ML19329E848

Summary of Application for CP

ML19329E848
Person / Time
Site:	Midland
Issue date:	10/30/1970
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
References
NUDOCS 8006180644
Download: ML19329E848 (122)

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, In the Matter of ) -

i CONSUMERS POWER COMPANY Docket No. 50-329 ,C Docket No. 50-330 .

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SUMMARY

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l THE ATTACHED FILES ARE OFFICIAL RECORDS OF g DIVISION OF DOCUMENT CONTROL. THEY HAVE BEEN gh hh ggi COMT CHARGED TO YOU FOR A LIMITED TIME PERIOD AND g MUST BE RETURNED TO THE RECORDS FACILITY R@h l BRANCH 016_.

CHAR PLEASE DO NOT EO OUT THROUGH THE M AIL. REMOVAL OF ANY SEND DOCUMENT]

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DEADLINE RETURN DATE

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October 30, 1970 _

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,l TABLE OF CONTETS l

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2.0 BACKGROUND

l 2.1 Consumers Power Company 2.2 Applicant's Participation in the Development of Nuclear Power 23 Need for This Plant 2.4 Considerations Entering Into Selection of This Plant 25 Process Steam To Be Supplied to Dov Chemical Company 30 THE EVOLUTION OF DESIGN OF THE MIDIAND FACILITY 31 Reactors 3 1.1 Shippingport - The First Commercial PWR ,

3 1.2 Indian Point I - An Early Large Commercial PWR  !

I 313 Connecticut-Yankee - A Subsequent Large PWR 3 1.4 Robert Emmett Ginna - A Large PWR Which Recently Began Operation 315 Oconee 1 - A Modern Large PWR 3 1.6 Midland Units 1 and a 32 Evolution of Engineered Safeguards 33 Reactor Buildings 4.0 ENVIRONMENTAL QUALITY 4.1 Introduction 4.2 Air Quality h.3 Water Quality 4.4 Solid Wastes h.5 Iand Use and Aesthetics 4.6 National Environmental Policy Act 50 OTHER GOVERNMENTAL AGENCIES 6.0 SITE CHARACTERISTICS 6.1 Location 6.2 Population 6.3 Land Use 6.4 Meteorology 6.5 Surface Water Hydrology 6.6 Groundwater Hydrology 6.7 Site Geology 671 General 6.72 Salt Solution Cavities 6.8 Seismology 69 Foundations 70 DESCRIPTION OF MIDLAND PIANT 71 Introduction

,, 72 Reactor and Reactor Coolant System

( 73 Reactor Building 74 Engineered Safety Features 75 Instrumentation and Control i

O . TABLE OF CONTENTS (Contd) 76 ~~ ~ frical Systems Elec T.7 Auxiliary Systems T.8 Steam and Power Conversion Systems 79 Process Steam System 7 10 Cooling Pond 7 11 Radioactive Waste System T.12 Shielding 8.0 SAFETY ANALYSIS -

90 COMPARISON WITH OTHER PWR 10.0 ENVIRONMENTAL MONI'IORING PROGRAM 10.1 General Radiological Surveillance 10.2 Aquatic Ecological Surveillance 10 3 Aquatic Radiological Surveillance 11.0 TECHNICAL QUALIFICATIONS 11.1 Consumers Power Company 11.2 Bechtel 11 3 The Babcock & Wilcox Company 12.0 QUALITY ASSURANCE AND QUALITY CONTROL 12.1 Introduction 12.2 Consumers Power Company 12 3 Bechtel 12.4 The Babcock & Wilcox Company 13 0 COMMENTS ON ACRS REPORTS 14.0 FINANCINL C2.LIFICATIONS 15 0 COMMON DEFENCE AND__ SECURITY

16.0 CONCLUSION

APPENDICES A COMPARISON OF PWR DESIGN PARAMETERS B FIGURES C RESEARCH AND DEVEIOPLENT PROGRAMS OF INTEREST TO ACRS FIRST IDENTIFIED IN EARLIER CASES k.

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1.0 INTRODUCTION

2 This document, prepared and filed by Consumers Power Company 3 (hereinafter sometimes called " applicant"),, is a sum =ary of the 4' application filed by the Company for permits from the Atomic Energy 5 Commission to construct a two unit nuclear. power plant, on a site 6 in Midland Township, Midland County, Michigan.

7 This document has been prepared in accordance with the provisions 8 of 10 CFR, Part 2, Appendix A,Section II (e), as a-summary of the 9 Company's

• application, including a summary description of the re-10 actors, an evaluation of the considerations important to safety and a 11 comparison of the proposed reactor design with the design of the re-12 actors previously approved or built. To assist interested members of

( - 13 g the public in understanding the complex technical material dealt with 14 in this report,-applicant has endeavored to use nontechnical language 15 whenever possible.

16 To enhance the usefulness of this summary, applicant has included 17 within it summaries of reports rendered by the Advisory Committee on 18' Reactor Saf9 guards on the application pursuant to Section 29 of the 19

~ Atomic Energy Act of 1954 as amended (42 USC Section 2039), and infor-20 mation incorporated in the docket in accordance with the National Enviwn-21 mental Policy Act of 1969, P.L.91-190.

• The application and other pertinent documents currently in the record of this proceeding may be examined at the Grace Dow Memorial Library, 1710 W St Andrews Road, Midland, Michigan.

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'l In addition to a summary of the application, this document contains

'2 a summary of applicant's program to consult with and obtain coc=ents of 3 interested Federal, state and local government officials in planning for 4 the Midland Plant. In cases where applicant is required t'o obtain 5 state and local approvals or. permits in connection with the design or 6 construction of the plant, or approvals from other Federal agencies, this 7 sumary describes the status of applicant's efforts to obtain such approvals.

8 Some of the information in this su==ary with regard to applicant's 9 programs to consult with or obtain regulatory approvals of government agen-10 cies other than AEC is not in the record of this proceeding, or is based 11 on more recent information. Applicant is prepared to establich such in-12 formation by appropriate evidence at the hearing to whatever extent is

, 13 determined to be appropriate in the course of this proceeding.

14 Applicant is prepared also to substantiate any other information 15 summarized in this sumary, or in the application as amended, by appro-16' priate testimony in the hearing as may be appropriate for determination 17 of any issues before the Atomic Safety and Licensing Board.

18 As set forth more fully in the ensuing sections of this succary, 19 two nuclear units and appurtenant facilities, which have been designated 20 as the Midland Plant (hereinafter sometimes called "the plant"), will 21 occupy an 1190 38-acre site on the southerly shore of the Tittabavassee 22 River opposite the industrial complex of The Dow Chemical Company at 23 Midland. The units are presently scheduled to begin ce=mercial operation 24 in November 1974 and November 1975 Both units are conventional pressurized

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1 vater reactors (PWR) fesigned together to generate about 1,300,000 kw 2 of electricity. They are similar to many other pressurized water 3 reactors which are in operation or under construction.

4 The plant will provide a base load electrical capacity which will 5' be needed by the applicant and by the Michigan Power Pool to meet pro-6 Jected electricity requirements. The plant will also provide the process 7 steam requirements of The Dow Chemical Company.

8 The plant is being designed and all planning is being carried out, l 9 with preservation and enhancement of the environment as important con-lo siderations. Assurance of safety has been of paramount consideration 11 in selecting a design for the Midland Plant. The plant is being de-  !

12 signed to keep radioactivity in plant effluents at the lowest practi-13 cable level and far below concentrations which independent government 14 authorities find to be satisfactory. The small quantities of radio-15 active vaste materials in liquid effluents will be well below concen- l 16 trations which such authorities find to be satisfactory for drinking 17 water. The cooling pond will protect against discharge of deleter-18 ious heat. All aspects of plant design, construction and operation  !

19 vill be carried out in accordance with applicable requirements of 20 Federal, state and local government agencies.

21 One important environmental consideration in planning for the 22 plant is the fact that operation of the plant will enable the Michigan 23 Power Pool to shut down a number of older plants which consume 2h fossil fuel. ~The new plant will also enable The Dow Chemical Company to.

23 shut down all of its existing steam plants which now burn fossil fuel.

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t As stated in the September 23, 1970 letter from the Federal Power Commission 1

2' to the Atomic Energy Commission concerning the applicant's proposed Midland 3 Plant:

4 "Under the prevailing fuels situation, if the Company had considered a fossil-fueled plant as a substitute for the two Midland nuclear units, the plant would of necessity have been a coal-fired plant and, as such, its operation would have added particulate and gaseous pollutants to the atmos-phere. The nonpolluting characteristics of nuclear generation can be expected to aid in maintaining air quality in the Company's service area."

5 Construction of additional electric generating capacity to meet 6 projected increased reqtdrements is a responsibility of Consumers 1

7 Power Company as a public utility under the laws of the State of l 8 Michigan. Falf111 ment of that responsibility is essential to the 9 health and economic well-being of the citizens of Michigan and surround-10 ' ing areas, as well as to the national defense. The Midland Plant is 11 important in meeting these goals and applicant believes it will do so in a 12 manner which is consistent with the highest environmental and ecological i 13 considerations.

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1 2.0' BACKGROUND 2^ - 2.1 Consumers Power Company 3 Applicant is a co=bination electric and natural gas utility 4 serving more than 1,000,000 electric customers in the lower peninsula 5 of Michigan. Applicant's electric service area covers approximately 6 27,100 square miles and contains a population of approximately 7 3,200,000. Principal cities receiving electrical service are Battle 8 Creek, Bay City, Flint, Grand Rapids, Jackson, Kalamazoo, Midland, Maskegon, 9 Pontiac and Saginav. Industries in the territory served by applicant 10 . include automobile and automobile equipment, primary cetals, chemicals, 11 fabricated metal products, pharmaceuticals, machinery, oil refining, 12 paper and paper products, food products and a diversified list of other 13 industries.

(~ 14 The maximum net demonstrated capability of applicant's intercon-15 nected system, which system is shown in Appendix B, Figure 2, exceeds 16 3,700,000 kilowatts. The net maximum demand on the interconnected sys-17 tem was 3,377,275 kilowatts on December 15, 1969 In order to serve 18 this load, applicant presently operates ten fossil fuel steam-electric 19 plants, one nuclear steam-ele :tric plant, fourteen hydroelectric plants, 20 seven gas turbine peaking plat ts and one internal combustion plant. In 21 addition, applicant and The Detroit Edison Company are the principal 22 membars of the Michigan Power Pool. Applicant is directly or indirectly 23 interconnected with other utilities in Michigan, Indiana, Illinois 24 and Ohio as a member of the MII0 group. The MII0 group consists in 25 varying capacities of' applicant, The Detroit Edison Company, Indiana &

26 Michigan Electric Company, The Toledo Edison Company, Northern La 27 Indiana Public Service Company, Ohio Power Company and Commonwealth 2-1

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F' 1 Edison Company. There are several agreements among various of 2 the companies in' regard to operations, facilities and coordination.

3 The agreements provide for the construction of interconnection 4 facilities and for the terms and conditions on which power is inter-5 changed. Applicant is interconnected through The Detroit Edison 6 Company with The Hydro-Electric Power Commission of Ontario. Applicant 7 is also a party to the East Central Area Reliability Coordination Agree-8 ment along with eighteen other parties. The purpose of this agreement 9 is to augment reliability of the parties' bulk power supply through 10 coordination of phnning and operation of generation and transmission 11 facilities. Applicant sells electric energy at wholesale to a number of 12 municipalities, rural electric cooperatives and invester-owned public 13 utilities.

14 . Applicant and The Detroit Ecison Company operate under an agreement 15 which provides fer pooled operation of their systems, coordination of 16 systems planning, design and construction, the rendering of mutual assist-17 ance during emergencies, and the effecting of maximum economy of produc-18 tion in providing the electric Aver requirements of each system. Each 19 system is thus dependent on the other system, as well as on its own, 20 having new generation available on schedule. In addition, both The 21 Detroit Edison Company and applicant must be concerned with the construc-22 tion programs of utilities in neighboring states and such utilities in 23 turn are concerned with and affected by their construction programs. To 24 this end, the Michigan Power Pool has made interconnection with utilities 25 in Indh m and Ohio and is coordinating its construction and operation 26 through the MII0 group and the East Central Area Reliability Agreement.

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c 1 The advantage of interconnection and coordinated construction and

2. operation is that savings can be made in investment for reserve 3 capacity; additional energy supplies can be obtained during time of 4 emergency; at times, more economical energy can be obtained from 5 . neighboring systems; advantage can be taken of the fact that peak loads 6 occur et different times on different systems; required maintenance 7 can be carried out at more convenient times and economies of scale 8 can result because of the ability to construct larger plants.

9 2.2 Applicant's Particitation in the Development 10 of Nuclear Power 11 Applicant owns and operates the Big Rock Point Nuclear Plant

-12~ located near Charlevoix, Michigan. This plant is a boiling water 13 reactor with a nameplate capacity of 75 MWe. Applicant received a con-(

14 struction permit for the Big Rock Point Plant dated May 31, 1960, a 15 provisional operating license, dated August 30, 1962, permitting opera-16 tion at power levels up to 157 MW t , and a final operating license dated 17- May 1, 1964, permitting operation at power levels up to 240 MW t

. Since 18 startup in 1962, the Big Rock Point Plant has produced net in excess 19 .of 2,423,164 MWhe. This reactor was one of the early privately owned 20 . reactors for the production of electric power.

21 Applicant has constructed the Palisades Nuclear Plant located near 22 South Haven, Michigan and is presently engaged in a public hearing on l

23 its application for an operating license for that plant. The Palisades 24 . Plant is designed to have an initial capacity of about 700 MRe with an 25- ultimate capacity of about 821 MWe.

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I 1 Applicant has been a pioneer in the use of nuclear power for elec-2 tric generation and has participated in the design, construction and 3 operation of nuclear plants for over ten years. It has a large and ex-4 perienced body of personnel that have operated nuclear plants and have 5 overseen the design and construction of nuclear plants.

6 23 Need for This Plant 7 The demands on applicant's system have doubled over the last ten 8 years and are expected to more than double over the next ten. It is neces-9 sary for an electric utility to forecast its demand well in advance and to 10 start planning and building new generation long before it is needed. It is 11 estimated that applicant's peak loads vill be 5,130 megawatts in the winter 12 of 1974-75 and 5,600 megawatts in the vinter of 1975-76. In order to meet 13 this projected demand and to provide the 17-18* percent reserve margin 14 which applicant considers desirable, applicant is planning additions to its 15 present net demonstrated capability of 3,731 megawatts. These addi-16 tions to capacity, in addition to the Palisades Plant, include an 17 1,800 megawatt pumped storage project for operation in 1973 which 18 is jointly owned with The Detroit Edison Company; the two-unit 19 1,300 megawatt Midland Plant for operation of one unit in 1974 and 20 the other in 1975, and two approximately 660 megawatt oil-fired in-21 termediate load units for operation, respectively, in 1975 and 22 1976. Of the units planned for this time period, only the Palisades 23' and Midland Plants are base load units, ie, designed to operate

• This 17-18 percent margin is an increase from the 15 5 percent margin previously considered desirable by Consumers Power Company. The in-crease results from the increased outage problems experienced by large generating units and difficulties of scheduling maintenance of such large units with only a 15 5 percent reserve margin.

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. ;" 1 more tha'n 50 percent of the time. 'Both the pumped storage plant and

-2 ^ the' oil-fired units are designed to operate less than 50 percent of the 3 time to meet various levels of peak demand. Both Midhnd units ar'e 4 needed on schedule if app 2.icant's system is to maintain an adequate 3 margin of reserves to' assure rbliable electric service to the State of 6- Michigan.

l 7_ While the Federal Power Commission's (FPC) comment to the AEC on

8 the environmental aspects of the Midland Plant, dated September 23, 1970, 9 does not reflect the latest data available to applicant or the current 10 construction sche 6,ias available to applicant, FPC's assessment of the 11 need for the Midland Plant is accurate:

12 ". . ., the load supply situations in the Pool . . .

should be judged in the light of recent experience with availability of hrge new generating units.

- During the period of October 1973 and February 1975,

l. the Pool is planning to place on'the line two fossil-fueld steam units with capacities in excess of 650 megawatts and a nuclear upit with a capacity in ex-cess of 1,000 megawatts [in addition to the Midland Plant] . The indicated reserve ~ margins assume avail-

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, ability of these units as planned. The recent history of large fo.ssil and nuclear _ steam generating units, however, reveals that during the first few years of-initial operation, the availability of hrge units can be highly uncertain. On this account, we feel that the indicated reserve me.rgins of the Michigan Power Pool are not excessive and that prudence dic-tates the need of the Midland Units 1 and 2 as a bulwark against the possible unavaihbility of existing capacity."

j 13 In addition, the Federal Power Commission stated:

lh- "It is evident, therefore, that if Consumers Power

, is to meet expected loads . . .,-reliance cannot be pheed on import of required firm power in lieu of construction of Midland Nuclear Units 1 and 2."

