ML20235C658
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{{#Wiki_filter:. _ _ - - _ ffy'(( fIf.S$u?- * '$b ne 1 e nn ,,,,'i[*.*f hYJs. e.an SAFETY EVALUATION BY THE DIVISION OF REACTOR LICENS1NC U. S. ATOKIC ENER0Y COMf!SSION IN THE MATTER OF THE CINCINNATI GAS AND E1.ECTRIC COMPANT WILLIAM N. EU9tER NUCLEAR POWER STATION i UNIT WO. I DOCKET No. 50-358 3 l i l t l 8709240526 870921 PDR FOIA MENZ87-111 PDR' i.I L#D
F1 r-i \\ lj TABLE OF CONTENTS i 1 ,[ T ab l e o f Con t e n t s.................................................. i Abbreviations...................................................... il 1.0 INTRODU CT I ON............................................... I e 2.0 SITE AND ENVIRONMENT....................................... 7 2.1 Coog raphy and Demography.............................. 7 2.2 Meteorology........................................... 10 2.3 Nydro1ogy............................................. Il 2.4 Coology, Seismology, and Soil Mechani cs............... 15 2.5 Environmental Radiation Monitoring and Ecology........ 19 2.6 Railroad, Rive r, and Air Traf fic...................... 22 2.7 Con c l us i ons........................................... 23 3.0 FACI LI TY D ES I GN............................................ 23 3.1. Reactor Design........................................ 23 3.1.1 Genera 1........................................ 23 3.1.2 Nuclear Design................................. 26 3.1.3 Tho rna l and Hydraulic Design................... 28 3.1.4 Re a c t o r Ls t e rn s 1 s.............................. 30 3.2 Re a c t o r............................................... 36 3.2.1 Sys tem Quali ty Croup Classificat ion............ 36 3.2.2 Reactor Coolan t Pressure Bounda ry.............. 38 3.2.3 Reactor Core Support St ruc tures-Des ign......... 47 3.2.4 Fra c t u re Toughnes s Cri te ri a.................... 47 3.2.5 Reactor Vessel Materials Surstillance Program.. 49 3.2.6 Le ak De t e c t i on................................. 49 3.2.7 I ns e rv i c e In s pe c t i on........................... 50 3.2.8 Reactor Coolant System Sensitized Stainless Stee1........................................ 51 3.2.9 Foreign Procurement............................ 52
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l t I i 11 3.3 Containment........................................... 53 3.3.1 Ce n e ral Con t ai nme n t De s i gn..................... 53 3.3.2 P ri ma ry con t a i nme n t............................ 53 3.3.3 Se conda ry Con t ainme n t.......................... 60 I i 3.4 Enginee re d S a f e ty Fea t ures............................ 63 e 3.4.1 Core S t andby Cooling Sys t em.................... 63 1 3.4.2 Res idual Heat Removal Sys tem (RHRS)............ 80 l 3.4.3 Post-LOCA Hydrogen Control and Containment Inerting..................................... 83 l 3.4.4 Long Te rm Cooling Wate r Supply................. 84 l 3.5 Protection, Control, and Emergency Electric Power Systems............................................. 85 1 3.5.1 cenera1........................................ 85 3.5.2 Prote ct ion Sys t ems Gene ric Items............... 87 3.5.3 New De s i gn I t e ms............................... 90 3.5.4 Electric Power Systems......................... 94 3.6 Radwaste system....................................... 99 3.6.1 Genera 1........................................ 99 3.6.2 Liquid Wastes.................................. 99 3.6.3 Ca s eo us Wa s t e s................................. 102 3.6.4 Solid Wastes................................... 104 3.6.5 Con c l us i on s.....................................105 3.7 Aux i l i a ry S y s t ems..................................... 105 3.7.1 Ge n e r a 1........................................ 105 3.7.2 New ruel Storage............................... 106 3.7.3 Spent Fuel Storage............................. 106 3.7.4 Service Water System........................... 108 3.8 Station Structures and Shielding...................... 110 3.8.1 Classification of St ructures and Equipment..... 110
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3.8.2 Structural Analysis and Design................. 111 3.8.3 Hechanical Analysis and Design................. 116 3.8.4 Se i s m i c Qu a l i ty As s uran c e...................... 118 l ) = =.
Page 4.0 PLANT SAFETY ANALYSIS...................................... 119 4.1 Genera1............................................... 119 4.2 Los s-o f-Cool an t Accide nt....................... 120 4.3 Fuel Han dling Accide n t............................... 121 4.4 Con t rol Rod D rop Ac ciden t............................ 4.5 123 Main Steam Line Break Accident. 4.6 Instrument Line or Process Lin 127 Conc 1 us f orm.................. e B re ak............... 4.7 128 129 5.0 QUALITY ASSURANCE.......................................... 129 6.0 TECHNICAL QUALIFICATIONS AND CONDUCT OF OPERATIONS... 132 7.0 EMERCENCY PLANNINC......................................... 136 i 8.0 ~ CONFORMANCE TO CENERAL DESIGN CRITERIA..................... 1 137 l 9.0 REPORT OF THE ADVISORY Cottf!TTEE ON REACTOR SAFECUARDS 138 l { 10.0 COPHON DEFENS E AND S ECURITY........................... 140 11.0 FINANCI AL QUALI FICATIONS................................. 3 141 12.0 CON CLUS I ONS............................................... 143 l i e l I LIST OF TABLES i 1 3.1.1 Major Design changes Incorporated into the 1 Wm. H. Z imme r S t a t i on.................................... i 3.1.2 24 Compa ris on o f BWR Design Pa rame te rs........................ l 3.1.3 25 Thermal an'd Hydraulic and Nuclear BWR Desi l AEC Code Class ification................... gn Parame te rs.... 3.2.1 29 1 3.3.1 37 Compa ris on of Containment Design Pa rame ters................ I, 54 1 [ l l l u. i.{ h t
ABBREVIATIONS .ACI American Concrete Institute ACRS Advisory Connaittee on Reactor Safeguards ADS Automatic Depressurization System AEC United States Atomic Energy Coussission ANC Aerojet Nuclear Corporation ANS American Nuclear Society ANSI American National Standard Institt.te APED Atomic Power Equipment Department (GE) ASCE Americar. Society of Chemical Engitseers ASME American Society of Mechanical Engineers ASTM American Society for TestinE and Materials ATWS Anticipated Transient Without Scras 2 BTU /hr-ft British Thermal Units per hour per square foot BWR Boiling Water Reactor CFM-CFS Cubic feet per minute or per second Ci/sec, cc Curies per second or per cubic centimeter s CSCS Core Standby Cooling System CSS Core Spray System CVG Creater Cincinnati Airport DBA Design Basis Accident DBE Design Basis Earthquake d-c direct current i 4
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a a { DRL Division of Reactor Licensing DWT Dropweight Test ECCS Emergency Core Cooling System FLECHT Full Length Emergency Cooling Heat Transfer Program ft feet FSAR Fina fety Analysis Report = 3 acceler.t.fon, 32.2 feet per second per second GDC AEC Ceneral Design Criteria For Nuclear Power Plant l Construction Permits CE Ceneral Electric Company i spa gallons per minute HEPA High Efficiency, Particulate, Air NPCI High Pressure Coolant Injection System EPCS High Pressure Core Spray System IEEE Institute of Electrical and Electronics Engineers in inch 4 kips thousand pounds k effective multipliestion factor (for the nuclear 'If fission process) ak/k/*F reactivity (ak/k) change per degree Fahrenheit = kV kilovolts kW/ft kilowatts per foot lb pound IDCA Loss-of-Coolant Accident 1 j. I i i 4
Loss-of-Fluid Test Low Pressure Coolant Injection System Low Pressure Core Spray . Low Populstion Zone Minimum Critical Heat Flux Ratio Mean Sea TAvel Main Steam Line Main Steam Line Isolation Valve million gallons per day assimum permissible concentration - miles per hour 'l megawatt days per ton g messwatts, electrical megawatts, thermal Nil Ductility Transition National Desanic and Atmospheric Administration not positive suction head Nuclear Steam Supply System Operating Basis Earthquake Probable Maximum Flood Parts per million Preliminary Safety Analysis Report pounds per square inch pounds per square inch gauge i
QA Quality Assurance QC Quality Control R&D Research and Development RCICS Reactor Core Isolation Cooling System RERS Residual Heat Removal System SDB Safety Design Basis SCTb Standby Cas Treatment System SWS Service Water System USCS United States Geological Survey w/o we1Rht percent uti/mi sterocurie per milliliter pC1/see microcurie per second 10 CFR AEC. Title 10. Code of Pederal Regulations Part 2 AEC Rules of Practice Part 20 AEC Standards for Protectica Against Radiation Part 50 AEC Licensing of Production and Utilization Facilities Part 100 AEC Reactor Site Criteria 4 9 9 s 9 l a I
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1.0 INTRODUCTION
h= This report is the Atomic Energy Commaission's (the Commission) [- safety evaluation of the application by The Cincinnati Cas and h Electric Company (CC&E), Columbus and Southern Ohio Electric R Company (C&SOE), and The Dayton Power and Light Company (DPL) lF (the applicants) for licenses required for the construction and gt lg,g operation of the proposed Wm. H. 21sumer Nuclear Power Station L (limer station, plant, facility). The Zimmer station will [ consist of a single-cycle, forced circulation, boiling water reactor unit, constructed near Moscow, Clermont County, Ohio, 25 alles southeast of Cincinnati, on the Ohio River. The Cincinnat.1 Cas and Electric Company is responsible for the design, construction, j and operatico of the Zimmer Plant and is also authorised to act as agent for C&SOE and DPL in all details of construction, including [ licensing. E The Zimuner facility will be designed for an initial power level 5 g of 2436 W thermal and a net electrical output of 807 W. All safety sy.tcas and analyses have been evaluated at the design f power level of 2540 W thermal and 840 W electric. Sargent and Lundy will perforin the architectural engineering services. The [ Ceneral Electric Company will design, fabricate, and deliver the [ single-cycle, boiling water nuclear steam supply system (NSSS), j E and will also fabricate the first core of nuclear fuel. The I )
construction contractor will be Kaiser Engineering Inc. The turbine generator unit will be supplied by the Westinghouse Electric Corporation. A single hyperbolic, natural draft, cooling tower l will be used to dissipate waste heat to the atmosphere. Cooling l water for the service water system and askeup for the cooling tower is drawn from the Ohio River. All return water is routed to a settling pond prior to its return to the river. The design of the Willista H. Zinser Nuclear Power Station is similar to boiling water reactors previously approved for con-struction. A comparison of some of the design characteristics of the Zimmer facility with those of similar facilities is afven in Section 3.0. The secondary containment, the reactor building, will provide an additional barrier to prevent the release of airborne radioactive materials, substantially the same as for other BWR stations. The primary containment design is similar to that of Shoreham and Limerick Nuclear Power Stations wherein the "over-under" vapor suppression containment concept is used. The containment differs in that the drywell and wetwell will be constructed from pre-stressed concrete. The gaseous and liquid radwaste system will be designed to limit the processed effluents to an annual average concentration for i routine discharges of less than 1% of 10 CFR Part 20 limits. L
1 ) l 1 I Our technical safety review and evaluation of the facility have been based on the applicant's Preliminary Safety Analysis Report i l (PSAR) and subsequent amendments thereto, During our review of the application, we held meetings with representatives of the appli-l i cants and their consultants to discuss the facility and 'the techni-cal material submitted. A chronological listing of the meetings and other significant events is given in Appendix A to this { evaluation. As a result of our review, we required a number of l changes to be made in the facility design. These changes are f described in the applicants' 21 amendments to the PSAR and are 5 discussed in appropriate sections in this report. All of these o documents are available for public inspection at the U.S. Atomic Energy coussission's Public Document Room,1717 N Street, N.W., Washington, D.C. and at the Clermont Cosmty Library, Batavia, Chio. Our technical evaluation of the preliminary design of the facility was accomplished with the assistance of consultants. The reports of our consultants on meteorology geology, seis-i {i l aclogy, fish and wildlife, and structural design are included an appendices to this report.
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The Atomic Energy Commission's Advisory Connaittee on Reactor n Safeguards (ACRS) has also reviewed the application. An ACRS j subconsnittee with representatives of our staff made a site visit i, ,li l i ~
1 Li . on August 27, 1971. Subsequent ACRS subcoussittee and full commit-tee meetings culminated in the ACRS report to the Commission that is included as Appendix B. This safety evaluation suussarises the results of the technical evaluation of the application for a construction permit. Princi-pally, the following matters were included in the reviews a. We reviewed the population distribution and use characterts-tics of the site environs, and the physical characteristics of the site including seismology, meteorology, geology, and hydrology to determine that these characteristics had been given proper consideration in facility design and that these characteristics conformed to the Commission's siting criteria (10 CFR Part 100) taking into acccunt both engineered safety features and site characteristics, b. We reviewed the applicants' preliminary facility design and planning for construction, fabrication, testing, surveillance, and expected performance of facility structures, systems, and components important to safety to assure that they will con-form with the Coassission's General Design Criteria (Appendix A, 10 CPR 50), appropriate codes and standards, and the Coarnission's Quality Assurance Criteria. e_. _ + _ _ _ _ - _ _ _ _ _ _ _ _ _. _ _ _ _ _ _. _ _ _ _ _ _ _. _ _... - _
[} 'l -~5 - 1 We evaluated the expected response of the facility to the c. various anticipated operating transients and to a broad spec-trum of postulated accidents, and determined that the poten-tial consequences of a few highly unlikely postulated accidents (design basis accidents) would.szceed those of all other accidents considered. We performed conservative analyses of these design basis accidents to determine that the calculated potential offsite exposure doses, that might result in the very unlikely event of their occurrence, j would not exceed the Commission's guidelines for site accept-5 ability given in 10 CFR Part 100. l l d. To determine that the applicants and their contractors are properly preparing for continuing safe operation of a completed facility, we evaluated the preliminary planning for conduct + of facility operation, the proposed organiastional structure, the concept for achieving industrial security, and the plan-ning for emergency action to protect the general public in the unlikely event of an accident requiring such action. We also evaluated the applicants' stated technical qualifications for operating and technical support personnel. I
_ l ) e. We evaluated the preliminary design of systems provided for 1 control of the radioactive effluents from the facility to 1 . determine that these systems can control the release of radio-active wastes within the limits of the Commission's regula-tions (10 CFR Part 20) and that the applicanta plan to operate the facility in such a manner as to reduce radioactive releases to levels that are in the order of 1% of 10 CFR Part 20. Many features of this facility will be similar to those we have already evaluated and approved for other reactors now under con-struction or in operation. To the extent feasible and appropriate, we have made use of our earlier evaluations to facilitate eepedi-tious review of features substantially the same as those considered previously. Whenever this has been done, the appropriate section of this report identifies the facilities involved. Our safety evaluations of those other facilities are available for public inspection at the Coussission's Public Document Room. The review and evaluation of the proposed facility at the con-Y struction peref t stage is only the first rtage of the continuing review of the design, construction, and operation of the facility. Prior to the issuance of an operating license, we. will review the final design to determine if all the Comission's safety re-quirements have been met. The facility would then be operated o
m ~ q,. L i-only in accordance with the terms of the operating license, the Coussission's regulations, and under the continued surveillance of the coussission's s egulatory staff. Based on our evaluation of the application to construct Unir No.1 of the William H. 21samer Nuclear Power Station so presented in the subsequent sections of this report, we have concluded that the applicants can construct and operate the proposed facility without endangering the health and safety of the public. 2.0 SITE AND ENVIRONMDrT 2.1 Ceostraphy and Demography 6 r i The William H. Eineser nuclear power station site (the site) la i r situated on a 635 acre tract of land in Washington Township, Claremont County, Chio, and is located approximately 24 miles r southeast of Cincinnati, Ohio and 1/2 mile north of Moscow, Chio, on the Ohio River. The minimum exclusion distance will be 380 r-1 E. meters and the established low population zone (LPZ) distance = [ is 4,827 meters. There are no missile sites within 300 miles of E-the site. l The topography in the western portion of the site is relatively f level at about 500 feet above mean ses level (MSL) and is primarily 5 farm land. The remainder of the site is hilly and partially wooded. i k w
- with the highest eleva-lon running up to 800 feet MSL. A r s11 stream with a bed elevation of 470 feet NFL runs east to west i
through t.e site to the river. The reactor plant grade will be at 520 feet MSL. The nearest population center with a population in excess of s 25,000 people is Covington, Kentucky which is located 20 miles northwest of the site. New Richmond, Ohio, with a 1970 population of 2,650 (projected to be 2,400 by 1985) is within 10 miles of the site. The cumulative population distribution in this same ares (0-10 mile radius) is very low, with 23,000 persons in the 1970 census and is projected to be 24,100 by the year 1985. The population distribution within the LPZ for the 1970 and 1985 (estimated) census is less than 1,700 and 1,800, respectively. 8 One third of the LPZ population lives in %> scow, Ohio. The 1970 census inf ormation provided by amendment No. 21 indicates that population centers within 25 miles of the site, such as Moscow, New Richmond, Covington, and Cincinnati, have a decreasing popula-tion trend and the projected 1985 general population within this same area will increase only slightly. s There are no residences or f acilities located within the exclusion radius (380 9 tters), except a 'small manufacturing plant employing i
a labor force of approximately 50 persons. The applicants have made legal arrangements with the owner of the manufacturing plant, to evacuate the area immediately or take whatever action is deemed necessary to prevent exposure to these employees. Several transportation facilities provide limited access to the site. Normal access to the station will be U.S. Route 52 which traverses the site about 1/2 mile east of the plant location. An access road with a bridge over route 52 vill be provided. There j is no direct rail access to the site from the Ohio side of the rive r. The Chesapeake and Ohio Railroad line is located on the Kentucky side of the river within several hundred feet of the shore. A barge unloading and landins facility will be constructed to handle large. equipment. Comunercial barge and boat traffic exists on the river throughout the year. The applicants have stated that there are no recreational activities near the site except for pleasure boating on the Ohio River. Based on the evaluation of the population distribution in this i l region, and on our evaluation of the calculated potential offsite doses discussed in Section 4.0 of this report, we conclua that 1 I the proposed minimum exclusion and low population zone distances l l meet the 10 CFR Part 100 guidelines. u t
j i .i 2.2 Meteorology The applicants have presented five years of meteorological data from the Creater Cincinnati Airport. (CVC) which is located in Covington, Kentucky. The prevailing wind at the airport is from the south-southwest with an average wind speed of 4.0 meters per second. The data indicate that cals conditions occur approximately 3% of the time, and Pasquill type C conditions occur 5.44% of the time with an average wind speed of 0.75 meter per second.' The applicants are conducting a meteorological program at the plant site. The short-term onsite meteorological data collected, approxi-mately 2 months, were. compared with the data collected at 0,reater Cincinnati Airport. From this comparative analysis the applicants concluded that a Pasquill stability Class F accompanied by a wind speed of 0.5 meter /second was justified (for the 5 percentile mete-crological conditions used in the short term accident analysis). Our review of the applicants' comparison between wind speed at the site and vind speed at the airport showed that the airport speed was 2.5 times the average onsite speed. The wind speed was cor-rected for this dif ference. The accident meteorological diffusion conditions assumed by the Regulatory staff were therefore based on Pasquill stability Class F conditions and wind speed of 0.2 i meter /second. At the operating license review, the applicants and the Cournission's Regulatory Staff will evaluate the wind and temperature differential f l
~~ ~ f (AT) data obtained from the onsite meteorological program. A review of the onsite meteorological data at that time will demon-strate the adequacy and ccmservatism of the values used to oeter-mine both routine and postulated accident exposure doses at the site boundary. 2.3 Hydrolony The plant is on the east side of the Ohio River about one mile north of the village of Moscow and approximately 25 miles southeast of Cincinnati. The netural groted at the site is about 40 feet i above the normal reservoir level of Markland Lock and Dam (eleve-tion 455 ft MSL). Plant grade is elevation 520 f t MSL. The historical flood of record (1937) has been estimated to have produced a maximum runoff rate of 830,000 cfs and a river level of approximately 515 ft MSL, which is about 60 feet above " normal" river level. The applicants have estimated that a probable maxf-num flood (PMF) would produce a peak runoff rate of 1,980,000 cfs and a corresponding river level of about 546 ft MSL. Considering the possibility of wind-wave action the applicants have established an elevation of 550 ft MSL as the level for which flood protection will be provided. The applicants have also evalmated the effects of the potential failure of upstream dams. They have concluded that those located along the mein stem of the Ohio River are too low to cause a higher flood level at the site than the PMF. Their L
12 - i investigation of the many tributary reservoirs indicates the relative location and size would allow flood waves caused by failures to be almost fully attenuated before reaching the site, ) and in no came cause a flocd level Sreater than the PMF. The applicante wi s' p:avide flood protection for all plant struc-tures to elevation 521 f t MSL. Safety-related plant structures will be permanently flood protected to c'ev tion 546 f t MSL. They have also innstigated the ef fects of potential wave action above the flood protection level and have concluded waves may reach approximately elevation 550 f t MSL. Safety-related structural openings will be protected to an elevation of $50 f t MSL. The applicants have stated that the essential cooling water require-ments for Unit I will be 28 cfs. The historical minimum instanta-neous low flow of 2100 cfs in the ares was recorded at Louisville. Kentucky, on August 20, 1930. Since then dans have been constructed which can be counted on to augment low flow for all but the most severe drought conditions. The applicants have concluded that suf ficien t !!ow exists in the Ohio River for emergency requirements. They u.ve also evaluated the ef f ects of zero flow condition and have concluded that sufficient storage existe. in Markland Reservoir above the proposed service water intako elevation of 439 f t MSL to provide suf ficient water supply storage for safe shutdown. The 9
l I applicants have also evaluated the extreme low flow condition in the Ohio River (equivalent to the local minimum flow of record in the area) coincident with complete loss of Markland Dam, Their analysis indicates that the resulting water level would be of l sufficient height to assure adaquate suction head on the pumps to l maintain the safe shutdown conditions. ~ Cround water in the area is drawn from aquifers near the surface. i,I .1 The upper twenty to thirty feet of material is unconsolidated silty and clayey sand. Beneath the alluvium and to depths ranging from sixty to ninety feet beneath the surface are located sand and gravel glacial outwash deposits. Both the alluvium and glacial outwash deposits provide a usable ground water resource. Perched water table conditions were found where a few discontinuous thin silt and clayey lenses occur above the normal water table. Bedrock composed mostly of limestone and shale are found at depths below t! ninety feet. Both rock formations are relatively 1spermeable, and are not generally used as water supply sources since well yields are generally less than one gallon per minute. These rock deposits also act as aquitards against downward water movement. The applicants have estimated permeabilities decrecsing from 463 l 3al/ day /ft to 122 gal / day /ft between depths of 30 to 60 feet, res pe c t ively. Surface infiltration is low and has been estimated i by the applicants to be about 0.4 gol/ day /f t Ground water
- gradients are normally toward the river, although at high river levels flow reversal occurs. Eleven towns or water districts within 25 miles of the site serve between 500 and 14,000 persons at rates ranging free 0.05 to 1.57 mad. Four public users, generally down gradient from the site, are located at distances ranging from about 5 to 17 miles from the sita. They are, however, along the banks of the Ohio River, which can be expected to intercept most ground water flow passing beneath the site. The closest public well is on the same bank of the Ohio River as the site and about three siles upstream. Thors are 10 private wells within one mile of the site, and a total of 12 walls within three alles of the site. The applicants have determined that the public or private wells presently in the direction of ground water movement will r.ac be affected by conditions at the site. 1 The applicants have investigated the effects M both surface and ground water spills of radioactive material. They have concluded that a release of the entire contents of the liquid radwaste collector, floor drain and chemical waste tanks (35,900 gallons) without any holdup in the plant would encounter low permeability in the layer of clayey silt immediately beneath the site. In addition, the ion exchange capacity of the soil can be. expected i to reduce the activity by as much er 80 percent before such a
) l i I. slug would reach the Ohio River. Finally, by assuming mixing of . the activity with an average 24-hour Ohio River flow of 10,000 cfs (which is a minianas seven day flow condition), the applicanta calculated that the effective concentration at the Cincinnati water intake 25 miles downstream would increase to less than 10 pC1/cc. We conclude that the flood protection to be provided for the facility is acceptable and that sufficient water is available in the Ohio River for safe shutdown purposes. Based on our independent snalysis we agree with the applicants that no exist-ing public or private ground water or surface supplies facilities are likely to be affected by conditions at the site. f 2.4 Coolont seismolosty and soil Mechanics The site is located within the Central Stable Region of North America, an area characterized by broad, circular-to-oblong t erosional domes and sedimentary basins. A system of arches con-I nect the domes and separate the basins. The site lies near the top of the Cincinnati Arch. At the site, the Precambrian crystal-line rocks are mantled by approximately 3,500 feet of sedimentary T rocks of Paleozoic age. Bedrock consisting of Ordovician shale T 9 and limestone ranges in depth from 83 to 88 feet beneath the Ohio q l River Valley alluvium. l'
i i o ; 0 s i Faulting has not been identified within the sedimentary strata in the vicinity of the site. The nearest kruwn fault to the site is the Maysville Fanit approximately 30 miles southeast cf the site. The age of this fault is considered to be Paleozoic, and definitely older than middle Tertiary. Ha9r f aulting in the eastern Missourt-southern Illinois-western Kentucky region, more than 200 miles from the site, is not struc-curally or tectonically related to the geologic structure at the site. In addition and of greater consequence to the site is the Cincimati Arch, also ref erred to as the Indiana-Ohio Plat form, which has a history of earthquake activity. The applicants attempted to demonstrate that geologic conditions are such that the seismic activity at Anna, Ohio, should not be related to the proposed site location. We and our seismic consultants, the United States Geological Survey (USGS) and the Seismology Division of the National Ocean Survey of the National Oceanic and Atmospheric Administration (PDAA, formerly the U.S. Coast and Geodetic Survey) did not agree with the applicants' contention that the seismic activity at Anna can be precluded f rom occurring at the plant site. Due to the lack of geological information about the geologic structure of the Anna area and the Cincinnati Arch, our consultants (the USGS and NOAA) rScommend, and we concur, that the plant be designed for ground motion accelerations of 0.20g for the DBE and 0.10g e I
17 - for the OBE. The' applicant h.** agreed to design the plant to these acceleration values. The consul t..nt.;' reporta from the USGS and NOAA are attached as Appendix C and Appendix D, respectively. 2.4.1 Foundation Entineerina The proposed ground acceleration values listed above were acceptable to use on the basis of the applicants' proposal to remove unaccept-ably poor soil from underneath the entire plant area to an approxi-mate depth equal to 450 feet MSL and backfill with cohesionless material compacted to a relative density of 85%. 1he stability of the service water pipeline, the pump house, and - the river bank in the event of liquef action of soil above elevation 450 f t MSL adjacent to the engineered fill under a DBE and some l ) intermediate flood conditions were evaluated. We requested that the applicants perform analyses to show that the major plant struc-tures including the pump house would remain stable, that the service water pipeline would remain functional, and the river bank would not slide endangering the service water system, assuming simultan- . eous conditions of a flood elevation at 508 f t MSL and a DBE. The L applicants' analyses indicated that the major Class I structures and service water system would remain functional under these conditions. However, the assurance given in the form of computed ] 1 factors of safety were unacceptable.
