ML19319E158

Summary Description of CP & OL Application. Contains Descriptions of Site & Environ & of Facility,Safety Analyses,Tests,Insps & Qualifications of Util Personnel

ML19319E158
Person / Time
Site:	Rancho Seco
Issue date:	08/16/1968
From:	SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:
References
NUDOCS 8003310677
Download: ML19319E158 (50)

Text

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i BEFORE THE UNITED STATES ATOMIC ENERGY COMMISSION In the Matter of SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION, UNIT NO.1 Docket No. 50-312

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SUMMARY

DESCRIPTION OF APPLICATION FOR REACTOR CONSTRUCTION PERMIT AND OPERATING LICENSE August 16, 1968

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TABLE OF CONTENTS

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1.

INTRODUCTION...

1 2.

DESCRIPTION OF SITE AND ENVIRONMENTAL CHARACTERISTICS WHICH INFLUENCE DESIGN.

5 2.1 Location.

5 2.2 Population.

5 2.3 Meteorology.

6 2.4 Hydrology..

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2.5 Groundwater.........

7 2.6 Geology...

7 2.7 Seismology.....

7 2.8 Environmental Radiation Monitoring 8

3.

DE6CRIPTION OF RANCHO SECO NUCLEAR GENERATING STATION...

9 3.1 Introduction 9

3.2 Reactor and Primary Coolant System 9

3.3 Reactor Building 13 3.4 Engineered Safeguards.

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3.5 Instrumentation and Control.

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3.6 Electrical Systems 18 3.7 Auxiliary Systems.....

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3.8 Steam and Power Conversion System.

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21 3.9 Radioactivity Control Systems.

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SAFETY ANALYSES

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TESTS, INSPECTIONS, AND QUALITY CONTROL 25 i

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4 TABLE OF CONTENTS (continued) page

6. 'RESEARCH AND DEVELOPMENT PROGRAMS.............

28 32 7.

TECHNICAL QUALIFICATIONS....

7.1 Sacramento Municipal Utility District.

32 34 7.2 Lechtel Corporation 7.3 Babcock and Wilcox Company..............

35 7.4 Other Consultants....

36 8.

COMMON DEFENSE AND SECURITY,..

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CONCLUSION.

39 APPENDICES APPENDIX A - List of References.............. A-1

. B-1 APPENDIX B - Figures..

' APPENDIX C - Qualifications of Witnesses...

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IlfrRODUCTION 4

2 This document is a Susmary Description of the Application,'as 3-amended by Amendments 1-through 4. of Sacramento Muncipal Utility 4

District (SMUD) (referred to as "the Applicant") for a construction 5

permit and facility license to construct and operate the Rancho

-6 Seco Muclear Generating Station, Unit 1, to be located at its Rancho c

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- Seco site in the southeast portion of Sacramento' County, California.

8 This Summary Description includes information on the Rancho Seco 9

site and environment, a description of the Rancho Seco Nuclear 10 -

Generating Station, analyses of the safety aspects of the plant, a i

11 summaary of quality assurance procedures, the research and develop-12 ment programa necessary for the final design, the technical qua-13' lifications of the Applicant and its principal contrectors, and 14 considerations relating to the common defense and security of the 15 United States.

16 This Svamary Description also constitutes a portion of the 17 prepared testimony of the Applicant for the public hearing on its 18 application for a construction permit and is sponsored collectively 19 by the following witnesses representing the Applicant and its 20' contractors:

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1 NAME ORGANIZATION TITLE 2

' John J. Mattimoe SMUD Asst. Chief Engineer and 3

Project Manager 4

Dallas G. Raasch SMUD Project Engineer 5

Ronald C. Stinson, Jr.

SMUD Plant Superintendent 6

Jey R. McEntee Bechtel Corporation Manager, Nuclear Power and 7

Desalting Engineering 8

George S. C. Wang Bechtel Corporation Chief Nuclear Engineer 9

Roy A. Norry Bechtel Corporation Project Engineer 10

' William G. Bingham, Jr. Bechtel Corporatien Project Nuclear Engineer 11 Robert T. Schemer Babcock & Wilcox Project Manager 12 Robert E. Wascher Babcock & Wilcox Manager, Nuclear Safety 13 Engineering Section 14 IJames M. Cutchin IV Babcock & Wilcox Senior Engineet. Licensing 15 The Rancho Seco nuclear generating unit will employ a pres-16 surized water nuclear steam supply system furnished by the Babcock 17

& Wilcox Company (referred to as B&W") and is similar in design 18

'to the nuclear steam supply systems which are being furnished by B&W 19 to the Duke Power Company for its Oconee Nuclear Station (AEC 20 Docket Nos. 50-269,-270, and-287) and the Metropolitan Edison Company-l l

21 for the Three Mile Island Nuclear Station (AEC Docket Jo. 50- 269).

22 A construction permit authorizing construction of the Oconee facil-23 ities was issued in November 1967 and a construction permit autho-24 rizing construction of the Three Mile facility was issued in May 25 1968 pursuant to Section 104 (b) of the Atomic Energy Act of 1964, 26 as-amended. The nuclear steam supply system will operate initially l

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at core power levels up to 2452 MWe, which corresponds to a gross 2

electrical output of about 850 MWe. An ultimate core output of 3

2568 MWt is expected, and all steam and power conversion equipment 4

is designed accordingly. All plant safety systems, including con-5 tainment and engineered safeguards, are designed and evaluated for 6

operation at this higher power level. The higher power level is 7

also used in the analyses of postulated accidents to establish the 8

suitability of the site under the guidelines set forth in 10 CFR 100.

9 The Applicant's construction permit application, including the 10 amendments thereto, has been reviewed by the staff of the Atomic 11 Energy Commission, which nas prepared and published a safety anal-12 ysis of the Application. The Advisory Committee on Reactor Safe-13 guards (referred to as "ACRS") has also reviewed the Application, 14 as amended through Amendment No. 4, and reported its findings co 15 the Chairman of the U. S. Atomic Energy, Commission in a letter 16 dated July 19, 1968. The ACRS concluded "the proposed plant 17 can be built at the Rancho Seco site with reasonable assurance that 18 it can be operated without undue risk to the health and safety of 19 the public." The AEC staff concluded similarly. Both the AEC Staff 20 Analysis and the ACRS report identify certain aspects of the design 21 with respect to which further technical information must be submitted l

22 or with respect to which research and development must be conducted

.pr' or to the issuance of an operating license. Research and develop-23:

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24 ment items are discussed in Section 6 herein.

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1 The principal architectural and engineering criteria which will 2

govern the plant design are set forth in Section 1.4 of Volume I of 3

the Applicant's Preliminary Safety Analysis Report. These criteria 4

together with the engineered safeguards and other incorporated r

5 systems provide assurance that the proposed Rancho Seco Nuclear 6

Generating Station, Unit 1, can and will be constructed and operated 7

at the proposed location without undue risk to the health and safety 8

of the public.

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2.

DESCRIPTION OF SITE AND ENVIRONMENTAL CHARACTERISTICS WHICH 2

INFLUENCE DESIGN 3

2.1 Location 4

The Rancho Seco Nuclear Generating Station will be constructed 5-

-in the southeastern part of Sacramento County, State of California.

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The site of Rancho'Seco is located 26 miles north-northeast of 7

Stockton and 25 miles southeast of Sacramento as shown in Figure 1, 8

Appendix B.

The site and immediate vicinity is shown in Figure 2, 9

Appendix B.

10 All land comprising the site is exclusively owned by SMUD 11 except for 320 acres in Sections 33 and 34. The land is flat to lightly rolling at an elevation of approximately 200 feet MSL.(1) 12 13 The use of land around the. site to a distance of 5 miles is almost 14 -

exclusively agricultural as shown in Figure 4, Appendix B.

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. should be noted that there are no dairy cattle within this 5-mile f

1 16 radius.(2) j 17 2.2 Population 18 The site exclusion. area, which is under control of the Appli-19 cant, has a minimum radius of 2,100 feet. There are 12 residents 20 within 1 mile and 174 residents within a 5-mile radius from the 21 -

reactor building. The distance to the boundary )f the low population 22 zone has been established as '5 miles.(3) The nearest population 23 center of 25,000 or more is Lodi, located 17 miles south-southwest 24' of.the' reactor building. Other population centers of 25,000 or more L

25 within a 50-mile radius of the site are Sacramento, Stockton, and 26 Modesto.

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Recreational. activities in the inanediate areas surrounding

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2 the site are minimal at the present time and are not expected to 3

change extensively in.the future.

4 2.3 Meteorology i

5 The site meteorology has been extensively investigated to pro-6_

vide a preliminary assessment of environmental consequences of i

7 routine and accidental releases of radioactivity. In general, the i

8 site meteorology is that of the Great Central Valley of California (5) 9 with infrequent storms and limited rainfall during the winter and 10 mostly cloudless skies in the summer. A preliminary study was made 11 of the site atmospheric diffusion characteristics, utilizing con-12 servative meteorological conditions.(0 An extensive, continuous 13 meteorological monitoring program for the site was initiated in A ril 1967.