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l~ The Federal Power Commission, also, recognized the unavailability of 2 practical fossil-fired and hydroelectric alternatives to the Midland 3 Plant.

4 The Midland Plant is, therefore, necessary to meet peak loads and 5 to help assure a reliable supply of electric power. In addition, the 6 plant is necessary to assure continued growth and development in the 7 State of Michigan. The addition of the Midland Plant vill enable

-8 applicant to provide electricity for industrial, commercial and residen-9 tial development and will aid the applicant in maintaining a rate 10 structure which is competitive with the cost of energy in other states.

11 This in turn vill aid the state in attracting and retaining industry.

12 Also, low cost process steam vill enable Dow to maintain its growth in

( 13 Michigan. In addition, the availability of electric power in Michigan 14 vill enable applicant to aid utilities in other states during short-term 15 _ emergencies. Operation of the Midland Plant vill enable the Michigan 16 Power Pool and'Dow Chemical Company to retire older, less efficient 17 fossil-fueled units with resulting economies and with resulting benefits 18 to the environment. For all of these many reasons, the Midland Plant is 19 a desirable and necessary addition to the applicant's system and to the 20 electric supply system of the State of Michigan.

21 2.4 _ Considerations Entering Into Selection of the Plant 22 The Midland Plar'. features, including the location, grew out of a 23 - mutual need of The Dow chemical Company and Consumers Power Company to 1

2h meet their respective and increasiag energy demands. Dow, which had '

25 traditionally generated all of its own process steam and electrical

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1 requirements at Midland with dual-purpose generating plants, was 2 ' becoaing-less competitive for the manufacture 'of energy intensive 3 products due to prohibitive fossil-fuel cost escalation. In order k to take advantage of the fuel economies of a large nuclear plant 5 but lacking the projected load to utilize the full capacity of such 6 . a plant, Dow requested Consumers Power Company to consider an allo-7 cation of Dov's process steam requirements as part of a Consumers 8 Power Company facility. Consumers Power Company's system energy 9 capacity forecast for 1974-75 indicated a need for $ncreased gener-10 ating capacity in the Midland-Bay City-Saginaw load center area.

11 It was thus natural for Dow and Consumers Power Company to look 12 toward a dual-purpose nuclear facility to meet their respective 13 energy needs.

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Ik Inherent in the considerations given to the general features 15 of the plant, was the long-time Dow experience with steam extraction 16 which in' turn gave promise of minimizing the plant waste heat. This 17 feature gave-rise to a plant efficiency similar to the most efficient 18 fossil-fueled plants, and thus appreciably reduced the normal nuclear 19 plant heat rejection.

20 Tre air pollution resulting from the current Dow generating 21 plants will be eliminated. The use of standard components to enhance 22 reliability was a requisite in the Midland Plant.

23 By :onsidering the mutually compatible energy needs of Dow and 24 Consumees Power Company, a nuclear-fueled steam extraction plant using

- 25 a. cooling pond designed to dissipate the total plant heat rejection is l( , 26 the most appropriate means of meeting these needs.

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1 2.5 Process Steam To Be Supplied to Dow Chemical Company 2 The steam to The Dow Chemical Company will be supplied frcm the 3 secondary side of tertiary system heat exchangers. A complete system 4 of shell and tube evapurators vill be installed for both the high-5 pressure and low-pressure steam supplies. The design flows to Dow 6 are h00,000 lb/hr of the high-pressure and 3,650,000 lb/hr of the 7 low-pressure steam.

8 The feedwater in the tertiary loop from which the process steam 9 vill be produced comes from Lake Huron. At any given time, the feed-10 vater stream vill consist of condensate return from Dow, and makeup 11 vater from the lake which is demineralized before being transmitted 12 to applicant.

( 13 It is a criterion for the process stea= that no radioactivity be 14 added to the feedwater by the process of transforming it to steam in

' 15 the tertiary system heat exchangers. To meet this criterion, the 16 process steam will be separated from the reactor coolant water, which 17 contains radioactivity, by two barriers: a steam generator and a 1 tertiary system heat exchanger. In addition, the radioactivity in 19 the process steam vill be compared with the radioactivity in the makeup 20 vater. If the activity level in the steam indicates radioactivity 21 leakage to the, steam from the secondary system, the flow of process 22 steam from the specific tertiary heat exchanger producing the steam

-23 vill be teminated.

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4 1 30 THE EVOLUTION OF DESIGN OF THE MIDLAND FACILITY 2 As suggested in Appendix A,10 CFR Part 2, this portion of the 3 applicant's summary is a " discussion of the evolution of the proposed 4 reactor Sesign, including associated engineered safety features, from 5 the design of reactors which have previously been approved or built."

6 As will appear from this discussion and the information set forth in 7 other sections of this summary, the principal design features of the 8 Midland ??h ,t are based upon designs incorporated in earlier plants.*

9 31 Reactore 10 3 1.1 Shippingport - The First Commercial PWR 11 The first nuclear reactor constructed to generate electricity for 12 commercial uses was Shippingport. Design of the plant began in 195.

(4 13 and it first produced electricity for commercial use in 1957 14 Shippingport is a four 'oop PWR. Its initial core was about 15 6 feet high and 6-1/2 feet in diameter, was composed of zirconium-clad 16 fuel plates and rnds and produced approximately 335 MW . It employed t I 1T 32 cruciform rods for control and was contained in a reactor vessel 18 approximately 32 feet tall and 9 feet in diameter. Each ecolant loop )

19 contained one reactor coolant pump and one steam generator. This 20 - reactor operated at a pressure of approximately 2000 psia and an average 21 temperature of 520 F and produced sufficient steam at approxin:ately 600 22 psia to generate approximately 90 MWe.

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• For a comparison of important parameters for the Midland Plant with those for a number of other PWR, see Section 9 0 and Appendix A of this summary.

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1 Tb- Shippingport reactor at the time of its t3 sign was a reason-2 able extension of known reactor technology which hat been obtained 3 principally from the Navy nuclear program. The plant demonstrated 4' that commercial generation of electricity by nuclear power plants was 5 feasible and it pointed the way for the design of future PWR. Speci-6 fically, it showed that:

7 1. A light water-cooled and moderated nuclear reactor fueled 8 with slightly-enriched uranium could be used to reliably 9 produce steam to generate electricity.

10 2. Normal central station-type turbine generator equipment 11 could be coupled with a nuclear reactor.

12 3 Fuel element burnups of 3000 to h000 MWD /MIU were obtainable.

( 13 4. The use of concrete for shielding was practical.

Ik 5 The use of 600 psi saturated steam in a turbine generator 15 was practical.

16 3 1.2 Indian Point I - An Early Large Commercial PWR 17 A construction permit was issued by the Atomic Energy Commission 18 to Consolidated Edison for Indian Point I in May of 1956. This reactor 19 began commercial operation in 1962.

20 Indian Point I is also a four-loop PWR. Its original core was 21 composed of'120 box-type fuel elements, each of which was 6 inches 22' square, 8 feet long and contained stainless steel-clad thorium dioxide 23 ' fuel rods. It produces approximately 585 MWt , empl ys 21 cruciform 24 rods for control and is contained in a 9-foot, 9-inch diameter 25 stainless-ch, carbon steel reactor vessel. Each coolant loop k.

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[ -1 contains two reactor coolant pumps, and one steam g9nerator. The 2

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reactor. operates at a nominal pressure of approximately 1550 psia 3 and an average temperature of 495 F and produces sufficient steam

.4 at approximately 420 psia such that after it is superheated to 1000 F 5 by oil-fired superheaters approximately 275 MWe are generated.

6 313 Connecticut-Yankee - A Subsequent Large PWR 7 Construction of this plant began in May of 1964 and commercial 8 operation was achiev*d in 1967 9 Connecticut-Yankee, also a four-loop PWR, has a multiregion core 10 composed of 157 fuel assemblies of 204 stainless steel-clad uranium 11 dioxide fuel rods each. The core is arranged in three concentric fuel 12 regions. At the end of each fuel cycle of approximately one year's

( 13 length, one-third of the fuel assemblies is removed, the remaining 14 two-thirds is rearranged and a fresh one-third is inserted. The fuel 15 cycle management program permits the achievement of high average fuel 16 burnups of approximately 25,000 MWD /MrU.

17 The reactor produces approximately 570 MWe and employs 45 cluster 18 rod assembliev for control. The design of these assemblies is sig-19 nificantly different from that of the bladed cruciform rods used in 20 previous reactors. It consists of twenty neutron absorber rods joined 21 together at the top by a spider-like bracket. The absorber rods travel 22 in guide tubes located within the fuel assemblies. This cluster control

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23 rod significantly decreased the neutron power peaking that was exper- l 24 ienced with the use of cruciform control rods. To further improve the 25 reactor fuel cycle and decrease neutron power peaking, chemical shim 1

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1 control is also employed in the connecticut-Yankee reactor. This 2 shim control consists of adding a small quantity of boric acid to 3 the reactor coolant. The boric acid serves two functions: (1) It 4 increases the suberitical margin of the core on shutdown and (2) it 5 gradually and precisely compensates for the thermal activity changes 6 due to fuel depletion and fission product accumulation as it is slowly 7 removed over the course of one fuel cycle.

8 The highly successful operation of the Om.necticut-Yankee PWR and 9 also the similar san onofre plant have been significant factors in es-10 tablishing the confidence that utilities now nave with regard to nuclear 11 power plants being able to produce electrical power safely and efficiently.

12 3 1.4 Robert Emmett Ginna - A Large PWR Which Recently

. . 13 Began operation 14 Ginna achieved commercial operation in early 1970. It is a two-4 15 loop PWR with a core power rating of 1455 MW and 18 8Pable of a gross t

16 output of approximately 496 MWe, Its 12-foot long core is made up of 17 121 fuel assemblies, each having 179 Zirealoy-4-clad fuel rods arranged 18 in 14 x 14 arrays and containing uranium dioxide pellets. Sixteen of 19 the rod positions in each fuel assembly are occupied by control rod 1

20 guide tubes and one position is occupied by an instrument tube, l

21 The initial core loading was divided into three roughly concentric 22 regions. New fuel will be loaded into the center region at each yearly l 23 refueling.

24 Reactor control is provided by a combination of cluster rod assem-25 blies and chemical shim. Of the 33 control rods, 29 are full-length and b

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1 4 are part-length. The full-length rods have a total reactivity worth 2 sufficient to shut the reactor down from any operating condition even 3 with the highest worth rod stuck in its fully withdrawn position. The 4 four part-length control rods are provided to flatten axial neutron flux 5 variations and to dampen out any spatial xenon oscillations that may re-6 sult from power changes.

7 The plant is designed to accept a step-load change of 10 percent 8 and ramp-load changes of 5 percent per minute between 15 percent and 100 9 percent of full power. The plant is designed for automatic load fciloving, 10 and partial load rejection without, trip.

11 In-core instrumentation in the form of remotely positioned ion 12 chambers provides neutron flux distribution information to the plant

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14 315 Oconee 1 - A Modern Large PWR Not Yet in Commercial Operation 15 Duke Power Company's Oconee Units 1, 2 and 3 received construction 16 Permits in 1967 In september 1970, Oconee 1 received favorable ACRS  !

l 17 review to commence operation and is scheduled to produce commercial l 1

18 power by mid-1971. l l

19 Midland Units 1 and 2 are the ninth and tenth reactors of a series 20 of Babcock & Wilcox Company (B&W) reactors of essentially identical 21 design.* Oconee Unit 1 is the first of this series. It is a two-22 loop PWR with an ultimate core power rating of 2568 MWe and a net i 23 electrical output of 886 MWe. Each loop contains a straight tube once-24 through steam generator and two reactor coolant pumps. A pressurizer vno

• The 8 prior units are listed on Lines 4e.-N" on Page 11-3 f

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1 connected to one of the two loops maintains the reactor coolant in a

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2 subcooled state. Each of the two reactor coolant outlet pipes contains e 3 a calibrated flow tube which is used to measure the reactor coolant flow 4 rate during operation. The reactor vill operate at a nominal pressure 5 ~ of 2200 psia with an average coolant temperature of 580 F. The reactor 6 has a 12-foot long core made up of 177 fuel assemblies, each of which 7 contains Zircaloy-4-clad fuel rods. The fuel rods contain uranium di-8 oxide pellets and are arranged in 15 x 15 arrays. Sixteen of the rod 9 positions in each fuel assembly are occupied by control rod guide tubes 10 and one position (the center one) can accept an instrument tube.

11 The reactor vessel consists of an 171-inch ID cylindrical shell 15 approximately 8-1/2 inches thick.

( 13 Reactor control is provided by a combination of cluster rod assem-14 blies arid chemical shim. Of the 69 control rods, 61 are full-length and 15 8 are part-length. In-core instrumentation consists of an assembly of 16 self-powered neutron detectors and a thermocouple.

17 3 1.6 Midland Units 1 and 2 18 An application for construction permits for Midland Units 1 and 2 19 was filed with the AEC in January 1969 Unit 1 is expected to produce 20 commercial power and process steam in 1974.

21 The reactors for these units are essentially identical to those of 22 the eight B&W units scheduled to precede them into operation. The Mid-23 land units are described in section 7 of this document.

24 The basic design of the fuel assemblies, control rod assemblies, 25 control rod drives, steam generators, and instrumentation has been f

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1~ confirmed by proof-testing performed by B&W and the actual components 2 will be proven in operation on earlier B&W designed plants.

3 32 Evolution of Engineered Safeguards k Engineered safeguards systems are defined as that equipment, of 5 either static or dynamic design, which is utilized in the event of a 6 reactor accident to limit the environmental consequences of that 7 accident.

8 The history of engineered safeguards features for commercial pres-9 surized water power reactors began when the Shippingport project was 10 authorized in July 1953 Seventeen years of design experience have 11 been accumulated in safeguards design and the current engineered safe-12 guards equipment installed on present-day plants is the result of design

( 13 evolution and extensive research and development. During the course of 14 these years, the improvements and capabilities of engineered safeguards 15 features have maintained or increased the protection provided the public 16 against the potential releases of fission products resulting from postu-17 lated reactor accidents.

18 Today there are about 60 commercial pressurized water power reactors 19 that have been designed, constructed or are in operation. All of these 20 plants have engineered safeguards systems to provide protection to the 21 public against radiation hazards.

22 Shippingport Engineered Safeguards Features 23 The Shippingport plant , which was designed for 335 Wg, contained 24 some of the engineered safeguards features presently utilized in current 25 designs. A brief summary discussion of these is as follows:

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l 1 1. The reactor system was housed in a steel container designed 2 to withstand the energy release from a loss-of-coolant accident.

3 The resultant design pressure was 52.8 psia. This building was 4 leak tested and showed that the leakage rates of a fraction of 5 1 percent per day were achievable.

6 2. The reactor plant had a safety injection system to supply 7 emergency core coolant in the event of a loss of integrity 8 of the reactor coolant system.

9 This system contained two 1500 gpm pumps that were automatically 10 actuated when the reactor coolant system pressure decreased to 500 psia 11 and was designed to reflood the core, thereby preventing significant 12 melting or release of fission products.

-r 13 Engineered Safeguards at Later Plants 14 With the advent of higher performance reactors, the single safety 15 injection system concept of the Shippingport plant was supplanted by a 16 combination of high-pressure injection systems and low-pressure injection 17 systems to provide emergency coolant to the core in the event of the loss 18 of reactor coolant system integrity. These systems provided a wider range 19 of coolant replacement capability in both flow and pressure to insure 20 adequate core cooling for a complete range of rupture sizes. Typical of 21 such designs was the Yankee plant (2) which had a 770 psia high-pressure 22 injection system and a 300 psi low-pressure injection system, utilizing 23 boric acid for criticality control.

24 The inability of the natural heat sinks within the reactor building 25 to absorb sufficient quantities of heat to prevent overpressurization of

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1 the reactor building structure required the use of supplemental cooling 2 as plant size increased. Both emergency building cooling equipment and 3' spray systems were utilized to prevent overpressurization during the 4 course of the postulated accident. Typical plants which have utilized 5 these concepts during this evolutionary period are Connecticut-Yankee 6 plant and the Ginna plant.

