ML19319D555

Summary Description of Application for Licenses

ML19319D555
Person / Time
Site:	Crystal River
Issue date:	07/16/1968
From:	US ATOMIC ENERGY COMMISSION (AEC)
To:
References
NUDOCS 8003170677
Download: ML19319D555 (58)

Text

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h UNITED STATES OF AMERICA ATOMIC ENERGY COMMISSION In the Matter of FLORIDA POWER CORPORATION (Crystal River Unit 3 Docket No. 50-302 Nuclear Generating Plant)

NegDInrurf ggy

SUMMARY

DESCRIPTION OF APPLICATION FOR LICENSES July '16, 1968 -

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i TABl E_OF CollTEtlTS Page 1.

IflTRODUCTI0ft 1-4 2.

DESCRIPTION OF SITE Atl0 ENVIR0flMErlTAL CHARACTERISTICS WHICH INFLUEtlCE DESIGft 2.1 Location 4

2.2 - Population 4

2.3 Meteorology 5

2.4 Hydmlogy 5-6 2.5 Groundwater 6

2.6 Geology 6-7 2.7. Seismology 7

2.8 Environmental Radiation flonitoring 7-d 1

3.

DESCRIPTION OF CRYSTAL RIVER PLNIT UNIT 3 4

3.1 Introduction 8-9 3.2 Reactor and Primary Coolant System 9-11 3.3 Reactor Building 12-13 3.4 Engineered Safeguards 13-14 4

3.5 Instrunentation and Control 14 -15 3.6 Electrical Systems 15-17

3. 7 ~ Auxiliary.Sys tems 17 1

'3.8 - Steam and Power Convers ton Sys tem 18 3.9 ' Radicactivi ty Contml Sys tems 18-19 4.

' SAFETY NIALYSES 19-20

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TABLE OF CONTENTS (cont'd)

Page 5.

TESTS, INSPECTIONS, AND QUALITY CONTROL 20-22 6.

RESEARCH AND DEVELOPMENT PROGRAMS 22-25 4

7.

. TECHNICAL QUALIFICATIONS 7.1 Florida Power Corporation 26-27

7. 2 Babcock and Wilcox Company 27-28 7.3 Gilbert Associates, Inc.

28-29 7.4 J. A. Jones Construction Company 29 L

8.

COMMON DEFENSE AND SECURITY 29-30 1

4 9.

CONCLUSION 30-31 i

APPENDICES i

APPENDIX A - List of References APPENDIX B - Figures APPENDIX C - Professional Qualifications of 4

Expert Panel Witnesses t

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IflTRODUCTI0tl 2

The purpose of 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 summary includes Amendments 1 through 5 7

and Supplements 1 tarough 3 to the original application which was 8

filed on August 10, 1967 and assigned Docket No. 50-302.

9 The Summary Description will provide infonnation on the environ-10 mental aspects et the site, description of the Crystal River Plant 11 Unit 3 design, conclusions from the safety analyses perfonned, descrip-12 tion of the Quality Assurance Program and its audit of the construction 13 quality controls, areas of continuing research and development efforts, 14 technical qualification of the Applicant and its contractors and

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15 conclusions of the Applicant in matters of common defense and security 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 Summary Description contains an Appendix A -

19 List of References as the text contains numerous 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 minimum of figures 23 to orient and amplify the word descriptions in the summary.

24 This Sunmary Description will constitute a portion of the pre-25 pared testimony 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 f tr. J. T. Rodgers, fluclear Project 29 flanager and. Director - Power Engineering 8. Construction.

1 To assist Mr. Rodgers in answering questions on cross-2 examination by the Board or another party, the following technical 3

witnesses repmsenting the Applicant, its engineers and i_ts con-4 tractors will make up a panel of technical expert witnesses whose 5

unprepared testimony will become a part of the Applicant's testimony 6

before the Board.

7 These persons are:

8 Name Organization Ti tle 9

Donald J. Rowland Florida Power Corporation Mechanical Engineer 10 E. Robert Hottens tein Gilbert Associates, Inc. Project Manager 11 Morton I. Goldman NUS Corporation Vice President 12 Carl E. Thomas The Babcock & Wilcox Co. Project Manager 13 Robert E. Wascher The Babcock & Wilcox Co. Man age r, Nuclear Safety 14 The educational and professional qualifications of the above 15 persons are included in Appendix C - Qualifications of Expert Panel

~16 Wi tnesses.

17 The Crystal River Plant Unit 3 nuclear generating unit will 18 employ a pressurized water nuclear steam supply system furnished 19 by The Babcock & Wilcox Company (also herein referred to as B&W) 20 and is similar in design to the nuclear steam supply systems which 21 are being fi:rnished by B&W to the Duke Power Company for its Oconee 22 Nuclear Station (AEC Docket Nos. 50-269, 270 and 287) and to the 23 Metropolitan Edison Company for its Three Mile Island Nuclear 24 Station (AEC Docket No. 50-289).

A construction p snnit authorizing 25 cons truction of the Oconee facilities was issued 1 i November,1967 26 and a construction pennit authorizing construction of the Three 27 Mile Island Nuclear Station was issued in May,1968. The nuclear i

28 s te am supply sys tem will operate initially at core power levels up

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I to 2452 MWt, which corresponds to a gross electrical output of about 2

855 MWe. An ultimate core output of.2544 MWt is expected, and all 3

steam and power conversion equipment is designed for a correspond-4 ing gross plant output of 885 MWe, Plant safety systems, including containment 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 amendments 11 and supplements thereto, has been reviewed by the staff of the Atomic 12 Energy Commission, which has prepared and published a safety evalua-13 tion of the Application. The Advisory Conmittee 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 Commission in a letter dated May 15, 1968. The ACRS concluded that -

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 similarly. The ACRS letter identified one item in the design for 21 which further development was requested. This item concerns off site 22 potter supply and is elaborated upon in this Summary in Section 3.6.

23 Other matters warranting consideration by the Applicant as mentioned 24 in the ACRS letter have been responded to in PSAR, Volume 4, Supple-25 ment 2, Question 5 (answer).

26 The principal architectural and engineering criteria which 27 govern the plant design are set forth in Section 1.4 of Volume 1 28 and in Supplement No. 2, Informal Question 4 of Volume 4 of the 29 Applicant's Preliminary Safety Analysis Report.

Design and 30 construction of the facility in accordance with these criteria 31 together with the redundant engineered safeguards systems provide 1

assurance that the proposed Crystal River Plant Unit 3 can and 2

will be constructed and operated at the proposed location 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 completion of Construction as December,1971 and the 6

latest completion of construction 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 1 and 2 already exist (see Appendix B, 12 Figure 1). This site is directly on the Gulf of Mexico, in the 13 Northwestern extrenes of Citrus County, Florida, approximately 7-1/2 14 miles NW of the town of Crystal River and 70 miles from Tampa, 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 mmoteness, with the Gulf of 19 Mexico on the West 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 8, 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 title of the site.(1) t 1

2.3 Meteorology 2

Meteorology in the mgion of the Crystal River P1 ant site has 3

been evaluated to provide a basis for preliminary detennination of 4

design criteria for stona protection, and a preliminary assessment 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. No tornadoes have been reported in Citrus County 8

for the period of record 1916-1966. Pmvailing wind directions are 9

seasonal, and are predominantly northeasterly during the autumn and 10 southwesterly during the spring.