P 14 15 2.4 Hydrology 16 The site is not intersected by any streams but has excellent drainage'at all times without danger of flooding.( )

17 Since the 18 Rancho Seco Nuclear Generating Station will have a closed-loop 19

. cooling system, radioactive liquid waste will not be released at 20 the site, and consequently,' the possibility of contamination of 21 local water supplies does not exist. Streams and lakes within a 22 50-mile radius from the reactor building are shown in Figure 2.4-1 23 of the Preliminary Safety Analysis Report (PSAR)..

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'l 2.5 Groundwater 2

Groundwater in the Rancho Seco area is stored chiefly in the

~3' Mehrten formation and is part of the Sacramento Valley Ground Water 4 ~ Basin. The normal water table is in excess of 140 feet below natural 5

ground surface.(8) The groundwater movement ic to the southwest 6

with a hydraulic gradient of 10 feet per mile. The quality of 7 _ groundwater is high and meets U. S. Public Health Department standards.

8 It has less than 200 ppm solids with less than 50 ppm hardness (CACO ).

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2.6 Geology 10 The site is located in the low hills at the edge of Sierra 11 Nevada Mountains and is founded on the Pliocene Laguna Formation, 12 underlain by an estimated 1,500 to 2,000 feet of tertiary or older 13 sediments deposited on a basement complex of granitic to metamorphic 14 rocks.. The obtained site exploration data indicates the unfaulted 15 nature of the se.diments and their suitability as a foundation upon 16 which to build the proposed nuclear power generating station.(10) 17 2.7 Seismology 18 There is no indication of faulting beneath the site. The 19 nearest fault system, the Foothill Fault System, is located about 20 10 miles to the east of the site and has been inactive since the 21

~ Jurassic Period, some 135 million years ago. The active faulting

-22 along which historically large earthquake shocks have originated

~23 are the Hayward and San Andreas Faults, some 70 and 89 miles to the 24 west respectively, and the faults 80-plus miles to the east, beyond 25 the' Sierra Nevada Range.(11,12) 7

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There is no reason to anticipate fault propagation in the site 2

area.. Earthquake shaking may occur as the result of shocks along 3

' distant faults but due to their relatively distant origin and 4

nature of the foundation material beneath the site, ground accel-5 erations of no greater than 0.05 g should occur during the life of 6

the plant.. A conservative value of 0.13 g will be used for the 7.

design. The ability of the plant to be safely shut down will not be 8

impaired in the hypothetical event of a maximum horizontal ground acceleration of 0.25 g.(13'14) 9 10 2.8 Environmental Radiation Monitoring 11 An environmental radiation monitoring program will be con-12 ducted at the site co establish existing background radiation levels 13 and to detect any changes which may occur. Surface, well, and

~ 14 rain water; air; milk; soil and silt vegetation; and fish and animal 15 samples will be collected and analyzed for grocs alpha and gross 16 beta-gamma activity. If any significant amount of activity is 17 found, the samples will be analyzed for specific radio nuclides.

18 Sampling points will be located both on-site and off-site.(15) i-8 u

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DESCRIPTION OF RANCHO SECO NUCLEAR GENERATING STATION 2

3.1 Introduction 3

A description of plant features and layout, as well as an 4

evaluation of plant safety are set forth in the Application, as 5

amended. Because final design of the station is not yet complete, 6

the plant description emphasizes the concepts, guidelines, and 7

criteria which will govern final design. The station will consist 8

of a reactor building, an auxiliary building (including control room e

9 and radwaste area), a turbine. structure, a fuel storage building, 10 a shop and warehouse, an administration building, cooling towers, a 11 switchyard, and various other auxiliary structures and equipment.

12 A plot plan of the Rancho Seco Nuclear Generating Station, indicating 13 the general station layout, is shown in Figure 3 of Appendix B.

Table 14 1.3-1 in the Application sets forth a compariso'n of the design para-15 meters of the proposed Rancho Seco Nuclear Generating Station with 16 the Duke Power Company's Oconee Units 1, 2, and 3; Florida Power and 17 Light Company's Turkey Point Units 3 and 4; and Florida Power Corpo-18 ratica's Crystal River Plant Unit 3.

The following is a summary of 19 the principal features of the plant which are significant to safety 20 considerations.

21 - 3.2 Reactor and Primary Coolant System 22

The reactor for the Rancho Seco Nuclear Generating Station is 23 of the pressurized water type. It has an initial rating of 2452 MWe, 24 corresponding to a gross electrical output of about 850 MWe. 16)

25. The nominal operating pressure for the reactor is 2185 psig, with L

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an average temperature of 579 F.

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

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The reactor-core is approximately 129 inches in diameter, with an' active. height of 144 inches.(10) It is made up of 177 fuel as-4-

5 semblies, each consisting of a 15 by 15 array of rods enclosed in 6

.a-square, stainless steel, perforated envelope. The array of rods 7

consists of 208 zircaloy tubes containing uranium dioxide,16 con-8 trol rod guide tubes,-and a center tube available for an in-core d

9 instrumentation assembly.(19) There are approximately 201,520 pounds 10 of uranium dioxide in the. ore.(17) 11.

The. thermal and hydraulic design limits of the core are con-12' servative and are consistent with those of other pressurized water 13 reactors currently in operation or under construction.(17,20) 14 Core reactivity is controlled by a combination of 69 movable 15

' control rod assemblies and a neutron absorber dissolved 'in the 16' coolant. The control rods are an alloy of silver-indium-cadmium 17~

encapsulated in stainless steel. The dissolved neutron absorber 18 is' boric acid.(21) 19 The control rods are used for short-term reactivity control 20 associated'with the changes in power level and also.with changes 21'

'in fuel burnup between periodic adjustments of dissolved boron l22 concentration.I The reactor:can be shutdown by the movable 23.

control rods from any power level-at any tLae.(23) Each movable

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-control rod assembly contains 16 control pins, and is actuated by 2

a separate control rod drive mechanism mounted on the top head of 3

the reactor vessel. Upon trip, the 69 control rod assemblies fall into the core by gravity.f 4}

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Systems are provided so that the concentration of dissolved 6

neutron absorber in the reactor may be adjusted. to maintain the 7

reactor shutdown at room cemp3rature and to provide a safe shut-8 down margin during refueling.(25) The concentration of dissolved y

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absorber is reduced to compensate for long-term reactivity changes, 10 burnup of fuel, and buildup of fission products over the core cycle.

11 The core is contained within a cylindrical reactor vessel 12 having the dimensions of 14 feet 3 inches inside diameter and 37 13 feet 4 inches in overall inside height. The vessel has a spherically-14 dished bottom head with a bolted, removable, spherically-dished top 15 head.(26) The reactor vesel is constructed of carbon steel with 16 all interior surfaces clad with austenitic stainless steel. The 17 reactor vessel is manufactured under close quality control, and l

18 several types of nondestructive tests are performed during fabri-19 cation.- These tests include radiography of welds, ultrasonic 20 testing, magnetic particle examination, and dye penetrant testing.(2D 21 During operation, specimens of reactor vessel materials will 22 be placed in the reactor adjacent to the inside surface of the 23 reactor vessel. These specimens will be subject to irradiation 24 similar to the shell of the reactor vessel. They can be removed i

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l' 3 periodically an'd tested.to ascertain the effects of radiation on the 2 -. -reactor vessel material.(28);

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.Two coolant loops are connected to the reactor vessel'by

'4: (nozzles located near the top of the vessel..Each loop contains

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!5- -one. steam' generator two motor-driven coolant pumps, and the inter-

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6 connecting piping. The reactor coolast-piping is carbon steel clad

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'on'the inside surface with austenitic ;tainless steel'.( '}

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8 coolant is ptamped from'the reactor through each steam generator.and

,v 9 (back..:o.the reactor inlet by two 88,000 gpm centrifugal pumps located at the outlet of each steam generator.(30) 4 11.

The steam ' generator is a vertical, straight-tube-and-shell heat

.12 ' exchanger which produces superheated steam at constant pressure

.- 13 over the power range... Reactor coolant flows downward through the tubes, and steam is generated on the shell side.(31) i 14!

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~ The reactor coolant pumps are vertical single-speed, shaft- '

16 sealed units having.bottoni suction'and horizontal discharge. Each 17 pump has!a separate' single-speed top-mounted motor, which is con-

,,3 nected to the pump by a shaft coupling.( }

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The pressurizer. a vertical! surge tank approximately half-i20 ' filled with reactor coolant and' half-filled with steam, is connected 21>

to the reactor coolant system to contro1Lsystem pressure. The 22 ~ operating pressure ofLthe system:is maintained by operating electric s'

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immersion heaters to increase pressure or.by spray (ng reactdr coolant-s p

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. ater int 9 the steam within this pressurizer tank to reduce 1

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pressure. Self-actuated safety relief valves connected to the 3

pressurizer prevent overpressurization of the reactor coolant

- 4 system.(32) 5 3.3 Reactor Building

.6 The reactor building'is designed to completely enclose the 7

reactor coolant system and portions of the auxiliary and engineered 8

safeguards systems (see Figure 5, Appendix B).