7 To meet the AEC's siting criteria (10 CFR 100) as plants became 8 larger, either greater exclusion distance or engineered safeguards 9 systems capability must be extended to include reduction of iodine con-10 centration within the post-accident reactor building atmosphere to within 11 acceptable limits defined by the site characteristics.

12 Two basic types of iodine and particulate fission product removal

( 13 systems have been incorporated in past designs. one of these is a 14 filter system which removes particulates and absorbs iodine on charcoal 15 'as the reactor building air is recirculated through the filtering system.

16- The second employs the reactor building spray droplets to impact par-17 ticulate matter and wash the iodine from the steam-air atmosphere. The 18 .vashing is accomplished by a chemical reaction between the iodine and 19 the chemicals contained in the spray droplets.

20- A typical plant design which included filters for removal of par-21 titulate material and iodine absorption was Connecticut-Yankee, which was 22 licensed for construction in 1964. 1 23 Spray removal systems using a sodium thiosulfate reagent for iodine ,

l 24 removal were first incorporated on Metropolitan Edison's Three Mile 1 25 IslandPlant(3) ,

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1 The most recent engineered safeguards feature incorporated to pro-2 vide core protection during the postulated loss-of-coolant accident is 3 the _ static core flooding system. Thus, the dynamic injection systems 4 were supplemented with fast-acting core flooding systems. These systems 5 contain emergency coolant in pressurized tsinks directly connected to the 6 reactor coolant system. These tanks of coolant are separated from the T reactor coolant system by check valves that open when the reactor coolant 8 pressure decreases below the pressure in the tanks. The net effect is a 9 rapid insurge of coolant to the core that is proportional to the system 10 demand for coolant replacement. Thus, a large leak would cause rapid 11 reactor coolant system depressurization and maximum flow from the flooding 12 system. Smaller ruptures have lower rates of depressurization and lower

.( 13 flow rates from the core flooding tanks. These static devices respond 14 rapidly to the needs of the reactor for coolant makeup and do not require 15 an external source of power.

16 This type of engineered safeguards feature was first incorporated ,

i 17 on the Consolidated Edison Indian Point II(4) and Duke Power Company '

18 Oconee plants an,. has also been incorporated in applicant's Palisades 19 Plant.

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20. Engineered Safeguards Features of Midland Plant 21 The Consumers Power Company't. Midland Plant Units 1 and 2 incor-22 porate engineered safeguards features that have been the result of plant 23 design improvements over the past several years. Those features which 24 are incorporated to satisfy the siting requirements for the Mid1_nnd plant 25 are described in detail in applicant's Preliminary Safety Analysis Report.

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1 The reactor is housed in a lov leaka6e (0.1 percent by volume per 2 day) steel-lined prestressed concrete reactor building. Automatic iso-3 lation of lines entering the reactor building is effected in the event 4 of an accident requiring isolation. The reactor core is protected 5 against fuel melting during loss-of-coolant type accidents by emergency 6 core cooling systems. These systems are the core flooding system, the T high-pressure injection system, and the low-pressure injection system.

8 Radioactive iodine released into reactor building during a postulated 9 loss-of-coolant accident is removed by a building spray system that 10 utilizes a sodium thiosulfate solution to rapidly and irreversibly react 11 with the iodine before it can leak to the environment.

12 Supplemental reactor building cooling is provided by two independent

(- 13 systems, a reactor building cooling system and the reactor building spray 14 system. These prevent post-accident building overpressurization and pro-15 vide for rapid depressurization of the building, thus reducing the leakage 16 rate to the environment.

17' Summary 18 The evolution of the reactor designs to larger sizes since 19 Shippingport has prompted the development of improved engineered safe-20 guards features to provide environmental protection against the conse-21 -quences of a majo'r fission product release associated with a postulated 22 loss of reactor coolant. These have evolved from the orderly analysis of 23 design needs and the designs are based upon a wealth of experimental data 24 that verifies design requirements and achievable performance.

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l The engineered safeguards features incorporated on the Midland plant 2 are essentially identical to those used on other reactor plants presently 3 approved for construction. Accordingly, the engineered safeguards fea-4 tures used for this plant represent fully engir.eered designs that have 5 been thoroughly evaluated and found to provide protection of the public 6 well within AEC guideline values.

7 33 Reactor Buildings 8 The Midland reactor buildings are posttensioned reinforced concrete 9 structures. In the design of these buildings, full advantage is being 10 taken of the experience gained in the review of similar designs with 11 the AEC for the Florida Power and Light Company's Turkey Point Plant, 12 Wisconsin-Michigan Power Company's Point Beach Plant, Duke Power 13 Company's Oconee Nuclear Station, Arkansas Power and Light Company's 14 Russel1Yille Plant, the Sacramento Municipal Utilities District Rancho 15 Seco Plant and the Palisades Plant, as well as reactor building designs i

16 by others which meet the same functional requirements. All the foregoing 17_ plants which have similar posttensioned reactor buildings have received 18 construction permits. Further, the Palisades and Point Beach reactor 19 buildings have been successfully leak tested and pressure tested at 20 115 percent of their design pressure levels. The reactor building is j i

21 designed to contain in excess of the maximum pressure resulting from 22 the most severe postulated loss-of-coolant accident. The reactor building 23~ integrity is assured by means cf a 1/4-inch steel liner plate which com-

- 24 pletely encloses the interior of the reactor building. The liner plate 25 is provided with a leak chase system which consists of steel channels 12 w

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1 velded over the liner plate seams. These channels are continuously 2 pressurized to assure that any leaks which may occur would be inward.

3 -All-liner plate penetrations are provided with a pressurized boundary E at a pressure greater than the building design pressure.

5' Table 3 3-1 ) compares the principal differences of the Midland 6 reactor building with reactor buildings of other plants which received

-7 construction permits. It should be pointed out that the 170-wire tendon 8 to be used for the Midland reactor building was developed for the pre-9 stressed concrete reactor vessel for the Fort Saint Vrain plant and 10 stressing equipment has provided 1000 cycles of load applications, more 11 than sufficient to stress all tendons on this reactor building, in one 12 test series alone.

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2 Table 3 3-1 Comparisons With Other Reactor Buildings Consumers Power Co Consumers Power Co Arkansas Power & Light Co Midland Plant Palisades Plant Russellville Plant 1- Diameter H6'-0" n6'-O" H6'-O" 2 Design Pressure 67 Paig 55 Paig 59 Psig 3 Wan Thickness 3'-6" 3'-6" 3 '-9" 4 Dome Thickness 3'-0" 3'-0" 3'-3" 5 Ievel of Prestress '1.2P-1 5P(1) 1 5P 1 5P 6 Wires per Tendon 170 (Maximum)- 90 186 7 Size of Wire 1/4" 1/4" 1/4" 8 Buttresses 3 6 3 Y

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Sacramento Municipal Utility District Florida Power & Light Co Wisconsin Michigan Power Co

Rancho Seco Turkey Point Point Beach 9 Diameter 130'-o" n6'-o" los '-o" x 10 Design Pressure 59 Paig 59 Psig 60 Paig

11 Wall Thickness 3'-9" 3' .0" 3'-6"

! 32 Dome Thickness 3 ' -0 " 3'-3" 3'-0" 13 Level of Prestress 1.2P-1 5P(1) 1 5P 1 5P 14 Wires per Tendon 55 Strands 90 90 15 Size of Wire 1/2" (7 Wire) 1/4" 1/4" 16 Buttresses 3 6 6 1

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References

1. The Shippingport Pressurized Water Reactor - 1958 - Addison-Wesley Publishing Co, Inc.

2. Part B License Application AEC Docket No. 50-29 " Technical Information and Final Hazards Summary Report."

3. Metropolitan Edison FSAR (Docket No. 50-289)

h. Indian Point No. 2 (Docket No. 50-2h7) 5 Oconee No. 1 (Docket No. 50-269)

6. PSAR, Section 5 1.

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e 1 4.0 ENVIRONMENTAL QUALITY 2 4.1 Introduction 3 In designing the Midland Plant, applicant is making every effort 4 to consider its impact on the envirenment, having as a goal the preser-5 vation and enhancement of environmental quality.

6 To help assure that this objective is met, air and water motltoring T programs are being established for the plant. These programs are des-8 cribed in Section 10 of this summary.

9 4.2 Air Quality 10 Trace amounts of gaseous fission products, less than a few percent 11 of the AEC limits, will be released to the envivonment. The radioactive 12 waste gases will be held up in the vaste gas decay tanks and filtered

! 13 to limit the off-site release to as low a level as practicable.

14 43 water Quality 15 To reduce the heat load on the Tittabawassee River and comply 16 with Michigan water quality standards, an 880-acre cooling pond will be 17 constructed adjacent to the plant. The pond is designed to accept 100%

, 18 of the vaste heat from the plant. It will be necessary to flush solids 19 from the pond to the river. However, this process merely returns to the 20 river solids present in the water drawn into the pond from the river which 21 have concentrated in the pond; and there will be no net addition of 22 solids to the river from the Midland Plant due to this process.

23 Due to evaporation of water from the pond, some water vapor will be 24 released to the atmosphere. Experience at similar facilities indicates

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1 that fogging will occur only over the immediate vicinity of the pond.

2 There vill be Jess release of water vapor from the pond than if  :

l 3 cooling towers were used to accept the entire vaste heat load, h There will be a controlled release of rcdioactive nuclides to 5 the river from the plant's liquid radioactive waste system. This sys-6 tem processes radioactive liquid waste through tanks, filters, de-7 mineralizers and evaporators to reduce radioactivity to levels as low 8 as practicable within AEC limits prior to discharge. This release will 9 be less than a few percent of the AEC limits.

10 h.4 Solid Wastes 11 Radioactive solid vastes, such as demineralizer resins, spent 12 filter elements, clothing, mgs, and other solid vastes will be packaged

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13 in suitable containers and shipped offsite for disposal by an AEC-licensed 14 contractor. Spent fuel will similarly be shipped offsite to a fuel re-15 processing facility.

16 h.5 Iand Use and Aesthetics 17 ,

The Midland Plant's functional design will be blended harmoniously 18 with the surroundings. The plant, which will occupy an 1190 38-acre 19 site along the south shore of the Tittabawassee River, is located directly 20 across the river from The Dow Chemical Company's industrial complex. The 21 plant structures will be located directly adjacent to the river. Dow 22 waste ponds are located nearby. The 880-acre cooling and storage pond 23 vill occupy an area generally to the south of the plant structures, as 2h shown on Appendix B, Fig 4. Applicant will plant trees and shrubs along h-2

v 1 the pond dike to screen it from a residential subdivision near the 2 - western boundary and from a road along the south boundary. The 3 center of the pond dike will be set back from the site boundary a

< 4 distance of not less than 160 feet and a security fence will surround 5 the dike.

6 Because there will be no bulk fuel storage and handling facili-7 ties gr high stacks, the plant will be streamlined in appearance in 8 comparison to fossil-fired power plants, see Appendix B, Fig 1. Noise 9 from the plant during normal operation will be minimal and not detec-10 table at the site boundary.

11 4.6 National Environmental Policy Act 12 The National Environmental Policy Act (NEPA) requires, among 13 other things, that all agencies of the Federal government include 1

14 in every recommendation or report on major Federal action signifi-15 cantly affecting the quality of the human environment, a detailed 16 statement by the responsible official on the following: l 17 "(1) the environmental impact of the proposed action, (ii) any adverse environmental effects which cannot be avoided should the proposal be implemented, (iii) alternatives to the prcrosed action, (iv) the relationship between local short-term uses of man's environment and the maintenance and enhance-ment of long-term productivity, and (v) any irreversible and irretrievable commitments of resources which would be involved in the proposed action should it be implemented."

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1 Prior to making such a detailed statement, the agency is required to 2 consult with and obtain coments of any Federal agency which has 3 Jurisdiction by law or special expertise with respect to environmental h impact.

5 In compliance with the policy of the Atomic Energy Comission, 6 applicant has prepared and filed with the Comission an environmental 7 report providing information on the above five points. The Atomic 8 Energy Comission has circulated it for coments to Federal agencius 9 having jurisdiction or special expertise with respect-to environmen-10 tal impact and, by publication of notice in the Federal Register, 11 35 FR 12795, August 12, 1970, has made the report available to local 12 agencies. Coments have been received from the Federal Power Comis-( - 13 sion, the Department of Agriculture, the Department of Housing and 14 Urban Development, the Department of Defense and Michigan Department 15 of Natural Resources.

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1 50 OTHER GOVERNMELTAL AGENCIES 2 Consumers Power Company intends to acquire all necessary regu-3 latory approvals for construction and operation of the Midland Plant.

k It has sought to fully cooperate with all governmental agencies having 5 an interest in the plant and has attempted to keep all agencies fully 6 informed. As part of this policy of cooperation, Consumers Power has 7 served copies of the application with all amendments as they were 8 filed on the Mayor of the City of Midland, on the Supervisor of 9 Midland Township, and on the Chairman of the Board of Commissioners 10 of Midland County and has served six copies on the State of Michigan 11 to be distributed to interested agencies.

.12 Besides applying to the Atomic Energy Commission for a permit

~k 13 to construct and a license to operate the Midland Plant, Consumers 14 Power on' June 9,1970 filed with the Michigan Water Resources Commis-15 sion a Statement of New or Increased Use of Waters of the State for 16 Waste Disposal Purposes and on August 28, 1970 applied to the Michigan 17 Air Pollution Control Commission for authority to construct and operate 18 air pollution control equipment at the Midland Plant. Following a 19 public hearinE held on August 10, 1970, the Water Resources commis-20 sion issued an Order of Determination, dated October 15, 1970, imposing 21 specific limits on the proposed use of the water of the Tittabavassee 22 River by Consumers Power. The Michigan Air Pollution Control Commission 23 is considering Consumers Power's application and is expected to act in 2h the near future.

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1 Consumers Power received, following a public hearing, approval 2 for its proposed use of land in the pond area from the Midland Town-3 ship Zoning Board of Appeals and received a building permit from 4 the Midland Tevnship Building Inspector. Consumers Power has provided 5 information in relation to construction of the plant to the City Com-6 mission of Midland, the Midland Township Board, the Midland County 7 Board of Commissioners, the Midland County Drain Commissioner, the 8 Midland County Road Commission, the Michigan Departments of Public 9 Health, Inbor and Natural Resources, the Michigan Public Service 10 Cecmission, the Michigan Water Resources Commission, the Governor's 11 Study Committee on Atomic Energy, the US Coast Guard and the US Army 12 Corps of Engineers.

I 13 Consumers Power has received miscellaneous approvals relating to 14 the plant from several'of these agencies and vill work closely with 15 these and any other agencies having an interest in the plant in 0.-der 16 to secure all necessary approvals and to assure that these govern-17 mental agencies are fully informed as to the status of the project.

18 The Midland Township Board by resolution on Septen er 9, 1970 19 let it be known that it encouraged the installation ef the Midland 20 Plant at the plant site located in Midland Township. The Midland 21 County Board of Commissioners by resolution on August 11, 1970 sup-22 . ported the location and development of the Midland Plant in Midland 23 County. On August 26, 1970, the Midland County Road Commission by 24 resolution supported the location and development of the Midland 25 Plant in Midland County.

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l 6.0 SITE CHARACTERISTICS 2 6.1 Incation 3 The Midland Plant is to be constructed on the south shore of h the Tittabawassee River adjacent to the southern limits of the City 5 of Midland, Midland County, Michigan. The site and immediate vi-6 cinity are shown in Appendix B, Figures 3, 4 and 5(1) .

7 All land comprising the site will be controlled by applicant ( .

8 This includes all land bounded by the Tittabawassee River bank on the 9 north and east, on the south by Gordonville Road and on the west by 10 farmland and scattered residences.

11 6.2 Population (3) 12 The plant exclusion area has a minimum radius of 0 31 mile. The

( 13 exclusion distance to the south site boundary, encompas nns the pond, lh is 1.12 miles.

15 The distance to the boundary of the low population zone is 1 mile, 16 within which there is estimated to be 38 permanent residents plus 17 portions of The Dow Chemical Company industrial complex.

18 The population distribution is shown in Appendix B, Figure 6. The 19 siting and safeguards provisions of the plant are in accordance with 20 the AEC guidelines set forth in 10 CFR Part 20 (" Standards for Protec-21 tion Against Radiation") and 10 CFR Part 100 (" Reactor Site Criteria").

22 63 Land use 23 Generalized descriptions of the land use within five m!.les of 24 the site are given below . Present use of land was determined by 6-1

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1 examining geological survey or and aerials of the area. Future use 2 was based on past development, ipaustrial trends and suitability of 3 land for various purposes 4 North - (0-15 miles) - Heavily industrialized. Dov's main industrial 5 complex lies in this area. Room for industrial 6 development in NNE section.

