ML20235F159
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I MM A bR 7 t)py April 4, 1968
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SAFETY EVALUATION BY THE-DIVISION OF REACTOR LICENSING U. S. ATOMIC ENERGY COMMISSION _
IN THE MATTER OF_
CONSUMERS PUBLIC POWER DISTRICT COOPER NUCLEAR STATION NEMAHA COUNTY. NEBRASKA DOCKET No. 50-298 8709280436 870921 M ZB - 1 PDR
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TALLE OF CONTENTS i 4
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1.0 INTRODUCTION
1 2;0 DESCRIPTION AND DISCUSSION OF PRINCIPAL PLANT FEATURES 2
- 2.1 Rasctor Desian 3 1
2.2 Emernancy Core Coolinn Systems- 7
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2.3 Containment 9 2.3.1 Primary Containment 9 2.3.2 Secondary. containment 12 .
2.3.3 Containment Structural Design. 13 2.4 Instrumentation 14 2.4.1 . Reactor Prote. tion System Power / Flow Instrumentation 15 2.4.2 Protection System for Actuation of the Engineered Safety Features 17 3.0 IMPORTANT SAFETY CONSIDERATIONS RELATED TO THIS FACILITY 19 !
3.1 Site -
19 3.1.1 Site Description 19 3.1.2 Meteorology 20 3.1.3 Geology and Seismology 21 i
3.1.4 Hydrology 22 3.1.5 Waste Disposal 23 3.1.6 Environmental Monitoring 24 3.2 Emergency Power 24 3.2.1 Off-Site Power 24 3.2.2 On-Site Power 25 t
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-ii-Page 4.0 ACCIDENT ANALY3IS 27:
4.1 General 27 4.2 Loss-of-Coolant Accident Inside the Drywell 28 4.3 Steam'Line Break Accident 29 4.4 Rod-Drop Accident at Hot-Standby 30 4.5 Refueling Accident l
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4.6 Conclusion 31 5.0 REQUIREMENTS FOR FURTHER TECHNICAL __I_NFORMATION 31 5.1 Development Program _of,,Signi_ficance for all Large Water-Cooled Power Reactors 32 5.2 Development Program of Significance for Boiling Wale,r Reactors 34 5.3 Ar_eas Reauirina Further Technical Information 38
'6.0 STATION DESIGN WITH RESPECT TO THE 70 GENERAL DESIGN CRITERIA 42 7.0 REPORTS OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 43 8.0 TECHNICAL QUALIFICATIONS 44 9.0 COMMON DEFENSE AND SECURITY 45
10.0 CONCLUSION
S 46 Appendix A - Report of Advisory Committee on Reactor Safeguards ;
Appendix B - Chronology - Regulatory Review Appendix C - Report of U. S. Weather Bureau Appendix D - Report of U. S. Geological Survey Appendix E - Report of U. S. Coast & Geodetic Survey Appendix F - Report of Nathan M. Newmark Consulting Engineering Services Appendix C - Report of U. S. Fish and Wildlife Service I
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1.0 INTRODUCTION
The Consumers Public Power District (CPPD, applicant), by application dated July 26, 1967, and subsequent amendments, has requested a license to construct and operate a boiling water reactor, designated. Cooper Nuclear Station, on a
. 1090-acre site, 885 acres of which are located in Nemaha County, Nebraska, and the remaining 205 acres in Atcheson County, Missouri, opposite the Nebraska ~
station site.
It is proposed to operate the reactor initially at core power. levels up to the_ design _ level of 2381 MW thermal. However, the nuclear steam supply system is designed for 2500 MW thermal, the power level at which the applicant antici-paces the reactor will. ultimately prove capable of operating. Accordingly .the j applicant, the Atomic Energy Commission's regulatory staff, and the Advisory Com-mittee on Reactor Safeguards have analyzed and evaluated the capacity of the engineered safety features, and the accident consequences assuming a powsr level
.of 2500 MWt. The thermal and hydraulic characteristics of the. reactor core were analyzed and evaluated at 2381 MW thermal. Before operation of Cooper fitation at any power level above 2381 MW thermal will be authorized, a further safety evaluation by the Atomic Energy Commission's regulatory staff will be made to
.- assure that the reactor can be operatead safely at the higher power level.
The technical safety review of the propoaed plant has been based on the applicant's Preliminary Safety Analysis Report (PSAR) and six subsequent amend-i ments, all of which are contained in the application. In the course of our l
l review of the material submitted, we held a number of meetings with the applicant and its contractors, the General Electric Company, and the Burns and Roe Corpo-ration, to discuss the proposed plant and to clarify the technical material sub-mitted. In addition, the Advisory Committee on Reactor Safeguards (ACRS) has u
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also considered this application and has met and discussed it with us and the applicant. The report of the ACRS is attached as Appendix A. A chronology of the n.eetings and principal correspondence is given in Appendix B. Reports by our consultants on meteorology, hydrology and geology, seismicity, seismic design, and marine ecology are attached as Appendices C through G, respectively.
The review and evaluation of the proposed design and construction plans of the applicant at the construction permit stage of the proposed unit is the first stage of a continuing review by the Atomic Energy Commission's regulatory staff of the design, construction, and operating features of Cooper Station. Prior to issuance of an operating license, we will review the final design to determine that all of the Commission's safety requirements have been met. The unit would then be operated only in accordance with the terms of the operating license and the Commission's regulations and under the continued scrutiny of the Commission .
regulatory staff.
The issues to be considered, and on which findings must be made by an atomic safety and licensing board before a construction permit may be issued, are set forth in the Notice of Hearing published in 33 FEDERAL REGISTER.
5174, dated March 29, 1968.
2.0 DESCRIPTION
AND DISCUSSION OF PRINCIPAL PLANT FEATURES In this section, we describe and discuss the principal systems of the Cooper Naclear Station facility relating to plant safety. The Cooper Station nuclear steam supply system is similar in design to that of the Tennessee Valley Authority's Browns Ferry Nuclear Power Plant (Docket Nos. 50-259 and 50-260), and the Philadelphia Electric Company's Peach Bottom Atomic Power
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~3-Station _ Unit Nos. 2 and 3, all of which were previously authorized for construc--
tion.by the Commission. We have compared the Cooper! safety-related system com-ponents to those same components incorporated into t..e design of the Browns Ferry plant.
The~ differences between the Cooper and Prowns Ferry plant designs are attributable to the different site features, to miner changes in the nuclear l
-steam supply system, and to normal differences'in the approach to plant design by different architect-engineers. The differences which have safety significance are the deletion of the baffles in the Suppression chamber; the deletion of the .
jet pump equalizer line; and the design of the airicek doors and the high tempera-ture primary' containment penetrations.
These differences, at well as the fundamental similarities to the Browns Ferry plant, are discussed in the sections immediately following. I 2.1 Reactor Desian ]
The reactor to be employed in the Cooper Station unit is a single cycle. l forced circulation, boiling water reactor producing steam for direr.t use in the steam turbine. The fuel consists of uranium dioxide pellets contained in sealed t
zircaloy fuel rods. The core, consisting of fuel red assemblies, is lodated I within a domed cylindrical shroud, separated from the reactor vessel by an annulus The annulus contains jet pumps through which feedwater is passed to the core, and inlet and outlet connections for the two recirculation loops. The jet pumps are driven by the recirculation flow from variable speed centrifugal pumps located in the two recirculation loops.
. Water, which serves as both the moderator and coolant, enters the bottom j of the reactor core snd flows upward through the fuel assemblies where boiling L
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dryers located within the~ reactor primary vessel abote the domed shroud. The
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steam passes through the main steam lines to the turbine. The sep.arated water
. mixes with the incoming feedwater and is returned to.the core via the jet pumps.
Control of the reactor is accomplished 'by two -means:
(1) ' Bottom-entry cruciform control rods are moved vertically within the reactor core by individual control drives. The drives are the same hydrau-4 lically operated, locking piston type employed in other General Electric boiling water reactors since Dresden 1,. including the Browns Ferry reactors.
(2) Variable frequency motor generator sets power the recirculation pump-motors for pump speed control. Changing the pump speed changes the recircu-lation flow rate, which in turn changes the' reactor power level. Up to 20 percent. variation in power can be achieved in this manner.
Except for plant total thermal output, 2381 MWt and 3293 MWt for Cooper and Browns Ferry, respectively, the reactors have almost identical core thermal designs. The small dif ferences in the core power density and the average and maximum heat flux are attributed to the differences in the core diameters and resultant limitations in filling out the core diameters with standard fuel l
elements. These differences are not considered to be of safety significance.
l A tabulation of the principal thermal design parameters of these two plants is given in Table 2.1.
In General Electric plants previously reviewed, the two jet pump headers were connected by an equalizer line, the purpose of which is to improve plant flexibility and availability if one reactor coolant recirculation system pump were inoperative. The applicant elected to delete the equalizer line in this
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-; -COMPARISON OF THERMAL DESIGN, COOPER STATION -- BROWNS FERRY Parameter (at rated power) Cooper Station ' Brown's Ferry Thermal Output, Rated (MWe) 2381' 3293 Total. Reactor Core Flow Rate (1bs/hr) 74.5 x 10 6-102.5 x 10 6
' Core Power Density (kw/1)- 51.2 50.8 2
Average Heat Flux (BTU /hr-ft ) 164,500 163,200 2
Maximum Heat Flux (BTU /hr-ft ). 427,820 425,000 Average Linear. Power. (kw/ft) 7.1 7.1 Maximum Linear Power (kw/ft) ~18.5 18.4 Critical-Heat Flux' Ratio 2 1.9 1 1.9 Average Fuel Temperature (*F) 1100 1100 Maximum Fuel Temperature (*F) 4380 4380 Core Average Exit Quality (%) 13.4 13.6 Assumed Fuel Damage Limits (1) Minimum Critical Heat 1.0 1.0 Flux Ratio (2) Linear Power (kw/f t)' 28 28 I
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plant. Operation without the equalizer with only one of the recirculation pumps running will limit, plant operation to 65% of rated power-(1548 MWt). The appli-cant has accepted this at an operational limitation.
4 Deletion of the equalizer line relates to plant safety in the following 1 manner. By eliminating the equalizer line, the effective area of' a recircula-- 'i tion line rupture is reduced since blowdown through the equalizer line is now -
impossible. Further, the logic network required to identify the unbroken reactor coolant recirculation system line for safety injection can be simpli-fied. As a direct result of the reduction in effective break' area, and conse-l quent slower blowdown time, the flow area of the drywell/wetwell vents has been reduced. This reduction represents no change in basic containment design criteria in that the same ratio of break area / vent area that has been used in previous BWR designs has been maintained. The ratio of break area / vent area is a parameter used to establish peak drywell pressure in a pressure suppression containment and has been established from tests conducted by General Electric at the Pacific Gas and Electric Company's Moss Landing facility.
In our judgment, the applicant has considered all possible safety related '
considerations of deletion of the equalizer line and thus we conclude that deletion of the equalizer line is acceptable in Cooper Station's design.
The fuel damage limit criteria for normal operation of Cooper Station are the same as those specified for Browns Ferry; 1.e., a minimum critical heat flux ratio (MCHFR) of 1.0 and a linear neat generation rate equal to 28 Kw/f t. In all anticipated operational transients evaluated, the controlling f ael damage limit is the critical heat flux criterion rather than the linear heat generation rate criterion. For the rated power, the critical heat flux is 1.9 times the
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. peak heat flux (MCHFR = 1.,9). The applicant calculates that the MCHFR drops from 1.9 to 1.3 for the most severe operational transient (tripping both recircula- ,
tion pumps without scram). We agree with General Electric that the difference between MCHFR = 1.3 and MCHFR = 1.0 is an adequate allowance for uncertainties in calculating MCHFR from quantities measured during operation of the reactor.
We find these thermal design criteria and parameters acceptable for a construction permit. However, for further assurance that fuel rods will not be overheated during normal operation at rated power and during anticipated tran-sients, additional analytical and experimental verification will be required prior to the operating license stage. Confirmation of core thermal performance will be required of all boiling water reactors of the Browns Ferry class, which includes Cooper Station. The General Electric Company is presently conducting experimental programs to obtain this confirmation. The results of these programs 4 i
are expected to become aval'.able prior to the date proposed for initial operation l of Cooper Nuclear Station. Details of the experimental programs are discussed in Section 5.0 of this Safety Evaluation.
t 2.2 Emergency Core Cooling Systems At least two different core cooling systems are provided to remove stored and decay beat, and to prevent fuel clad melting over the entire spectrum of postulated primary system loss-of-coolant accidents up to, and including, the rupture of a recirculation line. Such cooling capability is available even with the loss of normal off-site a.c. power.
The cor.ceptual design and design baser; for the Cooper Station emergency core cooling system (ECCS) are the same as those for Browns Ferry, and other
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.n recent' General Electric reactors. The ECCS' is an integrated system composed of the core spray ' system, high pressure coolant injection system (HPCI),. low pressure; coolant injection system (LPCI), and auto-relief (automatic pressure
,, relief); system.