It is a reinforced 9

concrete structure in the shape of a cylinder with a shallow domed

. 10 roof and a_ flat foundation slab. The cylindrical portion is pre-11 stressed by a post-tensioning systen, consisting of horizontal and 12 vertical tendons. The dome has a three-way post-tensioning system.

13 The foundation slab is reinforced with conventional bonded rein-14 forcing. The entire scructure is lined with welded steel plate, 15 1/4-inch minimum thickness, to provide vapor tightness. The founda-16 tion mat will be bearing on dense sandy silt and will be approxi-

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mately 10 feet thick.-

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-1 The building 'is designed to sustain safely all internal and 2

external loading conditions which may reasonably be expected to 3

occur doring theilife of the station or which could result from the 4

postulated design base accident to the reactor's primary coolant 5. system. The tendon system used in the structure is of the un-6

. bonded type with a protective compound used to prevent corrosion.

7 The reactor building is so designed that, with the engineered 8

safeguards systems provided, any leakage of radioactive materials 9

to the environment will result in doses well within AEC's 10 CFR 10 100 guidelines for any of the postulated accidents. The integrated 11 leak rate at design pressure will not exceed one tenth of one percent 12 by volume, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.(33) 13 Prior to operation of the facility, the reactor building will 14

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

'15 The etructural integrity test will be conducted at 115 percent of 16 design pressure, and the leak rate test will be conducted at design

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17 pressure. Deriodic leak rate tests will be performed to assure 18 integrity of the reactor building. A tendon surveillance capability 19 will be available to provide assurance that the tendons are free 20 from harmful corrosion and that excessive steel relaxation has not 21 taken place.

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~3.4 Engineered Safeauards

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. Engineered safeguards are ~ provided to fulfill the following 3

functions in the unlikely event of an accident:

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4 a.' Minimize the release of fission products from the fuel to 5

the reactor building atmosphere i

6 b.

Ensure reactor building integrity and reduce the driving 7

force for building leakage 8

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Remove fission products from the reactor building 9

atmosphere 10 The engineered safeguards systems can be grouped into emer-11 gency core-cooling systems, reactor building cooling systems, and 12 fission product control systems.

13 The emergency. core cooling systems contain both passive 14 flooding systems and pumping systems. The passive flooding system 15 consists of two pressurized core flooding tanks which automatically 16 discht.rge borated water into the reactor vessel in the event 17 the reactor system pressure drops below 600 psi. The pumping system 18

' consists of two completely -independent sub-systems. Each sub-system I

19 contains both a high pressure and a low pressure injection pump.

20 Either sub-system, in conjunction with the core flooding tanks, is 21 capable.of protecting the core for any size leak up to and including 22' the double-ended rupture of the largest reactor coolant pipe.

23 Either sub-system can supply coolant directly from the borated water 124-. storage tank or by recirculation from the reactor building sump o

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1 through heat exchangers which cool it before it is returned to cool 2

the core.

3 The reactor building cooling system, which is made up of two 4

separate and independent heat removal systems, limits the pressure 5

in the reactor building following a loss-of-coolant accident. One 6

system contains four separate fan and cooler units. The other 7

system contains redundant spray headers which spray low temperature 8

borated water into the reactor building to cool it.

Each of these 9

systems without the other has the heat removal capability to main-10 tain the reactor building pressure below its design pressure.(36) 11 Control of fission products following a loss-of-coolant acci-12' dent is provided by the reactor building itself and by a second 13 separate engineered safety feature for limiting release of fission 14

- products from the reactor building. The second'means for fission 15 product control is the iodine removal spray system which utilizes 16 chemicals mixed in the reactor building spray water to absorb the 17 iodine released from the reactor coolant system and render it un-18 available -for leakage. from the reactor building. The reactor 19 building and the iodine removal chemical spray system will limit 20 radiation doses at the exclusion radius and low population zone boundary to values within the 10 CFR 100 guideline values.I3 }

21 22 3.5 Instrumentation and Control

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A complete and dependable network of instrumentation and con-24 trols will be provided to ensure the safe operation of Rancho Seco 16.

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1 Nuclear Generating Station. The reactor protective system monitors 2

parameters related to safe operation and shuts down the reactor if 3

an operating lbnit is reached.(38) 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. Alarms (39) are pro-6 vided to alert the operator of abnormal operating conditions, and 7

interlocks (40) are provided to prevent abnormal operations which

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9 The nuclear instrumentation system monitors reactor power from 10 startup level through 125 percent of full power. There are separate, 11 overlapping instrumentation channels for the startup power range, 12 the intermediate approach to power range, and the power operation 13 range.(41) A control ~ system automatically monitors reactor system 14 conditions and the load requirements on the turbine-generator unit, 15 and adjusts reactor power, steam generator feedwater flow and the 16 turbine throttle for safe, efficient operation.(' )

17 The engineered safeguards protective system monitors plant

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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 start-22-up, operate and shutdown the nuclear unit will be placed in the 23 centrally. located control room. There will be sufficient information l

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display and alarm monitoring to ensure safe and reliable operation 2

under normal and accident conditions.

'3 3.6 Electrical Systems 4

The design of the electrical systems for the Rancho Seco 5

Nuclear Geserating Station is based on providing the required elec-6 trical equipment and power sources. to ensure safe, reliable opera-7 tion and safe, orderly ' shutdown of the unit under all normal and 8

emergency conditions. Three sources of power, each possessing 9

various degrees of redundancy are available to ensure a supply of 10 electrical energy to the station safety systems under all accident 11 conditions, including the loss-of-coolant accident as outlined 12 below:

Five 230-kv transmission lines can supply power for the 13 a.

14 unit auxiliary load dhrough the start-up transformers 15 connected to the 230-kv switchyard. Each start-up trans-16 former will be arranged to provide a source of power to 17 the nuclear service buses.

.18 b.

The main generator will continue to supply the station 19 auxiliary load upon abrupt separation from the rest of ths 20 230-kv system.

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Upon loss of all sources of power described in (a) and (b) 22 above, power will be supplied from two automatic, fast 23 start-up diesel engine generators. These are sized so that r

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1 either can carry the required engineered safeguards load.

2 The unit will generate. electric power at 22 kv, which will be 3

fed through an isolated phase bus to the unit main transformer 4

where it will be stepped up to 230-kv cransmission voltage and 5

delivered to the switchyard. The 230-kv switchyard, in turn, is 6

linked to the existing transmission network by five 230-kv circuits 7

using two different routes. SKUD's transmission system is fully 8

integrated with the Northern California network.

9 3.7 Auxiliary Systems 10 Auxiliary systems are provided to supply reactor coolant make-11 up and seal water, to cool the reactor during shutdown, to cool 12 components, te ventilate station spaces, to handle fuel, and to 13 cool spent fuel.

14 Reactor coolant makeup and seal water is suppled by the make-15 up and purification system. This system, which also serves the 16 engineered safeguards function of providing high pressure emergency 17 core coolant, maintains the proper coolant inventory in the primary 18 system, maintains the seal water flow, adjusts the concentration 19 of dissolved neutron absorber in the reactor coolant, and maintains proper water chemistry.(

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The decay heat removal system cools the reactor when the 2

reactor system is depressurized for maintenance or refueling. This 3

same system serves the engineered safeguards functions of providing 4

low pressure emergency core coolant and of recirculating borated 5

water to cool' the core in the event of a loss-of-coolant accident.(45) 6 The chemical addition and sampling system adds boric acid to 7

the reactor coolant system for reactivity control, potassium hydrox-8 ide for pH control, and hydrogen and hydrazine for oxygen control.

9 This system is also used to take reactor coolant and steam generator 10 water samples.(46) 11 The cooling water systems maintain temperatures throughout the 12 equipment and structures of the station.I47) Appropriate normal 13 ventilation systems are provided in the station.(48) 14-A fuel handling system (49) provides the means for safe, 15 reliable handling of fuel from the time it enters the station as new 16 fuel until it is shipped from the station as used fuel. Irradiated 17 fuel is handled under water at all times until after it is placed 18 into a shipping cask. The water provides a radiation shield as 19 well as a reliable source of cooling for the irradiated fuel assem-20 blies. A spent fuel cooling system maintains the temperature and 21 purity of the spent fuel storage pool water within acceptable limits.(50) i 20

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3.8 Steam and Power Conversion System 2

The steam end power c6nversion system is designed to recove 3

the heat energy generated in the reactor core by producing steam 4

in the two steam generators. This heat energy is converted to 5

electrical energy by the turbine-generator. Cooling towers will 6

dissipate the thermal energy rejected by the turbine condenser.

