features were found :o be functionally the same as recent BWR's 'and are acceptable. Hence, the evaluation was limited to a review of areas where new infomation was obtained. These areas were: a. Post Accident Monitoring Equipment b. Environmental Testing c. Engineered Safety Features fastability d. Cable Installation 7.1.1 Post Accident Monitoring Instrumentation We have reviewed the applicant's description of instrumentation for i f monitoring and recording key operational variablea during and after i a postulated loss-of-coolant accident. We find that instrumentation I provided for this function is acceptable. 7.1.2 Environmental Testina The applicant has identified electrical equipment located in the primary con:ainment which is required to operate during and subsequent to an accident. The qualification test procedures and test conditions to assure that the components inside containment will perform as required were adequately identified.
I ? - 7 8.- 7.2 Electrical Power Systems l 7.2.1 Offsite Power Limerick Units 1 and 2 will be interconnected to the Pennsylvania-me. Jersey-Maryland system through 220 kV and 500 kV transmission , systems. Power from Unit i feeds a 220kV seatstion and power from Unit 2 feeds a 500 kV substation. The substations are approximately 3000 feet apart and interconnected by a 500-220 kV bus tie transformer and transmission line. The transmission line is on its own independent set of towers. Each substation is arranged in a breaker-and-a-half configuration. Five transmission lines emanate from the station, all on individual sets of towers. j Dro lines leave the 220 kV substation en separate rights-of say and three lines leave the 500 kV substation, one on one right-of-way and two on another right-of-way. Sufficient distance exists between any two rights-of-way that assures any one event will not physically affect any' other one. Two startup auxiliary power transformers are provided, one located at the 220 kV substation and the other located at the 500 kV substation. Each transformer is capable of j supplying power to either unit. The applicant has completed trarmient stability studies that haye simulated the loss of each of the Limerick generators and the loss of the largest generator on the 5C4 kV grid. The results have shown that loss of offsite power vould not occur under these conditions.
79 - Our review of the preliminary design indicates that General Design Criteria No.17 of 10 CFR Part 50 can be met. We conclude that the offsite power system will be, acceptable. l 7.2.2 Onsite Power he engineered safety feature and safs shutdown loads are divided between four 4 kV emergency buses for each unit s'uch that the operation of any three for a unit will supply minimum safety requirements for that unit. Each bus in both units can be supplied power from the 500 kV startup auxiliary transformer or the 220 kV startup auxiliary tran4former. hese supplies have independent sad separate auxiliary transformers for reducing the voltage. A third supply to each bus is from a diesel generator. A total of four diesel generators are provided for both units and each is an ' e asd to two emergency buses, e. buc 'for each unit. Each die-sel generator is started automatical?.y of an emergency core cooling system initiating signal from either unit or by loss of offsite power. The accident loads are automatically sequenced on each diesel generator. The dieaal generators will have a continuous rating greater than the electrical safety loads. Each diesel generator is housed i i 1. individually in a reinforced concrete Class I seismic structure + above the design flood level. Each diesel generator will also be [ a self-sustainion entity with its own independent lube oil, fuel i oil, jacket water,1tbe oil cooling, air starting and control j
l I ' Four diesel fuel oil storage tanks are provided, one sys tems. for each diesel unit. Each tank has sufficient fuel for operating its associated diesel unit for seven days. The 115 volt a-c systeam provided for safety are arranged with two physically separate rametor protection system buses for each unit. ne 125 and 250 volt d-c power supplies for each unit consist of l-two independent 125/250 volt, 3-wire systems. Each system has two 125 volt batteries, each with its own charger and distribution panel. We battery chargers are capable of keepihg the batteries fully charged and supplying the d-c system loads simultaneously-ne two 125 volt batteries for each systes are located in a Here are four rooms on the station, each have j separate room. i f ceparate and independent supply and exhaust ventilation systess with two fire dampers. De batteries will be sized to supply essential loads for a period of greater than four hours on loss of its battery chargers during any plant operating or incident condition. This d-c power supply arrangement provides for adequate separation and independence of redundant supplies for both units. We find the preliminary design of the onsite power supply 1 l accept able. The applicant will meet the intent of the Coenision's 1 Safety Guides 6 and 9. Safety Guide 6 is concerned with "Indepen-dence Between Redundant Standby (Onsite) Power T - es and Between 'l l l
-. I I e l
- Instrumentation, controls, and emergency power equipment that are j
required to be functional during or after a design basis earthquake will be analyzed or vibration titsted to assure that the design requirements will be met. A13 equipment supplied by General Electric that is required to be functional will be tested and all equipment supplied by Bechtel, where practicable, will be tested. We conclude that the criteria for environmental testing is acceptable. 7.1.3 Enmineered safety Feature Testina Periodic testing of the engineered safety features requires the same degree of on-line testability as is required for the reactor trip system. The applicant's commaitment to this requirement is satis factory. 7.1.4 Desian Criteria for Cable Installation the applicant has documented his design criteria for cable installation in Supplemen: 4 to the Preliminary Safety Analysis Report. Pb110 wing the review, we concluded this criteria fe acceptable. e' =F
) Other Distribution Systems." Safety Cuide 9 is concerned with " Selection of Diesel Generator Set Capacity for Stan6y Power Supplies." 8.0 AIDCILIARY SYSTD6 8.1 General I The auxiliary systems are described in Section 10 of the Preliminary Safety Analysis Report. These varied systems normally provide plant services auxiliary to the production of electrical power. We have revieued the safety related design concepts of these systems and found them acceptable. The safety reisted aspects of systems which received special attention are discussed in the following sections. 8.2 Radioactive Waste Systems 8.2.1 _ General The applicant's design objective for gaseous and liquid radwaste discharge is to reduce the radioactivity in effluerats such that that the annual average concentrations are less than 1% of 10 CFR Part 20 lim *ts for each systems. The liquid and solid radio Ivo waste disposal systems are similar to those provided fo other facilities now under construction. However, the gaseous radioactive waste control and disposal system represents a significant departure from earlier plants since a cryogenic (Iow tegerature) system will be used to provide much longer holdup i i
I e
- i. and decay of the radioactive gases prior to dischstge. Another improvement will be the use of water from the condensate storage f
tank as a source for che steam used in sealing the main turbine gland seals instead of extraction steam which has a higher radioactivity content. The radweste and off-gas buildings are designed as Class I (seis-mic) structures. Bey are rainforced concrete and structural 4 steel buildings'largely below plant grade. Concrete exterior walls are waterproofed and designed for bydroctatic effects as l l necessary. 8.2.2 Liquid Radwaste System here are separate subsystems for the collection and processing of the liquid waste consisting of: drainage or " drains" from plant t equipment, floors, laundry, and those areas involving use of chemicals. He subsystems can be interconnected through existing cross-connections to provide flexibility in processing. l Me majority of liquid waste comes from equipment and floor drains that generally are of higher purity (lower conductivity) than the 1aundry or chemical drains (high conductivity). De equipment and floor drains are normally filtered, demineralized, and returned to the pri :ry coolant via the condensate storage tank. The chemical ,= and laundry drains are normally discharged to the environs after treatmen t. Wo radwaste evaporators will be available for j