h i 3 Our consultants, Newmark and Associates, performed independent analyses of the Class I structures assuming the same conditions described above. The resulting analyses showed that the Reactor and Auxiliary Buildings would move laterally and the service water 1 pipel s which is to be supported on pile bents would move toward the river due to drag from the liquefied soils tending to move g downslope (towards the river). Since the pump house is so massive and because it will be supported on a caissc'n founded at bedrock, it was felt that its stability is assured. Also, that portion of the service water inlet at the river will be protected by a sheet pile bulkhead. The bulkhead design would be acceptable provided it is designed for earth pressures at rest, for dynamic pressures, and that these pressures be assumed to act during low water levels. Our consultants indicate that the stability of the reactor build-ing and auxiliary buildings would be assured if a clay blanket that envelopes the foundations of these structures was provided. This clay blanket would prevent excessive pore water pressures and consequent sliding of structures in the event of a DBE and high water. The applicants have a8 reed to provide the clay blanket and to make observations prior to the operating license review both inside at.. outside the clay blanket; piezometric levels and rate of change i
~ I l l of water levels under partial flood conditions will determine whether pumping will be needed to prevent saturation of the founda-tion soils within the cisy blanket. Provisions for installing the pumps have been incorporated into the facility design, if required. Our consultants recommended, and we concur, that batter piles be added to the service water pipeline design to provide for lateral forces both downslope and transverse of the order of 360 kips per bent downslope and 100 kips or less in the transverse direction. l The applicant has agreed to provide the number of pile bents and battered piles that are required to resist these forces. The Newmark and Associates report is provided as Appendix E. j 2.5 Environmental Radiation Monitorina and Ecoloav The applicants have submitted their proposed environmental radia-tion monitoring program to be implemented at the Zimmer facility I s The proposed environmental radiation monitoring program will determine the levels of radioactivity that exist at the site and in the surrounding environment. The program will be initiated two years prior to fuel loading and will continue for at least two years during plant operations. This program will include I t i i ih
h collection and radiometric analysis of airborne particulate, surface water, bottom sediments, bot tom organisms, slime, well water, precipitation, fish, soil, vegetation, milk, wildlife, and miscellaneous food items, on-and off-site in both directions along the Ohio River. The area within 10 miles of the site will be monitored for natural background levels of airborne activity. Air samplers will be placed in population areas of greater than 100 people within five alles of the site and in population areas of 1000 or greater from five to ten miles from the site. Our consultant, the Fish and Wildlife Service of the Department of Interior, has provided a report on their review of the applicants' Environmental Radiation Monitoring Program, a copy of this report is provided as Appendix F. The consultant's report has been trans-mitted to the applicants for comment. The applicants indicated that applicable suggestions and recommendations received from the ) Fish and Wildlife Service and any other interested agencies would be incorporated into the Environmental Radiation Monitoring Program. The topography in the vicinity of the site is relatively level with sparsely wooded sections. The land is predominantly used for farm-i i t ing; information on the types of acreage ard yields of pertinent local crops has been provided. The applicants have already [ 1 1' Pj o m
l l
- performed an ecology study of the area in the vicinity of the proposed site. The purpose of the' study was to determine the bio-logical and relevant physicochemical conditions near the site prior to construction in order to assess the environmental impact of the facility.
The use of the cooling tower and settling pond is expected to minimiae the therms 1 effect on the aquatic ecology. The ecology study was performed in July through October,1970 and ) included daily and seasonal distributions, and abundance of vari-i ous life forms of the terrestrial and aquatic communities. The aquatic community study determined the composition of the plant and anissi components and the physical and chemical characteristics of the aquatic ecosystem at the site. The species studied includes Benthos, Periphyton, Phytoplankton, and Fish. The terrestrial study documented plant and animal diversity, distribution, and relative abundance of soils, vegetation, invertebrates, amphibians and reptiles, birds, and massaals. The data accumulated also will be used in the pre-and operational environmental radiation monitor-ing program proposed for the site. Based on our review of the information provided about the environ-mental radiation monitoring and ecological program, we and our consultants, the Fish and Wildlife Service of the Department of the Interior, conclude that the applicants '* programs are act.eptable. f
2.6 Railroad. River, and Air Traf fic There is no railroad cecess to the plant from the Ohio side of the river. The Chesapeake and Ohio Railroad traverses the Kentucky side of the Ohio River, within several hundred feet of the shore. The applicants plan to provide e barge unloading and landing facil-ity to accommodate shipments that cannot be transported by truck. The railroad car to barge transfer point will be located on the Kentucky side of the River. The existing transfer facility at the J. M. Stuart Station (a fossil plant owned by applicants) will be used or a new installation will be constructed opposite the facility. The Ohio side landing facility will also be constructed at the River's edge near the plant. The applicants have analyzed the potential effects of accidents ( l involving barge traf fic and its associated cargo. Included in the analyses were effects of barge traffic and potentially dangerous cargo such as toxic chemicals, explosives and flammable materials. The applicants concluded and we agree that the facility could be maintained in a safe condition following postulated accidents associated with the barge traf fic. Two small airport facilities are located about 10 and 15 miles north of the site. These airports do not offer scheduled commer-cial flights and their use is limited to one-engine light aircraf t f
__ _ _ _ _ _ _ _ _ _ for business or personal use. The nearest air facility that accomunodates large commercial aircraft is the Greater Cincinnati Airport located about 30 miles northwest of the facility, gased on guidelines used on previous reviews, we have concluded that the probability of an aircraft crash affecting the Zinsner facility is sufficiently low that no special design provisions are needed. 2.7 Conclusion We conclude that the exclusion distance and low population sone radius are acceptable; that the naturs1 phenomena are being ade-quately considered; that the meteoro.ogical and environmental monitoring programs are acceptable; and that the site is accept-able for construction of the William H. Zisumer Nuclear Power Station. 3.0 FACILITY DESIGN 3.1 Reactor Desian 3.1.1 General The reactor design of the Zimmer Nurlsa. Power Station, in many respects, is the same as that of several other BWR plants previ-ously reviewel. Tables 3.1.1 and 3.1.2 provide a comparison of major reactor design parameters for a number of these facilities. Although Table 3.1.1 identifies some differences between the Zimmer and Hatch facilities, comparison of the tabulated data f a b
- Table 3.1.1 M i j o r thi gn Ch an pe r, I n c o r po r a t e d into the Wm. H. Z ienne r Statinn l a Pe rane t e r Vm. H. Z insne r Edwin I. Hatch
- Ee ctor O.
.it Rer i rc u la t ion Loor Design Flow Control Throt cling W1th An Recirculat ton Pump Additional Valve Speed l Pump Flow Rate, spe/ pump 33,8 C 45,200 I Total Core Flow Rate, 6 6 lbs /hr 78.5 x 10 75.5 x 10 Wominal Pipe Diameter taches 20 28 Core Spray Systesi Number of Systems 2*a 2 Pussp Flow Rate ud Pressure, spe (psid)*** 4625 (119) 4625 (120) 1330 (1130) High Pressere Coolant Injectton Systen Coolant Injeet ton Mode Sprmy Flood Injection N thod Directly into core Indirectly into core via core spray sparger via feedweter sparger Flow Rate, spe (pald) 1330 (1130) 5000 (150-1130) Fussp Ntive Type Motor with sep-Stese Turbine arate Diesel generator tw Pressure Coolant Injection System Nu. of Pumps 3 4 Flow Rate, spe/ pump (psid) 4970 (20) 7700 (20) Loop Selection Logic None Sense Break Location Injection N thod Top of Core Recirculat ton Loop Piping Prieurv Containment Concept Over li Under Pressure Pressure Suppression Suppression (Lightbulb and torus) Construction Type Prestressed Concrete ASME Steel Pressure Steet lined Vessel Dryvell Geomet ry Frustum of Cone Light Bulb Shaped Pressure Suppression Chamber (PSC) Geometry Cylindrical Torus PSC and Dryvell Internn! l Design Pressure, psig 45 56 j i
- The Edwin 1. Hatch Nuclear Tasility has the same equipment arrangement as other BW facilities such as Brwns Ferry, Peach Bottom, and Enrico Fermi-2.
- A high pressure core spray system is included and serves as a redundant low pressure core spray system.
u
- psid - pounds per square inch differential between reactor vessel and primary containment.
1
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t r s o 4 a ) P W M ( / / ( k e c n y or M ( t t u u a t a o n S l e d ( t u u r t e e e e i i ( p n r u p p B B r p m r t g a r e p t t ( ( u a a s n ms e w t u u t l i f l p i ) e p d w o u o o x x a a d o u e l e t e i s s o p o u u r t c t o p s t s p l l l l e o e r r p og y p p l l a a f f p t r e i C/) S/ l a a n m m o t c t sr s 0 a m c r r t t e r c e e e ypeyp 3 m r i e e a a t e m r j b ob a o 1 r e r h h e e a d omr o 1 wt e h t t t h h r n i f f nl upl h t c e p e d o o a( nS( t t e m e m e t l t ( a n l u g u g a n a e r r S I eI d g e m a m a w g v d e e CS rC 0 e i i r i r d i i i b b ePD oP 3 t s t x e x e e s u s m m rHACL 3 a e e a v a v e e q n u u o 1 R D N M A M A F D E I N N C l
- _ _ _ _ - _ _ _ - _ _ shows that the nuclear fuel, reactor vessel and the thermal-hydraulic characteristics (also refer to Table 3.1.3) are basically the same. Our review and accera sace of the nuclear, thermal and hydraulic design was based primarily on the comparison with previously reviewed and approved boiling wate'r reactor characteristics. 3.1.2 Nuclest Design The nuclear fuel and its arrangement in the Zissner core is sub-stantially the same as that of previously approved plants (refer to Paragraph 3.1.3). The nuclear reactors compared in the Tables use reactivity concrol sys tems that are all identical. Control of the core reactivity is provided through movable cruciform control rods and a variable recirculation flow rate which automatically accessmodates demand for increase or decrease in power. A standby liquid control system is provided that injects into the core a solution of water and sodium pentaborate, a strong neutron absorber. The method used to regulate the recirculation flow rate through the core has changed significantly. Previous BVR's utilf red a variable speed recircula-tion pump while Zimmer is the first nuclear plant to utilize a flow control valve and a constant speed pump in the recirculation loop piping system. Mechanical and electrical devices associated with limiting positive reactivity insertion into the core include l control rod velocity limiters, pre-set control rod operating i
..,.. - _ _ _ _ _ _ _ - _. patterna, limited control rod drive speed capability, a control rod worth' minimizer computer progras and its rod blocks (to inhibit selection, withdrawal, or insertion of out-of-sequence roda during startup, shutdown, or low power operation), and con-trol rod drive housing supports. The rod patterns administrative 1y permitted will limit individual rod wortha to less than 0.01 Ak/k. The nuclear fuel la designed for 19,000 MWD /T (averaged over the initial core load). serichment in U-235 for the fuel charge will not be known esactly until about eighteen months before the initial fuel loading. The highly enriched fuel rods will be designed and made with large diameter end plug-shanks that can fit only in the proper location in the upper tio plate of the fuel assembly. The design permits lower enriched rods to be placed in the locations for higher enriched rods accidentally, but this would result in a fuel assently with less than the standard fissionable loading. The applicants have indicated that the General Elmetric Company will obtain inservice data on use of the fuel. An analysis of the data vill provide the inservice operating limits for reactor opera-tion. We conclude the information submitted provides a suitable basis to expect safe performance of the reactivity control mechanisms and the nuclear fuel under normal and accident operat-ing conditions of the Zinner facility. 4 I N
t i 3.1.3 Thermal and Nydraulic Design The thermal and hydraulic characteristics of the Zimmer and Hatch reactor cores are nearly identical. They are very similar to the cores of Newbold Island, Browna Ferry, and Fesch Sottom. Table 3.1.3 provides a comparison of characteristics for these facilities. Core cooling systems for the facilities are identical two recirculation loops and twenty jet pumps fer each reactor. The recirculation flow rate for the Zimmer facility is less than the flow rate of similar SWR facilities. However, the tots! developed head is t.igher so that the total jet pump flow is about the same. This could af fect the jet pump performance characteristics. This area is being invest 15ated to assure that the jet pump design falls within the experimental design limitations. During normal steady-state operation the thermal hydraulic design of the Zimmer core will assure a minimum critical heat flux ratio (NCHFR) of 1.9 or greater and the maximum linear heat generation rate will be safn-tained below 18.5 kW/f t. The average core power density will be 51.2 kW/ liter. General Electric submitted a topical report " Analysis of Anticipated Transients Without Scram" (NED0-10349) in Ma rch, 19 71. Our review of this report led to the conclusion that tripping of the recir-culation pumps, for some transient situations, would provide
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I additional plant protection. The applicants have stated that they will provide the capability to trip the two recirculation pumps on a high pressure signal to assure acceptable limits and consequences of an ATWS in the Zismer plant. We plan to review the specific means of implementing this feature during cons true-tion of the facility. We conclude that the thermal and hydraulic design provides adequate safety margin to protect against fuel failure during normal steady-l state operations and abnormal operational transients. 3.1.4 Reactor Internals 3.1.4.1 Design The design of the reactor vessel internals will be in accordance with Section III of tthe ASME boiler and Pressure Vessel Code. The material used for fabrication of most of the reactor vessel internals will be solution-treated unstabilized Type 304 austenitic stainless steel conforming to ASTM specifications. Weld procedures and welders will be qualified in accordance with the ASME Boiler and Pressure Vessel Code. The reactor vessel internals will pro-vide a floodable volume in which the core can be cooled in the 0 event of a break in the nuclear system process barrier external l to the vessel. Deflections and deformations will be limited to allow proper function of the control rods and Core Standby f i 9
d-Cooling Systems thereby assuring safe shutdown of the plant and removal of decay heat. These capabilities of the reactor vessel internals have been examined under conditions of the loss-of-coolant acciden t, the steam line break accident, thermal shock following LPCI or HPCS reflooding of the inner voluum during a LOCA or MSL i ) break, and earthquake. Forces and induced stresses resulting from these events are within the system capabilities. For normal design loads of mechanical, hydraulic, and thermal } I origin, including anticipated plant tranatents and the operating 4 basis earthquake, the reactor internal components other than core support structures will be designed to function within the appro-l priate normal and upset operating condition category stress limit r criteria of Article 4. Section III of the ASME Boiler and Pressure m L
- t Vessel Code. We find these code limits acceptable.