In The purpose of the HPCI system is to provide coolant addition to the reactor
-vessel for primary coolant system break sizes smaller than those for which the i <
, core spray cooling system and the low pressure coolant injection system are -
designed. The HPCI system uses a' steam turbine-driven pump and is designed to achieve.its function without reliance on electrical power other than the station
, . battery system. The steam-driven pumps pump water from the pressure suppression n (
pool. For intermediate size breaks, the HPCI will depressurize the pressure q vessel'to a low enough' level so that the core spray system or the low preseure Wi- ?
j coolant injectiot, system can function. A description of this function of the ny 4a
- HPCI system is presented in Section 5.2.
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The primary'stystem relief valves will function as an automatic depressurizer,
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under#certain,cond'itions,-following loss-of-coolant accidents. During normal
.r operation,,these valves open automatically on a reactor overpressure of approxi-V matelyj.00 psi and then close at a lower, pre-set, pressure level. An additional fuhetion of these valves will be to open and remain open below this pre-set closing pressure, when signalled to do so, folleving a loss-of-coolant accident. Tria " remain open" signal will be triggered by simultaneous signals from high drywell pressure, loss of reactor water level, and non-operation of either the. reactor feedwater system or HPCI system. By remaining open, the relief valves will reduce the reactor primary system pressure down to the point where the LPCI system and/or the reactor core spray cooling system can accomplish i
reflooding of the reactor core.
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i The reactor core spray cooling system uses water'from the pressure suppres-l
. sion chamber pool to cool the core,- witti the water being distributed directly to 1-the reactor. core by two spray headers mounted inside the plenum above the-reactor core.
' The low-pressure coolant injection (LPCI) systen is provided as an l
o additional independent,' redundant means of , removing stored and decay heat from L
- the reactor core following a loss-of-coolant accident. Either the core spray L
'or LPCI. systems will serve to restore the water level in the reactor primary vessel to at least 2/3 the core height and to maintain this water level by making up boilof f, or. any leakage from the core shroud.
Additional discussions of the. emergency core cooling systems are presented i
in Sections 5.2 and 5.3.
3 Containment 2.3.1 Primary Containment-The Cooper Station primary containment design incorporates the same pressure suppression concept and spray cooling provisions that have become standard for General Electric boiling water reactors, including Browns Ferry. The primary containment is designed to accommodate the peak transient pressures and tempers- 1
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I tures resulting from a failure of the reactor primary coolant system equivalent '
in size to the circumferential rupture of one of the main recirculation pipes. ,
In conjunction with the reactor building, the primary containment has the added function,' in the event of a loss-of-coolant accident, of limiting potential fission product 'of f-site releases to values below 10 CFR Part 100 guidelines.
The primary containment design consists of a dryell and a pressure suppression chamber (torus). The reactor vessel, the reactor coolant j
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, . recirculating loops, and other branch connections of, the reactor primary system are located;in the drywell. The pressure suppression system consists of a
' chamber.which contains 87,660 cu. ft. of water. It also includes a connecting vent system b'etween the drywe11~and water pool.
The drywell is a steel pressure vessel with'a spherical lower portion, and
?' na cylindrical upper portion. The pressure suppression chamber 'is- a steel q pressure-vessel.in the shape'of a torus located below and encircling the drywell.
U The design pressure of eachivessel is 56 psig. A vent system connects the dry-well.to the suppression chamber and. terminates below the water-level in the pressure suppression chamber, so that in the event of a reactor system pipe failurs in the drywell, the released steam passes directly to the suppression pool water where.it is condensed. This transfer of energy'to the water pool rapidly. reduces'the pressure in the drywell and, accordingly, limits the amount of subsequent leakage from the primary containment. Provisions are made for the removal:of' heat from within the primary containment to maintain integrity of the -
containment system indefinitely following any accident un t.o a design basis loss-of-coolant accident.
The design pressure for the drywell and suppression chamber for Cooper Nuclear Station was established on the basis of the Moss Landing pressure Ll j~ suppression chamber tests; i.e., 56 psig for the drywell and 35 psig for the l -
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suppression chamber.- Baffles were included in the original suppression chamber to prevent short term overpressure of some 6 psig as observed on the 1/4-scale '!oss Landing tests. However, to simplify pneumatic testing of the I
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primary containment, the suppression chamber design p: essure has~been' increased 3
from 35.ps'ig.to;56'psig (based on code allowances for a maximum internal f i
pressure of 62 psig). This allowed the exclusion of de baffles from thel ^
' Cooper containment since even if the 6 psig observed overpressure were to o'ccur, j
. the. design pressure of the suppression chamber would :st be exceeded. Further.
there is more recent experimental evidence from additional full scale. testing that the overpressure would not occur in a full-scale geometry. A description
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of the testing and the corresponding analysis is give: in Amendment 1 of the !
PSAR. Another reason for having baffles was to prevent excessive turbulence in ;
the water'in the torus during blowdown, resbiting in possible uncovering of the downcomers. The tests at Hoss Landing conducted by General Electric have shown that the baf fles were not required to prevent azimuchs1 sloshing or other fluid l ,
disturbances. The uniform placement of the vents in the suppression chamber plus the venting action, which would be' a low frequency loading, tends to pro-duce only small amplitude waves. In view of the foregoing, we conclude that removal of the baffles in the suppression chamber does not diminish the margins of safety associated with the full baffled design of previous BWR's.
The suppression chamber pool temperature of 130*T quoted in the original PSAR after four hours of Reactor Core Isolation Cooling (RCIC) system operation was based on an RCIC mode of operation similar to that of the Dresden 2 plant.
- - The modified Cooper design (included in Amendment 1) using the RHR heat exchanger as a condenser results in a suppression cha
- Ser pool temperature after four hours of RCIC operation of 110'F. This results in a 60"F rather than 40'F f margin in suppression pool temperature based on a maximum allowable temperature L
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of 170'F following a design-basis loss-of-coolant accident blowdown. To take advantage of this increased temperature differential, the minimum volume of the 3 1 suppression chamber water was reduced to 87,660 fe . This was based on utiliz- 1 ing'only 50*F of the available 60*F temperature differente. We believe this
. reductica in water- volume is acceptable since it represents a refinement in quantitative analysis and does not reduce any safety margins in overall cooling requirements following loss-of-coolant accidents.
In the Browns Ferry plant, double bellows are used for the pipe penetra-tions; whereas for the Cooper Station, the designer has selected a single !
bellows pipe. penetration design. However, in t.he single bellows penetration, a '
seal arrangement is provided to permit periodic leakage testing of the bellows.
The design of the' single bellows penetration takes into account the simultaneous
- stresses associated with normal thermal expansion, live and dead loads, seismic loads, and loads associated with a loss-of-coolant accident. This same single bellows design is being incorporated in the Peach Botto. Units. We see no "
evident advantage of double bellows over' single bellows penetration and find the single bellows penetration preliminary design acceptable.
The Cooper unit will have single seals on each dryvell airlock door com-pared to the double seals on the Browns Ferry units. We believe that because the integrity of the airlock can be demonstrated first by pressurizing between doors and then by pressurizing the containment to test both doors, the design is acceptable. This same design is being incorporated in the Peach Bottom units.
1 2.3.2 Secondary Containment {
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As in the Browns Ferry and other General Electric plants, the reactor building for the Cooper Station unit, together with the standby gas treatment i f
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system and a stack, provide the secondary. containment barrier when the primary containment is in service. The ' reactor. building also serves as the primary con-tainment ' structure during periods when' the drywell . (primary containment) is open for servicing, such as during refueling. . The reactor. building will consist of
. poured-in-place, reinforced concrete exterior walls up to the refueling floor.
- f. Above the level of the refueling floor, the building structure will be steel frame with insulated metal siding. To provide leak-tight integrity, the siding is designed with' caulked interlocking vertical joints and overlapping horizontal joints.
2.3.3 Containment Structural Desian-The primary and-secondary containment structures are classified as Class I structures. The applicant has proposed that the seismic design for the buildings be based on dynamic anayses using acceleration response spectrum curves corres-ponding to a 0.10g " operating basis earthquake."* With this earthquake, stresses in the structural members will be below the allowable working stress values. The design is such that a safe plant shutdown can be made for a 0.20g
" design basis earthquake,"** even though the stresses in some members may approach the yield point. These seismic design bases have been found acceptable by our seismic design consultant, Nathan M. Newmark Consulting Engineering Services (Appendix F). Further, this consultant accepts the damping values proposed by the applicant for the design of the containment under maximum and design earthquake loads. We agree that these bases are acceptable.
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- The " Operating Basis Earthquake" for a reactor site causes the vibratory ,
ground motion for which all features of the facility necessary for continued i operation are designed to remain functional. The maximum ground accelera-tion of the Operating Basis Earthquake is equal to at least one-half that of the Design Basis Earthquake which for Cooper Station is 0.10g.
- The " Design Basis Earthquake" for a reactor site causes the vibratory ground motion for which all features of the facility necessary to protect the health and safety of the public are designed to remain functional. The ,
Design Basis Earthquake is the largest earthquaka postulated for the site which for Cooper Station is 0.20s. j ay
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=i The reactor building or secondary containment, is to be designed for an internal negative pressure of 0.25-inch of water with respect to the outside atmosphere under neutral wind conditions with the emergency ventilation system functioning.. Under.these circumstances, any leakage.would be,into the building L and releases would be from the plant stack. . 'The building is also designed to be
- able to withstand an internal pressure of seven. inches of water without struc-tural failure,'and without pressure relief. The structural steel frame of the reactor buil'ing d will be designed to withstand the tornadic force resulting from a tangential' wind velocity of 300 mph, a' pressure drop of 3 psi in three : seconds, j and a transverse velocity of 60 mph (Amendment 5). - However, the reactor build-ing metal siding and roof . decking will~ be designed for normal wind loading.
When this design velocity is appreciably exceeded, the siding and decking may be blown off, exposing the refueling floor and parts of the reactor building not required for safe shutdown. _The structural design will assure that under the 300 mph wihd loading, the structure will not collapse, nor permit the crane to fall, nor otherwise endanger critical safety systems required for safe shutdown-of the plant. Even if the primary containment should become exposed, it will be able to withstand the 300 mph wind velocity.
Considering the foregoing, as well as the advice of our seismic design i
consultant, Nathan M. Newmark Consulting Engineering Services, we believe that this proposed approach to the structural design of the containment structures relative to seismic and tornado effects is acceptable.
2.4 Instrumentation ThereactorprotectionsystemIscram)andthedesignprotectionsystemwhich J actuates the engineered safety features have been' reviewed previously in the i
Quad Cities, Vermont Yankee, and most recently in the Peach Bottom application
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l and found to be acceptable. The areas which differ in design from those previously reviewed are:
- a. Reactor protection system power / flow instrumentation.
- b. Protection system for actuation of the engineered safety features.
The intent of our review.was to' ensure th'at the above mentioned systems meet the proposed Institute of Electrical and Electronics Engineers (IEEE) .
Standard for Nuclear Power Plant Protection Systems.
2.4.1: Reactor Protection System Power / Flow Instrumentation The power range monitoring system consists of 124 Local Power Range Monitor (LPRM), 6 Average Power Range Monitors (APRM), 2 Rod Block Monitors (RBM), and 2 Recirculation Flow Converters.
The LPRM channels provide power distribution monitoring (indication and alarm) and provide signals to the APRM channels. The LPRM's also provide signal to the RBM channels by means of an automatic switching matrix. The switching matrix selects signals from the LPRM channels nearest the control rod selected to be withdrawn.
Each APRM channel will average the signals from eight to twelve LPRM channels for the purposes of bulk power monitoring and automatic reactor core protection. Specifically, the APRM protects the core through a fixed high flux scram trip. Three APRM channels will be connected to each of the two reactor protection system logic channels. A trip in any one of three APRM channels in each of the logic channels will cause a reactor scram trip. The APRM also pre-vents core operation at excessive bulk power levels with reduced recirculation flow. This function is provided by rod block actuating trips which vary auto-matica11y with recirculation flow. A trip of any one of the six APRM channels l
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results iri a control rod block. -In summary, with reduced recirculation flow the APRM system can now perform only a rod block function at pre-determined power
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l 1evels. , Reactor scram is actuated by a fixed high flux scram setting (120%
power).'.We'have investigate'd the adequacy of the proposed system to prevent fuel damage at steady ~ state-reduced recirculation flow and believe the proposed systems
. are adequate. However, we have asked the applicant to continue reviewing the adequacy of the systems for anticipated transients at reduced recirculation flow.
Rasul'ts of this evaluation will determine the required APRM functions. This area can, in our opinion, be resolved during the operating license review.
The RBM is designed to generate trip signals which are used to automatically inhibit control rod' withdrawal when local power distribution may leak to fuel damage. The RBM system will, through the automatic switching astrix, average the output si;nals t from the LPRM channels nearest to the control rod selected for withdrawal. Rod withdrawal is delayed for about one second to permit the RBM average signal to be calibrated to read the same as the pre-selected reference APRM channels. Rod withdrawal will be inhibited if the resultant signal exceeds trip' set points which are automatically varied with recirculation flow signals provided by the recirculation ~ flow converters.
The recirculation flow converters provide reference signals which are used to bias the APRM and RBM systems as described above. Also, each flow converter contains'a comparator circuit which accepts a signal from..its own converter and the other converter. The resultant difference signal from each comparator will actuate a trip circuit which prohibits control rod withdrawal. The trip logic I
is one-out-of-two.