7 This cycle, including the necessary equipment to achieve safe 8

and reliable operation, is similar in concept and design to turbine-9 generator cycles in successful use for many years.

10 3.9 Radioactivity Control Systems 11 Radioactive gaseous, liquid, and solid wastes in the station 12 are handled by the waste. disposal systems. These systems contain 13 the equipment necessary to safely collect, process, and prepare 14 for disposal the radioactive wastes which result from reactor 15 operation. These systems are designed to minimize the release of 16 radioactive material from the station to the environment and will 17 maintain releases below the limits of 10 CFR 20. No radioactive 18 liquid waste will be released to the environment.

19 A process radiation monitoring system monitors effluent 20 released to the environment and provides an early warning of pos-21' sible equipment. malfunction or potential radiological hazard. The 22 radiation monitoring system includes a combination of' continuous-

. 23 ' automatic-monitoring and periodic sampling.

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. Shielding throughout the' station ensures that radiation doses 2

to the general public.and to operating personnel during normal 3

operation are well within the limits of 10 CFR 20.

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SAFETY ANALYSES 2

Potential malfunctions or equipment failures have been ana-3 lyzed to provide a safety evaluation of the Rancho Seco Nuclear 4

Generating Station..This evaluation demonstrates that the public e

5 will not be exposed to radiation in excess of the limits established 6

in the-AEC's regulation for siting requirements, 10 CFR 100, even 7

in the very unlikely event that one of the accidents postulated in 8

the Application should occur.(51) 9 Two categories of malfunctions or equipment failures have been 10 analyzed: those in which the core and coolant boundaries are pro-11 tected, and those in which one of these boundaries is not effective 12-and standby safeguards are required. The core and coolant boundary 13 protection analysis shows that in the event any of the postulated 24 malfunctions were to occur,.the normal protection systems operate 15 to maintain the integrity of the core and 'of the coolant boundary.(52) 16 The standby safeguards analysis demonstrates the capability of the 17 engineered safeguards systems to assure protection of the public 18 for postulated malfunctions in which the normal protective systems may not maintain the integrity of the core and coolant boundary.($ }

19 20 These analyses show that for all credible malfunctions the radiation

~21 exposure to the general public is well below the limits prescribad 22

-in 10 CFR 100.

23 Of the postulated equipment failures, a loss-of-coolant acci-24 dent is the most severe. Emergency core cooling systems are pro-25 vided to prevent clad and fuel damage that would interfere with 23 y

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continued core cooling for reactor coolant aystem failures up to 2

and including the comple.te severance of ihe largest reactor coolant l

3 pipe. The core cooling system ensures that the core will remain in 4

place and intact.(54) The reactor building spray or emergency 5

cooling units maintain the integrity of the reactor building,(55) 6 The iodine removal sprays in conjunction with the reactor building 7

assure that the public is protected from radiation and radioactive 8

material.(56) Emergency electrical power is available on-site to 9

ensure operation of these systems even if all external sources of 10 electric power to the plant are assumed to be unavailable at the 11 time of the accident.(57) 12 Results of the safety analyses show that, even in the unlikely 13 event of a loss-of-coolant accident, no core melting will occur.(56) 14 However, in order to demonstrate that the operation of a nuclear 15 power station at the proposed site does not present any undue hazard 16 to the general public, a hypothetical accident has been analyzed in-17 volving release of 100 percent of the noble gases, 50 percent of the 18 halogens, and 1 percent of the solids in the fission product inventory.

19 The analysis evaluated both the direct radiation exposure and the 20 potential total dose to the thyroid from the inhalation of fission 21 products which are assumed to leak from the reactor building. The 22 low leakage rate of the reactor building and the iodine removal spray 23 system reduce the potential radiation dose to the thyoid to below 24 the 10 CFR 100 guidelines even in the event of such a hypothetical occurence. (58) 25 24

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1 5.

TESTS INSPECTIONS AND QUALITY CONTROL 2'

Pressure containing components of the reactor coolant system 3. will_be designed, fabricated, inspected, and tested in accordance 4

with Section III, Nuclear Vessels, of the American Society'of 5 ~ ' Mechanical Engineers Boiler and Pressure Vessel Code. The piping 6

will meet the applicable provisions of PowerLPiping USA Standards 7

and associated nuclear code cases., Non-destructive testing, 8

including radiography, ultrasonic, magnetic particle, and liquid 9

penetration examinations will be performed during fabrication of

' 10 the nuclear vessels.

11 Auxiliary systems and equipment will be designed, fabricated, 12 and tested to the appropriate provisions of recognized codes and 13 standards of organizations such as the American Society of Mechanical 14 Engineers, American Society for Testing Materials, USA Standards 15 Institute, and Institute of Electrical and Electronics Engineers.

16 A comprehensive field testing program will be conducted to 17 ensure that equipment and systems perform in accordance with design 18 criteria.

19 The reactor building will be designed and built in accordance

. 20 - with applicable portions of the Building Code Requirements for 21 Reinforced Concrete, ACI 318-63; Specification for Structural Con-22 crete for Buildings, ACI 301-66; AISC Manual of Steel Construction; 23 ASME Boiler and Pressure Vessel. code,. Sections III, VIII, and IX.

25

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1 Materials and workmanship will be inspected to ensure compliance 2

with appropriate codes, specifications, and standards. Materials 3

to be inspected and tested include concrete, liner plate, pre-4 stressing systen materials, hatches, penetrations, structural and 5

reinforcing steel.

6 The reactor building will be structurally tested at 115 percent 7

of design pressure by pneumatic test. In addition, it will be leak 8

tested to ensure compliance with a maximum allowable gross leak rate 1

9 of one' tenth of one percent by volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the design 10 pressure.. Provisions have been included for in-service pressure 11 testing of equipment and personnel hatches and other penetrations.

12 Consideration has been given ed the inspectability of. the 13 reactor coolant system in the design and arrangement of components.

14

. Access for inspection of the reactor coolant system includes access 15 for visual examination by direct or remote means.

16 SMUD's Architect / Engineer and its Construction Manager, its 17 contractors, and subcontractors will perform the necessary quality 18 assurance functions to provide a safe and reliable installation. In 19 addition, SMUD will retain independent consultants in the area of 20 quality ' assurance to monitor the adequacy of quality control pro-

'21

'cedures followed in the design, fabrication, construction, erection, i

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transportation, and testing of appropriate reactor components and 2

structures. Experienced and trained personnel qualified in all areas 3

of specialization necessary to assure conformance with high quality 4

standards in unstruction of the station vill be available.

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RESEARCH AND DEVELOPMENT PROGRAMS 2

The nuclear steam supply system for Rancho Seco is similar 3

.in concept to several projects already in operation, und t con-

. 4 struction, or:recently licensed by the Atomic Energy Commission.

5 Thel preliminary design is based on technical data which has been 6.~' developed in the ' nuclear industry and on data developed by B&W 7

which is specifically related to the Rancho Seco design. To com-8 plate the final detail design of some components additional technical 9

information will be obtained.

. 10 The following are the areas of the plant design in which addi-11 tional technical data will be developed to finalize design details.

i-12 a.

Once-Through Steam Generator 13 The design of the once-through steam generator is based on 14 experimental work on boiling heat trancier and data

15 obtained by B&W in full length model tests of the unit.

16-The testing of a prototype unit has been completed but 17 not yet documented.

It' includes performance, mechanical, vibration and blowdown tests, and control system develop-19 ment. The results have confirmed the analytical predic-20 tions of performance, and sufficient data on the perfor-21 mance and structural design has been obtained from opera-22 tion of the test models to finalize the design of the 23' steam generators. (59)- :b.

Control Rod Drive Unit

~

25 LThe design of the control rod drive mechanisms is based 26 on a principle which has been used in operating reactors 28

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and which has been extensively tested by B&W. - Test pro-2 grams have-included full scale prototype testing under 3

no-flow conditions, full scale prototype' testing at 4

~ operating conditions, and components testing. Testing of 5_

a prototype mechanism was carried out for a full-life 6

. cycle of strokes and trips, and major design parameters

'7 were confirmed. Life cycle testing is being repeatedi

'8 using a pinion gear of improved material. Data from 9

these test programs will be used in the final design of 10 the control rod, its guide structure, and'the control rod 11 drive mechanism.( }

-12 c.

In-Core Neutron Detectors 13 The performance and longevity of the self-powered detectors 14 is being demonstrated by detectors installed in the Babcock 15 and Wilcox Test Reactor and in the Big Rock. Point Nuclear 16 Power Plant.(

The: tests have demonstrated that the de-17 detectors perform successfully. Tests are being continued in order to demonstrate detector. longevity. At the present time,

)

'19 the Big Rock Point detectors have accumulated operational 20 experience equivalent to three years of operation.

21 d.