7 - (15-4 miles) - contains the populated residential, comercial 8 community of Midland. Area is nearly saturated 9 with buildings. Little growth expected.

10 -(4-5 miles) - Primarily residential. This area is expected 11 to develop into a residential suburban comunity.

12 East -(0-1 mile) - Heavily industrial, area saturated. ,

( 13 -(1-3 miles) - Sparsely . populated residential area. Many 14 forested acres, scattered faming. Expect 15 industrial growth 1-2 mile section. This is 16 another area for suburban development.

17 South.-(0-1 mile) - Iand contained within site.

1.8 - (1-3 miles) - Mostly forested lands, some faming. Poten-19 tial area for industrial expansion.

20 - (3-5 miles) _ Primarily faming, scattered nonfarming 21 residents, very few comercial establishments.

22 West - (0-1 mile) - Industrial property owned by applicant and Dov.

23 -(1-2 miles) - Mostly suburban residential, light faming.

24 Boom for suburban residential development.

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1 - (2-5 miles) -

Primarily agricultural, sparsely populated.

2 Some forested land.

3 The area 5-50 miles from the site is primarily used for farming, 4 -where not forested,except for the industrial co=munities of Bay City, 5 Flint, and Saginav. A large portion w tne area northeast of the 6 site is public land used for hunting and camping.

7 6.4 Meteorology 8 The site meteorology has been extensively investigaten to provide 9 an assessment of environmental consequences of routine and accidental 10 releases of radioactivity. The topography at the site is comparatively 11 flat with elevations ranging from 600 to 634 feet above mean sea level.

12 It is estimated that there are no significant changes in the topography

(~ 13 greater' than 50 feet within 50 miles. Thus, topographically induced or 14 altered winds should not be important at this site.

15 Temperatures of 90 degrees or higher occur on an average 14 times 16 per summer and of zero or lower on an average of 6 times during the 17 winter. Mean annual precipitation at Midland is 29.8 inches. Rainfall 18 is greatest in June, averaging 3 15 inches. Snowfall totals 33 3 inches 19 during an average vinter. Cloudiness is greatest in the late fall and 20- winter. Prevailing vind direction is southwest and average hourly velo-21 city is greatest in the early spring and lowest in late summer and early 22 fall. A study was made of the site atmospheric diffusion characteristics, 23- utilizing conservative meteorological conditions (5) . Extensive meteor-24 - ological data was available from the Tri-City Airport at.l Dow. A 25 meteorological program for the site vill co=mence at least two years 26 .before fuel loading.

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1 6.5l Surface' Water Hydrology 2- Surface water investigations were made in connection with the 3 safety aspects of the plant. These included the magnitude.of pos-4 sible floods and possible failure of upstream dams .

5 .The maximum recorded flood since 1907 reached a stage of 610.0 6 feet in 1916( }.

7 The probable maximum flood discharge at the plant site has been 8 ~ conservatively calculated to be about 270,000 cfs which includes the

-9 effect of upstream dam failures. This discharge is calculated to re-10 sult in a probable maximum flood level of 632 feet at the plant site 11 which level has been used as a basis for plant design.

12 The plant area vill be filled to elevation 634 and thus the vital

( 13 installations are protected against water damage in the imHkely event 14 of a probable maximum flood.

15 6.6 Groundwater Hydrology (N 16 .The presence.of a thick, impermeable clay member has produced two 17 , hydrologic conditions at the site. They are:

18 1. . A perched water table in the sand above the clay.

19 2. An artesian aquifer in the sand and gravel. underlying the clay.

~20 The perched water table. is in the upper sand which generally is 21 less than 50 feet deep in the borings. The quantity of water in these 22 - _ surface sands is limited and is not a source of domestic supply in the-23 -site area. Snell domestic supplies are obtained from the underlying

24. f confined aquifer.

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1. u 1 The thick, imper =eable clay layer, which acts as the confining 2 media pruterting the upward flow of the artesian water, also prevents 3 the downcard percolation of surface water from the plant area.

4 Water wells located within the cooling pond area have been sealed 5 to insure that cooling pond water does not seep into the domestic supply 6 and that no artesian groundwater leaks into the cooling pond water.

7 6.7 site Geology (8) 8 671 General 9 Site investigations were conducted to evaluate the geologic con-10 ditions that are pertinent to the design, construction and operation of 11 the Midland Plant. These investigations indicate that the Midland area 12 is underlain by two types of glacial deposits. These are a loose glacial

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13 lake deposit overlying a dense glacial till. The plant is founded on 14 the dense till. These deposits overlie bedro::k which is the Pennsyl-15 vanian Saginav formation composed of nearly flat-lying shales, sand-16 stones, and siltstones. The glacial deposits vary in thickness from 17 350 to 360 feet.

18 672 Salt Solution Cavities 19 Underlying the site are thin salt beds encountered in the 20 Devonian formation of the Detroit River section at a depth between 21 4100 and 4300 feet. The Ibv chemical company has mined these salt 22 beds for some time ar' Sas also extracted brine from the Devonian 23 sylvania sandstone at a depth of 5100 feet. overlying the area 24 mined are dense beds of limestone, dolomite, shale and sandstone.

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1 The maximum surface subsidence has been investigated by two 2 separate consulting firms, Woodward-Clyde & Associates and General 3 Analytics, Inc. One firm is experienced in foundation investigations, 4 and the other is experienced in mine settlement investigations. Experts 5 with these firms reviewed the information from the Dow brine well mining 6 operations. The data include drilling records and drill corings of 7 the rock strata overlying the salt solution cavities, accurate records 8 of materials removed and records of ground surface elevations.

9 An independent calculation of surface subsidence was made by each 10 consultant. In one case, the consultant used an analytical approach 11 based on est1I:uted cavity openings and the known elastic properties of 12 the surrounding rock strata. The other consultant used an empirical 13 method developed to predict surface subsidence of mines in England.

14 The conclusion reached by both consultants was that for the salt wells 15 in the area of the plant site, the maximum surface subsidence which 16' could be expected would be less than 1 inch (9) .

17 .

Furthermore, if subsidence were to occur, it woula affect a 18 large area due to the spacing of the wells, their grest depth and 19 the competency of the overlying flat-lying sedimentary rocks. Such 20~ a broad regional subsidence would have no significant effect on sur-21 face structures, which cover a scall area compared to the probable 22 subsided area. The negligible predicted maximum subsidence has been 23 confirmed by the r corded surface elevations over the past 10 years.

24 The Dow Chemical Company has observed no measurable surface subsidence 25 - due to the salt solution extraction. Lo further insure the safety of 6-6

1 the Midland Plant, Dow will not drill any new brine or salt wells 2 within one-half mile of the plant site.

3 _6. 8 seismoloh(10) k The utdland site is located in a region of slight seismic ac-5 tivity. .dthough earthquakes have been felt in this region of the 6 United States, Midland experienced all with low intensity. The maxi-7 mum intensity experienced at the pzoposed Midland nuclear site as a 8 result of any historic earthquake is V on the Modified Mercalli scale.

9- Intensity v corresponds to a surface acceleration of 0.03 g on 10 Hersberger's (1956) curve. For design of the plant Class I structures,*

11 a conservative value of 0.06 g will be used for the operating basis 12 earthquake, and 0.12 g acceleration will be used for the design basis

( 13 earthquake fer safe plant shutdown.

Ik No faults are mapped in the surficial deposits of the southern 15 peninsula. The active Keweenaw fault exists approximately 325 miles 16 northwest of the site on the Upper Peninsula. This fault, however, 17 is not related geologically to the structural province of the Michigan 18 Basin and its activity is not important in the evaluation of the Midland 19 site. A less important fault zone about 240 miles northwest of the site 20 in the vicinity of the Menominee Range does not appear to be active.

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1. Class I structures, systems and equipment are those whose failure could cause uncontrollable release of reactivity or those essential for immediate and-long-term operation following a loss-of-coolant accident.

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1 6.9 Foundations 2 The reactor buildings and lower portions of the auxiliary building 3 are at elevations such that the foundations are established on the 4 stiff-to-hard cohesive soils which underlie the site. These soils 5 are considered to provide excellent foundation support without exces-6 sive settlement under both static and dynamic conditions of loading ( .

7 These Class I structures are founded on earth-supported mat foundations.

8 The south portion of the auxiliary building has its base at el-9 evation 610 while the existing ground surface soils in this area vary 10 between elevation 605 and elevation 612. The surface soils in this 11 area are loose sands of variable thickness which do not provide suit-12 able foundation support. Consequently, these soils are to be removed k 13 down to the underlying very stiff-to-hard cohecive soils and will be ik replaced with controlled compacted granular or cohesive fill. All 15 other Class I structures or components will also be founded on the 16 stiff, hard cohesive soils cr on engineered compacted fill material.

17 Furthermore, in the area of the Turbine Building, all loose sands, 18 soft or compressible clay soils, and organic soils will be excavated 19 and replaced with suitable fill material.

6-8

a - - -

(

References *

1. PSAR, Sections 1.2.1 and 2.2.1

2. PSAR, Section 2.2.2 3 PSAR, Section 2.2 5

h. PSAR, Section 2.2.6 5 PSAR, Appendix 2A

6. PSAR,' Appendix 2B 7 PSAR, Section 2.6

8. PSAR, Section 2 5 3 9 The Woodward-Clyde & Associates report is appended to Item 2.17 submitted by Amendment No 7; it follows Amendment No 5 in Volume II of the PSAR. The General Analytics, Inc report var submitted in

(. Amendment No 10, PSAR Volume III.'

10. PSAR, Section 2 7

11. PSAR, Section 2.8.4 3

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1 70 DESCRIPTION OF MIDLAND PIANT

.( 2 71 Introduction 3 A description of plant features and layout, as well as an eval- >

b untion of plant safety, is set forth in the application and its amend-e 5 ments. The plant description emphasizes the concepts, guidelines and 6 criteria which will govern final design.

7 The station will consist of two reactor buildings, an auxiliary 8 building (including fuel storage, control room and radwaste area), a 9 turbine structure (housing two turbines), an administration building 10 (including offices, a shop and warehouse), a water storage and cooling 11 pond, a switchyard and various other auxiliary structures and equipment.

'12 Arrangement of the Midland Plant is shown in Appendix B, Figure 3 13 Section 9, together with Appendix A, presents a comparison of the 14

{ nuclear steam supply system design parameters of the proposed Midland 15 Plant with those of Duke Power Company's Oconee Unit 1; Commonwealth 16 Edison Company's Zion Plant and the Sacramento Municipal Utility 17 Company's Rancho Seco Unit 1 Plant.

IN The following subsections are a summary of the principal features 19 of the plant which are significant with respect to safety considerations.

20 72 Reactor and Reactor Coolant System-21 The reactors for the Midland Plant are of the pressurized water 22 type. 'Ibey have an initial core rating of 2452 14i teach and together 23 vill produce approximately 21 x 106 lb/kW steam for production of elec-2b tricity and process steam ( . The nominal operating pressure for the 7-1

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'l reactor is 2185 psig, with an average core outlet temperature of 604 F .

2 The reactor coolant system is designed for a pressure of 2500 psig at a 3 nominal temperature of 650'F(3) . The nuclear steam systems are essentially 4 identical to eight _otber B&W units which have previously received construc-5 tion permits.

6 The reactor core is approximately 129 inches in diameter, with 7 an active height of ihh inches ( }. It is made up of 177 fuel assem-8 blies, each consisting of a 15 x 15 array of fuel rods held in place 9 by mechanical spacer grids. The array of fuel rods consists of 208 10 cirealoy tubes containing uranium dioxide,16 control rod guide tubes, 11 and a center tube available for an in-core instrumentation assembly .

12 There are approximately 207,500 pounds of uranium dioxide in the

( 13 core (6) ,

14 ~The thermal and hydraulic design limits of the core are conser-15 vative, and are consistent with those of other pressurized water re-16 actors currently in operation or under construction '

17 Core reactivity is controlled by a combination of h9 movable 18 control rod assemblies, a neutron absorber dissolved in the coolant, 19 and burnable absorber rod assemblies ( }. The control rods are an ,

l 20 alloy of silver-indium-cadmium cncapsulated in stainless steel. The l l

21 dissolved neutron absorber is boric acid. The burnable absorber rods I 22 are sintered A1 0 -B C en apsulated in stainless steel (9}

23 g .

-23 Eight part length axial power shaping rod assemblies are pro- ,

2h vided to thwart any tendency toward axial power shifts resulting from 25 a redistribution of xenon O'10) .

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~l- The control rods are used for short-term reactivity con-2 trol associated with the changes in power level and also with 3 changes in fuel and burnable absorber burnup between periodic ad-4 justments of dissolved boron concentration (ll) . The reactor can be 5 shut down by the movable control rods from any power level 5 at any time. Dich movable control rod assembly and axial power 7- shaping rod assembly contains 16 control rods, and is actuated 8 by a separate control rod drive mechanism mounted on the top 9 head of the reactor vessel. Upon trip, all control rod assem-10 blies fall into the core by gravity causing immediate reactor 11 shutdown.

12 The control rod drive and axial power shaping rod drive s 13 mechanisms are sealed drives of the roller nut type in which a 14 lead screw coupled to the control rod assemblies is axially driven 15 by the rotary motion of a pair of roller nut segment arms.

16 The segment arms which are within the pressure housing are part of 17 the drive motor and are electrically driven by the motor stator 18 which is external to the pressure housing. The segment arms are 19 held in engagement with the lead screw whenever the drives 20 are electri.sily energized. The reactor trip signal or loss of 21 power to the control rod drives causes the roller nut se pect 22 arms to disengage from the lead screw causing the rod assemblies 23 to fall into the core. The axial power shaping rod drive

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1 mechanism is modified so that the roller nut assembly will not dis-2 engage from the lead screw on reactor trip or loss of electrical 3 power .

h Seventy-two of the fuel assemblies will be utilized as locations

13) Each assembly.has 16 burnable 5 for burnable absorber rod, assemblies .

6 absorber rods, a stainless steel spider, and a coupling mechanism for 7 positive coupling with the fuel assembly hold-down latch ( ,

8 The concentration of dissolved neutron absorber in the 9 reactor coolant may be adjusted to control relatively slow 10 moving reactivity changes and to provide a safe shutdown margin

.11 where a cooldown to reactor building ambient temperature is 12 required ( .

13 The core is contained within a cylindrical reactor vessel having Ik internal dimensions of lh feet 3 inches in diameter and 37 feet h 15 inches in height. The vessel has a spherically dished bottom head 16 with a bolted removable spherically dished top head . The re-17 actor vessel is constructed of carbon steel with all interior surfaces 18 clad with austenitic stainless steel. The reactor vessel is manu-19 factured under close quality control, and several types of nondestrue-20 tive tests are performed during fabrication. These tests include 21 radiography of welds, ultrasonic testing, magnetic particle exardn- .

22 ation, and dye penetrant testing . Specimens of reactor vessel 23 materials vill be placed in the reactor adjacent to the inside surface 24 of the reactor vessel. During operation, these specimens will be T-4 mm, w- --w 7

1 subject to irradiation similar to the shell of the reactor vessel. A 2 portion of the specimens will be removed periodically and tested to 3 ascertain the effects of radiation on the reactor vessel material .

k The two coolant loops are connected to the reactor vessel by 5 nozzles located near the top of the vessel. Each loop contains one 6 steam generator, two motor-driven coolant pumps, and the interconnec-7 ting piping. The reactor coolant piping is carbon steel clad on the 8 inside surface with austenitic stainless steel (19) . Reactor coolant 9 is pumped from the reactor through each steam generator and back to 10 the reactor inlet by two 88,000 gpm centrifugal pumps located near 11 the outlet of each steam generator ( } . The reactor coolant pumps are 12 vertical, single-stage, shaft-sealed units having bottom suction and 13 horizontal discharge. Each pump has a separate, single-speed, top-14- mounted motor, which is connected to the pump by a shaft coupling.

15 The steam generator is a vertical, straight-tube-and-shell Leat 16 exchanger which produces superheated steam at constant pressure over 17 the power range. Reactor coolant flows downward thraugh the tubes 18 and steam is generated on the shell side ( 1) .

19 The pressurizer, a vertical surge tank approximately half-filled 20 with reactor coolant and half-filled with steam, is connected to the 21 reactor coolant system. The operating pressure of the system is main-22 tained by, operating electric immersion heaters to increase pressure or

23. by spraying reactor coolant water into the steam within this pres-2h surizer tank to reduce pressure. Self-actuated safety relief valves

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1 connected to the pressurizer prevent overpressurization of the reactor 2 cool ant system .