Under design basis accident conditions, which include the combined loads froes a recirculation line break or a steam line break plus the design basis earthquake, the reactor internal components will - l be designed to the criteria submitted in Appendix C of the PSAR, ~ as roodified by Amendments 7 and 13. For these conditions, the deflections of the fuel channels and control rod housings, are also limited to assure control rod operability and the preservation of 1.
core geometry to permit emergency core cooling. The modified design criteria submitted for the Ziammer reactor internal components are consistent with comparable code emergency and faulted operating condition category limits and the criteria which have been accepted for all recently licensed plants. We find these criteria acceptable. 3.1.4.2 Dynamic System Seismic. Operatina and LOCA Analysis _ seismic loading or, the reactor internals will be determined by means of a normal mode-time history analysis. We find this proce-dure acceptable. Dynamic loading due to normal and upset operating conditions will be computed by means of quasi-dynamic methods based on the measured vibration response of similar reactor dseigns. We are presently reviewing a summary of SWR vibration test data with design predic-tions submitted as Amendment 19 (Proprietary) to the Quad-Cities docket and referenced in Amendment 7 to Zimmer. The additional { i information required to complete our review of Quad-Cities Armend-ment 19 will be submitted as a topical report. Final sceeptance of the procedures proposed for determining the dynamic loading due to normal rnd upset operating conditions for the Zimmer reactor j internals will await completion of our review of Quad-Cities Amendment 19. When we complete this review, the results will be applied to Zinumer. If necessary, confirmation of the ID
~ ) w l procedures employed to determine the normal and upset condition design dynamic loadings for the Zimmer reactor internals structures can be obtained during the pre-operational vibration test program for this plant by the use of additional instrumentation and analyses beyond that currently contemplated. This matter is discussed in l. more detail in Section 3.1.4.3. Design loadings for the postulated loss-of-coolant accident (LOCA) will be determined by computing the response of each structural l i i member to the calculated peak pressure differential applied as an t equivalent static load. In response to our concerns regarding the i validity of this static analysis the applicants have stated that the natural frequency of the BWR internal structures is more than ten times the calculated frequency of the LOCA loads thus assuring no significant dynamic amplification. On.he basis of the informa-tion submitted by the applicants we find the analy*ical method from ~ which their evaluation was made is acceptable. ) The nuclear steam supplier (CE) is reviewing the hydraulic loads that might be experienced by the guide tubes that surround the y control rods, in the unlikely event of an accident. We will review the information that the applicants are developing and will complete I our review of the adequacy of the control rod guide tube design f k margins prior to completion of plant construction. In the event s h1 b
. it appears that additional design margin is required to provide further assurance that the guide tubes will not collapse under accident loads, such margin could be provided by increasing the guide tube wall thickness. Subject to the verification of the l adequacy of the design margins of the guide tubes, we find this j \\ to be acceptable. l 3.1.4.3 Vibration Control The nuclear steam supply system supplier, General Electric, has stated that other plants with similar internal supports will be vibration-tested prior to the coupletion of Zinner and that these prototype tests will identify the potential problem areas, if any, for the Zimmer plant. Zimmer is the first of the CE-69 product line plants with changes in the recirculation system piping and equipment and with design changes to the core standby cooling i i: i systes. Additional information has been requested to allow a complete evaluation of all the prototype testing accomplished by General Electric. The use of prototype results are valid j only if the analytical methods and test procedures employed for the prototype have been confirmed by an acceptable preoperational a vibration test program for reactors of the same product line. (] a
- N 1
1 4 8
Since a prototype plant has not been designated for Zineser and since, as noted previously, the additional information required to complete our review of Quad-Cities Amendment 19 has not been sub-mitted, the applicants have agreed to provide the capability to conduct either a confirmatory type or prototype vibrations! test
- program, special measures during design and construction of the facility will provide the flexibility to implement the test program.
4 This program should measure the responseO of the reactor internals to determine the flow-induced forces and the related dynamic fore-6 ing functions for all significant modes of normal reactor operation. The data obtaired by these measurements on reactor internals should be sufficient t.o verify that the steady state and cyclic stresses in the components, as determined by analyses, are within the accept-able design limits set forth in the design specifications and code requirements and that the results meet the acceptance criteria of the vibration test program. In the event that the Zimmer plant is not the first of the CE 1969 product line to receive an operating license and has. conducted a satisfactory vibration test program, we may relax our require-ments, but as a minimum a confirmatory vibration testing program l will be required. e Frequency and magnitude of vibration (in terms of displacements, velocities, and accelerations). j u___________________
l i 3.2 Reactor 3.2.1 Sys tem Quality Croup Classificat ions l l The applicants' have applied a system of code classification groups to those pressure-containing components that will be part of the reactor coolant pressure boundary and other fluid systems important to safety. Those classification groups generally correspond to the tentative code classification groups A, B, C and D developed by the regulatory staff. The codes applicable to the componenta in each of the clnasification groups are identified in Table 3.2.1. We and the applicants are in general agreement on the code classi-ficath groups for the reactor coolant pressure boundary and the majoris, of those fluid systems important to safety. yor those systems, portions of systems, or components where the applicants' claas!!! cation grouping differs from oura, the applicanta have upgraded their classifications to a quality level substantially equivalent to the opplicable staf f classification code groups. Clarification of these upgraded classifications is documented in Appendix A.0 of Amendment 6 and in Amendment 7 to the PS AR. The ASME code for pumps and valves does not adequately cover pump and valve pressure retaining castings for lines over 2 inches up to and including 'A inches in systems which are classified A & B l l and in Group D for main steam and turbine bypass lines. To ec,ve r 4
k I !!i w" ' i I t r I eI t n o e I rI n e a d I uV e l v 0 o t s l t a i C s n sna n rv u 1 s e t eov e oi q tE al s rii l u F 1 nMsl a ) a P t u a 0 q 3 eS pCv sd e c q v 0E r l Am i dil D d eE i 1 o R a usu ru nS u D r vtP eq aqt p a r q A o 0 Iif vE di a u .o E W1 S uarl nl o re r ed1 r A 1 Nqroar a e AEDfVo t rb C l o o 6 1 so iC n ,9 3 ~ - l o o 0 0 t el Bl i 2 5B R s nl a es 6 6 e s ech s E si - I E t v p li s t Ms v I IS S n l m at i S ei P PN N e a u vrs n AVD A AA A1 V P i ad U upr q a r e ecd e eI s n p i n w rI t3 k
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r r e t nS rut s 7 a t exb o C ed ion ehoN na Es( r E a f l o wi l c el 1 Me t c iC t s bti0 emC 3 S ve ei i s o C 0ai as0h e Al t bm.l d Bl 2 nb c 1 t r. B ao rd p r es 6ia i sD i1 tvn l eep a Ess mT l eAhu-I f t l pt a d Msa I a n S el Pxn poWieSqT N ado a u pd t S rno ht t e a AVC AEi ACA wRN A DaF soih t S nt t s s s t s n d eI pe re4i n rI me aos a uI uS pd d s P ee sn 1 t nbi e eo r. so f ? d ri s oI aiyi W o Pt s f ct ac C c a s ame A d e l es gr p f nS C d a nuns o p a ol igo u CC) niie y o re 7 a if t s y r r ed Es( ans o I a G l o 1 Me t onh a iC 3 S ve ecit mm o u Bll A Al t r m y B ao - rao S es tvn eoxt E s s I f t r e Msa S ad o ue t S el N r no szt n AVC A DaF siie r esal c r ra i pet v t g r d rei n e i e n E l enu e rs se h e a T l h eq n ul P g pg g s O AWpe w se as saa n s e N' ss 5 rk ork i pv es 1 on mon p ml )a re - t a t t a i ua C PV 0ST AST P Pv ( l I I1 I[I' l l)! l
these castings adequately, volumetric examination (radiography or ultrasonic testing)- will be necessary. Where size or configuration does not permit volumetric examination, surface exantination (mas-metic particle or liquid penetrant testing) may be substituted. The applicants have agreed to conform to the ASME code for pumps and valves including the added testing outlined above. The Group Classification Diagrains, given in the PSAR, Figures A.2-1.1 and A.2-1.2 are acceptable; however, the applicants have agreed to provide piping and instroentation diagrams identifying A the detailed boundary limits of each classification group during m "= the construction stage. e- ? We find that the system quality group classification as specified = by the applicants and supplemented by provisions for upgrading 5 quality levels, and additional nondestructive examination require- = ment discussed above are acceptable for this facility. -= mi m 3.2.2 Resetor Coolant pressure Boundary 3 3.2.2.1 Design _= The reactor coolant pressure boundary will be a Class 1 (seismic) g system designed, fabricated, and inspected in accordance with the g-requirements of the applicable codes delineated abc,ve in System 3 t __f Quality Croup Classifications. The stress limit criteria specified / Y k y E Y
. for the notaal mad upset operating condition categories of the applicable codae will apply for all nornwl loads and anticipated transients including the operating baats earthquake. The design, fabrication, and inspection criteria are consistent with those accepted for all recently reviewed plants of this type and we find them acceptable. Under the loads that result from the design basis accident, the design basie earthquake, and the combination of these, the capo-nonts of the reactor coolant pressure boundary will be designed to the applicable emergency and faulted operating condition category limits of the appropriate codes. Where the appropriate codes do not provide explicit design limits for these operating condition categories, these components will be designed to the criteria sub-i mitted in Appendix C of the PSAR. The criteria of Appendix C as modified by Amendments 12 and 13 are consistent with comparable component code criteria. We find these criteria acceptable. 3.2.2.2 pipe Whip Criteria The applicants state that they will examine all lines which are part of the reactor coolant pressure boundary for their potential l to whip. In no case will reactor coolant system pipe whipping be allowed to cause any damage to the reactor coolant system, essential
- __ _ -_ ____-- - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ equipment or to the containment. Pipe whipping will be prevented by installing restraints and/or placing check valves or closed - motor-operated valves as close to the reactor vessel as possible. 1 Protection against the effects of longitudinal as well as circum- ) farential breaks will be included in the design. We find the pipe whip criteria acceptable. l 3.2.2.3 jia_in Steam Line isolation Valve Laskate l In operating 3WR's, leakage through the main steam line isolation valves OtSLIV) following a postulated loss-of-coolant accident relies on the low leakage characteristic of the valve. The applicants will provide a main steamline isolation sealing system that will preclude placing complete reliance on the low leakage characteristics of the valves. Several sealing concepts are being considered by the applicants and a study to select the best pos-l 1 sible MSLIV sealing system is being performed. The proposed design will be submitted for our review and approval prior to j incorporating this design feature into the facility. Any portion of the MSL utilized as part of this sealing system will be designed to the seismic requirements of Group A piping outlined in para-graph 3.2.1 above. We will complete our review of the adequacy of the sealing system selected prior to completion of construction Y .of the facility. We conclude that this is acceptable. d i lj i 1 0 l
3,2.2.4 Reactor Coolant Recirculation Loop' Flow Control Valve Zimmer will use a flow control valve in conjunction with a constant speed pump to vary the recirculation loop flow rate. In previous . SWR's, variable speed pumps were used to perfom this same function. l l Each of the two recirculation loops in the Zissner Station will have a motor-driven constant speed, vertical, concrifussi pump which develops 845 feet of head at a flow rate of 34,000 syn. The flow control valve in each loop will be 20 inches in diameter and is of the ball type. Valve motion will be accomplished by an electro-hydraulic actuator. Each loop will also contain two 20-inch-diameter globe valves located in a bypase line around the flow control valve. The flow control valve will be operable in either the manual or the automatic mode. The automatic mode will be o'perable over a 35% range, nominally from 65% to 1002 power. Various interlocks will be provided that pr'otect the pump against cavitation. These interlocks will cause the pump to trip, prevent pump startup, cause the control valve to close, and/or cause the pump discharge block valve to close. Those interlocks that will have a safety implication are discussed below. E The valve and actuators are being developed currently. Tests have been performed on an eight-inch-diameter valve and the data are t l E
l l l 1 42 - f l l l being used to design the full-size valve.. The valve actuator and 1 some electronic components currently are being developed and tested. l The testa include performance capability, f ailure modes, and the i determination of the maxiumum valve stroking speeds (as limited by the hydraulic design). A fully-assembled valve, actuator, and control system will be tested prior to operation. These testa will be conducted in a test loop under actual operating temperatures and pres s ures. The safety evaluation of the recirculation loop flow control con-sidered valve malfunctions, and the interaction of valve motion with a loss-of-coolant accident. The applicants have analysed the consequences of the flow control valves in both recirculation i t loops opening at maximum speed and closing at maximum speed. Thermal limitations were not exceeded in either case. The applicant also analysed valve cycling. It was found that a high flux scram would occur before thermal limits are approached. v In the event that the flow-rate-sensing signal in one recirculation l- } I loop were lost while operating in the automatic mode, the valve in the other loop would receive a signal to open. This situation was analyzed at various power levels by the applicants. Again it was found that satisfactory thermal astgins were msintained. I i 4
l l \\.II l I\\ Motion of the control valve and the pump' discharge block valve j would not be precluded following a loss-of-coolant accident (LOCA). The normal response of the controls and interlocks would cause the control valve to close at a rate of 10 percent /second following a thCA, and if a failure of a valve actuator hydraulic line were to occur, the valve could close at a rate of 20 percent /second. These 3 valve motions would have the effect of decreasing the effective size of postulated breaks in the recirculation lines. This, in turn, would reduce the rate of reactor pressure decrease and thus i affect the performance of the CSCS. The applicants have analysed for these potential effects. The analysis assumes that the flow l 1 control and/or block valve closes on the broken and unbroken j; recirculation loop and the simultaneous failure of the HPCS system . f';I 4 e l 1. occurs. The result of this worst case analysis indicates that the $1 resulting cladding temperatures are less than the design basis j. accident maximum allowable temperature (2300*F) with no valve l: f, closure. However, the applicants will provide fully independent instrumentation and associated power supply which will prevent the p l! k recirculation flow control and block valves from closing during a f* lhCA. t j f; c We intend to continue the evaluation of the design of the control [ valve, the circuitry, and testing program during construction of the facility. m .. ) L 1, T4
- 3.2.2.5 Primary System Pressure Relief The objective of the pressure relief system will be to listit any over pressure of the reactor coolant boundary (reactor vessel and recirculation lines) that might occur from abnormal operational transients. In addition, the automatic depressurisation feature of the system will be used in the event' of small primary systein breaks to depressurize the primary system for low pressure coolant injection system operation.- Eight safety-relief valves will be provided that discharge to the suppression pool and perform the following functions:
(a) in the relief mode the valves will be opened by a signal genersted by high primary system pressure to limit primary system overpressure to a point below the self-opening I pressure of the safety and safety-relief valves. (b) in the safety 1 l l mode the valves will be self-actuated by primary system high pressure to augment safety valve capacity, and (c) to automatically depressurize the primary system. There will be five safety valves that discharge directly to the drywell and function in conjunction with the safety-relief to prevent overpressurization of the primary /, system. The safety-relief valves will be spring-loaded valves ) equipped with air cylinder operators that permit remote, manual or g automatic opening of the valves at pressures below their scif- .4 }iI opening set pressure. The safety valves will be the conventional I spring-loaded type. j i I. Sn
_ _ _ _ - The designed safety valve capacity for this system is based on a pressure rise resulting from a main steam flow stoppage (turbine trip) at operating conditions, turbine pressure 980 peig (103% rated), no steam bypass of the turbine mad a reactor scram due to high pressure. The analysis indicates that a design safety valve and safety-relief valve capacity of approximately 100% of reactor rated flow is capable of maintaining an adequate pressure margin (approximately 60 pet) below the peak ASME Code allowable pressure of 1375 peig. Eight of the thirteen' safety and cafety-relief valves will limit the peak pressure of the reactor coolant bondary to 1375 peig. Based on the above, we conclude that the systes design is acceptable. 3.2. 2. 6 Seismic Destan of Main Steam Lines The main steam line OtSL) piping from the reactor pressure vessel out to and including the MSL outer containment isolation valve will be destaned, fabricated, and inspected to the requirements of Class I (seismic) including Quality Group A. The main steam lines from the outemost valve up to the turbit.a. hcwever, will be designed, fabricated, and inspected to the requirements of Class II (seismic) including Quality Group D*. Consequences of postulated failures of this portion of the main steen line are presented in Section 4.0 of this report. Although the calculated radiological consequence of such accidents are well below the 10 CFR Part 100
l i guideline values, they are significant. For this reason, we and the ACKS have required that the main steamline be upgraded to a seismic I classification to further reduce the probability of a steamline-break accident. The Main Steam Line (MSL) from the outmost MSL containmarnt isole-tion valve up to and including the main stop and control valve assembly and all branch lines 2-1/2 IFS inches diameter and larger up to and including their first valve (including their restraints) will be designed by the use of an appropriate dynamic seismic analysis to withstand the CBE and DBE loads, in combination with other appropriate loads, within the limits of the ANSI B31.1 piping code and the PSAR Croup B requirements for OBE and DBE 4 j loading combinations. The analysis will either confirm that the main stop and control valve assembly and branch lines terminal stop valves, including their directly associated supporting struc-- tures connected to the turbine building, are rigid anchors (with natural frequencies above 33 eps) for the MSL, or that these valve l assemblies and support structures will be included in the dynamic 3 t seismic analysis of the MSL. The pressure retaining portions of the main stop and control valve assembly and the branch line termi-f nal valves will be designed to withstand the OBE and DBE loads l fl - [ within the PSAR Croup B requirements. The dynamic input loads { for design of the MSL will be derived from a time-history modal gn } I analysis (or an equiStlent method) of the Auxiliary, Reactor, and applicable portions of the Turbine Buildings. .j The Class II Turbine Building housing portions of the N5L any undergo some plastic deforestion under the DSE; however, the i plastic deformation will be limited to a ductility factor of 2 i and an elastic multi-degree-of-freedos system analysis will be used to determine the impet to the MSL. The stress allowables and the associated deformation limits fo'r piping and supporting structures, including related portions of the Turbine Buildings, will be in accordance with the pSAR Croup a requirements for OBE and DSg loading combinations. 3.2.3 Reactor Core Support Structures-Desian For all operating condition categories, i.e., normal, upset, energency and faulted, the core support structures will be designed to stress, deformation and fatigue limit criteria which are consis-tent with the criteria of the code for core support structures currently in preparation by ASME. We find these criteria acceptable. 3.2.4 Fracture Toushness Criteria The reactor vessel will be designed in accordance with the ASME Soiler and Pressure Vessel Code, Section III. Current ASME Section i 111 Code rules permit that a vessel be pressurized only above a i temperature equal to the sum of the Nil Ductility Transition (NDT) ~
A- { ~ temperature plus 60*F. The h7f temperature, according to paragraph N-331 of the Code, can be obtained by eit.her the dropweight test h (DVf) or the Charpy V-notch (C ) impact test. y Recent fracture toughness test data indicate that the current ASME I Code rules are not always sufficiently conservative, and may not guarantee adequate fracture f oaghness of ferr* tic materials. While I the Charpy V-notch tests continue to be useful in measuring the upper shelf fracture energy value, the C specimerns, generally, y do not predict correctly the KIrf temperature. The latter, therefore, must be obtained from other tests, such as the DWT test. Quite often, l l siso, considerable difficulty exists in defining from the C test ] y curves the transition temperature region in which fracture tough-l ness of ferritic materials increases rapidly with temperature. In addition, this transition temperature aegion shif ts to higher tem-peratures when the thickness of the specimen tested la increased (sise effect). We have discussed the proposed AEC fracture toughness criteria with the applicants and advised them that adequate fracture toughness data will be required to establish appropriate heatup and cooldwn limits for this plant at the time of the operatinR license review. In addition, we informed the applicants that f racture toughness data for all pressure-retaining ferritic components of the reactor n _ _ _ _ _ - _ _ - _ _ _ _ _ - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ - - - - - - - ' - ^ ' - - ~ ^ ' ~ ~~
49 - i coolant preusure boundary would have to be provided. The data submitted by the applicants meet the current requirements of Section 111 of the ASME Code, but are not adequate to establish compliance with the proposed AEC fracture toughness criteria. We intend to review the fracture toughness data available at the time of operating license review, and we will apply the AEC fracture toughness criteria to establish appropriate heatup and cooldown limits for this plant. 3.2.5 Reactor Vessel Material Surveillance Program i The material surveillance program is consistent with ASTME-185-66 I which has been accepted on previous similar BWR plants. The reactor vessel material surveillance program submitted by the applicants includes provisions with respect to total number of specimen capsules placed in the reactor vessel, number of capsules scheduled to be withdrawn and tested, archive material available for additional specimens if required later in the service life of the vessel, and material chemistry documentation. We conclude that this program will adequately monitor radiation-induced changes in material fracture toughness properties of the ferritic materials of the Zisner reactor vessel, during its service life. 3.2.6 Leak Detcetion The applicants' proponed reactor coolant pressure boundary leak detection system within the drywell has been improved with the k l
b i i .t l. l' use of airborne parr*culate sangling equipment. The proposed L -l equipment is similar to that now u:;ed at Dresden 2 and 3 and is considered acceptable for leak detection. The air sampling points f in the drywell and suppression chamber will be monitored continu-f ously by drawing samples outside the containment and measuring l gross beta activity. The drywell air sampling system will supple-t 1 ment other leak detection equipment that permits measurement of drywell pressure, temperature, and sump level. Air samples taken mar.ually from specific areas will provide a means to determine the i approximate location of the break. j Temperature, pressure, and flow sensors with associated instrumen-l tation and alams will be provided beyond the limits of the reactor 1 l coolant pressure boundary for leek detection of vital fluid-carrying systems external to the primary containment. We conclude that the leaksge detector systems proposed by the applicant are acceptable. 3.2.7 Inservice Inspection 3.2.7.1 Inservice Inspection Program for Reactor Coolan! Pressure Boundary The ac'ceas for inservice inspection of the reactor coolant pressure boundary will be in compliance with t' ' ASME Boiler and Vessel Code; Section XI: Rules for Inservice Inspection of Nuclear Reactor Coolant Systems. Access will be provided for each spplicabic i
I component uithin the reactor coolant pressure boundary in accordance with the requirements for inspection given in Table 15-261 of Section II. The applicants have provided the necsesary equipment for both remote and contact inspection of the reactor coolant pressure boundary. A procedure has been prepared to coordinate the equipment development wit!4 the reactor plant design. Procedures and systems are being developed for recording and gathering the inservice inspection data and for comparing the data taken in inservice inspections with that taken in the baseline inspection. We conclude that the inservice inspection program is acceptable. 3.2.7.2 Inservice Inspection Pronram for Croup B & C Fluid Systems The applicants' program includes provision for inservice inspection. for the Group B and C fluid systems, to the maximum extent femalble. Engineered %sfety features will receive functional testa and inspection to assure integrity and operability. We conclude that the Group B and C Fluid Systems will be designed to permit periodic inspection and that the proposed access provisions are acceptablu. 3.2.8 Reactor Coolant System Sensitired Stainless Steel The applicants state that the sensitization of non-stabilized stainless steel during fabrication vill be avoided. The precautions used to atsure this include rapid cooldown from
1 1 i ~ $2 - solution heat treatment temperature, the use of low carbon stainless steel where exposure to temperatures above 800*F will be experienced due to postweld heat treatment operations, the' i use of low carbon welding electrodes, and the limiting of weld-ment interpass temperatures to be below 350*F. Non-sensitised l stainless steel nossles and pipe safe-ends will be affixed af ter the final vessel stress relief. We conclude that the planning i ') I l to avoid sensitization of sustenitic stainless steel during the ] I j fabrication period is acceptable. 3.2.9 Foreign Procurement J The applicants have not yet selected all of the suppliers of resetor coolant pressure boundary components. The appiteants have stated that no foreign procurement is anticipated and that if foreign j procured components become necessary, a tabulation of components procured frw foreign vendors would be provided for our review. 3.2.10 Electroslan Welding The applicants were asked to p'rovide a tabulation of all componente using the electroslag welding process. In response to our query, the applicants stated that the use of electroslag welding is not anticipated. However, should the process be used, the applicable components and process information will be identified and provided for our review. i
- - - _ - - _ - - - - - - 3.3 containment
- 3. 3.1 General C<mtainment Desian The containment systems will include the primary containment using i
the pressure suppression concept and the secondary confinement l which includes the reactor building, its recirculating (atmospheric ventilation) system, and the standby gas treatment system (5 CTS). The containment configuration will be similar to that for the Limerick and Shoreham nuclear power stations. The drywell will be a steel-lined prestressed concrete vessel in the shape of a frustum of a cone. The vapor suppression chamber will be a steel-line! prestressed concrete right circular cylinder located directly beneath the drywell. The drywell and wetwell will be separated by a reinforced concrete floor penetrated by eighty-eight vent pipes. A low-leakage Reactor Building will surround the primary containment to serve as a secondary barrier. A comparison of the containment l design parametets for the Zinser Nuclear Power Station with those r of other BWR's is presented in Table 3.3.1. Both primary and secondary containments will meet, among others, the criteria for Class I seismic design. Structural aspects of the containment design are discussed in the Station Structures and Shielding Section. (Section 3.8.) i s 3.3.2 Primary Containment The vapor suppression concept for the' reduction of pressure inside the primary containment following a LOCA has been used in the 1
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. Zimeser design as in other BWR facilities. The drywell and wetwell configuration will be of the "over-upder" type, f.e., the drywell will be constructed above the wetwell and together they will form a erntinuous, single structure for the primary containment. The codes utilised in the design of the primary containment are ACI 318-63, " Building Code Requirements for Ra^1aforced Concrete" and the ASME Soiler and Pressure Vessel Code, Section !!!, Subsection 5. The latter code governs design of the drywell head, locks, penetrations, and other steel struc-tures having pressure vessel functions. Where appropriate, the latter code also applies to the steel liner; however, the liner plate is designed to function only as a leak tight membrane. Containment liner material will be A-516 steel, Grade 60 to SA 300. Connecting the drywell and wetwell will be eighty-eight straight I pipe vents, which will be approximately forty-four feet long and two feet in diameter. The vents will project a short distance above the reinforced-concrete drywell floor and will extend into the suppression pool to provide a flow path for uncondensed steam
- Into the water.