Core Thermal and Hydraulic Design 22 The PSAR as originally submitted contained,'in Section 3, 23 an evaluation of the core thermal-capability in which the

~

24 heat transfer limits were predicted based on a correlation -

25 of-experimental DNB (Departure from Nuclear-Boiling) data r

f 29 a.

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'l-developed by The Babcock & Wilcox Company. In erder to 2

completely substantiate the B&W :orrelation additional 3~

research and development data is necessary. -These re-4 quirements are described in the PSAR.(02) 5 Subsequent to submittal of the original PSAR, core thermal 6

performance was also evaluated using the W-3 correlation 7

for predicting DNB. This correlation is available in the 8

literature and has been used and found acceptable in j

9

. establishing thermal design limits for other large pressur-10 ized water reactors. The thermal evaluation using the i

.11 W-3 correlation is also presented in the PSAR and its 1

12 supplements. With the use of this correlation, vessel model 13 flow tests are necessary to substantiate operation of the 14-plant within acceptable thermal limits. Flow testing which

. 15 demonstrated acceptable flow distribution for the rated 16 power level without internal vent valves in the model has

~17 been completed. Flow testing with internal vent valves

.18 installed and with open internal vent valves must still be 19 performed.

20

- e.

Emergency Core Cooling and Internals Vent Valves 21 Analytical evaluation of the effects of blowdown forces on 22 the internals and of the ' performance of the internal vent i

23 valves installed in the core support shield to insure

'24 adequate covering of the core by e,mergency coolant is in 25 progress. A prototype of these valves will be tested to 30

~

e lI demonstrate their operating characteristics.(63) 2 f.

Fuel Failure L3 A' study, including testing, is underway to assure that 4

there are no failure uechanisms which might interfere 5

with the ability of the emergency core cooling systems 6

to' accomplish their objective. The results of the work 7

to date demonstrate the ability of the design to accommo-8 date potential fuel failure mechanisms. This work will be 9

continued to assure that. fuel rod failures will not signif-10-icantly affect the ability of the emergency core cooling system to prevent clad melting.(64) 11 12 g.

Xenon Oscillations 13 11e possibility of the occurrence of xenon oscillations 14 throughout core life is being evaluated. If it is deter-15 mined tha't'such oscillations may occur, appropriate design 16 changes to eliminate or control the oscillations will be 17 incorporated. (65) The design of a means to eliminate or 18 control such oscillations is being carried out in parallel 19 with the studiet of the possibility of such oscillations.

20 h.

Chemical Spray Additive

.21 SMUD agrees to participate in one of the two R & D programs 22 on radio iodine spray renoval system additives presently 23 reviewed by AEC.

,31

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7.

TECHNICAL QUALIFICATIONS 2

7.1 Sacramento Municipal Utility District 3

SMUD is a public agency organized pursuant to laws of the 4

State of California, it'has sole responsibility for supplying elec-S tric energy to consumers within the capital city of Sacramento and 6

surrounding area, totaling 650 square miles. Operation commenced 7

'in January 1947. SMUD's power generation facilities consist of 8

five hydro plants with a capacity of 480 megawatts. When completed 9

in 1970, SMUD's hydro power development on the upper American River f

10.. will total 630 megawatts.

11 SMUD will be responsible for the design, purchasing, construc-12 tion and operation of Rancho Seco Nuclear Generating Station, Unit 1.

13 For large hydro plant construction it has been a policy not to use 14

" turnkey" contracts, but'to retain a competent engineering firm to 15 design the plant and to manage the various contractors who construct 16 the plant. Equipment is purchased by SMUD and construction contracts 17 are let by SMUD. This practice has been successfully followed for 18 all' of the District's major generating facilities now in service or 19-planned. The design and construction of the Rancho Seco nuclear

.20'

. plant will be handled in this same manner.

'21 Experience has shown that to accomplish project construction

22. satisfactorily in the above manner, SMUD must have within its own

-23 organization an overall engineering. competence in the project field.

32

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1 SMUD's assistant chief' engineer, Mr. John Mattimoe, who is 2-experienced in both ' design and construction of large power projects 3 - and who supervises the District's major project design and con-4 struction activities, has been assigned the responsibility for this 5 - project within SMUD's organization. Under Mr. Mattimoe, a nuclear 6

plant engineering group is 'being established having competence in 7

the necessary disciplines. This will be a permanent group which 8

will be expanded as may be required to perform the same work on e

9 future plants.

10 -

SMUD's management recognizes the need for a proper understand-11 ing of the project by management, and for the appropriate buildup 12 and training of engineering and operating staff. Key management 13 personnel have been involved in the program from its inception.

- 14 The present general manager and chief engineer started investigating 15 the potential of nuclear power in 1956. Other management and engi-16 neering personnel have enrolled in nuclear engineering courses and 17 participated in various nuclear seminars. The staff currently as-18 assigned to licensing also includes a number of management personnel, 19 who have received special courses on the plant systems.

20-Recruitment and training programs have been developed to 21 assure that the project engineering staff will be prepared to han-22 ' die the design and construction program in a safe and efficient 23 manner, see PSAR Fig. 1C-1.

33 A=a ww n--eA.'

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'A properly trained and licensed force with experienced super-2.visory personnel to operate Rancho Seco is to be formed in the five-3 years between now and 1973 when -the plant is to be completed. The 4

employees required will be recruited and trained well in advance of l

l 5

the scheduled plant start-up with key personnel at the plant site 6

during most of the construction period in accordance with the 7

schedule set out in Fig. 12.3-1 of the PSAR.

8 7.2 Bechtel Corporation 9-Bechtel Corporation has been retained by SMUD as Architect /

10 Engineer and Manager of Construction for the Rancho Seco project.

11 Working closely with SMUD, Bechtel is responsible for project 12 studies and conceptual design, specification of material and ser-13 vices, project detailed design, construr: tion management and assis-

~14 tance in plant testing and start-up.

15 Bechtel Corporation has been continuously engaged in construc-16 tion or engineering activities since 1898. For the last 20 years, 17 Bechtel has been active in the fields of petroleum, power generation 18 - and distribution, harbor development, mining and metallurgy, and 19 chemical and industrial processing.

20 Since the close of World War II, Bechtel has been responsible

~

21 for the design of over 165 power generating units, representing more 22 than 38 million kilowatts of new generating capacity, which includes i

h a

34 i

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1' units of 'the largest and most modern types. Of this number, more

' 2

~ than 11 million KWe-is produced by.20 nuclear-fueled units.

3 For over 18. years, Bechtel has been engaged in the study, 4

design and construction of nuclear installations. Their experience 5

includes design or~ construct on, or both, of such facilities as d

6-

accelerators,. nuclear research laboratories, hot cells, experimental t

7.

reactors, ~ and nuclear fuel processing plants, as well as nuclear 8

Power plants. A summary of experience is listed in the Application.

9-7.3 Babcock and Wilcox Company 10 B&W's participation in the development of nuclear power dates 11' from the Manhattan Project. B&W's nuclear activities are broad and 12 include applied research to develop fundamental data; design and 13 manufacture of nuclear systems, cores, and components; and design, 14 manufacture, and erection of complete nuclear stean generating 15 systems. Through the E&W Company's several divisions, a wide range 16 of. equipment for nuclear application is designed and manufactured.

17 The B&W Company's major nuclear contracts, in addition.to a sub-4

- 18' stantial percentage of components for the nuclear Navy, have included 19 Indian Point No. 1; NS Savannah; Advanced Test Reactor; Oconee Nuclear 20 Station Units 1, 2, and 3; Three Mile Island Nuclear Station; Crystal 21-River Plant Unit 3; 'and four other units in various stages of licensing

' 22: - in addition to the Rancho Seco Nuclear Generating Station.

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~2 SMUD will retain independent consultants to monitor the quality 3

assurance programs in all phases of the project.

4 SMUD has contracted with NUS Inc. to perform specific studies 5

and checking in the nuclear area and plans to continue this practice 6

throughout the life of the project.

7 SMUD has also used'a number of specialist consultants on the 1

8 project and will continue to do so when their services are required.

9 -- These include Dr. Perry Byerly - seismologist, Dr. George Housner -

10_, seismic design, Meteorological Research Inc. - meteorology, Roger 11-Rhodes - geologist, T. Y. Lin and Associates - prestressed conc /ete i

I 12 design, Dr. A. H. Mattock - concrete design and Dr. J. L. Shapiro -

~ 13 nuclear design.

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COMMON DEFENSE AND SECURITY.

2 iThere is no indication that construction and operation of the 3

Rancho Seco Nuclear Generating Station will in any way be inimical

_'4 to the common defense and security of the United States.

As ' stated in the Application, SMUD is a California municipal

'5-6 utility with' statutory authority.for the production, transmission, 7

and sale of. electric energy. All of th'e directors and principal 8~

officers are citizens of the United States and SMUD is not owned, 9-controlled,'or dominated by an alien, a foreign corporation, or a

- 10 foreign government.