3 The reactor vessel, steam generator, and pressurizer will be de-h signed, manufactured and tested in accordance with Section III of the 5' ASME Code. 'I'he reactor coolant piping will conform to USASI B317 and 6 the reactor coolant pump casings will be manufactured in accordance 7 with Section III of the ASME Code, where applicable 3) ,

8 73 Reactor Building 9 The reactor building is designed to completely enclose the re-10 actor coolant system and portions of the auxiliary and engineered 11 safeguards systems. It is a reinforced concrete structure with a 12 post-tensioned system in the shape of a cylinder with an elipsoidal-( 13 domed roof and a flat foundation slab. The cylindrical vall and dome 1k are prestressed by a post-tensioning system, consisting of steel ten-15 dons. The building will have three vertical and horizontal top and

~6 1 bottom buttresses to which tendons will be anchored. The foundation 17 slab is conventionally reinforced with high-strength reinforcing 18 steel.. The inner surfaces of the entire structure are lined with 19 welded steel plate,1/h-inch minimum thickness, which functions as {

20 a leak-tight membrane. The foundation mat will be bearing on the j 21 hard, stiff blue clay and will be approximately nine feet thick. I 22 The building is designed to sustain safely all internal and ex-23 ternal loading conditions which may be expected to occur during the 24 life of the station or which could result from postulated accidents 25 to the reactor's primary coolant system (24. The tendon system used 7-6

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i 1 in the structure is of the unbonded type with a protective compound 2 used to prevent corrosion.

3 The reactor building is so designed that, with the engineered h safeguards systems provided, any leakage of radioactive materials to 5 the environment vi .1 be well within the AEC's guidelines (10 CFR 100) 6 for any of the postulated accidents. The integrated leak rate at 7 design pressure will not exceed one-tenth of one percent by volume, 8 within 2h hours. Furthermore, the reactor building will be designed 9 to resist flotation due to the maximum probable flood.

10 Prior to operation, the reactor buildin6 will be subjected to 11 a structural integrity test and leak rate test. The structural in-12 tegrity test will be conducted at 115 percent of design pressure.

13 Periodic leak rate tects will be performed to assure integrity of 14 the reactor building. A tendon surveillance program will provide 15 assurance that the tendons are free from hamful corrosion and that 16 excessive steel relaxation has net taken place .

17 Reactor building materials and workmanship will be inspected to 18 ensure compliance with appropriate codes, specifications, and standards.

19 Materials to be inspected and tested include concrete, liner plate, 20 prertressing system materials, hatches, penetrations, structural and 21 reinforcing steel .

22 Provisions have been included for in-service pressure testing of

., 23 perscrmel hatches and other penetrations (2I) .

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1 7.4 Engineered Safety Features 2 Engineered safety features for each nuclear unit are provided 3 to prevent or mitigate the effects of accidents which result in 4 release of radioactivity.

5 The engineered safety features fulfill the following functions 6 in the unlikely event of a serious loss-of-coolant accident:

7 a. Protect the fuel cladding.

8 b. Maintain reactor building inegrity.

9 r. Reduce the driving forces for reactor building leakage.

10 d. Remove fission products from the reactor building atmosphere.

11 The engineered safety features can be grouped into emergency 12 core cooling systems, reactor building cooling systems, and fission

( 13 product control systems ( .

14 The emergency core cooling systems contain both passive flooding 15 systems and pumping syste=s. The passive flooding system consists 16 of two pressurized core flooding tanks which automatically discharge 17 borated water into the reactor vessel in the event that the reactor 18 system pressure drops below 600 psi. The pumping system consists 19 er two completely independent subsystems. Each subsystem contains 20 both a high-pressure and a low-pressure injection pump. Either sub-21 system, in conjunction with the core flooding tanks, is capable of 22 protecting core integrity for any size leak up to and includir4 a 23 double-ended rupture of the largest reactor coolant pipe. Either 24 subsystem can supply coolant directly from the borated water storage 25 tank or by recirculation from the reactor building sump through heat 26 exchangers which cool it before it is returned to cool the core (29) ,

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1 The reactor building cooling system, which is made up of two 2 separate and indenendent heat removal systems, limits the uressure 3 in the reactor building following a loss-of-coolant accident. One h system contains four separate air cooling units. The other system 5 contains redundant spray headers which spray low temperature horated 6 water into the reactor building to cool it. The spray water also 7 contains chemical additives as described below. Each of these sys-8 tems without the other has the heat removal capability to maintain 9 the reactor building pressure below its design pressure level (30) ,

10 The primary control of released fission products following a 11 pcatulated loss-of-coolant accident is provided by the reactor build-12 ing. A second means for controlling released fission products is the

( 13 iodine removal spray system which consists of redundant subsystems lh which utilize chemicals mixed in the reactor building spray water 15 to absorb the iodine released from the reactor coolant system and 16 render it unavailable for leakage from the reactor building. For the 17 postulated maximum hypothetical accident the reactor building and 18 either one of the two redundant reactor building spray systems will 19 limit radiation exposures at the exclusion boundary and low population 20 zone boundary to values within AEC safety guidelines (10 CFR 100)(31) ,

21 75 Instrumentation and Control 22 Redundant networks of instrumentation and controls will be pro-23 vided to ensure the safe operation of the Midland Plant. The reactor 24 protection system and the engineered safeguards actuation system are 25 designed to meet the requirements of the proposed " Standard for Nuclear 26 Power Plant Protection Systems," IEEE-279, Rev 10. The engineered T-9 1

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1 safety features actuation system monitors plant conditions and 2 automatically initiates operation of tue engineered safety features 3 systems, if required (32) . This system initiates emergency core 4 cooling systems on signal of high reactor building pressure or low 5 reactor coolant system pressure. The reactor protection system 6 monitors parameters related to safe operation and shuts down the 7 reactor if an operating limit is reached . The operating limits 8 are well below the safety limits. Shutdown will be accomplished by 9 interrupting power to the control rod drive mechanisms and allowing 10 the contre! rods to drop into the reactor core. Alarms (34) are pro-11 vided to alert the operator to abnormal operating conditions, and 12 interlocks (35) are provided to prevent operations which could lead

[ 13 to potentially hazardous conditions.

14 The nuclear instrumentation system monitors reactor power 15 from start-up level through 125 percent of full power operation.

16 There are separate overlapping instrumentation channels for the 17 source range, the intermediate range, and the power range .

18 Following proven power station design philosophy, all control 19 stations, switches, controllers, il indicators necessary to start 20 up, operate, and shut down the nuclear unit will be placed in the 21 centrally located control room. There will be sufficient information 22 display and alarm monitoring to ensure safe and reliable operation v 1 1

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k 1 under normal and accident conditions. Special emphasis will be given 2 to maintinaing control room integrity during accident conditions (37) ,

3 For the unlikely contingency that occupancy of the control roon k .may be temporarily denied, the capability will be available for taking 5 the plant to, and maintaining the plant in a safe shutdown condition 6 from other locations in the plant .

7 The instrumentation signals sent to control and safety circuits 8 from common transmitters are made fully independent by the use of iso-9 lation amplifiers. The effectiveness of these isolation amplifiers 10 has been demonstrated by analysis and by actual test of prototype 11 equipment.

12 A radiation monitoring system monitors radiation levels in selected 13 plant areas and in plant effluent released to the environment, provides l

1k an early warning of possible equipment malfunction or potential radiolog-15- ical hazard. The radiation monitoring system inclu u continuous auto-16 matic monitoring and is supplemented by periodic sampling (39) ,

17 7.6 Electrical Systems l

18 The Midland Plant will generate electric power at 22 kV and 24 kV, 19 respectively. This power will be red through separate isolated phase 20 buses to the unit main transformers where it will be stepped up to 345 kV 21 transmission voltage and delivered to the switchyard on separate over-22 head lines. The units and associated switchyard are part of the 1

23 Consumers Power Company integrated electric system, which is described 24 in Section 2.1 of this sunuary.

25 The design of the electrical systems for the Midland Plant units 26 is based on providing the required electrical equipment and power I

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-x 1 sources to ensure safe, reliable operation and safe, orderly shutdown 2 of the unit under any normal or emergency conditions. The following 3 sources of power are available to provide redundancy and to assure a 4 supply of electrical energy to the station safety systems under all 5 credible circumstances :

6 1. The five 345 kV transmission lines which terminate at the 7 Midland Plant switchyard provide start-up and standby power 8 through the 345-138 kV step-down substation,138 kV Start-Up 9 Line and Start-Up Transformer No 1.

10 2. Start-Up Transformer No 2 provides an alternate off-site 11 power source via a separate 138 kV line from a consumers 12 Power Company substation serving Dow.

( 13 3 Each unit will continue to supply power to its own aux 11-1h iary systems in the unlikely event of a trip separation 15 from the transmission system.

16 4. Upon loss of all sources of power described in 1, 2 and 17 3 above, power will be. supplied from the two automatic, 18 fast start-up diesel engine generators. These are sized 19 so that either can carry the engineered safeguards load for 20 one plant plus the safe shutdown load for the other unit.

.21 5 The station battery is sized to provide a safe. and orderly 22 shutdown of each unit in the event that all a-c power is 23 lost.

24 These normal, standby and emergency sources of auxiliary elec-25 trical power supply redundant plant auxiliary, engineered safeguards i -

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2 distribution systems.

3 77 Auxiliary Systems 4 Auxiliary systems are provided to supply reactor coolant make-5 up and pump seal water, to cool the reactor during shutdown, to cool 6 components, to ventilate station spaces, to handle fuel, to cool spent 7 fuel, and to adjust the concentration of various chemicals in the re-8 actor coolant ( 1) .

9 Reactor coolant makeup and seal water is supplied by the makeup 10 and purification system. This system, which also serves the engineered 11 safeguards function of providing high-pressure emergency core coolant, 12 maintains the proper coolant inventory in the primary system, maintains 13 the seal water flow, adjusts the concentration of dissolved neutron 14 absorber in the reactor coolant and maintains proper water chemistry } .

15 The decay heat removal system cools the reactor when the reactor 16 system is depressurized for maintenance or refueling. This same sys-17 _ (.ces the engineered safeguards functions of providing low-pressure 18 emergency core coolant and of recirculating borated water to cool the 19 ' core in the unlikely event of a loss-of-coolant accident .

20 The chemical addition system adds boric acid to the reactor coolant 21 system for reactivity control, potassium hydroxide for pH control, and 22 hydrogen and hydrazine for oxygen control .

23 The cooling water systems maintain temperatures throughout the 24 equipment and structures of the station . Appropriate normal and 25 emergency ventilation systems are provided in the plant .

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1 A fuel handling system provides the means for safe, reliable 2- handling of fuel until it is shipped from the station as used fuel.

3 Irradiated fuel is handled under water at all times until after it is 4 placed into a shipping cask. The water provides a radiation shield 5 as well as a reliable source of cooling for the irradiated fuel assem-6 blies. A fuel pool cooling system maintains the temperature and purity of the spent fuel storage pool water within acceptable limits }

7 .

8 7.8 Steam and Power Conversion System 9 The steam and power conversion system is designed to accept 10 steam from the nuclear steam system. One portion of this heat 11 energy is converted to electrical energy by the turbine generators.

12 A second portion of the heat energy is used in the process steam

/ 13 reboilers to generate process steam for Dov. The circulating 14 vater system, utilizing cooling pond water, vill dissipate the 15 balance of the heat energy which is rejected by the turbine 16 condensers.

17 79 Process Steam System 18 Steam from the steam and power conversion systems is used in 19 indirect tertiary heat exchangers (reboilers) to generate process

-20 (tertiary) steam for Dov. The function of the rebo11ers is to pro-21 vide complete physical separation between the turbine plant cycle and 22 the process steam delivered to Dow(' }. ,

23 7 10 cooling Pond l 24 Turbine condenser cooling water is stored in a cooling pond. The  !

25 pond is sized to allow for all water losses from the pond, without make-26 up from the Tittabavassee River, during the three-month period when the 1

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1- river flow is low ( . The emergency service water pond is a depressed 2 area in the cooling pond, which will provide water storage for re-3 actor decay heat removal in the unlikely event of a dike failure (52) ,

h 7 11 Radioactive Waste System 5 Radioactive gaseous, liquid and solid wastes in the station are 6 handled by the vaste disposal systems. These systems contain the 7 equipment necessary to safely collect, process and prepare for dis-8 posal or reuse of the radioactive vastes which result from reactor 9 operation. These systems are designed to minimize the release of 10 radioactive material from the plant to the environment and will main-11 tain releases well below the limits of 10 CFR 20(53) ,

12 The ~ gaseous and liquid radvaste releases from the plant are an-13 ticipated to be as lov or lower tLan other plants in operation or 14 under construction, except that the 11 quid radvaste release vill not 15 be as low as that of the Rancho Seco Plant which is not situated near 16 a waterway and cannot discharge liquid vastes. The maximum expected 17 radvaste release rates from the plant are as follows:

18' Gaseous 88,600 Curies / Year (30-DayHoldup) 19 Tritium 6,300 Curies / Year

.20 Other Liquid Wastes 0.1 Curies / Year 21 The above release rates are based upon 1 percent failed fuel, up to 22 1 gpm leakage into the turbine cycle and 30 percent tritium release from 23 the fuel . The quantities for failed fuel and steam generator 24 leakage are based upon current operating experience representing max-25 imum levels that might be expected. The quantity level shown for tritium

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1 release is a conservatively high assumed value pending further op-2- erating experience with the current generation of chemical shim, 3 zirealoy fuel-clad PWR. Some estimates have been as low as 1 percent 4 with a corresponding decrease in the tritium vaste discharge rate.

5 T.12. Shielding 6 Shielding throughout the plant, primarily in the form of heavy

.7 concrete walls, insures that radiation exposures to the general public 8 and to operating personnel are within the limits prescribed by the 9 AEC (10 CFR 20)(55} .

References

1. PSAR, Table h-5

2. PSAR, Table 3-1 3 PSAR, Section 4 3 1

4. PSAR, Table 3-2 5 PSAR, Table 3-1

6. PSAR,. Table 1-2 7 PSAR, Section 3 2 3

8. PSAR, Section 3 2.2; Table 3-2 9 PSAR, Section 3 2.4 3 7; Table 3-22

10. PSAR, Section 3 2.h.3 6; Figure 3-59

~

11. PSAR, Sections 3 2.2.1.2, 3 2.4 3 5; Figure 3-58

12. PSAR, Section 3 2.4 3 2 13 PSAR, Table 3-22; Figure 3-4

14. PSAR, Seetion 3 2.k.3.T '

15 PSAR,. Sections 3 2.2.1.2, 3 2.2.1 3 b; . .

16. . PSAR, Section h.2.2.1 and Table h-3 7-16 v

P s

17 PSAR, Sections 4.5 1 and 4.5 2

18. PSAR,-Section 4.4 3 19 PSAR, Table 4-12

20. PSAR, Section 4.2.2.4 and Table 4-7

21. PSAR, Section 4.2.2 3 and Table h-5

22. PSAR, Section k.2.2.2 and Table 4-4 23 PSAR, Section 4 5  !

24. PSAR, Section 5 1.1.2 25 PSAR, Section 5 1.4 3

26. PSAR, Sections 5 1 3 and 5 1.4 27 PSAR, Section 5 1 5 3

28. PSAR, Section 6 29 PSAR, Section 6.1.2

30. PSAR, Sections 6.2 and 6 3

31. PSAR, Section 14.2.2.4

32. PSAR, Section 7 1 33 Ibid

34. PSAR, Section 7.6 3 35 PSAR, Section 7 2 3 2 36.~ PSAR, Section 7 3 1 37 PSAR, Section T.6.5

38. PSAR, Items 716 and 719 (Following Amendment No 5, volume II)

39. . PSAR, Section 7 5

40. PSAR, Section 8-

. 41. PSAR, Section 9 T-17

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42. PSAR, Section'9 1 43 PSAR, Section 9 3

44. PSAR,'Section 9 2 45 PSAR, Sections 9 5, 9.6, 9 7 and 9 8

46. PSAR, Section 9 12 47 PSAR, Section 9 9

48. PSAR, Section 9.4 49 PSAR, Section 10.

50. PSAR, Item 11.00 (Following Amendment No 9, volume III)

51. PSAR, Section 10.2 5

52. PSAR, Section 9 7 2 53 PSAR, Section 11

54. PSAR, Item 11.1 (Following Amendment No 5, volume II) 55 PSAR, Section 5.4.2

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1 8.0 SAFEfY ANALYSIS 2 Potential malfunctions or equipment failures have been analyzed 3 to provide a safety evaluation of the Midland Plant. The results of 4 these analyses show that the effects of the radioactivity released to 5 the environment, even in th2 very unlikely event that such an accident 6 should occur, are well within the guidelines established by the AEC 7 (10 CFR Part 100)( .