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-17 The applicant has stated that the design v1' 11 comply with the proposed IEEE
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Standards for Nuclear. Power Plant Protection S' y stems. Our review of the pre-'
liminary design indicates that the proposed system will meet the proposed standard ,
l 1The sdaquacy of the system ' depends upon the manner in which ' channel independence ' j I
and isolation are impismented in the final design. We will review this aspect il I
of the final design in detail at the operating license application evaluation. j i
2.4.2 Protection Systes for' Actuation of the Ennineered Safety Features Recent design changes in Low Pressure Coolant injection (LPCI) and Core; Spray systems have resulted in modifications to the control instrumentation for' i these systems. The dual logic protection system which initiates these systems, however, remains unchanged.
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The equaliser line interconnecting the recirculation loops has been elimi- 1 nated by a recent design change. As a result of this change, the instrumentation and circuit logic used to detect the undamaged flow path for coolant injection has been modified. The modified system is described below.
The LPCI sensing circuit is started upon receipt of low-water level or pri-mary containment high pressure signals. A ten-second delay is provided to per-mit sufficient time for the reactor coolant recirculation flow to decay and the flow direction sensors to determine flow direction. The flow direction sensors are differential pressure gages connected between each of the five jet pump riser pipes and the reactor vessel. The sensors will be arranged in an equivalent
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one-out-of-two-times-two logic scheme similar to the scram system. These sensors will indicate a positive or' negative signal. A positive signal indicates that flow is into the reactor vessel (undamaged loop) and a negative signal w
J indicates that flow is away from the reactor vessel (damaged loop).
j Upon identification of the undamaged flow path and after the time delay, the ,
valves in this loop and the LPCI injection valve receive the signal'to close and open respectively.
The applicant has stated that this system will-be designed to meet a single failure criterion. We'believe that the applicant can and will design this por-tion of the system to meet.this' criterion.
The core spray system is actuated simultaneously with the LPCI system and
' the same diversity of actuating signals as described above fo'r-LPCI actuation
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apply to this system. The Cooper design represents a change from recent BWR 4
designs in that both loops of the core spray system inject coolant concurrently Recent BWR. designs permitted one loop upon receipt of the actuating signals.
to'be actuated initially and if it did not provide full' flow, the second loop would be actuated.-
Because of the inherent redundancy and diversity of the actuation signals, we believe that the instrumentation can be designed such that no single failure will disable this system.
All instrument channels required for reactor protection and the actuation Also, of the engineered safety features are testable during reactor operation.
there is complete separation of control and safety functions.
The applicant has stated that a means will be developed to easily distinguish wires, and components from similar components. which protection system cables, are not related to protection. We believe that the applicant's stated intentions L* are adequate for the construction permit review.
The applicant will provide test data to confirm that th'ose electrical coms ponents, instruments, and cables located in the primary containment will be f'
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i capable of performing their function during and subsequent to any accident for the length of time required. This is discussed in more detail in Section 5.2 )
'of this report. We believe this commitment is adequate for the construction permit review. The data will be evaluated during the operating license review. !
The applicant has stated that the reactor protection system and the'instru-mentation which actuates the engineered safety features are being designed to -
the proposed IEEE Standard. Our analysis of the preliminary design indicates that these systems can be built to satisfy the proposed IEEE Standard.
3.0 IMPORTANT SAFETY CONSIDERATIONS RELATED 'IV THIS FACILITY In our evaluation of this application, we have given special considera-tion to a number of site and design features which have safety implications related primarily to the Cooper unit. These safety considerations are discussed in the following sections.
3.1 Site 3.1.1 Site Description The site is in Nemaha County, Nebraska, and Acheson County, Missouri. The reactor will be located on the west bank of the Missouri River, between the villages of Brownv111e and Nemaha. The site consists of 1,093 acres of land owned by the applicant and has an exclusion area radius of about 2200 feet. The i nearest residence is approximately 3500 feet from the reactor. Within one mile of the proposed reactor, there are four permanent residents; within two miles, there are 51 permanent residents; within ten miles, there were 3,499 residents in year 1960; within twenty miles, there were 27,028 residents in
' year 1960. The nearest population centers with more than 25,000 residents k
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(Lincoln, Nebraska, and' St. Joseph,. Missouri) are at least sixty miles from f
the site.: The villages of Brownville (pop. 243) and Nemaha (pop. 250), Nebraska,*
are.2-1/4 miles northwest'and 3 miles southwest,~respectively, of the site.
The' applicant predicts that the population within a.40-mile' radius,.an area that is predominantly used for farming, will. decrease between 1960 and 1980. Con-
.sidering the proximity of Brownv111e'and Nemaha to the Cooper site, the low-population zone radius as ' defined in 10 CFR Part 100 is 3 miles. Because of the remoteness of the site from any population center'containing more than 25,000 residents, . it is unnecessary to determine a population center distance.
In'the Accident Analysis section of this report, we have shown that radiation doses. co the population would be within.10 CFR-Part 100 guidelines at the site boundary and the low population zone if an accident.were to occur.
3.1.2 Meteoroloav The atmospheric diffusion rates espected in mid-continental United States, are applicable to the site area. The applicant has presented a range of parame-ters for estimating the amount of diffusion to'be expected.under various atmos-pheric conditions et the site. We believe that the diffusion parameters proposed by the applicant are adequately conservative except for the assumption that the radioactive plume resulting from the steam line break rises above ground level with Le curbulence or mixing. In our analysis of the steam line break' accident, we assumed that the radioactive plume did not rise above ground level. At the request of the staff, the Environmental Meteorology Branch (Weather Bureau), ESSA, has also reviewed the site meteorology. Its comments are included in Appendix C. ,
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. The applicant will erect.a meteorological tower at the proposed site with
." instrumentation to measure wind speed, wind direction, and temperature at
-various heights.- We~ believe that data to be 'taken in the' new meteorological monitoring program will'show that the atmospheric. diffusion' assumptions used -
by us to estimate accident' doses are conservative. The data taken should also provide an acceptable basis-for establishing routine gaseous release limits ati
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the operating, license stage. Cooper Station plant design also will meet. tornado design criteria. This has been discussed in Gection 2.3.3.
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3.1.3 Geoloav and Seismoloav Test borings at 'the proposed site show that! the site is characterized by:
alluvialandglacialdepositsoversedimentaryrocks'andthatthedistancefrom-l the surface to rock is approximately 70. feet. This has been confirmed by the U. S. Geological Survey. The report of the Geological Survey is included in Appendix D.
The potential for' liquefaction of soil sub-base that resulted from the in-situ soil conditions will be eliminated by corrective design provisions and foundation construction procedures (A:nendment 2). In summary, these include:
- 1. Extensive de-watering of the area.
- 2. Excavation to bedrock.
- 3. Sheet piping and cellular cofferdams to minimize water seepage.
- 4. Structural fill, supported by the bedrock, of granular material com-pacted to varying relativa densities ranging from 75% at bedrock to 85% at the final ground surface. The fill will be placed beneath and i
around the major structures on the site. l
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- 5. An extensive program of measuring soil properties prior to and during placement of the fill to verify design' values and to ensure'up-seepage.
within the fill does not loosen the material.
- 6. A reinforced concrete mat foundation, supported by the fill, for d
major structures.
A rigorous quality control program has been specified to assure that the necessary provisions and procedures are followed 'during construction. ,
1 The applicant has used a horisontal acceleration of 0.10g~ at the rock sur-face and at the base of the structures as the design criterion for the elastic response of major structures, and a horizontal acceleration of 0.20g both at the rock surface and at the base of the structures as the design criterion for safe and orderly shutdown of the reactor and to retain functional integrity of major structures. Further discussion of the structural design of the plant relative to seismic effects is in Section 2.3.3. Our seismic consultant, the U. S. Coast and Geodetic Survey (USC&GS), and also the U. S. Geological Survey (USGS) agree that these acceleration values are adequate for the site. The reports of the USGS and the USC&GS are included in Appendix D and E, respectively.
3.1.4 Hydroloav The reactor will be located on the river side of the Missouri River levee, more than 275 miles downstream from the nearest flood control dam, Gavins Point Dam. There are no dans downstream of the site. The maximum river level recorded i
prior to the installation of river controls is 899 feet MSL. The maximum river level recorded subsequent to river controls is 895.4 feet MSL. These data are based on studies by the U. S. Army Corps of Engineers. The ' elevation of the i
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- leves is 902 feet MSL and the elevation of the plant grade sd11 be 903.'5' feet MSL. , If flood water were to rise to the' elevation of the levee, further rise would'be . retarded by the flow of water into a six-mile wide valley at the site.
If a major flood control dam (Oahe or Fort Randall) were to fail at' the time of
< the maximum probable flood', the river elevation would not exceed .906 feet MSL, based on data supplied by the Corps of Engin'eers (Amendment 2). The waters
( = released by the dam would arrive at the site at least three days after dam failure, which is ample time for the applicant to provida an additional two and one-half feet of flood protection above plant grade.
Lowest river flow' occurs during the winter. The requirements for plant cooling are about one-third of the. lowest river flow. Service water capability !
will be designed for safe shutdown under the most critical conditions of ice'
- and low water; that is, the elevation of the intake of the service water pumps stil be such as to result in three feet of free water for these conditions.
3.1.5 Waste Discosal ..
The stated ~ design objective of the Cooper rad-vaste systems is that the release of radioactive effluents will not exceed the limits' of 10 CFR 20. Liquidi wastes will be collecced, treated, stored, and re-used or disposed in batches at a rate dependent upon the nature and concentration of radioactive material. .
Gaseous radioactive wastes released by the condenser ejector will be directed i
through a 30-minute holdup line, will pass through absolute filters, and then l i
be released via the stack. Liquid and gaseous wastes will be sampled and moni- i tored to maintain control over the release of radioactive material, and to assure that off-site concentrations are within the requirements of Part 20. ;
Our review of the site characteristics'does not reveal any specific problems with respect to dilution and dispersion of liquid and gaseous effluents,
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. and we do not foresee any difficulties in controlling disposal of these wastes-.
1 during routine operatior. to meet the requirements of the Commission's regula- J tions, 10 CFR Part 20. j 3.1.6. Environmental Monitorina l:
-The applicant will begin an environmental radiation monitoring program j about two years'before the scheduled startup of the reactor. The program will consist of measurements of (1) background radiation levels, and (2) radioactive
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materials in samples of air, river water,. river bottom sediment, ground water, . ;
soil, vegetation, milk, and fish. In our opinion, the scope.of the environ- i i
mental monitoring program is adequate to enable an evaluation.of the effects of reactor operation on the environment. . In addition,'we hiva requested the comments of the Fish and Wildlife Service on the impact of this plant on the environs. The applicant's program of environmental monitoring is consistent-with the Service's reconsnendations and will provide an adequate basis for evaluating the data from post-operational monitoring. A copy of the Fish and Wildlife report.is' included in Appendix G. .
3.2- Emeraency Power 3.2.1 Off-Site Power The Cooper Station generating capacity represents an appreciable fraction of the total Consumer.s Public Power District generating capacity. The Cooper 800 MW unit,1 e -- , will be integrated into the interconnected power systems
[ .of the Midwest. The plant will be connected to four 345 KV transmission lines that provide direct interconnection with (1) the Mid-Continent Areas Power
_ Planners (MAPP), (2)'the Iowa Pool, (3) the Mo-Kan Pool, (4) the Nebraska 1
Public Power System, and (5) the 0maha Public Power Districte These systems
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The applicant has. made extensive studies of. system stability, load flow, . 1 o .. .
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' capacity' reserve requirements, . and switching surges. Thestudiesindicatethat,]
i in the event'of a' trip'of the unit, the system may be perturbed, but'off-site- ;
power to the station will not:be lost.
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1 In addition to the startup transformer and its transmission line feeder l initially. proposed, in Amendment No. 3 the applicant agreed to provide an' l
additional startup transformer with a separate transmission'line feeder, so that two separate sources of off-site startup and emergency power are avail-able. Additional studies are being made as to-the reliability of the proposed-sources of power to these two startup transformers.
We believe that the off-site electric system for Cooper Nuclear Station ;
J is adequate since two redundant' sources of off-site power are being furnished and that no single failure'should prevent. power from being supplied to the minimum engineered safety features from off-site sources.
3.2.2 On-Site Power The engineered safety feature loads and the loads required for safe shutdown are connected to two 4160 volt critical buses. Redundant loads are connected.
to the critical buses such that each bus can independently supply power to meet minimum engineered safety feature and safe shutdown requirements. In the event of an accident, all the required equipment on both buses will be utilized, but minimum requirements would be met with only one operating bus. Each bus will be isolated ' from the other bus and operate independently of it. The two diesel-generators on each bus will, be sized to start and supply the necessary loads
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- i that are required.for the accident condition. In the ACRS letter, the recom-mandation was made that further consideration of this system be made.to avoid !
the. need for diesel generator synchronization. - We will pursue this' as part of four continuing review of Cooper Station to assure that the Committee's recom- =
nandation is set. Each diesel-driven generator ,will be provided with its own-L. . .
d.sy-tank containing sufficient fuel for eight hours of operation at maximum load.
