ML19317G413
| ML19317G413 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 07/16/1968 |
| From: | US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML19317G410 | List: |
| References | |
| NUDOCS 8003030817 | |
| Download: ML19317G413 (73) | |
Text
{{#Wiki_filter:A " ~ ~ ~ - UNITED STATES OF AMERICA ATOMIC ENERGY COMMISSION l t . yf 4 s In the Matter of i F1.0RIDA POWER CORPORATION (Crystal River Unit 3 Docket No. 50-302 Nucicar Generating Plant) l 1 c:.t A w r Tg-gue M ya 4 w. k). ~ f SlM4ARY DESCRIPTION OF APPLICATION l3-t
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,.. n 4 ,4 TABLE OF CONTENTS (cont'd) r 'et w Page 5. TESTS. INSPECTIONS. AND QUALITY CONTROL 20-22 6. RESEARCH AND DEVELOPENT PROGRAMS 22-25 4 7. TECHNICAL QUALIFICATIONS 7.1 Florida Power Corporation 26-27
- 7. 2 Babcock and Wilcox Conpany 27-28
. i 't 7.3 Gilbert Associates. Inc. 28-29 7.4 J. A. Jones Construction Company 29 8. COMMON DEFENSE AND SECURITY 29-30 9. CONCLUSION 30-31 APPENDICES ? APPENDIX A - List of References y APPENDIX B - Figures !.v7 APPDIDIX C - Professional Qaalifications of Expert Panel Witnesses s..g ':.t cy;y ,4:4 47. ? y 4 w'Y i 9. 3.p 4% . ei; 4:s, u:.; k*" .F 3.,., 4-:* -i D.. L f;Q4d < tins?@MM,1fhy b)'S 'tS G+W'N.Q,: if WiQ ~ 7c.' f,'/,UN;'Li." f 3 ly NJ9 '% % :.d( 9[Qs 3.s.%..*w'-["ff:*p.C, ~Qpr; ?W l?;/]g)W'%,9 3.s x,.iQQ 0g3QQ,g.Q.;s %dzbs Why; Q cx v pqmpx&: .c,..g% o M Js,,,* &,y p ggg,# a g gy$v;ai QMlX^,,m+ z_; A 9 2n 4ie ;n y*h &..LQ. Q ?::3 .: m a: ..,, ~ o .m.. VP.*71 %u:wgn _w r., g e. . ;in -~j:Q. ,. q~; b.~ f;,s ;M v n fr ',3.. :y ::*,. ;;;. ] t 2.; ei k 1yy py 4,pclp ~;;Q).
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. un i 2 The purposa cf this document is to present a Summary Descrip-3 tion of Florida Power Corporation's Application filed under Section 4 104b of the Atomic Energy Act of 1954, as amended, for necessary 5 licenses to construct and operate Crystal River Unit 3 Nuclear 6 Generating Plant. This sumraary includes Amenduents 1 through 5 7 and Supplements 1 through 3 to the original application which was 8 filed on August 10,1%7 and assigned Docket No. 50-302. 9 The Sunnary Description will provide infonnation on the environ-10 mental aspects of the site, description of the Crystal River "lant Unit 3 design, conclusions from the safety analyses perfonne., descrip-11 t
- j 12 tion of the Quality Assurance Program and its audit of the construction 13 quality controls, areas of continuing research and developnent efforts, 14 technical qualification of the Applicant and its contractors and 15 conclusions of the Applicant in matters of common defense and securlty 16 and also the health and safety of the public with regards to the con-17 struction and operation of this nuclear facility.
18 In addition the Suinary Description contains an Appendix A - 19 List of References as the text contains nunerous references to the 20 Preliminary Safety Analysis Report where definite and detailed 21 description of the plant, its design or the Applicant's position 22 may be located. Appendix B - Figures contain a r:inimum of figures 23 to orient and amplify the word descriptions in the sunmary. 24 This Sunmary Description will constitute a portion of the pre-25 pared tes timony of the Applicant to be presented at its hearing 26 before the Atomic Safety and Licensing Board and is therefore being ) 27 sponsored by a Florida Power Corporation (herein sometimes called 28 the Applicant), technical witness fir. J. T. Rodgers, fluclear Project 29 lianager and Director - Power Engineering & Construction. -, M [b } h ki [ m : a m m[ w ; g( w x y p: m hl( hk $n w e n w; w, g g x ngm a sy genxw.wyayce mp a g. w a g;hbb.$gengw&0fhW h jf 3 ' h yh; m^5?' .hk$f m tQ x m:;gg:p&cu(p)ewlh%gg g%.g - yg-r gc 3 6 % q)+J.2e;g%cJ1 W g " Jn _' " ' c om nx l u b22 We '-~ i ) I
1 in assist rir. Modgers in answering questions on cross-2 oxamination by tha Board cr ans:mr party. tha f;11owing t:chnical 3 witnesses renresenting the Applicant, its engineers and its con-4 tractors will make up a panel cf technical expert witnesses whose I unprepared testimony will becone a part of the Applicant's testimony 5 6 before the Board. 7 These persons are: 8 flame Organization Ti tle s<ove &,,,,cyn. 9 Donald J. Rowland Florida Power Comoration. 10 E. Robert Hottenstein Gilbert Associates, Inc. Project Manager I 11 Morton I. Goldman ' NUS ' Corporation Vice Pmsident 12 Carl E. Thomas The Babcock & Wilcox Co. Project Manager 13 Robert E. Wascher The Babcock & Wilcox Co. Manager, Nuclear Safety 14 The educational and professional qualifications of the above 15 persons are included in Appendix C - Qualifications of Expert Panel 16 Witnesses. 17 The Crystal River Plant Unit 3 nuclear generating unit will 18 employ a pressurized water nuclear steam _ supply system furnished 13 by The Rabcock & Wilcox Coinpany (also herein mferred to as B&W) f 20 and is similar in design to the nuclear steam supply systects which l' 21 are being furnished by B&W to the Duke Power Coenpany for its Oconee tbsclear Station (AEC Docket Nos. 50-269, 270 and 287) and to the 21 Metropolitan Edison Company for its Three Mile Island Nuclear 9V N Station (AEC Docket No. 50-289). A construction penai t authorizing 25 ennstruction of the Oconee facilities was issued in November,1967 6 avid a construction penni'. authorizing construction of the Three 27 Mile Island Huclear Station was issued in May,1968. The nuclear ?, 29 s ti a.: supply system will operate initially at core power levels up a e f ,,q ec [. ) .y pg y= - g m g- [ yq,,,p h b gy g,,&re;n N M Qy p %q, n pn jmahh.g - - ,.-r a es.- m.,y*;%gl** 2 C mw J::; A;ND N;y'w/.1$hNW?gQ p m@;y@q@;m:f g;gg n 3
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J 1 to 2452 *t, dich c:rresponds to a gross si;ctrical output cf about 2 855 We. An ultimata care cutput cf 2544 Nt is cxpected, and cil 3 steam and power conversion equipment is designed for a correspond-I 4 ing gross plant output of 885 We. Plant safety systems, including 5 contairrnent and engineered safeguards, have been evaluated and will 6 be designed for operation at the higher power level. The higher 7 power level is also used in the analyses of postulated accidents, 8 establishing the suitability of the site under the guidelines set 9 forth in 10 CFR 100. 10 The Applicant's licensing application, including the amendnents 11 and supplements thereto, has been reviewed by the taff of the Atomic d 12 Energy Commission, which has' prepared and published a safety evalua-13 tion of the Application. The Advisory Comittee on Reactor Safeguards 14 (referred to as ACRS) has also reviewed the complete Application, 15 and reported its findings to the Chairman of the U. S. Atomic Energy 16 Comission in a letter dated May 15, 1968. The ACRS concluded that - 1 17 "the proposed reactor can be constructed at the Crystal River Plant 18 site with reasonable assurance that it can be operated without undue 19 risk to the health and safety of the public." The AEC staff concluded 20 s imilarly. The ACRS letter identified one item in the design for 21 which further development was requested. This item concerns off site 22 power supply and is elaborated upon in this Sumary in Section 3.6. ?' Other matters warranting consideration by the Applicant as rentioned 24 in the ACRS letter have been responded to in PSAR Volume 4, Supple-25 nent 2, Question 5 (answer). 26 The principal architectural and engineering criteria which 27 govern the p] ant design are set forth in Section 1.4 of Volume 1 28 and in Supplement No. Informal Question 4 of Volume 4 of the p f Applicant's Preliminary Safety Analysis Report. Design and 29 30 construction of the facility in accordance with these criteria i 31 together with the redundant engineered safeguards systems provide ) e 14 J aw nmA.wm s
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1 asstranc] that tha propostd Crystal River Plant Unit 3 can and 2 till be constructed and oper ted at th2 proposed 12 cation without 3 undue risk to the health and safety of the public. 4 The Applicant's Construction schedule tentatively sets the date 5 of earliest cogletion of Construction as December,1971 and the 6 latest completion of cor.struction date as June,1972. 7 2. DESCRIPTION OF SITE AND ENVIRONMENTAL CHARACTERISTICS WHICH 8 INFLUENCE DESIGN. 9 2.1 Location 10 The Crystal River Plant Unit 3 will be constructed at the same 11 location where coal-fired units I and 2 already exist (see Appendix B, 12 Figure 1). This site-is directly on the Gulf of Mexico, in the 13 Northwestem extrenes of Citrus County, Florida, approximately 7-1/2 14 miles NW of the town of Crystal River and 70 miles from Tawa, Florida. 15 The Applicant owns 4.738 acres including a wide access strip 16 provided for railroad, road, and transmission line right-of-way 17 extending eastward from the plant to U. S. Highway No.19. The 18 site region is characterized by its remoteness, with the Gulf of 19 Mexico on the 'a'est and with gradually rising terrain from tidal 20 swampland to gently rolling hills, some 16 miles to the East. 21 2.2 Population 22 The exclusion area, which is under the ownership.and control 23 of the Applicant, has a radius of 4,400 ft. (see Appendix B, Figure 2). 24 The nearest residence is 3-1/2 miles from the reactor building. 25 The low population zone has a radius of five miles. The nearest ~ 26 population center of 25,000 or more is Gainesville, Florida, 27 which is located 55 miles NNE of the site.(1) r 4s* M '^ M D Qff( TN }ff 0; k'; 'f h f{h} f l ]R7p'M;M hMg 4MMa& MyWigMsg 7 y din
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g 1 2.3 f%teorology W 2 Meteorology in the region of the Crystal River Plant site has g been evaluated to provide a basis for preliminary deteruination of 3 design criteria for stona protection, and a preliminary assessment 4 5 of routine and accidental radioactive gas release at the site. 6 Tropical storms are less a hazard in this region than in most other 7 areas of Florida. Lo tornadoes have been reported in Citrus County a for the period of record 1916-1966. Prevailing wind directions are 9 seasonal, and are predominantly northeasterly during the autumn and 10 southwesterly during the spring. Winds seldom persist from one 11 direction for longer than twenty-four hours. ] 12 In general, dif fusion of waste gases in the atmosphere is good. 13 Wind direction is useally highly variable, and wind speeds are sel-14 dom extrenely low.(2) 15 2.4 flydmlogy g 16 The two streans in the vicinity of the site are the Withlacoochee 17 River and the Cys tal River. The plant site is located approximately 18 3.8 miles south of the mouth of the Withlacoochee and about the same 19 distance north of the nouth of the Crystal River. The Withlacoochee i 20 is the major stream, having a drainage area at its entrance into the 21 Gulf of Mexico of appruximately 2,000 square miles. The discharge of 22 the Withlacoochee due to rain runoff is augmented by a base flow of b 23 grounhater runoff and artesian spring discharges.(3) The Crystal i 24 River is much smaller than the Withlacoochee River, with its major 25 discharge consisting of artesian spring discharges. 26 There are no public water supplies in the area of the plant 27 and all surface and underground water flow is from the plant toward g 28 the Gvi f of Mexico, 29 Frtm the studied infonnation and data the hydrology of the 30 site is concluded to be mos t favorable. > J MEiQg%a$%@d@h@y%jp@k!hj: iW33M95 5 hpkM M $k b h ny4winn$$a$NOh!![8dMp3$hN ' 7Mh lyl MM kd3$$54d!D . @ym dh ' iQ gpyh&wwiq sw:Appmcmdw
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1 tsecausa cf its cxposed location on th] Gulf cf Mexic3, the h 2 facility is theoretically cxposed t3 maximum intensity h:rricanes. 3 The general line of approach for hurricane protection will include g h 4 full protection against hurricane winds, flood tides and wave 5 action for components which must function for a safe and orderly 6 shut-down of the nuclear unit. This protection will be provided 7 for any intensity hurricane up to and inclu' ding the Proba* le Maximum - o 8 Hurricane (PM1) based on parameters established by the applicable 9 portions of the revised Environmental Sciences Services Administra-10 tion Criteria.(4) 11 2.5 Groundwater 12 ' Groundwater at the site occurs under watertable conditions, as ~ ~ ' 13 opposed to quite general artesian conditions throughout most of the 14 s tate. 15 The groundwater table occurs at a depth of approximately 8 to 16 9 feet below ground surface at a plant elevation of approximately 17 90 feet (mean low water = 88 feet). Groundwater levels observed 18 in drill holes were recorded to fluctuate, with little lag time, in 19 response to tidal variations. 20 No local flooding occurs from rainfall because of the site-21 proximity to the Gulf and infiltration. Af ter almost instant 22 infiltration, a westward sloping hydraulic gradient exists so 23 that site originated water flows into the Gulf of Mexico.(5)' 24 2.6 Geologg 25 The s.ite is located on gently southwesterly dgping biogenic 26 carbonate (limestone and dolcnite) rocks which have been-differen-27 tially dissolved along the most pervious zones of the rock, result-pd 28 ing in a network of general vertically oriented dissolved zones 29 (solution channels). The closest faulting occurs a distance of 30 three miles to the east but' stratigraphic correlation and continuity D $N5hhiNhf' kh N hhh hk s % M 9 d s: p +hdib u yp w g ;g er p y$h {#W sp ~ Y sp MiMM M ~ &;ph ;e wa@,,j%@;y#wya&dh,m#aga@gm RmnM m yen && M - rust:c%e;wn xN h n $in N Y $ i,W $ 500 $ A fL J w.x a . 5.?.$ & &;. p?$_Q. W-W: Tl - ADWM ^T Y -~ b