8 Two categories of malfunctions or equipment failures have been 9 analyzed - those in which the fuel rods and reactor coolant system 10 pressure boundary are protected and those in which one of these bar-11 riers to the release of fission products is not effective and engi-12 neered safeguards are required. The core and coolant boundary pro-( 13 tection analysis analyzes those abnormalities that are either inherently 14 terminated or require operation of the normal protection systems in order 15 to maintain the integrity of the fuel rods and/or the reactor coolant 16 system (2) . The standby safeguards analysis analyzes those accidents 17 in which one or more of the nominal protection systems are not effective 18 and therefore requires the operation of standby safety features (3) ,

19 The results of these analyses show that the effects of radioactivity 20 releases to the environment for all credible malfunctions are well with-l 21 in the AEC safety guidelines (10 CFR 100).

22 Of the numerous malfunctions evaluated, a loss-of-coolant accident 23 would be the most severe. Emergency core cooling systems are provided 24 to prevent clad melting for entire spectrum of reactor coolant system 25 failures ranging from the smallest leak to the complete severance of

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1 the largest reactor coolant pipe. The core cooling systems insure 2 that core integrity will be maintained } . Either the reactor building 3 sprays or the air recirculation and cooling system will maintain the

.4 integrity of the reactor buildiag, assuring that the public will be 5 protected against potential radiation hazards from the postulated 6 accident (5) . Emergency electrical power is available on-site to T insure operation of these systems even if all external sources of elec-

-8 tric power to the plant are assumed to be unavailable at the time of 9 the accident .

10 Results of the safety analyses show that, Sven in the event of a 11 loss-of-coolant accident, no core melting will occur (5} . However, in 12 order to demonstrate that the operation of a nuclear power station at 13 the proposed site does not present a hazard to the general public, a 14 "caximum hypothetical accident" has been analyzed assucing a release 15 from the fuel of 100 percent of the noble gases, 50 percent of the halo-16 gens and 1 percent of the solids in the fission product inventory. Fifty 17 percent of the halogens then further plate out. To have such a gross 18 release of fission products, one must postulate a multitude of failures 19 in the engineered safeguards systems (I) , therefore, the accident is I 20 not regarded as credible. Even given these assu=ptions, however, the 21 low leakage rate of the reactor building and the iodine removal spray 22 system limit the potential radiation dose to the thyroid and the whole l 23 body to well below the AEC safety guideline values (10 CFR 100)( .

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-References

1. PSAR, Section 14 i

2. -PSAR, Section 14.1

'3 PSAR, Section 14.2-

4. PSAR, Section 6.1

5. -PSAR, Section 14.2.2 3 2

6. PSAR, Section 8.2 3 7 PSAR, Se'etion 14.2.2.4-

8. -PSAR,. Table 14 (

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L 1' 90 COMPARISON WITH OTHER PWR 2 Appendix A is a comparative list of important design and operating 3'

parameters of the Midland Plant and of several other nuclear units which 4 are scheduled to precede the Midland Plant into operation. The other 5

units are Oconee Unit 1, Rancho Seco Unit 1, and Zion Units 1 and 2.

6 The Midland, Oconee, and Rancho Seco units are being supplied by B&W;.

7 the Zion units are being supplied by Westinghouse.

8 The-initial core rated power, nominal operating' pressure, operating 9 temperature, total coolant flow rate, average thermal output at rated 10 power, average heat flux at rated power, minimum departure from nucleate 11 boiling ratio (DNBR), and number of fuel assemblies are identical for 12 the B&W units. The lesser number of control rod assemblies in the Midland 13 units compared to the other B8W units is the result of a reduction in 14 requirements for inserted control rod assemblies for equilibrium and trans-15 ient xenon control. The differences in fuel burnups among the B&W plants 16 is primarily the result of differences in plant operating requirements 17 specified by the utilities involved.

18- A comparison of the number of coolant loops, coolant pumps, steam 19 generators, and engineered safeguards features is also included. The l 20 loop arrangements and emergency core cooling systems capabilities for the 21 B&W units are identical. Though the loop arrangements for the B&W and l 22 the Zion units are basically different, the emergency core cooling systems 23 for all the units are generally similar.

24 The design of the Midland units compares favorably with the design 25 of the other units presented. Similarities with other reactor building  !

m 26 designs are compared in Section 3 of this summary. In addition, certain 27 safety features of the Midland units are similar to those of the Zion and l

l 9-1

- ~ - - - .

. . . . - i

.(-

1 Indian Point units. These include chemical reagent sprays and pro-2 vision for charcoal filters, failed fuel detection, provisions to

-3 flood the reactor vessel cavity after a loss-of-coolant accident, 4 penetration pressurization equipment, equipment to handle radiolysis 5 of water after accidents and means to withstand mechanical forces and 6 pressure transients in the vessel cavity.

7 The design of each of the units is based on infomation developed 8 from operation of commercial and prototype pressurized water reactors 9 over a number of years. The Midland Plant designs are based on this 10 existing power reactor technology and have not been extended beyond the 11 boundaries of known information or operating experience.

12 Prior to the time Midland Unit 1 is scheduled to go into commercial 13 operation, more than ten operating years' experience is to be available 14 from the eight B&W units of essentially identical design which are 15 scheduled to precede it into operation.

16 In addition, the safety of the Midland units for operation at their 17 design power level is further attested to by the fact that the ACRS in its 18 letter to Chairman Seaborg of September 23, 1970, stated that the com- 6 19 mittee had completed its operating license review of oconee Unit 1 and 20 concluded that there is reasonable assurance that Oconee Unit 1 can be 21 operated at power levels up to 2568 MW . The design of the Midland t

22 reactors .is essentially identical to the Oconee 1 reacto-c design.

9-2

-. = - - -

3, ,

1 10.0 ENVIRONMENTAL-MONITORING PROGRAM 2 10.1 General Radiological Surveillance 3 A preoperational environmental radiation survey will be con-

-4 ducted to determine the existing levels of radioisotopes in the environ-5 ment surrounding the plant, to note the use of the environment in the o area and to sample those substances that are either directly consumed 7 by man or involved in the ecological food chain of man.

8 Both the preoperational and the operational monitoring programs 9 vill sample air, water, aquatic life, including fish, and river sediment.

10 External gamma exposure to man will be estimated by placing both film 11 and thermoluminescent dosimeters in the environment. Discharges will 12 be sampled and monitored before leaving the plant.

13 Gross beta analyses will be routinely performed on all samples.

14 When gross beta counts exceed predetermined levels for each type of 15 sample, however, additional specific isotope analyses will be performed.

16 All river water samples will be analyzed for tritium and all air samples 17 will be analyzed for I-131. Air samples will be collected on a weekly 18 basis, and other materials will normally be collected monthly.

19 A reference area method of general radiological surveillance 20 vill be employed. This type of surveillance program utilizes a statis-21 tical comparison between sample analysis results of two sets of sta-.

22 tions. ,0ne set of eight stations, commonly called the inner ring, 23 will be located within about two miles from the plant, while another 24 set of three stations, called background stations, will be located 25 approximately 20 miles from the plant. Statistical differences between 7

10-1

. . ~ . ... - . .. . - -

s 1 the survey results at the inner and outer rings will permit a deter-2- .mination as to whether any increase in radioactivity is of plant origin, 3 and will permit an estimate to be made of any dose to man or buildup 4 in radioactivity in the food chain. The survey will be sufficiently 5 flexible to adjust to new environmental ,onditions and to take into 6 account knowledge gained from the sample cellection and analyses.

7 Tentatively 5 additional air sampling stations and 12 additional 8 thermoluminescent dosimeter stations are planned to be added at the 9 time the plant goes into operation.

10 10.2 Aquatic Ecological Surveillance 11 An ecological survey is planned to detect oossible effects of 12 the cooling pond discharge or other plant vastewaters on the Titta-13 bawassee River. Preoperational surveys will be conducted to provide 14 baseline data for comparison with conditions after the plant is in 15 operation.

16 Fish and bottom organisms populations will ce sampled at each 17 site as indices of ecological conditions, since they show not only 18 direct response to environmental change but also reflect variations 19 in productivity and other biotic parameters.

20 Sampling for the effects on the ecology of the Tittabawassee 21 River will be done twice a year; in spring and in late summer or fall.

22 These dates will probably be made to coincide with two of the dates 23 set for radiological sampling. The spring sampling period will 24 indicate the degree of change from the previous years and will sample 25- the fauna at a time when maximum seasonal benthic diversity is expected.

( -

l 10-2

c.

'k 1 . The late summer or fall sampling vill be done i= mediately after the 2 most critical periods'of low water and highest temperature and before 3 sufficient time has passed for recovery from previous summer weather

h. to mask the effects of these most adverse natural conditions.

5 To facilitate the ecological studies, ecological surveillance 6 sites are, with the exception of one additional station upstream of 7 the pond discharge, at sites which are designated for aquatic radio-8 logical surveillance. Four stations are upstream of the pond dis-9 charge and four stations are on the Tittabawassee River below the 10 point of discharge.

11 It is anticipated that at least two years of preoperational 12 data vill be collected, and that the studies will be continued after

( 13. the plant is in operation. The operational surveys will essentially 14 be a continuation of the preoperational surveys in order to easily 15 compare data.

16 10 3 Aquatic Radiological Surveillance 17 A special aquatic radiological survey is planned to determine le (1) the -background levels of radiation in the aquatic invertebrates 19 and fish in the Tittabavassee River and in Saginaw Bay before the 20 Midland Plant is in operation, (2) the sources, if any, of radionu-21 clides discharged into the Tittabawassee River and Saginaw Bay before 22 the plant is in operation, (3) the amounts and types of radionuclides 23 in the Tittabavassee River and Saginaw Bay waters after the plant 24 is in operation, and (4) the concentration factors and types of 25' radionuclides found in aquatic invertebrates and fish in the Titta-26 _ bawassee River and Saginaw Bay after the plant is in operation.

10-3

1 The radiological surveillance program will include analyses of 2 samples of water, bottom sediment, bottom organisms and fish at each 3' site. Except for the Saginaw Bay stations, samples will be collected 4 four times a year, approximately on a quarterly basis. In Saginaw Bay, 5 samples will be taken biannually (spring and fall) and will include, in 6 addition to the other parameters, samples of the various species of 7 resident waterfowl.

8 Three sampling stations are on tributaries above the Midland Plant, 9 -to provide data on upstream conditions for the reference analysis; five 10 stations are on the Tittabawassee River below the plant; two stations are 11 on other tributaries to the Saginaw River, which would not be affected 12 - by plant discharges, but are necessary to inventory radioactivity charac-s 13 teristics further downstream; one station is on the mouth of the Saginaw 14 River, and approximately 15 stations are in Saginaw Bay. Samples will 15 also be collected within the plant cooling and storage pond.

16 If the preoperational survey reveals any significant pattern in 17 the radiological parameters, changes in the program, such as the elimina-18 tion of particular samples or the inclusion of additional samples, may 19 be considered.

1 Reference

1. PSAR, Amendment No 5, Item 2.1 (T_

l 10-4

1 11.0 TECHNICAL' QUALIFICATIONS 2 11.1 Consumers Power Company 3 Applicant has over h8 years of experience in the operation of 4- conventional electric generating plants. Applicant's General Office 5 -staff includes a number of persons associated with the Midland Plant 6 project who have previous experience in the nuclear field, including 7 experience in the design, construction and operation of applicant's

~8 75' megawatt boiling water nuclear plant at Big Rock Point near 9- Charlevoix, Michigan and the design and construction of applicant's

-10 2,h50 megawatt thermal Palisades Plant near Covert, Michigan.

11 The operating staff will consist of approximately 78 full-time 12 employees ( . The majority will come from existing generating o 13- plants on the Consumers Power Company system.

14- The major training for supervisory personnel will take place 15 at the Big Rock Point and Palisades Plants. Consumers Power Company 16 personnel vill participate in the design phases of the Midland Plant 17 and may also receive simulator training at a B&W training facility.

18 11.2- Bechtel 1

1 19 Bechtel Corporation and its affiliate, Bechtel Company, have i 20 .been retained by Consumers Power Company' as Architect / Engineer and 21 Constructor for the Midland project.

22 Working under-the overall direction of Consumers Power Company '

23 and in close coordination with B&W, Bechtel will design or furnish, 24 or both,' all portions of the plant except the turbine generators and .

25 the equipment and fuel supplied by B&W. Bechtel will also construct

)

26 the_ entire plant.

11-1 ,

m- -

---n,n

,m 1- Bechtel has been continuously engaged in construction or engi-2 neering activities since 1898. For more than 20 years, Bechtel has 3 been active in the fields of petroleum, power generation and distri-h bution, harbor development, mining and metallurgy, and chemical and 5 industrial processing.

6 Since the close of World War II, Bechtel has been responsible 7 for the design of over 166 power generating units, representing 8 more than 43 million kilowatts of thermal generating capacity, which 9 includes units of the largest and most modern types. Of this number, 10 more than 15 million kilowatts are produced by nuclear-fueled units.

11 For over 20 years, Bechtel has been engaged in the study, de-12 usign and construction of nuclear installations. Its experience in-

.(

i 13 cludes design or construction, or both, of such facilities as accel-14 erators, nuclear research laboratories, hot cells, experimental 15 reactors and nuclear fuel processing plants, as well as nuclear power 16 plants. A summary of experience is listed in the application (3) ,

17 11 3 The Babcock & Wilcox Company (B&W) 18 B&W's participation in the development of nuclear power dates 19 from the Manhattan project. B&W's nuclear activities include applied 20 research to develop fundamental data, design and manufacture of nu-21 clear systems components, and design and manufacture of complete 22 nuclear steam generating systems. Through the B&W Company's several 23 divisions, a wide range of equipment for nuclear application is de-24 signed and manufactured. The B&W Company's major nuclear contracts, fg a

11-2

..u_ __.

1 'in addition to a' substantial percentage of components for the nuclear 2 Navy, have included:

3- Indian Point Unit 1 h NS Savannah 5- Advanced Test _ Reactor 6- .Oconee Nuclear Station Units 1, 2 & 3 7 .Three. Mile Island Units 1 & 2 8 crystal River Unit 3-

-9 Rancho Seco Unit 1

10. Arkansas Nuclear One Station Unit 1 11 Midland Units 1 & 2 12 Davis-Besse Unit 1 13 TVA Station (Undesignated) Units 1 & 2 14 With the exception of the last three contracts, all units have re-15 ceived construction permits.

r S-References

1. Midland Plant - Preliminary Safety Analysis Report, Section 12.1

2. Midland Plant - Preliminary Safety Analysis Report, Section 12.2 3 General Information, Appendix D k'. General Information, Appendix C 11-3

-, , - , - - - r -

I 1

l 1 12.0 QUALITY ASSURANCE AND QUALITY CONTROL

-2 12.1 Introduction 3 Applicant has initiated an extensive Quality Assurance (QA) 4 Program to help assure that the Midland Plant is desig -4 and

5. constructed so that it will operate efficiently and reliably and 6 without undue risk to the health and safety of the public and the 7 plant personnel. The QA Program encompasses all phases of the de-8 . sign, construction, and operation. A program plan ( ) has been 9 developed outlining how the QA Program will be implemented. Over-10- all responsibility for the program rests with applicant.

i 11- Bechtel vill perfom Quality Assurance of plant engineering, 12 shop fabrication (including the nuclear steam system, except for r

5-13 -fuel),fieldfabricationandconstruction. Bechtel is also respon-14 sible for providing Quality Control of receiving, storage, installation 15- and erection at the plant site.

16 The Babcock & Wilcox Company has a Quality Assurance Program 17 for.the nuclear steam system. The scope of the B&W program includes

~

18- the design, fabrication, shop testing and shipment of the components 19- they furnish.

20 12.2 Consumers Power Company 21 Applicant will actively participate in a program or surveil- i 22 lance to assure that Bechtel and B&W implement their respective QA

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12-1

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1 Programs during the design, procurement and construction phases 2 of this project. Applicant's surveillance program vill include 3 audits, inspections and witnessing of tests at vendor shops. Audits 4~ vill also be performed on Bechtel home office activities and at 5' the plant site during construction.

6 The direction and coordination of applicant's QA Program from 7 design through construction is the responsibility of a Quality 8 Assurance Engineer (QAE). The QAE will conduct or supervise audits 9 and inspections of Bechtel home office, vendor shops and site audits.