r 4 processing liquid with relatively'high radioactivity or chemical impurities. The liquid wastes from any subsystem may be evaporated; the distillate is normally recycled to the condensate storage tank for re-use while the evaporator concentrates are processed through the Solid Radwaste System. This evaporation and concentration pro-cess will reduce the discharge of contaminated liquid. This con-forms with the objective of processing and re-using most liquid wastes within the plant. Liquid wastes being discharged to the environs are handled on a batch basis. The liquid batchu are held for a period of time to allow complete mixing, sampling, analyzing, and processing for transfer to the condensate storage tank, solid radwaste storage, or to the environs. Discharge of any processed liquid is accom-panied by dilution in the plant circulating water discharge ' pipe. 7be expected annual release of radioactivity in liquids will be I fif teen mil 11 curies from the activity that may be contain:4 in chemical and laundry drains. If discharge of floor drains is required because of high chemical content, this liquid ney be decontaminated in the evaporators, if required, and the annual average radioactivity concentration in the plant effluent will still be below 1% of the annual average concentration limitation of 10 CFR Part 20. 1
b'* 4 The appli. ant' Atudy of operat iny, vhperience of boiling water reas t or., such as the KRf1 reactor in Germany and the Dresden I rear.cor'provided the basis for consideration of tritium. A Pubite Health Service (PHS) publication on Dre:.: den I indicates that the average radwaste release rate for tritium in 11guld for this 200 Mw(e) reactor was 0.05 nicrocuries per second and that the calculated ratio of (radcaste) tritium rele.tse rate to tritium production rate in the fuel was 0.001.II} Tritium in the coolant in produced frnm neutron.8ntera:tions with deuterium and from dif fusion or leakage of fissfor. product tritium through the fuel cladding. The amomt of fission product tritium in the fuel, while l i greater in quantity than the neutron interaction sources, does not contribute significantly to the total quantity in the coolant because of the fuel cladding integrity and the low rate of tritium diffusion through the Zircalloy cladding.. This latter conclusion was stated both by the PHS in its report ( } and by D. C. Jacobs in a monograph on tritium sources. The applicant initially used this information to conclu.fe that about 5 9 pC1/second production rate for each reactor under maximum fuel leakage conditions would be the maximum. However, subsequent operating experience from the I '" Radiological Surveillance Studies at a Boiling Water Nuclear Power Reactor." U.S. Department of Health, Education, and Welfare Public Health Service - Environmental Health Section, March 1970, pagea 85 and 86. ( Ibid, page 13. ( "Sou rces o f Tr i t ium.... " D. C. J ac obs US AE C/ TI D pub l ica t i on, 1968, page 22. i 4
ts-KRB s t a t i on inditated t h s. t numerous fuel f ailui es had little etteit on the t r t iutn release rate. The applfrant accordingly reduced th. t rit ium f ormat ion rate to that rest.l t ing f rom neut ron-deu t e rium in terac t ion, 0.3 mi c rocurits pe r s,econd, which is bel.eved to oc (bservation and measurement of radioactivity the only saajor source. levels and production rate for tritium will be fo? lowed during reactor operation to assure that technical specification limits, developed at the FSAR stage, are not exceeded and that c cm c en t ra-tions dischcrged f rom the f acility are in accordance with the Cosmaission's requirements for limits as low aa practicable. The liquid radwaste system will be arranged so that any leakage ar.d/or spillage is retained inside concrete enclosures or dykes. Liquids are returned to the liquid radwaste system through the radwaste d rain system. The enclosures and drain systes have the capacity to handle a major leak in the Isrgest tank without per-mitting radioactive liquid to escape of fsite, If a spill of liquid at t.he center of the site occurred, the low permeability of onsite soils and underlying rock would provide holdup as follows: f our and one-half years to the edge of the site- (2500 f t) and one and one-half years to the Schuylkill River (1000 f t). In suimmary, we have reviewed the applicant's design cbjectives and concept for the liquid radwaste system and found the system ac cep t ab l e. We shall establish opetating limits, in accordance I
} " wi th the Connission 's regulations, in the Technical Specifications I at the operating license stage of review. 8.2.3 . Caseous Radwas te Sys tem The gaseous radwaste system has the main functions of recon 61ning the radiolytically formed but nonradioactive hydrogen and oxygen, g condensing the reselting steam, providing for holdup for radio-active decay of short-lived isotopes, and removing radwaste krypton and xenon from the process steam and providing for their } storage. To accomplish these functions, a catalytic recombiner subsys tem, a cryogenic (low-temperature) treatment subsys tem, ath. a storage subsystem will be provided. The source of radioactive gases that will be processed is the off-gases from the normal operation of the main steam condenser. These gases include i fission product noble gases (present only if fuel leaks occur), setivation gases, radiolytically formed hydrogen and oxygen, and j i the air that normally leaks into the condenser. The volumetric j flow rate of these gases is such that the radioactive gas flow rate is negligible by comparison with the others; the total aspected off-Fas flow rate is about 279 cubic feet per minute (SCFM) per unit. The hydrogen-oxygen catalytic recombiner subsystem for each tritt inclades a preheater, recombiner, and af t.cr ce'enser. Excess
6 hydrogen will be injected into the off-gas flow upstream of the
- recombiner to assure that the oxygen is completely removed; with-out oxygen there will be no flammable. mixture.
Research and development of catalytic recombiners is being supported by the applicant and other utilities. The cryogenic treatment subsystem provides for radioactive decay of the short lived isotopes and removal of the krypton and manon from the process s tress. The subsystes will include retreatment to reduce oxygen, 00 and water vapor, a cryogenic distillation 2 column to remove krypton and zenon, and a three hour cryogenic delay column for decay of nitrogen 13 and residual neble gases. The liquified krypton and zenon are collected in the column re-boiler while the purified gaseous radwasta stream is released. The effluent la continuously monitored. A common standby cryo-i genic treatment train will be provided. The krypton and xenon rich liquid is removed batchwise from the i system for gasification and storage at ambient temperature in i pressure vessels. Procedural limitations for the talesse of the gases under controlled conditions will be developed and reviewed as part of the Technical Specifiestions at the operating license stage of the review process. I s i e ] ,.m
i - 4" l e The s'pplicant has indicated that the catalytic reconbiners of the gaseous treatment system will' trap halogent, so that no significant -i I release is expected. Experience with the KRB BWR gaseous waste i I j t treatment system indicates that this plating-out of iodines on metals and on hester elements in the recombiner is to be anticipated. However, charcoal and HEPA (high efficiency-particulate removal) i filters will be installed downstream of the recombiners to ensure removal of radioactive isotopes of gaseous fodine and radioactive particulate. The particulate content of off-gases will also be diminished by the scrubbing and filtering in af ter-condensers, water-sealed compressors, water separator retreatment sections, and distillation columns. The off-gas stream for processing does not include other sources l such as the of f-gases from the mechanical vacuum pump (operates l about 10 times per year for four hours during startup). HPCI tur-I bine condenser testing, and the gases resulting from ventilation a of contaminated (low level) areas (ches labs and equipment rooms). The applicant maintains that the few occurrences and the low con-l centrations of radioactivity expected do not warrant attempting 4 to process these very high volumetric flows in the gaseous red-l 1 j was te treatment system. Since there will be a general reduction in the release of radioactive gases as a result of use of the cryogenic treatment subsystem, the contributions that were l i formerly minor f ractions of the gaseous releases and are still l l i 1 I
e e i ven low, have 'become relatively more significant. Accordingly, in keeping with the goal of reducing releases to levels "as low as practicable" we will continue our evaluation of these low level sources during the construction period to determine 'the' practical-1 I icy of their elimincion. l l Sin;:e the applicant will operate the gaseous radweste treatment system continuously and will use condensate water for steam to seal the turbine glands, we consider the applicant's concept for control, collection,.and processing of gaseous radwaste a signifi-cant contribution to fulfilling the requirement to release radio-activity at a level as low as practicabis. The applicant will provide and we will review details on the design and development of the total system as it progresses. 8.2.4 Solid Radwaste M sten The solid radwaste system is housed in the radwaste building and is designed for processing wet waste from water treatment and cleanup systems or radwaste processing, liquid concentrate from the was te evaporator, rend dry wa.<tn such as filters, ragn, clothing, and equipment parts. Radioactive materials in these solid forms will be prop 1rly protected and packaged for shipment to an authorized disposal site. The processing and packaging of solid radwaste from filters and deaf neralizers is dependent upon its level of radioactivity. 1.ov e' 4 I