Winds seldom persist from one 4

11 direction for longer than twenty-four hours.

12 In general, diffusion of waste gases in the atmosphere is good.

13 Wind direction is usually highly variable, and wind speeds are sel-14 dom extrecely low.(2) 15 2.4 Hydrology 16 The two streans in the vicinity of the site are the Withlacoochee 17 River and the Crystal 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 mouth of the Crystal River.

The Withlacoochee 20 is the major stream, having a drainage ama at its entrance into the 21 Gulf of Mexico of approximately 2,000 square miles. The discharge of 22 the Withlacoochee due to rain runoff is augmented by a base fic< of 4

23 groundwater runoff and artesian spring discharges.(3) The Crystal 24 River is much smaller than the Withlacoochee River, with its major 25 Cischarge 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 28 the Gulf of Mexico.

- 29 From the studied infonnation and data the hydrology of the 30 site is concluded to be most favorable..

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1 Because of its exposed location on the Gulf of Mexico, the facility is theoretically exposed to maximum intc: sity hurricanes.

2 3

The general line of approach for hurricane protection will include 4

full protection against hurricane winds, flood tides and wave action for components which must function for a safe and orderly 5

6 shut-down of the nuclear unit. This protection will be provided 7

for any intensity hurricane up to and including the Probable Maximum 8

Hurricane (PMi) based on parameters established by the applicable 9

portions of the revised Envimnmental 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.

The groundwater table occurs at a depth of approximately 8 to 15 9 feet below ground surface at a plant elevation of approximately 16 17 90 feet (mean low water = 88 feet).

Groundwater levels observed in drill holes were recorded to fluctuate, with little lag time, in 18 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 Geology 25-The site is located on gently southwesterly dipping biogenic 26 carbonate (limestone and dolomite) rocks which have been differen-27 tially dissolved along the most pervious zones of the mck, msult-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 1

of seismic profiles negate the possible existence of subsurface 2

faults at the site.

Regional tectonic elements are inactive and 3

present no threat to the structural integrity of local geology.

It is concluded that geologically the site rock mass is competent 4

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 some 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 12 Scale). The maximum ground motion at the site due to this earth-13 quake did not exceed 0.025g. The Crystal River Plant Unit 3 Class I 14 structures, components and systems will be designed to withstand an earthquake based on a maximum horizontal ground acceleration of 0.05g 15 16 and a maximum vertical component of 0.033g as an added margin of 17 safe ty. The ability of the piant to be shutdown safely will not be 18 impaired, however in the event of an earthquake based on a maximum 19 horizontal ground acceleration of 0.10g and a maximum vertical 20 component of 0.067.(7) 9 21 2.8 Environmental Radiation Monitoring A pre-operational environmental radioactivity monitoring program 22 23 will be conducted in order to determine the magnitude of the radio-24 activity in the envimnment surmunding the nuclear reactor site 25 and to study fluctuations in the radioactivity levels 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 environmental radioactivity levels that may possibly be attributed 29 to the Crystal River Plant Unit 3 operation. -

1 A comprehensive sampling program will be initiated at -least 2

. two years prior to Unit 3 startup. The collected samples will 3

consist of Gulf and well water, soil, air particulate, animal 4

' thyroids, fish, shellfish and bottom sediments. A post-operational 5

environmental program will be similar to the pre-operational program 6

with the sampling 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 re ommendations to assure protection and improvement of c

11 marine resources around the Crystal River Plant.

Florida Power 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 fomulating an acceptable pre-16 operational environmental radioactivity monitoring 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 set forth in the Application, 23 as amended.

The 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 mactor building, an auxiliary building, 27 a turbina building, a contml mom, a fuel storage building, a 28 service building, a substation, and various other auxiliary struc-29 tures and equipment. A plot plan of the Crystal River Plant Unit 3 indicating the general plant layout is shown in Figure 1, Appendix B.

1 2

A cutaway drawing of the reactor, containment, and structure arrange-3 nEnt is shown in Figure 3, Appendix 8.

Table 1-2 in the Application 4

sets forth a tabular comparison of the design parameters of the pro-5 posed Crystal River Plant Unit 3 with the Metropolitan Edison Com-6 pany's Three Mile Island tiuclear Station, Duke Power Company's 7

Oconee Units 1 and 2, and Florida Power and Light Company's Turkey 8

Point Units 3 and 4.

Following is a sunmary of those prinicpal 9

features of the plant which are considered significant to safety.

10 3.2 Reactor and Primary Coolant System 11 The reactor for the Crystal River Plant Unit 3 is of the 12 pressurized water type.

It has an initial rating of 2452 MWt, 13 corresponding to a gross electrical output of 855 fNe.(9) The 14 nominal operating pressure for the reactor is 2185 psig, with an 15 average temperature of 579 F.

The reactor coolant system is 16 designed for 2500 psig pressure and 650 F temperature.(10) 17.

The reactor core is approximately 129 inches 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 mds enclosed in 20 a square, stainless steel, perforated envelope. Each assembly con-21 sists of 208 zircaloy tubes containing uranium dioxide,16 control 22 rod guide tubes, and a center tube available for an in-core instru-23 nentation assembly.(12) There are approximately 201,520 pounds of 24 uranium dioxide in the core.(10) l 25 The thermal and hydraulic design limits of the core are conser-26 vative, and are consistent with those of other pressurized water 27 reactors currently in operation or under construction.(10) & (13) 28 Core reactivity is contmiled by a combination of 69 movable 29 control rod assemblies and a neutron absorber dissolved.in the I

coolant. The control rods are an alloy of silver-indium-cadmium-2 encapsulated in stainless steel. The dissolved neutron absorber 3

is boric acid.(14) 4 The control rods are used for short-tenn mactivity contml 5

associated with the changes in power level and also with changes 6

in fuel burnup between periodic adjustments of dissolved boron 7

concentration. (15) The reactor can be shut down by the movable 8-control rods from any power level at any time.(16) Each movable 9

control rod assembly contains 16 control pins and is actuated by 10 a sep6 rate control rod drive mechanism mounted on the top head of 11 the reactor vessel. Upon trip, the 69 control rod assemblies fall 12 into the core by gravity.(17) 13 Systems are provided so that the concentration of dissolved 14 neutron absorber in the reactor may be adjusted to maintain tne 15 reactor shutdown at room temperature and to provide a safe shutdown 16 margin during reN14".(18) The concentration of dissolved absorber 17 is reduced to e se for long-term reactivity changes, 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 4 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 28 examination, and dye penetrant testing.(20) During operation, speci-29 mens of reactor vessel materials will be placed in the mactor near 30 the inside surface of the reactor vessel.

These specimens are subject r

1 to irradiation similar to that to which the shell of the reactor 2

vessel is exposed. They can be removed periodically and tested to 3

ascertain the effects of radiation on the reactor vessel material.(21) 4 Two coolant loops are connected to the reactor vessel by nazzles 5

located near the top of the vessel.