10 Site surveillance is the responsibility of the applicant's Construc-11 tion Department Field QAE who functionally reports to the QAE. The 12 Field QAE and QAE will be supported by engineers from the Engineering, k 13 construction and operating Departments and other specialists within 4

14 the applicant's organization.

-15 A comprehensive field testing program for the Midland Plant 16 vill be conducted to insure equipment and systems perform in accord-17 an-a with design criteria. The applicant's Engineering Department is 18 respo aible for preparing preoperational test procedures and its 19 construction Department will be responsible for running the tests.

20 Quality Assurance of preoperational testing vill be performed by 21 applicant's Field QAE and QAE.

22. Applicant is preparing written procedures containing internal 23 instructions to employees with respect to performance of their respon- ,

24 sibilities under applicant's QA Program.

v 12-2

~.

r 1 12 3 Bechtel 2 The Bechtel Quality Assurance Program will be carried out 3 in accordance with Bechtel's Nuclear Quality Assurance Manual to 4- meet specific applicant requirements ( }. This manual describes 5 the overall program and identifies management and administratize 6 procedures and individual responsibilities. General instructions, 7 guidelines and checklists for inspections are contained in documents 8 developed for specific phases of the QA program.

9 1. Bechtel Procurement Department Inspection Manual.

10 2. Bechtel Field Inspection Manual.

11 3 Bechtel Field Procurement Procedures.

12 -Several levels of design review and approval are applied to C '13 significant design aspects of Bechtel work. These standard practices 14 include:

15 1. Checking and review of design work by members of the 16 Project Engineering team, other than those who originated 17 the work.

18 2. Review and approval by the Originating Engineer's 19- Design Group supervisor.

20 3 Review and approval by the Project Engineer.

21 4. Review and/or approval by the appropriate Chief Engineer 22 of certain key designs and calculations.

r 23 A design control. checklist'is prepared which identifies 24 drawings, specifications and other data which shall be reviewed by

, 'b i

12-3 l

1 Chief Engineers or Technical Specialists under the Chief Engineers' 2 direction. The Design Control Checklist is agreed on by the Project 3 Engineer and the cognizant Chief Engineers soon after initiation 4 of the project. -When the items identified in the Design Control 5 Checklist have been completed, the cognizant Chief Engineer will 6 -have the final review performed and execute a design control approval 7 signifying that necessary review and monitoring work have been com-8 pleted and the item is satisfactory from the standpoint of QA re-9 quirements.

10 Structures, systems and components to be covered by the QA Program 11 are identified by a Q-List which is prepared by the Project Engineering 12 team and reviewed and approved by Chief Engineers. Additional design con-

. 13 trol measures are implemented for items on the Q-List. The specific level lh of inspection and control afforded items on the Q-List are determined 15 by the Project Engineering tess through consultation with the Chief 16 Engineers and Bechtel's Technical Specialists. Factors considered 17 in establishing the degree of control include nature of the item, 18 importance of the item to plant safety and reliability, previous 19 experience with this or comparable items, capabilities of potential 20 vendors or subcontractors, requirements of applicable codes or standards 21 and provisions of the PSAR. Where required, Bechtel Engineering docu-22 ments include' specific procedures or requisites for the production and 23 QA of the item or Bechtel vill request their preparation by the organi-2k zation responsible for manufacturing or erection.

12-4

r 1 In implementing the program in the construction phase, QA related 2 . responsibilities are assigned to the following personnel:

3 1. The Project Field Engineer supervises Quality Control k (QC) inspection at the jobsite. In carryi,ng out this 5 assignment, he assigns qualified Field Engineers to 6 perform QC inspections, he supervises the preparation 7 of inspection checklists, verifies accuracy and 8 completeness of inspection reports, ascertains that 9 defects are removed and insures that repairs are 10- carried out in accordance with applicable specifica-

.11 tions, instructions and procedures.

12 2. The Quality Control Engineer reports to and assists the

( 13 Project Field Engineer in carrying out QC inspection 14 responsibilities. He normally is assigned responsibility 15 for review of inspection reports, coordination, training

. 16 and advising Field Engineers performing QC inspection 17 assignments, coordination of testing laboratories and  ;

t 18 overall detailed execution of field inspection and 19 maintenanceofthefieldQC/QAfiles.

20 -3. Field Engineer-Inspectors carry out the inspection 21 assignments and are responsible for completing the 22 appropriate inspection forms. Field Engineer-Inspectors t 23 function on a disciplinary basis, eg, mechanical equip-i 24 ment, civil-structural, electrical-power, instrumentation 25 control, velding-metallurgy. The number of inspectors

(?

U 26 assigned depends upon the requirements of the variable 12-5

.,- .,a, -.- . .-. - -- -

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1 QC inspection workload and construction schedule.

2 They have access to all design drawings, applicable 3 codes and sampling and testing procedures pertaining 4 to their inspection assignments.

5 4. The QA Engineer is the field representative of the 6 Project Engineering team and receives supervision and 7 technical support by the Project Engineer. However, 8 he is assigned by and administratively reports to the 9 Manager - Quality Assurance. He provides QA surveil-10 lance of engineering, QC and construction activities in )

11 the field. He reviews and accepts inspection reports,

U2 audits the permanent field QA/QC documentation files

( 13 and monitors the overall QC program. The QA Engineer

]

14 has the authority to stop the work for which Bechtel  !

15 Corporation is prime contractor in the event of noncom-1 l

16 formance with drawings, specifications and procedures i

17 established for structures, system and units on the Q-Idst. j l

18 He serves as field contact with a;1 'icant's Quality Assurance 19 organization and others concerned with Quality Assurance 20 in the field.

21 -12.4 The Babcock & Wilcox Company 22 B&W has a comprehensive Quality Assurance Program covering the design, 23 procurement, fabrication and testing of nuclear steam systems and fuel for 24 the Consumers Power Company Midland Plant ( . The program is administered

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f 1 by.a separate Quality Assurance organization whose responsibility is 2 to assure the implementation of the program. The program establishes 3 and maintains standards of quality through the development and imple-h mentation of requisite methods .and procedures. In addition, the pro-5 gram is organized and administered so as to meet the provisions of 6 the AEC's Quality Assurance Criteria for Nuclear Powerplants (3) ,

7 As a main contractor to Consumers Power Co=pany, B&W is respon-8 sible, through its project manager, for developing the necessary speci-9 fications and/or associated purchase documents, which include Quality 10 Control requirements, and for insuring that such requirements are 11 followed for the shop fabrication of the nuclear components in B&W's 12 scope of supply. B&W vill also develop recommendations for site

( 13 storage and installation of all such components.

14 The project manager is assisted in this effort by the manager of 15 B&W-Nuclear Power Generation Department (NPGD) Quality Assurance, who 16 reports directly to higher management to insure that he has sufficient 17 organizational freedom to identify problems affecting quality and to 16 insure that solutions are obtained. However, the project manager is 19 responsible for coordinating the planning and scheduling of the i 20 B&W-NPGD Quality Assurance effort with other organizations and con-1 21 tractors involved in the Midland Plant Project. Further, B&W will 1

l 22 provide erection consultation at the site throughout the period of 23- installation of the nuclear steam system. B&W will maintain a Quality i

)

24 history file for the work that B&W and its subcontracters perform and, l 25 soon after a component is shipped to the site, will furnish to applicant 26 copies of the portions of the file that apply to the component.

12-7

_.-.m ,;

1 BW's QA provisions for Midland Plant cover the design, pro-2 'curement, fabrication and testing of the BW nuclear steam systems, 3 and fuel, including necessary quality control requirements. The 4 policies and procedures specified in this Plan are coordinated with 5 the Quality Assurance activities of other departments of the Power 6 Generation Division.

References

1. Midland Plant PSAR, Appendix 1B

. 2. Midland Plant PSAR, Amendment 6

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3 10 cFR Part 50, Appendix B I

4 i i i_

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7-1 13 0- COMMENTS ON ACRS REPORTS 2' .In its reports of June 18, 1970 and September 23,'1970 to 3 the' Chairman of the Atomic Energy Commission favorably. reviewing 4 the Midland Plant for construction, the Advisory Committee on 3 Reactor Safeguards (ACRS) identified certain areas for further 6 consideration. These areas as they appear in the ACRS letters 7 are listed below, together with applicant's comments thereon.

f

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t 13-1

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f 1 13 1 "A large volume of liquid chlorine is maintained in a refrig-erated storage vessel about one mile from the Midland plant control room. The applicant is continuing his study of the consequences of a major accidental release of chlorine from this vessel. He has included in his criteria for the design of the control room the ob-jective of finding a practical method of maintaining the concentration of chlorine in the control room atmosphere below the eight hour thres-old limiting value (TLV) of 1 ppm for the most serious conceivable chlorine accident. The Committee believes that adequate air purifi-cation facilities should be provided in the control room ventilation system to reduce chlorine concentration to the eight hour TLV of 1 ppm so that operators can work without respiratory equipment during an extended. chlorine emergency. This matter should be resolved during construction in a manner satisfactory to the Regulatory Staff."

Answer:

2 Applicant will review this design during construction with the 3 Regulatory Staff. A system design vill provide for supplying air to k the control room through a filtration system so as to meet a chlorine 5 concentration criterion of 1 ppm and eliminate the necessity for

( 6 recpiratory equipnent.

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13-2

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"The applicant has stated that he will provide additional

~

1 13 2 evidence obtained by' improved multi-node analytical techniques to assure that the emergency core cooling system is capable of limiting core temperatures to the limits established at present. He will also make appropriate plant changes if the further analysis demon-strates that such changes are required. This matter should be re-solved during construction in a manner satisfactory to the Regulatory Staff. The Committee wishes to be kept infomed."

• Answer:

2 MW analyzed the ability of the Midland emergency core cooling 3 systems to limit fuel clad temperatures to acceptable values following 4 a loss-of-coolant accident and reported the results of these analyses 5 inthePSkR. Recent work by AEC consultants raised a question about 6 whether the analytical techniques used in the analyses reported in 7 the PSAR were detailed enough.

.8 CRAFT, MW's multinode computer code for analyzing loss-of-coolant 7

-( 9 accidents, will be used to reevaluate the Midland emergency core cooling 10 system'(ECCS) design. 'This work vill be completed in time to make any 11 necessary changes in the ECCS design, should the analysis show that the 12 current design will not limit the fuel clad temperature to acceptable 13 values.. i i

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13-3 a, .. . - .. u.. - .. . _, -.-

1 13 3 "The safety injection system for the Midland plant is actuated ,

by either low reactor pressure or high containment pressure signals.

However, of these two, the reactor is tripped only by the low reactor pressure signal. The Committee believes that provision also should be made to trip the reactor by the hiSh containment pressure signal."

Answer:

2 A diverse backup to the low-reactor-coolsnt-system-pressure 3 trip signal such as a power-flow, low-flow, or high-reactor-building-4 pressure trip signal will be provided for in the design of the Midland 5 reactors.

k.

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. 41 13-4

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1- 13 4 "The applicant plans to develop more detailed criteria for the installation of protection and emergency power systems together with appropriate procedures to maintain the physical and electrical independence of the redundant p rtions of these systems. The.

Committee believes that these criteria and procedures should be

! reviewed.and approved by the Staff prior to actual installation."

Answer:

2 Applicant will review the detailed criteria and procedures

-3 with the Staff and obtain its approval prior to actual installation.

f l

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I i

13-5

1 -13 5 "The applicant considers the possibility of melting and subse-quent disintegration of a portion of a fuel assembly because of flow starvation, gmss enrichment error, or from other causes to be remote.

However, the resulting effects in terms of local high temperature or pressure and possible initiation of failure in adjacent fuel elements are not well known. Appropriate studies should be made to show that such an incident will not lead to unacceptable conditions."

Answer:

2 B&W has completed a study of fuel and clad temperatures and 3 internal pressures for a fuel rod undergoing sustained departure 14 from nucleate boiling due to flow starvation. The results of that 5 study demonstrated that such an incident would not lead to unaccept-6 able conditions and were submitted to Division of Reactor Licensing 7 (DRL) in B&W Topical Report, BAW-100ll+, " Analysis of Sustained 8 Departure From Nucleate Boiling Operation."

y 9 A B&W study to examine the effects of misplacing a fuel 10 pellet, a fuel rod or a fuel assembly will be completed during 1971 11 and the results will be reported to the DRL staff.

13-6

i 1 13 6 "The committee believes that consideration shoulu 've given to the utilization of instrumentation for prompt detection of gross failure of a fuel element."

Answer:

2 Tne Midland Plant will employ instrumentation to promptly detect 3 gross failure of a fuel element. This instrumentation vill be similar 4 to that approved for the applicant's Palisades PJant. The instrumenta-5 tion vill continuously monitor the gamma radioac+1vity in the reactor 6 coolant after a sufficient decay time to reduce the concentration of 7 very short-lived, nonfission product radionuclides such as N-16, i'

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l 13 7 "The committee has commented in previous reports on the develop-ment of systems to control the buildup of hydrogen in the containment which might follow in the unlikely event of a major accident. The applicant proposes'to make use of a technique of purging through filters after a suitable time delay subsequent to the accident. However, the Committee recommends that the primary protection in this regard should utilize a hydrogen control method which keeps the hydrogen concentration within safe limits by means other than purging. 'Ihe capability for purging should also be provided. The hydrogen control system and pro-visions for containment atmosphere mixing and sampling should have redundancy and instrumentation suitable for an engineered safety fea-t *,re . The Committee wishes to be kept infomed of the resolution of

'is matter."

Answer:

2 Applicant has accepted the Committee recommendation to utilize 3 a hydrogen control method other than purging, in accordance with the 4 above. The method expected to be employed vill utilize hydrogen-5 oxygen recombiners.

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-l' 13 6 "The Committee recommends that the applicant accelerate the study of means of preventing common failure modes from negating scram action and of design features to make tolerable the consequences of failure to scram during anticipated transients. The applicant stated that the engineering design would maintain flexibility with regard to relief capacity of the primary system and to a diverse means of re-

.ducing reactivity. This matter should be resolved in a manner satis-factory to the Regulatory Staff during construction. The Committee wishes to be kept informed."

Answer:

2 B&W has completed a study of common failure modes. The results 3 of this study have been submitted to DRL in B&W Topical Report, 4 BAW-10019, " Systematic Failure Study of Reactor Protections Systems."

5 B&W has completed a study of anticipated operational transients 6 without trip and reported the results to DRL. Applicant has been'ad-7 vised that the Regulatory Staff is considering issuing guidelines for

( 8 the further study of operational transients without trip. B&W is 9 vaiting for DRL to issue the guidelines before proceeding further.

10 The Midland Plant design does not preclude the possibility of 11 increasing the primary system relief capacity or of providing diverse 12 means of reducing reactivity. However, the common failure modes and 13 anticipated transient without trip studies completed to date do not 14 indicate the need for either revision co the reactor system design.

(;

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1- 13 9 "Other problems related to large water reactors have been identified by the Regulatory Staff and the ACRS and cited in previous ACRS reports. Die Committee believes that resolution of these items should apply equally to the Midland Plant Units 1 & 2."

Answer:

2 The 'other matters referred to in this section of the ACRS 3- letter are discussed in Appendix C, "Research and Development Programs

!+ of Interest to'ACRS First Identified in Earlier Cases."

O d

t 13-10

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l 13 10 "The Committee believes that the proposed system' of reboilers will provide substantial additional assurance that leakage of primary system radioactivity into the export steam can be maintained at an .i extremely low and insignificant level and that the export steam can '

be maintained essentially at natural background levels. The detailed procedures for monitoring and control of the reboiler ' system should be developed during construction in a manner satisfactory to the Regu-latory Staff. .The Committee wishes to'be kept informed."

Answer:

2 Applicant vill develop detailed procedures for monitoring and 3 control of the reboiler system. These vill be developed during con-4 struction in a canner satisfactory to the Staff.

I 13-u j

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FIGURE IC")

MIDLAND PLANT UNITS I AND 2 ,

NUCLEAR POWER STATION PROJECT QUALITY ASSUR ANCE ORGANIZ ATION i

' CNART DELOW SHOWS RELATLONSHIP SETWEEN ENGINEERING. Mt0CUREMENT. AND CONSTRUCTION.

CONSUMERS POWER COMPANY SECHTEL BECHTEL

/ PROJECT ENG CONST SUPT \

C P Co QUAL ASSUR ENG

\ PROCUREMENT _ SUBCONTR-6 VENDORS

_ SHOPINSPEC _ M ATRL .SUPR.