i i specific activity (LSA) and high specific activity (HSA) materials each havt a special type of container designed to acco.wdate the material and remote filling methods used. The LSA material is f ed f rom hoppers or waste concentrate storage tank to downcomers I that lead to one of the two 2-f t diameter openings on the top of a 200 cubic feet container. Absorbent material is added to the I liquid LSA radwaste containers in the empty s torage area. Con- { cainer filling of HSA material is accomplished by use of a filter i drum positioned in the same fashion as a LSA container. The RS A { container la filled by pumping a HSA slurry through the filter drum where a filter bag removes and retains the solida inside the drum. Closure of the LSA container with S5-gallon drum lida is accomplished by a remotely controlled capping device. Closure of the HSA filter is achieved by inserting plugs into the hose con-i nections used for filling. k i i The solid radvaate system is similar to that satisfactorily used I l in other BWR facilities. The concept of design and operation is l l acceptable. .I 6.3 Spent Fuel Storage ,I Each reactor unit has a spent fuel storage pool capable of I l accommodating safely 150 percent of the full core load of fuel i 1 as we sblies. Each pool is lined with stainless steel. No inlets. i outlets, or drains are permitted that night allow the pool to be '? drained below spproxim.itely 10 f t above the top of the active fuel. Lines extending below this level are equipped with syphon breakers, check valves, and other anti-drainage devices. De spent fuel pools and the spent fuel storage racks are designed as Class I seismic structures. We find these criteria to be acceptable. A con 61 nation of physical arrangemer.t of pertinent structures and administrative procedures for spent. fuel cask handling will ensure that the cask cannot fall into the spent fuel storage pool or affect other equipment required for a safe plant shutdown. De reliance on administrative procedures to achieve this design goal will be described further in che Final Safety Analysis Report. 1 We find this approach to be satisfactory. 8.4 Emeraency Service Water (ESW) and Residual Heat Removal Service Water (RNRS) Systems he Emergency Service Water (ESW) System is designed to provide cooling water from the Schuylkill River intake structure to emergency equipment to be used during loss of offsite power conditions or a loss-of-coolant accident. D e Residual Heat Removal Service Water (RHR5W) System is designed to provide cool-ing water during post-accident conditier.a for heat removal from the RER eystem (e.g., LPOIS mode; and supply water if needed for post-accident flooding of the core or primary containment. Poth systems and their related structures including the pump house and
l 92 - pumps are.of Class I seismic design and operable at river flood stage or low flow. De pump house is to be designed with water-i tigh t compartments that prevent any single failure from flooding all ESWS cr RHRSWS pumps. We find the design criteria for these sys tems acceptable. 8.5 Main control Room Ventilation System he ventilation and air conditioning system for the main centrol room will be designed to Class I seismic standards. A radiation monitor will provide the initial signal, on high radiation in the fresh air intake duct, that will result in the automatic diversion I of the air flow to a clean-up system or alternately, depending on [ the outcome of the applicant's study described in section 9.6, to a closed recirculation sys tem. Should smoke be detected in the ducta, a purge with 100 percent outside air may be manually i actuated. I i We have concluded that the design of the control room ventilation 1 j system is acceptable. 9.0 ACCITENT ANALYSIS 9.1 General A number of design basis accidents have been evaluated in the course of our review of the Limerick Station. Pertinent details of each design basis accident are discumised in the following st6 sections. J A
. e i The appif cant initially did not use the assumptions outlined in Safety cuide #3 to calculate the design " base case" accident ex-l pos ures. However, following our request for such ' calculations, the applicant prepared data on exposures using AEC assuusptions for each postulated accident, except that the applicant used a filter efficiency of 99% for the recirculation system filter and for the Standby Cas Treatment System (SBCTS) filter. We have assumed lower efficiencies for these filters as descrf. bed in the folicnring sections. Our calculations account for the Reactor Building Recirculation Sys tem, the mixing and cleanup of T sa'.or Building atmosphere, and us,e of the Standby Cas Treatoent System (SBCTS) wherever appro-priate to the accident. For the loss-of-coolant accident, the meteorological values were chosen conservatively since onsite meteorological information was not available. Type F and 0.5 meter per second (stable) wind conditions were utilised during the early hours following the accident. Meteorology for the time periods following the first few hours of the accident was based on an atmospheric diffusion model for ground level release that la described in Safety cuide 3. All of the accidenta result in calculated exposure doses well below the 10 CTR Part 100 guidelines.
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e 9.2 los s-o f-Cool an : Accident (IDCA) The assumption of the complete severance of the suction line to the primary water recirculation pump is an assumed accident used to evaluate containment capabilities and exposure doses at Limerick and other boiling water reactors. The calculation of liquid mass flow rate for the blowdown of the reactor vessel la based 'on an effective break area of about 4.82 sq. f t. that re-sults from the summing of effective flow areas for the following fluid sources: recirculation lice (3.667 sq. ft.), twenty jet-pump nozzles (1.076 sq. f t.), and the reactor water cleanup line (0.087 sq. f t.). A discussion of our review of the analytical model for the blowdown and costsiamaat pressure calculation and also the calculation of expossre doses resulting from a ihCA is presented in the following paragraphs. 9.2.1 Analytical Model for Pressure Suppression The applicant had used the General Electric Company's standard i procedure for calculating reactor coolant bloudown and the 1 subsequent primary containment pressure transients. peak calcu- { Isted drywell pressure and peak calculated dryvell floor differ-ential pressure depend upon many assumptions for the thermal and hydrodynamic flow phenomena associated with reactor vessel blow-down and the suppression of pressure by steam condensation in the wetwell pool. We questioned the conservatism of some of the assumptions used in the calculations. The moe t s ignifi car.t a rea
_ e of concern was the' analytical nodel' used for the vent flow phenomenon. Hence, the applicant and in turm, General Electric Company, were asked to justify or refine their a alytical model used to calculate the primary containment pressure transient following a loss-of-coolant accident. In response, the General Electric Company refined the analytical model. The mos t signifi-cant change is the use of a mixture density term in the vent flow n; model to account for the mixture of water, gas (air), and steam which flows through the vents. Bis refinement of the calcula-tional model has been documented in Amendment 5 to the Preliminary Safety Analysis Report and in the General Electric Company's Topical Report NEDO-10320 "The GE Pressure Suppression Containment Analytical Model," April 1971, and NEDO-10320, Supplement 1, l Msy 1971. 9.2.2 LOCA Exposure Dosec The assumptions include conservative values representing the j I capabilities of the Primary Reactor Containment, the Reactor Building, the Reactor Building Recirculation System, and the Standby Cas Treattsent System. These assumptions include: a release rate from the Primary Containment of 0.5% of its free air voltsee per day; mixing of effluent in only 50% of the Reactor I Building volume; and filter efficiencies as described in sections j 1 l C