1 cr s;ismic profil;s negat2 th] possibla cxistence cf subsurface ,_). 2 faults at th2 site.. Regional t:ctonic olements are irctive and 3 Present nn threat to the structural integrity of local geology. L) 4 It is concluded that geologically the site ruck mass is competent 5 to support safely a nuclear station.(6) 6 ' 2. 7 Seismology 7 The State of Florida is seismically inactive, and the' closest 8 area to the site of significant seismic activity is Charleston, 9 South Carolina soae 330 miles distance from the site. Attenuation 10 data available for this area indicates that the site experienced 11 an observed intensity no higher than Intensity V (Modified Mercalli. N l'1E Scale). The maximian ground motion at the site due 'to tnis earth-13' quake did not exceed 0.025g. The Crystal River Plant Unit 3 Class I 14 structures, components and systens will be designed to withstand an 15 earthquake based on a maximtsn horizontal ground acceleration of 0.05g 16 and a maximum vertical conponent of 0.033g as an added margin of 17 safety. The ability of the plant to be shutdown safely will not be 18 impaired, however in the event of an earthquake based on a maximtsa 19.horizor.tal ground acceleration of 0.10g and a maximisn vertical 20 conponent of 0.067.(7) 9 21 2.8 Environmental Radiation Monitorino 22 A pre-operational environmental radioactivity monitoring program 23 will be conducted in order to detennine the ma'gnitude of the radio-24 activity in the environnent s'ervunding the nuclear reactor site 25 and to study fluctuations in the radioactivity lewis prior to the 26 operation of the nuclear generating plant.. The information obtained 27 will serve as a guide and baseline in evaluating any changes 'in 28 environcental radioactivity levels that may possibly be attributed e 29 to the Crystal River Plant Unit 3 operation. ( 1 ) v j nura,m e MMd$$w#$$a$55%Migj@+k,d,ki h, ~s Fjyj[y,r , ] mm mmm wn~
1 'A corgrehensive sampling program till ba initiat:d ct 1:ast (v) 2 two y::ars pricr to Unit 3 startup. Th2 c ll:ct.d samples will 3 consist of Gulf and well water, soil, air particulate, animal O 4 thyroids fish, sheiirish and bottom sediments. A post-operational ~ 5 envimnmental program will be similar to the pre-operational program 6 with the sarnpling and analyses schedule related to the level of 7 activity found in the environmental samples.(8) 8 The U. S. Department of' Interior Fish & Wildlife Service has on 9 February 12, 1968, in a letter to the AEC made several study and 10 survey recomendations to assure protection and improvement of 11 marine resources around '.2 Crystal River Plant. Florida Power 1
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s 12 Corporation will implement the programs suggested by this letter. 13 Preliminary discussions have been held with appropriate Florida 14 State Agencies on the scope of the program. Continuing discussions 15 will be held with State agencies in formulating an acceptable pre- - 'VO 16 operational environmental radioactivity nonitoring program. The 17 resulting program will be reviewed periodically to assure maximum 18 effectiveness. 19 3. DESCRIPTION OF CRYSTAL RIVER PLANT UNIT 3 20 3.1 Introduction 21 A description of the plant features and layout as well as an ~22 evaluation of the plant safety are s'et forth in the Application. 23 as acended. Tne Preliminary Safety Analysis Report describes in 24 detail the criteria to.be used in establishment of the final plant 25 design even though design itself is incomplete at this time. The 26 station will consist of a reactor building, an auxiliary building, 27 a turbine building, a, control room, a fuel storage building, a 28 service building, a substation, and various other auxiliary struc-29 tures and equipment. A plot plan of the c ystal River Plant Unit 3 r (a3 8- .-a u w b $ h dhdSh ddE@EN.MN$!@$$$Eid3E%
inoicating rna genbrst ' plant ' layout'is showiiffigum 1,~ Appendix B. ~ ~ ~ a n\\ .( J 2-A cutaway drawizg cf th2 re ct:r, cont';irunent. 'and structure crrange-3 nent is shown in Figure 3, Appendix 8. Table 1-2 in the Application ~ O. 4 sets forth tabui:.r comParisoa of the desion Para eters of the Pre-5 posed Crystal River Plant Unit 3 with the Metropolitan Ediscn Com. ^ 6 pany's Three Mile Island Nuclear Station, Duke Power Conpany's 7 Oconee Uni ts 1 and 2, and Florida Power and.Ligh't Ccapa'ny'stTurke'y- - ,.8 Point. Units 3 and.4.. Following is a suumary of those prinicpal - ~ 9 features of the plant which am considered significant to safety. 10 3.2 Reactor and Primary ~ Coolant System 11 The reactor for the Crystal River Plant Uni.t_3 is of the N '*12 ~ ~ 'rssurized ' water type. It ha's an inittal rating of^ 2452 HWt, pe 13 corresponding to a gross electrical output of 855 fWe.(9) The 14 nominal operating pressure for the reactor is 2185 psig, wi.th an 15 average tenperature of 579 F. The reactor. coolant system is 16 designed for 2500 psig pressure and 650 F tenperature.(10) 17 The reactor core is approximately 129' inch'es in' diameter, with 18 an active height of 144 inches.(11) It is made up of 177 fuel 19 assemblies ~, each consisting of a 15 x 15 array of rods enclosed in 20 a squart, stainless steel, perforated envelope. Each assembly con-21 sists of 203 zircaloy tubes containing uraniun dioxide,16 control 22 rod guide tubes, and a center tube available for an in-core instru-23 nentation assembly.(12) There are ap;,ruximately 201,520 pounds of 24 uranium dioxide in the core.(10) 25 The thernal and hydraulic design limits of the core are conser-26 27 ~ vative, and are consistent with those of other pressurized water reactors currently in operation or under construction.(10) & (13) Q~ 28 Core reactivity is controlled by a combination of 69 mnvable 29 control rod assemblies and a neutron absorber dissolved in the m I
'yy 1 collant. Th3 control rods 'are_.an alloy cf silv;r-indium-cadnium - i )' 2 enc psulated in statal;ss steel. Th; dissolved neutron?absorter1 u-3 is boric' acid.(14) 4 The control rods are,used for. short-tenn reactivity co-ml 5 associated with the changes in power level and also with changes 6. in-fuel burnup between periodic adjustments!of dissol'ved boren ~ ~ ~ 7 concentrati6n.(15) The' rea'c't'o'r 'c'an be'sfiut down by the movable 8 control rods from any power leval at any time.(16) Each movable 9 control, rod assenbly contains 16 control pins and is actuated by ~ ~ ~' 10 a separate control rod drive mechanism mounted on the top heat a f_ li the, reas tor. vessel. Upon -trip, the 69 control rod assemblies' fall J' g x 12 f'nto' t'Ycore by gravity.'(17) h 13 Systems are provided so that the concentration of dissolved 14 neutrun absorber in the reactor mAy be. adjusted to maintain-the -- ----- 15 g reactor shutdown at room temperature and to provide a safe shutdown .16-margin during. refueling.(18) The concerit' ration' of dissolved abqorber, _ ~ '17 is reduced to conpensate for' long-term reactivity chances, burnup 18 of fuel and buildup of fission products over the core cycle. 19 The core is contained within a cylindrical reactor vessel hav-20 ing the dinensions of 14 feet 3 inches inside diameter and 37 feet 21 a inches in over-all height. The vessel has a spherically dished 22 bottom head with a bolted removable spherically dished. top head.(19) 23 The reactor vessel is constructed of carbon steel with all interior 24 surfaces clad with austenitic stainless steel. The reactor. vessel 25 is manufactured under close quality control, and several types of 26 nondestructive. tests are perfonned during fabrication. These tests 27 include radiography of ' welds, ultrasonic testing, magnetic particle O 8
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I to irradiation similar to that to which the shell of the reactor 2 vessel is exposed. They can be removed periodically and tested to ascertain the effects of radiation on the reactor vessel material.(21) 3 4 Tuo coolant loops are connected to the reactor vessel by nozzles 5 located near the top of the vessel. Each loop contains one steam 6 generator, two motor-driven coolant punps and the interconnecting 7 piping. The reactor coolant piping is. carbon steel clad on the inside g sm I- '8 surface with austenitic stainless steel.(22) Reactor coolant is punped 9 from the reactor through each steam generator and back to the 10 reactor inlet by twe 8Fs,000 gpa centrifugal pumps located at the 11 outlet of each steam generator.(2/; i 17 The steam generator is a vertical, straight-tube-and-shell heat exchanger which oroduces superheated steam at constant pres-13 1 14 sure over the pwer range. Reactor coolant fims downward through 15 the tubes, and steam is generated on the shell side.(24) 16 The reactor coolant puits are vertical, single-speed, shaft-17 scaled units having bottom suction and horizontal discharge. Each 18 pump has a separate, single-speed, top-mounted notor, which is 19 connected to the pump by a shaft coupling.(23) i3 The pressurizer, a vertical surge tank approximately half-filled 21 with reactor coolant and half-filled with steam is connected to the 22 reacter coolant sys tem to control system pressure. The operating 23 pressure of the system is maintained by operating electric inersion 24 heaters to increase pressure or by spraying reactor coolant water 25 into the steam within this pressurizer tank, to reduce pressure. 26 Self-actuated safety relief valves connected to the pressurizer 27 prevent overpressurization of the reactor coolant system.(25) .g l I s :qsoanMTEf7MW"N5MA PM#
o 1 3.3 Reactor Building d 2 The reactor building is designed to enclose completely the 3 reactor coolant system and portions of the auxiliary and engineered 4 safeguards systems.(26) It is a reinforced concrete structure in 5 the shape of a cylinder with a shallow domed roof and rising from a 6 flat foundation slab. The cylindrical portion is prestressed by a 7 post-tensioning BDRV system consisting of horizontal and vertical I 8 tendons. The dome has a three-way-post-tensioning system. The 9 foundation slab is reinforced with conventional mild steel. The 10 entire structure is lined with welded steel plate 3/8 inch thick l 1 11 wall and dome and 1/4 inch thick bottom, to pmvide vapor tightness. 12 The concrete foundation mat will be appmximately 9 feet thick with 13 a 2-foot thick concrete slab above the bottom liner plate. The 14 cylinder portion will have an inside diameter of 130 feet, a wall 15 thickness of 3 feet 6 inches, and a height of 157 feet. The shallow 16 dome roof will have a large radius of 110 feet, a transition radius 17 f feet 6 inches, a thickness f feet and a height from spring O 18 line to apex of approximately 32 feet 4 in.(27) 19 The building is designed to sustain safely interial and. 20 external loading conditions which may reasonably be expected to 21 occur during the life of the plant or which could result from the 22 worst postulated accident to the reactor's primary coolant system. 23 The tendon system used in the structure is of the unbonded type 24 with a protective compound used to prevent corrosion.(28) 25 The reactor building is so designed that, with the engineered 26 safeguards systems provided, the leakage of radioactive materials 27 to the environment will result in doses well within AEC's 10 CFR 28 Part 100 guidelines for any of the postulated accidents.(29) The Q 29 integrated leak rate at design pressure (55 psig) will not exceed 30 1/4 of one percent by weight of the contained volune of air in 24 i 31 hours. (30) ) = acww.mrymeRNFPMMMMMMW'M'MROM"" l
c ! p 1 Prior to operation of the facility the reactor building will 2 be subjected to a structural integrity test and leak rate test. i 3 The structural integrity test will be conducted at 115 percent of 4 design pressure, and the leak rate will be conducted at design 5 pressure. Periodic leak rate tests will be performed to assure 6 integrity of the reactor building.(31) A tendon surveillance 7 capability will be available to provide assurance that the tendons 8 are free from harmful corresion and that excessive steel relaxation f., [ 9 has not taken place.(32) l 10 3.4 Engineered Safeguards 11 Engineered safeguards are provided to fulfill che following ~ 12 functions in the unlikely event of an accident: 13 a. Minimize the release of fission products from the fuel 14 to the reactor building atmosphere. g 15 b. Insure reactor building integrity and reduce the driving 16 force for building leakage. 17 c. Remove fission products from the reactor building 18 atmos phere. 19 The engineered safeguards systems can be grouped into an 20 emergency core coolirg system, reactor building cooling systems 21 and fission product removal systens.(33) 22 The emergency cor= cooling systen. cor.tains both passive flood-23 ing and punping equipment. The passive flooding equipment consists Q 24 of two pressurized core flooding tanks which automatically discharge 25 borated water into the reactor vessel in the event that the reactor c'6 system pressure drops below 600 psi. The pur: ping equipment consists l % w, u1mwrumenwr;gwmmu.;# mms &gw-g
1 of two completely independent sub-systems. Each sub-system contains 2 both a high pressure and a lw pressure injection pump. Either sub-3 system, in conjunction with the core flooding tanks, is capable of 4 protecting the core for any size leak up to and including the double-5 ended rupture of the largest reactor coolant pipe. Either sub-system 6 can supply coolant directly from the borated water storage tank or by 7 recirculation from the reactor building sump through heat exchangers 8 which cool it before it is retumed.to. cool the core.(34) 9 The mactor building cooling system, which is made up of two 10 separate and independent heat removal systems, limits the pmssure 11 in the reactor building following a loss-of-coolant accident. Ona 12 system contains three separate fan and cooler units. The other 13 system contains redundant spray headers which spray low temperatum 14 borated water into the reactor building to cool the reactor building 15 atmosphere. Each of these systems independently has the heat mmoval 16 capability to maintain the reactor building pressure below its design 17 pressure level.(35) 18 Control of fis -ts following a loss-of-coolant acci-19 dent is provided by .or building itself and by a second 20 separate engineered safety -feature for limiting mie'ase of fission 21 products from the reactur building. The second neans for fission 22 product contml is the iodine removal spray system which utilizes 23 sodium thiosulphate mixed in the reactor building spray water to 24 absorb the iodine released from the reactor coolant system and ren-25 der it unavailable for leakage from the reactor building. The 26 reactor building and the iodine removal chemical spray system will j 27 limit radiation doses at the exclusion radius and low population i l 28 zone boundary to values within the 10 CFR 100 guideline values.(36) i 29 3.5 Instrumentation and Control 30 A complete and dependable network of instrumentation and con-31 trols will be provided to insure the safe operation of the Crystal O, I l d mnnw num w=m w w w ww w=