11 The Application contains no restricted or other defense infor-12 -mation and Applicant has agreed that it will not permit any individ-13 ual-'to have access to Restricted Data until the Civil Service Com-14. mission shall have'made an investigation and report to the Atomic 15 Energy Commission on the character, associations and loyalty of 16 - such individual, and-the Atomic Energy Commission shall have deter-17 mined that permitting such persons to have access to Restricted

-18 Data will not. endanger the common defense and security.

4 19

.As a licensee, Applicant will be subject to regulations of the 20 Atomic Energy'Conmission relating to the transfer of and account-4 21-ability-for special. nuclear material in its possession. Recent 22' amendments to-the AEC-Rules and Regulations (10 CFR 50.60) under J

.23 which the AEC will discontinue allocating quantities of special

'24-nuclear material to reactor licensees evidence that such material-37 a.

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' war or: national emergency the AEC may order the recapture of special-3. nuclear material, as well as the operation of any licensed facility.

4 (10 CFR 50.103)-

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CONCLUSION 2-

- On the basis of the foregoing and the Application, the 3:

Applicant respectfully submits that:

SMUD's Applicatien, as amended, describes the proposed 4

a.

5

. design of the Rancho Seco Nuclear Generating Station, Unit 6

No.1, including the' principal architectural and engineering 7

criteria for the design, and identifies the m&jor features 8

or components incorporated in the plant for the protection 9

of the health and safety of the public.

10 b.

The Application, as amended, identifies the technical and 11 design information necessary to complete the final safety 12 analysis. Such information can reasonably be lef t for 13 later consideration and will be supplied in the final 14 safety analysis report.

15 c.

Safety. features which require further research :nd develop-16 ment, and the research and development programs to be 17 carried out, are identified in Section 1.5 of the Applica-18

-tion.

The research and development program is reasonably 19 designed to resolve any questions associated with such 20 features at or before the latest date stated in the Appli-21 cation for completion of construction of the facility.

22 d.

Taking into consideration the characteristics of the site

.23-and environs and the proposed design of the Rancho Seco 24~

Nuclear Generating Station, such facility can be constructed

. 25~

and operated within the limitations established by 10 CFR 20, 39

.- ~, _ -.. ~ - - - -

T.

i I

within the site criteria set forth in 10 CFR'100,.and 2

.without' undue risk to the health and safety of the public.

3 e.

The Applicant is technically qualified to design and con-

' 4 struct the proposed-facility.

5 f.

The issuance of a construction pemit for the Rancho Seco 6

Nuclear Generating Station will not be inimical to the 7

causson defense and security of the United States or to the 8

-health and safety of the public.

J e

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i APPENDIX A LIST OF REFERENCES J

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" W m4 APPENDIX A LIST OF REFERENCES 1.PSAR, Volume I, Section 2.2.1

. 2 PSAR, Volume I, Section 2.2.5.1 3

PSAR, Volme V, Appendix 14A, Question 14.A18 4

PSAR, Volume I, Section 2.2.4 5

PSAR, Volume IV, Appendix 2A 6

PSAR, Volume I, Section 2.3 7

PSAR, Volume I,'Section 2.4.1 8

PSAR, Volume IV, Appendix 2C 9 - PSAR, Volume I, Section 2.4.6.4

- 10 PSAR, Volume I, Section 2.5 11 PSAR, Volme I, Section 2.6 12 PSAR, Voltane IV, Appendix 2D 13 PSAR,' Volume V, Appendix 5A 14 PSAR, Volume IV, Appendix 2H, Question 2H.3 15 PSAR, Voline I,'Section 2.8 16 PSAR, Volume I, Section 1.2.2 17 PSAR, Volume I,' Table 1.3-1 18 PSAR, Volume'I, Table 3.2-2

- 19 PSAR, Volume I, Table 3.2-1 20

'PSAR,: Volume 1, Section 3.2.3.

21~

PSAR,.Voltane I, Section 3.2.2.1.2 and Volume II, Section 7.2.2.1 221 PSAR, Volume I, Table 3.2-6 and Volume I, Figure 3.2-1 23' PSAR, Volume I, Section 3.2.2.1.3 4

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~ 24 PSAR, Volm e I, Section 3.2.4.3.2 25 PSAR, Volme II, Section 9.2 26 PSAR, Volume II, Section 4.2.2.1 27

. PSAR, Volume II, Section 4.1.4.4 28 PSAR, Volume II, Section 4.4.3 29 PSAR, Volume II, Section 4.2.5

+

30 PSAR, Volume II, Section 4.2.2.4 31 PSAR, Volume II, Section 4.2.2.3 32 PSAR,~ Volume II, Section 4.2.2.2 33

.PSAR, Volume II, Section 5.5.1.2 34 PSAR, Volume II, Section 6 35 PSAR, Volume II, Section 6.1 36 PSAR, Volume II, Section 6.2 37 PSAR, Volume III, Section 14.3.9 38 PSAR, Volume II, Section 7.1

' 39 PSAR, Volume-II, Section 7.4.3 40-PSAR, Volume II, Section 7.2.3.2 41 PSAR, Volume II, Section 7.3.1 42

'PSAR, Volume II, Section 7.2 43 PSAR, Volume II, Section 7.1.2.2 44 PSAR, Volune II, Section 9.1 45 PSAR, Volume II, Section 9.5 46 PSAR, Volume II, Section 9.2 47 PSAR,-Volume II, Section 9.3 48

-PSAR, Volume II, Section 9.7 9

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49 PSAR,' Voltane II, Section 9.6 50 PSAR, Voltane II, Section 9.4

-51 PSAR, Voltane III, Section 14 52 PSAR, Voltune III, Section 14.1 53 PSAR, 'Voltane III, Section 14.2 54-PSAR, Voltar.e II, Section 6.1 55 PSAR,.Voltane II, Section 6.2

.56 PSAR, Voltane III, Section 14.2.2.3 57 PSAR, Voltane II, Section 8.2.3 58 PSAR, Voltane III, Section 14.3.8 59 PSAR, Voltane I, Section 1.5.7 60 PSAR, Voltane I, Section 1.5.6 61 PSAR, Voltane I, Section 1.5.8 62 PSAR, Voltane I, Section 1.5.2-63 PSAR,- Voltane I, S'ections 1.5.5 and 1.5.9

- 64' PSAR, Voltune I, Section 1.5.3 65 PSAR, -Voltane I, Section 1.5.1 66 PSAR, Voltane I, Section 1.5.10 4

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O APPENDIX C QUALIFICATIONS OF WITNESSES e

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. l' EDUCATIONAL AND PROFESSIONAL QUALIFICATIONS

'2-JOHN J. MATTIMOE

'3

-ASSISTANT CHIEF ENGINEER 4

SACRAMENTO MUNICIPAL UTILITY DISTRICT 5

1.

My r.ame is' John J. Mattimoe. My residence is 2057 Wakefield 6

'Way, Sacramento,' California. I am employed by the Sacramento 7

Municipal Utility District.as Assistant Chief Engineer. I 8

am also' the Project Manager for the District's Rancho Seco 9

Nuclear Power Generating Station.

'10 2.

I served in the U. S. Army Corps of Engineers from March 1943

.11-through September 1946, being discharged as a First Lieutenant.

12 3.

-I graduated from Stanford University in 1947 with a Bachelor l

13 of Science degree in civil engineering.

.14 4.

- From-1947 through 1956 I. was employed by the Pacific Gas and t

]

15 Electric Company in their General Construction Department. I 16 had responsibilities in the construction of their major hydro-17'

' electric developments on the Feather River.

18

.5.

In 1956 I joined the Sacramento Municipal Utility District and -

have had responsibility for all new major construction. In 20 June 1958 I was assigned to the District's Upper American 21

_ River Project, and in August 1962, I became the Assistant. Project Engineer on the District's hydroelectric project.

a a

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1-6.

In-January 1963 I became Manager of the Hydroelectric Construc-12 tion Department, and in February 1966 I became Manager of the L3L Hydro and General' Construction Department..

tr 17.

In' November 1967 I was appointed.co the' position of Assistant 5

Chief Engineer, having responsibility for all engineering, all -

6

. major construction, and all nuclear power construction.

7 8.

LI have been a Registered Professional Engineer in the State of

-8 California since 1958 and I an a member of the American Society 9

of Civil Engineers.

~

e I

i i

1 7

LC-2J

l 1e

. EDUCATION AND PROFESSIONAL QUALIFICATIONS -

DALLAS G, RAASCH

'3

-PROJECT ENGINEER,. RANCHO'SECO NUCLEAR G2NERATING STATION 4

' SACRAMENTO MUNICIPAL UTILITY DISTRICT-

=5-1.

My name is Dallas G. Raasch. My residence is 3929 Orangewood 6

Drive, Fair Oaks, California.