Each loop contains one steam 6

generator, two motor-driven coolant punps and the interconnecting 7

pi ping. The reactor coolant piping is carbon steel clad on the inside 8

surface with austenitic stainless steel.(22) Reactor coolant is pumped 9

from the reactor through each steam generator and back to the 10 reactor inlet by two 88,000 gpm centrifugal pumps located at the 11 outlet of each steam generator.(23) 12-The steam generator is a vertical, straight-tube-and-shell 13 heat exchanger which produces superheated steam at constant pres-14 sure over the power range. Reactor coolant flovs downward through 15 the tubes, and steam is generated on the shell side.(24) 16 The reactor coolant pumps are vertical, single-speed, inaft-17-sealed units having bottom suction and horizontal discharge. Each 18 pump has a separate, single-speed, top-mounted motor, which is 19 connected to the pump by a shaft coupling.(23) 20 The pressurizer, a vertical surge tank approximately half-filled 21 with reactor coolant and half-filled with steam is connectad to the 22 reactor coolant system to control system pressure. The op: rating 23 pressure of the system is maintained by operating electric immersion 24 heaters to increase pressure or by spraying reactor coolant water 25 into the steam within this pressurizer tank, to reduce pr"ssure.

26 Self-actuated safety relief valves connected to the pressurizer 27 prevent overpressurization of the neactor coolant system.t25)

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1 3.3 Reactor Building 2

The reactor building is designed to enclose completely the 3

reactor coolant syst.n 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 BBRV systec consisting of horizontal and vertical 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 11 wall and dome and 1/4 inch thick bottom, to provide vapor tiohtness.

12 The concrete foundation mat will be approximately 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 of 20 feet 6 inches, a thickness of 3 feet and a height from spring 18 line to apex of approximately 32 feet 4 in.(27) 19 The building is designed to sustain safely internal 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 carrosion.(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 29 integrated leak rate at design pressure (55 psig) will not exceed 30 1/4.of one percent by weight of the contained volume of air in 24 31 hours3.587963e-4 days <br />0.00861 hours <br />5.125661e-5 weeks <br />1.17955e-5 months <br />. (30) 4 4

1 Prior to operation of the facility, the reactor building will 2

be subjected to a structural integrity test and leak rate test.

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 I'

8 are free fmm harmful corrosion and that excessive steel relaxation 9

has not taken place.(32) 10 3.4 Engineered Safeguards 11 Engineered safeguards are provided to fulfill the 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.

P 15 b.

Insure reactor building integrity and reduce the driving 16 force for building leakage.

1 17 c.

Remove fission products from the mactor building 18 atmos phere.

19 The engineered safeguards systems can be grouped into an 20 emergency core cooling system, mactor building cooling systems 21 and fission pmduct removal systems.(33)

I 22 The energency core cooling system contains both passive flood-23 ing and pumping equipment. The passive flooding equipment consists 24 of two pressurized core flooding tanks which automatically discharge 25 borated water into the mactor vessel in the event.that the reactor 26 system pressure' drops below 600 psi. The pumping' equipment consists s 7 n

1 of two completely independent sub-systems. Each sub-system contains 2

both a high pressum and a low pressum 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 mtt.rned to cool the core.(34) 9 The reactor building cooling system, which is made up of two 10 separate and independent heat removal systems, limits the pressure 11 in the reactor building following a loss-of-coolant accident. One 12 system contains three separate fan and cooler units. The other 13 system contains redundant spray headers which spray low temperature 14 borated water 1,nto the mactor building to cool the reactor building 15 atmosphere. Each of these systems independently has the heat mmoval 16 capability to maintain the reactor building pmssure below its design 17 pmssure level.(35) 18 Control of fission products following a loss-of-coolant acci-19 dent is provided by the reactor building itself and by a second 20 separate engineered safety featum for limiting release of fission 21 products from the reactor building. The second means for fission 22 product control 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 27 limit radiation doses at the exclusion radius and low population 28 zone boundary to values within the 10 CFR 100 guideline values.(36) 29 3.5 Instrumentation and Contml 30 A complete and dependable network of instrumentation and con-31 troh will be provided to insure the safe operation of the Crystal 1

River Nuclear Generating Unit. The reactor protective system moni-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 interrupting power to the control rod drive clutches and allowing 4

5 the control rods to drop into the reactor core.(38) Alarms (39) 6 are provided to alert the operator to abnomal operating conditions, 7

and interlocks (40) am provided to prevent abnormal operations 8

which could lead to potentially hazardous conditions.

9 The nuclear instrumentation system monitors reactor power from 10 startup level through 125 percent of full power operation. There 11 are separate overlapping instrumentation channels for the startup 12 power range, the intermediate approach to power range and the power 13 operation range.(41) A control system automatically monitors mactor 14 system conditions and the load requirements on the turbine-generator 15 unit, and adjusts reactor power, steam generator feedwater flow and 16 the turbine thruttle for safe, efficient operation.(42) 17 The engineered safeguards protective system monitors plant 18 conditions and automatically initiates operation of the engineered 19 safeguards systems, if required.(43) 20 Following proven power 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 infomation 24 display and alarm monitoring to insure safe and reliable operation 25 under normal and accident conditions.

26 3.6 Electrical Systems 27 The design of the electrical systems for trystal River Plant 28 Unit 3 is based on providing the required electrical equipment and 29 power sources to insure safe, reliable operation and safe, orderly I

shutdown of the unit under all nonnal and emergency conditions.(44) 2 The main unit itself is designed to withstand a full load dump with-3 out a trip-out, and as explained elsewhere, the mactor 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 redundancy 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 auxiliaries 12 required for full load on the turbine-gene.ator.

This 13 transformer will serve as the nonnal scurce of power for 14 safeguards equipment.

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 same site, can 17 supply power to this transfonner.

18 b.

In response to the ACRS concern expressed in the ACRS 19 letter dated May 15, 1968 to the Chai rman of the U. S.

20 Atomic Energy Commission regarding FPC compliance with 21 Criterion 39, the Unit No. I and 2 startup transformer 22 will be connected to supply a redundant feed to the 4160 23 Volt engineered safeguard busses.

24 c.

Upon loss of all sources of poaer described in a & b above, 25 power will be supplied from two quick-starting diesel-26 generator units connected to safeguards busses. The 27 diesels am sized so that either can carry the required 28 engineered safeguards load. A pmliminary estimate of 29 the rating of each emergency generator is 2850 kw.

1 The unit will generate electric power at 22 kv, which will be 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 turn, is linked to Applicant's 5

existing transmission network as follows: the 230 kv substation is 6

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 9

Florida and one to North Pinellas. There will be no transformation 10 tie between the 500 kv and the 230 kv substations at the Crystal 11 River Generating Plant.

12 3.7 Auxiliary Systeras 13 Auxiliary systems am 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 in the reac-21 tor coolant, and maintains proper water chemistry.(45)

The decay heat mmoval system cools the mactor when the reactor 22 23 system is depressurtzed for maintenance or mfueling. This same 24 system serves the engineered safeguards function of recirculating 25 borated water to cool the core in the event of a loss-of-coolant 26 accident.(46)

The cooling water systems maintain temperatures throughopt the 27 28 equipment and structures of the plant.(47) Appropriate normal 29 ventilation systems are provided in the plant.(48)

I 1

A fuel handling system (49) provides the means for safe, 2

reliable handling' of fuel from the time it enters the plant site 3

as new fuel until it is shipped from the plant site as used fuel.