.w,. l 5.1. 2.1 and 5.1. 2. 2. The calculation of finnion products re-1 } leased durir.g the U)CA is bas ed on assumpt ions that the total core is involved, and that 2002 of the noble gases and 252 of the halogens are r.tIcased to the 'in terior of the Reactor Containsu t. A ground level release height, a wake factor of 0.5, a seactor Building area of 2200 m, and dif fusion parameters discussed in sesection 9.1 were used for atmospheric diffusion calculations. 1 Exposure doses (gens) calculated for the teole body and for the thyroid at the Exclusion Area Boundary are 3.4 and 15.0 for two bours exposure and at the Law Population Zone soundary are 11.0 end 58.0 for 30 deys esposure. These values are well within the 1f aits of 10 CFR Part 100. 9.3 Fuel Handling Accident (Refueling Accident) i i This calculation is based an an assumed accident used for other boiling water reactors as well as Limerick. The reactor has been i shutdown for 24 hours for refueltag, the reactor vessel-head is l off, and the cavity above the reactor vessel fa flooded to the l 1evel of the spent-fuel pool. While transferring an irradiated fuel assembly, the assembly la dropped by the ref ueling equipment onto the top of the reactor cars f rom the maximum height allowed by th e eq uipmen t. The dissipatio.s of the potential energy of the dropped fuel assembly results in damaging 111 fuel rods which have experienced 1.5 timeu the average core neutron flux. The event
occurs unde.vster and 901 of the released halogens remain in the water. No plateout of halogens is assumed. High radiation sig-nata initiate the Reactor Building Encirculation System and the SBCTS, Other assumptions are as outlined above for the IDCA. Esposure doses (Rams) calcul'ated for the whole body and for the thyroid at the Exclusion Area Boundary are 1.6 and 7.8 for two hours esposure and at the law Population Zone Soundary are 1.5 and 3.5 for 30 days exposure. These values are well within the limita of 10 CFR Part 100. 9.4 Control Rod Drop Accident As accident model considered for other boiliss water reactors as teoll as I.inerick ta used as the basis for calculations. 1he acci-dent occurs 30 minutes af ter a shutdown from esteasive power oper-ation uhile the reactor is being restarted from " hot atenary." We assume that a control rod stuck to che fully-inserted position. The rod-drive is disconnected from the control rod and retracted. Then, the control rod falls out of the core. This rapid removal of the control rod could result in a Imrse, but short-duration increase in power in those fuel elements surrounding the control l rod. This increase in pouer could canase rupture of fuel rods duc l to overheating. Based upon the postulated worst case, i.e., the i' reacter critical at 10% full power and coolant temperature at j 547'", the analysis indicate that 330 fuel rods could fail. We l assume that the 330 fuel rods are in a region of the core that I d k l ).-
experienced 1. 5 t imen the averag.* core.' lux. All of the fission product ncLle gases are released from the damaged fuel rods. Of the halogens released f ree these damaged rods, a f raction is re-tained in the water such that 5% of the halogens contained in the damaged fuel reach the turbine-condenser volume where 502 plates-out. The rammining airberne activity la assumed to leak from the condenser at a rate 0.5% of the turbine-condenser free volume (250,000 cu. f t.) per day. This activity is then assumed to ex-filtrate from the turbine building (a ground level release). A wake factor of 0.5 and turbine building area of 1050 m are used with LOCA meteorology for dif fusion calculations. Exposure doses (Rams) calculated for the whole body and for the thyroid at the Exclusion Area soundary are 0.8 and 15.0 for two j bours exposure and at the low Population Zone Soundary are 0.5 j and 15.0 for 30 days exposure. The exposure doses for this acci-dont are well within the limits of 10 CFR Part 100. 5.5 Main steam Line (pst) Break In evaluating this secident, it was postulated that one of the four main steamlines ruptures outside of the primary containment while the reactor is operating at rated power. Steam and water from the reactor then escapes through both ends of the broken pipe to the environment. The amount of radioactivity raleased was calculated from the activity of noble gases and individual fadine radioisotopes that pass through the main steam line durig ~
_ - _ _ _ _ _ _ _ _ _ _ - _ _ _ - - - - _ _ _ _ - - - _ - e e the maximum allowable closure time for the MSL isolation valves. Usint assumptions comparable to those given in Safety cuide 5 and an assumption for coolant activity corresponding to a high off-gas discharge rate of 1.0 curf e/second af ter 30 minuces holdup, the applicant calculated exposure doses (Rass) for the whole body and thyroid at the Esclusion Area Soundary of.613 and 6.0 for two hours exposure, and at the Imv Population Zona soundary of.0079 and 1.1 for 30 days exposure, nese esposure doses are well within the guideline values of 10 CFR Part 100. The Regulatory staff has not rec.alculated independently the doses for this event at this stage of the review process. The offsite consequences of a main steamline break are dependent upon the radioactivity con-centration, primarily todine, in the coolant. Sue, the conse-quences are susceptible to operational control, namely the setting of limits on allowable iodine concentration in the coolant. We will assure that tne Technical Specification limits for todine concentration in the coolant will be restricted so that the cal-culated exposure following an assumed Mit break will be well within f.he 10 CFR Part 100 ' guideline values. 9.6 Control Room Exposure Doseg Durina Accidents Using the design basis of 5 Rem uhole bMy and 30 ? us thyroid axposure doses as the criteria, the applicant has described maa- ) i sures taken to keep exposures within the limits. The exposure t analysis performed accounts for radiation exposure throughout the s i f
I f r - 100 - i rourse of a pos t ulated 1.0CA. TSe applicant's analysis gave 0.44 Rex (whole body) integrated doce in the control rma for 10 days continuous occupancy with a primary containment leak rate of 0.635Z of the volume per day. Assuming shif t operation with j 1 1 normal rotation of personnel, their entrance and exit to the con-l l trol room, no credit for breathing apparatuc, and other accident l r.onditions, the applicant stained an exposure dose for 30 days i to a control room operator of 1.7 Rees (whole body) and 0.24 Rom (thyroid). 'If personnel are esposed to unfiltered outside air due I to a control room air purge operation for the period of one hour, the applicant's calculation results in a whole body dose of 0.16 Rae and a thyroid dose of 3.2 x 10' Rom. Our cair.ulations of the control. room esposure doses dif fer from ,i the applicants' because of differences in the analytical models for computing the release of airberne radioactivity from the I Reactor Building, uncertainties in the analysis of meteorological I phenomena, and differences in filter efficiencies. Other
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j assumptions for our calculation of the inhalotion dose (thyroid i exposure dose) are generally the same as the applican:'s. Our I calculated thyroid exposure dose is higher than allowed by the 1 { current criterion (30 Ren maximum). Accordingly, the applicant t has begun a study to dett rmine ways to reduce this dose to accept-able levels. Presently, an acceptable solution (used on *,he Nwbold Island facility design) is to provide a capability for l
101 - ) isolating cont rol soon ventilation (1) reci rculating the I control room air, without external make-up, through filters for nonradioa*:tive clean-up and reuse of the air for as Icog as i possible (estisuted 100 hours before C0 buildup contaminates air) { 7 i af ter dacecting airborna radiation in the fresh air intake, and (2) using the emergen:y diesel generators pcwer supply for opera-ting control room air cooling equipment. The applicant's proposed I solutions to this problem will be reviewed further during the con-l struction period. We will assure that the solution conforms to the exposure criteria. 9.7 Main steam Line Isolation Valve Leakane i If one isolation valve fails to close in one main steam line fo11 cuing a loss-of-coolant accident, leakage chrough the other i (closed) valve is an uncertainty that may result in exposure doses uhich may approach 10 CFR Part 100 guideline values. The low flow rates which can caune exposure problems are such that the uncertainty remains in current MSL 1ao2ation valve design and I testing concepts. Hence, the applicant has aareed to provide an acceptable method of sealing the MSL laolation valves during the pos t-LOCA period. .% section 4.2.4 of this report for details. O - ~ ~ ~ a i j