Each vent opening will be shielded by a steel deflector plate to prevent overloading any single vent by direct flow from a pipe break near that particular vent. The steel 4
-~- ~ t . i deflection plate will double as a temporary seal during the pressus! 3 l test of' the drywell floor and also to prevent foreign objects from i entering the vent. The vents will be structurally tied together by supports to provide resistance against forces which will be f developed during the postulated LOCA. Vacuum breakers will be provided to equalise the static pressures between the suppression chamber and the drywell and to provide a centro 11ed return flow ( path from the suppression chamber to the drywell to assure design operation of the suppression chamber in the event of a small steam f 1 l leak. The applicants have supplied the results of the LOCA containment i pressure transient based on the new vent flow model. The new model s is documented in the General Electric Company Topical Report NEDo-10320 "ne General Electric Pressure Suppression Containment i Analytical Model," April 1971, and Supplement I to NEDO-10320, ) May 1971. i De significant areas of our primary containment review are dis-cussed below: (a) Post-IDCA Containment Pressure: As a result of the continuing review of the method of calculating the post-LOCA peak drywell containment pressure', a revised analytical model of the con-tainment pressure transient was developed. The applicants 1 t I t I
_ _ _ _. ) recalculated the LOCA peak drywell pressure using this refined I g, I vent flow model. The recalculated peak drywell pressure and -l peak deck differential pressure increased only slightly (less l than 2 psig) over the earlier calculations. The pressure-i increase for Zimmer will be small because the suppression i i system will have a large vent area to break area ratio (twice as such' as for Limerick), which makes the vent flow model I less sensitive to the size of the primary system rupture. An extensive review of the containment design and blowdown model led us to the conclusion that a 152 pressure margin should be added to the peak calculated drywell pressure and -a 30%' pressure margin should be added to the calculated peak deck differential pressure. The margin will provide for a limited number of unknowns such as a somewhat different acci-dont from the design basis accident and errors associated l with the calculation and assumptions used. These design margins were applied to the Zimmer containment system and the following containment design ps.rameters were found accept able : Calculated Design Margin Containment Peak 38 45 20% pressure psig Peak Deck Dif ferential 15 25 55% pressure psig
4 I
- l-l 1
l ' l (b) Drywell Deck Desig,n: The drywell deck must be designed to listic the likelihood of bypass of blowdown steam from the drywell directly into the air chamber above the suppression pool. Short circuiting of the pool would produce significantly higher pressures in the suppression chamber and the drywell. Several potential bypass areas have been identified. These are the periphery joint of the drywell finor to containment wall, joints between the downconers and the dr>well floor, and cracka in the reinforced concrete floor. The concrete drywell floor will have minimal amounts cif cracking as a result of 60 Kips per foot of hoop prestresoing on the slab. The loads due to prestressing and thermal expansion will tend to seal the containment wall periphery joint and any developed crack. The downcomer vents that penetrate the dryvell floor will be equipped with welded seal plates and anchored to the concrete at the top and bottom of the slab. In addition they will be epcxy-coated to improve bonding capability and reduce leakage potential. Based on the drywell deck design, we conclude that the potential for bypass leakage is small. (c) Wetvell-to-Drywell Vacum Relief Valves: The relief valves will be positioned in the downcomers and are designed to withstand dynamic loads associated with a 1,0CA. Howe ve r, i 4
, the valves are possible leakage paths from the drywell to the suppression chamber air space. The valves will be continuously monitored and remotely operable for testing from the reactor building. Failure of a valve in an open position I 2 could usult in a potential bypass area of about 0.42 f t -with a consequent 4 psi increase in peak containment pressure. This pressure increase is well within the design margin of the containment. (d) Dryvell Deck 1,eakane Test: A leak rate test program will be developed to verify the leak tightness of the concrete deck. We will review the detailed test program during the operating license review to assure that an acceptable test program is developed. (e) Primary Containment Penetrations: Penetrations of the primary containment will be in accordance with current design criteria. The applicants' instrument line isolation system is designed in accordance With AEC Safety Cuide Number 11 and, therefore, 1s acceptable. (f) Atmosphere to Primary Containment Vacuum Relief: The concrete containment structure is designed to accommodate pressures of approximately 2 psi less than atmospheric pressures. The
ultimate capability of the containment structure will be in excess of 5 psi (9.7 paia internal). Studies showed that I the maximum cooldown rate of the post-IDCA drywell atmosphere by containment spray produces a 2.5 psig pressure differential. I Accordingly, the applicants are not providing vacuum relief valves between the'inside of the primary containment and the reactor building atmosphere. We find this acceptable. 3.3.3 secondary containment The secondary containment structure (Reactor Building) is designed i to limit release of airborne radioactive materials and provides { l l for a controlled release of building atmosphere so that offsite ) doses from the nostulated design basis accidents will be well below 10 CFR Part 100 guideline values. The Reactor Building will enclose I the reactor and its primary containment. The Reactor Building i exterior walls and superstructures up to the refueling floor will 1 j be constructed of reinforced concrete. Above the level of the refueling floor, the building structure will be fabricated of industrial steel members, insulated siding, and a metal roof. Joints in the superstructure panelling will be caulked during installation to assure leak tightness. Penetrations of the Reactor Building are designed to have leakage characteristics con-sistent with leakage requirements of the. entire building. The 1 design criteria is to provide a leaktightness that would limit l k 1 q
7: - L inleakage to 100% building voluse per day at 1/4 inch water (vacuum) while operating the Standby Cas Treatment System. The following sections present information on design features and . equipment which contribute to the leakage control capability of the Reactor Building. i i 3.3.3.1 Reactor Building Recirculation System The reactor building recirculation system will be provided to assure that mixing of the reactor building atmosphere occurs in the event of a postulated loss-of-coolant accident or the refuel-ing accident. Within ten seconds after receipt of an appropriate signal (high pressure in the drywell, high radiation in the reactor building exhaust ventilation duct, or manual-initiated signal from the control room) the reactor building will be isolated froma the outside atmosphere. Both the recirculation system and the Standby Cas Treatment System will start automatically on isolation of the reactor building. The recirculation of reactor building air, following an isolation signal, utilizes the normal ventilation systes ductwork which has been sealed to prevent outleakage from the building. The normal ventilation fans will be shutdown and one of the two 100% capacity redundant reactor building recirculation fans will be started. The fan auction will draw air from the areas above and below the
l i ~ refueling floor and discharge the air throughout the reactor f building. The flow rate in 'the main duct will be about 80,000 lj CFM. A small fraction, about 2300 CFM, of air will be exhausted I*i to the SCTS in order to. maintain the alight negative pressure [ in the reactor building. I f 3.3.3.2 Standby Cas Treatment System l The Standby Cas Treatment System will consfat of two parallel [ j process systetas designed to meet seismic Class I requirements. Each process system will have a minimum capacity of 100% of I l reactor building volume per day. Each filt e r train of the I, dual process systems will haves a desister for removing excess . moisture; a prefilter capable of removing 80 to 85 percent of f-particulate; electric heating coils to reduce the relative humidity of the ses entering the charcoal filters to less than 70 percent; a high efficiency particulate filter (HEPA) capable of removing 99.97 percent of particulate matter that is 0.3 micron or larger in size; an iodine filter (impregnated activated carbon bed) capable of removing not less than 99 percent of iodides; a mixing device, utilizing the vortex principle to assure a howgeneous mixture of air and the filtered gases; an additional iodine filtet identical to the one above; and an additional HEPA filter identi-cal to the one described above. Both s tandby gas t reatmen t trains f
L l (' will be started automatically following receipt of an appropriate signal (high drywell pressure, high radiation or by manual activation). The applicants have indicated that the iodine filter efficiency of each filter will not be less than 99.9 percent as a result of the charcoal bed increased depth and due to a gasketless, welded seas type design that olisinates the bypass of air around the I charcoal bed. The new gasketless deep bed filter design has been evaluated and a filter efficiency of 99% for two filters in series was found acceptable. We conclude that the design of the SCTS is acceptable. q 3.4 Enzineered Safety Features l 3.4.1 ' Core Standby Coolina Systems 3.4.1.1 Introductf,g The AEC Regulatory Staff has conducted a general reevaluation of ') the emergency core cooling systems (ECCS)* for light water reactors. Analytical methods and models were the principal areas considered ~ in the review. Experimental results accursulated over the past
- The ECCS for the GE 1969 product line plants is referred to as the core standby cooling systems (CSCS).
i 64 - 1: I several years and their applicability to these methods and models I were also examined. Technology associated with analyses of CSCS performance capability j has developed substantially in the past five years. As a conse - 'I quence of this expanding techrology, improvements in analytical techniques have been evolving. Some of these improvements were i j incorporated into the evaluation techniques used for facilities already licensed for operation and for other facilities for which ~ 1 the regulatory review was essentially complete. Recently, addi-tional changes and improvements have been' made in these analytical ' l techniques for use in analyses of loss %f-coolant accidents. ij These changes and improvements are documented in topical reports s4mitted by the General Electric Company. As a result of these changes and because of the results of some preliminary safety ) research experiments, we have reevaluated the effectiveness of \\ core standby cooling systems in light water reactors. Recent experiments (December 1970) performed by the Aerojet Nuclear Corporation (ANC)* in support of the Loss of Fluid Test (LOTT) program revealed that certain then-current analytical techniques and assumptions were inadequate to describe certain phases of a blowdown experir.e' t in a small-scale test ri ;. 1h es e expe rir.en ts, n t i
- Formerly the Idaho Nuclear Corporation I
m i ~ r l l 1 'which were part of a series designated as the Semi-scale Blowdown tests, were intended to provide data that could be used to check certain analytical techniques to be used in conjunction with design of the LOFT program. The LOFT program and associated separate effects tests (including the Semi-scale Blowdown tests) are designed to investigate loss-of-coolant phenomena for pres-surized water reactors. We have reviewed the results of the Seal-scale Blowdown tests and have concluded that although they provide useful input to the continuing development of analytical techniques, the results cannot be applied directly to power reactors in general, and that specifically, they are not relevant to the evaluation of the perfomance of core standby cooling systems for boiling water reactors. On June 19, 1971, the AEC issued an Interim Policy Statement con-taining interim acceptance criteria for the performance of core standly cooling systems in light-water nuclear power plants. Section IV of the Interim Policy Statement states than an accept-able evaluation model for General Electric reactors is given in Appendix A, Part 2 of the Statement. The applicants and CE per-formed additional loss-of-coolant calculations using the criteria and assursptions set forth in Appendix A, Part 2. This section dis-cusses the results of our review of the Zinsner CSCS in accordance with the Interim Policy Statement.
G~ - 1 -t l y ? 1.. { 3.4.1.2
System Description
l CE's 1969 product 'line CSCS is significantly different from the CSCS for other BWR plants (1967 product line) for which constructic, permits and operating licenses have been issued. A comparison of ~ the core standby cooling systems for '69 and '67 product line l plants is presented in Table 3.1.1. The 1969 product line core f standby cooling systems consist of the High Pressure Core Spray Systes (HPCS), the Automatic Depressurization System. (ADS), the Low Pressure Core Spray System (1.PCSS), and the Low Pressure Coolant Injection System (LPCIS) which is one mode of operation of the.RHR System. The various systems are initiated automatically j by a high drywell pressure signal or a low reactor vessel water level signal, with the exception of the ADS system which requires coincidence of the two signals. The CSCS is designed to provide adequate core cooling for the complete break spectrum up to and including the design basis break which is the complete double-ended circumferential rupture of a recirculating pipe. The design b sis break for Zimmer is 2.2 f t which is significantly smaller than the approximate 5 f t design basis break for the 1967 product line plants. E The new HPCS system will combine the function of the former high l pressure coolant injection (HPCI) system and one of the core f M
l L - 67'- spray systems (CSS) of previous BWR plants. A single motor-l driven pump and associated pumping and instrumentation will be used to perform both functions if offsite power is lost. Power [ 4 for the NPCS pump will be provided by a separate diesel gerarator. l HPCS injection coolant will enter the vessel and be piped to a sparser crer the reactor cora via two entry points near the top of the shroud. Nozzles spaced around the sparser will spray the water over the top of the core and into the fusi assemblies. This injection mode, through the core spray sparger rather than through the feedwater sparger, represents a significant departure from previous BWR designs. The system's core l cooling capabilities are designed to function over a wide range of reactor coolant system pressures and break mises. For small breaks that do not depressurine the reactor vessel rapidly, the system will asintain reactor water level and depressuriae the vessel. For large breaks, rapid depressurisation will occur and the NPCS will cool the core in the spray cooling mode untH suffi-cient inventory is accumulated to terminate the transient. The pump characteristics are selected to satisfy requirements for both the high pressure and low flow rate for small breaks and los pressure and high flow rate for large breaks. When the cooling system is activated, the initial flow rate will bc i
I i .I l - established by primary system pressure. As the reactor pressure .l falls, the flow rate will increase until required core spray flow rate is achieved when the vessel reaches 200 paid (dif-ferential pressure between the reactor vessel and the primary containment). The HPCS systems NPSH requirements will be met l without reliance on a pressurized primary containment. j -{ i The ADS will remain the same as for previous BWR systems. The system will. rapidly reduce reactor pressure and enable the low pressure standby cooling systems to function. The ADS will utill e six of the. relief-ssfaty valves in the nuclear pressure relief I i system. These valves are actuated by coincident signals of low I l water level and high drywell pressure. The ADS will not be acti-vated unless oither the 1.PCS or LPCI system is operating. t The 1.PCS system will consist of a centrifugal pump powered by either normal of fsite power or by the onsite diesel-generator. The system will be identical to other BWR LPCS systems except that only one loop will be provided. The HPCS system operating in the low pressure mode will serve as a redundant core spray loop. When the reactor vessel pressure is low enough, watet from the suppression pool will be piped to a spray sparger above the core. This sparger will be separate and distinct from the HPCS ( l sprny sparger. The 1.PCS pump will be located in the reactor t 4 L___ i
building below the water. level in the suppression pool to amoure positive pump suction. NPSH requirements can be met without-reliance on a pressurized containment. The LPCI system will inject water from the suppression pool directly into the core region through three separate nozzles to flood the core. Previous BWR designs used the recirculation loop piping to inject the LPCI cooling water. The LPCI system will be initiated by the sans signals as the HPCS and LPCS systems and will operate inde-pendently to cool the core. The low pressure coolant injection system is one mode of operation of the Residual Heat Removal (RHR) system. The pumping system is designed to provide both-adequate head and coolant flow capacity to meet flooding requirements for the entire break spectrum. The use of suppression pool water will establish a closed loop for recirculation of LPCI water. Because I the LPCI will supply water directly to the reactor vessel, the recirculation loop selection logie used on previous BWR designs l to sense break locations will not be required. All of the subsystems listed above require the availability of either offsite or onsite power. While the reliability of the CSCS performance would be improved if offsite power is available. 2 i
r l j - 70 l the performance of the CSCS is evaluated assuming only onsite electrical power is available. In addition, the performance capability of the CSCS is shown to be adequate assuming a failure of any active component within the CSCS. This single failure criterion is applied coincident with the assumed loss of of fsite
== power. 1: 3.4.1.3 Discuss ten of CSCS Review
== The procedures used by CE to analyse the consequences of a LOCA depend upon the particular break size and the location being evaluated. It has been shown by CE that the 'wors t-case' situa-tion (i.e., highest cladding temperature) arises for a break in the coolant recirculation lines because they have the potential for ecusing the coolant mass loss from the vessel to be mere rapid and more extensive than for break,s in other lines in the reactor coolant system that would carry either steam or two phase fluid mixtu'res. Based upon our review of the CE analyses we have found that the ' worst-case' situation, with regard to assessing the performance of the CSCS. would be for an instantaneous break of a large recirculation outlet line. For 'the purposes of analyses, the changing thermal and hydraulic phenomena that are annociated with a design basis loss-of-coolant accident (LOCA) may be described in five phases: (a) temperature changes and heat removal during reactor blowdown with associated flow coastdown, (b) achievement of critical heat flux at any . point on the fuel rod cladding and associated teurerature rise of fuel and clad material. (c) lower plenum flashing causing a temporary resurgence of core flow,* (d) temperature rise of fuel and cladding with diminished cooling and complete depressuri-zation, and (e) temperature changes and heat removal during CSCS operation. The analysis of each of these phases of the 14CA originally were performed using calculational models and techniques different from the models and techniques currently used by the app 11canta (and GE). Since the original analysis additional information and results of tests related to the per-i fo. mace of the CSCS systems also have become available. As discussed in the following sections, we have met with CE on many occasions to review details of the codes, models, and analyses. in addition, independent checks of certain portions of these cal-culations using different codes have shown reasonably good agree-ment with the CE results. The first phase of the IDCA is the short-term blowdown during which energy is removed from the core by coolant passing through the core and exiting through the postulated break, causing the reactor coolant system pressure to decrease rapidly. Initially conditions are nearly the same an during normal operation and nucleate boiling continues undisturbed. In the nucleate boiling
- This phase occurs when the fluid in the lower plenum reaches a saturation condition resulting in a rapid expansion of the fluid causing a large flow increase in the core region.
l I ... i !i 4 regime, a heat transfer coef ficient of about 3 x 10 stu/hr-f t2*F s q i 1 is calculated to exist during this' period of time which is slightly less than 9 seconds. ~ A short time later, the, core flow and system { press' era decrease sufficiently that nucleate boiling cannot be sus- } ~ l-tained and the heat transfer rate decreases markedly. In previous l j l analyses a " dry-out" model was used to calculate the time at which I i the degradation,in heat transfer occurs. This model was based on 9 l the results of tests in which the flow in a heated test section was stopped simultaneously with initiation of depressurization. Because the flow in a BWR core af ter a pipe break is expected to c l coast down rather thaa to stop instantly, degraded heat transfer will occur later than would be predicted using the " dry-out" model. In the current analyses, the time at which the departure f rom the 'l nucleate boiling occurs is determined using empirical correlations i based on the results of steady-state critical heat flux (CHF) tests. l ,l Additional information received from CE confirms the conclusion l l that the use of a steady-state correlation is appropriate. CE has performed transient tests to demonstrate the validity of their steady-atate CHF correlation during transient conditions. The tests were performed in which flow and pressure reduced separately and concurrently. Comparison of the CilF measured in the tests with the CllF predicted by the steady-state correlation show that the use of the correlation gives results that are conservative. l r _ _ _ _ _ _ _ _. _ _ _ = _.. _
i j : The short-tern blowdown phase ends when the coolant flow through the core is assumed to stop as the water level in the downconer region reaches the inlet of the jet pump. Even though flow actually would continue at a very reduced rate, for conservatism in the analysis, the flow in the core is assumed to stop at this time. During this period of f1w stagnation, the heat t rans fer coefficient used in the analysis is assumed to be equal to zero. This implies that no heat transfer by conduction or convection from the fuel rods to the coolant occurs and the 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 i phase; which exists for about 3 seconds. The CE analysis yields 4 a maximum calculated cladding temperature during this period of i -1 about 1500*F. Similarly, CE calculates that the pressure in the I primary system during this second phase of the postulated LOCA is continuing to decrease because of mass loss through the tireak, and i the depressurisation rate during the short-term blowdown regime is of the order of 10 psi /sec. We conclude that the use of a heat transfer coefficient equal to zero for this period is very conservative. In the third phase of the LOCA, lower' plenum flashing occurs. 1 This is a flow phenomenon during the blowdown wherein a sudden l -l l 1 1 ~
f' f I i transient increase in the core flow begins a few seconda af ter [ the core flow has decayed to near zero. The increase in core l flow results when the liquid level in the vessel drops below the i -{ recirculation line suction nozzles. causing the flow out the break to change from a liquid phase to a steam phase and increasing the t rate of depressurization of the system (to about 30 to 40 psi /sec). i This rapid depressurisation results in a rapidly changing thermo-t dynamic state of the fluid in the primary system. Because the l fluid in the lower plenum beneath the core was initially in a subcooled state (by about 24 Beu/lb), it does not change thermo-L dynamic state during early blowdown as does the rest of the fluid system; however, when the system pressure decreases to the level where this fluid flashes to steam a large increase in steam flow through the core results. This period of the 1,0CA is thus called " lower plenum flashing." Calculation of flows, temperatures and pressures during this phase depends on the knowledge of the flash-ing process, the effect of flow maldistribution, the resistance to flow of a two-phase mixture through the core and jet pump dif-l' i fusers, and the rate of blowdown through the break. ) i During this period of increased core flow, CE assumes that nucleate I boiling is reestablished and that relatively large heat t rans fe r coef ficients result. Although nucleate boiling amy be reestablished.