I am employed by Sacramento 7

Municipal Utility District as Project Engineer on the Rancho A

'8 Seco Nuclear Generating Station.

9 2.'

In this position I am responsible for assisting in the 10

. licensing effort and coordination of the technical aspects

' 11 of the Project.

12

-3.

I graduated from the Univeristy of Minnesota with a Bachelor 13 degree in Electrical Engineering in 1950. I-have also taken 14 post-graduate cources in nuclear physics, nuclear engineering, l

15

. controls systems and industrial management.

16

.4.

In 1950 I joined the Helix Irrigation District as an Electrical 17:

Engineer and subsequently became Superintendent of Electrical 18

. and Pumping, responsible for the design, construction and 19

. operation of all pumping stations, corrosion control systems 20 and communication' systems.

21 2 5.:

In 1957 I. joined the-Atomic Power Division of Westinghouse

~

-22T Electric ~ Corporation as an Instrument Engineer and partici-

'pate'd in the training of instrument technicians and construc-23I

~ 24 tion activitfes.in the Naval AIW Program at cl$e National-125

-Reactor Test Station in Idaho. In 1959 as a Senior Engineer,

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l-tI'was'in_ charge'of all test instrumentation at the AIW 2

. Facility and also served as a Lead Engineer in the startup

~

.3

~ activities for'the AIW reactors.

4-6.

In 1961 I joined the Aerojet-General _ Corporation as a Design 5

EEngineer in _ the ~ Solid Propellant Division in Sacramento.

In

.6 1962 I became Supervisor of Process Instrumentation, respon-7-

sible for the design, construction follow and startup of 8

several major production facilities in the Solids Rockets 9

. Division. In 1964 as Supervisor of_the Standards Laboratory, 10 I was responsible for the calibration of all equipment used 11-in the Quality Control programs throughout the Sacramento plant.

12 7.

In lyoS I joined the Sacramento Municipal Utility District as 13 an Electrical-Engineer and was assigned to the Hydro-Generation -

14 Division to resolve the technical problems on the existing 15 facilities, and to review the design of all new facilities to 16 improve the reliability of the hydro system.

17

8..

I an a member of the Institute of Electrical and Electronic

'18 Engineers and the Instrument Society of America.

19 9.-.I am a registered Professional Engineer in the State of

~ 20' '

' California.--

r

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r

1 EDUCATIONAL AND PROFESSIONAL QUALIFICATIONS 2

RONALD C. STINSON, JR.

3 NUCLEAR PLANT SUPERINIENDENT 4

SACRAMENTO MUNICIPAL UTILITY DISTRICT 5

1.

My name-.is. Ronald C. Stinson, Jr.

I reside at 9444 Durango 6'

Way, Elk Grove, California, and am employed by The Sacramento 7

Municipal Utility District as Nuclear Plant Superintendent.

8 2.

I received a Bachelor of Science Degree in 1953 and an tis 9

(Nuclear Engineering) in 1961, both from Texas A&M University.

)

i 10 3.

From 1953 to 1958, I was on active duty in the U. S. Army.

11 During that period I was involved in the rocket and nuclear 12 weapons programs.

i

-13 4.

From 1958 to 1961, I.was a graduate student at Texas A&M.

14 During this period I held vartaus positions at the Nuclear 15 Science Center, the AGN 201 Reactor Lab, and the Hot Lab.

16

'5.

From 1961 to 1964, I was employed by the Gener:1 Electric 17 Company at.Hanford, Washington. During this period I held 18 various positions including Shift Supervisor - Reactor Opera-19 tions, _ Training Specialist, and Reactor Analys :.

20 6.

From 1964.to.1966, I was employed by the Geneial Electric 21 ~-

Company at Vallecitos Atomic Laboratory as Development Engineer 22 And Manager of Nuclear Safety Compliance. In the latter posi-23~

tion,:I.was Manager of a unit which had responsibility to i

C-5'

1 review and approve all experiments and design changes, and to 2

maintain all relationships with the AEC and other regulatory 3

agencies.

I was a member of the Laboratory Safeguards Group.

4 7.

From 1966 to 1968, I was employed by the General Electric 5

Company, Atomic Power Equipment Department as Project Engineer 6

for Dresden 2, 3, and Quad Cities 1 and 2 with prime project 7.

responsibilities-for the Nuclear Steam Supply Systems.

8 8.-

Since February 1968 I have been employed by the Sacramento 9

Municipal Utility District as Nuclear Plant Superintendent.

10 -

9.

.I su a member of the American Nuclear Society.

i C-6 l

-1 EDUCATIONAL AND PROFESSIONAL QUALIFICATIONS JAY R. McENTEE 2

3 MANAGER OF NUCLEAR AND DESALTING ENGINEERING 4

.VERNON DIVISION 5

BECHTEL CORPORATION 6

1.

My name is Jay R. McEntee.

I reside at 5338 Barrett Circle, 7

Buena Park, California. I am employed by the B tchtel Corpora-8 tion as Manager of Nuclear and Desalting Engineering.'.

9 2.

In this position, I am responsible for engineering management 10 of nuclear and desalting projects pndertaken by the Vernon 11 Division of Bechtel Corporation.

12 3.-

I graduated from Cornell University, Ithaca, New York in 1948 13 with a Bachelor of Science Degree in Mechanical Engineering.

14 4.

Between 1948 and 1952, I was employed in the Research and 15 Development Department of a refrigerator manufacturer and 16 later as a field engineering supervisor for a manufacturer of 17 flue gas heat-recovery equipment.

18 5.

I joined the Bechtel Corporation in 1952 and have held assign-19 ments as Engineer, Group Leader, Project Mechanical Engineer, 20 and Project Engineer for the design of various major fossil-21 fueled central power stations for California and Arizona

~

22 utilities.

C-7

. ~ _.

'1-6.

In 1962 I was assigned as Project Engineer in responsible 2

charge of Bechtel's design of the San Onofre Nuclear Generating 3

Station. This assigtunent continued until' my present assign-4 ment, starting late in 1966.

~

5 7.'

-I am registered ~as a Professional Engineer in the~ State of 6

California. I an a m' ember of the American Society of 7

Mechanical Engineers and the American Nuclear Society.

C-8 3

~

1 EDUCATIONAL AND PROFESSIONAL QUALIFICATIONS 2

GEORGE S. C. WANG 3

CHIEF NUCLEAR ENGINEER 4

VERNON DIVISION-5.

BECHTEL CORPORATION 6

1.

My name is George S. C. Wang. My residence is at 1900 7

Calle Alegria, Fullerton, California 92633.

I am employed 8

by Bechtel Corporation, Vernon Division.

I am Bechtel 9

Corporation's Vernon Division Chief Nuclear Engineer.

10 2.

-In this position I am responsible for nuclear design and 11 licensing efforts in Bechtel Corporation, Vernon Division.

12 3.

In 1957, I graduated from the University of Michigan with a 13 Bachelor of Science Degree in Civil Engineering.

In 1959 I 14 received a Masters of Science Degree in Civil Engineering and a Masters of Science Degree in Nuclear Engineering.

15 4.

Since joining Bechtel Corporation in 1962, I have worked on 16 every nuclear project in the Vernon Division of Bechtel 17 Corporation. These projects include the San Onofre Nuclear 18 Generating Station project, the Metropolitan Water District 19 Dual Purpose Power and Desalting project, the AEC SNAP-8 Flight 20 Configuration Test Facility Project, the Advanced Nuclear 21 Reactor study, and the GPU/AI-Fast' Breeder Power Plant project.

22 5.

I am a member of American Nuclear Society and American 23 Society of Civil Engineers.

C-9

.i.-

.1 EDUCATIONAL AND PROFESSIONAL QUALIFICATIONS 2

ROY A. NORRY 3

PROJECT ENGINEER, NUCLEAR POWER'AND DESALTING GROUP i

4-VERNON DIVISION 5

BECHTEL CORPORATION 6

1.

1My name is Roy A. Norry. My residence is at 32 Esperanza, 7

Sierra Madre, California 91024.

I am employed by Bechtel 8

-Corporation in the Nuclear Power and Desalting group.

I am 2

9 Bechtel. Corporation's Project Engineer for Sacramento 10 Municipal Utility District's Rancho deco Nuclear Generating 11-Station.

12 2.

In this position I am responsible for the overall coordination

. 13 aof the' engineering phases and licensing effort associated 14 with.those areas of the design for which Bechtel Corporation i.

15 is responsible.

16 3.

I graduated from the University of Wales'in 1946 with a 17 Bachelor of Science degree in Electrical Engineering.

18 4

Between 1953 and 1960'I worked-on the design of thermal power

.19 plants in England and Canada.

20-5.

I joined Bechtel Corporation in 1960 as an electrical engineer 21 and subsequently became project electrical engineer on the 22 Etiwanda project-(2 x 350RMW thermal) and San Onofre (450 MW

. 23

' nuclear).

24 6.

I-am a registered. professional engineer in the State of Cali-

.25 fornia.