4 Irradiated fuel is stored and handled under water at all times 5

until after it is placed into a shipping cask. The water provides 6

a radiation shield as well as a reliable source of cooling for the 7

irradiated fuel assemblies. A spent fuel cooling system maintains 8

the temperature of the spent fuel storage pool water within accept-9 able limits.(50) 10 3.8 Steam and Power Conversion Sys tem 11 The steam and power conversion system is designed to remove the 12 heat energy generated in the reactor core by producing steam in the 13 two steam generators. This heat energy is converted to electrical 14 energy by the turbine-generator. This cycle, including the necessary 15 equipment to achieve sa#e and reliable operation, is similar in con-16 cept and design to. turbine-generator cycles in successful use for 17 many years.(51)

-18 3.9 Radicactivity Control Systems 19 Radioactive gasecus, 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 ti on. These systems are designed to minimize the mlease of radio-24 active material from the plant to the environment and will maintain 25 releases below the limits of 10 CFR 20.(52) & (53) 26 A process radiation monitoring system monitors effluent released 27 to the environment and provides an early warning of possible equipment 28 malfunction or potential radiological hazard.

The radiation monitoring 29 system includes a combination of continuot s automatic monitoring and 30 periodic sampling.(54) & (55) 4,

~

1 Shielding throughout the unit insures that radiation doses 2

to the general public and to operating personnel during nomal 3

operation are well within the limits of 10 CFR 20.(56) 4 4.

SAFETY ANALYSES 5

Potential malfunctions or equipment failures have been analyzed 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 8

radiation in excess of the limits established in the AEC's regulation 9

for siting requirements,10 CFR 100, even in the very unlikely event 10 that one of the accidents postulated in the Application should occur.(57) 11 Two categories of malfunctions or equipment failures have been 12 analyzed; those in which the core and coolant boundaries are protected 13 and those in which one of these boundaries is not. effective and standby 14 safeguards are required. The core and coolant boundary protection 15 analysis shows that, in the event any of the postulated malfunctions 16 were to occur, the normal protection systems operate to maintain the 17 integrity of the core and of the coolant boundary.(58) The standby 18 safeguards analysis demonstrates the capability of the engineered 19 safeguard systems to assure protection of the public for postulated 20 malfunctions in which the normal protective systems may not maintain 21 the integrity of the core and coolant boundary.(59) These analyses 22 show that for all credible malfunctions the radiation exposure to 23 tt e gneral public is well below the limits prescribed in 10 CFR 100.

24 Of the postulated equipment failures, a loss-of-coolant acci-25 dent is the most severe. Emergency 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 29 pi pe. The core cooling system insures that the core will mmain in 30 place and intact.(60) The reactor building spray or emergency cool-31 ing units maintain the integrity of the mactor building,(61) and r

I the iodine removal sprays in conjunction with the reactor building 2

assure that the public is protected from radiation and radioactive 3

material. (62) Emergency electrical power is available on-site to 4

insure operation of these systems even if all extemal sources of 5

electric power to the plant are assumed to be unavailable at the 6

time of the accident.(63) 7 Results of the safety analyses show that, even in the event of 8

a loss-of-coolant accident, no core melting will occur.(64) Howeve r, 9

in order to demonstrate that the operation of a nuclear power station 10 at the proposed site does not present any undue hazard to the general 11 public, a hypothetical accident has been analyzed involving release 12 of 100% of the noble gases, 50% of the halogens and 1% of the solids 13 in the fission product inventory. The analysis evaluated both the 14 direct radiation exposure and the potential total dose to the thyroid 15 from the inhalation of fission products which leak from the reactor 16 building. The low leakage rate of the reactor building and the 17 iodine removal spray system reduce the potential radiation dose to 18 the thyroid to below the 10 CFR 100 guidelines even in the event 19 of such a hypothetical occurence.(65) 20 5.

TESTS, INSPECTIONS, AND QUALITY CONTROL 21 Pressure containing components of the reactor coolant system 22 will be designed, fabricated, inspected, and tested in accordance 23 with Section III, Nuclear Vessels, of the American Society of 24 Mechanical Engineers Boiler and Pressure Vessel Code. The piping 25 will meet the applicable provisions of Power Piping USA Standards 26 831.1.0-1955 and associated nuclear code cases.(66) Nondes truc-27 tive testing, including radiography, ultrasonic, magnetic particle, 28 -and liquid penetration examinations will be perfonred during fabri-29 cation of the nuclear vessels.

30 Auxiliary systems and equipment will be designed, fabricated, 31 and tested to the appropriate provisions of recognized codes and i

i 1

standards of organizations such as the Arrarican Society of Mechani-2 cal Engineers, American Society for Testing Materials, USA Standards 3

Institute, and Institute of Electrical and Electronics Engineers.

4 A comprehensive field testing program will be conducted to 5

insure that equipment and systems perform in accordance with design 6

cri teria. Tests will be perfonned both before and after fuel load-7 ing and criticality.(67) 8 The reactor building will be designed and built in accordance 9

with applicable portions of the Build ng Code Requimments for 10 Reinforced Concrete, ACI 318-63; Specification for Structural 11 Concrete for Buildings, ACI 301-66; AISC Manual of Steel Construc-12 tion; ASME Boiier and Pressure Vessel Code, Sections III, VIII, 13 and IX.(68) Materials and workmanship will be inspected to insure 14 compliance with appmpriate codes, specifications, and standards.

3 Materials to be inspected and tested include concmte, liner plate, 16 prestressing system materials, hatches, penetrations, structural 17 and reinforcing steel.(69) 18 The mactor building will be structurally tested at 115 percent 19 of design pressure.(70) In addition, it will be leak tested to 20 insure compliance with a maximum allowable gmss leak rate of one 21 fourth of one percent by weight of the contained volume of air in 22 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (71) at the design pressure. The penetrations will be 23 periodically pressurized to design pmssure to demonstrate their 24 leak tightness.(72)

Consideration 'as been given to the inspectability of the 25 26 reactor coolant system in the design and arrangement of components.

Access for inspection of the reactor coolant system includes access 27 28 for visual examination by dimet or remote means. c.

~

1 The Applicant's contractors and equipment suppliers will pro-vide required quality control functions, procedures, and techniques 2

3 to assure manufacture and construction in accord with the plant 4

design and specifications furnished to the Applicant by its 5

architect / engineers, Gilbert Associates, Inc.

6 In addition, the Applicant, through its Power Engineering and Construction Department, will provide a quality assurance program 7

8 using as appropriate its own personnel, those of Gilbert Associates, 9

and/or qualified independent testing agencies to monitor contractors 10 and suppliers compliance with the above quality control requimments.

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 14 and specifications.

15 6.

RESEARCH AND DEVELOPMENT PROGRAMS 16 The nuclear steam supply system for the Crystal River Plant 17 Unit 3 is similar in concept to several pmjects already in opera-18 tion, under construction, or mcently licensed by the Atomic Energy 19 Connission.

The pmliminary design is based on the technical data 20 which have been developed in the nuclear industry and on data deve-21 loped by B&W which is specifically related to the Crystal Piver 22 Plant Unit 3.

To complete the final detail design of some compon-23 ents additional technical information will be obtained.

24 The following am the areas of the plant design in which addi-25 tional technical data will be developed to finalize design details.

26 a.

Once-Thmugh Steam Generator 27 The design of the once-thmugh steam generator is based on 28 experimental work on boiling heat transfer and data obtained 29 by B&W in full length model tests of the unit.

The testing 1

of a pmtotype unit has been completed but not yet documented.

2 It includes perfomance, mechanical, vibration and blowdown 3

tests, and contml system development. The results 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 models to 7

finalize the design of the steam generators.(73) 8 b.