I - 102 - 9.8 _ Instrument Line or Process Lins Break Concern with the consequences of a break outside primary contair-i ment of an "unisolatable" lina prompted the applicant to adopt the suggested methods of the Cosmicsion's Safety Guide No.11, { \\ l " Instrument I.ines Penetrat.ng Primary Containment." To reduce the I consequences of the failure of the small diameter instrumentation lines penetrating primary containment as well as the failure of any stessline, we will establish, at the time of the operating license review, the limit on r-imary coolant concentration of radiotodine to arsure that such releases to the atmosphere will I l r.:t encoed calculated thyroid desas at the site boundary in excess l of 30 Res. Such a limit in the Technical Specifications combi.ned with the applicant's agreement to conform to Safety Guide No.11 i result in an acceptable solution to the instret line or process ij line break probles. f \\ ( i 9.9 Cryonenic Treatment subsys tem Accident The 61te boundary (suclusion area) exposure doses rmulting free t a complete instantaneous loss of the equilibrium fission product .i inventory of the cryogecic system has be calculated by the I applicant. The thyroid dose la meg!1gible and the whole-body dese was 0.113 Ram. These exposurse are well below the limits of 10 CFR Part 100. Exposure doses at the low populatiori sone distance would be less. Assumptions used were a release of fis-sion products at ground level in tha building wake where aseteorol-ogy conditions of Pasquill Type F diffusion and a 1 meter /second S +
- 10 3 - - wind speed occur No credit for dec ay wsw taken. We find these assumptions and the results to be acceptable. 10.0 COWDUCT or OrtRATIONS 10.1 Technical Ous11fications The philadelphia Electric Cosf any is responsible for the design, construction, and operation of the Limerick Genersting Station. The utility has acquired esperience in the design, construction, and operat.on of numerous fossil fuel power plants, hydroplanes, diesel-engine and gas-turbine unita, and other nuclear power plants. peach Botton Atomic Power Station Unit No.1, a gan-cooled nuclear facility constructed and operated by che applicant, continues to provide nuclear esperience for Philadelphia Electric Company personnel. the applicant is also constructing Peach Sotton Units 2 and 3; both units are boiling water reactors and similar to those proposed for Limerick. The applicat maintains an gagineering and Research Department staffed by several hundred engineers. As of June 1,1971, fif ty-j ] three engineers in Philadelphia Electric (bapany were assigned full or partial responsibility for the Lianerick project. This staff includes personnel who have acquired training and experience j i in many phases of nuclear projsets undertaken by the utility and j other organiastions. This technical staff is responalble for the f review and appr al of desiso features of the plant. This staff i j I { ........-__.,a.u.- --l+
l J - 104'- will prepare and conduct, with consultant aid, a quality assurance program and will follow field construction of tu plan: until its completics (see section 11 for details). He applicant has obtained the services of an Engineer-Constructor-(Bechtel Corporation) abo provides engineering and construction services and integrates the items supplied by the General Electric R Company with the items supplied by others for the remainder of the plane. The General 31ectric Csepany will supply the nuclear steam styply system, the soclear fuel, and technical personnel for guid-I asce and support of the applicant during start-up operations. Other se6centracters and consultants have been engaged to provide i ampertise la fields of a specialized nature such as meteorology, { j hydrology, seismology, geology, and envircassatal radiation c { i -( maattering. j pollovf og our review, vs conclude that the applicant md its i I= contractors are technically quellfied to design and construct the ) Limerick facility. \\ 10.2 Ormanisation of plant Maamement De responsibility for plant. operations is assigned to the Station Operating Department that has direct responsibility for operating all of philadelphia Electric's large power generating facilities. Se Superintendent at the plant has prime responsibility for safe and reliable plant operation. Organised under the Superintendent _________....-.m
- 105 - T e are two groups: the Operating Group and the Staff Croup. As their aanse imply, the Operating Group personnel are la attendance at the station seven days a week. 24 hours each day on a shif t baats. 1 A normal shif t will consist of eight men appropriately trained and experienced. Se staff Croep providae support in technical, main-tenance, and administrative duties. In defining plant personnel qualifications, the proposed " Standard for gelection and Trainias of Personnel for Weclear Power Plants (ANS-3)" was used as a guide. Organizational esperience is opera-ting Pead Botten Onit 1 with a plant organization stallar to the one proposed for Limerick provided addittomal smidance. We have compared the Limerick personnel geslifications with the ANS-3 requirements and comelude that the proposed personnel m11fice-tions for Limerick are acceptable. The Limerick superintend
- t. assistant superintendent, and other key peroommel will be assigned at least three years prior to initial fuel loading so that they may participate in the design and construction phases and the formal trainias program.
Other personnel to receive trainias will be assigned at least twenty -l l months prior to initic! fuel 1oading. The proposed plant personnel fl training program will follow the pattern previously used by Philadelphia Electric Company. A comprehensive and continuing training program and schedule has been adequately described. l \\ e 0% ..m I
- 106 - 10.3 Operatian Procedures Wittes operating procedures will be developed for the plant. Deviattoos from approved precedures will be permitted only in accordance with the procedures este11 abed ender che Adadaistrative Secties of the Technical Specificatises. All approved procedures i' will be available is the plast control rees and will inclede et l least: l Standsed Operettag Procedores Enerpsecy Procedures l Igdetenanos Praead== w Refeeling Procedures t Blocking Procedures Technical Specifications I 5ealth Physics Procedures e 10.4 Startup. Prooseratise. and puser Tests Plass for testing have been enlimented. The several difforent categories of tests are discossed la the following sesections. 10.4. 1 Construction and startue tests Bechtel Carperstiam and/or Osseral Electric Company personnel er their sabeestractors under technical direction of techtel or M [ will perform these tests, most of which require formalised pro-i ,) eedures, reports, and approvals. Philadelphia Electric Company will provide " surveillance" over these tests. Examples of these l tests aret containment final leak rate test, system hydrostatic test, calibration and setting initial trip set points on instro-f mento, and relief / safety volve adjustoent. '~ ~ im___
- 107 - 10.4.2 Freos.erational tests Included within this category are the construction and startup tests just discussed and also tests of plant systems such as plant electrical systems, instrument air systems, and makeup I unter supply systems. Such testing occurs before completico of construction and provides opportunity for traintag plant operators. The applicant has listed the major systems and indicated that de-tailed test procedures for each must be prepared by General Electric or sechtel for review and approval by the utility's management. I i 10.4.3 Startue and power test prosram some systems cammot be checked out during the earlier phases of J the test progran, but can be tested proper 2y only with the reactor I at power. Baclear characteristics of the zeectivity control system, therani and Wres11c characteristics of the primary
- system, and 1
~ eentros monitorias capabilities will be chect ed with the reactor j critical. The applicant's plant staff will perform these tests under the technical directime of a General Electric Company startup Specific, approved startup procedures will restrict testing I crew. to the objective of proving that the plant is capable of operating safely and satisfactorily up to rated power. i m b +w ,y
r-108 - 's Q 10.4.4 Conclurt ons Satisfactory planning for tests has been made. ue shall reviev .I .g tPe startup tests and planned startup personnel staffing at the i operating license stage in greater detail. { 'e 10.5 Emergency Planning and Plant Security 10.5.1 Emergency plans The applicant has used the guidance in Appendix E, " Emergency for. T4cilitics" 10 CTTs Tet t 50 ( in presenting a description of energency planning for the facility. 1 a The emergency plans have as "their primary objective the protec- ) f tion of the health and safety of the general public and of station pers onnel." Detailed emergency procedures impicmenting.the emer-gency plans will be written for specific energency situations dealing with a spectrum of accidenta, related both to radiation and non-radiation incidents. The applicant hat, begun plans in coordination with the responsible local, State, and Federal i g officials and will continue to consult with such officials during l 4 the development of the final emergency plan. Security measurt.s and actions associated with the p evention and control of civil (! t disorders and acts of sabotage are treated as non-radiation emer-gencies. We conclude that the applicant's emergency planning ~ 4 conforms satisfactorily with the guidance and criteria provided l [ d i i I 3 .-.-a i !