~ 75 - there is insuf ficient experimental evidence to support this asstssp-tion. The efore, we asked CE to perform the analysis with the assumption that only stable' film boiling occurs, with greatly reduced values' of heat transfer coefficient, as determined by the Croeneveld correlation, during the period of lower plenum flashing. This assumption is in accord with Appendia A, Part 2 of the coussission's June 19, 1971, Interim Policy. statement on the core standby cooling systems. Following the period of lower plenus flashing it is conservatively asstmed that no convection cooling occurs. Heat generation, pro-duced by the radioactive decay of the fission products, and therent radiation among the fuel rods causes the core to heat up. The resuate presented in the applicants' August 4,1971, submittal show the calculations of fuel clad temperatures in the core for four fuel rod groups. We have reviewed the calculations for this r period and conclude that the predicted thermal responses calculated during this phase are conservative. Although the loss of water level or the increase in drywell pressure resulting from a pipe break is sensed issnediately and the CSCS is 1 )- signaled to start, the actual injection of water by the low pressure systems does not occur for about 30-40 seconds. This time period is required for the diesel generators to start and accept load, the reactor pressure to fall below the CSCS pump discharge pressure and
a l} a i l the CSCS pumps to come to full flow. Water is injected into the
- 1 reactor through both the LPCI system and the core spray system.
l 1 CE has recalculated the fuel rod temperatures during the design i basis LOCA, using a new core spray model and heat transfer correls-j tion based on the data from spray cooling testa of electrically-1 l heated-full-length fuel assemblies containing rods that were clad t with either stainless steel, or sitconium (FLECHT tests). The 28 l .h results of these tests of full-length assemblies are susewrised i l [ in the report submitted by CE and titled "Ef fects of Cladding t Temperature and Material on CSCS Performance" (NEDO-10179). The i purpose of the tests was to demonstrate the effectiveness of top spray and bottoo flooding in cooling full-sised BWR sesemblies l' at temperatures and power levels representative of the Zisawr core. 4 In order to accomplish this purpose nearly 150 tests were run at conditions and with bundles that were as nearly representative of toe expected accident conditions and core configurations as possible. The tests did not simulate the blowdown phase of a LOCA, but used the expected conditions at the time of initiation of the low pres-sure core standby cooling systems as initial conditions for the Most of the tests were run using top spray and bottom tests. l flooding i. concurrently. 1 l i
j}y. i s e ' i: / 1 ~ II " l \\ I'; 5 q y j CE has devtloped a " heat transfer cor' relation based on data obtained re in the stainless steel bundle tetts. This correlation is reported i in the CE topical report "less-of-Coolant Accider,t and Emergency j Core Cooling Models for General Electric Boiling Water Reactors" 1 (NEDO-10329). The correlation acenats for the reduced heat trans-for is, the central pins of an assembly more realistically than ] 7 i \\ previous correlation, predicts the clad temperatures more accur-I stely, and forms the basis for applying' the stainless steel test i results to Zircoloy bimdles. lie have reviewed the GE correlation '4 sed its application, and a core spray' sedel developed independently J i 'by the Aarojet Nublear Corporation (ANC). In both sedels, heat f transfer by theran1 radiation from the rods is the controlling phenomenon in"cocling of the rods, and heat transfer by convection plays only a oms 11 part. The convection portion of the heat transfer is inferred from the g results of the tests on stainless steel bundles by subtracting d the thermal radiation heat flux from the total heat flux as measured in the ce.its. The thermal radiation flux is calculated using the i geometrical view factors and experimentally determined emissivity. In the CE analysis a value of 0.6 was used for the emissivity of both the stainless steel rods and channel box. Subsequen t testa l-made to measure the emissivity of the stainless steel bundles indi-cated that.a value of P,.9 would be appropriate. Since the use of l l \\ l s V
/ s. a . ya. i l the larger value of test bundle emissivity results in smaller values of the empirical convective heat transfer coefficient, we have required CE to perform the analysis of the accident using cor.vective heat transfer coefficients derived using the larger ~l l value of emissivity. This matter is treated in accordance with I Appendia A. Part 2 of the Commission's Interim Policy Statement on .I I core standby cooling systems, since the cooling is mainly_ by thermal radiation from the fuel to the channel box, which is the outer container of the fuel bundle, the peak fuel clad temperatures q reached are strongly influenced by the temperature of the Zircaloy l channel. Until the channel is wetted, by action of the spray, it b is at a high temperature and is a relatively poor heat sink. CE and ANC use different heat transfer coefficients derived from the same data to calculate the cooling of the channel prior to quenching. GE uses a correlation based on theoretical analysis to calculate the time to wet or quench. ANC estiastes a larger quench time from available data. In order to assure that the channel quench times are conservatively calculated, we have required the calculation of the fuel rod temperatures using a channel quench time which is 60 seconds longer than that determined by the theoretically-based (Yamanouchi) correlation. This requirement is in accordance with Appendix A Part 2 of the Consnission's Interim Policy Statement on core standby cooling systems. t
, The peak cladding temperature calculation assumes that cooling by spray action persists until the accumulation of water is sufficient to terminate the temperature transient by flooding action. The applicants and CE have performed these analyses for the entire pipe break spectrue, up to and including a double-ended severance of the largest pipe of the reactor coolant pressure boundary. In the limiting case of a postulated double-ended break of a primary coolant system recirculation loop pipe with the simultaneous failure of the LPCI system, the calculated maximum fuel clad tempera-ture is 1800*F, using the AEC evaluation model described in Appendia A, Part 2 of the interim Policy Statement. For smaller breaks the calculated maximum clad temperature is less; for example, a 2 break of 0.05 ft would result in a calculated maximum clad tempera-ture of 1400*F. The cladding-water chemical reaction is calculated. to be less than 0.12% of the total amount of cladding in the reactor a for all break sizes. 3.4.1.4 Conclusions We conclude that the design of the 71mmer core standby cooling system is acceptable and complies with the Interin Policy State-ment issued on June 19, 1971. The analysis shows that the conse-1 quences 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
80 -- 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 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. 3.4.2 Besidual Heat Kemoval System (RHRS) The RHRS is designed for five modes of operation as discussed below.. The system will consist of two heat exchangers, three main system q pumps, three service water ptasps, and associated valves, piping, controls and instrumentation necessary to form three independent closed loops. The pumps will be arranged and located to assure that adequate NPSH is available for all modes of operation. The RER systes loops are not interconnected; thus a failure in l one loop cennot cause a failure of another. In addition, two of the three independent loops and associated equipment will be located in a separate protected area of the reactor building to reduce l the possibility of a single physical event causing the loss of I' I the entire RHR system. The separation provided by the above arrange-l l ments assures that the various operating modes of the RHR system t' I - will always be funettonal. These modes are as follows: i 3 e i i f I I.
I 1 1 (a) The shutdown cooling and reactor vessel head spray dperating mode will be placed in operation during normal shutdown and cooldown. The shutdown cooling system will be capable of l i maintaining the nuclear system at 125'F during reactor refueling and servicing. The head spray will maintain l saturated conditions in the reactor vessel head volume by l i condensing steam being generated by the hot vessel and internals. This assures a high water level which limits thermal stresses in the vessel during cooldown. (b) The suppression pool cooling mode of operatios will be placed in operation immediately following the design basis loss-of-coolant accident to limit the suppression pool temperature to less than 170*F. This suppression pool temperature pro-vides assurance that complete condensation of blowdown eteam will be obtained. (c) The containment spray cooling system will provide additional redundancy to the standby core cooling systems for post scci-dent conditions. Spray headers located in the drywell and above the suppresi,lon pool will be capable of remving the energy from the drywell atmosphere by condensing the water vapor. The coolant spray will be collected in the bottom of i
1 1 I i the drywell until the level reaches the downcomers vents and drains back to the suppression pool. A spray ring in the sup-l pression chamber cools any noncondensible gases collected in \\ t!.. 'ree volume above the pool. The spray headers of the RHR5 will not become operable until the low pressure conditions are met. (d) The RHR system operating in the condensing mode will be only one machod of operation of the reactor core isolation cooling system. The amount of decay heat that can be dumped to the suppression pool during RCICS operations will be limited by the maximum allowable temperature of the pool (170*F). The RCICS turbine steam exhausted to the suppression pool will result in a 3'F/hr temperature rise. Therefore, decay heat can be transferred to the service water instead of the pool by operatinR the RHR heat exchanger as a direct steam condenser. r The cool condensate is either dumped to the suppression pool or returned to the suction side of the RCICS pumps. (e) The low pressure coolant injection (LPCI) system operating mode will be used to maintain the reactor vessel coolant inventory following a LOCA. We conclude that the design of the RHRS is acceptable.
4 3.4.3 Post-IDCA Hydrogen Control and Containment'Inerting The applicants cited " Hydrogen Generation in gWit's" as discuswd in 1 Dresden 3 Amendment 23 as a basis for venting the containment through the standby gas treatment system as a means for the control of hydrogen accumulation in the primary containment following a loss-of-coolant accident for limiting the concentration to pre-clude the potential hazard. They also stated that they are follow-ing the program underway by the General Electric Company to develop i a hydrogen-oxygen recombiner for resolution of potential hydrogen hasards.- We and the ACRS (see Appendix B) ~1nformed the applicants that we require a system that provides for mixing, sampling, and control of hydrogen concentration in the primary containment fol-lowing a LOCA due to metal water reaction and radiolytic decompost-tion of water. The hydrogen control system should be designed to maintain the hydrogen concentration within acceptable limits using the assump-tions listed in the AEC Safety Guide 7. " Control of Combustible Cas Concentrations in Containment Following a Loss-Of-Coolant Accident." The applicants' containment purging capability is an acceptable backup system for a hydrogen control system designed to satisfy Safety Guide 7. In amendment 21, the applicants have l agreed to incorporate these provisions in their design. We I find this consnitment acceptable, a-
I i 1 i ) l 84 - I, I s l As indicated in previous BWR reviews, inerting of the primary l containment atmosphere will provide additional margin in the con-j l l trol of the concentration of hydrogen produced by metal-water I f reactions and radiolysis in the event of a 14CA. This will extend ( l the time available to effect actiona to cope with the gas evolution.- The applicants' design includes the provision for inerting the ) { primary containment. In amendment 21 the applicants agreed to j inert the containment. We find this commitment acceptable. 1 . i 3.4.4 Lona Term Coolina Water Supply Heat removal capability during operation, for safe shutdown or I during the accident mode of operation (including loss of outside l 1 power) will be provided by means of the combination of the Servies Water System (SWS) and the Reactor Building Closed Cooling Water System (RBCCWS). The Service Water System will also remove heat rejected by the Reactor Building Closed Cooling Water System. The SWS will consist of four motor-driven horizontal centrifugal pumps, which take auction from the Ohio River. Three of the four pumps are required for normal operation, while one pump has the f capacity to cool the shutdown plant and keep it in a safe condition af ter a loss-of-coolant accident. In Amendment No.15, the appli-cants modified the design of the SWS to provide for a ' split header system to assure adequate cooling capability even in the event of a failure of an active or passive component. u__--_-__ i
1 1 l The RBCCWS will be a heat sink for various essential and non-easential I equipment located in the reactor building and in the auxiliary bay. Three 50% capacity pumps and two 100% capacity heat exchangers will be providad. In Amendment No.15, the applicants modified the design of the RBCCWS to provide for a split header system to assure adequate cooling in the event of postulated failures of pressure components. One pump and one heat exchanger will be sufficient to provide the cooling capability for plant shutdown following a LOCA concurrent with loss of off-site power. The RBCCWS will be monitored con-tinuously to detect leakage from reactor associated systems and components. i The SMS and RBCCWS which supplies essential equipment will be designed and constructed in accordance with Class 1 (Seismic) t and Quality Croup C requirements. I We have concluded that the provisions made to assure a continuous supply of long term cooling water in the event of a design basis accident are acceptable. In addition, the potential for leakage 1 ) of activity to the Ohio River 1s-acceptably low. 3.5 Protection. Cont rol and Emergency Electric Power Systems _ 3.5.1 Ceneral In our review of the reactor trip and control systems, the engineered safety feature circuits, and the emergency electric power system, ii
( , l ) the Cosumission's General Design Criteria (CDC), the Proposed IEEE Criteria for Nuclear Power Plant Protection Systems (IEEE-279) L dated August 1968, and AEC Safety'Cuides 6 and 9 served, where applicable, as the bases for evaluating the adequacy of these designs. On the basis of our review, we found that a part of the reactor protection instrumentation, installation criceria, environmental testing of control equipment, control systems, and the ins trv-mentation that initiates and controls the engineered safety features were the same as those we reviewed for the Pilgrim Nuclear Power Station and are acceptabic. The following discussion concerns matters that are new design areas, items unique to the Zimmer station, and protection system generic items which were given further attention as a result of our previous review of other BWR facilities. Specifically, the areas reviewed are as follows: s A. Protection System Generic Items 1. Independence of Redundant Channels 2. Capability for Periodic Testing 3. Incident and Accident Surveillance Instrumentation 4 Annunciation of Engineered Safety Feature Bypasses
1 i l l 1 5. New Design Items 1. Control Systems Solid State Manual Control System a. i b. Solid State Rod Position Circuitry Recirculation Loop Valve Control System c. 1 I 2. Reactor Trips Deletion of Condenser Low Vacuum Trip and /Jdition a. of APIM Startup Reactor Trip at 15% Power 3. Core Standby coolins System Initiation and control l Instrumentation C. Itema Unique to the Zimmer Station 1. Emergency Electric Power Systems i 2. Physical Separation Criteria for Protection and Emergency Power Systems s 3. Seismic, Radiation, and Environmental Testing 3.5.2 Protection Systems Generic Items 3.5.2.1 Indepetidence of Redundant Channeln One of the applicants' safety design bases requires: "There shall be sufficient electrical and physical separation between channels and between logica monitoring the same variabic to pre-vent environmental factors, c1cet rient t ransients, and physical )
.l ! from impairing the ability of the system to respond even ts co rre c t ly." Our review of the protection system design revealed several areas that do not satisfy this requirement, nor those of IEEE-279. The concern is the connection of. redundant protection channels to single switches and terminal boards in the control room panels. This compromises the protection channel independence and increases their vulnerability to single random failures. The app!! cants have identified four instances where redundant protection system channels are connected to a single switch and have agreed to provide some additional physical separation for these protec-J tion system switches. The method and type of additional separation will be reviewed again at the operating license review. This commit-asnt is acceptable. 3.5.2.2 capability for Periodie Testing Previous BWR designe did not include the built-in capability to permit periodic testing at power. Testing uf the reactor pro-tection and engineered safety feature circuits required manual interference with circuits (removal of fuses), or the perturbing of reactor parameters to assess the operability of the f stru-mentation. The reactor protection system and the engineered safety feature circuit designs for 2f amer will include the inherent capability for complete periodic testing during power operation. t_______-------------- J
i i r. - 89'- 3.5.2.3 Incident and Accident surveillance Instrumentation The BWR reactor protection and engineered safety feature instru-4 mentation channels generally use blind sensors and, therefore, do not provide continuous readout in the control roon of the para-meters being monitored. The applicants will prcwide redundancy of instrtaments monitoring vital parameters necessary to assess and mitigste consequences resulting from anticipated operational occurrences without loss of function, and they will comply with the requirements of IEEE-279. We consider diverse means for monitoring plant conditions (such as rod position determination by neutron monitoring, rod position indicators and computer read-out of parameters) as meetin's the requirement for redundancy. We find this acceptable. l 3.5.2.4 Annunciation of_ Enmineered Safety Feature Bypass i Our review of the Zimmer design revealed that annunciation of the bypass of an engineered safety feature resulting from a deliberate operator action was not included. The app 11canta stated that they considered administrative control as an effective and ade-quate means to identify these bypasses. We informed the applicant i that we do not consider administrative controls an acceptable snbstitute for the indications required by IEEE-279 and that we will require that the design of these circuits include control i h
) y 1 room annunciators to indicate wnenever operator actions result in the loss of an F.SF function or a reduction in system redundancy. The applicant has agreed to incorporate this feature into the design of the systas. - 1 3.5.3 New Desimp Items t 3.5.3.1 Control systems The reactor manual control system will act on the electro-hydraulic control rod drive system to position control rods in the core. I i This system design has been modified from relay logic control cir-evitry to self monitoring solid-atste circuitry. The electro-hydrauli drive system, however, will remain unchanged from previous designs. ) l This change in design will substantially reduce field wiring and should improve operability through the self monitoring feature being incorporated. Our evaluation is primarily concerned with assuring that the reactor protection system is capable of preventing fuel damage with control system failures or during operational l occurrences. l The applicants have identified the primary design changes in the proposed solid-state circuitry from those for the previous BVR relay logic design. He has stated that single failures will not result in transients that are beyond the capability of the reactor protection system. Additionally, he has described the prototype qualification tests that have been completed and those still to be performed. 1
- We have concluded that the applicants' design criteria and planned testing program are satisfactory. The technique of using sequential coding signals is new to the nuclear power plant control system designs and therefore we will perform an in-depth evaluation of the design during the operating license review.