Contml Rod Drive Unit 9

The design of the contml rod drive mechanisms is based on 10 a principle which has been used in operating mactors and 11 which has been extensively tested by B&W. Test programs 12 have included full scale pmtotype testing under no-flow 13 conditions, full scale pmtotype testing at operating 14 conditions, and components testing.

Testing of a proto-15 type mechanism was carried out for a full-life cycle of 16 strokes and trips, and major design parameters wem 17 confi rmed.

Life cycle testing is being repeated usinti 18 a pinion gear of improved material.

Data from these test 19-programs will be used in the final design of the control 20 rod, its guide structure, and the contml rod drive 21 mechanism.(74) 22 c.

In-Core Neutmn Detectors 23 The performance and longevity of the self-powered detectors 24 is being demonstrated by detectors installed in the Babcock 25 and Wilcox Test Reactor and in the Rig Rock Point Nuclear 26 Power Plant.

The tests have demonstrated that the detectors 27 perform successfully. They are being continued in order to 28 demonstrate their longevity, f t. the present time, the Big 29 Rock Point detectors have accumulated operational experience 30 -

equivalent to three years of operation in the Crystal River 31 reactor.

The Babcock & Wilcox Test Reactor detectors have 32 accumulated an equivalent of two years operation.

1 d.

Core Thennal 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 correlation 5

of experimental DNB (Departum fmm Nuclear Boiling) data 6

developed by The Babcock & Wilcox Company.

In order to 7

conpletely substantiate the B&W correlation additional 8

research and development data is necessary.

These sequi m-9 ments am described in the PSAR.(75) 10 Subsequent to submittal of the original PSAR, core thermal 11 performance was also evaluated using the W-3 correlation 12 for predicting DNB. This correlation is aseilable in the 13 literature and has been used and found acceptable in 14 establishing themal design limits for other large pres-15 surized water reactors. The thermal evaluation using the 16 W-3 correlation is also presented in the PSAR and its 17 supplements. With the use of this correlation, vessel 18 model ficw tests are necessary to substantiate operation 19 of the plant within acceptable thennal 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 e.

Emergency Core Cooling and Intemal Vent Valves 26 Analytical evaluation of the effects of blowdown forces on 27 the intemals and of the performance of the intemal vent 28 valves installed in the core support shield to insum 29 adequate covering of the core by emergency coolant is in 30 pmgress. A prototype of these valves will be tested to 31 demonstrate their operating characteristics.(76) i I

(

f 1

f.

Fuel Failure 2

A study, including testing, is underway to assure that 3

there am no failure mechanisms which might interfem 4

with the ability of the emergency com cooling systems to 5

accomplish their objective. The results of the work to 6

date demonstrate the ability of the design to accommodate 7

potential fuel failure mechanisms. This work will be 8

continued to assure that fuel rod failums will not affect 9

significantly the ability of the emergency core cooling 10 system to prevent clad nelting.

11 g.

Xenon Oscillations 12 The possibility of the occurrence of xenon oscillations 13 throughout core life is being evaluated.

If it is 14 determined that such oscillations may occur appropriate 15 design changes to eliminate or control the oscillations 16 will be incorporated.(77) The design of means to eliminate 17 or control such oscillations is being carried out in 18 parallel with the studies of the possibility of such 19 oscillations.

20 h.

Sodium Thiosulphate 21 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 25 to remove and retain iodine effectively and to demonstrate 26 the compatibility of the chemical with plant materials is 27 in progress.(78) -

e 1

7.

TECHNICAL QUALIFICATIONS 2

7.1 Florida Power Corporation 3

Florida Power Corporation is responsible for the design, pur-4 chasing, construction, and operation of Crystal River Plant Unit 3.

5 This practice has been successfully followed for all of the Company's 6

major generating facilities now in service or planned.

7 The Applicant has 68 years experience in the design, construc-8 tion, and operation of electric generating plants.

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-14 ing mobile units) for a total net electric generating 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 18 unit and 4 gas turbine engine driven units of 30 MW each at two sepa-19 rate plant locations.

20 7.1.1 Power 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 studic^, design specifi-25 cations, procurement, engineering 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)

(

1 The msponsibility for establishing and executing a nuclear 2

training program and maintaining liaison with the Power Production 3

Department is a vital function and responsibility of the Power 4

Engineering staff.

5 Power Engineering is under the direction of the Manager of 6

Power Engineering who is responsible to the Nuclear Project Manager.

7 7.1.2 Power Construction 8

Florida Power Corporation has a staff of graduate engineers to 9

provide the construction management and engineering skills including 10 quality contml required during the course of power plant construction 11 and is msponsible for construction management of the Crystal River 12 Nuclear Plant.

13 With its nucleus of six experienced construction engineering 14 and supervisory personnel, the Applicant's construction staff 15 manages all activities of some 500 power plant construction 16 workers employed by subcontractors to the Applicant performing 17 construction work. An Applicant owned or leased inventory of 18 essential and modem construction equipment is kept in madiness, 19 and the latest construction techniques are employed, including 20 computer processed CPM analysis of the construction activities 21 and computerized cost accounting and cost contml. The construc-22 tion section is under the direction of the Manager of Power 23 Construction who is responsible to the Nuclear Pmject Manager.(80) 24 7.2 The Babcock and Wilcox Company 25 B&W's participation in the develapment of nuclear power dates 26 fmm the Manhattan project.

B&W's bmad nuclear activities include 27 applied research to develop fundamental data; design and manufacture 28 of nuclear system, cores, and components; and design, manufactum, 29 and entction of complete nuclear steam generating systems. Through 1

B&W's divisions a wide range of equipment for nuclear application 2

is designed and manufactured. The B&W Company's major nuclear 3

contracts, in addition.to manufacture of a substantial percentage 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, Thme Mile Island Nuclear Station and ' other units in 7

various stages of licensing in addition to Crystal River Plant 8

Unit 3.(81) 9 In addition to supplying the Crystal River Plant Unit 3 nuclear 10 steam system, B&W's Power Generation Division will emct this equip-11 ment and be responsible for the quality control procedures mquired 12 during emetion phases.

13 7.3 Gilbert Associates, Inc.

14 Gilbert Associates, Inc. has been retained by the Applicant as 15 the Architec'c-Engineer for this project.

They will fumish plant 16 layouts, diagrams, and system arrangements and provide specifications 17 for major items of equipment and systems. Provision is made for 18 consulting services required as well as resident engineering per-19 sonnel' during construction.

20 Gilbert Associates, Inc., engineers and consultants, was organized 21 in 1906 and has its main office at Reading, Pennsylvania.

Since 1942 22 GAI has been responsible for the design of over 110 thermal generat-23 'ing units, both fossil and nuclear, mpresenting more than 16,000,000 24 kilowatts of capacity. Design experience includes mheat cycles,

'S once-through boiler units, and supercritical units in ratir.gs up to

.6 900,000 kilowatts. At present GAI has over 8,000,000 kilowatts of 27 generation under design.

28 Gilbert projects since 1950 include complete programs of nuclear 29 power development involving analysis of sites, complete evaluations.-

+.

1 of proposals, contract and fuel program assistance, preparation of license applications, complete plant design and procurement.(82) 2 3

7. 4 J. A. Jones Construction Company 4

The J. A. Jones Construction Company will provide general 5

contractor service; for Crystal River Plant Unit 3.

Through a subsidiary company, Livsey & Company, 'nc., the mechanical equip-6 7

ment and piping erection will be accomplished.