TVo common fuel oil storage tanks are also provided and have sufficient capacity for seven ' days operation. : Equipment will be arranged 'to provide the maximum practical separation between switchgear sections, power centers, . control centers,-
and similar equipment groups, as well'as circuits. Means will be provided for i
rapid location and isolation of system faults. Two separate duct banks will be installed between the diesel-generators and the 4160 volt critical buses to maintain separation between the cables for each critical bus.
Two seperate station ' battery systems will be furnished' each consisting of .two batteries, one operating at 250 volts for large motor loads and one operating at.125 volts for control and semil motor loads, with battery chargers.
A loss ov voltage on one battery will cause an automatic transfer of load to the corresponding battery on the alternate system. The batteries are located in two ventilated battery rooms, and a ground indicating and annunciator system will be provided.
We believe that the on-site electric system for Cooper Nuclear Station is adequate.since no single failure should prevent power from being supplied to ;
the minimum engineered safety features from on-site sources.
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- 4.0 ACCIDENT ANALYSIS ,
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- 4.1' General
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.The design-basis accidents analysed in the Cooper PSAR are the same.as those q investigated in the Browns'Farry, Vermont Y&nkee, and Peach' Bottom Units No. 2 4
and 3 applications. . The only major difference in the radiological consequences l of these accidents between plants is attributed to the differences in site features and meteorology. The doses calculated by the staff are. listed in the following table and are within 10 CFR Part' 100 guidelines. Our assumptions concerning each accident are discussed in.the following sections. As usual forJ reactors designed by General Electric, the doses calculated ~ by the applicant are significantly less than those listed in the following table.
TABLE 4.1 MAXIMUM OFF-SITE DOSES RESULTING FROM DESIGN BASIS ACCIDENTS 2-Hour Dose-(Rems) 30-Day Dose (Rems) 2,162 Foot Exclusion At Low-Population Radius Radius of 3 Miles Whole Body Thyroid Whole Body Thyroid Refueling Accident
- 2 25 (1 15 Main Steam Line Break (1 20 <1 2 Accident Loss-of-Coolant Accident
- 2 60 4 190 Rod-Drop Accident (1 10 (1 2.2
- Two-hour dose calculated at point of maximum dose -- 6400 feet, because l of elevated release.
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Stack effluents will be released at 1,215 feet MSL; the plant grade will be at 903.5 feet MSL; and the highest ground'at the exclusion area site boundary is approximately 890 feet MSL.,At 6400 feet west of the reactor, the highest g
' round elevation is 1100- feet MSL; the effective stack height for. this ground is' assumed to be 115 feet. In calculating the maximum two-hour dose beyond the exclusion radius that would result from the refueling and loss-of-coolant accidents, we assumed a downwind distance of 6400 feet, an effective stack
. height of 115 feet, . Pasquill Type-F.dif fusion conditions, and one-meter-per-second wind speed. In calculating the maximum two-hour dose beyond. the exclusion radius that would result for the main steam line break accident and rod-drop accidents, we assumed a ground level release, a 11,000 ft2 ,gf,cggy, building wake area, a downwind distance of 2162 feet Pasquill Type-F diffusion I conditions, and one-meter-per-second wind speed. Other factors used in our j i
calculations for the particular accidents are discussed in the following' sections.
4.2 Loss-of-Coolant Accident Inside the Drywell The loss-of-coolant accident is assumed to be a double-ended severance of one of the coolant recirculation lines. In calculating the doses resulting from this accident, we assumed TID-14844 release fractions, a drywell leakage rate of )
0.635 percent. per day, drywell leakage immediately into the standby gas treatment system, 90 perceac erriciency of the iodine filter, and release via the plant stack at 1,215 feet MSL.
We conservatively assumed that the drywell leakage immediately enters the E
! standby gas treatment system because insufficient dats are available to .)
i-determine the degree of mixing in the secondary containment. If uniform mixing I
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and 100% per day leakage of the secondary containment were factored into the analysis, the two-hour doses beyond the exclusion radius would be reduced by a factor of about 10 to 20.
4.3 Steam Line Break Accident In this accident, we assumed'that one of the four main steam lines outside of the drywell and reactor building is broken, and that in the five seconds before' closure of the steam line isolation valve, a total of 52,000 lbs. of steam and water'would be released'to the environs. We assumed that the concen-tration of activity contained in the reactor water would be ten times greater-than the concentration stated in the application; that the activity per pound -
of steam was the same as the activity per pound of water; and no rise in the steam cloud as it moved to the exclusion radius and beyond.
The calculated doses resulting from the steam-line-break accident are a function of the closing time of the steam line isolation valves, and the assumptions used in the blowdown acdel. Based on the applicant's blowdown model, no fuel cladding perforations would occur for a valve closing time of three seconds. Thus, the activity released would be that contained in the reactor water prior to the accident. For a longer valve closing time, an excessive
-number of rods would perforate, increasing the fission product release activity.
The steam line isolation valves have been designed with the capability to close in a minimum of three seconds. The General Electric Company will conduct proto-type tests to verify isolation valve closure times under simulated accident con-i ditions as discussed in Section 5.2.
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' 4.4 Rod-Droo' Accident at Hot-Standby
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This accident is assumed to occur.30 minutes after shutdown from an' extended
-period of operation'at 2500 MWt. A rod is assumed to drop out of the core while at hot-standby, and the resulting excursion is assumed to damage 330 fuel rods to the extent that they release all their noble gas and half of their iodine fission products.- The 330 rods damaged were calculated by the applicant as being the number'of rods in the core whose fuel enthalpies exceeded 170 cal /gm during -l
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the excursion. Fuel melting is considered to occur in the range from 220 to 280 cal /ga. Of the quantities released, all. the noble gases and 10%' of the iodines were assumed to flow through the steam line to the'hotwell where 50% 'of the lodines were assumed to plate-out.
The applicant has informed us that the released fission products will cause the main steam line radiation detector to alarm, and.the signal will trip the condenser vacuum pump and close the isolation valve on the exhaust side of the vacuum pump. A remote manually operated isolation valve will also be installed on the exhaust side of the vacuum pump. The applicant has estimated a transit time of six to seven minutes for the fission products to flow from I'
the steam line detectors to the condenser. The applicant considers, and we agree, that this is ample time. to close the manual valve if automatic, isolation fails.
Following condenser-vacuum pump isolation, the condenser will lose vacuum and fission products may diffuse from the condenser. We assumed that because of barometric pressure changes, fission products are released from the condenser at a rate of one percent per day.
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. . .. l 4.5 Refuelina Accident 1l This accident is a'ssumed to occur 24. hours after shutdown from an extended period of. operation. During ,'a fuel handling operation, a fuel bundle is assumed to fall onto the core with sufficient force to physically damage ~445' )
fuel rods. Twenty percent of the noble gases and ten percent of the iodine
>from these fuel rods are released into the pool water. Ninety percent of the iodine is assumed to be retained in the pool water. The calculational basis for predicting damage of 445 rods can be found in Section XIV of the PSAR. One
. half of'the iodines reaching the secondary containannt are assumed to plate-out within the building. Further, the noble gases and iodines remaining airborne within the building'are assumed to be discharged to the environs through the standby gas . treatment system and stack within two hours. Ninety percent of the ,
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,4 . 6 Conclusion ,
These accidents represent the spectrum of potential plant casualties and in our judgment are representative of the upper limit of what could occur in
.each case. For each of these accidents, we have calculated off-site doses within the 10 CFR Part 100 guidelines and consider our calculated doses to be acceptable-for this plant.
5.0 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION A number of areas requiring further analytical, experimental, design development, or testing efforts to substantiate the adequacy of system design and safety features of the present generation of boiling water reactors similar to Cooper Station have been identified. Some of these matters are pertinent 4
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-4 not only to boiling water reactors, but.to all large water-cooled power reactors, y particularly those employing high performance' cores similar to the Cooper '
reactor. There are other development areas which are of significance solely to 'l 4
. boiling water reactors. In the following sections, we have differentiated between these two categories. In addition, we have identified four areas in which further information will be required as the facility design progresses )
'but'which we feel are now. adequate.
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' 5.1 Development Program of Significance for all Larne Water-Cooled Power Reactors The development areas delineated in this category are:
(1) Linear Heat Generation Rate Fuel Damane Limit' A linear h'st e generation rate of 28 kilowatts per foot is used by the -
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applicant.as a fuel element damage-limit. In Amendment 2 of the PSAR, the applicant outlined a fuel elen.ent test program which will cover the worst anticipated transient heat generation rates, and maximum expected
- fuel burnup. Test fuel rods have been operated at various linear heat generation levels, and have verified calculational models. Additional work is planned which includes experience with high burnup of fuel (20,000 to 30,000 MWD /T) and long-term operation at high linear heat genera-tion rates, of capsules as well as complete fuel assemblies. These tests cover the T ' rum of anticipated operating conditions of Cooper Station and thus we believe that the work done to date and anticipated will solve outstanding questions in this area. The results of this test program are j expected to become available in 1969, prior to the date proposed for init;>1 operation of Cooper Station.
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. (2) ~ Local Fuel Meltina Resultina from Inlet Coolant Orifice Blockane The applicant has undertaken a research and development effort in this l
area. Preliminary results have indicated that flow blockage during normal operation is local in nature, and cannot propagate to affect the remainder of.the core.' Additional analytical and experimental work will be conductedL to confirm the results of these preliminary studies. Program completion is scheduled for early 1969.
(3) Effect of Fuel Clad Failure on Emergency Core Coolina Based upon analytical and experimental work done to date, clad per-
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foration occurs at a localized area of a fuel rod. Perforation is caused by high internal pressure and the point at which perforation occurs is random, depending upon a weak' point in the rod. Further experimental and analytical work will be continued in order to confirm and further refine the understanding of this fuel damage model. This work will include further perforation tests of fuel cladding under various conditions of temperature, pressure and metal ductility, further heat transfer analysis of fuel bundles under accident conditions, and other tests as appropriate. A report of the results of these tests is scheduled for the end of 1968.
Based upon the work done to date and the scope and schedule for the test programs, we believe there is reasonable assurance that this area will be satisfactorily resolved prior to the date propos,ed for initial operation of Cocper Nuclear Station.
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5.2 Development Program of Significance for Boiling Water Reactors The development areas delineated in this category are:
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(1) Core Spray Effectiveness Analyticalandtestworkiscurrentlyfunderwaytooptimizethecore spray systems for the General Electric boiling water reactors. Application l of the core spray system to the Cooper Station reactors will be based on i
the results of this development work. Appendix E of the PSAR provides a summary of the core spray test program.
To date, core spray tests have been conducted at fuel bundle powers greater than expected in Cooper, and at water flow rates lower than that which will be provided. The latest results on core spray tests have been l reported in Amendment 2 of the PSAR. These latest 49-rod fuel bundle tests indicate that wetting of both sides of a fuel bundle channel by the core spray flow can reduce peak clad temperatures significantly. Future experi-mental work will include testing at higher fuel temperatures and using zircaloy rather than stainless steel clad so as to more closely simulate actual reactor condition. Tests have also been done and will continue on spray distribution over a simulated reactor core. In view of the effort expended on this matter to date and the plans for continued work scheduled through 1968, we believe that this matter will be resolved prior to the date proposed for initial operation of Cooper Station.
(2) Steam Line Isolation Valve Testing General Electric Company is currently developing a program to test the function and closure time of main steam line isolation valves under simu-laced accident conoitions. Three specific programs have been planned (Amendment 2). The first includes small scale tests to observe the 5
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phenomenon of high speed steam'bwing stopped by~ valve closure; the second is a full. scale tesc wheretthe steam flow Lrat'E" through the valve is increased
, over normal flow;-and the third program 111 simulate accident situations where the -valve is subjected' to conditions of -high steam flow with water
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entrainment. Success in these programs;is' the ability of the valves. to r_r close.in a required period of nine.
The results of these tests scheduled for the end of 1968 are expected-to satisfactorily demonstrate the performance characteristics of the steam isolation valves. We expect that- this matter:will be satisfactorily resolved prior to the date proposed for ini$1afhperation of Cooper Nuclear 'I Station.
(3) Adecuacy of HPCI System as a Deeressuriser
_The principal function of the HPCI system is to maintain water inven-tories sufficient to assure core coqling for postulated small breaks. For intermediate breaks around 0.3 square feet, it serves to depressurise the pressure vessel to a low enough level so that.the core spray system or low pressure injection system'can reach rated flow. The HPCI system is designed to pump 4250 gpm into the rszetor pressure vessel within a re[ctor pressure range of about 1100 psig to 150 peig. J We'are continuing to follow the analytical techniques for predicting 1
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vessel depressurisation using the HPCI system. In Amendment 2 of the PSAR, i it was shown that over the spectrum .of liquid break sizes, for the HFCI-LPCI combination, a minimum mixing efficiency of about 85% was, required to prevent clad melting. However, in our opinien, the small difference between the calculated peak clad temperature for the HPCI-LPCI combination and the melting temperature of 3370'F does not allow adequate margin for uncertainties.
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The applicant has stated t$at the required mixing efficiency for the HPCI system will be obtained by optimizing the feedwater sparger design; i.e., by using u large number of small sparger holes to maximize the HPCI exit velocity and jet surface area, and by aiming the sparger holes to
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maximize wetted film area. This proposed optimization of the feedwater
' sparger design would be based on an analytical model which has not been verified. Therefore, the General Electric Company has formulated an
, experimental test program to determine HPCI mixing efficiencies. It is planned that the proposed tests will be completed in 1968. We believe that the principle of using the HPCI to depressurize the reactor has been adequately demonstrated. The experimental program should demonstrate the feasibility.