'C-10~

t '

n~

n

-v-,

1 7.

I am a member of the American Nuclear Society, and the 2

Institute of Electrical and Electronic Engineers.

1 l

i C-11

1 EDUCATIONAL AND PROFESSIONAL QUALIFICATIONS 2

WILLIAM G. BINGHAM, JR.

3 PROJECT NUCLEAR ENGINEER, NUCLEAR POWER AND DESALTING GROUP 4

VERNON DIVISION 5

BECHTEL CORPORATION 6

1.

My name is William G. Bingham, Jr.

I reside at 5192 McComber 7

Road, Buena Park, California 90620.

I am employed by the 8

Bechtel Corporation in the position of Project Nuclear 9

Engineer.

10 2.

In this position, I am' responsible for nuclear design and 11 licensing efforts in Bechtel Corporation for the Sacramento 12 Municipal Utility District's Rancho Seco Nuclear Generating 13 Station.

14.

3.

I graduated from the University of California at Los Angeles, 15 California in 1957 with a Bachelor of Science Degree in 16 Engineering.

17 4

I joined Bechtel Corporation in 1957 and held assignments 18 as Engineer for the design of various fossil-fueled central 19 power stations.

-20 5.

In 1961 I was assigned as nuclear engineer on the SNAP-8 21 Flight Configuration Test Facility and subsequently as 22 nuclear engineer on the San Onofre Nuclear Generating Station 23-where'I was responsible for shielding of reactor and associ-24

. ated equipment and various phases of safety analysis.

i C-12

. - ~...

" a4 r

a

1..'

'l-6.

I am registered as a Professional Engineer in'the State of 2

California.

I am a member of the American Nuclear Society 3

and~the American Society of Civil Engineers.

l

)

4 l

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l 1

EDUCATIONAL AND PROFESSIONAL QUALIFICATIONS 2

' ROBERT T. SCHOMER

-3.

PROJECT MANAGER, NUCLEAR POWER GENERATION DEPARTMENT 4

POWER GENERATION DIVISION 5

THE BABCOCK'& WILCOX COMP /.NY 6

1.

My name is Robert T. Schomer. My residence address is 3416 7

Landon Street, Lynchburg, Virginia, 24503.

I am employed by 8

The Babcock & Wilcox Company,-Power Generation Division, 9

Nuclear Power Generation Department, as a Project Manager.

l 10 2.

I served in the U.S. Navy Reserve from June J43 through 11 January 1946,' Lt(jg) rank - honorable discharge.

\\

12 3.

I was graduated from the Stevens Institute of Technology in 13 1940 with a Degree in Mechanical Engineering.

I obtained ea 14_

MSME from Stevens Institute _of Technology in 1949.

I attended 15 Oak' Ridge School of Reactor Technology 1950-1951, 16 4

In 1948 I started my nuclear energy experience by working 17 for Gibbs & Cox, Inc., on the design and development of 18 nuclear submarines.

19 5.

In 1953 I joined the Babcock & Wilcox Company in the Atomic 20 Energy Division as an application engineer.

21 6.

From 1955-1963 I was a project manager on various projects

-22' including the Liquid Metal Fuel Reactor, the N. S. Savannah 23 Upgrading Program and the German Nuclear Ship project.

C-14

0:

1 7.

From 1963-1967.I was Development Programs Manager responsible 2:

for the Division's R&D Programs.

3 8.

In 1967.I was transferred to the Nuclear Power Generation 4

Department of the Boiler Division as a Project Manager in

~ 5 charge of the Rancho Seco Nuclear Generating Station, Unit 6

.No. 1, for the Sacramento Municipal ~ Utility District.

l i

i C-15' s

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a e

1

.FDUCATIONAL AND PROFESSIONAL QUALIFICATIONS 2-

' ROBERT E. WASCHER 3

. MANAGER, NUCLEAR SAFETY ENGINEERING SECTION 4

. NUCLEAR POWER GENERATION DEPARTMENT.

5 POWER GENERATION DIVISION

-6 THE BABCOCK & WILCOX COMPANY

. 7-

. 1.

My name is Robert E. Wascher. My residence is 1916 Eastwood 8

Lane, Lynchburg, Virginia, 24503.

I am employed by The 9

Babcock & Wilcox Company, Power Generation Division, in the 10 Nuclear.. 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

' 14 Technology.

15 3.

Upon graduation I joined the Oak Ridge -National Laboratory 16 as an Associate Development Engineer responsible for the 17 development of mechanical components for homogeneous nuclear 18 reactors.

19 4

In 1955 I was commissioned an officer in the U.S. Navy.

20 During my naval!-service, I' served as the Navy Liaison Officer

'21

.in'the Army Package Power Reactor Program.

I was also 22 assigned to the Navy's Bureau of Yards and Docks with responsi-23 bility for nuclear engineering problems of the Bureau.

~

In 1958 I joined The Babcock & Wilcox Company as-a Nuclear

- 24 5.-

25

. Engineer with responsibility for the safety analysis of the 26~

' Consolidated Edison Company's Indian Point No. 1 Nuclear Plant.

i 1C-16

.wa

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.~..

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.a.-.

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1

~In.1959 I was appointed Supervisor of the Safety Analysis 2.

Group with responsibility for safety analysis of nuclear 3-plants designed by B&W.

In-1964 I became Chief of the Opera-4 tional Analysis Section with responsibility for reactor and 5

system dynamic analysis, reactor control analysis, plant 6

performance, and 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 11 Safety Committee, a committee responsible for periedic review 12 of the operation of the N.S. Savannah.

From 1962 to 1966 I 13 was also Chairman of B&W's Nuclear Development Center Safety 14 Review Board. -In 1966, I was appointed to the Atomic Energy 15 Commission's Advisory Task Force on Power Reactor Emergency 16 Cooling.

In addition, I am a member of the American Nuclear 17 Society: and the Atcaic Industrial Forum's Safety Committee.

18-8.

I am-a. registered Professional Engineer in the State of 19

Virginia, i

l C-17

_p 2 :=:..z.

= :. = - =.

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' EDUCATIONAL AND PROFESSIONAL. QUALIFICATIONS-

' JAMES M. CUTCHIN, IV

'3 ~

SENIOR ENGINEER, LICENSING GROUP.

4.

NUCLEAR SAFETY ENGINEERING SECTION 5

NUCLEAR POWER GENERATION DEPARTMENT

6 POWER GENERATION DIVISION 7

THE BABCOCK & WILCOX COMPANY

-81 1...

My name is James M.' Cutchin, IV. My home address is 6028 4

9

Rhonda Road,' Lynchburg, Virginia.

I am employed by the 10 Babcock & Wilcox company in the Nuclear Power Generation a

11 Department of their Power Generation Division as'a Senior 12-

. Engineer.

13 2.

I received a' Bachelor of Science Degree from the United States 14 Military Academy in 1955. After serving over four and a half 15 years in the United States Air Force, I was discharged as a 16 Regular First Lieutenant and Instructor Navigator.

I entered 17 North Carolina State University and received a Master of 18 Science Degree in Mechanical Engineering in 1962.

19 3.

From June 1962 to June 1964, I was an Associate Engineer in the 20 Thermal Analysis Section of the Atomic Energy Division of The 21 Babcock - &f Wilcox ' Company.

I performed thermal and hydraulic 22.

analyses of various components for the Advanced Test Reactor.

23.

During the first half of 1964, I served as the Company's

--24 representative in the area of core thermal and hydraulic 25

~ design to the' German Consortium designing the reactor for the 26

' nuclear ship, OTTO HAHN.

J C-18.

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.1 4.

From June 1964 to October 1965, I was an Engineer in the 2

Division's Preliminary Design Section where I was responsible 3

for the thermal and hydraulic design and analysis of cores for 4

various nuclear reactor concepts.

5 5.

In October 1965, I was assigned to the Thermal Analysis Group 6

of the Division's Core Engineering Section as Lead Engineer 7

for the daermal and hydraulic design of cores for Central 8.

Station reactors.

9 6.

From July 1966 to October 1967, as a Senior Engineer in the 10 Thermal Analysis Group of the Nuclear Power Generation Depart-11 ment's Reactor Engineering Section, I participated in the 12 thermal and hydraulic design and analysis of the cores and 13 other components for Duke Power Company's Oconee reactors and i

14 Metropolitan -Edison Company's Three Mile Island reactor.

15 7.

Since October 1967, as a Senior Engineer in the Licensing Group 16 of the Nuclear Safety Engineering Section, I have been involved i

17 in the licensing activities associated with Metropolitan 18 Edison Company's Three Mile Island Nuclear Statica and 19 Sacramento Municipal Utility District's Rancho Seco Nuclear i

20 Generating Station.

21 8..

. I am a member of the American Society of Mechanical Engineers,

~

22 the American Nuclear Society, the Society of the Sigma Xi and 23 the Tau beta Pi Association.

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