This most highly qualified construction finn brings to this 8

project its significant nuclear project experience over the past 9

10 yea rs.

Besides work on the Oak Ridge gaseous diffusion plant and 11 the plutonium production reactors at Hanford their latest work is the completion of the Jersey Central Oyster Creek Unit #1 under 12 13 contract to General Electric.

14 J. A. Jones Construction Company has developed and established quality control procedures, techniques and testing skills as a result 15 16 of this extensive experience with the AEC. This background will bring to the Crystal River project a more comprehensive program for 17 18 overall quality control.

19 8.

COMMON DEFENSE AND SECURITY 20 There is no indication that construction and operation of the i

21 Crystal River Plant Unit 3 will in any way be inimical to the 22 common defense and security of the United States.

23 As stated in the Application, Florida Power Corporation is a 24 Florida Corporation engaged as a public utility in the production, 25 transmission, and sale of' electric energy. All of the directors 26 and principal officers of the Company are citizens of the United 27_ States and the Company is not owned, controlled or dominated by

- 28 an alien, e foreign corporation, or a foreign government.,

f

l 1

The application contains no restricted or other defense 2

information and Applicant has agreed that it will not permit 3

any individual to have access to Restricted Data until the Civil 4

Service Comission shall have made an investigation and mport 5

to the Atomic Energy Commission on the character, associations 6

and loyalty of such individual, and the Atomic Energy Commission 7

shall have deterinined that permitting such persons to have access 8

to Restricted Data will not endanger the common defense and 9

s ecuri ty.

10 As a licensee, Applicant will be subject to regulations of the 11 Atomic Energy Commission relating to the transfer of and accountability 12 for special nuclear material in its possession. Recent amendments 13 to the AEC Rules and Regulations (10 CFR 50.60) under which the AEC 14 will discontinue allocating quantities of special nuclear material 15 to reactor licenses evidence that such material is no longer scarce.

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.

CONCLUSION 20 On the basis of the foregoing and the Application, the Applicant 21 respectfully submits that:

22 a.

Florida Power Corporation's Application, as amended, des-23 cribes the proposed design of the Crystal River Plant 24 Unit 3, including the principal architectural and engineer-25 ing criteria for the design, and identifies the major fea-26 tures or components incorporated in the plant for the 27 protection of the health and safety of the public.

28 b.

The Application, as amended, identifies the technical or 29 design information necessary to complete the final safety 1

analys is. Such information can reasonably be lef t for 2

later consideration and will be supplied in the final 3

safety analysis report.

4 c.

Safety featums which mquire further research and develop-5 ment, and the research and development programs to be carri-6 ed out, am 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 msolve any questions associated with such featums at or 10 before the latest date stated in the Application for comple-11 tion of construction of the facility.

12 d.

Taking into consideration the characteristics of the site 13 and environs and the proposed design of the Crystal River I

14 Plant Unit 3, such facility can be constructed and operated 15 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 l

19 construct the proposed facility; and 20 f.

The issuance of a construction pennit for the Crystal River 21 Plant Unit 3 will not be inimical to the conmon defense and 22 security or to the health and safety of the public.

APPENDIX A LIST OF REFERENCES f

+

Appendix A LIST OF REFERENCES (1) PSAR, Volume 1, Section 2, Paragraph 2.2.4 (2) PSAR, Volume 1, Section 2, Paragraph 2.3.1 (3) PSAR, Volume 1, Section 2, Paragraph 2.4.1 (4) PSAR, Volume 1, Section 2, Paragraph 2.4.2 (5) PSAR, Volume 1, Section 2, Para:raph 2.4.5 (6) PSAR, Volume 1, Section 2, Paragraph 2.5 (7) PSAR, Volume 1, Section 2, Paragraph 2.6 PSAR, Appendices, Appendix 21 (8) PSAR, Volume 1, Section 2, Paragraph 2.7 PSAR, Volume 4, Supplement 1, Questions 5.8 & 6.1 (9) PSAR, Volume 1, Section 1, Paragraph 1.1 (10) PSAR, Volume

, Table 1-2 (11) PSAR, Volume 1, Table 3-2 (12) PSAR, Volume 1, Table 3-1 (13) PSAR, Volume 1, Section 3, Paragraph 3.2.3 (14) PSAR, Volume 1, Section 3, Paragraph 3.2.2.1.2 PSAR, Volume 2, Section 7, Paragraph 7.2.2.1.2 (15) PSAR, Volume 1, Table 3-6 PSAR, Volume 1, Figure 3-1 (16) PSAR, Volume 1, Section 3, Paragraph 3.2.2.1.3 (17) PSAR, Volume 1, Section 3, Paragraph 3.2.4.3.2 (18) PSAR, Volume 2, Section 9, Paragraph 9.2 (19) PSAR, Volume 1, Section 4, Paragraph 4.2.2.1 (20) PSAR, Volume 1, Section 4, Paragraph 4.1.4.4 (21) PSAR, Volume 1, Section 4, Paragraph 4.4.3 (22) PSAR, Volume 1, Section 4, Paragraph 4.2.5 (23) PSAR, Volume 1, Section 4, Paragraph 4.2.2.4 A-1

Appendix A (24) PSAR, Volume 1, Section 4, Paragraph 4.2.2.3 (25) PSAR, Volume 1, Section 4, Paragraph 4.2.2.2 (26) PSAR, Volume 2, Section 5 (27) PSAR, Volume 2, Section 5, Paragraph 5.1 (28) PSAR, Volume 2, Section 5, Paragraph 5.1.2.1 PSAR, Volume 2, Section 5, Paragraph 5.1.2.8 (29) PSAR, Volume 2, Section 5, Paragraph 5.5 (30) PSAR, Volume 2, Section 5, Paragraph 5.1.2.2 (31) PSAR, Volume 2,-Section 5, Paragraph 5.6.2.1 (32) PSAR, Volume 2, Section 5, Paragraph 5.6.2.2 (33) PSAR, Volume 2, Section 6 (34) PSAR, Volume 2, Section 6, Paragraph 6.1 (35) PSAR, Volume 2, Section 6, Paragraph 6.2 (36) PSAR, Volume 3, Section 14, Paragraph 14.2.2.4 (37) PSAR, Volume 2, Section 7, Paragraph 7.1.1 (38) PSAR, Volume 1, Section 3, Paragraph 3.2.4.3 (39) PSAR, Volume 2, Section 7, Paragraph 7.4.3 (40) PSAR, Volume 2, Section 7, Paragraph 7.2.3.2 (41) PSAR, Volume 2, Section 7, Paragraph 7.3.1 (42) PSAR, Voluna 2, Section 7, Paragraph 7.2 (43) PSAR, Volume 2, Section 7, Paragraph 7.1.2.2 PSAR, Volume 2, Section 7, Paragraph 7.1.3.3 (44) PSAR, Volume 2, Section 8 (45) PSAR, Volume 2, Section 9, Paragraph 9.1 (46) PSAR, Volume 2, Section 9, Paragraph 9.5 (47) PSAR, Volume 2, Section 9, Paragraph 9.3 (48) PSAR, Volume 2, Section 9, Paragraph 9.7 A-2