- 109 - i' L in ' Appendix E,10 CFR Part 50. Additional review of emergency pin.nning will be accomplished at the operathg If cense weary of the review process. Y I 10.5.2 Pl ant security. The applicant has included aspects of plant security or security 1 neuurea in the facility Deergency Plans. Security measures will include perimeter fencins, single point access mder centrol of a==curity gu.sd on 24-huur duty, u.= of Lodse. fus Mr cunni and locking of outside doors on buildings. The applicant will prepare detailed procedures that will be integrated into existing company emergency plans for plant security. We conclude that ade-quate description of security planning has been presented by the I applicant. Additional review will be accomplished at the opera- { ting license stage of the review process. i , ( 10.6 J_ndependent Safety Review of Operator Actions l ) All work in the operating plant will need to be approved by either f j i o the Station Superintendent or properly designated authority. Thero , 4, will be written operating procedures approved for plant use. A Plant Operation Review Committee and th Operating and Safety Review Committee are organizations at the plant and at the utility headquarters, respectively, to provide control over the initiation 4 of in-house tests, changes in operation procedures, advice on any e I l d i f
1;t. e - 110 - .. ? A u.wsual or unreviewed safety questien pursuant to 10 CTR 50.59 M and review of any violations of the Technical Specifiestions. f,[ N One member of the Operating and Safety Review C.iaittee will be M requir,ed to be a qualified consultant independent af PhilaNph;a 3 j j-. Electric Company. 9 The other mer?rars of this committee will be engloyees of the utility but not menbers of the Limerick Cenera-ting Station staff and will be qualified per section 4.6 of the 4 3 prepeced " Standard for Scicctica sad Tr&ining vf Toiswencifor Nuclear Power Plants." These committees vill meet to oversee objsetively, or. a scheduled basis and as needed, the performance of the plant personnel. We conclude that the plans for objective M I} review of operater actions are acceptable. 11.0 QUALITY ASSURANCE (QA) The Philadelphia Electric Company described their Quality Assuranc (QA) Program in Appendix D of the t.fearick PSAR. Our review resulted in requests for additional information that have be n r7 U e documented. This added information coverad a broad band of topics 4 within the Quality Assurance Program and generally responded s adequately to our questions. The applicant's orge' ' tion fo r the Limerick facility includes a Philadelphia Electrh Manager QA Project, who is responsible for coordinating the QA Prograa and is authorized to stop construction work when necessary to assure good construction practice. PEco will have QA Engineers l-4 4, at the construction site to audit cone'.ruction and installation z 9 e - - - ~.. d
- 111 - 93 \\ vork; these onsite auditors will rcpert directly to the PE P.anager, QA Project. PECo has retained HPR Assoelates to assist . ty PEco in monitoring the QA Frogram. P,; Se quality control progrens of component manufacturers , subcon-trectors, and Bechtel Corporation (the architect-engineer fira) are normally under the audit / surveillance of the applicant's con-f tractor group responsible for that portion of plant design His audit is in addition to the QA program of the individual manufte- .g turer or site constructor. General Riectric Con;,any providas the Nuclear Steam Supply Sy (17535) mod will be responsible for quality catrol requirements 3-for this equipment to include: manufacture; site receipt, in-i* spection and storage; ano site installation for the "assential" components under its scope of supply. In Appendix D of the PSAP., the CE quality assurance program is describeJ further. His pro-describM is consistent with our requirements and is gram ab accept able. 4 [ The architect-engineer firm, Bechtel Corporation, is the plant design contractor, the procurer of all non-NSSS equipment ,and the responsible onsite contractor for construction of the Li merick facility. Bechtel Corporation has provided a quality assurance c program covering these responsibilities in accordance with our requirements. This program is described in Appendix D to the PSAR ] 1 i en
1: 1 - 112 - 9A,3
- Ge i.
'f. AEC th; ectors fr ca kgion I, Division of Compliard.e. nsde en st inspection at the applicant's off'ce in Decer.ber 1970 to deter- )[ t ,a mine the adequacy of 1*T ementation of quality assurant.e criteria l in Appendix B,10 CFE Part 50. The Cong11suce inspectors found ~ ~ at that time that the applicant's implementation of its Quality l Assurance Program was not satisfactory. As a consequesa of that .I 1 3 inspection, the applicant made constitwats for corrective action, j re:;':erted a re-inspectice, end engaged M'R Aaanef atea en advia* thes in the organization, writing, sad implementing the details a-of their Quality Assurara Program. o tr* l since the December 1970 Compliance inspection, the applicant has j reorganized the QA personnel and prepared additional i. documentation $ [ of procedures for QA actions. The requested QA reinspection by Region I, Division of Compliance, was made in April 1971 and fi-resulted in an overall finding of satisfactory response to the y)3 ( guidelines set forth in the above-mentioned documents. In our Y view, the applicant's current Ruality Assurance Program is acceptable. i 12.0 TEQiNICA1, SPECIFICffEONS 1 0 Dur rev'tv of the applicant's proposed Technical Specifications vas lir.ited to the coverage and depth required at this stage of the licensing process (10 CFR 50.34). We had noted the lack of: a section on Administrative Controls; a listing of important } l \\
- 113 -'. m. Q i [ MAL and/or uncersm verds 'or phrases with their definitions; a plan for including details on the plant " Inservice Inspection Prgcri"; 3 a plan for describing the cwcept and procedures for a Site j Eny' utal Menitoring Program; and a section which included h Id s* tier. otditions for operation and surveillance requirements 1 i we physics or reactivity effects. The applicant has recti-IW *.:ese omissions by atating plans in each of these areas. Per instance, the applicant's Inservice Inspection Program will I be covered in the Tacimical Specifications by listing components and ports to be examined, the method of examination, and the fre- ~ quency with whidt the examination will bc conducted. The Technital Specifications vill ha reviewsd in dcpth at the operatir.g lica:.se i y stage of the review. ? 13.0 BPORT OF 111E ADVISORY COMGTTEE ON REAcr0R S Folleving its 136th meeting on August 5-7, 1971, the Advisory N Commitcee cn Reactor Safeguards (ACRS) reported on the rasulta of its construction permit review of the Linarick Generating I 5.:stion in a le*ter dated August 5.1971. A copy of this letter ( is atta-hed as Appendix 8. S We have discussed its content with the applicant. Based upon our discussions and our review of Amendments to the PSAR submitted subsequent to the ACRS letter's.] receipt, we consider that the appli: ant is being responsive to the recommendations presented in the ACRS lectar. Most of ths l}," items in tha 4CRS Ictter ( including the reconnendations, have been ?j i r 1 m. .n _a ^
o I, ? - 114 - t i SI ( discussed elsewhare in this safety evaluation. However, a brief [ presentation on these items including references to the pertinent h ', sect 'ons of this evaluatien, is given below: ,.O 13.1 Process Line Breaks (section 9.8): 1he ACRS mentioned the ap;11 cant's concept for using bicw-out panels that relieve the Reactor Building at the internal air t pressure cf 7 inches of water to protect engineered safety equip-1
- V g