l (3 1 River fluclear Generating Unit. The reactor pmtective system moni-O 2 tors parameters related to safe operation and shuts down the mactor 3 if an operating limit is reached.(37) This will be accomplished by ~ 4 interrupting power to the control rod drive clutches and allowing 5 the control rods to drop into the reactor core.(38) Alarms (39) 6 are provided to alert the operator to abnormal nperating conditions, 7 and interlocks (40) are provided to prevent abnormal operations 8' which could lead to potentially hazardous conditions. .s a 9 The nuclear instrumentation system monitors reactor power from s 10 startup level through 125 percent of full power operation. There 11 are separate overlapping instrnmentation channels for the startup 12 power ranne, the intermediate appro,.ch to power range and the pwer 13 operation range.(41) A control system automatically monitors reactor i 14 system conditions and the load requirements on the turbine-generator 15 unit, and adjusts reactor pwer, steam generator feedwater flow and Q 16 the turbine throttle for safe, efficient operation.(42) 17 The engineered safeguards protective system monitors plant 18 conditions and automatically ini+iates operation of the engineered 19 safeguards systems, if required.(43) 20 Following proven pwer station design philosophy, all control 21 stations, switches, controllers, and indicators necessary to startup, 22 operate, and shut down the nuclear unit will be placed in the cen-23 trally located control room. There will be sufficient information 24 display and alarm monitoring to insure safe and milable operation 25 under normal and accident conditions. 26 3.6 Electrical Systems 0 27 the desisa or the eiectricei systems for crvstei aiver eieat i 28 Unit 3 is based on providing the rvquired electrical equipment and a 29 power sources to insure safe, reliable operation and safe, orderly l 0 i kxxmmw=wwwmwwwwe+mrew-v
I shutdown of the unit under all normal and etergency conditions.(44) 2 The main unit itself is designed to withstand a full load dLanp with-3 out a trip-out, and as explained elsewhere, the reactor control 4 system is designed to run back to 15 percent of full load steam 5 generation under these conditions without a reactor scram. Off-6 site and on-site sources of power, each possessing redimdancy are 7 available to insure a supply of electrical energy to the plant 8 safety systems under all accident conditions including the loss-of-9- coolant accident, as outlined below: 10 a. The Unit 3 startup transformer will be connected to the 11 230 kv substation and will be sized to carry the aux 111 aries 12 required for full load on the turbine-generator. This 13 transforver will serve as the normal source of power for 14 safeguards equipnent. Four 230 kv transmission lines - 15 two from Central Florida, and two from Curlew, as well as 16 either or both of Units 1 and 2 at the sane site, can l 17 supply po<er to this transforiner. 18 b. In response to the ACRS concern expressed in the ACRS 19 letter dated May 15. 1968 to the Chairman of the U. S. 20 Atonic Enerqy Comission regarding FPC cor pliance with 21 Criterion 39, the Unit No. I and 2 startup transforver 22 will be connected to supply a redundant feed to the 4160 1 23 Vol t engineered safeguard busses. s 24 c. Upon loss of all sources of pcwer described in a a b above. 25 pwer.<ill de supplied from two quick-starting diesel-26 generator units connected to safeguards busses. The ~, 27 diesels are sized so that either car carry the required ic1 28 engineered safeguards load. A preliminary estimate of 29 the rating of each emergency generator is 2850 kw. hj -! pa y h r n_ n.n.n n.e nnn+wan_ 1
1 The unit will generate electric power at 22 kv, which will be -.s 2 fed through an isolated phase bus to the unit main transformer where 3 it will be stepped up to 500 kv transmission voltage and delivered 4 to the substation. The substation, in tum, is linked to Applicant's 5 existing transmission network as follows: the 230 kv substation is C connected to the existing FPC transmission network by four circuits, 7 two going south to Curlew and two going east to Central Florida. 8,,,The new 500 kv substation will include one outgoing line to Central ~ Florida and one to fiorth Pinellas. There will be no transformation 9 i 10 tie between the 500 kv and the 230 kv substations at the Crystal 2 11 River Generating Plant. 12 3.7 Auxiliary Systems 13 Auxiliary systems are provided to supply reactor coolant makeup 14 and seal water, to cool the reactor during shutdown, to cool com-15 ponents, to ventilate station spaces, to handle fuel, and to cool 16 spent fuel. 17 Reactor coolant makeup and seal water is supplied by the makeup 18 and purification system. This system maintains the proper coolant 19 inventory in the primary system, maintains the seal water flow, 20 adjusts the concentration of dissolved neutron absorber 17 the reac-21 tor coolant, and maintains proper water chemistry.(45) The & cay heat reinoval system cools the reactor when the reactor 22 23 system is depressurized for raintenance or refueling. This same 24 system serves the enginecred safeguards function of recirculating 25 borated water to cool the core in the event of a loss-of-coolant 26 accident.(46) a The cooling water systems maintain temperatures throughout the j 27 28 equipment aad structures of the plant.(47) Appropriate nomal 29 ventilation systems are provided in the plant.(48) 3 i O o 3 " i
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1 A fuel handling system (49) provides the means for safe, m e 2 reliable handling of fuel from the time it enters the plant site ] g 3 as new fuel until it is shipped from the plant site as used fuel. i 4 Irradiated fuel is scored and handled under water at all times [ 5 until after it is placed into a shippin; cask. The water provides f 6 a radiation shield as well as a reliable source of cooling for the 1 7 irradiated fuel asserblies. ' A spent fuel cooling system mal'ntains j ~ 1 1 8- > the temperature of the spent fuel storage pool water within accept- 'j 9 able limits.(50) D N 10 3.8 Steam and Power Conversion System 3 11 The steam and po<er conversion system is designed to remove the 12 heat energy generated in the reactor core by producing steam irt the ~ 13 two steam generators. This heat energy is converted to electrical 'j 14 energy by the turbine-generator. This cycle, including the necessary 15 equipment to achieve safe and reliable operation, is similar in con-
- .a g 16 cept and design to turbine-generator cycles in successful use f.r 17 nany years.(51) 1 L
o 18 3.9 Radicactivi ty Control System [
- 9 Radioactive gaseous, liquid, and solid wastes in the station 20 are handled by the waste disposal systems. These systems contain 21 the equipment necessary to collect, process, and prepare for safe 22 disposal of the radioactive wastes which result from reactor opera-23 tion. These syste
- :s are designed to minimize the release of radio-24 active material from the plant to the environment and will maintain 25 releases belod the linits of 10 CfR 20.(52) & (53) 26 A process radiation ronitoring system conitors effluent released 27 to the envirtnment and provides an early warning of possible equipment
, ) h 28 nalfunction or potential radiological hazard. The radiation ronitoring q 29 system includes a corbination of continuous automatic ronitoring and f 33 periodic sa@ ling.(54) & (55) x Nh w
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3 1 Shielding throughout the unit insures that radiation doses N 2 to the general public and to operating personnel during normal J M 3 operation are well within the limits of 10 CFR 20.(56) y & J 1h 4 4. SAFETY NIALYSES 'lJ 5 Potential malfunctions or equipnent failures have been analyzed v.s] 6 to provide a safety evaluation of the Crystal River Plant Unit 3. 7 This evaluation demonstrates that the public will not be exposed to ~ M
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radiation in excess of the limits established in the AEC's regulation f-{- 9 for siting requirements,10 CFR 100, even in the very unlikely event t 10 that one of the accidents postulated in the Application should occur.(57) 's 11 Two categories of malfunctions or equipment failures have been L-g J 12 analyzed; those in which the core and coolant boundaries are protected f 13 and those in which one of these bomdaries is not effective and stan&y [ 14 safeguards are required. The core and coolant boundary protection 15 analysis shws that, in the event any of the postulated malfunctions 16 were to occur, the normal protection systems operate to maintain the f 17 integrity of the core and of the coolant boundary.(58) The stan&y g.: 18 safeguards analysis demonstrates the capability of the engineered ( 19 safeguard syste'ts to assure protection of the public for postulated F 20 malfunctions in which the nonnal protective systems may not maintain 21 the integrity of tt ' core and coolant boundary.(59) These analyses M 22 shw that for all credible malfunctions the radiation exposure to ef 23 the general public is well belm the limits prescribed in 10 CFR 100. c 24 Of the pnstulated equipaent failures, a loss-of-coolant acci- [ 25 dent is the most severe. Erergency core cooling equipment is pro- ~( 26 vided to prevent clad and fuel damage that would interfere with , }; 27 continued core cooling for reactor coolant system failures up to 28 and including the complete severence of the largest reactor coolant ,{ g 29 pipe. The core cooling system insures that the core will remain in g 30 place and intact.(60) The reactor building spray or emergency cool-g 31 ing units maintain the integrity of the reactor building,(61) and k y __ yp=
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b 1 the iodine removal sprays in conjunction with the ret.ctor building 2 assure that the public is protected from radiation and radioactive ?:t 3 materla1.(62) Emergency electrical power is available on-site to w.V 4 insure operation of these systems even if all external sources of ((k!f 5 electric power to the plant are assumed to be unavailable at the [ 6 time of the accident.,(63) ?? .. Q };. 7 Results of the safety analyses snow that, even in the event of ~ l 1 8 a loss-of-coolant acdidentf no" core melting will occur.(64) However. 9 in order to Jemonstrate that the operation of a nuclear power station j F 10 at the proposed site does not present any undue hazard to the general 11 public, a hypothetical accident has been analyzed involving release .9 12 of 100% of the noble gases. 50% of the halogens and 1% of the solids . iN fi 13 in the fission product inventory. The analysis evaluated both the i
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14 direct radiation exposure and the potential total dose to the thyroid M 15 from the inhalation of fission products which leak from the reactor g-15 bui lding. The icw leakage rate of the reactor building and the g ~ "; 17 iodine removal spray system reduce the potential radiation dose to M[ 18 the thyroid to below the M CFR 100 guidelines even in the event 19 of such a hypothetical occurence.(65) O:n tM 20 5. TESTS, INSPECTIONS, AND QUALITY CONTROL 21 Pressure containing cor:ponents of the reactor coolant system 22 will be designed, fabricated, inspected, and tested in accordance ( 23 with Sc tion III, Nuclear Vessels, of the Arerican Society of h-24 Hechanical Engineers Boiler and Pressure Vessel Code. The piping ,M 25 will r:eet the applicable provisions of Power Piping USA Standards 25 B31.1.0-1955 and associated nuclear code cases.(66) Nondes truc-Q 27 tive testing, including radiography, ultrasonic, magnetic particle, 28 and liquid penetration examinations will be performed during fabri-g 29 cation of the nuclear vessels. ~. Q mg_ 30 Auxiliary systems and equipment will be designed, fabricated, 31 and tested to the appropriate provisions of recognized codes and w-M I i . :p F'
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1 standards of organizations such as the Anerican Society of Mechani-Jl 2 cal Engineers, American Society for Testing Materials, USA Standards Ak-3 Institute, and Institute of Electrical and Electronics Engineers. 1h M_ h 4 A comprehensive field testing program will be conducted to M]- f 5 insure that equipment and systems perfonn in accordance with design 6 cri teria. Tests will be perfonned both before and after fuel load-hg 7 ing and criticality.(67) ,~ g@h 8 The reactor building will be designed and built in accordance 9 with appilcable portions of 15 Building Code Requirements for p 10 Reinforced Concrete, ACI 318-63; Specification for Structural ]3g 11 Concrete for Buildings ACI 301-66; AISC Hanual of Steel Construc-12 tion; ASME Botier and Pressure Vessel Code Sections III VIII, 'g 13 and IX.(68) Materials and wortmanship will be inspected to insure 9fjf 14 compliance with appropriate codes. specifications, and standards. 15 Materials to be inspected and tested include concrete. liner plate, ,[ 16 prestressing system materials, hatches, penetrations, structural 59 17 and reinforcing steel.(69) G.'! 18 The reactor building will be structurally tested at 115 percent FEM 19 of design pressure.(70) In addition, it will be leak tested to M 20 insure cocpliance with a maximum allowable gross leak rate of one sg 21 fourth of one percent by weight of the contained volume of air in 'i{ 22 24 hours (71) at the design pressure. The per.etrations will be Q 23 periodically pressurized to design pressure to demonstrate their ] 24 leak tightness.(72) 25 Consideration has been given to the inspectability of the M 26 reactor coolant system in the design and arrangement of components. _ [@M Access for inspection of the reactor coolant system includes access 27 28 for visual examination by direct or remote means. g ' Gi: w ME O "y' i c ex- ,C $,yr m;y, 3;y n:5L G i s h M D D K JiYk U 2 E I S T W 2 5 Kk~ W M M