s Appendix A (49) PSAR, Volume 2, Section 9, Paragraph 9.6 (50) PSAR, Volume 2, Section 9, Paragraph 9.4 (51) PSAR, Appendices, Appendix 1A, Section 1 PSAR, Volume 2, Section 10 (52) PSAR, Volume 3, Section 11, Paragraph 11.1 (53) PSAR, Volume 4, Supplement 1, Question 6.1 (54) PSAR, Volume 3, Section 11, Paragraph 11.1.2.4 (55) PSAR, Volume 4, Supplement 1, Figure 5.8-5 (56) PSAR, Volume 3,.Section 11, Paragraph 11.2.1 (57) PSAR, Volume 3, Section 14 (58) PSAR, Volume 3, Section 14, Paragraph 14.1 (59) PSAR, Volume 3, Section 14, Paragraph 14.2 (60) PSAP., Volume 2, Section 6, Paragraph 6.1 (61) PSAR, Volume 2, Section 6, Paragraph 6.2 (62) PSAR, Volume 3, Section 14, Paragraph 14.2.2.3.5 (63) PSAR, Volume 2, Section 8, Paragraph 8.2.3 (64) PSAR, Volume 3, Section 14, Paragraph 14.2.2.3 (65) PSAR, Volume 3, Section 14, Paragraph 14.2.2.4 (66) PSAR, Volume 1, Section 4, Paragraph 4.1.5 (67) PSAR, Voleme 3, Section 13 (68) PSAR, Volume 2, Section 5, Pdr y raph 5.1.2.4 (69) PSAR, Volume 2, Section 5, Paragraph 5.6 (70) PSAR, Volume 2, Section 5, Paragraph 5.6.1.2 (71) PSAR, Volume 2, Section 5, Paragraph 5.6.1.3 (72) PSAR, Volume 2, Section 5, Paragraph 5.6.2.1 (73) PSAR, Volume 4, Supplement 1, Question 1.4 A-3

Appendix A (74) PSAR, Volume 1, Section 1, Paragraph 1.5.2 PSAR, Volume 4, Supplement 1, Question 1.4 (75) PSAR, Volume 4, Supplement 1, Question 1.4 (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 (79) PSAR, Volume 1, Section 1, Paragraph 1.6, Figure 1-12

& Appendices, Appendix 1A, Paragraph 1.4 (80) PSAR, Volume 1, Section 1, Paragraph 1.6, Figure 1-12,

& Appendices,. Appendix 1A, Section 1 (81) PSAR, Appendices, Appendix 1A, Section 2 (82) PSAR, Appendices, Appendix 1A, Section 3 A-4

a APPENDIX B FIGURES i

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e e

APPENDIX C PROFESSIONAL QUALIFICATIONS OF EXPERT PANEL WITNESSES

w 1

EDUCATIONAL AND PROFESSIONAL' QUALIFICATIONS

-2

-DONALD J. R0WLAND 3~

MECHANICAL ENGINEER 4

POWER ENGINEERING AND CONSTRUCTION DEPARTMENT 5

FLCRIDA POWER CORPORATION 6

1.

My name is Donald J. Rowland. My residence is 6401 - 31st 7

Avenue North. St. Petersburg, Florida.

I am employed by Florida 8

Power Corporation, Power Engineering and Construction Depart.

9-ment as a Mechanical Engineer.

10 2.

I graduated from Auburn University in 1958 with a Bachelor of 11 Mechanical Engineering degree.

12 3.

In 1958, I accepted a position at Florida Power Corporation in 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 17 West Coast Nuclear Group - East Central Nuclear Group, gas 18 cooled reactor.re:earch and development project. My duties were 19 to maintain liaison contact between the project and Florida 20 Power Corporation and to assist General Nuclear Engineering Cor-21-poration in design activities.

22 5.

Ir. 1961, I was reassigned to field supervision on power plant 23' construction projects an'd progressed through positions of re-24-sponsibilityfin plant construction and design engineering.

4 s

J

' C-1 2

DONALD J. R0WLAf40 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 mechanical and nuclear design engineering associated with 4

new generating plants.

C-2

1 EDUCATIONAL AND PROFESSIONAL QUALIFICATIONS 2

E. ROBERT HOTTENSTEIN 3

PROJECT MANAGER, NUCLEAR ENGINEERING DEPARTMENT 4

POWER DIVISION 5

GILBERT ASSOCIATES, INC.

6 1.

My name. is E. Robert Hottens tein. My residence is Old State 7

Road, Oley, Pennsylvania 19547.

I am employed by Gilbert

-8 Associates, Inc. as a Project Manager.

In this position, I 9

am responsible for the overall technical direction and 10 related administrative effort that emcompasses the licens-4 11 ing and engineering efforts associated with the design of 12 Crystal River Plant Unit 3 for which Gilbert Associates, 13 Inc. is responsible.

14 2.

I was graduated from the Pennsylvania State University in

'15 1950 with a Bachelor of Science Degree in Mechanical 16 Engineering. From 1955 to 1959 I studied nuclear physics 17 and nuclear engineering at North Carolina 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.

25 5.

From-1955 to 1959, I was assigned to the Reading Offic 26' where I participated in nuclear plant studies for the 27 U. S. AEC, U. S. Air Force, and public utilities.

j C-3

- - _ ~

E. ROBERT H0TTENSTEIN j.

1 6.

From 1959 to 1962,'I was assigned to the Saxton Experi-2 mental.eactor Project.

For this project, I was res-3 ponsible for the design and start-up of the radioactive 4

waste disposal facility.

i 5

7.

From 1962 to 1965, I participated in the General Public 6

Utilities Oyster Creek Proposal Evaluation Program and 7

did nuclear siting studies for the Pennsylvania Power 8

and Light Company.

9 8.

From 1965 to 1966, I was assigned to the Robert Emmett 10 Ginna Nuclear Station Project for the Rochester Gas and 11 Electric Company as Project Nuclear Engineer.

'12 9.

From 1966 to 1967, I was the GAI Project Engineer on the i

13 Florida Power Corporation Nuclear Plant Siting Study, 14 Nuclear Steam System - Specification and Evaluation Team.

15

10. From 1967 to present, I have served as the GAI Project

~16 Manager 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 Nuclear Society, and a Licensed 19 Professional Engineer in the State of Pennsylvania.

4 i

a I

.I f

b C-4

1 EDUCATIONAL AND PROFESSIONAL QUALIFICATIONS 2

MORTON 1. GOLDMAN 3

VICE PRESIDENT AND GENERAL MANAGER 4

ENVIRONMENTAL SAFEGUARDS DIVISION 5

NUS CORPORATION 6

1.

My name is Morton I. Goldman. My address is 1730 M Street, 7

N.W., Washington, D. C., 20036.

I am Vice President and 8

General Manager of the Environmental Safeguards Division of 9

NUS Corporation and have served in this capacity since January 10 1966.

I am responsible for all site evaluations, safety analyses, I

11 waste management system design and environmental program develop-12 ment conducted by this Division. This has included the evaluation 13 of site and environmental safety factors for a number of nuclear 14 and tossil fueled plants in this country and abroad including the 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 Company), Surry (VEPCo), Three Mile Island 18 (Metropolitan Edison), Prairie Island (NSP) Burlington and Salam 19 (PSE&G), Zion (Commonwealth Edison), Kewaunee (WPSCo), Calvert 20 Cliffs (BG&E), Diablo Canyon (PG&E), and Beaver Valley (Duquesne 21 Light Company).

22 2.

I was graduated from the New York University in 1948 with the

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 Master of Science degree in Nuclear Engineering and in 1960 J

26 a Do'ctor of Science degree, all from the Massachusetts Institute 27 of Technology.