mene from sacessive steam wile swintaining mapooute doemas *pcIuw 10 CFR Part 100 guides. E 13.2 class I (seismic) Main Steaaline (EL) Desians (section 4.2.5) The criteria for design of the EL out to and including the tarbine i stop valve and other appurtenances to the ML's was set forth. The 1 l -4 applicant has accepted and is implementing the criteria. .1 i 13.3 EL Isolatica valve (EL-IV) Leahme: (sections 4.2.4 and 9.7) The ACRS mentioned the applicant's couaiteent to providing for 3 the seal of ML-IV's folicwing a loss-of-coolant accident. 13.4 Reactor W.ields (section 5.1.1) The ACRS stated details of design and construction of the shield. f 1 The use of magnetite concrete between the circular steel plates, reinforcing near openings, and a capability for the entire shield to withstand internal pressure ad jet forces were mentioned. The applicant has adopted these and other conservatism in the shield design in a satisfactory manner. ] I i ... ~ _ . ~.......... 3y. 1 ~
i - 11S - l e 13.5 Emeraency Core CoolinfmSys em: (section 6.1) The ACPS tentioned the .icant's conwitrent to providing improved HPCI and LPCI systema to provide increcsed system reliability and lowered peak fuel' clad temperatures. Details of these improvements will be reviewed at the FSAR state of the i e review. { 13.6 Radioactive Waste Disposals (section 8.2) The ACPS steted that the applicant had provided radionettve waste i i disposal systems that include featuret beyond th we normally provided in befling water reactor plants. We # ' continue to follow the-developaset, during the constructive pariod, of the radioactive vasta processing and dispossi systus for the facility. 1 13.7 ht?'cipated Transiento Without Seras (A'gisl: The ACRS took note of the applicant's acceptance of the CE concept i for providing ultimate negative reactivity to the core for reactor li f, ' shutdcwn and control after an ATWS event (discuased in General 1 Electric Company Topical Report NEDO-10349, March 1971 - ATVS). i9 Tha CE concept involves automatic tripping of the two recircula-lj tion pumps on coincidence signals of high neutron flux and high Ll (; l ( reactor pressure, followed by injection of a liquid poison into the core. This enneept is acceptable to the Regulatory Staff. j The ACRS recommended further evaluation of the sufficiency of d[i; this approach and the specific means of implementing the pump i trips. The ACKS recommended resolution of this item with the 4l 1 1 f !f .. ~.... _ ~.. - -. = l
h '1.16 - Fe;;ulatory Staf f during the construction period. We will fully I rese!ve this itta prior to issuante of an operating license. .g 13.8 Hydrogen Conrrol and Containment Inertina: (sections 6.2.1 f, 6.2.2) - While.the ACRS noted the applicant's commitment to provide a system for control of hydrogen in the ' primary containment following the unlikely event. of a loss-of-coolant accident, the Committee further ~ i f commented on the need to inert the atmosphere of the primary. containwnt to meet the current AEC criteria in Safety Guide 7. I' I Re applicant has made commitments to meet the criteria of Safety - 2 Cuide 7 and also to provide an inerted atmosphere in the primary containment. We will follow the appif cant's development of the 4 4erogen control system during the construction period and will, /5 .f n during the review of technical specifications, est slish operating ? / j limits on herting. Q 13.9 Maximum Cround Acceleraticut (section 2.4) 1 g he ACRS recommended that added construatism be applied in seismic design by increasing; the DBE saximum horizontal ground acceleration to 0.15s. ne applicant has accepted and is designing Class I (seismic) systems or components with this new value. s il 14.0 CON.r0RJWLCE TO GENE 8tAL DESIGN CRITERIA i Based upon our evaluation of the preliminary design and the design 3 J criteria for the 1.imerick Cenerating Station, ifnits 1 and 2, we ! t [ +.ep.epassg e-+ e ' mes., +.e.,g + 4
- 4 4 e.we.6 6
- dlMeim**N
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Mk 1Q .j -11i- .+ h 'c.< ' conclude that the applicant will provide a final f acility dealgn that will meet the latest AEC Ceneral Design Criterie 4 15.0 COMON DEFENSE ANDJ ECURIQ l %e applicant states that the eetivities to be conducted would be 6 jj-within the jurisdiction of the United States and that all of the directors and principal officers of the applicant are United Statt l citi:en:. h'c find nousing in the applica6 ion to suggest ta'st ene applicant is owned, controlled or dominated by an alien, a fore corporation or a foreign government. The act:.vities to be conducted f# do not involve any restricted date, but the applicant has agreed safoguard any such data wttich signt become involved in accordan j with the regulations. Thc applicant will obtain fuel as it is needed l from sources of supply available for civilit., purpeses, so that no r y diversion of specini nuclear material frcr military purposes is 1 ( involved. For these reasons, and in the absence of any information 5-to the contrary, we have fcund that the activities to be perfo ~ will not be inimical to the common defense and security. f. r aed E h 16.0 FINANCIAL QUALIFICATIONS I j I We Comission's regulations which relate to financial data rd .l F information required to catablish financial qualifications for a rt applicant for a facility construction permit are 10 CFR 50 33(f) { and 10 CPR 50, Appendix C. J J j __ -. - * - - ~ " ' ' ' '
g, EM ,.g 118 - US L?: f We have reviewed the financial information presented 4 the ' j. application and Amendment No. 8 thereto of Philadelphia Electric Company to construct the Limerick Generating Station, Units 1 l and 2. Based on this review, we have concluded ' hat Philadelphia t Electric Company la financially qualified to design and construct these proposed facilities. Our conclusion is based on the following facts and considerations: ihe applicant estimated the costs of construction of the a. plant, including transmission facilities and other associated costs, and initial reactor cores will total about $818.0 million. The details of these estientes are contained in the application and are summarized below !j. Total ~ nuclear plant capital. costs $717.2 million Transmission facilities and other associa tad. costa 4.8 Inventory costs of initial cores for y both units 96.0 j Total 1818.0 million M We have reviewed the detafis of the estimated plant capital ' k'. }. costs for constructf.on and have found them to be reasonable t e u. The applicant estimates its total construction expenditures for the five year period 1971-1975 will amount to ap;coxima ly i $2,26'5.8 million. ib I (
M - 119 - l +I I: 6 .t fi 4 b. Construction of the nuclest plant will be financed by {'
- y Philadelphia Electric Company in the normal course of j
~f finar,cing its plant construction prc. gram. Such financing is ( obtained' from internally generated funds, principally unappropriated earnings and provision for depreciation, from 1 i sale of debt and/or equity securities, and frota short-ters loans needed to meet requirements on a temporary basis. 2 Based on the applicant's record of earnings and provision for I
- b depreciation and other accruals over the past six years and on I
the resaonable assua7 tion of the continuation of these earning 3 levels ever the years 1971 through 1976, it is reasonshie to i 4 4 except that internal sources can provide, af ter dividends, $60 I to $70 million per year, or's total ranging between $360 and $400 million for the six-year period. In view of the applicant's j resources, the strength of its financial position, the high regard held for its bond issueo, and its proven ability to 'm /RK 'ji borrow on a short-term basis, it is our opinion that issuance i ;. ; of securities can he relied upon with reasonable assurance to ] supply the balance of funds required to finance the design and I j construction c: the timerick units as well as additional i J generating capacity. 4 Philadelphia Electric Co. is soundly financed and has c. i significant resources at its commanu. .J of December 31, 1970, i,k I l l Y j- 'j g l 4
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- t I