(4) Engineered Safety Features -- Electrical Eculoment Inside Containment Electrical equipment which must operate inside the primary containment in an accident environment is limited to isolation valve operators and cables. Where practical, the valves are designed to fail "as is" or closed (safe failure). A circuit failure after the valve has closed will be a safe failure. In addition to designing the equipment to withstand the accident environment long enough to operate the valves, the applicant will perfoca ouvitonmental testing. In Amendment 2 of the PSAR, the applicant stated that the manufacturers will test a sample of cable and a motor operator of the type to be installed in the Cooper Station primary l containment. The tests will demonstrate that the material and equipment l
l will survive the accident conditions of simultaneous pressure, temperature, and humidity for a period of time essential for their operation.
(5) Control Rod Worth Minimizer The applicant has stated that the basic system will have been tested, installed, and operated on a number of General Electric boiling water reactors prior to use at Cooper Station. A prototype system was installed in early 1965 in Dresden Unit 1 for test purposes.
We expect that the operating data that will be forthcoming from these reactor plants will be sufficient to determine the adequacy of the rod worth minimizer for Cooper Station.
(6) Jet Pump Development Considerable analytical and test work has been completed on the jet pump system for reactor coolant recirculation to establish its basic design characteristics. Additional development programs in progress, and planned, are summarized in Appendix D of the PSAR.
This development program, and the fact that this device will have been operated on other reactors prior to its application on Cooper Station will be adequate to determine its capability.
(7) Rod Velocity Limiter The rod velocity limiter, which is desig:ed to limit the free-fall velocity of a control rod is being tested. This device will also have been tasted during the pre-operational test phase in other boiling water reactors 4 prior to its application in Cooper Station.
(8) In-Core Neutron Monitor System In-core startup and power neutron detectors have been developed to 3 I
reduce neutron source requirements and to improve neutron flux monitoring
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I' capability in the startup and power ranges. Testing of these devices _ is presently being done in the Consumers fower Company's Big Roc'k Point. :
E reactor.. The applicant has stated that the in-core detectors inJ this - '
l reactor have given excellent results and, demonstrated ' satisfactory sensi- I tivities in' repeated counting cycles through suberitical and critical 1
operation and have' demonstrated counting ability during hot startup af ter a scram.. A life test will also be conducted to demonstrate the feasi-bility of leaving the: chambers in the high flux regions continuously._ i l
Identical chambers will have been in operation in boiling water reactors for several years before the first unit of this plant is operations. A.
complete report on this item is scheduled in 1968.
Because of the experience described, satisf actory in-core testing will have been conducted to demonstrate the adequacy of these monitors
prior to operation of Cooper Station.
5.3, Areas Reauirina Further Technical Information In this section, we will discuss safety issues which are of significance for most current General Electric boiling water reactors. Our intent in dis-cussing these matters with respect to this application is to indicate the status ;
of our continuing technical evaluation of this type of reactor.
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f (1) _ Failure of Passive Components of ECCS 1 The Cooper Nuclear Station emergency core cooling system design (ECCS) is i
the same as that provided for the T>rowns Ferry plant, the Vermont Yankee a
plant, the Peach Eottom 2 and 3, and all of the Dresden 2 class plants. The design function of the ECCS is to provide the capability of preventing core melt for the full spectrum of primary system breaks. To accomplish this function, there are provided multiple, but independent systems, consisting
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'i of the high pressure coolant injection (HPCI), the low pressure coolant injection (LPCI), the core spray, and auto-relief systems. With the loss ;
1 of off-site power, certain portions of the ECCS also have the' function of assisting in the removal of core decay heat.
The torus is the. primary water source for the core spray and residual heat removal system, and the alternate water source for the H'CI system which utilizes the condensate storage tank as the primary water source. ;
Isolation valves are located at the suction and discharge side of each pump. The torus water distribution system and ECCS pumps are located on 9 the lowest level of the reactor building. To assure that a leak in the torus or the distribution system to the pumps would not floow other ECCS equipment, each ECCS pump will have a separate suction line from the suppression chamber. The pumps of the individual systems will be spatially and structurally isolated from one another. In the event of leakage fror an active component of a system, the compartment housing the components will f i
serve to segregate and isolete the leakage and provide means of detection. !
Passive failures could potentially flood an individual pump compartment or could result in flooding of the concrete call in which the torus is 1 situated. The design will be such that only a single pump would be lost due !
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to flooding and also that NPSR for the remaining pumps will not be lost due i
to loss of torus water. )
f We believe that this design reasonably protects the ECCS from passive !
l failures which could negate ECCS function.
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'I (2)- Thermal Sh'ock The effect of thermal shock on the reactor vessel and its appurtenances q induced by_ injection o'f emergency core cooling water into the higher tempera-ture reactor system has not 'yet been completely, analyzed. The applicant has partially responded in Amendment 2. He will perform a complete I analysis to demonstrate that the primary and secondary stresses fros' thermal shock through operation of the emergency core cooling system will meet the criterion outlined in their response, which are:
- 1. Primary stress shall be within the applicable limits per the ASME Code Section III.
- 2. Secondary stress shall be determined and the associated strains-shall be limited so that the capability to safely shut the plant down is not compromised.
The completed analysis will be available within the next several months. We anticipate that the results of this analysis will verify pre-liminary results and show that the effects from thermal shock on the vessel and its appurtenances from a single event can be tolarated and that functional capability will not be lost. If the analysis does not verify the preliminary results, design steps will be required to minimize or eliminct- 7etential problems.
(3) Interchannel Flow Stability In Amendment 3 to the Cooper Station PSAR, the applicant referenced a technical paper titled " Technology of Boiling Water Reactor Stability Analysis" as being applicable to.the Cooper design. This paper presented
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3-a report on the technological status of the' dynamic. analysis of boiling i
water reactors. . The analytical model (FABLE II) which has been developed ,
to assess reactor power stability and inter-channel hydrodynamic stability 4 is summarized and evaluated by comparison to Garigliano Nuclear Reactor and KRB Nuclear Reactor rod oscillator data. The power density of these:
two reactors is lower than that in Cooper Station.
With the collection of additional supporting data relative to the General Electric second generation, high power density' core, we believe that the FABLE II code should be' adequate for the analysis of inter-channel flow stability. However, we believe that additional analyses.are required to delineate both the onset-of interchannel flow stability and the margin to onset during reactor operating conditions..
The General Electric Company has indicated that they are continuing their studies on interchannel flow stability and will keep us informed of 1
their findings as they become available. We intend to continue our con-sideration of this matter. . Additional analytical results and reactor operating data are expected to become available prior to the date proposed for initial operation of Cooper Naclear Station.
(4) In-Service. Inspection Provisions w e being incorporated in the design of Cooper Station to facilitate inspection of selected areas of the interior of the reactor vessel and its components, as recommended in the report, APED-5450. This report, titled " Design Provisions for In-service Inspection," was sub-mitted by the General Electric Company to the regulatory staff on May 4, 1967. We will revP v the applicant's in-service inspection program at the operating license stage.
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I (5) . Primary System Leak Detection -
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Detection of. leaks in the primary system will be accompli'hed s by J
D monitoring the sump level in the containment vessel, by monitoring the d$ T ;)
l of the cooling water of the containment. vessel' heat exchangers,.and by monitoring the containment vessel pressure. By these means, a range of leaks of. lass L than 1-gpa.up to 40-gym can be detected. Selected areas in j
the reactor building in the vicinity of the RCIC and RHR equipment will I also be monitored.
We believe that the development programs, including those proposed by the
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applicant and those being performed at other facilities, are reasonably designed to resolve any safety questions ~ associated with the above features of Cooper Nuclear Station and will provide the data necessary to construct Cooper Station in accordance with the criteria and specifications set forth in ths' PSAR.
6.0 STATION DESIGN kiiTH RESPECT TO THE 70 GENERAL DESIGN CRITERIA i
In November 1965, the Commission published its General Design Criteria for Nuclear Power Plant Construction Permits, and on July 11, 1967, published in the FEDERAL REGISTER its revised General Design Criteria taking into account comments received on the initial criteria and further development of the criteria by the regulatory staff. The applicant in Amendment Nos. 2 and 3 cross-referenced the information as presented in the application with the criteria. We have evalu-ated the application in conformance with the revised criteria and have concluded l that except as noted below, the proposed unit conforms to the revised criteria.
Recognizing that the proposed revised criteria may be modified again as a
. result of comments by interested parties, the staff intends to review the pro- i posed unit at the operating license stage in light of the criter,ia as formu-
. lated at that time. !
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l The capability of satisfying criterion 35 which relates to prevention of brittle fracture in the pressure vessel and the other parts of the primary system was discussed at length with the applicant and commented upon by the ACRS.
The applicant states that its pressure vessel will meet the' requirements of Criterion 35 as presently phrased, and has assured that the remainder of the primary system will meet the intent of the criterion. Details of the design approach to be followed by the applicant in meeting Criterion 35 for the balance of the primary system will be resolved between the applicant and the staff.
7.0 REPORTS OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS In a letter to the Commission, dated March 12, 1968, the Advisory Com-mittee on Reactor Safeguards (ACRS) reported on the proposed' Cooper Nuclear Station. A copy of this letter is attached as Appendix A.
In its' letter, the ACRS commented specifically on.the emergency on-site power system and Criterion 35 which relates to the coolant pressure boundary NDT requirement. These have been discussed in Sections 3.2.2 and 6.0, respectively. The Committee also noted that the items of concern enumerated in the Browns Ferry report apply similarly to the Cooper Nuclear Station. These items have been discussed in this report and will be resolved to the satis-faction of the staff and the ACRS prior to the issuance of an operating license.
The ACRS letter concluded ". . . the Committee believes that the proposed facility can be constructed with reasonable assurance that it can be operated without undue risk to the health and safety of the public."
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8.0' TECHNICAL QUALIFICATIONS
~ The~ CPPD.will be the sole' owner and operator of the Cooper Nuclear Station.
The applicant has~had extensive experience with conventional steam. generating plants. It also participated.in the construction and operation of the Hallam Nuclear Power Facility and thus employs a number of men with considerable' I previous nuclear operating and maintenance experience. The. nucleus of the 1 ' operating group of. Cooper will be composed of licensed operators and senior operators from Hallam. All of the staff and operators will also receive train-ing in boiling water reactor technology. The general training program will
-inilude academic training, programs conducted by the nuclear systems manu-
.facturer, on-site classroom instruction, and'on-the-job training at other boiling water reactor plants. .
Based on tho' experience that has been attained with Hallam and the other considerations above, we believe that CPPD possesses adequate technical depth and experienca to follow the design and construction of the Cooper Station to
' assure it follows the design.
The nuclear contractor, the General Electric Company, has been engaged for j
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a number of years in design, construction, and operation of boiling water reactors. Operating reactors with which GE has been associated include the Commonwealth Edison Company's Dresden Unit 1, ths Pacific Gas and Electric Company's Humboldt Bay Plant, and several power reactors now operating in ,
foreign countries. The company is also engaged in the design of such plants as the Dresden Units 2'and 3, the TVA plants, the Philadelphia Electric Company's Peach Botton Units 2 and 3, and many other domestic plants. i c j
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, The Burns and Roe Corporation, the applicant's construction contractor,.
has demonstrated its knowledge of nuclear and related technology by participa-- l s
tion in the design and construction ofthe Jersey Central Oyster Creek nuclear
- power generating stationh the 'Hanford'New' Production Reactor, and the Shipping- 'I port Atomic Power Plant. The corporation'has also participated in the design and construction of a number of large fossile fueled electric generating plants..!
' i On the basis of the abov's considerations, we believe that the Consumers Public Power District and its contractors, the General Electric Company, and the Burns and Roe Corporation, are suitably qualified to design and construct Cooper Nuclear Station.
9.0 COMMON DEFENSE AND SECURITY The application reflects'that the activities to be conducted would be within.the jurisdiction of the United States and that all of the directors and principal officers of the applicant are American citizens. We find nothing in the application or otherwise to suggest that the applicant is owned, controlled _ i or dominated by an alien, a foreign corporation, or a foreign government. The activities to'be conducted do not involve any restricted data, but the appli-cant has agreed to safeguard any such data which might become involved in accordance with paragraph 50.33(j) of 10 CFR Part 50. The applicant will rely upon obtaining fuel as it'is needed from sources of supply available for civilian purposes, so that'no diversion of special nuclear material from military purposes is involved. Por these reasons and in the absence of any information to the contrary, we have found that the activities to be performed
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will not be inimical to the common defense 'and security.
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10.0 CONCLUSION
S Based on the proposed design of the Consumers Public Power District's Cooper Nuclear Station, on the criteria, principles and design arrangements .
, I for systems and components.thus far.describedv which include all of the l important safety items, on the calculated potential consequences of routine and .'