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1 The Applicant's contractors and equipment suppliers will pm-s l vide required quality control functions, pmcedures, and techniques 2 3 g to assure manufacture and construction in accord with the plant 4 design and specifications fumished to the Applicant by its 5 architect / engineers, Gilbert Associates, Inc. 6 In addition, the Applicant, thmugh its Power Engineering and Construction Department, will provide a quality assurance program 7 8 using as appropriate its own personnel, those of Gilbert Associates, c 9 and/or qualified independent testing agencies to monitor contractors j 10 and suppliers compliance with the above quality control requirements. ] 11 This quality assurance gmup will include experienced and trained / 12 personnel qualified in areas of specialty necessary to assure con-13 formance with the high quality standards dictated by plant design e" 14 and specifications. g 15 6. RESEARCH AND DEVELOPMENT PROGRAMS [/j 16 The nuclear steam supply system for the Crystal River Plant j 17 Unit 3 is similar in concept to several pmjects already in opera-jc 18 tion, under construction, or recently licensed by the Atomic Energy p 19 Connission. The preliminary design is based on the technical data y 20 which have been developed in the nuslear industry and on data deve- [ 21 loped by B&W which is specifically related to the Crystal River ,07 22 Plant Unit 3. To conplete the final detail design of some compon-23 ents additional technical information will be obtained. 24 The following are the areas of the plant design in which addi-( 25 tional technical data will be developed to finalize design details. 3$ ,. E 26 a. Once-Throuch Steam Ganerator ~L g 27 The design of the once-through steam generator is based on l j,~ 28 experimental work on boiling heat transfer and data c6tained 29 by B&W in full length codel tests of the unit. The testing } 6 'l$ " q kmeswm u.;<wmCW M ?MM#N""~
1 of a prototype unit has been completed but not yet documented. 2 It includes perfomance, mechanical, vibration and blowdown 3 tests, and control system development. The msults have 4 confimed the analytical predictions of performance, and 5 sufficient data on the performance and structural design 6 has been,obtained from operation of the test modals to 7 finalize the design of the steam generators.(73) 8 b. Control' Rod Drive Unit 1 I 9 The design of the control rod drive mechanisms is baseo on 10 a principle which has been used in operating reactors and 11 which has been extensively tested by B&W. Test prograw -t 12 have included full scale prototype testing under no-flow 13 e
- i conditions, full scale prototype testing ai operating 14 conditions, and components testing. Testing of a prcto-7 15 type mechanisc was carried out for a full-life cycle of l
16 g strokes and trips, and major design parameters were 17 con fi rred. L.ife cycle testing is being repeated using 18 a pinion gear of improved material. Data from these test "W 19 programs will be used in the final design of the control 'Q 20 rod, its guide structure, and the control rod drive Q 21 mechanism.(74) Y!. 22 c. In-Core Neutron Detectors .y The performance and longevity of the self-powered detectors 23 a L 24 is being demonstrated by detectors installed in the Babcock g .i 25 and Wilcox Test Reactor and in the Big Rock Point Nuclear 'A 26 Power Plant. The tests have demonstrated that the detectors
- f 27 perfom successfully. They are being continued in order to y
28 demonstrate their longevity. At the present time, the Big g 29 Rock Point detectors have accumulated operational experience V.i 30 equivalent to three years of operation in the Crystal River ,M 31 reactor. The Babcock & Wilcox Test Reactor detectors have vLg 32 accumulated an equivalent of two years operation. > :cs M A b[h' ~ mwwmmmwm mmmM"*"" "
1 d. Core Themal and Hydraulic Design 2 The PSAR as originally submitted contained, in Section 3, 3 an evaluation of the core thermal capability in which the 4 heat transfer limits were predicted based on a cormlation 5 of experimental DNB (Departure from Nuclear Boiling) data 6 developed by The Babcock & Wilcox Company. In order to 7 cogletely substantiate the B&W cormlation additional 8 research and development data is necessary. 'These ' require-9 ments are described in the PSAR.(75) 10 Subsequent to submittal of the original PSAR, core thermal 11 perfomance was also evaluated using the W-3 correlation 12 for predicting DNB. This correlation is available in the 13 literature and has been used and found acceptable in 14 establishing themal design limits for other large pies-15 surized water reactors. The themal evaluation using the g W-3 correlation is also presented in the PSAR and its 16 17 supplerents. With the use of this correlation, vessel 18 nodel flow tests are necessary to substantiate operation 19 of the plant within acceptable themal limits. Flow test-20 ing which demonstrated acceptable flow distribution for 21 the rated power level without intemal vent valves in 22 the model has been completed. Flow testing with internal 23 vent valves installed and with open internal vent valves 24 must still be done. 25 Etercency Core Coolino and Intemal Vent Valves e. 26 Analytical evaluation of the effects of blowdown forces on 27 the intemals and of the perfomance of the intemal vent 28 valves installed in the core support shield to insure Q 29 adequate covering of the com by emergency coolant is in 30 p mgress. A prototype of these valves will be tested to 31 demonstrate their operating characteristics.(76) O 4 y D M :,_._.,,-_-- _.. "-s n e s q p~~T W-WMWMiF7Ii ' ME#"
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1 f. Fuel Failure O 2 A study, including testing, is underway to assure that 3 g there are no failure nechanism: which might interfere 4 with the ability of the emergency core cooling systems to 5 acconplish their objective. The results of '.he work to 6 date demonstrate the ability of the design to acconynodate 7 potential fuel failure mechanisms. This work will be B continued to assure that fuel red failures will not affect 9 significantly the ability of the emeryency core cooling 10 system to prevent clad melting. 11 g. Xer,on Oscillations _ / 12 The possibility of the occurrence of xenon oscillations 13 throughout core life is being evaluated. If it is 1 14 determined that such oscillations may occur appropriate j 15 design changes to eliminate or control the oscillations O ""' be 'acorporated-(77) '"* das'9" ' =""a' 'a ' " ='"*** 17 or control such oscillations is being carrie. cut in IB parallel with the studies of the possibility of such 19 oscillations. 20 h. Sodium Thiosulphate 21 i One of the radiological protection systems of the Crystal 22 River Plant Unit 3 provides chemical sprays into the 23 reactor building to remove iodine under accident conditions. 24 Testing to demonstrate the ability of the chemical sprays 1 25 t to remove and retain todine effectively and to demonstrate 26 the compatibility of the chemical with plant materials is 27 in progress.(78)
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1 7. TECHNICAL QJAL*IFICATIONS 9 2 7.1 Florida Power Corporition 3 Florida Power Corporation is responsible for the design, pur-4 chasing, construction, and operation of Crystal River Flant Unit 3. 5 This practice has been successfully followed for all of-the Company's 6 major generating facilities nw in se~rvice or planned. 7 The Applicant has 68 years experience in the c'esign, construc-8 tion, and operation of electric generaag pi. ants. 9 At present, Florida Power Corporation operates eight steam-10 electric generating plants containing a total of 23 units with a 11 net capability of 1,512,000 kilowatts, one hydroelectric plant 12 with a capacity of 8,400 kilowatts, and two internal combustion 13 generating units with a total capacity of 2,000 kilowatts (exclud-Q 14 ing robile units) for a total net elcctric generatirg capability 15 of 1,522,400 kw. 16 -The Applicant has under construction at the Crystal River Plant, 17 Unit 2, which is a new 510,000 kw coal fired steam electric generating i unit and 4 gas turbine engine driven units of 30 ni each at two sepa-18 l 19 rate plant locations. 20 7.1.1 P_cwer Engineering 21 The Applicant will provide engineering direction, management, 22 and technical supervision for all systems, equipment and structures 23 which comprise the Crystal River Nuclear Plant. A staff of graduate 24 engineers will be responsible for engineering studies, design specifi-h 25 cations, procurement, enginee-ing approval and coordination, and 26 construction engineering liaison, in each engineering discipline 27 relating to the design and construction of the Crystal River Nuclear 28 Plan t. (79) N ! ( = M M E X m " 2~~DLM3m;rrmM
9 The responsibility for establishing and executing a nuclear 1 2 training program and maintaining liaison with the Power Production O ""*"*"'""""'"d"'""" ' '"' " "*" 4 Engineering staff. 5 Power Engineering is under the direction of the Manager of Power Engineering who is responsible to the fluclear Project Manager. a 7 7.1.2 Power Construction _ 8 Florida Power Corporation has a staff of graduate engineers to 9 provide the construction rnanagenent and engineering skills including 10 quality contml required during the course of power plant construction 11 and is responsible for construction nanage,ent of the Crystal River 12 fluclear Plant. 13 With its nucleus of six experienced construction engineering Q 14 and supervisory personnel, the Applicant's construction staff 15 manages all activities of sone 500 power plant construction 16 workers employed by subcontractors to the Apolicant performit g 17 construction work. An Applicant owned or leased inventory of 18 essential and modem construction equipment is kept in readiness, 19 and the latest construction techniques are employed, including 20 computer processed CPM analysis of the construction activities 21 and computerized cost accounting and cos t contml. The construc-22 tion section is under the direction of the ibnager of Power 23 Construction who is responsible to the fluclear Pmject Manager.(80) 24 7.2 The Babcock and Wilcox Company 25 BAW's participation in the developr:ent of nuclear power dates 26 fmm the Manhattan project. B&W's broad auclear activities include O 27 ePPiied 'esearch to deveioP rundameatai eeta: eesi a eaa meaurecture 9 i 28 of nuclear system, cores, and compor.ents; and design, manufacture, 29 and erection of complete nuclear stea, generating sys tems. Through l0, F ~-lit, b v 6
1 B&W's divisions a wide range of equipment for nuclear application 2 is designed and manufactured. The B&W Coppany's rejor nuclear "' '" 'dd' " * * " " ' ' ' ' " ' " ' ' '"'"' "'"** e i 4 of components for the nuclear Navy, have included Indian Point No.1, ( 5 NS Savannah, Advanced Test Reactor, Oconee Nuclear Station Units 1, 6 2 and 3. Three Mile Island Nuclear S,tation and 3 other units in 7 various stages of licensing in addition to Crystal River Plant 8 Unit 3.(81) J 9 In addition to supplying the Crystal River Plant Unit 3 nuclear 10 steam system BAW's Power Generation Division will erect this equip-11 pent and be responsible for the quality control procedures requimd 12 during erection phases. s l l 13 7.3 Gilbert Associates, Inc. I 14 Gilbert Associates, Inc. has been retained by the Applicant as g 15 tha Architect-Engineer for this pmject. They will furnish plant 16 layouts, d grams, and system arrangenents and provide specifications 17 for major items of equipment and systems. Pmvision is made for 18 consulting services required as well as resident engineering per-19 sonnel during construction. I i 20 Gilbert Associates. Inc., engineers and consultants, was organized ?1 in 1906 and has its rnain office at Reading, Pennsylvania. Since 1942 r 22 GAI has been responsible for the d2 sign of over 110 thennal generat-23 ing units, both fossil and nuclear, representing rore than 16,000,000 24 kilowatts of capacity. Design experience includes reheat cycles, 25 once-thmugh boiler units, and supercritical units in ratings up to 26 900,000 kilowatts. At present GAI has over 8,003,000 kilowatts of 27 generation under design. 28 Gilbert pmjects since 1950 include corplete programs of nuclear 29 power development involving analysis of sites, complete evaluations i, O i % mmmmmme m.m eme,mm---
1 of proposals, contract and fuel prt, gram assistance, preparation of s 2 license applications, comlete plant design and procurement.(82) .;1 1 3
- 7. 4 J. A. Jones Construction Company 4
The J. A. Jones Construction Co@any will provide general .q 5 contractor services for Crystal River Plant Unit 3. Through a 6 subsidiary cogany, Livscy & Comany, Inc., the trechanical equip-7 rent and piping erection will be accomplished. a 8 This most highly qualified construction firm brings to this S 9 project its significant nuclear project experience over the past 10 yea rs. Besides work on the Oak Ridge gaseous diffusion plant and j 11 the plutonium production reactors at Hanford their latest work is 12 the completion of the Jersey Central Oyster Creek Unit #1 under ~ 13 contract to General Electric. u g J. A. Jones Construction Company has developed and established 14 15 quality control procedures, techniques and testing skills as a result 16 of this extensive experience with the AEC. This background will q 17 bring to the Crystal River project a more co@rehensive program for 18 overall quality control. 19 8. COMMON DEFENSE AND SLCURITY 20 There is no indication that construction and operation of the 21 Crystal River Plant Unit 3 will in any way be inimical to the 22 cocumn defense and security of the United States, v 23 As stated in the Application, Florida Power Corporation is a Flcrida CorpGration engaged as a public utility in the production, 24 25 transmission, and sale of electric energy. All of the directors g 26 and principal officers of the Co pany are citizens of the United 27 States and the Comany is not owned, controlled or dominated by 28 an alien, a foreign corporation, or a foreign govemment. .10 1 . wnw ma mw ~ -mm um=m- =--
s i 1 The application contains no restricted or other defense j 2 information and Applicant has agreed that it will not permit 3 any individual to have access to Restricted Data until the Civil q Service Conrnssion shall have made an investigation and report ) 4 5 to the Atomic Energy Cocnission on the character, associatioas
- i j
6 a-d loyalty of such individual, and the Atomic Energy Cc :nission I 7 shall have determined that permitting such persons to have access t' 8.: to Restricted Data will not endanger the common defense and 'l 9 securi ty. l 10 As a licensee, Applicant will be subject to regulations of the 11 Atomic Energy Connission relating to the transfer of and accountability i 12 for special nuclear material in its possession. Recent amendments j
- j 13 to the AEC Rules and Regulations (10 CFR 50.60) under which the AEC 14 will discontinue allocating quantities of special nuclear rnaterial 15 to reactor licenses evidence that such material is no longer scarce.