l - 120 - t o; 4$ i. e:.y p ! cash and net receivables totaled $75.0 million. Los;-term debt represented. 54.3% of the total capitnif ration and 51.7% Ih[ of the net investment in utility plant. The applicant's Dun k L FC f and Bradstreet credit rating is 5A1 (the highest category) [ ] r ind Moody's Investors Service rates the company's first ) Lr ; mortgage bonds as Asa (gilt-edge). A i> i Operatis.g revenue of $504.4 million for 1970 ves up 48% over i ff 1965, and net income, after taxes, of $68.4 million was up ) g 21% over 1965. The volume of electric energy sales over the g same six years has increased 36% to 22,813 million kilowatt h 4 hours in 1970. The pertinent financial rat:,no indicate a sound financial position, and these are in line with ratios 1 i of the electric utility industry as a whole. A summary saalysis I reflecting the ratios and other pertinent data for the applicant is attached as Appendix I. In brief, these ratios I{ ~i 7 as of December 31,1970 are: long-term debt to not utility } plant .52; net plant to capitalization - 1.05; proprietary I ratio .41; operating ratio .79; rate of gross earnings l y on total ir. vestment - 6.07%; rate of earninge on stockholders' l } 5. equity - 8.0%; times interest earnad on long-t -tm debt - 2.5; . hy and retained earnings - $239.5 million, d h 17.0 (_ DNCLUSIONS Based on the proposed design of the Limerick Generating Station; t on the criteria, principles, and design arrangements for the )#e k - j
's j. ./l - 121 - l ]i Qi .4 aT systoms and components thus far described, which include all 3;- .q ji-of the irportaat safety items and 5 n the calculated potential consequences of routine and accidental release of radioactive [ materials to the environs; on the scope of the develo;nent si programs which will be conducted; and ce the technical competence of the applicant and the principal contractors; we have concluded k. .[~ that, la accordance with the provisions of paragraph 50.35(a), 10 CFR part 50 and parasraph 2.104(b) 10 CFR Part 2: o 1. The applicant has described the proposed design of the facility, including the principal architectural and engineering criteria ( wt for the design and has identified the major features or ,hy components for the protection of the health and safety of s the public. l 2. Such further technical or design information as may be required l
- ?l to complete the safety analysis and which can reasonably be j
- r" left for later considerations, will be supplied in the final
) Jkl ' '] safety analysis report. a I 3. Safety features or components requiring development have been .p identified by the applicant. The applicant's development pro-t k gree will be conducted to resolve any wafety questions associ-d i i 3k b I i ated with such features or components. 4 j i ,k J i \\y s c IM l,hf , 'N 1 f Li. ( _ j El fj
I - 122 - i 1I 1. On the basis of the foregoing, there is reasc.aable assuranca f that (i) such safety questions vill be satisfactvily resolved at or before the latest date stated in the appli-cation for completion of construction of the proposed facility and (ii) taking into consideration the site criteria 8 contained in 10 CFR Part 100, the proposed facility can be constructed..! operated at the proposed location without 0 edue risk tc the health and safety of the public. i De applicant is fin.mcially and technically qualified to design and construct the proposed facility. 6. Se issuance of perndts for the construction of the facility will not be inimical to the common defense and security or f to the health and safety of the pelic. l t 1 j l I i _______-__________.m._--
J 123 3-t wMy APPF.NDIX A f;.L ! @'~ OTRONOLOGY OF EVENTS ~ f 1. February 26, 1970 Submittal of Preliminary Safety Analysis M, i and. License Application .dl 2. Hay 13,1970 ] Initial meeting with applicant to discuss review schedule and areas for review 3. Jme 10,1970 Meeting with applicant to discuss eite ralated matters, radioactive vaste control ? and disposal, and radiation shielding f 4. June 19,1970 RMm6 tor Environmou;.:' Feport from applicant 5. June 25,1970 Request for additional infomation from. applicant 6. Jme 30-Jul 1,1970 Meetine with applicant te discuss reactor fuel and vassal internals, primary coolant i Q syste=, containment, CSCS, and safety analysis 7. August 6,1970 Request for additional information from W applicant I 8. August 19, 1970 Meeting vich applicant to discuss struc-tural analysis including seismic design ,;M and forces due to IACA, vinds, tornadoes, missiles, and thermal stresses l { 9. Septanbar 15, 1970 Request for additional information from applicant
- 10. Septed er 23, 1970 Meeting with applicant to discuss pre-liminary technical specifications, conduct of operation, emergency planning, and instrienentation and control
'4
- 11. October 6,1970 Submittal of Supplement No. I to PSAR
- 12. October 10, 1970 Request for additional information from applicar,t k
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h - ID - L -13. Oc tober 13, 1970 Request for additional information fne [ applicant i 14. October 30, 1970 Subnittal t,f Environceatal Report by . applicant _~ 5. Novaber 10, 1970 ACRS Subcoenittee site visit
- 16. November 24, 1970 Submittal of Supplement No. 2 to PSAR t
- 17. Deceeer 2,1970 Request for additional inic,:mation froa app 11 cant 18.
Decad er 7, 1970 S:tmittal of a request for construction omenytton
- 19. December 11, 1970 Submittal of Supplement No. 3 to PSAR 2's, January 14-15, 1971 Meeting with applicant to discuss response to varied questions as presented in Supplements 1, 2, and 3 4
21. February 8,1971 Request for additional information from applicant
- 22. February 24, 1971 submittal' of Supplement No. 4 to ?SAR i
- 23. March 2,1971 Meeting with applicant to diseuas analytical model for calculation of poet-LOCA containment pressure transient 24.
March 11, 1971 Request for additional information from applicant 25. Ma. ch 12,1971 Request for additional information from applicant ) 26. t'.a rch.%, 19 71 Mee'.ing with applicant to discuss aseuda iters prior to ACRS subcommittee meeting 27. March 31,1971 Meeting of applicant and DRL staff with the ACRS subcommittee for Limerick Generating Station. i
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125 - '! 8 y# @s
- 28. April 7,1971 Submittal of informtion for anti-trust
' j j g-review by Depart.wnt of Justice c { l Ef
- 29. April 30,1971 Submittal of Supple: neat No. 5 to PSAR 4]
j q a
- 30. May 6,1971 ACRS meeting with DRL J
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- 31. May 26-27,1971 Meeting with applicant to discuss l
queations on a variety of matters yj ?
- 32. June 11,1971 Semittal of Suppleant No. 6 to PSAR
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- 33. June 15,1971 Sthmittal of hsendment No. 7 to PSAR 2 l
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- 34. June 17,1971 Request for additional inforantion from ji applicant (update of financial status g
and data)
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- 35. June 18, 1971 Meeting with applicast to discuss set. ale L.I design, selected structural design require-4 ments, and status of review p$
- 36. July 16, 1971 Submittal of Amendment No. 8 co PSAR
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- 17. July 18, 1971 Submittal of Amendment No. 9 to PSAR i
- 38. July 28,1971 Meeting with applicant to discuss agenda
- J items for ACRS subcommittee 39.
Tiily 25,1971 Submittal of information ce site k., 9J construction plans j
- 40. July 29,1971 ACRS subcommit tee meeting to perforp general review and discuss outstanding 9
items prior to fill committee meeting N
- 41. July 30,1971
.j Submittal of Amendment No.10 i i
- 42. July 30,1971 Gubmittal of ECCS evaluation (included later in Amendment No.11) 1
- 43. August 5, 1971 ACRS (Full Consmittee) meeting f
I
- 44. Augaat 10, 1971 ACRS letter to AEC, report on Limerick
-3 i } .s
- ,9 w
w t r 1 1 fa L
.o - 126 - 9 i i
- 45. August 10, 1971
- beting with applicmt to diacuas implementativa of ACRS reco.wendations
- 46. september 16, 1971 Semittal of Amendment No.11 to PSAR
- 47. september 22, 1971 54mittal of Amendment No.12 to PSAR i
- 48. October 8,1971 84mittal of Amendment No.13 to PSAR l
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