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accidental release of radioactive material.s to the environs, on the scope of ' ;
the development. program which will be conducted, and on the technical competence. I of the applicant and the principal contractors, we have concluded that, in accordance with the provisions of paragraph 50.35(a),10 CFR Part 50 and
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2.104(b) 10 CFR Part 2:
(1) The applicant has described the proposed design of the facilities, including the principal architectural and engineering criteria for the design i and has identified the major features or composants for the protection of the health and safety of the public;.. .
(2) Such further technical or design information as may be required to ]
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complete the~ safety analysis and which can, reasonably be left for later con- 1 sideration, will be supplied in the final sa'fety analysis report.
i (3) Safety features or components, which require research and development l have been described by the applicant and the applicant has identified, and there will be conducted a research and development program reasonably designed to )
I resolve any safety questions associated with such. features or components; l
(4) On the basis of the foregoing, there is reasonable assurance that l (1) such safety questions will be satisf,actorily resolved at or b.efore the
' latest date stated in the application for completion of construction of the 1
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. l 47 proposed facility and [2] taking into consideration the site criteria contained ,
in 10 CFR Part 100, the proposed f acility can be constructed and operated at the proposed location without undue risk to the health and safety of the public; j (5) The applicant is technically qualified to design and construct the proposed facility; and (6) The issuance of a permit for the construction of the f acility will not be inimical to the common defense'and security or to the health and safety of the public.
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APPENDIX A l 1
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
. ' UNITED STATES ATOMIC ENERGY COMMISSION '
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WASHINGTON. D.C. s0548 MAR 1,21968 ; -
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Honorable Clenn T. Seaborg ,
qttairman -
i Uj S. Atomic Energy Conunission shington, D.'C, 20645
Subject:
?1EPORT ON COOPER NUCLEAR STATION
Dear Dr. Semborg:
At its ninety-fifth meeting, on March 7-9, 1968' the Advisory Committee en Reactor Safeguards completed its. review of the ' application by the ,
i posumers' Public Power District of Nebraska for autliorization to con-el:ruct its Cooper Nuclear Station. The project was previously considered 4p ACRS Subcomunittee meetings held on October 10,*1967 at the plant site,
. s 29, 1968 in Washington, D. C; During its review, the
$adonFebruary ommittee had the. benefit of discussions with representatives of Consumers
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6 Public Power District, General Electric Company, Burns and Roe Inc., and .
the AEC Regulatory Staff. The Coaunittee.also studied the documents
- listed. ,
The Cooper. Nuclear Stati'on will be located about 60 miles south of Omaha, Nebraska, on the west. bank of the Missouri River. The plant is designed to produce 238L We (801 We). The reactor in this plant is a high power Jensity boiling water unit, with design features similar to those of the
- previously reviewed Browns Ferry Nuclear Power Station reactors.
The soils ne'ar the, river at this. location consist primarily of saturated,,
poorly compacted sands which may be susceptible to the phenomenon of
. liquefaction if shaken by a sufficiently severe earthquake. The' applicant will improve the foundation support of the plant by excavation to within a few feet of bedrock, dewatering, replacing, and compaccing the sand to ,~
.a density sufficient to prevent liquefaction if such an earthquake should occur. The reactor containment will be placed upon a concrete mat which, in turn,,will rest .upon the. compacted sand.
The Committee recommends that the applicant give further consideration 'to the design of the emergency on-spa power system to avoid the need for synchronization of the diesel-driv' en generators.
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l Discussion of General Design Criterion Number 35 (10 CFR 50.34 proposed) has' occurred in connection with this review. The manner in which the- 1 i intent of this criterion will be met for the Cooper Nuclear Station should )
be re' solved between the applicant and the AEC Regulatory Staff. j l
fThe Convaittee, in its report i:o you of March '14,1967 on the Browns Ferry )
' Nuclear Power Station, called attention to a number of matters that warrant !
jf:areful consideration with regard to'reactora of this type and power den- ,
.(sity. These matters apply similarly to the Cooper Nuclear Station.
The Advisory Conunittee on Reactor Safeguards believes that the items men-tioned.above can be, resolved during construction. The Comittee believes that, if due atteation is given to the foregoing conunents, the propo' sed facility can be constructed with reasonable assurance that it can 'be oper-
. -ated without undue risk to the health and safety of the public.
Sincerely yours,
.A?
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~, , Carroll I. Zabel Chairman .
References ,
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- f. Consumers Public Power District License Application for the Coc,per Nuclear Station, dated July 26, 1967;' Preliminary Safety Analysis L Report, Volumes I, II, and III.
- 2. Amendment No. I to License Application, dated December *1,1967.
,3 . Amendment No. 2 to License Application, dated January 25, 1968.
4 Amendment No. 3 to License Application, dated February 10, 1968.
- 5. Amendment. No. 4 to License Application, dated February 10, 1968.
- 6. Amendment No. 5 to License Application, dated February 20, 196,8.'
- 7. Amendment No. 6 to License Application, dated March 2,1968.
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f APPENDIX B CHRONOLOCf '- REGULATORY REVIEW OF THE .
' CONSUMERS PUBLIC POWER DISTRICT'S PRELIMINARY SAFETY ANALYSIS REPORT o .i 1
- 1. July 27, 1967 Submittal of Preliminary Safety Analysis Report.
2... August. 30, 1967 Heating with applicant to discuss the liceneing schedule.
- 3. September 28,.1967 Meeting with applicant to discuss site related subjects and changes in the plant design to be described in ;
Amendment-1.
- 4. October 9, 1967 Heating with applicant to discuss _the station structural design as related to local geology and seism 61ogy.
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T5. October 10, 1967 ACRS Subcommittee visit to sita.
- 6. October-24, 1967 Heeting with applicant to discuss containment and components, control and instrumentation, and electrical' power supply systems. ,
- 7. December 1, 1967 Submittal of Amendment No.-1 to PSAR, revised pages reflecting some plant design changes.
- 8. December 20, 1967 Lett'er from DRL to applicant requesting additional information.
- 9. January 25, 1968 Submittal of Amendment No. 2 to PSAR: answers to DRL 4 request for additional information of December 20, 1967.
- 10. February 10, 1968 Submittal of Amendment No. 3 to PSAR; corrected pages '
to be inserted in Amendment No. 2; supplemental information to cla-1"* and correct material previously submitted.
- 11. February 10, 1968 Submittal of Amendment No. 4 to PSAR; information on f L- ,
the Low Pressure Coolant Injection System Logic. f i
- 12. February 14, 1968 Heeting with applicant to discuss Amendments 2 and 3. ]
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- 13. February 20, 1968 . Submittal of Amendment No: 5 to PSAR; corrected pages to be inserted in Amendments 2 and 3;. supplemental L
information to clarify and correct information pre-viously submitted. 3 d
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- 14. February'29, 1968 ACRS Subcoannittee meeting.
- 15. March 2, 1968 Submittal' of Amendment No. 6 to PSAR; clarification of " Functional Loading," " Deformation," " Design Criteria-for Battery Systems," " Design Criteria for Critical Piping and Structures," and " Design Tables."
- 16. March 8, 1968 ACRS' meeting.
- 17. March 12, 1968 ACRS letter to Commission on Cooper Nuclear Station.
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i APPENDIX C 1
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Comments on Cooper Nuclear Station Consumers Public Power District ,
Preliminary Safety Analysis Report Volumes I,11 and III dated July 27, 1967 Prepared by Environmental Meteorology Branch Air Resources Laboratory Environmental Science Services Administration ,
i September 29, 1967 The site location in the central plains of the United States would indicate a well ventilated region and a surface inversion frequency between 30 'to 40 perecent of the time, depending upon season.
The meteorological diffusion evaluation methods - including that for the main steam line break - are identical to.those used for the Browns Ferry Reactor and numerous other similar reactors. Our comments now, as then, are similar, namely, the questionable validity of using a strictly empirical cloud rise equation to compute a steam cloud rise to 3800 feet and the crucial assumption that the intial distance of 300 feet between plume centerline and the ground is maintained.
The topography of the site and surrounding area was not adequately discussed and no topographic map could be found to assess the possi- /
bility of the plume centerline being closer to the ground than 300 feet. l I
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3-6-6e APPENDIX D Consumers Public Power District Cooper Nuclear Station Site Nemaha County, Nebraska AEC Docket 50-298
. Hydrology
\ The site is located on the right bank of the Missouri River 532.5 miles upstream from its confluence into the Mississippi River between Brownsville and Nemaha. The Missouri valley is about 6 miles wide in this reach and is about 150 feet lower than the crest of the bordering bluffs. The river bed has been realigned and ' levees on both sides of the river which were overtopped during the 1952 flood have been' restored.
The highest recorded flood in this reach of the Missouri occurred in April 1952 when a discharge of 414,000 cfs (eubic feet per second) was ;
observed at the Nebraska City gage about 30 miles upstream from the site. !
The valley was flooded from bluff to bluff from Yankton, S. D. to St.
Joseph, No. except for a few areas where levees, were not overtopped. The levee on the right bank of the river at the site was not overtopeed but the. valley on the left. bank was inundated. The flood is reported to have reached a stage of 899 feet asl (above mean sea level). The recur-rence frequency of the 1952 flood is estimated to be somewhat smaller !
than once in a hundred years. Since 1952 major reservoirs on the Missouri River have been completed which would provide flood storage to reduce the probable frequency of occurrence of major floods. However, floods of considerably greater magnitude than that of 1952 could not be completely controlled. f The Engineering Division of the Omaha District of the Corps of Engineers has furnished the applicant with the following computations and estimates.
- 1) Discharge over the Fort Randall Das spillway due to the maximum probable flood entering the reservoir has been computed as 620,000 cfs. 1
- 2) Maximum probable discharge at the site due to a rainctorm on the drainage area' upstream from the site was estimated to be 600,000 cfs; corresponding stage at site was esti-mated to be 902-903 feet asi.
- 3) Discharge due to a maximum probable rainstorm upstream from the site coincident with a failure of Fort Randall Das I
was estimated to be 1,200,000 efs; the corresponding stage !
was estimated to be 905 306 feet as1.
- 4) The time required for's flood wave to travel from Fort Randall i Dam to the site was estimated to be at least.3 days. ]
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l The applicant has chosen 903.5 feet asi as the flood protection level i for all structures essential to the reactor, with the provision that temporary protection could be made upon warning of high flows acrose Fort Randall Dam.
The level of flood protection chosen appears reasonable prodided that I the proposed temporary measures can also be put into effect after the levee on the right bank of the river is overtopped or breached.
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The analysis of th'e geology of the Cooper Atomic Power station, Nebraska, as presented in Asc Docket No. 50-298 and supplements, was reviewed and 1 compared with the available literature. The analysis appears to be carefully derived and to present an adequate appraisal of those aspects of the geology that would be pertinent to an engineering evaluation of
-the site .
Although there are no ide' n tifiable local geologic structures that could be expected to ' increase thd earthquake potential of the site, there are two regional ~ geologic structures that could be seismically significant to the site. These are the Nemaha anticline, a northward-trending, j major structural feature of the midcontinent, and the associated Humboldt j fault. The'Humboldt fault, which trends northward frau Kansas through j southeastern Nebraska along; the steep east limb of the anticline, approaches to within about 20 miles west of the site at its nearest
. point.. Several earthquake epicenters appear to be located in the immed-inte vicinity of these two structural features. )
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In order to assure that the compacted granular fill will have the specified physical properties necessary for.che adequate and safe support j of the major plant, structures, it is essential that the fill be emplaced under' continuous and competent engineering controls, s
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APPENDIX E REPORT ON THE SIS ' SEISMICITY 1
POR THE COOPER NUCLEAR STATION, NBEASKA At the- request 'of' the Division of Reactor Licensing of the Atomic Energy Commission, the_ Seismology Division of
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the Coast and Geodetic Survey has evaluated the seismicity of the area around the proposed reactor site between Browns-ville and Nemaha, Atchison County, Nebraska, and has reviewed ,
the similar analysis made by the applicant in the "Prelimi-nary Safety Analysis Report" of the Consumers Public Power District. The applicant's report of the seismicity of the area is comprehensive and satisfactory for evaluating the seismicity factor of the site. /
Based upon the review of the seismic history of the site and the surrounding area and the reisted geologic con-siderations, the Coast and Gecdetic Surv6y agree with the applicant' that an acceleration of 0.10 g on rock would be adequate for representing earthquake disturbances likely to . !
occur within the lifetime of the facility. In addition, we agree that the applicant's proposal to design for an accel-eration of 0.20 g on rock would adequately protect the l' ,
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2-j facility for ground motion from the maximum earth. quake likely'- .
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$ l to affect this site, i U. S. Coast and Geodetic Survey Rockville, Maryland 20852 December 12, 1967 l 3
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j APPENDIX P I
J REPORT TO AEC REGULATORY STAFF ADEQUACY'0F THE STRUCTURAL CRIT'ERIA POR 1
COOPER NUCLEAR STATION' l CONSUMERS PUBLIC POWER DISTRICT (Docket No. 50-298) t i
by )
if N. M. Newmark, j W. J. Hall and A. J. Hendron ]
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March 1968
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NATHAN M. NEWMARK
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Consulting Engineering Services
- 1114 Civil Engineering Building I Urbana, Illinois 61801
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APPEND 1X F E
ADEQUACY OF THE STRUCTURAL CRITERIA FOR THE
. COOPER NUCLEAR STATION
- by i
-N. M. Newmark, W. J. Hall and A. J. Hendron, Jr.