(g 16 Moreover, in the event of a state of war or national emergency the 17 AEC may order the recapture of special nuclear material, as well as 18 the operation of any licensed facility. (10 CFR 50.103). 19 9. CONCLUSI0fl i 20 On the basis of the foregoing and the Application, the Applicant 21 respectfully submits that; j 1 22 a. Florid t Power Corporation's Application, as amended, des-23 cribes the preposed design of the Crystal River Plant i 24 Unit 3, iv.luding the principal architectural and engineer-25 ing criteria.Sr the design, and identifies the raajor fea-26 tures or conponents incorporated in the plant for the 27 protection of the health and safety of the public. g h ] 28 b. the Application, as amended, identifies the technical or f 29 design information necessary to conplete the final safety 1 hi. 9 1q m s mm m=am mmmeu=== w=*!"!!!em
l It 9 1 analysis. Such information can reasonably be lef t for 2 later t insideration and will be supplied in the final 3 safety analysis report. 4 c. Safety features which require further msearch and develop-g 5 rent, and the research and development, programs to be carri-6 ed out, are identified in Volume 1. Section 1.5, Volume 4, 7 Question 1.4, and Supplement 2. Question 6 of the Application. 8 The research and development program is reasonably designed 9 to resolve any questions assoc.iated with such features at or 10 bei..re the latest date stated in the Application for comple-11 u, tion of construction of the facility. ).' l 12 d. Taking into consideration the characteristics of the site 13 and environs and the proposed design of the Crystal River 14 Plant Unit 3, such facility can be constructed and operated 15 9 within the limitations established by 10 CFR 20, within the 16 site criteria set forth in 10 CFR 100, and without undue 17 risk to the health and safety of the public. 18 e. The Applicant is technically qualified to design and 19 construct the proposed facility; and ,q 20 f. The issuance o'f a construction permit for the Crystal River 21 Plant Unit 3 will not be inimical to the comon defense and 22 securit." or to the health and safety of the public. i 1 Y i O y u Ti -g ...10 'y 7 i y m yy,c.m m;m mnanre m me e me* ~
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\\ i it 4' I APPENDIX A LIST OF REFERENCES l O k e a y 4. s $e T-l g :- O t r O ' s \\ - _,:, g i ?.cr; 2-x. t wxe2+ ware =c% =?:uts 12gpK~:::=vm=m ~-- t l t
1 ? Appendix A j LIST CP REFERENCES f' (1) PSAR, Volume 1, Section 2, Paragraph 2.2.4 g (2) PSAR, Volure 1, Section 2, Paragraph 2.3.1 Ig' (3) PSAR, Volur.e 1, Section 2, Paragraph 2.4.1 ,if (4) PSAR, Volume I, Section 2. Paragraph 2.4.2 !9 (5) PSAR, Volume 1, Section 2, Paragraph 2.4.5 (6) PSAR, Volume 1, Section 2, Paragraph 2.5 ? (7) PSAR, Volume 1. Section 2, Paragraph 2.6 f PSAR, Appendices, Appendix 21 2(4 (8) PSAR, Volurre 1, Section 2 Paracraph 2.7 l 'E .I PSAR, Volume 4, Supplement 1. C -tions 5.8 & 6.1 'A i D (9) PSAR, Volume 1. Section 1 Para ph 1.1 li 7 (10) PSAR, Volume 1, Table 1-2 ?! l 4 (11) PSAR, Volume 1, Table 3-2 \\
- 4I h y
(12) PSAR, volume 1, Table 3-1
- (:
(13) PSAR, Volume 1, Section 3. Paragraph 3.2.3 17 (14) PSAR, Volurre 1, Section 3, Paragraph 3.2.2.1.2 PSAR, Vo!ume 2. Section 7, Paragraph 7.2.2.1.2 , f\\ (15) PSAR, Volume 1 Table 3-6 ] PSAR, Volume 1, Figure 3-1 ~I (16) PSAR, Volume 1, Section 3, Paragraph 3.2.2.1.3 l r lJ (17) PSAR, Volume 1 Section 3. Paragraph 3.2.4.3.2 ,f , [;.;; (18) PSAR, Volume 2. Section 9. Paragraph 9.2 .i gJ (19) PSAR, Volume 1 Section 4, Paragraph 4.2.2.1 ./ (20) PSAR, Volume 1, Section 4, "aragraoh 4.1.4.4
- 3 (21) PSAR, Volume 1, Section 4,
'aragraph 4.4.3 y (22) PSAR, Volume 1, Section 4, Paragraph 4.2.5 ,.u (23) PSAR, Volume 1. Section 4. Paragraph 4.2.2.4 i O ~ A y, r : 2 m : m a r m :.T l N r W ; W e
o Appendix A (24) PSAR, Volume 1, Section 4. Paragraph 4.2.2.3 .n (25) PSAR, Volume 1, Section 4, Paragraph 4.2.2.2 wMh (26) PSAR, Volume 2. Section 5 hp 7 (27) PSAR, Volume 2. Section 5 Paragraph 5.1 ?h y$ (2C) PSAR, Volume 2, Section 5, Paragraph 5.1.2.1 PSAR, Volume 2, Section 5. Paragraph 5.1.2.8 {@D "4 (29) PSAR, Volume 2', 'Section 5, Paragraph 5.5' W-Ig (30) PSAR, Volume 2. Section 5. Paragraph 5.1.2.2 w (31) PSAR, Volume 2,-Section 5, Paragraph 5.6.2.1 NT (32) PSAR, Volume 2 Section 5, Paragraph 5.6.2.2 -N (33) PSAR, Volume 2 Section 6 %g (34) PSAR, Volume 2, Section 6, Paragraph 6.1 %-1 (35) PSAR, Volume 2 Section 6 Paragraph 6.2 (36) PSAR, Volume 3. Section 14, Paragraph 14.2.2.4 Y,$ (37) PSAR, Volume 2, Section 7, Paragraph 7.1.1
- p 9
(38) PSAR, Volume 1 Section 3, Paragraph 3.2.4.3
- - g,i_
(39) PSAR, Volume d Section 7 Paragraph 7.4.3 L. 1; (40) PSAR, Volume s', Sectinn 7, Paragraph 7.2.3.2 A N (4.1 ) PSAR, Volume 2. Section 7, Paragraph 7.3.1 ,MW (42) PSAR, Volume 2. Section 7, Paragraph 7.2 3 9w (43) PSAR, Volume 2. Section 7 Paragraph 7.1.2.2 '] PSAR, Volume 2, Section 7, Paragraph 7.1.3.3 k (44) PSAR, Volume 2, Section 8 V4 (45) PSAR, Volume 2, Section 9, Paragraph 9.1 ,, ' h (46) PSAR, Volume 2, Section 9. Paragrept u F q 4 (47) PSAR, Volume 2, Section 9, Paragraph 9.3 q ,y (48) PSAR, Volume 2, Section 9 Paragraph 9.7 &9 a a: Nj. A-2 5 n = A S; P -c276 E 0"b m m - m - m wwwwpenn;mwm me
Appendix A (49) PSAR, Volune 2, Section 9, Paragraph 9.6 b4,, (50) PSAR, Vnlume 2, Section 9, Paragraph 9.4 3'; h (51) PSAR, Appendices Appendix 1A, Section 1 J/f PSAR, Volume 2, Section 10 9% 4 (52) PSAR, Volume 3, Section 11, r'aragraph 11.1 r.i jv (53) PSAR, Volume 4 Supplement 1, Question 6.1 A tyy (54) PSAR, Volume 3. Section 11, Paragraph 11.'1.2.4 '_.C.# (55) PSAR, Volume 4, Supplement 1. Figure 5.8-5 M. (56) PSAR, Volume 3..Section 11, Paragraph s1.2.1 ] A .J (57) PSAR, Volume 3, Section 14 =mk (58) PSAR, Volume 3 Section 14, Paragraph 14.1 M (59) PSAR, Volume 3. Section 14 Paragraph 14.2 G JG. (60) PSAR, Volume 2 Section 6, Paragrap: 6.1 [@ m M.c W (61) PSAR, Volume 2, Section 6, Paragraph 6.2 3M $[, (62) PSAR, Volume 3 Section 14, Paragraph 14.2.2.3.5 un[jg (63) PSAR, Volume 2 Section 8, Paragraph 8.2.3 Uh (64) PSAR, Volume 3, Section 14, Paragraph 14.2.2.3 dT (65) PSAR, Volume 3 Section 14, Paragraph 14.2.2.4 Mm-(66) PSAR, Volume 1, Section 4, Paragraph 4.1.5 O? (67) PSAR, Volume 3. Section 13 m Q (68) PSAR, Volume 2, Section 5, Paragraph 5.1.2.4 (69) PSAR, Volume 2, Section 5, Paragraph 5.6 t, y; (70) PSAR, Volume 2. Section 5, Paragraph 5.6.1.2 ,y g (71) PSAR, 'Iolume 2. Section 5, Paragraph 5.6.1.3 (72) PSAR, Volume 2. Section 5, Paragraph 5.6.2.1 . :2 J }.]'r (73) PSAR, Volume 4, Supplement 1. Question 1.4 'sq g m. Qdg. A-3 's@ ^* ' K T & T k R L W n T '} }"_ M ? IN? a?A
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Appendix A (74) PSAR, Valume 1. Section 1, Paragraph 1.5.2 PSAR, Volume 4. Supplement 1, Question 1.4 (75) PSAR, Volume 4 Supplement 1 Question 1.4 O _t (76) PSAR, Volume 4, Supplement 1. Question 1.4 (- (77) PSAR, Volume 1, Section 3 Paragraph 3.2.2.2.3 ? (78) PSAR, Volume 4, Supplement 1. Question 1.4 I (79) PSAR, Volume 1. Section 1, Paragraph 1.6, Figure 1-12 I ? & Appendices. Appendix 1A, Paragraph 1.4 24 (80) PSAR, Volume 1. Section 1, Paragraph 1.6, Figure 1-12, z. & Appendices. Appendix 1A, Section 1 \\f (81) PSAR, Appendices, Appendix 1A, Section 2 (82) PSAR, Appendices, Appendix 1A, Section 3 l -g k$r h n-a . :5 T ,j 5% lr? An; g e 1 q% O ll A-4 e . - *'f:l? ] 55.'s") K???