/

METALLURGY & WELDING ,_________________j I SECTION

\ START-UP ENG ______---__-----s .

\ QUAL ASSUR ENG N DESIGN GP SUPVRS JOB F.NG 2.

• FIELD SUPTS e v TESTING L'ABORATORIES FIELD ENGINEERS coNem oc wCLOG cmtATRL NECH PIPING WE LDING ARCH ELEC INSTR r .,--v v - - ~ s m ym e -m m ~ g g- w

O '

s QUALITY ASSURANCE PROGRAM i

ENGINEERING PROCUREMENT DESIGN MA NUFACTUR ING CONSTRUCTION l

I CONSUMERS CONSUMERS g CONSUMERS p

1 N BECHTEL $$

EF BECHTEL E 5

p--- BECHTEL 3 ,

OIE 3 E l

• SUB l

BBW SI BaW

' El @3 -

3 I __ CONTRACTORS l

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OTHER l OTHER E l VENDORS VEN DORS l l

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AE C INSPECTION lI l-- AE C INSPECTION l STATE CODE  ! STATE CODE INSURANCE i INSURANC E

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i FIGURE -2

1

(-

1 14.0 FINANCIAL QUALIFICATIONS 2 Applicant has estimated that the total cost of the Midland 3 Plant, including the first-core fuel, is $394,827,000(1} . All items 4 included in the categories comprising the plant cost estimate are the

'5 'same as those which would be included in the relevant electric plant 6 and nuclear fuel inventory accounts of the Federal Power Commission's 7 Uniform System of Accounts prescribed for Public Utilities and Licen-8 sees'(Class A and Class B).

9 Applicant will finance the Midland Plant as an integral part of 10 its normal construction program, using funds internally generated and 11 from the sale of securities, in the same general manner as it finances 12 other plant additions.

f Reference -

1. General Information, Page 6, as Amended by Amendment No 13

(

14-1 t , -

n

(

l 15 0 COMMON DEFENSE AND SECURITY

2. Applicant is a Michigan (;.orpo12 tion and conducts its 3 busaress solely within the State of Michigan. Its directors and 4 principal officers are all citizens of the United States. The 5 Company is not owned, controlled or dominated by an alien, a foreign 6 corporation or foreign government .

7 The Company has agreed that it will not pelett any individual 8 to have access to restricted data until tne Civil Service Commission 9 has made an investigation and report to the AEC on the character, 10 associations and loyalty of such individual, and the AEC has deter-11 mined that permitting such person tc., have access to restricted data 12 vill not endanger the common defense and security ( .

(' 13 As stated in Section 57 of the Atomic Energy Act of 1954, ,as 14 amended, and 10 CFR, Part 70, Section 70.42, the Cempany may not 15 transfer special nuclear material except as authorized by the AEC.

16- The Company will, as a licensee, be subject to AEC regulations re-17 quiring accountability for and physical protection of fissionable 18 material (3) ,

References

1. General Information, Paragraph (d)

2. General Information, Paragraph (i) 3 10 CFR 70, Paragraphs 70 51 Through 70 56; 10 CFR 73 D

15-1 e

e ,, t - " , ,-

r v ,

e- ,- -e#,-

(,

v.

1

16.0 CONCLUSION

-2 On the basis of the foregoing and application, the applicant 3 respectfully submits that:

4 1. Applicant has described the pmposed design of the 5 Midland Plant facilities, including, but net limited to, 6 the principal architectural and engineerirg criteria for 7 the design, and has identified the major features or 8 components incorporated therein for the protection of the 9 health and safety of the public.

10 2. The application, as amended, identifies the technical and 11 design infomation necessary to complete the final safety 12 analysis. Such information can reasonably be left for later 13 censideration and will be supplied in the final safety analy-(

ll+ sis report.

15 3 Safety features which require further research and develop-16 ment, and the research and development programs to be carried 17 out, are identified in Section 15 of the PSAR. The research 18 and development program is reasonably designed to resolve any 19 safety questions associated with such features at er before 20 the latest dates stated in the application for completion of 21 construction of the facilities.

22 4. Taking into consideration the characteristics of the site 23 and environs and the proposed design of the Midland Plant

{2:

w 3

16-1

l 1 facilities,and the site criteria contained in 10 CFR, 2 Part 100, such facilities can be constructed and op-3 ersted at the proposed location without undue risk to l' 4 the health and safety of the public.

5- 5 Appli: ant is technically qualified to design and con-

'6 struct the proposed facilities.

T 6. . Applicant is financially qualified to design and con-8 struct the proposed facilities.

9 7 The israance of permits for the construction of the 10- proposed Midland Plant facilities will not be inimical 11 to the common defense and security of the United States 12 or to the health and safety of the public.

.(;

~.

.16-2

APPENDIX A COMPARISON OF PWR DESIGN PARAMETERS ( }

~

Midland

• Rancho *** Zion

- Unit 1 Seco **Oconee Unit 1 or 2 Unit, 1 Unit 1 or 2 1 Initial Core Rated Power, MWt 2452 2452 2452 3250

. 2 Nominal Operating Pressure, Psia 2200 2200 2200 2250 3 Nominal Coolant Inlet Tempera-4 ture, F 555 555 555 539 5 Average Coolant Temperature 6 Rise in Core, F 50 50 50 69 6 6 7 TotalCoolantFlowRate,Lb/Hr 131 3x10 131 3x10 1313x10 135x10 8 Average Thermal Output at 9 Rated Power, kW/Ft 54 5.4 54 67 10 Maximum Thermal Output at 11 Design Overpower, kW/Ft 19 2 19 2 19 2 21.2 12 Average Heat Flux at Rated Power, Btu /Hr-Ft2 163,725 163,725 163,725 207,000 13

.14 Minimum DNBR at Rated Power - 2.21(W-3) 2.21(W-3) 2.21(W-3) 1.81(W-3) 15 Minimum DNBR at Design Overpower 1 71(W-3) 1 71(W-3) 1 71(W-3) 130(W-3) 16 Number of Fuel Assemblies '

177 177 177 193 17 Equilibrium Core Average Fuel 18 Burnup, mwd /MIU 27,490 28,200 28,200 33,000 19 Number of Control Rod Assemblies 57 69 69 61

'20 Total Control Rod Worth, % ok/k 8.0 10.0 10.1 7 21 Number of Reactor Coolant Icops 2 2 2 h 22 Number of Reactor Coolant Pumps 4 4 4 4 2 2 2 4 23 Number of Steam Generators 24 Number of High Head Safety 25 Injection Pumps 2 2 2 2 26 Number of Low Head Safety 27 Injection Pumps 2 ,

2 2 2 f

28. Number of Core Flooding Tanks 2 2 2 4 29 Number of Reactor Building Air

30. Coolers 4 4 3 5 31 Number of Reactor. Building 32 Spray Pumps 2 2 ~2 3

• As Proposed for the Rancho Seco Unit 1 FSAR (50-312) l

• • As Extracted From the Oconee Unit 1 FSAR (50-269)

• • • As Extracted From the Zion PSAR (50-295 and -304)

Note: Differences between this Appendix and Midland PSAR, Table 1-2, i

'('. are due to the use of more recent information than was available at the time Table 1-2 was prepared.

A-1

APPEIIDIX B FIGURES

.1 Figre 1 - Architects' Rendering of the Midland Plant 2 Figure 2 - Consumers Power Company System Map 3 Figure 3 - Station Arrangement 4 Figure 4- - Site Plan 5 Figure 5 - Plant Site and Adjacent Area 6 Figure 6 - Population - 10 Miles 7 Figure 7 - Consumers Power Co=pany - Midland Organization Chart 8 Figure 8 - Steam Generator 9 Figure 9 - Fuel Assembly r _10 Figure 10 - Pressurized Water Reactor 11 Figure 11 - Nuclear Steam System 12 Figure 12 - General Arrangement 599' 13 Figure 13 - General Arrangement 614' o

14 Figure 14 - General Arrangement 634' 15 Figure 15 - General Arrangement 659' l

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CONSUMERS POWER COMPANY 34,02 % POPULATION DISTRIBUTION (39,700) 0-50 MILES ,

f l6Q k & 9 6 sHIAWASSEE Co. l ,

CONSUMERS POL ,XMPANY lEl/WD FIANT Organization Plant Superintendent i

Assistant Plant

• Laperintendent

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, Reactor Chemical & Instrunent & , General

• Shift Engineer Radiation Control Protection Engineer Engineers (3) Supervisors (g)

Electrical Mechanical Engineer

Supervisor Supervisor l i

General Technicians Technicians ** Control Engineer (7) (3) -

Operators (114 )

l 4

Electrical Mechanical Technician Clerks R::pa n (1) 2)

A"*Ili"#I Repairug Operators (12)

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• Pre-critical AEC License Total Plant Personnel (78)

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APPENDIX C RESEARCH AND DEVEIfPMENT PROGRAMS OF INTEREST TO ACRS FIRST T,DENTIFIED IN EARLIERJ ASES 1 The nuclear steam supply system for the Midland Plant is similar 2 in concept to several projects already in operation, under construc-3 tion or recently licensed by the Atomic Energy Commission. The pre-4 liminary design is based on technical data which has been developed 5 in the nuclear industry and on data developed by B&W which is specif-6 ically related to the Midland Plant nuclear design.

7 Most of the work reported in this section consists of proof-8 testing of engineered designs, confirmatory tests to confirm ana-9 lytically predicted conditions, or analytical studies to evaluate 10_ design or accident conditions.

11 The following summarizes the status of the research and develop-12 ment items listed in the PSAR:

13 1. Once-Through Steam Generator Tests 14 The ' design of the once-through steam generator is based on 15 experimental work on boiling heat transfer and data obtained 16 by B&W in full-length model tests of the unit. The testing 17 of a prototype unit by B&W has been completed. It included

18. performance, mechanical, vibration and blowdown tests, and 19 control system development. The results have confirmed the 20 analytical predictions of performance, and sufficient data 21 on the performance and structural design has been obtained 22 from operation _of the test models to finalize the design C-1

N.,'

l' of the steam generators and to confirm the analytical model 2 developed for steam generator depressurization. The results 3 of the tests are reported in B&W Topical Reports ,3) ,

4 2. Control Rod Drive Line Test 5 The design of the control rod drive mechanisms is based on 6 a principle which has been used in operating reactors and T which has been extensively tested by B&W. Test programs 8 have included full-scale prototype testing under no-flow 9 conditions, full-scale prototype testing at operating 10 conditions, including flow, and components testing. Proof-11 testing of a prototype mechanism was carried out for a full-12 life cycle of strokes and trips at optimum and 100 percent 13 misalignment conditions, and major design parameters were 14' confirmed. Data from these test programs confir=ed de-15 sign requirements indicating that rod drop time require-16 ments were met, that excessive wear of components did not 17 occur and that corrosion was not significant. These test I 18 data have been reported in B&W Topical reports ',5 ) , ,

19 3 Self-Powered Detector Tests 20 The testing to demonstrate the performance and longevity 21 of the self-powered detectors in the B&W vo;t reactor and 22 in the Big Rock Point Nuclear Power Plant has been completed.

23 The tests have demonstrated that the detectors perform accord-124 .ing to specifications and are capable of measuring neutron

- 25 flux locally in a PWR environment over a period of several C-2

r~,

1 years with a relative accuracy of t5 percent. At the pres-2 ent time, the detectors have accumulated operational experi-3 ence equivalent to approximately six years of full power 4 operation in the Midland Plant reactors. The test results 5 are reported in a B&W Topical Report ( .

6 4. Thermal and Hydraulic Programs T Core thermal performance was evaluated using the W-3 corre-8 lation for predicting departure from nucleate boiling. This 9 correlation is available in the literature and has been used 10 and found acceptable in establishing thermal design limits 11 for other large PWR. With the use of this correlation, 12 only the vessel model flow tests were necessary to substan-13 tiate operation of the plant within acceptable thermal limits.

14' These flow tests have been completed and have demonstrated 15 acceptable flow distribution for the rated power level. They 16 are reported in a B&W Topical Report (T) ,

17 Other thermal and hy'draulic work being done by B&W is directed 18 towards improvement in future core designs.

19 5 Blowdown Forces on Internals 20 The loads on the reactor and internals following a LOCA and 21 the resultant stresses and deflections in the reactor internals 22 have been analyzed for a skirt supported vessel at another 23 site, and the results are reported in two B&W Topical reports '9} .

24 Portions of these analyses will be repeated for the Midland

.25 Plant vessels and site seismic characteristics and reported to 26~ the AEC staff.

C-3

e~s v

'l' 6. . Fuel Rod Clad Failure 21 A B&W program has been conducted to investigate loss-of-3 coolant accident fuel-clad failure mechanisms in order 4 to insure that none will interfere with the ability of 5 ~the emergency core cooling systems to accomplish their 6 objectives. The program involved testing and analytical 7 phases. Parametric tests to investigate possible mechan-8 isms of cladding failure including eutectic formation, 9 cladding embrittlement, and cladding swelling and per-10 foration have been carried out. Results indicate that the 11 emergency core cooling systems will effectively cool the 12 core, even if substantial fuel rod swelling occurs. This

>( 13 work is described in a B&W Topical Report (10) ,

-14 7 Xenon Cscillations 15 The posribility of the occurrence of xenon oscillations 16 throughout the core life is being evaluated analytically 17 by B&W( . A modal analysis and one, two and three dimen-18 sional calculations have been carried out for a core design 19 .similar to the Midland Plant to evaluate axial, azimuthal, 20 and radial oscillations including methods for controlling 21 possible oscillations (1 '13'14) Xenon oscillations are 22 primarily.an operational problem, not a critical safety 23- problem, because the oscillations would be- slow (25-30 hours) 24 and can.be controlled by operator action. Confirmatory-C-4

O..e 1 analyses will be carried out on the Midland Plant core de-2 sign and will be reported to the AEC staff.

3 8. Iodine Removal Spray -

4 The design of the iodine removal spray system is based on 5 information obtained from R&D Programs conducted by B&W and

6. others. These programs have demonstrated the ability of the 7 chemical sprays to effectively remove and retain iodine and .

8 the stability and chemical compatibility of the spray solu-9 tion. The results of these programs are reported in B&W 10 Topical Reports (15,16) ,

11 9 Internals vent valves 12 A test program, including hydrostatic, closing at zero pres-13 sure, pressure differential verification, funettonal handling, 14- vibration and prototype testing, has been completed and the 15 test program demonstrated that the vent valve design will 16 perform adequately during both normal operating and accident 17 conditions. The results of the tests on the full-sized valves 18 and on the prototype valves are reported in B&W Topical Re-19 ports .

References

1. PSAR,' Volume I, Section 1 5; PSAR Amendment No 5, Item 1.1.

2. BAW-10002, "Once-Through Steam Generator Research and Development Report"(PROPRIETARY).

3 BAW-10002, Supplement 1, "Once-Through Steam Generator Research and Development Report" (PROPRIETARY).

s. 4. BAW-10007, " Control Rod Drive System Test Program."

C-5

,Q 5 BAW-10007, Supplement 1, " Control Rod Drive System Test Program."

6. BAW-10001, "In-Core Instrumentation Test Program."

7 BAW-10012, " Reactor Vessel Model Flow Tests" (PROPRIETARY).

8. BAW-10008, Part 1, Rev 1, " Reactor Internals Stress and Deflection Due to Loss-of-Coolant Accident and Maximum Hypothetical Earthquake."

9 BAW-10008, Part 2, Rev 1, " Fuel Assembly Stress and Deflection Analysis Due to Loss-of-Coolant Accident and Seismic Excitation" (PROPRIETARY).

10. BAW-10009, "Effect of Fuel Rod Failure on Emergency Core Cooling" (PROPRIETARY).

11. PSAR, Volume 1, Section 1 5 1 and Section 3 2.2.2 3 and Appendix 3A.

12. BAW-10010, Part 1, " Stability Margin for Xenon Oscillations - Modal Analysis."

13 BAW-10010, Part 2, " Stability Margin for Xenon Oscillations - One-Dimensional Digital Analysis."

11+ . BAW-10010, Part 3, " Stability Margin for Xenon Oscillations -

Two- and Three-Dimensional Digital Analysis."

15 BAW-10017, Rev 1, "Research and Development Report in the Stability and Compatibility of Sodium Thiosulfate Spray Solutions - R&E Report" (PROPRIETARY).

16. BAW-10022, " Effectiveness of Sodium Thiosulfate Sprays for Iodine j l

Removal"(PROPRIETARY).

17 BAW-10005, " Internals Vent Valve Evaluation" (PROPRIETARY).

4 a.

C-6

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