INTRODUCTION This report concerns the adequacy of the containment s'tructures and compo-nants, nuclear piping and reactor internals, for the Cooper Nuclear Station
, for which application for a construction permit and facility license has br en made to tho' U. S. Atomic Energy Commission (AEC Dc,cket No. 50-298) by i tle Consumers Public. Power District. The facility is .co be located in 4 Nomaha County, Nebraska, on the west bank of the Missouri River, 3 miles-
.NE of Nemaha,, Nebraska', and 10 miles SE of Auburn,' Nebraska.
Specifically.this report is concerned with'the evaluation of the design criteria that determine the ability of the containment system, piping and'
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reactor internals to withstand a design earthquake acting simultaneously with other applicable loads forming the basis.of the design. The facility also is to be designed 'to withstand a maximum earthqaake simultaneously-
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with other' applicable. loads to the extent of insuring safe shutdown and containment. . This report is based on information and criteria set forth in the Preliminary Safety Analysir, Report (PSAR) and amendacnts thereto as
' listed at the end of this report. We have participated in discussions with the AEC Regulatory Staff, and the applicant and its consultants, in which many of the design criteria were discussed in detail.
DESCRIPTION OF THE FACILITY F 'Ihe Cooper Nuclear Station is described in the PSAR as a single-cycle forced-circulation boiling water reactor for producing stes.m for direct use in a steam turbine. The nuclear unit is to be furnished by the General Electric Company and is designed for an initial power output of 2381 MWt !
(778 MWe net). . In most respects the design for this nuclear system is sub-stantially similar to the TVA Browns Ferry Nuclear Power Station.
The primary containment system, which houses the reactor vessel and the recirculation system, consists of a drywell, a pressure s,uppression chamber, (shaped like a torus and containing a large pool of water), isolation
, . valves, containment cooling , systems, and other service equipment. The dry-t wel1~is a steel pressure vessel with a spherical lower portion 65 feet in diameter and a cylindrical upper portion 35 ft. -7 in. in diameter. The over-all height of the drywell is approximately 111 f t. 3
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- j The reactor building provides secondary containment, and consists of cast- l
. in-place reinforced concrete exterior walls up to the' refueling floor and .
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of steel frame with insulated metal siding' above~ this floor. This siding j Q.; Lis installed with-sealed ' joints. j
~ The geologic strata' in the region in~ order of. increasing depth are soil
' deposits, sedimentary rocks, and deep basement igneous rocks. At che site, the alluvial deposits in the flood plain vary in thickness from 62 '
to 71. feet,. and are of two major sub-types of different geologic origin, namely surficial fine-grained soils and underlying sands. . The surficial' soils consist of meander-belt and back-swamp deposits, ranging'in thick-ness from 10 to'25 feet. .These deposits include silty sand,-sandy silt, silty clay and clay and are encountered in localised pockets and.in com-
. plex combinations. The underlying sands are of'either alluvial or glacial outwash deposits or both,' with lenses of clay, coarse sand and fine gravel
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distributed irregularly throughout the deposit. 'The bedrock immediately under the site consists of horizontally bedded shale and limestone'with occasional thin. coal seams.- '
i The closest evidence of faulting is 20 miles to the west, the Nemaha anti-cline and its associated Humboldt-fault. Neither.of these is recent.
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l There is no evidence at the site of either a: fault or other bedrock dis-
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-continuity.
1 SOURCES OF STRESSES IN CONTAINMENT STRUCTURE I AND CLASS I COMPONENTS The primary containment-system, which includes the drywell, vents, torus, and penetrations is to be designed for the following conditions as noted
. in the PSAR; pressure suppression chamber and drywell design pressure,
+56 pois and -2 psi; design' temperature of drywell and suppression cham-bar, 281*F. It is noted on page V-2-3 of the PSAR that the drywell is to be designed for live and seismic loads imposed coincidentally on the shell with the design pressure and temperature noted.
All structures will-be designed to withstand a wind velocity of 100 mph.
Where failure of equipment and structures could affect the operation and function of the primary containment and reactor primary system the unit will be designed to assure safe shutdown under possible short-term tornado loadings with a wind velocity of 300 mph.
As noted in Section XII of the PSAR, the seismic forces to be used in the structural' design of Class I buildings and equipment will be based on an earthquake with a maximum ground acceleration of 0.1g. It is further noted that the combined stresses resulting from functional loadings and
'" from an earthquake having a maximum ground acceleration of 0.2g will be such that safe shutdown can be achieved.
The reactor building, which comprises the secondary containment system, is listed as a Class I -- critical structure. The reactor building is 2
)
m _ _ _ _ . . _ _ . ._ m
-3
. i:o be' designed to withstand an internal pressure of 7 in. of water (about one-quarter poi) without structural failure and without pressure relief. i
. In other respects.the design follows the same practice as noted for the-primary containment system. .
Other Class I' equipment and structures are to be designed to withstand.
the same seismic forces' and functional loads noted for the primary and secondary containment systems, with the exception of several special conditions noted in.the PSAR and discussed later in this report.
. On page XII-2-7 of the PSAR it is noted that " Class II structures and equipment will be designed to resist effects of seismic loads with the horizontal base shear coefficient as determined from the. Uniform Building Code, or taken as 0.103,.whichever is' greater." Such design will employ a one-third allowable-increase in basic stress. All such equipment will be bolted or fastened so that it will not be displaced if friction is non-exis tent. Since generally the base shear coefficient of 0.10g will govern, this is more conservative than the Uniform Building Code, and.
therefore we concur. 1 COMMENTS ON ADEQUACY OF DESIGN l
Foundations
. It.is noted in the PSAR' that the reconunended foundation scheme consists of reinforced concrete" mat foundations for the major. structures (reactor build-ing, turbine-generator building, pump house, and discharge facilities), and 4 isolated spread footings and slabs on, grade for the appurtenant structures.
The mat and the spread footing foundations are to_ be supported by a struc-tural fill extending from the bottom of any particular foundation to the l bedrohk surface. In Amendment 2 on page A-II-25 the applicant suggests i that the in-situ soils be replaced with a compacted structural fill or be I compacted in-situ at the following relative densities as defined on pages A-II-7 and A-II-8 of Amendment 2: (1) Dr= 85 percent from the final ground surface at EL 903 to EL 855;- (2) D = 80 percent from EL 855 to EL 830; and (3) D = 75 percent from EL 830 to Eedrock surface at approximately EL 820.
Ihese# values of relative density provide a factor of safety of 1.5 against liquefac'tion.
,> In addition, the construction of this fill will be controlled by the insta-lation of piezometers, a carefully conceived and executed plan of control-ling the placement density, and will be verified by borings made af ter the structural fill is complete. The details of this portion of the program .
j are outlined in Amendment 2. !
We concur in this approach to the foundation construction.
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Seismic' Design Criteria In conneccion with the design of the Class I structures and components, we epucur in the approach proposed of a basic design for a design earthquake with a maximum horizontal ground acceleration of 0.10g and a maximum earth-quake of 0.20g maximum ground acceleration with provision for safe shutdown. i These-values are consistent with those recommended by the U. S. Coast and Geodetic Survey (Ref. 3).
As noted in Section XII of the PSAR, the maximum horizontal acceleration and maximum vertical acceleration will be considered to act simultaneously.
The vertical ground acceleration will be assumed to be equal to one-half the horizontal ground acceleration ard, where applicable, stresses will be added directly. We are in agreement s (th this approach, since the appli-cant has stated in Amendment 6 that the addition of the stresses as noted refers te addition of all streasse arising from dead load, live load, pres- ;
sure, thermal effects, etc.-
The damping values which are to be employed in the design are given in
' Table XII-2-1 and further elaboration on the damping value for concrete is given in Section XI of Amendment 2. We are in general agreement with the values listed to be employed in the design and, unless noted other-vise, assume that these values will be used for both the design and maxi-mum earthquake design conditions.
A table of allowable stresses for the primary containment is given in Section XII of the PSAR and we are in agreement with the value of 69 percent of yield for primary membrane plus bending for the load'combina-tion involving the design seismic loading. For the maximum earthquake loading condition it is noted in the table that the combined earthquake a and functional load stresses probably will not exceed yield stress. How- _j ever, where calculations indicate that a structure or piece of equipment '
will be stressed beyond yield point an analysis will be made to determine that its energy absorption capacity exceeds the input from the earthquake i and, further, that resulting deflections or distortions will not prevent I the proper functioning of the structure or piece of equipment and will I not endanger adjacent structures or components. A similar approach is advocated in Table XII-2-3 wherein allowable stresses are given for the reactor bui.di.;. :.'a are in agreement with the general approach adopted !
for the primary containment and reactor building. We believe it is acceptable provided that the deformation limits to be selected result in an adequate margin of safety.
The earthquake design spectra are presented in Figs. I-1 through X-4 of Appendix A, Vol. 3, of the PSAR, and further elaboration on the choice of the response spectra are presented in the foundation design section of Amendment 2. The spectra to,be employed are patterned after the 21 July 1952 Kern County Earthquake (N69W) recorded at laf t, California.
We concur in the use of this spectrum for the site, j l
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1 The design of the intake structure, which is a Class I structure, is discussed in detail in Section VII of Amendment 2. From the information presented in the PSAR in Section'XII, it appears that the foundation for the pump house will also be a structural fill. On the basis of the infor-nation presented-with regard to the foundation, flood damage and possible damage by collision with. river traffic, ice, and by seismic forces, we believe that the design approach being undertaken is satisfactory.
The design of the penetrations is covered in answer to questions pre- q sented in Sections XII and XIV of Amendment 2. It is noted that for the maximum design earthquake condition the stresses resulting from the load combination given can be greater than the yield strength; however, it will be established that energy absorption capacity exceeds energy input.
It is noted lacer in the same section that the deformation limit asso-ciated with the allowable stress exceeding the yield strength is an inelastic deformation which will not violate containment integrity.
This statement does not provide a complete basis for judging the margin of safety. However, we are in agreement with the general approach pro-vided that the definitive limits of deformation to be selected result in an adequate margin of safety.
The answer.to the question concerning control room instrumentation and connections (Section XIII of Amendment 2) indicates that control room instrumentation, including panels required for safe plant shutdown, will I be designed to withstand seismic effects including tilt. Further, the j applicant has stated in Amendment 6 that the batteries and battery sup-ports (Class I items) will be designed to withstand the expected seismic loadings. We are in agreemert'with the applicant's approach .in these j areas.
The discussion in Section XXI of Amendment 2 on the emergency core cooling system indicates that the ECCS pumps will be made watertight to the maxi-mum level contemplated to protect the pumps from flooding. We concur in this approach.
For the reactor vessel and internab, a discussion of the design criteria is given in Section XIX of Amendment 2. A statement is made there as follows: "The probability that a major earthquake should occur coincident with other major accidents becomes vanishingly small. Consequently, load-ings under this hypothetical condition have not been studied in detail." i It is not clear to us that such a loading condition is inconceivable. Some l elaboration has been provided in Amendment 6. The application states that l Class I items will be designed for combined loading including concurrent I operating, accident, and seismic loads. The criteria presented by the i applicant are generally acceptable provided they result in an adequate margin of safety.
1 It is noted in the PSAR in several places that the isolation valve design is l under study and that tests are to be conducted. We concur in this approach. I 1
d 2
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, ~6-4 CONCLUSIONS 4
In keeping with the design goal of providing serviceable structures and components with a reserve of strength and ductility, and on the basis of the information presented, we believe that the design criteria out-lined for the containment and other Class I components can provide an adequate margin of safety for seismic resistance.
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REFERENCES j
- 1. - " Preliminary Safety Analysis Report - Vol. I, II, and III," Cooper i Nuclear Station, Consumers Public Power District, 1967. .
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- 2. " Preliminary Safety Analysis Report - Amendments 1, 2 'and 6 " q Cooper Nuclear Station, Consumers Public Power District, 1967, 1968. 1
- 3. " Report on the Seismicity of the Cooper Nuclear Station Site," U. S.
Coast and Geodetic Survey, Rockville, Maryland, December 13, 1967. ;
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. APPENDIX C IN T4nP6Y REFCR TO:
d h UNITED STATES
- h,yo [._.... i \
DEPARTMENT OF THE INTERIOR 4.E.;-f FISH AND WILDLIFE SERVICE
% *' WASHINGTCN. D.C. 20240 C 21 ty, i
Mr. Harold L. Price Director of Regulations U. S. Atomic Energy Commission
-Washington, D. C. 20545
Dear Mr. Price:
This is in response to questions raised by your staff on the cor:=ents ;
of the Fish add Wildlife Service concerning the application by Consumers Public Power District for a construction permit for the "
proposed Cooper Nuclear Station, Nemah County, Nebraska, AEC Docket '
No. 50-298.