The control rod position.information circuitry design has been i changed free relay logic to solid-state circuitry. The motiva-tion for this change is also to reduce field wiring and instal-lation problems. The rod position detection and display remain ic unchanged from previous designs except for an apparent increased reliance on the plant computer. j \\ l The applicants have stated that the use of the computer has no safety significance on the plant. We plan tg investigate this aspect of the use of the computer further during the operating license review. s The recirculation loop flou control has changed from that used in previous BWR designs (see section 3.2 of this report). Accordingly, the design of the recirculation pump and loop isolation valve interlock circuitry have also changed. The circuitry for this new flow control system uses an APRM flux (protection system) signal to derive an error signal from the i master power controller which is fed to the flux controller. The
e I flux controller then changes flow in each loop (and reactor power. by modulating the control valve. Our review of the proposed syste indicates that the APRM flux signals used for control are deJignec in accordance with the requirements of Section 4.7 of IEEE-279 and are acceptable. 1 ) 3.5. 3.2 Resetor Trips The applicants' proposed design delstes the direct reactor trip i derived from instruments monitoring condenser low vacuum. An i indirect reactor trip is provided by a condenser low vacuum signal ~ which closes the turbine stop valve. We informed the applicants that although the turbine stop valve meets the current criteria, they must provide a condenser vacuum instrument monitoring system that satis fies the requirements of IEEE-279. The applicants have agreed to provide an additional reactor trip that satisfies this requi remen t. A condenser low vacuun signal will be used to close the MSL isolation valves which, in closing, trips the reactor. We conclude that the reactor trips seet the appropriate require-ments and are acceptable. Additionally, the applicants plan to modify the APRM channels by extending their effectiveness into the startup range and includ-i ing a reactor t rip at 15% power. In previous BWR designs, the t f t 4
__ _ I ) Ar*0f channels were made effective only in the run mode. The I i applicants have stated and will provide docusen6mtion that these
- * ~
d changes have no safety significance and are being included in the d design to preclude the necessity for recalibration of the Inter - mediate Range Monitors (IRM), due to changes in responses result-ins from control rod motion in the vicinity of the igm detector. This consitment is satisfactory for the construction permit review l and will be followed closely during the operating license r view. 3.5.3.3 gre Standby coolinz System (cscs) Initiation and cont rol Ins t rument ation The Core Standby Cooling System was designed by the applicant to meet the requirements of IEEE-279 and -308; the General Design Criteria; and General Electric's criteria for separation,'tente-l bility, and qualification testing. j I The CSCS is to be divids into three redundant divisions such that the loss of any one will not preclude adequate core cooling over i the complete range of break sizes. This arrangement will resu?.t 'l in a balanced three diesel - three bus electrical distribution system. Additionally, the initiating sensors are uniquely assigned to each division, thereby avoiding the interconnections and assuring separation and independence of these redundant divisions. To pro-vide further assurance of separation, the location of sensors with i
/ t l 1. respect to elevations and azimuth around the containment was coca ide red. The releys and logic circuitry will be arranged in three panels, one for each division, This new C5CS design contains diversity is both the initiating signals and the subsystems. We conclude that this systes is designed to referenced criteria and is acceptable. 3.5.4 Electric Power systeso ).5.4.1 Of f eito Power The Zimmer Nuclear Power Station will be interconnected to the transatssion systes throegh 345 kV circuits. Power from the l unit's generator will be fed via a single circuit containias che mais step-up transforms to the 345 kV switchyard. The 345 kT evitchyard will be arranged in a breaker-and-a-half config.sration. Two 345 kV transmission circuits will emanate from the switchyard. A third 345 kV circuit will be added prior to plant operation. The 345 kV system (through the swite3 yard and reserve auxiliary transformer) represents one source of offsite power to the ensi-neered safety feature and safe shutdown loads. This of f site power source will be automatically connected to the emergency buses sup-plying these loads. A second source of offsite power will be sup-plied from the 69 kV transmission system via a circuit which ecenects 1 1
L , to the emergency buses through a separate reserve auxiliary t rans former. This second source of offsite power will also be l. capable of supplying all engineered safety features and safe shutdown loads. The applicanes have eemptsted atudles concerning the stab 111ty of their system with respect to the effects of 'the loss of the unit of this station or the largest remaining generating unit. From these studies they have concluded that offsite power to the ensi-J seered safety features will not be lost. t Our review of the independence of the offsite power circuits between the grid and the emergency buses revealed that a single i l failure of a 345 kV line or tower could also result in the loss of the 69 kV circuit, since a portion of the 69 kV circuit crosses beneath the 345 kV lines. The applicants have stated that they I can reinstate a second independent 69 kV line in less than 30 min-utes and that an analysis will be provided to demonstrate that the facility can withstand the loss of all off-site and on-site power for a period of two hours. Based on the above, we conclude that the offsite power systems satisfy the CDC 17 and is acceptable. 3.5.4.2 Onsite Power '1he design of the standby (onsite) emergency power system utilizes the split bus concept in accordance with AEC Safety Guide 6. '!he l 1 1 1
i ! redundanc engineered safety feature equipment t rains are divided into three divisions with each division assigned to an emergency 4,160 volt bus. (kie diesel generator set is connected to each b us. The three redundant diesel generator sets are each.individw-ally housed in a seismic Class I room such that there is complete physical separation between units. Auxi!!ary systema for sach of these machines are physically and electrically independent of each other and housed is Class I rooms. The fuel supply systen will be of suf ficient capacity for operating all three diesel engines at required loads for 7 days. The applicants have stated that the criteria for selection of diesel generators will include the regulatory positions of AEC 4 Safety Guide 9. Three d-c sys tems are to be provided. The 125 volt d-c systen v111 consist of three redundant batteries, each with its own charger and distribution system. Further, each battery is located in an independent Class I room with separate ventilation systems. A separate portable battery charger will be provided which can be substituted manually for any normally installed charger. The batteries will have a capacity to supply all assigned loads for a minimum of one hour. This 125 volt system is designed to be
_ compatible with the three-division engineered safety feature load grouping discussed previously. The remaining d-c systems consist of 1) two redundant 250 volt battery systems and 2) two redundant 24 volt battery systems. %e 250 volt system will supply power to larger d-c loads such as pumps and large valves. These redundant 250 volt battery systems will be assigned to two of the three engineered safety feature load divisions such that single failures will not preclude minimum safety engineered feature operation, The two redundant 24 volt battery _ systems supply power to the neutron monitoring system. The 250 volt and 24 volt battery system designs and installation will be identical to that of the 125 volt system. The applicants have stated, and we agree, that they will meet the requirements of CDC 18 with respect to providing the capability for periodic testing and inspection of the emergency power systems. s 3 We conclude that the design features of the standby (onsite) power systems (both a-c and d-c) for the Zinner facility satisfy the regulatory positions of AEC Safety Guides 6 and 9 and the require-ments of GDC 17 and CDC 18.ind are acceptable.
'k 1. 3.5.4.3 Cable Installation Criteria for Protection and Electric Power Systems i The applicants have dxteented the criteria for cable installation, deatsn, selection, and routing as well as the criteria for identi. fication of safety-related switchgear, control panels, components q and interconnecting viring and trays. We find that these criteria are acceptable. 3.5.4.4 Seismic. Radiation. and Environmental Testina The instrumentation, control and electrical systes environmental I testing criteria indicate that all safety-related devices supplied by General Electric would be seismically tested with the exception of the primary pressure boundary devices which would be analysed. The applicants have agreed to test or provide test data for all devices supplied by Sargent and Lundy and the primary pressure boundary devices. We couelude that sufficient information will be available for verification of ' design criteria during the operating license review. l i The criteria also includes the qualification testing of all safety related equipment located inside the primary containment at the extreme combined temperature and pressure conditions of the design basis loss-of-coolant accident. We have reviewed the criteria and conclude that they are acceptable. l h I I i 1
l i j. The criteria for radiation qualification of all safety-related devices have been reviewed. We conclude that they are the same as those for previously licensed plants and find them acceptable. 3.6 RADWASTE SYSTDI 3.6.1 General The applicants have described the systes to be provided for the Zianer facility for treatment of gaseous and liquid radioactive { waste discharge. These systems have been designed to reduce the radioactive effluents such that the annual average concentrations released to the environment are less than 12 of 10 CFR Part 20 limits for each system. The liquid and solid waste disposal ) systems provided will be substantially the same as those systems provided for other BWR's. The gaseous radioactive waste disposal i eysten has been improwd over previous systems by providing an ) additional gaseous holdup, system that will decay krypton isotopes i s for 24 hours and menon isotopes for 18 days. This improvement in the delay system will provide for a much longer holdup to permit decay of short half-life radioactive gases prior to their release to the atmosphere. 3.6.2 Liquid Wastes The liquid radvaste system will collect, monitor, process, store, and provide controlled discharge of all radioactive liquid waste. The system components consist of storage tanks, waste deminer-alisers, filters, and evaporators similar to those used on other l J .O
i 4. r - 100 - BWR facili ties. The liquid radwaste treatment sys tem is designei to provide the maximum practical capability for recycling process wastes to the reactor system and thereby reduce the need to discharge radioactive 11gulds to the environment. This will be i t accomplished by separating the various liquid waste sources into three.nsin classifications with specialized treatment for each. These classifications are high purity, low purity, and A cheat cal waste. Cross connections among subsystems provide i l additional flexibility for processing the waste by alternste I methods. High purity (low conductivity) liquid wastes will be primarily i collseted from equipment drain sumps. Liquids from this source will be processed by filtration and ion exchange through the vaste filter and dominera11:ers and then transferred to the condensste storage tank for reuse as makeup water. Reprocessing of high conductivity or highly radioactive liquids will be accomplished by the demineraliser system or by the radwaste evaporators. Low purity (high conductivity) wastes from floor drain sumps will be co11ceted in the floor drain collector tank. These wastes gen-erally have low concentrations of radioactive impurities and are processed in the same manner as high purity wastes. 1 5' 3 I
i / u j - 101 -- i dl ' mexical and detergent wastes normally will be processed through 3: the radwaste evaporators. Chemical distillate will be desiner-alized before being discharged to the Ohio River and the concentrates! I will be processed by the solid ra&raste sys tem. Detergent wastes will be treated with anti-foaming agents prior to evaporation. De distillate will be reused in the laundry and only the excess will ~ ' be discharged from the plant. The plant is designed so that operation of the liquid waste disposal l system will be by manual start and automatic closure of the discharge line valve on a high radiation signal. Protection against acciden-j tal discharge will be provided by design redundancy,' instrumentation and alarus, and procedural controls. Liquid wastes that are dis-charged to the environs will be handled on a batch basis through the plants circulating water discharge piping. The liquid batches will be held for a period of time to allow for decay, complete mix-ing, sampling, analysing, and processing prior to the transfer to the condensate storage tank, solid radwaste storage, or to the environs. Based on the applicants' design criteria and process objectives to allow for maximum recycling of liquid radwaste, and on the 1 applicants' analyses of concentration, we conclude that the l l i f
s - 101 - 9-. Chemical and detergent vastes normally will be processed through }- the radwaste evaporators. Chemical distillate will be desiner-alized before.being discharged to the Ohio River and the concentrates will be processed by the solid radwaste system. Detergent wastes will be treated with anti-foaming agents prior to evaporation. The s distillate will be reused in the 1 sundry and only - the excess will be discharged from the plant. The plant is deatsned so that operation of the liquid waste disposal system will be by manual start and automatic closure of the discharge.) line valve on a high radiation signal. Protection against acciden-tal discharge will be provided by design redundancy. Instrumentation and alarme, and procedural controls. Liquid wastes that are dis-charged to the environa vill be-handled on a batch basis through the plants circulating water discharge piping. The liquid batches will be held for a period of time to allow for decay, complete mix-ing, sampling, analyzing, and processing prior to the transfer to the condensate storage tank, solid radweste storage, or to the environs. Based on the applicants' design criteria and process objectives to allow for maximum recycling of liquid radwaste, and on the applicants' analyses of concentration, we conclude that che i I i
1 - 102 - .l u I activity. releaseJ from the liquid radwaste system will be less I l l than II' of the 10 CFR Part 20 limits. i i 3.6.3 Caseous vastes 1 ,i The off-gas treatment system of the plant is designed to collect, i { process, moottor and dispose cf radioactiw gaseous wastas l generated in the operation of the plant. During nonnal operation ' j the gaseous radweste system will operate on a continuous basis wi. I effective monitoring and control provided to prevent releases of activity greater than 1% of the limits of 10 CFR Part 20. Gaseous redweste in SWR's are generated from two main sources; i noncondensible radioactive gases removed from the turbine gland 9 l seal condenser, and noncondensible radioactive gases renowd from the asin condenser by the air ejector. In the absence of fuel rod leaks, N-13 from the air ejector off-sases and 5-16, N-17 and N-19 from the gland seal system are the principal contributors to the environs radiation dose. If fuel rod leaks do occur, s the noble radioactive gases, menon, and krypton become the principal contributors. Sas11 quantities of off-gas from the turbine gland seal condenser j will be removed by a mechanical vacuum pump. Exhaus t from the puerp : and the off-gas from the turbine grand seal vill enter a common two i minute holdup pipe for decay of the shortlived radioactive con-j stituents prior to being discharged to the off gas vent.
) - 103 - The air ejector off-gas is the major source of gaseous radioactive wastes and consists of activated noble gases (primarily zenon and. krypton), radioactive halogens, and radiolytic dissociated hydrogen i i and oxygen. The off-gas treatment system will incande s'high-1 temperature catalytic recombiner to combine the diamociated bydrogen and aerygen continuously from the off-gas stream of air ejector. The off-gas effluent from the recombiner will be condensed sh - and passed through a 30 minute holdup for decay of short-lived isotopes. The off-gas will be processed further by a high-efficiency filter (NEPA) before it enters the additional decay system that provides a curJe redwtion factor necessary to meet their design baats of eM syeten Before being discharged from the off-gas vent, it will be passed throt 3h a final NEPA filter. The radiation levels at the off-gas vent will be monitored continu-ously by two detectors. Should the activity of the gas being seleased exceed the reduction capabilities of the processing system, an isclation valve will be automatically closed to prevent s further flow of gaseous effluent to the environment. The appli-cents have stated that all equipment will be used to the fullest extent possible at all times to achieve the design basis for the sys ten (1% of 10 CPR Part 20). As a design basis for the gaseous waste system, the applicants have used a noble gas input equivalent to an annual average off-gas
4.. n g '{ .; s 'r. ,g !.1.,s'- y Ii t, , f f- .i 104 - ,/ ' r .l t- ,L race.of 100,000 pC1/sec based on.a M-min decay only. The addf-A* , g )l, i ;' s tional heldup system will provide for a more. complete decay of all V. h . isotopes except long-lived krypton and zenons, thereby significant! ,' l 'e 't reducing.the release of radioactivity to the environment. I .s ) \\ Two types of additional decay systems are being consideres, by the applicants. One is an -ambient-temperature, charcoal bed system 4 j l and the other is a cryogenic tersperature system. The applicants j i ] j atste that either system is caFable of meeting their design. hasta. i At ther present time we have insufficient.information to perform an t I ' independent calculation to permit va to verify the conclusion. -j 4 { [ However, as a result of additional information, (the design of the / l 4 Newbold Island, Bailey, and Limerick of tsas systems), we have ... t Y > determined that the applicants can design an additional holdup ? 1 [ . system so that the release of radioactivity fropathe station will u. meet 1% of 10 CTR Part 20. The applicants are aware that we require additional inforinstion on the design of the' decay systems. They . h,q-have agreed to provide this information when the final design hr, deren selected, but before fabrication and installation of the system. We conclude that this commitment is acceptable. 3.6.4 Solid Wastes j The solid radwaste system is designed to collect, monitor, process. [ package and provide temporary storage faellities for wet and dry s i. u i: e b ff i. 'L 8 JL'al__ -
( 4 i 1 - 105 - ) radioactive solid waste for of f-site permanent disposal. We t wastes t resulting from spent filter demineralize., deep bed desineraliser resins, filter sludge, and evaporators concentrates will be passed
- y
' to waste' tanks that serve as storage and batching, f acilities. Dry f waste consis'es of air filters, paper, rass, contaminated clothing, tools and equipment parts. Cropressible dry wastes will be compacted to reduce volume and shipped in 55-ss11on steel drums in the. 1 same manner as the noncompressible wastes. Design and operation of the solid radwsste system do not involve any unusual safety problem not already previously considered on f 1 other SWR applications. l 3.6.5 conclusions Based on the applicants' coasmitments to utilise the equipment in l the waste treatment systems to the fullest extent possible, and to maintain the radioactivity levels in affluents to 1% of 10 CFR Part 20, we conclude that the system can be designed and operated to meet the requirements of the proposed Appendix I of 10 CTR Part 50 of the consiission's regulations. 3.7 Auxiliary Systems 3.7.1 Ceneral The design bases, safety considerations, and inspection and tescing requirements of the auxiliary systems have been reviewed and we .i 1
106 - have concluded that the design concepts are accentable. The following systems were found capable of performing their intended function: Tool and Servicing Equipment; Puel Fool Cooling and Cleanup System; Reactor Building Closed Cooling Water System; Turbine Building closed Cooling Water System; Fire Protection System; Heating, Ventilating and Air Conditioning Systes; Instru-ment and Service Air Systeme; Fortable and Sanitary Water Systems; Equipment and Floor Drainage Systems; Plant Process Sampling System; Communication Systems; Lighting Syste.s; Heating Boiler; I and Make-up Wster Treatment System. 3.7.2 New Fuel Storare The storage racks for new fuel will be sufficient to hold 30 percent of the full core fuel load. The storage racks will be designed and maintained with sufficient spacing between the new fuel assemblies to assure that the array, when fully loaded, will limit the effective multiplication factor of the array (k,gg) to not more than 0.90. The fuel storage racks will be designed to Class I seismic requirements. We conclude that the design provi-sions are acceptable. 3.7.3 Spent Fuel Storage The spent fuel s torage racks provide specially designed undervater storage space for spent fuel assemblies requiring shielding and/or 9 1
- 107 - cooling prior to shipment. Pool storage space to accommiodace 160 percent of the full core fuel load will be provided. The, spent fuel racka design provides for a suberitical multipliestians factor For norussi (k,gg) for normal and abnormal storage conditions. conditions, k,gg W 11 be equal to or less than 0.M. Mr abnoras! conditions, such as dropping equipment on the racks, earthagwakes and accidents involving refueling equipment, k,gg will not exceed 0.95. i The spent fuel storage pool will be lined with statsless steel to limit the possibility of pool leakage through seams and penetra-tions. Ilo inlets, outlets, or drains are permitted that might allow the pool to be drained below approximately 10 feet above the top of the active fuel. External lines. extending below this level will be equipped with syphon breakers, check valvss, and other suitable devices to prevent inadvertent pool drainage. The pool r will be provided with interconnected channel drainage paths behind the liner veld joints to prevent preseure buildup behind the liner plate, to prevent uncontrolled loss of contaminated pool water, and to provide expedient liner leak detection and measurements. A separate spent fuel shipping cask storage area vill be provided ' adjacent to the spent fuel pool. An interconnecting canal between these areas will permit underwater fuel transfer to shipping cask, u h
I I - 108 - The cask storage area vill be cons tructed of reinforced-concrete and will be lined with stainless s teel. The applicants have stated that the cask storage area is capels of withstanding the cask drop accident; however, if an accident should breach this area, drainage would not have an adverse effect on the spent fuel pool storage area. E The spent fuel storsgs racks, spent fuel pool and the spent fuel shipping cask storage area are designed as Class I seismic s t ruc t ures. The transferal of the crane and the shipping cask over the spent fuel pool will be pro'nibited by appropriate interlocks and administrative controls. We have concluded that the design of the spent fuel storage facility meets the AEC Safety Guide Number 13 " Fuel Storage Facility Design Basis" and, there-fore, is acceptable. I 3.7.4 Service Water Sys tem The service water system (SWS) is designed to supply cooling water during normal plant operation and for all emergency conditions. Service water is supplied to the following systems, equipment, and facilities located in the reactor butiding and the turbine building: Reactor Building Closed Cooling Water Heat Exchangers; Residual Heat Removal Heat Exchangers; Turbine Building Closed Cooling Water Heat Exchangers; Turbine Generator Oil Cooler Heat Exchangers; Turbine
- 109 - Generator Hydrogen Cooler Heat Exchangers; Turbine Generator Stator Water Cooling Nest Exchangers; Diesel Generator Cooling Nest Eachangera; Service Conneetiens in Reactor Building and Turbine Building; and Radwasta Facility. The SUS will consist of four borisontal centrifugal pumpe; three of 'the four will be required for normal operation and only one will be required for safe shutdown after an accident condition. ) The SW5 pumps will be separated into two flood-proof cubicles to esintain independence. The applicants' split header systes design provides adequate cooling capabilities in the event of a single failure of an active or passive component. Service water will supply the cooling requirements for the control 1 room air conditioning equipment and the air conditioning equipment for' the auxiliary equipment room following a combined generator trip and total loss of off-site a-c power or a loss-of-coolant r accident. The SWS and related structures will be Class I seismic design, capable of safe shutduwn at river flood stages, and operable at low river flow rates (see Section 2.3 Hydrology for details). The capability of the service water system to supply long ters cooling water following a loss-of-coolant accident is discussed in Section 3.4 of this report. y 1
- 110 -
- i Based on the above information, we conclude that the SWS is designed f
to' eapply adequate cooling water for all emergency conditions and, therefore, is acceptable, j 3.8 Station Structures and Shieldina i f 3.5.1 Class t rication of Structures and_ Eeutseent l' The applicants define Class I structures and componente as thoss l whose failure might cause or increase the severity of. a loes-of-1 I coolant accident, or result in an macontrolled release of excessive amounts of radioactivity, or whose function is vital to safe shutdown l 1 1 and isolation of the reactor. The Class I structures are designed ij to withstand the applicable seismic loads resulting from the seismic it ]' events and other applicable loads without loss of function. Class II L structures and equipment are those that are taportant to reactor operation, but are not essential to the safe shutdown and isolation of the reactor and whose failure could not result in a significant release of radioactive esterist. Class III are those that are not essential to the operation, shutdown, or isolation of the reactor and whose failure cannot result in the release of radioactive material. These definitions are considered to be acceptable for identifying those vital structures and components essential to reactor safety. The potential interaction of Class II or III structures or equipment With Class I structures is discussed in Appendix I.0 of the PSAR. This appendix has been reviewed and its criteria were found to be acceptabic.