C-5

c MORTON I. GOLDMAN

l. '

1 3.

From 1948 to 1949 I was a Research and Teaching Assistant at 2

the. Sanitary Engineering Research Laboratory, New York Univer-3 sity conducting research on water coagulation and assisting in 1

4 teaching sanitary cherastry and sanitary biology laboratory 5

courses.

6 4.

From 1949 to 1950 I was a Research Assistant at the Radioactivity 7

Research Laboratory, Sanitary Engineering Department at MIT con-8 ducting original research on removal of radionuclides from water i

9 by standard water treatment techniques.

10 5.

From 1950 to 1961 I was a Commissicaed Officer with the United 11 States Public Health Service, Div#sion of Radiological Health.

12 I was first assigned to the Radialogica)..ealth Training Section 13 from 1950 to 1954 as the engineer staff member lecturing on appro-14 priate aspects of radiological safety and waste disposal.

15

-- 6.

From 1954 to 1956 I was on loan to the Oak Ridge National Labor-16 atory as Chief of Soils and Engineering Section, Waste Disposal 17 Research Activities.

In this position I conducted and supervised

~ 18 research on disposal of radioactive wastes at Oak Ridge National 19 Laboratory.

20

.7.

From 1956 to 1959 I was assigned to MIT as Project Leader for the 21 Radioactive Waste Disposal Project of the Sanitary Engineering 22-Department and in-training in the Nuclear Engineering Department.

23-In the former capacity I initiated and supervised research on

[

24 novel methods of disposal of high..tivity fission product waste C-6

MORTON I. GOLDMAN 4

1 materials.

In addition, I served on the MIT Reactor Safeguards 2

Committee as its secretary.

3 8.

From 1959 to 1961 I was designated as Nuclear Installation 4

-Consultant with the Division of Radiological Health in Washington, 5

D. C.

In this capacity I provided technical consultation and 6

assistance to State Health Agencies and other Federal Agencies on 7

health. and safety problems associated with nuclear installations.

8 As part of my responsibility I served as the evaluator responsible 9

for the following nuclear plants: Yankee, Elk River, Indian Point, 10 Carolina-Virginia, Hallam, Pathfinder, Peachbottom and Humboldt 11 Bay.

12 9.

Since 1961 I have been with NUS Jorporation and active in all of 13 the environmental safety activities described earlier.

14

10. I am the author.and co-author of a number of papers on radiation

.151 and public health, nuclear safety and radioactive waste management.

16

11. I am a member of the American Society of Civil Engineers (ASCE),

17 the Water' Pollution Control Federation, the American Association 18 for the Advancement of Science, the American Nuclear Society and 19 the Air Pollution Control Association.

I am also a Licensed Pro-20 fessional Engineer in the State of New York and the District of 21 Columoia and a' Diplomate of. the American Academy of Environmental 22 Engineering in Radiation tiygiene and Hazard Control.

I am also 23 a memoer of Committee N18 " Nuclear Design Griteria" of the USA C-7.

r MORTON I. G0LbrAN i

l Standards Institute.

I served as the U. S. representative 2

on an expert panel of waste management practice at nuclear 3

power plants at the International Atomic Energy Agency in 4

Vienna.

T i

i t

i t

i r

1 i

f 5

i 5

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1 i

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C -

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E

+

1-EDUCATIONAL AND PROFESSIONAL QUALIFICATIONS 2

CARL EDWARD THOMAS 3

PROJECT MANAGER 4-NUCLEAR POWER GENERATION DEPARTMENT 5

POWER GENERATION DIVISION 6-THE BABC0CK & WILC0X COMPANY 7

1.

1 -

My name is Carl E. Thomas. My residence is Route 1, Madison 8

Heights, Virginia.

I am employed by The Babcock & Wilcox 9

Company, Power Generation Division, in the Nuclear Power Gen-10 eration Department.

11 2.

I graduated from the University of Chattanooga in 1954 with a I

12 B.S. in Engineering Physics.

In 1955, I graduated from the 13 Oak Ridge School of Reactor Technology.

From 1957 to 1962. I 14 participated part time in the University of Virginia Engineer-15 ing Graduate Study program.

16 3.

I served in the United States Army Air Corps from 1944 to 1945.

17 4.

In 1955, I joined The Babcack & Wilcox Company. My early assign-18 ments were associated with the Aqueous Homogeneous Reactor, the 19 Moderator Control Reactor and the liquid metal fuel ractor 20 concepts as a reactor physicist.

4 l '

21 5.

In 1960, I became Chief of the Operationel Analysis Section with 22 responsibility for reactor and system dyr amic analysis, re;ctor 23 control _ analysis, plant performance, and tafety analysis.

In 24 -

1964, I was appointed Director of the Savannah Technical Staff.

25 and later was designated Manager of Technical Services, Savannah 26 Technical Staff. My responsibilities in these capacities e

d C-9 5.

s n

-r

~

^

CARL EDWARD i..sMAS 1

included engineering, shipyard modifications to the Savannah 2

reactor plant, and safety and licensing activities.

In addition, 3

I served as nuclear advisor aboard the N. S. Savannah during 4

port visitation voyages to American and European ports.

5

6..In 1965, I was appointed Assistant Manager of the Reactor 6

Enginee-ing Department of the Atomic Energy Division.

In this 7

capacity, I was responsible for the engiuering of marine reactor 8

plants.

9 7.

In 1966, I joined the Contract Department and am now serving as 10 Project Manager for the B&W contract with Florida Power Corpora-11 tion for Crystal River Unit No. 3.

'l l

a 4

6 a-C-10.

c.

e.

'1 EDUCATIONAL AND PROFESSIONAL QUALIFICATIONS

'2 ROBERT E. WASCHER 3

MANAGER, NUCLEAR SAFETY ENGINEERING SECTION 4

NUCLEAR POWER GENERATION DEPARTMENT 5

POWER GENERATION DIVISION 6

THE BABC0CK & WILC0X COMPANY 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.

11 2.

I graduated from the Illinois Institute of Technology in 1952 12 with a Bachelor of Science Degree in Mechanical Engineering.

13 In 1953 I graduated from the Oak Ridge School of Reactor Tech-14 nology.

15 3.

Upon graduation I joined the Oak Ridge National Laboratory as 16 an Associate Development Engineer responsible for the development j

17 of mechanical components for homogeneous nuclear reactors.

18 4.

In 1955 I was commissioned 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 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 ^ompany as a Nuclear 24 Engineer with responsibility for the safety analysis of the 25 Consolidated Edison Company's Indian % int No. 1 Nuclear Plant.

C-ll

A 7

ROBERT E. WAST.ER I

1 In 1959 I was appointed Supervisor of the Safety Analysis Group

?

with responsibility for safety analysis of nuclear plants designed 3 ~

by B&W, In 1964 I became Chief of the Operational Analysis 4

Section with responsibility for reactor and system dynamic 5

analysis, reactor control analysis, plant performance, and 4

6

. safety analysis.

7 6.

In 1965 I was appointed Manager of the Nuclear Safety Section, 8

' my prest,.t pr ' tion. -In this position i am responsible for 4

9 safoty sr r

of the plants designed by B&W.

T 10

7. -During 1964 and 1965 I was Chairman of the N.S. Savannah Safety i

11 Committee, 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 Commission's J

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 Committee, i

18 8.

I am a registered Professional Engineer in the State of Virginia.

i 4

2 C-12

.-