In our letter of February 21, we recommended that the applicant be required to redesign the liquid ,aste disposal system to insure that )
radioactive wastes resulting fron accidental leaks or failure of the I waste sample and floor drain s:mple tanks would not reach the river. {
The applicant's statement in the Pre W % ery Safety Analysis Report I concerning the vaste holding tenb was interpreted to mean that the yearly maximum permissible concentration might be exceeded during an s l
accidental release of the tanks.
l We have been informed by your staff that this was a misinterpretation I
- and have been assured that the yearly =rxir.um per. issible concentra-tion would not be exceeded even during an accidental release. If this is, in fact, the situation, recommendation No.1, page 3 of our letter of February 21 voald not be necessary. We are concerned, however, that the maximam permissible concentration may not always
' guarantee that fish and wildlife vill be protected from long-range 4 adverse effects; therefore, we request your consideration of the l remaining recommendations. '
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- Sincerely yours, )
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C^~4ssioner ,
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' $J 'gtQ;, UNITED STATES m acru aura To l
- l; DEPARTMENT OF THE INTERIC.3 r
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- , ' _q FISH AND WILDLIFE SERVICE h*l]' U is.,c,g, WASHINGTON. D. C. 202/.G .% y a n
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Mr. Harold L. Price o n Director of Regulations EO *-
U. S. Atomic Energy Comission " A' O Washington, D. C. 205h5
Dear Mr. Price:
This is in reply to Mr. Boyd's letter of August 16, 1967, requesting our coments on the application by Consumers Public Power District, for a construction permit and facility license for the proposed Cooper Nuclear i Station, Nemaha County, Nebraska, AEC Docket Nc 50-298.
The proposed power station would be constructed on the Missouri River about 2 miles south of Brovnville, Nebraska, at river mile 532.5. A
- boiling water reactor designed for an output of 2,381 thermal megawatts and a net electrical output of 778 megawatts w,uld be used as a power source.
A radioactive waste disposal system and other facilities required for a complete and operable nuclear power plant will be provided.
Condenser cooling water would be pu ped from the Missouri River at the rate of 1,h50 efs in the sumer and 1,000 cfs in the winter. After
{
i absorbing radioactive and heat vastes, the water would be returned to the i Missouri River through a short flura. The applicant plans to conduct pre-and post-operational environmental radiological surveys, which would l include plant site.
conection and analysis of aquatic biota both above and below the Pre-operational surveys would begin at least two years before reactor start-up. Details of the surveys, such as sampling station sites )
and frequency of sampling, are not yet ava m ble.
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Under present flow regulation, a minimum flow of 31,000 cfs at Nebraska City has been maintained from March through September for navigation purposes. During the winter months, a flow of a: least 3,000 cfs is needed for water quality control. In actual in recent years have been 6,000 cfs or more. practice, the winter flows Under prolonged drought i
conditions, however, the navigation season win be shortened so that mini-mum flows for water quality control during the vinter can be maintained.
1 Fish species of importance for sport fishing in he Missouri River are I
sauger, channel catfish, flathead catfish, blue catfish, carp, and crappies.
Most of these fishes also contribute to the enmeial fishery. In addi-tion, fish from the Missouri ' River move into tributary streams, especially during spring, thereby contributing substantially to the fisheries in these streams. ,
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. The area influenced by the Missouri River in eastern Nebraska and western Missouri is an important migration route for ducks and geese.
It provides resting and feeding areas for many species of waterfowl, including blue geese, Canada geese, baldpate, teal, pintails, and I mallard. The Squaw Creek National Wildlife Refuge is about 28 miles I southeast of the project site in northwestern Missouri. Also, the Nebraska Game and Parks Commission owns and operates the Plattsmouth State Special Use Area for public waterfowl hunting at the confluence l of the Platte and Missouri Rivers, about 50 miles above the plant site. !
Three additional National Wildlife Refuges have been proposed for j acquisition and development on the river between Yankton, South Dakota, l and Sioux City, Iowa. These refuges, together with intensified State and Federal management, are expected to increase the number of water-fowl using the Missouri River Valley ani adjacent lands. ;
l It is stated in the Pre W w y Safety Analysis Report "If a tank were ,
to fail and the vastes discharged to the canal, the activity discharged l would not be different from that if they were discharged over a longer period within permissible concentration limits. Since Title 10, l Part 20 of the Code of Federal Regulations permits averaging of dis- i charges over a period of a year, the effect on the yearly average would I be small." These statements imply that the maximum permissible concen- l trations in the waste materials might be exceeded for short periods of l time.
Title 10, Part 20, Code of Federal Regulations established maximum permissible levels of radioactivity that can occur in drinking water I without harmful effects for man. However, operation within these limits may not a1 ways guarantee that fish ani wildlife will be protected from adverse effects. Radioisotopes of many elements are concentrated and stored by organisms that require these elements for their normal metabolic activities. Some organisms concentrate and store radioisotopes i of elements not normally required but which are chemically similar to elements essential for metabolism. In both cases, radionuclides are !
transferred from one organism to another through various levels of the s food chain as are nonradioactive elements. These transfers may result i in further concentration of radionuclides and a wide dispersion from the project area by migratory or wide-ranging fish, mammals, and birds.
To prevent the possibility of excessive concentrations of radionuclides in .the aquatic biota and associated fishes, mammals, and birds of the Missouri River,as a result of accidental leaks or failure of the waste sample and floor drain sample tanks, the liquid waste disposal system should be redesigned so that such accidental release of radioactive liquid wastes cannot reach the river.
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, 4 The pre- and post-operational radiological surveys planned by the applicant should include studies of the effects of radionuclides on selected organisms indigenous to the project area which require the waste elements, or similar elements, for their metabolic processes.
These studies should be planned in cooperation with the U. S. Fish and Wildlife Service, the Federal Water Pollution Control Administration, the Nebraska Game and Parks Commission, the Nebraska Department of Health, the Misasuri Department of Conservation, and the Missouri Water Pollution Board.
i If it is determined from pre-operational surveys that release of radio-active effluent at levels permitted under Title 10, Part 20, Code of Federal Regulations will result in harmful concentrations of radioactivity ,
in fish and wildlife, plans should be made to reduce the discharge of !
radioactive material to acceptable levels. Post-operational surveys should be conducted to evaluate the predictions based on the pre-operational surveys and to insure that no unforeseen damage occurs.
Because of the impo'rtance of the fish and wild 14 fe resources of the Missouri River in the area of influence both above and below the proposed site, it is imperative that every possible effort be made to protect these valuable resources from radioactive contamination. Therefore, it is recommended that the Consumers Public Power District be required to:
- 1. Redesign the liquid waste disposal system to insure that radioactive wastes resulting from accidental leaks, or failure of the waste sample ani floor drain sample tanks, will not reach the river. !
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- 2. Cooperate with the U. S. Fish and Wildlife Service, the Federal Vater Pollution Control Administration, the Nebraska
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l Department of Health, the Missouri Department of Conservation, and other agencies of the States of Nebraska and Missouri ,
concerned with the development of plans for radiological !
surveys.
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- 3. Conduct, or arrange for the conduct of, pre-operational radiological surveys of the environment including studies of the effects of radionuclides on selected organisms indigenous to the project area, particularly those that store radioactive isotopes. These studies should be conducted by scientists knowledgeable in the fish and wild-life field, and include, but not be limited to the following procedures:
Collect ani analyze samples for contained radioactivity as follows:
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- 1. Samples of water and sediments, fish, aquatic plants, mollusks, and crustaceans should be collected within 500 feet of the proposed reactor effluent outfall, as well as'frcm downstream stations.
- 2. Samples of biological material should be analyzed for both beta and gamma radioactivity. Water and sediment samples need be measured 'only for gamma radioactivity.
- h. Prepare a report of the pre-operational radiological survey and provide five copies to the Secretary of the Interior and five copies each to the' States of Nebraska and Missouri for evaluation prior to project operation.
- 5. Conduct post-operational radiological surveys, similar to thogs in recommendation 3 above, and prepare and submit reports every six months, during reactor operation or until it has been conclusively demonstrated that no significant adverse conditions exist. Submit five copies of the results of these reports to the Secretary of the Interior and to the appropriate State and Federal Agencies for evaluation.
- 6. Make modifications in project structure design and operations to reduce the discharge of radioactive wastes to acceptable levels if it is determined in the pre-operational or post-operational surveys that the release of radioactive effluent permitted under Title 10, Part 20, Ocde of Federal Regulations ,
will result in hamful concentrations of radioactivity in
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fish and wildlife. ,'
We understand that it is the Atomic Energy C W ssion's opinion that its regulatory autho:ity over nuclear power plants involves only those hazards associated with radioactive materials. However, we recomend and urge that, before the permit is issued, the danger to fish and wildlife resources from thermal po21ution and other potential hazards which may result from plant construction and cperation be called to the attention of the applicant. We recommend that the applicant be requested to discuss this matter with appropriate Federal and State conservation officials and to develop measures to minimize these hazards.
We are especially concerned with the possibility of damages to aquatic life from heated water discharged into the river from the proposed nuclear plant. Studies of the heat relationships involved in the con-denser cooling water systems made by the applicant indicate that the
, heated water discharged will be about 180 F. above the temperature of the river water at the intake. The site plan presented in the applica-tion delineates space for future expansion of facilities. Such an expansion would increase the discharge of heated water over the 1,000 to 1,h50 cfs resulting from the initial installation, h
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I During the navigation season, river flows are regulated to about -
31,000 cfs. Condanser cooling requirements win be generally about
. 1,h00 cfs at this time. This large amount of heated water win flow as a separate current for an undetermined width, depth, and dirtance downstream because its density will differ from that of the normal river water. It ' win eventually lose its identity through mixing and natural heat losses.
During the winter months, however, river flows may vary from 6,000 cfs i to as low as 3,000 cfs. With the colder water at this season, condenser i cooling will require only about 1,000 cfs or one-sixth to one-third of )
the river flow. If it were possible to have complete mixing of the j effluent with the river flow at the point of discharge, it is estimated 1 that the entire flow would be warmed from:3 to 60 F. above the normal i temperatures.
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. Proposed water quality standards limit thermal increases of the river '
flow to 50 F. Therefore, under extreme low-flow conditions, the opera-tion of the Cooper Nuclear Station could result in exceeding these proposed limits. Furthemore, these limits are proposed to cover the cumulative effects of all sources of thermal pollution enterirg the !
Missouri River. Future expansion of the Cooper Nuclear Station, together ]
i with the possibility.cf additional nuclear or fossil fuel plants on the river, win increase the severity of the thermal ponution.
Increased water temperature may not only be detrimental to fish and wildlife directly, but also may affect those resources indirectly l through changes modifying the environment. Higher water temperatures i Mminish the solubility of oxygen and decrease the availability of this I essential gas. The elevated temperatures increase the metabolic rate, respiration, and oxygen demand of fish and other aquatic life; hence 4 the demand for oxygen is increased under conditions where the supply is decreased.
Warmer water in the fan and winter may result'in attracting and holdir4 a substantial number of waterfowl in the project vicinity, which other-wise would have migrated southward. If waterfowl are held by this unnatural situation, their lives may be endangered when the weather l becomes severe and natural food supplies become depleted or unavailable. l l
The 3h5-KV transmission line to be constructed across the Missouri l River into Missouri is likely to be a hazard to migrating birds using the river. Especiany during periods of poor visibility, birds have been known to be killed by coniding with the transmission lines under i i
I conditions similar to those which will be established at the Cooper Nuclear Station. This problem may be intensified by concentrations of waterfowl in open water areas created by heated water discharged during I winter months.
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Because of the intricate ecological relationships depending on water temperatures and related physicochemical characteristics of the aquatic environment, detailed ecological surveys should be conducted by the i applicant before and af,ter plant, construction and operation. These surveys should be designed to measure the effect of the plant operation on fish and wildlife resources of the river, and should be planned in cooperation with the U. S. Fish and Wildlife Service, the Federal Water Pollution Control Administration, the Nebraska Game and Parks Commission, the Nebraska Department of Health, the Missouri Department of Conservation, the Missouri Water Pollution Board, and any other State or Federal agency concerned with such resourees. If it is determined by pre-operational surveys that heated water discharges would harm fish and wildlife resources significanthr, plans should be provided for facilities to reduce the tem-perature of thp effluent to acceptable levels. Post-operational surveys should be made]to evaluate the predictions based on pre-operational surveys, and to guard against unforeseen damcge.
In view of the Administration's policy to maintain, protect, and improve the quality of our environment, and most particularly the water and air media, we request that the Atomic Energy Commission urge the Consumers Public Power District to: i
- 1. Cooperate with the appropriate State and Federal agencies in developing plans for ecological studies; begin these studies at least two years before reactor start-up and continue them on a regular basis until it has been conclusively determined that no adverse effects on fish and wildlife resources will result from plant operation.
- 2. Meet with the abovc-mentioned Federal and State agencies at ,
frequent intervals to discuss new plans and to evaluate the !
result of ecological studies.
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- 3. Make such modifications in project structure design and operation, i including but not limited to facilities for cooling discharge I water, as may be determined necessary as a result of pre- j operational and post-operational surveys. i There is a good , potential for development of public-use facilities directly downstream from the plant. The Nebraska Game and Parks Commission and the {
Missouri Department of Conservation have indicated an interest in discussing ;
the possibL,, J developing public fishing and boat launching facilities l in this area in cooperation with the Consumers Public Power District. I Therefore, we request that the Atomic Energy Co=ission urge the Consumers Public Power District to consider such public-use facilities as may be
> jointly agreed upon by the two ageneks.
The opportunity comment on this project is appreciated.
( Sincerely yours, f
1 S.aa 3 Commissioner
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