- 111 - 3.6.2 Structural Analysis and Design 3.8.2.1 Structural Analysis The containment analysis to performed utilizing piste and shell analyses. 503-11 developed by Knolls Atonde Power Laboratory, m surrr developed by Prof. Kalnine of Lehigh University, PLFDf and TDICO developed by gargent & Lundy as proprietary programs and AFDf which was originally developed by Rohn and Naas will be used. The applicants have stated that all of these prcgrams have been validated by gargent & 1l undy by selecting appropriat.e i elasticity problems for which the solutions can be obtained by manual computation or other acceptable methods, and comparing the computer solution with the established solution. On this j basis we conclude that the analysis methods are acceptable. 3.8.2.2 Structural and Foundation considerate orts The plant is located on approximately 100 feet of clayey silts l l underlain by limestone bedrock. The major plant structure will be supported on met type foundations which will be placed on engineered fill. Such foundation structures are considered ade-quate to transfer the loads to the underlying soil. For the prestressed concrete portion of the containment structure, the applicants have specified acceptable allowable stresses as applicable to the various design loading combinations I l l
- 112 - including seismie loads. For the yield limit loading combinations p the applicants have specified the load factors to be utilized. The other Class I structural components and structures will be designed to speciflad allowable stresses undar working stress design methods and checked against yield limits for load combins-tions that includa the tornado or design basis earthquake. These procedures._ in general, comply with rules of ACI 318-63, except in certain details. The applicants have stated that the more conservative rules contained in a proposed revision of this code published in the Journst of Act dated February and September of 1970 will be used. The entire containment will be pressure-tested to 52 pets, which is 151 hisher than the desisn pressure. The drywell floor will be pressure-tested initially to demonstrate its capability to withstand 25 poi differential pressure. Thes e tests are consistent with the practice adopted on previous BWR applications such as Shoreham. The containment will be instrumented with stress meters, strain meters, and deformation meters along one meridian and at the equipment hatch. The structural instrtamentation for use before and during the strue-l tural acceptance test will cennist of stress and strain meters along wi th a sys tem for measuring deformations. These ins t ruments l
L - 113 - generally will be placed on two seridians with additional instruments at the large hatch. The instrumentation program is acceptable. A statistical evaluation program is underway by gargent and Imady to justify the tendon surveillance program. We will continas our review of this ares at the operating license review when the results of the applicant'a investigation and resulting surveillance t _ program are available. We consider this to be acceptable. The Class I steel-framed structures will be designed in accordance with the AISC Specification (1969). This specification provides design rules for the working stress method of analysis (part I of AISC) and the plastic design method of analysis (part II of AISC). ] i We have considered the safety mergins associated with these design sathods for steel and concrete structures and have found them to be acceptable for the fondation at this site. 3.8.2.3 Tornado Analysis Design wind loads are based on the 100-year recurrence as defined in ASCE-3269. These wind loads are computed to produce 53 psf I for heights greater than 150 feet. The effects of the design basis tornado are translated into forces on the structure by the
l h. i - 114 - y 1 tornado model. The model consists of the simultaneous application of a 300 mph rotational velocity and's 60 mph transitional velocity, with a 3 psi pressure drop in 3 seconds._ The. velocities considered by the model are converted into an equivalent static pressure uslag the equation in ASG paper No,' 3269 titled " Wind I Forces on Structures." 'We conclude that procedures to be used to design ths fat.ility assinet the effects of a tornado are i acceptable. 3.8.2.4 Sacrificial Shield The applicants have presented information in response to the staff's question 12.3.8 (Amendment 15) indicating that the jet 2 forces resulting from a longitudinal split (138 in ) g,,f,g,, systems that penetrate the sacrificial shield will not result la failure of the sacrificial shield. The applicant has also 1 indicated that the sacrificial shield could withstand a l 2.2 ft break within the pressure vessel annulus. Design of the l 4 sacrificial shield plugs around pipe penetrations will prevent I the plugs from becoming potential missiles for primary ruptures within the pressure vessel annulus. We conclude that this is acceptable. l 3.8.2.5 Materials and Construction The containment liner will be constructed from ASTM A-516 i Crade 60 steel. The reinforcing steel used in Class I structures o t
1 - 115 - will meet the requirements of ASTM A-415, Crade 40 with the structural steel and its fabrication for Class I structures meeting the requirements of one of the following ASTM specifica-tionst' A-36, A-441, A-572 A-53, A-7, A-354, A-325 and A-490. The prostressing steel will conform to AffM - A-421, Type R4. The structural concrete will be is conformance with ACT 316-63. The user testing of reinforcing steel will be performed on a frequency of one full stae reinforcing har for each 50 tons of each heat, or for each heat if the heat is ses11er than 50 tons. Based on the information contained in the PSAR, we conclude that the materials and construction techniques are acceptable for construction of Class I structures. The primary containment drywell and pressure suppression chambers will be constructed of prestressed concrete using the 58RV System of post-tensioning utillains parallel-lay, unbonded-type tendoes. The tendons will be ' fabricated from ninety 1/4-inch-diameter, cold-drawn, stressed-relieved, prostressing grade wires with each tendon individually encased in conduit. the criteria and loads are very similar to the Trojan Nuclear plant and others recently approved and are acceptable. i h
L - 116 - I l l ,3.8.3 Mechan f eal Analysis and Design I 3.8.3.1 Seismic Input 1 t l The applicants' seismic design response spectra provides for an amplification factor of 3.5 at 22 dampins for the period [ range from 0.15 to 0.5 see and amplification factors greater i , ;1 than 1 in the period range 0.15 to 0.033 seconds. The structure I l 6 I and equipment damping is in accordance with those recommended by l our consultant. N. Newmark (see Appendia E). The modified time I l histories to be used for component equipment design are adjusted j in amplitude and frequency to envelop the response spectra spect-fled for the sits. We and our seismic consultants conclude that l the seismic input criteria proposed by the applicants provide an j t u acceptable basis for seismic design. t 3.8.3.2 geismic System Dynamic Analyses .i Modal response spectrum multi-degree-of-freedom and normal nods-time history methods will be used for all Category I strue-tures, systems, and componeats. Coverning response parameters will be combined by the square root of the sum of the squares to obtain the modal anximum when the modal response spectrum method is used. The absolute sum of responses is used for in-phase, closely spaced frequencies. Flocr spect ra inputs to be used for design and test verification of structures, systems and components !.t I vill be generated from the normal mode-time history method. A lj-l 1 1
- 117 - vertical seismic-system dynamic analysis is being employed for all structures, systems, and components. We and our seismic con-sultants conclude that the seismic system dynsmic me thods and procedures proposed by the applicants provide an acceptable basia 1 for the seismic design. 1 3.8.3.3 Class I (Seismic) Wehenical Fluid Eystems. l All Class I systeme, components, and equipment outside of the l reactor coolant pressure boundary will be designed to sustain normal loads, anticipated transienta, and the operating basis earthquake within the appropriate code allowable stress limits and the design basis earthquake within stress limits which are comparable to those associated with the emergency operating con-dition category. We consider that these stress criteria provide an adequate margin of safety for Class I systems and components l that may be subjected to seismic loadings and are acceptable. 3.8.3.4 Seismic Ins trumentation In Amendment 12, dated June 11, 1971, the applicants have indicated I that the AEC Safety Cuide Number 12 on " Instrumentation for Earth-quakes" will be met. We evaluated this material and concluded that the instrumentation is in accordance with the Safety Guide and is acaeptable.
t - 118 - 3.8.3.5 Hissile Protection The design criteria presented by the applicant.a for the design of the miselle protection features of the Zisumer facility are accept-able in renard to both internal and e:, eternal nissles. In ternal arissle protectica will be provided by separation of vital reatm-dant equipment, use of missile shielding, and proper erfer.tation, and by. the consideration of design of potential missile sources. The external missile criteria specifically provide for protection of the containment, control room, and other vital plant features. In the unlikely event of a turbine failure or seneration of tornado missiles, the -pplicants' design criterion is that the plant can be shut down safely and maintained in the safe shutdown condi tion. This criterion is acceptable. The secondary confine-ment building proposed for Zimmer provides for external missile protection for the drywell and fuel pool that is comparable with i other BWR designs. We conclude that the missile protection criterion for the Zimmer Station is acceptable. 3.8.4 Seismic Quality Assurance The quality assurance requirements for Class I (seismic) structures, systems, and components have been specified by the applicants. We have concluded that these quality assurance provisions that were implemented for all items designated as seismic Class I I i V
1 - 119'- 'for design, comply with the requirements of Appendix 8, " Quality Assurance Criteria for Nuclear Power Plants" of.10 CFR 50, 4.0 PLANT 3AFETY ANALYSIS 4.1 Ceneral During the review of this plant, the' applicants and th's _ staff postulated and analysed a number of incidents and accidents. Four j accidents were.:ensidered in evaluating the capability of the ensi-- neered safety features. including containment, to protect the public from potential, though very unlikely, accidental releases of fission products from the reactor. These four accidents are (1) the contrel-rod-drop accident, for which it is assumed that. I the rapid withdrawal of a control rod causes rupture of fuel roda by overheating; (2) the fuel handling accident, in which it is assumed that fuel rods are ruptured as a result of dropping a fuel element during refueling operations; (3) the steam-line-break . accident, in which it is assumed that the rupt'ure of a steam line outside the containment results in the release of reactor water and steam to the environment; and (4) the loss-of-coolant accident (LOCA), in which it is assumed that the rupture of a reactor coolant pipe inside primary containment leads to the rupture of fuel rods due to overheating. These accidenta are considered as t design basis accidents (DBA's) and the calculated potential conse-quences of the DBA's exceed those of all other accidents considered. We performed conservative analyses of these design basis accidents to o t'
- 120 - assess the capability of the engineered safety features and containment to ecmtrol the possible escape of fission products from the facility. ' The two-hour dose at the site boundary and the duration of the accident dose at the Jow population sone distance were calculated for each accident. In our calculations we have used a conservative ' model for estimating the noteorological diffusion of the effluents and a conservative estimate of the internal building mixing that occurs. The results of our analyses vains these conservative assumptions are well within 10 CFR Part 100 guidelines. 4.2 Loss-of-Coolant Accident 1 In calculating the consequences of the postulated loss-of-coolant i accident, we have used the conservative assumptions presented in the 'Coussission's Safety Guide 3 entitled " Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-
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-t i Coolant Accident for Boiling Water Reactors." The primary con-1 ] tainment was assumed to leak at a constant rate of 1.5% of the containment volume per day at accident condittorts for the duration of the accident without consideration of the effects of decrease in pressure during the post-accident interval. The leakage is assuined to pass directly to the secondary containment where 50% l-I i f i t
- 121 - internal mixing of the free air-volume of the reactor building atmosphere would occur and would then be released at a contro14ed flow rate through the standby gas treatment system (which has a 991 filter efficiency for the two SCTS deep bed charcoal filters in series) which exhausts to the environment at an assumed worst case elevation (ground level). The recirculation and mixing credit i assumes-that activity leaking from the primary containment is mixed in the reactor building by means of the 80,000 cfm recir-I ) culation fan and associated ductwork. The resultant calculated LOCA dose levels at the exclusion radius are 102 Res to the thyroid and 3 peu to the whole body and at the -low population sone are 81 Rea to the thyroid and 3 Res to the whole body. These doses are well within the 10 CFR Fart 100 guidelines. 4.3 Fuel Handling Accident For this accident, it is assumed that a fuel assembly will be dropped by the refueling crane into the reactor core or spent fuel pool from the maximum height perinitted by the fuel handling equipment design, even though the reactor fuel handling equipment and pro-cedures are designed to reduce the probability of dropping a fuel element while refueling. In addition, operating limitations and u____.__.____
.i - 122 1l [ engineered safety features are provided to Ifmit the consequences -of the accident. Fuel handling will not be ellowed before a sini-i aum of 24 hours has elapsed after reactor shutdown, so that fission-product activity has decayed significantly. Should such an acci-l dont occur, the radiation detectors will automatically isolate l' the reactor building, shut off and isolate the normal ventilation i j system, and start the standby gas treatment system before ' fission p i 1 products can be exhausted by the normal ventilating systes. With 0 the operation of these engineered safety features, fission products. ] rolessed to the reactor building air will be mixed and filtered before release. i' In our evaluation of the refueling accident, we assume that the cladding on 111 fuel rods is damaged (equivalent to more than 2 fuel assemblies); the fission product gases from the damaged fuel rods are released to the water, and then escape to the reactor building; the damaged fuel rods were all used in a high power region of the reactor (1.5 times the average power); and that all the radioactivity is released within 2 hours. We also assume that 10% of the noble gases and 10% of the iodine gases in the damaged fuel rods are released to the pool water, and that 1% of the iodines and all the noble gases in the pool 1 1 ~- ^ - ~ ~ - __1
- 121 - water escape to the reactor building air.. The reactor buildling i i air is processed in th" sasse manner as in the 14CA accident. The resultant calculated dose levels at the exclusion radius are 2 Ren to the thyroid ar d to the whole body and at the low population sone are less than 1 Ram to the thyroid and to the whols body. ' These doses are well within the 10 CFR Fart 100 guidelines. 4.4 Control Rod Droo Accident For the postulated control rod drop accident, it is assumed that a bottom entry control rod has been fully inserted and has stuck ir. this position, the drive becomes uncoupled and withdrawn from the rod. Subsequently, it is assumed that the rod falls out of the core it.serting an amount of reactivity corresponding to the ) worth of the rod. The reactor is designed to reduce the probability of this accident and engineered safety features are provided to limit the conse-quences of the accident. For example, the control rod worth minimiser 1* designed to limit the reactivity worth of any control rod during the startup phase of reactor operation. The control rod velocity limiter will limit the velocity during free fall to less than five feet per second. The steam line radiation monitor will detect excessive radioactivity and isolate the main i l turbine and condenser by closing isolation valves in the condenser ] l I l ]
I i - 124 - l 1 1 l .i mechanical vacuur pump system before the radioactive steam can 1
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. travel from the detector to these isolation valves. Because of .1 the operation of these engineered safety features, the fission ,) products that escape to the environment would be only those t i-which leak from the isolated turbine and condenser. ) ] ) \\ l in evaluating the radiological consequences of this accident, we I have made assumptions based upon the applicant's analytical model as presented in the preliminary safety analysis report. I As discussed in the subsequent paragraphs, the analysee tech- ] niques for this particular accident are being revised by 2 i and, depending on the results 'of these analyses, we may require i l. modifications, in addition to those presently provided, to t ' mitigate the potential consequences. t J The radiological consequences of a rod drop accident when the t reactor is in the hot standby condition at zero power is the worst situation for this accident because a high energy release is cal- ) i culated for this condition and because a path for the unfiltered 1 release of fission products could exist through the mechanical vacuum pump on the condenser. However, an automatic isolation valve has been installed on the condenser vacuum pump which would be closed by a high radiation signal from the steam line monitor to confine fission products released from the fuel to the prican I
1 l 1:i - 125 - 1 system. The pump would also be tripped by the signals, thus providing a second barrier to thE release of fission products. For this accident, the most reactive control rod assembly (reactivity 4 worth of 2.5%Ak) is assumed to drop otst of the core 30 minutes after 4 shutdown, causing 330 fuel rods to e:sceed a ' calculated energy input of 170 cals/gs. 7t.ese rods were assumed to perforate releasing 100% of the contained noble gases and 50% of the contained halogens to the reactor coolant system. Of the halogens released from the affected rods, 90% are assumed to be retained -in the primary system and one-half of the rensining halogens are assumed to be removed by plate-out. All of the noble gases and 2.5% of the halogens are assumed to be released frem the primary system through condenser vacuum pump system to the atmosphere. A conservative ground level release was assumed, j with Pasquill Type F meteorological 4.onditions and a 0.2 meter per second wind speed, for the 2-hour doses at the site i., dary. At the low population zone, these conditions were assumed to con-t tinue for eight hours followed by spreading of the plume into a 22-1/2* sector for the period from eight to 24 hours. For this accident, the 24 hour time interval is the full course of the accident. The resulting doses are well within the 10 CFR { Part 100 guidelines. ) )
1 4 - 126 - N l . The Atomic Energy Commission has for some t!.me utilized the j -ij ~ Brookhaven National Laboratory (BNL) as its consultant as part of the - regulatory assistance program. For the past few months, i l personnel at BNL have been performing in' ependent calculations of d a casign basis control rod drop accident, As a consequence of I the work performed to date at SNL 12. appears that the model used ~ \\ i by ~ General Electric to evaluate the design basis control rod drop accident should be revised. Specifically, :he assumed rate of ( ~ negative reactivity insertion fros' control rod scram is not l j suitably conservative since it uses insertion characteristics ~.j now considered to be not readily attainable in large boiling ) l water reactors. In addition, the actual insertion rates are not i linear as assumed. 8 The techniques for analyses of the control rod drop accident are presently being revised by the General Electric Company. Included ir the revised analyses are, among other features, a change in the method for modeling the rate of negative reactivity insertion from a control rod scram. A description of the revised analyses p-and the results of the new analyses are expected to be available within the next few months. Upon receipt of this information, the staff will evaluate the adequacy of the revised model and the resul-e tant consequences of the postulated accident situation. If the consequences of any of the analyzed trans4=ots result in a peak t.' s
f- - 127 - calculated enthalpy of 280 calories per gram, or the radiological consequences approach the guideline values of 10 CFR Part 100, we will require modifications 'to limit the consequences within acceptable values. I 4.5 Main Steam Line Break Accident The break of a main steam line outside both the drywell and the reactor building represents a potential escape route for reactor coolant from the vessel to the atmosphere without passage through the reactor building and standby gas treatment system. This escape route would exist only for the few seconds required to i sense the break and close the main steam line isolation valves. The steam line break would be nonsed by either high steam flow or i increased temperature in the pipe tunnel if the break occurred ) in this region. The steam line isolation valves would be closed within 5.5 seconds af ter the steam lhs break is sensed. This closure time determines the amount of primary coolant that would be lost through the break. In calculating the consequences of a steam line break accident, we have followed the Commission Safety cuide 5 entitled " Assumptions Used for Evaluating the Potential Radiological Consequences of a Steam Line Break Accident for Boiling Water Reactors. ' We also assumed that the radioactive inventory in the reactor coolant water corresponds to that associated with an estimated of f-gas release
- 128 - rate of 0.1 curies /sec (meneured before the charcoal delay system) and that no fuel rods rupture during the accident. The escaped fission products would be dif fused by the same meteorological conditions as those described above. The calculated dose levels resulting f rom this accident at the exclusion radius are 16 Rem to the thyroid and less than 1 Rem to the whole body and at the low populat ion zone are 3 Ram to the thyroid and lema than 1 Res to the whole body. These doses are still well within the 10 C7R Part 100 guidelines. 4.6 Instrument Line or Process Line Break In addition to the Postulated Design Basis Accidenta described above,.we have also evaluated the break of. ana11 instrument or process line outside of the primary containment during normal operations.- Such an event represents a potential escape route for reactor coolant from the vessel tc the reactor building. To reduce the consequences of the failure of lines penetrating the primary containment and to reduce the probability of an unisolated line, the applicants will meet the Cossaiselon Safety Guide No. 11 entitled b " Instrument L!nes Penetrating Primary Reactor Containment." In our evaluation of this event, we used the same iodine concentra-tion in the prieury coolant as for the stean If ne break. Our walys.f s also assumes that the radiat f on 1cvels in the reactor 1 1
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1 - 129 - l.; building venL 61stion duct would be high enough to isolate the. normal ventilation system and to trip the SCTS so that the radioactivity would escape to the atmosphere af ter being filtered by the SCTS. O The consequence of this accident is less than the calculated doses ^ for the steam line break accident and therefore.'is well below the - 10 CFR Part 100 guidelines. 4.7 Conclusions We have calculated potential offsite doses for four postulated ' design basis accidents which hr the greatest potential for off-site consequences using asaupptions consistent with those which we have used in previous BWR reviews. Based on our evaluation we conclude that the calculated radiological doses that might result from any of the design basis accidents are well within the guideline values given in 10 CFR Part 100. 5.0 QUALITY ASSURANCE The applicants have described their Quality Assurance (QA) Program l to be implemented in the design and construction of the Zinumer plant. The proposed QA program will satisfy the criteria of Appendix B,10 CFR Part 50, and information identifying how the applicants intend to meet each of the 18 QA criteria has been addressed. i i I j l
- 130 - The applicants are responsible for implementing the over-all QA program. Other organizations involved are Sargent and Lundy, the architect engineers; the General Electric Company, supplier of.the Nuclear Steam Supply System;' and Kaisers Engineers, Incorporated the constructors. The organization for the Zimmer project consists of a project manager, an sssistant manager who supervises eight full-time engineers (two from each of four disciplines that conduct only design reviews) and a separate full-Il time office QA group consisting of a section head and four engineers. i I This office'QA group reviews drawings-and specifications to assure f that QA requirements, including design review by appropriate in-house B 8-engineers and outside contractors, have been conducted. The.appli-i cents also have a field QA organization, consisting of three engineers who perform field QA work, such as equipment receipt inspection and equipment installation checking. The applicants stated.that addi-tional part-time office QA personnel are available if required and that a number of additional field QA personnel would be assigned to the site. This proposed organizational arrangement of QA functions relative to the project function and the planned extent i of staffing is acceptable. The applicants' QA plan, the criteria of l Appendix B,10 CFR 50, and the draf t copy of their QA manual (a working docurrent to de fine in greater detail the QA plan and i 1 L l ]
i s - 131 - responsibilities of the principles involved in their QA program) I have been inspected by Region III, Division of Compliance, on July 7 and 8,1971. Based on the above information, we believe that the applicants' quality assurance program is acceptable to satisfy the requirements of Appendix B of 10 CFR Part 50. r The architect-engineer (Sargent & Lundy) is responsible for tha development of detailed specifications, drawings and coordination of the design interfaces. Sargent and Lundy's internal review program and the extent of its design review of CE-supplied equip-ment is acceptable. The General Electric Company's Quality Assurance program is identical to that performed on other BWR e A facilities and is acceptable. The constructor (Kaiser Engineers, %Q Inc. [KE1]) is responsible for conducting a Quality Assurance and Quality Control (QA/QC) Program that meets the requirements of = the CC&E QA program and the intent of Part 50, Appendix B, as applied to the functions of the constructor. The KEA QA/QC manager has the authority to suspend construction activities at any time work continuation could affect the safety related functions of the facility. Region III, Division of Compliance conducted a special inspection in January 1971 to review the KEI construction QA/QC program as related to the Zimmer project. Compliance found E p [
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e ~ ' t.. s 9 ' the Kaiser QA/QC programs ' acceptable and, more important, found 1 9~ that they desanstrate a sound knowledge of Part 50, Appendix B,. y requirements 'and fully understand the needs for implementing a b comprehensive QA/QC program during construction.