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h 9 o S* 6 APPENDIX C PROFESS 10'iAL QJALIFICATIO*iS 0F EXPERT PANEL WITHESSES 1 O O 4 r,anzm e 2m muramem rgxm p - i m
i 1 EDUCATI0flAL AND PROFESSIONAL QUALIFICATIONS 2 DONALD J. R0WLAND 9 3 I2ittMFeAL EE"t!EHt Sea'" Pe-e Faf'=e 4 POWER ENGI! LEERING AtiD CONSTRUCTION DEPARTMENT 5 FLORIDA POWER CORPORATION 6 1. My nane is Donald J. Rowland. My residence is 6401 - 31st 7 Avenue North, St. Petersburg, Florida. I am employed by Florida 1 8 Power Corporation, Power Engineering and Construction Depart- % or R.re - Gn ra,er. s 9 ment as a Nedhees:s4 i ..r. 10 2. I graduated from Auburn University in 1958 with a Bachelor of 11 Mechanical Engineering degree. 1 .1 12 3. In 1958, I accepted a position at Florida Power Corporation in i l 13 the Mechanical Engineering Department and was assigned to field 14 supervision on power plant construction projects. 15 4 In 1959. I was assigned to General Nuclear Engineering Corpora-16 tion as Florida Power Corporation's representative on the Florida i j 17 West Coast Nuclear Group - East Central Nuclear Group, gas 18 cooled reactor research and development project. My duties were I 19 -to maintain liaison contact between the project and Florida l' ' 20 Power Corporation and to assist General Nuclear Engineering Cor-1 21 paration in design activities. l 22 5. In 1961, I was reassigned to field supervision on power plant l 23 construction projects and progressed through positions of re-h 24 sponsibility in plant construction and design engineering.. k O 1, 0 1 C-1
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l i / 00fiAl.D J. R0WLN4D il 1 6. In 1966. I was assigned my present position of Mechanical 2 Engineer. In this position I am responsible for the direction 3 of rnechanical and nuclear design engineering associated with O 4 new generating plants. i 1 e b 'l .:'.l 1 ' O t 1 -i r 4 0 ' l O ' i 'f C-2 i l
E n 1 EDUCATIONAL A'lD PROFESSI0ilAL QUALIFICATIONS 2 E. ROBERT HOTTRISTEIN O 3 PROJECT MA'{AGER, fiUCLEAR ErlGIfiEER!flG DEPARIMENT 4 POWER DIVISION 4 5 GILBERT ASSOCIATES, Ific. q 6 1. My name ;is E., Robert Hottenstein. My residence is Old State 7 Road, Oley, Pin'n's'ylvania 19547. I am employed by" Gilbert ~ q 8 Associates. Inc. as a Project Manager. In this position I 9 am responsible for the overall technical direction and 10 related a:binistrative effort that emcompasses the licens-11 Ing and engineering efforts associated with the design of q 12 Crystal River Plant Unit 3 for which Gilbert Associates, 13 Inc. is responsible, n 14 2. I was graduated from the Pennsylvania State University in Q 15 1950 with a Bachelor of Science Degree in Mechanical 16 Engi nee ring. From 1955 to 1959 I studied nuclear physics 17 and nuclear engineering at fiorth Carulina State University. 18 3. In 1950, I joined Gilbert Associates, Inc. as a Mechanical 19 Engineer. 20 4. From 1950 to 1955 I was assigned to the Knolls Atomic 21 Power Laboratory. In this assignment my entire effort 22 was devoted to engineering analyses and details on the 23 primary coolant system for sodium and light water cooled 24 reactor plants for submarines. q ~ 25 5. From 1955 to 1959, I was ass ~gned to the Reading Office [ h 26 where I participated in nuclear plant studies for the 27 U. S. AEC, U. S. Air Force, and public utilities. 5i j ~ l7 C-3 c u --_nnnwu-m umm_=m,----
P b 3 E. ROCFRT HOTTE!lSTEIN '1 1 6. From 1959 to 196?, I was assigned to the Saxton Experi-2 cental Reactor Project. For this project, I was res-3 ponsible for the design and start-up of the radioactive S 4 waste disposal facility. h ej 5 7. From 1962 to 1965 I participated in the General Public a
- s ". i,
- r. 6 Utilities Oyster Creek Proposal Evaluation Program and 7
did nuclear siting studies for the Pennsylvania Power a 8 and Light Company. &) 9 8. From 1965 to 1966 I was assigned to the Robert Emett o 10 Ginna Nuclear Station Project for the Rochester Gas and 11 Electric Company as Project Nuclear Engineer. ') 1 12 9. From 1966 to 1967 I was the GAI Project Engineer on the 13 g Florida Power Corporation !!uclear Plant Siting Study, T, 14
- uclear Steam System - Specification and Evaluation Team.
15 10. From 1967 to present, I have served as the GAI Project 16 l4anager on the Crystal River Nuclear Plant Unit 3 Project. 17 11. I am a member.of the American Society of Mechanical 18 Engineers and the American !;uclear Society, and a Licensed 19 Professional Engineer in the State of Pennsylvania. g. p 'J i j h 7 0 3 ~
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4 1 EDUCATIONAL AND PROFESSIONAL QUALIFICATIONS g 2 MORTON I. GOLDMAN 1 3 VICE PRESIDENT AND GENERAL MANAGER 1 4 ENVIRONMENTAL SAFEGUARDS DIVISION 5 g NUS CORPORATION t y 9j 6 1. My name is,Morton I. Goldman. My address is 1730 M Street, 4Ji 7 N.W., Washington, D. C., 20036. I am Vice President and d 'S 8a.,.: General Manager of.the Environmental Safeguards Division of ~ k 9 NUS Corporation and have served in this capacity since January 10 1966. I am responsible for all site evaluations, safety analyses, n 11 waste canagement syste= design and environmental program develop-MJ 12 ment conducted by this Division. This has included the evaluation 13 of site and environmental safety factors for a nunter of nuclear 14 and fossil fueled plants in this country and abroad including the - g 15 following PWR plants: Trino Vercellese (ENEL, Italy) San Onofre 16 (SCE), Malibu (LADWP), H. B. Robinson (CP&L), Point Beach (Wis-17 consin-Michigan Power Cocyany), Surry (VEPCo), Three Mile Island { 18 (Metropolitan Edison), Prairie Island (NSP) Burlington and Salam 19 (FSE5G), Zion (Commonwealth Edison), Kewaunee (WPSCo), Calvert 20 Cliffs (BG&E), Diablo Canyon (PG&E), and Beaver Valley (Duquesne 21 Light Company). i 22 2. I was graduated from the New York University in 1948 with the .q 23 degree of Bachelor of Science in Civil Engineering. In 1950, I ,} 24 received a Master of Science degree in Sanitary Engineering, in 25 1958 a. vaster of Science degree in Nuclear Engineering and in 1960
- 9 26 a Docter of Science degree, all from the Massachusetts Institute
,[ 27 of Technology. 1 O W .a h.'1'MT A Abdi.2 ZQ h M L h ?
g MORTON 1. GOLDMAN K cl 1 3. From 1948 to 1949 I was a Research and Teaching Assistant at 2 the Sanitary Engineering Research Laboratory, New York Univer-1 g 3 sity condudting research on water coagulation and assisting in C,.[) 4 teaching sanitary chemistry and sanitary biology laboratory -32 J.J 5 courses. ?xi$ 6 4 From 1949 to 1950 I was
- c. Research Assistant at the Radioactivity
.4 k. 7 Research Laboratory, Sanitary Engineering Department at MIT con-j 8 ducting original research on removal of radienuclides from water y cf 9 by standard water treatment techniques. 2 a ['" 10 5. From 1950 to 1961 I was a Conrnissioned Officer with the United i[d) [ 11 States Public Health Service, Divisioa of Radiological Health. 10 12 I was first assigned to the Radiological Health Training Section j y g y 13 fronn 1950 to 1954 as the engineer staff member lecturing on appro- }ig 14 priate aspects of radiological safety and waste disposal, n , j 15 6. From 1954 to 1956 I was on loan to the Oak Ridge Natione; Labor- .x 16 atory as Chief of Soils and Engineering Section, Waste Disposal 17 Research Activities. In this position I conducted and supervised -M 18 research r,n disposal of radioactive wastes at Oak Ridge National ..9 19 Laboratory. j - :;j 20 7. From 1956 to 1959 I was assigned to MIT as Project Leader for the 1./ 21 Radioactive Waste Disposal Project of the Sanitary Engineering .f h 22 Department and in traini.g in the Nuclear Engineering Department. ,q s? 23 In the former capacity I initiated and supervised research on i 24 novel methocs of disposal of high activity fission product waste l 4+,; r.> l ? ti <S rr h n. T< T+ y ' m' Q3'~ ' ',; p s' ~;_q:gQ(.R Q.' gj Q4 'T7;^M[j Q_Qf VG ' 4%* %MYM' dei y 3
~..e MORTON I. GOLDMAN ~ 3r 1 materials. In addition, I served on the MIT Raactor Safeguards 2 Comittee as its secre ary. g_ g 3 8. From 1959 to 1961 I was designated as Nuclear Installation 73 !?X q.h 4 Consultant with the Division of Radiological Health in Washington, sT , Q-5 D. C. In this capacity I provided technical consultation and [$$f 6 assistance to State Health Agencies and other Federal Agencies on .? 7 .f4 health and safety problems associated with nuclear installations. 8 As part of my responsibility I served as the evaluator responsible [6 9 for the following nuclear plants: Yankee, Elk River, Indian Point, 15p; S,$, 10 Carolina-Virginia, Hallas, Pathfinder, Peachbottom and Humboldt [ 11 Bay. wg N 12 9. Since 1961 I have been with NUS Corporation and active in all of ' l?; q M 13 the environmental safety activities described earlier. Nh q~.( 14
- 10. I a:a the author and co-auther of a nur:ber of papers on radiation
.'.%4-15 ar.d public health, nuclear safety and radioactive waste managentnt. Q .h_ 16
- 11. I a:n a enember.of the American Society of Civil Engineers (ASCE),
Mg 17 the Water Pollution Control Federation, the American Association k 18 for the Advancement of Science, the A,erican Nuclear Society and -:n ((h 19 the Air Pollution Control Association. I am also a Licensed Pro-10 20 fEssional E..gineer in the State of New York ano the District of i 6'sf 21 Coluscia and a Diplomate of the American Academy of Environmental M'$ $ 22. Engineering in Itadiation ilygiene and Hazard Control. I am also 9 23 a cc::ser of Comittee H18 "Huclear Design Criteria" of the USA
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... ~. --, MORTON 1. GOLDMAN 1 Standards Institute. I served as the U. S. representative w
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on an expert panel of waste management practice at nuclear Q,l g 3 power plants at the International Atomic Energy Agency in 7 /. 4 Vienna. .2.f x ~. I. .>e..s >,'~;, , c4 .7
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<,e / 1 EDUCATIONAL AND PROFESSIONAL QUALIFICATIONS 2 CARL EDWARD THOMAS 3 PROJECT MANAGER
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4 HUCLEAR POWER GENERATION DEPARTMENT 91 W 5 POWER GENERATION DIVISION u# 6 THE BABC0CK & WILC0X COMPANY 4n Ni? 7 1. My name is Carl E. Thomas. My residence is Route 1. Madison
- q
.g 8 Heights, Virginia. I am employed by The Babcock & Wilcox 4:r %y_ y 9 Company, Power Generation Division, in the Nuclear Power Gen-s d% 10 eration Department. [w. N& 11 2. I graduated from the University of Chattanooga in 1954 with a a r._ 'h[ 12 B.S. in Engineering Physics. In 1955. I graduated from the IM 19 13 Oak Ridge School of Reactor Technology. From 1957 to 1962, I n[gs / 14 participated part time in the University of Virginia Engineer-15 ing Graduate Study program. m; Ok 16 3. I served in the United States Army Air Corps from 1944 to 1945.
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$$[ 17 4. In 1955, I joined The Babcock & Wilcox Company. My early assign-M 18 ments were associated with the Aqueous Homogeneous Reactor, the %4W 19 ' Moderator Control Reactor and the liquid metal fuel ractor g[ 20 concepts as a reactor physicist. 21 5. In 1960, I became Chief of the Operational Analysis Section with g 22 responsibility for reactor and system dynamic analysis, reactor tc j 23 control analysis, plant performance, and safety analysis. In T19 24 1964, I was appointed Director of the Savannah Technical Staff h 25 and later was designated Manager of Technical Services, Savannah w 26 Technical Staff. My responsibilities in these capacities by ,im a. y:g[ C-9 91 i i r --
e. o CARL EDWARD THOMAS 1 included engineering, shipyard modifications to the S.vannah 2 reactor plant, and safety and licensing activities. In addition.
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3 I served as nuclear advisor aboard the N. S. Savannah during $'5 ^ port visitation voyages to American and European ports. jQ1 4 .ag L.c Q"'y 5 6. In 1965. I'was appointed Assistant Manager of the Reactor 7 ~ (+$? 6 Engineering, Department of the Atomic Energy Division. In this i S 7 capacity, I was responsible for the engtuering of marine reactor
- R f@ff 8
plants. Q 1:pg ]:,3 9 7. In 1966, I joined the Contract Depactment and am now serving as .,p. 17-10 Project Manage or the B&W contract with Florida Power Corpora- .<, if $7;- 11 tion for Crystal River Unit No. 3.
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EDUCATIONAL AND PROFESSIONAL QUALIFICATIONS 2 ROBERT E. WASCHER 3 MANAGER, NUCLEAR SAFETY ENGINEERING SECTION 4 NUCLEAR POWER GENERATION DEPARTMENT h 5 POWER GENERATION DIVISION 6 THE BABC0CK & WILC0X COMPANY 1. ,c ., i.... . 7 1. My name is Robert E. Wascher. My residence is 1916 Eastwood 8 Lane, Lynchburg, Virginia, 24503. I am employed by The Babcock 9 & Wilcox Company, Power Generation Division, in the Nuclear 10 Power Generation Department. Il P. I graduated from the Illinois Institute of Technology in 1952 12 with a Bachelor of Science Degree in Mechanical Engineering. 1 13 In 1953 I graduated from the Oak Ridge School of Reactor Tech-14 nology. O 15 3. Upon graduation I joined the Oak Ridge National Laboratory as 16 an Associate Development Engineer responsible for the development 17 of mechanical components for homogeneous nuclear reactors. 18 4. In 1955 I was comissioned an officer in the U.S. Navy. During 19 my naval service, I served as the Navy Liaison Officer in the 20 Army Package Power Reactor Program. I was also assigned to the j 21 Navy's Bureau of Yards and Docks with responsibility for nuclear 22 engineering problems of the Bureau. 23 5. In 1958 I joined The Babcock & Wilcox Company as a Nuclear O 24 Engineer with responsibility for the safety analysis of the 25 Consolidated Edison Company's Indian Point No. 1 Nuclear Plant. O C-11 .z. + - - + - - I i
ROBERT E. WASCHER 1 In 1959 I was appointed Supervisor of the Safety Analysis Group 2 with responsibility for safety analysis of nuclear riants designed 3 by B&W. In 1964,I became Chief of the Operational Analysis 4 Section with responsibility for reactor and system dynamic 5' 7 analysis, reactor control analysis, plant performance, and i ~ 6 safety analysis. 7 6. In 1965 I was appointed Manager of the Nuclear Safety Section. 8 my present position. In this position I am responsible for \\ 9 safety and licensing of the plants designed by B&W. 10 7. During 1964 and 1965 I was Chairman of the N.S. Savannah Safety 11 Comittee, a committee responsible for periodic review of the 12 operation of the N.S. Savannah. From 1962 to 1966 I was also 13 Chairman of B&W's Nuclear Development Center Safety Review 14 Board. In 1966. I was appointed to the Atomic Energy Comission's 15 Advisory Task Force on Power Reactor Emergency Cooling. In 16 addition, I am a member of the American Nuclear Society and the 17 Atomic Industrial Forum's Safety Comittee. 18 8. I am a registered Professional Engineer in the State of Virginia. O O C-12 ...-..,...p p > #.,r. y 9.. ;.. %. ., wwvm.. w.. y9 r.op y.._.
265 t t t CHAIRMAN JENSCH: Does that conclude the direct 3, presentation from this witness? i 31 MR. EVERTZ: Yes, Mr. Chairman. CHAI:IMAN JENSCll: What is the suggestion of the par-4 ticipants as to cross-examination? Do you desire to await the I S I I l 4{ formal' tender of-the other-witnesses as outlined in Appendix C to Applicant's Exhibit No. 17 y f MR. EVERTZ: Yes. g Mr. Chairman, at this time I an prepared to proceed 8 i by calling the applicant's expert panel witnesses to conclude g i I
- l the sponsorship of Appendix C.
g CHAIRMAN JENSCll: Would you do that, please. g I MR. EVERTZ: At this time the applicant, Florida y h Power Corporation, would call its panel of expert witnesses i I to assist Mr. Rodgers in responding to questions directed to l g him on cross-examination and also, as stated, to complete the j g l sponsorship of Applicant's Exhibit No.1. The panel of expert g I witnesses consists of the following individuals: l I Mr. Donal1 J.Rowland, Florida Power Corporation, 19 Senior Power Engineer, -- CHAIRMAN JENSCH: As each of the names is called, l 21 l will the gentlemen come forward, please, and sit near l l t Mr. Rodgers? 1A MR. EVERTI: Mr. E. Robert Hottenstein, Gilbert 24 j Associates, Inc., Project Manager; O
266 l 1 Dr. Morton I. Goldman, NUS Corporation, Vice Presi-2 i dent; 8 Mr. Carl E. Thomas, the Babcock-Wilcox Company h 4 Project Manager; e And Mr. Robert E. Wascher, Babcock and Wilcox Company I s Manager, Nuclear Safety. I 7 CilAIRMAN JENSCH: Do you desire each of those 6 gentlemen to be sworn? 9 MR. EVERTZ: Ye.3, Mr. Chairman, these witnesses to ' should be sworn. i 18 n Whereupon, DONALD J. ROWLAND, 12 E. ROBERT HOTTENSTEIN, is, I MORTON I. GOLDMAN, to CARL E. T110 MAS, and es{ ROBERT E. WASCilER 16 17! vere called as witnesses and, having been first duly sworn, l I ' were exanined and testified as follows: g. i FURTHER DIRECT EXAMINATION 3, XZXZX 3{ BY MR. EVERTZ: l Q Gentlemen,.I show you the document attached to g I Applicant's Summary Description of Application as Appendix C, yj O entitled " Professional Qualifications of Expert Panel Wit-3 nesses," and I ask you if you, respectively, prepared that u o n y' portion of this document which contains your own professional i Is I 1: h w
f eb8 267 l t qualifications? I 1 A (Chorus of "Yes".) g 3 MR. EVERTZ: - Mr. Chairman, may the record show that 4 each witness answered in the affirmative? g i 5l CilAIRMAN JENSCll: The record will so show. i 6 _a MR.'EVERTZi' This ' question is directed to the en-9 i y tire group'of expert panel witnesses. BY MR. EVERTZ: Q A[etheanswers contained in this document entitled "Sim-'ary Description of Application" as Appendix C, i g I entitled "Pbfessional Qualifications of Expert Panel Wit-f g -n nesses," as to your personal qualifications true? u ? A (Chorus of "Yes".) u O .MR. EVERTZ: May the record show that each witne.s g gl answered in the affirmative. Cl! AIRMAN JENSCil: The record will so show. ( g 1 BY MR. EVERTZ: l p Q I now ask each of you whether or not you would he.c I 18 1 1a ; and now give fron the witness stand the sane statenents pe - taining to your respective professional qualifications as are 30 contained in this document attac:ted to applicant'r. Summaty Description of Application as Appendix C, entitind "Prefes-O siona1 qualifications of Expert ra=e1 Witnesses" ana,hich you have just.estified you have prepared? O A (Chorus of "res.3 0 w-
M 1 ' !., ED (9 { 268 . -7 !l MR. EVERTZ: Mr. Chairman, may the record show that ij f each witness answered in the affirmative? A CllAIRMAN JENSCil: The record will so show, h 4 ~ !!R. EVERTZ: Mr. Chairman, this document attached E to applicant's Summary Description of Application as Appendix r, : a o C entitled " Professional Qualifications of Expert Panel Wit-g l hnesses"hasbeenidentifiedandsponsoredbythegroupofex-1 4 8-pert panel witnesses. At this tine, the applicant will offer t[ into evidence the document attached to the applicant's Sun-tary Description of Application as Appendix C and entitled % e 1 it j " Professional Qualifications of Expert Panel Witnesses," and p
- g ;f ask that it be incorporated into the record of the proceeding
- s i as if read, pursuant to the provisions of Rule 2.743, Part 2
- .2 of the Commission's rules and regulations.
ss, And I request that this document be permitted to re , constitute the direct testimony of this group of expert panel 17 witnesses.. Is there any obj ection by the stafb Cl! AIRMAN JENSCII: a a MR. HADLOCK: No objection. g.I CliAIRMAN JENSCil: State of Florida? g ) MR. TURNBULL: No objection. 3, L C18 AIRMAN JENSCll: The City of Gainesville? .a O MR. FAIRMAN: No objection. Cl!AIRTIAN JENSCll: The request is granted, and Appen-a , h n ;. dix C may be considered as part of Applicant's Exhibit No.1. O e 9 e i h ya
\\ ~~-- I 4 1 MR. EVERTZ: Mr. Chairman, there are no further 'j questions of this group of expert witnesses at this time. May E I suggest that they remain seated and be available for ques- ^- 8 4 tioning at the appropriate time? g E' In addition to Mr. J. T. Rodgers and the panel of i 1 l expert witnesses,' we also have available a group of expert i s 7, bsckup witnesses. Mr. Rodgers will, if he deems it necessary ei refer to one of them for answering any question by the Board 4 1 or any other interested party which pertains to his technical j 9 to specialty. CifAIRMAN JENSCH: The request is granted. 1 What is the suggestion of the parties in reference u. is to cross-exaninction? Do you desire that the evidence of the O t staff be presented so that if the interrogation relates to a 14 i f-l 15 Particular subject on which the staff nay make a contribution that they would likewise be available at the time the subject to n; is bieng considered? What is the wish of the parties 7 MR. DUNN: Yes," sir, I would propose that the staff g ? put its 2 cet testimony on at this time. g gl CHAIRMAN JENSCII: IWs the staff have any objection? I MR. HADLOCK: I think that is a desirable way to y proceed, Mr. Chairman, a h CHAIRMAN JENSCll: Will you call your witnesses then, g J Staff Counsol, pleasef gg h MR. HADLOCK: Mr. Chairman, I would like to call 3 O l l L. I 1 4 i l
g t l 270 j A 3l Charles G. Iong, Denwood F..Lss, Jr., Patrick W. Howe, q g Gordon Burley, and Mr. John P. Emptist, and ask that they be 8 v 8 sworn at this time, Mr. Chairman. t-g 4l Whereupon, I 8 CHARLES G. LONG,
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s DEhYOOD F. ROSS, JR., 7 PATRICK W. HOWE, a GORDON BURLEY, and s l' JCHN P. BAPTIST
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were called as witnesses and having been first duly sworn, 3 H ',' were examined and testified as follows: l l. 12 ' CHAIRMAN JENSCH: Proceed, Staff Counsel. XEXEI 13! DIRECT EXAMINATION k' - g 14 MR. HADLOCK: Mr. Long, will you state your positit,n 'n is with the Atomic Energy Commission? MR. LONG: I an a Branch Chief, in charge of ore q Sa s* n' of the five Reactor Project Branches within the Division of Reactor Licensing.- to [ MR. HADLOCK: Did you prepare a statement cf your g, professional qualifiestions for use in this proceeding? g6 i MR. LONG: I did. at MR. HADLOCK: Is that statement true and correct to g-h the best of your knowledge and belief? a g4,l. MR. LONG: It is. MR. HADLOCK: Mr. Ross, will you indicate your D 0 h h n... - nu,-... m,n_ p -,,.,,,_.. nn.
pq ; eas e, l 271 -l 1.i .1 7 s position with the Atomic Energy Commission, please7 3 d g MR. ROSS: I am a Project Leader in the Reactor 8 Project Branch No. 3, Division of Reactor Licensing. I work 8 4 with Mr. Long. 4( g a .) A HR. HADLCCK: Did you also prepare a 31.atement of r your professional qualifications for this proceeding? a 4 7' MR. ROSS: I did. J al MR. HADLOCK: Is that statement true and correct to Q j 9 the best of your knowledge and belieff to MR. ROSS: It is. 7 C 11 MR. HADLOCK: Mr. Howe, will you state your name and ui state for the record your position with the Atomic Energy i. 13 Commission, please? O i 14 MR. HOWE: Patrick W. Howe, Assistant Branch Chief, Environmental Radiation Safety Technology, Reactor Technology. 16 l ta MR. HADLOCK: Did you also prepare a state:nent of q i i 17, your professional qualifications for this proceeding? r MR. HOWE: I did. i la j MR. HADLOCK: Is that statenent true and correct to g jo ' the best of your knowledge and belief? ~ MR. HOWE: It is. 21 Q MR. HADLOCK: Dr. Burley, would you state your g ~1 ~ position with the Connission, please? n DR. BURLEY: Reactor Engineer, Containmont Component 34, 3 Technology Sranch. i i i n
r g 272 fy t3 i r.tM i I MR. HADLOCK: Did you prepare a statement of your ] [hj h I professional qualifications? a DR. BURLEY: Yes, I did. %y $ 4 MR. IIADLOCK: Is that statement correct to the best 3 ,f a of your knowledge and belieff w.l 5 DR. BURLEY: Yes, it is. .M 7[ MR. HADLOCK: Mr. Baptist, will you tell us by whom
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you are employed and your position with that employer, pleasef a 4 ia 5 MR. BAPTIST: I am Project Leader of the Radio-f. biological Laboratory, Bureau of Commercial Fisheries, It /g if Beaufort, North Carolica. 3 12 CHAIRMAN JENSCH: Is that a Governmental unit? ~:jf MR. BAPTIST: It is under the Department of Inte..or,I Id la 4 0 [% CHAIRMAN JENSCH: Pro ceed'. l 14 sir. !. Q 15 m i A -C to l MR. HADLOCK: Did you prepare a statement of profes-m g y
- , n 17 i-sional qualifications, Mr. Baptist?