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L l Based on the above information, we have concluded that ".he p . applicants' over-all. Quality Assurance program for the Zinsser ([. 6.[ project is acceptable. s 7 6.0 TECHNICAL OCALIFIfg{0NS AND CONDUCT OF OPERATIONS . j/ s }It The Cincinnati Cag and Electric Co. is responsible for the design,. -{ construction, and operation of the facility' and is authorized to 'e act as the agent for C&SOE and DPL in all details of construction s-I ] including licensing. The applicants have no previous experience associated with generating electricity through the use of nuclear l e } fuel. Houcvsr, the utilities (specifically the CC&E Company) j have acquired considerable experience in the design,. construction, and operation of numerous fossil fuel power plants. The various subcontractors and consultants assisting the applicants have j b been actively involved in'the suelear energy field. ~ The technical support for the design and construction effort of j i the facility is provided by the CG&E Company's Engineering and l Construction Department staffed by over one hundred engineers. l
i - 133 - 1 Responsibility for the review and approval of design features of the plant rests with the CG6E technical staff. In addition, the I Cc4E staff will implement, with the assistance of consultants, a i quality assurance program and will. follow field construction of ) the plant until its completion. Complementing the applicant's technical staf f will be the engineer-ing staffs of: the architect-engineer Sargent and Lundy, which will provide the coordination of all aspects of procurement of l equipment, systems, and construction services through the use of specifications and drawings that will integrate the total station design; the General Electric Company which will provide the nuclear f steam supply systes, the nuclear fuel, and technical personnel for j guidance and support during start-up operations; the Westinghouse Electric Corporation which will supply the power conversion i equipment; and the constructor, Kaiser Engineers, Inc., which I will perfor1m the services of construction manager. Other sub-contractors and consultants have been engaged to provide expertise in fields of a specialized nature such as: meteorology, hydrology, seismology and geology. We conclude, based on the results of the Zimmer review, that CC&E retains a technically competent and safety-oriented engineering organization that can effectively manage, design and cons t ruct the Wm. H. Zimmer Nuclear Power St ation.
.6 c - 134 - I 'l The applicant's operations group will be headed by the operating supervisor who is responsible for the supervision of the plant operations. The Shif t Supervisor reports to the Operating Supervisor and will be in charge of the plant operation on a shift basis. The I l operating shif t will consist of Shif t Supervisor, Control (Licensed) 1 Operator, Assistant Control (Licensed) Operator, Auxiliary Operator, and j, ' Ass!stant Auxiliary operators. The auxiliary operator and assistant I will be qualified to perform the health physics technical activities. l l Five shif ts are provided in order to maintain adequate twenty-four i' hour per day, seven days per week coverage. The applicant 's operating shif t crew size, for single unit operation, is in accord-ance with AEC Safety Guide number eight, " Personnel Selection and Training. On this basis, we conclude that the CC4E Company will have an adequate station organization and capable staff with which to operate the Zimmer facility safely. With respect to personnel training and qualifications, the app 11-cant has established the minimum requirements for the selection, qualification, and training of key station personnel. These requirements are in accordance with ANSI N-18.1, " Proposed Standard for Selection and Training of Personnel for Nuclear Power Plants " l dated June 22, 1970. The CG&E Company nuclear training program I is basically the same as the one developed by the General Elect ric Company for BVR plants. The program is designed to meet the spect-fic needs of each man selected to perform a specialized task and
c - 135 - ) l to provide him with sufficient knowledge and experience to perform [ their function. We conclude that the training program is acceptable to provide the facility personnel with the required capabilities to operate the Zimmer facility safely. The applicant has stated that they will-establish a routine review function and an Operations Review Committee (ORC). Routine reviews will be conducted at the plant level under the direction of the Plant Superintendent. The ORC will be composed of se least five personnel who for the most part are not members of the operating organization. The ORC will meet at least quarterly and will be responsible for review of design changes involving unreviewed safety questions, and changes to the Technical Specifications or other changes which might affect nuclear safety. The Committee will also review facils ty ' operation, equipment, and personnel j performance to determine adherence to license requirements. The i results of such reviews will be reported to the Executive Vice President of CC&E. Based on the appiteant's commitment to meet DRL's Operational Safety Guide No. 4 " Supplementary Guide to the Contents of Tech-nical Specifications (Administrative Section)", we conclude thet the plans for an objective review and audit are acceptabic.
i - 136 - 7.0 EMERCENCY PLANNINC The applicants have presented in Amendment 11 'a description of the emergency plans that will be developed for the facility for use in the event of an accident involving the release of radioactive material to the environment. The plan will rely on the use of 'l the station organization with the Shift Supervisor responsible for initiating corrective action to limit the consequences of the l accident. Supplementing the normal operating crew for conditions that exceed the capability of the shift crew will be a designated ency organization.. In addition, an emergency coordinator <6 will be designated who will be in complete charge during the i emergency. At predetermined action levels the emergency coordinator I l will decide on the need and extent of offsite assistance required in case of an emergency involving the general public. The plan will be coordinated with local agencies auch as South C1.aremont q .{ Disaster Connaittee (formerly the New Richmond Flood Cosmaittee), s the Red Cross, law enforcement agencies (i.e., the Ohio State l Patrol, the Kentucky State Police, and the Claremont County Sheriff), 4 Fire Departments, U.S. Coast Guard, Ohio a. I Kentucky Health Depart-ments and other government agencies. Medical support will be coordinated through the applicant's Medical Director.
l - 137 - The applicant's description of the proposed emergency plan for the l l facility conforma satisfactorily with.the guidance and criteria 1 provided in Appendix E,10 CFR Part 50. The applicant has contacted the superintendent of the Richmond j School District with regard to evacuation procedures for the school. Frocedures for evacuating and quartering residents in the areas of Moscow and New Richmond, Chio for Ohio River Floods have been in existence for some time. The South Claremont County Disaster Coe-mittee (formerly the New Richmond Flood Committes) and the Red Cross have decided to increase their scope to include all types of disasters. The applicants have contacted the Red Cross and the South Claremont County Disaster Coenittee to discuss inclusion of the facility in the overall emergency plan for the local area. The applicant has docu-mented the results of agreements reached with the various agencies contacted. We conclude that the emergency plan proposed for the site and surrounding area is acceptable for the construction per-mit stage of review; however, we intend to continue our review of this matter during plant construction to assure that the procedures for implementation of the progras are adequate. 8.0 CONFORMANCE TO CENERAL DESIGN CRITERI A 1 Based on our evaluation of the preliminary design and design l criteria for the proposed Wm. H. Zinsner Nuclear Power Station. 0;
I - 138 - Unit I, we have concluded that the applicants plan to meet the intent of the General Design Criteria for Nuclear Power Plants, I published May 21, 1971, as Appendix A to 10 CFR Part 50 in the final design of the station. 9.0 REPORT OF THE ADVISORY CODMITTEE ON REACTOR SAFECUARDS (ACRS) The Advisory Committee on Reactor Safeguards (ACRS) met with 1 l representatives of the applicant and the regulatory staff to review the proposed Ms. E. Zimmer Nuclear Power Station at its i 137th meeting on September 10, 1971. In addition, steccusittee l.i ametings of the ACRS were held with the applicant and the regule-tory staff on August 27, 1971, at the plant site, and on I g September 1,1971, and again on September 8,1971, in Washington, D. C. A copy of the ACRS letter to the Commission, dated { September 17, 1971, concerning the CG&E application for a construction permit for the Wm. H. Zismer Nuclear Power Station, Unit 1 is attached as Appendix B. In its letter, the ACRS reported favorably on the application for a construction permit provided that the items discussed below are resolved during the construction of the facility. The ACRS recommended that the applicants design the main steam ) lines so that they retain their integrity during a design basis earthquake (DBE), and that the sealing system, installed to mini-I
- 139 - mise leakage through the MSL isolation valves, be designed as an engineered safety feature. These matters have been disctreed I with the applicants (see section 3.2.2 of this report). The appli-cants have stated that the final design of the systems will be sub-mitted to regulatory staff for review and approval. The ACRS stated that the automatic tripping of the recirculation pumps, to make tolerable the consequences of failure to scram during anticipated transients, represents a substantial improve-l ment for this facility. However, the ACRS recosamends that the applicants perfom further studies to evaluate the sufficiency of this cpproach and the specific means of implementing the pro-posed pump trip. The applicants have stated that a study would be performed and the results submitted to the regulatory staff for review. We will review the applicant's findings during the operating license review and keep the ACRS informed of the resolo-tion of this matter. The ACRS recommended that the primary containment be inerted during operations and that the proposed hydrogen control system, provided to control the concentration of hydro 8en in the primary contain-ment in the unlikely event of a loss-of-coolant accident, be designed to maintain the hydrogen concentration within acceptable b
1 4 l 140 - l l limits using the assumptions listed in the AEC Safety Guide 7, Control of Combustible Cas Concentrations in Containment Following a Loss-of-Coolant Accident. The applicants have stated that inert-ing provisions have been incorporated in their design and that the hydrogen control system will be designed in accordance with AEC criteris and submitted to the staff for their review and approval. The Advisory Conunittee on Reactor Safeguarcs has concluded that the items mentioned above can be resolved during construction of the facility and that, if due consideration is given to these J items, the William H. %lasser huclear Power Station, Unit 1, can be constructed with reasonable assurance that it can be operated without undue risk to the health and safety of the pub)f e. 10.0 CG900N DEFINSE AND SECURITY The application reflects that the activities to be condui ted will be within the jurisdiction of the United States and all of the directors and principal of ficers of the oppiteants are United States citizens. The applicants are not ovned, dominated or con-trolled by an alien, a foreign corporation, or a foreign govern-ment. The activities to be conducted do not involve any restricted data, but the applicant has agreed to safeguard any such data which night becoew involved in accordance vith the req u i rement s 4
L - 141 - ) of 10 CFR Part 50. The applicants will rely on obtaining fuel as it is needed from sources of supply available for civilian purposes so that no diversion of special nuclear material for stilitary purposes is involved. For these reasons, and in the absence of any information to the contrary, we have found that the activity to be performed will not be inimical to the comanon defense and security. 11.0 FINANCIAL QUALIFICATIONS i l The Comnission's regulations which relate to the financial data and information required to establish financial qualifications for an applicant for a facility construction permit are 10 CFR 50.33(f) ) and 10 CFR 50, Appendix C. The joint applications of the Cincinnati Cas and Electric Company, Columbus and Southern Ohio Electric l Ccapany, and the Dayton Power and Light Company, as amended (Amendment No. 20), and the accompanying certified annual finan-cial statements provide the financial information required by the Cossais sion 's regulations. Based upon the financial information presented, the estimated i cost of construction for Unit No. I of the proposed facility, j including inventory costs for the first core fuel is $307.000,000 l of which $287,000,000 is for the nuclear steam production plant
4 .l - 142 - ) l l i ti f h; cos t, and $20,700,000 is for associated transmission facility '? cos ts. We have determined that l the estimated costa of total pro-duction plant construction and transmission costa are reasonable. h We have reviewed the financial statements of the applicants to 4 l determine whether they are financially qualified to meet these estimated costs. - The applicants' operating agreement provides i that the three companies will own the Zinser station, Unit 1 as tenants in consson, with the tentative percentage shares of owner-ship being as follows: CG&E 40.0%,- C&SOE 28.5%, and DFL 31.5%. Construction costs will be shared accordingly. i The total costs to constru:t the Zimmer facility will be financed by the applicants as an integral part of its normal construction program, using funds generated internally and-from the sale of debt and/or equity securities and from short-tern loans when and as tsquired, in the same general manner as it finances other plant additions. An analysis of the applicants' financial statements over the past { five years (1966-1970) indicates a strong financial position, sound financing, significant resources, steady increase over the period in the level of operating revenues.and plant inves tment, l and a continued sustained level of earnings, r 4 0 6 (
- 143 - 1 Based on our evaluation, together with. the reasonable assumption that auch earnings will continue, the applicants ' excellent credit and high bond ratings and its. Proven ability to borrow on a short-term basis, we conclude that the applicants possess or can obtain the necessary funds to meet the requirements of 10 CF3t 50.33(f) with respect to the operation of the William H. Ziauner Nuclear Power Station, Unit 1. Copies of the staff's financial analysis-of each Company are attached as Appendix C.
12.0 CONCLUSION
S i 1 Based on the proposed design of the William H. Zinsner Nuclear Power Station; on the criteria, principles, and design arrangements for 'che systems and cessponents thus far described, which include all of the important safety items and on the calculated potential consequences of routine and accidental release of radioactive asterials to the environs; on the scope of the development pro-grams conducted; and on the technical campetence of the applicant and the principal contractors; we have concluded that, in accordance with the provisions of paragraph 50.35(a),10 CFR Part 50 and paragraph 2.104(b),10 CFR Part 2: 1. The applicants have descr'ibed the proposed design of the facility, including the principal architectural and engineering c
I - 144 - criteria for the design and has identified the major features. or components for the protection of the health and safety of. the public. 2. Such further technical or design information as may be required to complete the safety analysis and which can reasonably be 'left for later considerations, will be supplied in the final safeey analysis report. 3. Safety features or components, which require research and development, have been described by the applicants and the applicants have identified, and there will-be conducted a research and development program reasonably designed to resolve any safety questions associated with such features or components. 4 On the basis of the foregoing, there is reasonable assurance that (1) such safety questions will be satisfactorily i' resolved at or before the latest date states in the appli-cation for completion of construction of the proposed facility and (ii) taking into consideration the site criteria contained in 10 CFR Part 100, the proposed facility can be constructed and operated at the proposed location without undue risk to the health and safety of the public. f-t J
- 145 - 5. The appl. cant, are financially and technically qualified to l design and construct the proposed facility. 6. The issuance of permits for the construction of the facility will not be inimical to the comanon defense and security or to the health and safety of the public, i
_1 APPENDIX A Chronology For The Zimmer Application 1. April 7,1970 Submittal of PSAR Voltanes 1 thru 5 sad License /pplication for Units 1 and 2 (Docket Nos. 50-358 and ' 359). 2. May 12,1970 Initial meeting with the applicant to
- f discuss preliminary status of DEL review of 21sumer Station.
I 3. May 21, 1970 Meeting with applicant to discuss F i significance of Environmental Report I-and development of information provided in PSAR. i 4. July 30 & 31, 1970 Technical meeting with applicant to dis-cuss meteorology, hydrology, geology, seismology, general site items and l 1 con t ainment. I S. Augus t 7,1970 Technical meeting with applicant to dis-t cuss seismic design criteris, and structural analysis of Class I design for all forces (seismic, accident, wind, etc.). \\ 6. lugust 25, 1970 Technical meeting with applicant to discuss hydrology - maximum probable flood design. 7. October 13, 1970 Request for additional information mitted to applicant. 8. November 3,1970 Amendment No.1 - Submittal of geology, seismology and site foundation information. l 9. November 5,1970 Prelimin'ary ACRS Report transmitted to ACRS.
- 10. November 19 & 20, 1970 Technical meeting with applicant to discunn the station auxiliary systems and the quality i
assurance program. 11. November 20, 1970 Amendment No. 2 - Submitted information on hydrology and containment. l l l 4 lR
i, e I l l t . 1 BLANK PAGE I 2 i I I i i f
12. December 16 1970 Technical meer.ing with applicar.* t ; discuss instrumentation, control and elect rical systems. I 13. December 17, 1970 Amendment No. 3 - Revised PSAR to reflect change to a single unit design. 14. January 5,1971 Technical meeting with applicant to discuss seismic design values and liquefaction analys is. I
- 15. January 12, 1971 Amendment No. 4 - Received response to DRI.
reques t for additional information dated October 13, 1970, also included were revised accident analyses and design changes to SCTS system. 16. January 22, 1971 Applicant submitted Environmental Report.
- 17. January 26, 1971 Meeting with the applicant to discuss back-filling technique under the reactor building, and service water pipes (betseen reactor 3
l building and pump house). i 18. February 3,1971 Technical meeting with the applicant and Feb ruary 10, 1971 General Electric Company for a generic discussion of 1969 product line flow control valve, instrumentation, and control. 19. February 8,1971 Meeting with applicant to discuss liquefaction potential of soil under service water pipes and pump house. 20. February 12, 1971 Amendment No. 5 - Completed revisions of PSAR as outlined by Amendments 3 and 4. s 21. February 25 & 26, 1971 Technical meeting with applicant to discuss ECCS, pipe whip criteria and radiological consequences of accidents. 22. Feb ruary 16, 1971 Request for additional information transmitted to applicant. i I,
23. April 3,1971 Amendment No. 6 . Revised Section 7 of 'PSAR Instrumentation and Control and updated system classification of Pressure Boundary Integrity Criteria, Appendix A. s'
- 24. ' April 9,1971 Request for additional information on the balance of review items transmitted to the applicant.
- 25. April 19, 1971 Amendment No. 8 - Provided anti-trust i
information, i I
- 26. April 21,1971 Raquest for additional information on Environmental Raport transmitted to the applicant.
- 27. April 27,1971 Amendment No. 7 - Raceiwd response to DRL request for additional information dated
'] October 13, and February 16, 1971 and revised quality assurancs program. a
- 28. May 17,1971 Amendment No.1 '- Receiwd response to request for additional information dated April 21,1971.
- 29. May 17, 1971 Amendment No.10 - Provided information for notification to newspapers for legal notices. -
- 30. May 17, 1971 Amendment No. 9 - Revised PSAR to incorporate l
new Appendizes H.0 - Resolution of AEC-ACRS ) Staff Concerns, I.0 - Seismic Analysis of Critical Structures Systems and Components, and J.0 - Analysis of Under-Cround Service Water Pipes and Supporting Piles, j l
- 31. May 28, 1971 Amendment No.11 - Received response to DRL request for additional information dated April 9,1971, also provided primary con-tainment hydrogen, oxygen and fission product sampling system.
1
- 32. June 15,1971 Amendment No.12 - Received response to balance,of DRL request for additional inforination, also provided pipe whip criteria.
l i l t i I d d
.33. July 15,1971 Amendment No. 13 - Provided revision to DRL questions, and also provided conformance to I AEC ECCS interim acceptance criteria and flood protection capabilities. I j 34 July 16, 1971 Request for additional information on generic CSCS questions transmitted to the applicant.
- 35. July 23,1971 Amendment No.14 - Revised PSAR to incorporate additional _ details on inservice inspection j
program and radweste off-gas treatment system. .i j
- 36. July 27 & 28,1971 Technical meeting with applicant and General Electric Co. to discuss balance of instru-mentation, control, and electrical review.
I l
- 37. July 30,1971 Amendment No. 15 - Revised PSAR to incorporate t
split loop design of auxiliary system also included information on flow control valve, reactor nozzle failure, and primary contain-ment design and leakage. l 38.- August 4, 1971 Amendment No.16 - Received response to request for additional information dated ..] July 16,1971. s 't
- 39. Augus t 5,1971 Received consultate report from the Flah and Wildlife Servlet. U.S. Department of Interior.
- 40. August 9,1971 Technical meeting with applicant, DRL, DRS, and our consultants to discuss balance of review on reactor building and service water pipe soil foundation.
- 41. August 11, 1971 Amendment No. 17 - Provided an analysis to i
show that the recirculation flow control valves do not have a safety significance.
- 42. August 16, 1971 Report from the Fish and Wildlife Service, U.S. Department of Interior, was transmitted to the applicant.
43. August 23, 1971 Amendment No. 18 - Received responses on the site soil foundation discussed in the August 9,1971 meeting. f I I I
- 44. ' Augus t - 27, 1971 ACRS Subcommittee site visit meeting.
45. September 1, 1971 ACKS Subcommittee meeting to discuss the Core Standby Coo!!ng System and containment design.
- 46. September 8,1971 Amendment No.19 - Rec.!ved balance of I
facility responses and commitments necessary to resolve all matter.
- 47. September S,1971 ACES Subcommittee meeting to discuss hydrogen control system and outstanding items prior to the full committee meeting.
- 48. September 9,1971 Received consultants report from Nathan M.
Newmark Consulting Er.gineering Services.
- 49. September 10, 1971 ACRS full committee meeting.
- 50. September 16, 1971 Received applicants' cosaments to the Fish and Wildlife Service report.
51. September 17, 1971 Received ACRS report on Wm. H. Zimmer Nuclear Power Station.
- 52. September 23, 1971 ACRS Report (letter) was transmitted to the applicants for comments.
53. October 8,1971 Request for additional information (pertaining to updating their financial status) was trans-mitted to the applicants.
- 54. November 2,1971 Amendment No. 20 - Received applicants' response i
for updating their financial status.
- 55. November 17 & 18, 1971 Site visit to discuss environmental matters.
- 56. November 18, 1971 Supplement No. I to the ER - provided the applicants' response to agencies connaents on the Environmental Report.
57. December 3, 1971 Amendment No. 21 - Received applicants' concents to the ACRS Report (letter). l l l i
I )' 58. De cember 3,1971 Amendment No. 2 - Updated the Zimer Environ-mental Report 59. December 20, 1971 Req ues t for additional information on Envircm-mental Report transmitted to the applicants. 60. January 4,19 72 Supplement No. 2 to the ER - Received the applicants ' analysis for the Envirortment al Impact of Accidents. 1 4 i l , 8 s D l I l' 4 - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.}}