.I w% MR. BAPTIST: I did. '~} ts A MR. HADLOCK: Is it true and correct to the best 4 9l s l of your knowledge and belief? i . /j 2c t a MR. BAPTIST: Yes, it is. 21 7 g[ MR. IIADLOCK: Mr. Chairman, I nove at this time that j ,9 l h Q l y nj these statements of professional qualifications of Mr. Long, { ? i g h, Mr. Ross, Mr. Ilowe, Mr. Burley, and Mr. Baptist be incorporated l [l g w in the record at this point as if read. 'O 3 , :p t EN i: l s.. s. _ "L sM ' ' ?.tGEMEEMEMEM!EMMMfWEFJEWXMLT.TEP22Jm:M;'qshmm h
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-:'? 1 2iy CHAIRMAN JENSCH: Any objection from the applicant? h MR. EVERTZ: No objection. 29,l d1 P CHAIRMAN JENSCH: The State of Florida? 9llb. 1 S h
- l MR. TURNBULL:
No. M . -e _ r1 CHAIRMAN JENSCH: The City of Gainesville? ~# }3 l M 8I -MR. FAIRMAN: None. { 4, 7, CHAIRMAN JENSCH: The request is granted. The state-g ,;k, ment of professional qualifications of each of the witness to T #g\\ 4 which statements Staff Counsel referred may be incorporated 8 l within the tr.anscript as if read at-this place. 8' 'Ef JP ~8 (The professional qualifications of Messrs. Long, 2d l 12 Ross, Howe, Burley, and Baptist follow:) q A j$, e x. i lh 14 \\ s}i .. :- )
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PROESSIO'JAL C:7ALI?ICATIONS 'd6 f. ,..' P ; ss.. [j DIVISION 07 REACT 02 LTcs.ul.w 7, .,a ' p.f g I w Ch..lt) I as tha chief of osa of five Reactor Project 3 ranches in the DI. vision p _s :s ggj of 2cactor Licensing. In this pcsition, I am responsible for tha a= *ysis
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',3 a:d aveluation of tha nuclear safety aspects of melaar reactor facilities 'p V.4 tdJ ll assig=ad to the b:csch. Il l TH!j . ~~ l M; I attended A111c=ca' Collega in Cs= bridge Springs 2eansylvsnia from 1952 TV j 3 to 1954, cs a p:ce:g%^ ing student. I attended the University of Pittaturgh, 1 M ' Q Pit:sburgh, Pe=sylvasi:sfres 1954 to 1956, c d received a 33' degree in mechan-- ~ ~'# icci enginecting. I have completed severe.1 g rdcata courses in enginaaring
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10 %F %q In Jc:a 1956, I took a positics with.the Curtiss-Wright Corporc:ica in .~ ~ M1
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M fjij of vcricus receto: cc=pc= cats fc the ? ojects Pluto and Zeb c. As s. Iced it '4yM design o ginece, I ves respo:sible for design of severci "research recc ors
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As. a M project' caginec, I ucc =cspo:sihic for the desis=, fchricction a:d cor.struc-3 v,a G tics of *:ia Thcilcad Ecsec ch Recctor. .ry 1: Sepec=ber 1950, I resi;.cd =y pccitica c: C==:isc-U:ight Cc:pc:ctic=. .f+ I thcs cecep cd :.r. cppoi r..c : cs c recc:c engi=cer with the Division of -/ s Q . Recccc Licc=ci=3, US Atcmic '- q Cc=ission. In this position, I pc::ic=. v O ipcted in the safety rr. views of the %3 Sr.vc=:r.h; the Eu=boldt Ecy necy or c: w su 9 3... O. L .y N 3
I g) .k,%[i~j$y,i(g.3.ge,dkde fg g-fy} ,. o s,. (,.., ' -.m y r SW ^ b-5% h .m 2-1 N. a y# L Cureks. California; the Big Rock Point Reactor at Big Rock Point, F.ichigan; ~ the Elk River Reactor at Elk River, Minnc2:ta; and several resharch and tast' [ In April 1962, I left the coacission to 2ccept p position as resceors. ,g1
- i. was process engineer with the Internuclear Coc:peny in Clayt.., Micscuri.
g .a c:ponsibic fc the rccctor design cad p:ccces systems for the University l i [y of Missouri 2s:ccrch Resctor, Colu= Sic, %issouri. w t
- '[s In November 1962, I ratu.ne'd to the Division of Racetor Licensing as yj
, yg.; a recctor cs;incer. I h:va had the pri=.y responsibility for safapy review. l w of the Consolidsted Edisca Indica Point Unit No.1 at Indica Point, New Yory; I g l r4r j ths LcC:csse Eoiling W:tc; Reactor at Canoa Station, Monroe County, Wicconsin; y.-*. the Florida ?cuar and Light Cenpany Turkey Point kinics 3 and 4 ct Tc:2ey paint, I xv 9g 0:do County, Florida; tha HI Resec r,h Recctor'et C:.= bridge, Eassachusetts; t. g ua Geor;ia Tech Ocschech Reactor at W. sta, Georgia; End the Western new ~ 95 ^ l ~ w l 'y Yc k Kuclec Research Center Reactc; at Euffalo, Ncv York. /.I $b h.,IZf 6 r ^fy f
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.xJQ PATRICK U. HOWE [%y ) IFL PROFESSTIONAL QUALIFICATIONS l nf[whh ENVIRONMENTAL AND RADIATION SAFETY TECHNOLOGY BRANCH
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e;q N DIVISION OF REACTOR LICENSING l pp; I,Vf4 n Tj? Qc W My name is Patrick W. Howe. I sa Assistant Branch Chief. Environmental and (ONk E3 Radiation Safety Technology Branch of the Division of Reactor Licensing. I am ich S((h'k responsible for the supervision of the Braach's participation in evaluating proposed reactor sites and radiological consequences of postulated accidents. /D q3 q~ %,4 c. I was born in Wilmington, North Carolina. I graduated from The Citadel, ~., c% p,:$ Charleston, South Carolina, with a B.S. in Chemistry and have a certificate from ny g UCLA in Engineering Managementc am 39 Prom 1951 to 1956 I was employed at the Savannah River Plant by E. I. duPont
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];, de Nemours & Company, Inc. During this employment I co ducted the pre-operational %) environmental program, served as " construction liaison" on the construction of M?$ the Health Physics Laboratory Building and the 772-F Works Technical w;.u ~.m ?.n y!. Laborntory Building and was appointed as Laboratory Supervisor for the Plutonium 4.g's Quality Control Laboratories. %Q. Eh Upon leaving the Savannah River Plant I joined the Martin Comoany as Senior - ;@2 > y.s "yf Nuclear Engineer on Project Pied Piper - the predecessor to SRAP. With the i,TJ ' V;) completion of Project Pied Piper I accepted the position of Superintendent, Hesith N '.g O ** ""' ****'**' ""*'" ""' " *"*"' ' *** "" ""*"~" c"' ' <.u "A Company's Nuclear Puel Manufacturing Division. hi wMhh n~ P f-. qq:.y ,c hi - ,k,- ' MM.N. "J" Md" 1ER'ad"LMAA P.h 6*dWMM d mN-
.s ,./,4 g 4 y@ r iv 0;>- In late 1957 I joined the staff of the Health Cheetatry Department. E:p g Lawrence Radiation Laboratory, Berkeley, California, and in 1959 uns appninted hg jfL' Department Head. In this function I was responsible for the radiological @f% safety, industrial safety and support engineering requirements of a broad MA% m variety of LRL research programa. During the period of 1962 to 1966 I served 'I' t as a member of the Atomic Safety and Licensing Board. }; !- $w In June 1966 I resigned from the Lawrence Radiation Laboratory and joined [N William H. Brobeck and Associates as Director of Development. The firm of p WMB&A is principally engaged in the design of nuclear particle accelerations and g.. I (; p sechnical engineering research. I resigned this position to accept my present gg function as of September 1, 1967. g . v.%r1 I have authored er co-authored over fifteen publications in the field Qiyj Q of nuclear safety and engineering. syg g py In addition to my normal employment I have served as a consultant to over J W; ,$h twenty industrial, institutional and governmental agencies in the field of ik Ajih nuclear safety and engineering.
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l J~..,. I an a amenher of the Containsment & Component Technology Branch and . j4 l perform technical analyses, reviews, and evaluations in the area of 3 i At fission product release, propagation and removal as related to nuclear ,w power reactor safety. h% 34 Before joining the Atomic Energy Cosmaission in April 1967. I was
- g associated for 14 years with the National Bureau of Standarda and worked iy; on problems in the field of solid state chemistry. Prior to that. I vaa r: D-p._
.- s G d research associate With the Geophysical Laboratory of the Carnegie .Q Q9 w. Institution of Washington and worked in the area of chemical thermodynamics. hhd ! WQ I havc been principal contributor of more than thirty scientific research -:l papers or reports.
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-yl;9 My' formal education is as follows. ?%% @h A.B. Chemistry Tenple University 1948 .ym M.S. Physical Univ. of Maryland 1950 fh Chenistry sg' Ph.D. Physical Georgetown Univ. 1962 Se Chemistry
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) r i e, ' .m. ?. 7..?TIST ? 10?ESSIC':.*.. C'JA'.IFICATIC':S y. x>y- ?;, ?csition: Project Lecder, Vertebrate ?,roject, Pollution Studies Program, e) _ Rcciobiological Laboratory, Eureau of Co:snercial Fisheries, a 2caufort, North Carolins 25516 'N
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Education: T y A.S. Degree, Biology,1942; Maryville College, Maryville, Tenn. k. s
- b M.S. Degree, Siology,1947;~3acknell University, Lewisburg, Pa.
?' Ley [h Nuclear Methods Applied to Cceanography, two weeks,1961. Oak Ridge Institute of Nuclear Studies, Oak Ridge, Tenn. o Mobile Radioisotope Labcratory Course, two weeks,1962. ?ly Ock Ridge Institute of Nuc1 car Studies, Oak Ridge, Tenn. D 3:swf l Ret.ctor Safety and Hesseds Evcluation, 68 hours,1963. j U.S. Public Eealth Service, Rchert A. Taf t Sanitary Engineering ,gijg Center, Cincinnati, Ohio. . ?@kJ ,4 N Mans.gement for Supervi.sces, o..e week,1966. Bureau of ? ( l Co=cercial Fisheries Esployee Developccat Program. j' p-l Departmental Manager Levelopacnt Program, 5 months, 1967. M l Department of the Interior, 'Jashington, D. C. ? I a Experience: Ibh
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~,. Thirteen years of rcdiobicle;ical research on the uptake of @N; rcdio nuclides by marine organist.s, and the transfer of radionuclides through food cheins. Lfic.
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.u Six years of research on tne sof t-shell clam in Massachusetts. $h?e +%Vt -&t. ^ One year of research on :he A=erican oyster in Cheaapeake Bay. W.6:9 .qt Advisory or Consultant Activities: Q-d t $g During the past four year 6, h:ve assisted the Laboratory A Director in evaluating che possible effects of nuclear reactor 35 operation on fishery resour:es. .j Marine Biolo3y consults.nt in a cooperative radiological survey f3 of the Savannah River Ltuary with the U.S. Public Health, fg Service Division of Rcdiolo;ical Ecalth,1960. 34 ?-2? -e' -d M i 9
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I s 4 Scica;.ific Ta'.ks beforc Professiensi Societies: h ~ rt ,:g) t M,. Technic:1 p pers on cri;iac1 research were presented before l ] %17
- he Auerican Fisheries So:iety, Association of Southeastern
((' 'I Liolo;ists, and atlancic 3:carine Research Society. j %g 8 l A= ,J Mer.bership in Scica:ific Societics:
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F.calth Physics Society bnding). M ~ [N'h i. ..clantic Fishe-ics Bio'.adists-Secretary-Trt,gsurer, 1952. kh [ Atlantic ::stusrine Research Society. Secretiry-Treasurer, 1960. l A=crican Fisheries Socia:y. iQ.i; [.g g i ${ [ Publications: 9 w$,h 1 } Tuelve scientific publica: ions on upt ke sad retention of ';;g re.dionuclides by =srinc or2 nis=s; transfer of radionuclides rfpA through food chcins; :::.!:-shell cic=s; and horseshoe crabs. g' These papers have appe re-n U.S. Fish and Wildlife Fishery W Sulle: Ins, Transactica. a. :he A=arican Fisheries Society, Of fi U.S. Pahlic Mcsith Scr ti e Reports, Sacond National Sy=posium on Radioccology (in pr.:ss), and the Second International {lhgg y Conference on Wa:cr Pc'.la: ion Rasearch. ids we1 I t' e ?? l %.y8 u 'iM7 th e e riM l M 8 mg 47 [d ?, ml I g);f m )
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I I cbl5 274 I I CHAIRMAN JENSCH: Proceed. ~ i 1 if b MR. HADLOCK: Mr. Long, did you supervise and parti- - 3 {j cipate in the preparation of a document entitled " Safety j 4 g Evaluation" by the Division of Reactor Licensing, United States Atomic Energy Commission in the matter of Florida Power Corpora}- 5 l tien, Crystal River Unit 3 Nuc1 car Generating Plant, Docket C 1 i ? 7 i No. 50-302, and dated June 6th, 19687 I s.l MR. LONG: I did. 6 MR. liADLOCK: Did you also participate in the prepary e 10 tion of that dc.cument, Mr. Ross? 1 MR. ROSS: Yes, I did, j j j
- i ij HR. IIADLOCK:
F8r. Hese? 'I l ts @ MR. HOWE: Yes, I did. i
- u.,j MR. HAULOCK:
Mr. Burley? j ( i I tr [l DP. BURJ.EY: Yes, I did. f l!l 1.iR. 1:ADLOCE: I ask each of you, is that doctcr.at it i 7 true and correct to the best of your knowledge r.nd bel'.af? l 5 st, (Chorus of "yes".) [ J 4 MR. HADLOCK: i.re there any corrections to that i i e t' s, document, Mr. Long? I .g: MR. LONG: Yos, there is one correction. On page u 6, the third line from the botton of the page -- l Excusece,heforeproceeding,does! O C>ii1RMAN JENsCn: i' m t 24 the staff happen to have an extra copy? I do not seen to havej h
- c '
cy copy available. O I L .}}