ML19340B026
| ML19340B026 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 11/16/1959 |
| From: | Kirk R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML19340B024 | List: |
| References | |
| NUDOCS 8010170732 | |
| Download: ML19340B026 (26) | |
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[" # 4, UNITED STATES
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ATOMIC ENERGY. COMMISSION A
WASHINGTON 25, D. C.
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' inn e COMM0 WEALTH EDISON COMPANY
- DOCKHf NO. 50-10 h1EE,EEE License No. DPR-2, as emended 1.
This license applies to the dual-cycle, boiling water type res.ctor i:
designated by Comonwealth Edison Cor:pany (hereinafter referred to L
as "Cornonwealth Edison") as the "Dresden Nuclear Power Station" (hereinafter nferred to as "the facility") which is owned by Comonwealth Edison and located in Grundy County, Illinois, and described in Comonwealth Eiison's application attested March 31, 1955 and amendments to the application attested June 24, 1955, February 1, 1956, March 9, 1956, March 15, 1956, June 6, 1957, June'12, 1957, July 26, 1957, September 3,1957, November 5,1957, December 17, 1957, May 26, 1958, June 5, 1958, August 25, 1958,
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December 26, 1958, December 30, 1958, Jr.auary 6,1959, February 6, 15,)1959, (hereinafter collectively 1959, April 3, 1959, and May referred to as "the application" and for which Construction Permit No. CPPR-2 was issued by the Atomic Energy Comission (hereinafter refernd to as "the Comission") on May 4,1956 and amended on March 31, 1958.
2.
Subject to the conditions and requirements incorporated herein, and subject to the order of the Comission in reference to operations of the nuclear facility at the 315 mesawatt (thermal) power leve' the Ccumnission hereby licenses Comonwealth Edison:
i Pu suant to Section 104(b) of the Atomic Eau y Act cf 1954, as a.
amende, (hereinafter referred to as "the Act") and Title 10, CFR, Chapter 1, Part 50, " Licensing of Production and Utilization Facilities", to possess and operate the facility as a utilization facility; b.
Pursuant to the Act and Title 10, CFR, Chapter 1, Part 'l0, "Special Nuclear Material", to receive, possess and use 2280 kilograms of contained uranium 235 as fuel for operation of the facility; cnd c.
Pursuant to the Act and Title 10, CFR, Chapter 1, Part 30, "Licens-
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in g of Byproduct Material", to possess, but not to separate, such j
byproduct material as may be produced by operation of the facility, i
3 This license shall be deemed to contain and be subject to the conditions specified in Section 50.54 of Part 50 and secticn 70 32 of Part 70; is subject to all applicable previsions of the Act and rales, regulstions and orders of the Comission now cr hereafter in effect; a.nd is subject to any additional conditions specified er incorporated belev:
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Opnrating Requirementa (1) Comonwealth Edison shstil not operate the facility at a steady state power level in excess of 630,000 kilowatts (therr.al), and shall not operate the facility in excess of 315 meEexs.tts (themal) without further 0-der of the Comission.
(2) Subject to the provisions of this paragraph 3., Comonwealth
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Edison shall operate the facility only in accords ce with the design and pe:forn.sce specifications and operating limits r.nd procedures described in the application sad Appen m "A" to this license.
(3)
In any es.se where the procedures or specificatione de-scribed in the application era rot consistent with the requirements of this paragrsph 3. and Appendix "A" to this license, the requirements cratained in this pcrr.-
graph 3, and Appendix "A" shal govem.
(4) Comonwealth Edison shr11 not change or nodify the design or performance specifications or operating limite or pro-cedures described in Appeniix "A" to this license until after a description and hazards evalualion report of the Proposed change has been filed with the Comission by Crwnmweslth Edison end the Comission ehr11 have cuthor-ized such change in writing.
(5) Except with respect to the specifications, limits and procedures containei ir. Appendix "A", Co:::nonwee.lth Edison may change or modify the design or perfomance specifi-cations or operating limits or procedures described in the application only in accordsnee with the following procedures:
Comonwsalth Edison shall provide the Comiesion with a description and hazarde evslue. tion report of the proposed change.
If, within fifteen days efter the is,te cf ac-knowled ment by the Division of Licensing and Eegulation 6
of meipt of such report, the Comission does not issue any' notice to Ccamonwealth Edison to the contry, Comon-vec 1 Edison may mske evch charge without further tpprev61.
If, within fifteen days afts the ds.te of s,cknowledgment by the Division oi Licensing End Eeg21Stion of racsipt of such report, the C:.snissian notifies Comonwnith Edison that the bats de involved may be greater than or me*,eris11y different n'om those e'dvzed in the Eazarde Sa==sry Eeport, or that the proposed chsnge involves a materiel alteration of the facility, the ch oge shd1 not be made until siter such chege hs.s been authorized in writing by the Co:r.is-sion. If e, liceres sme**-t is necessary to E.uthorize the the repo-t submitted by Corr.onvetith Edison
-j proposed chsnEtt shC.1 be desned to constitute cn applici. tion fer E. license j
amniIn b.
As used in this paragraph 3., a prcposed chage j
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shall be deemed to involve hazarls which may be " greater than, or different from, those analyzed in the Hazards Sumary Report" if (1) the probability of any type of i
accident analyzed in the Hazards Sumav Report might be increased, or (2) the possible consequences of any type of accident analyzed in the Hazards Suey Peport might be increased, or (3) such change night create a
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credible probability of an accident of a type different from, and the possible consequences of which would not be of a lesser mgnitude than each of, the accidents analyzed in the Hazards Suma-y Report. The "Eazards Sumr7 Report" as used in this pars 6raph 3. is defined as the " Enclosure Section" attested June 12, 1957, the
" Pre 14*ing Hazards Sumey Report" attested September 3, 1957 and amendments 1, 2, 3 and 4 thereto respectively attested May 26, 1958, August 25, 1958, December 30, 1958, and February 6, 1959 and the " Operating Procedures and Emergency Plass" attested June 5, 1958 submitted by Comenwes.lth Edison.
b.
Records In addition to those otherwise required under this license sad applicable regulations, Camonwealth Edison shall keep the following records:
(1) Reactor operating records, including power levels and periods of operation at each power level.
(2) Records shming the radioactivity released or discharged into the tir or water beyond the effective control of Camonweelth Edison as messured at the point of such release or discharge.
(3) Records of ecargency shutioins, incluMng reasons therafor.
(4) Becords containing a description of each change made
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pursuant to pe-w h 3.a.(5) hereof.
c.
Reports (1) Co::conwee.lth Eiieon shs11 ma.ke en imsiiate report in writing to the Co:rireion of any si 11ficut indies. tion S
or occurrence of an ur. safe condition rele. ting to the operation of the facility.
(2) At least seven dQs prior to ec:cnencing each cf the initial operating phases listed below, Czaonwealth Edison shall file with the Cctrission a report de-scribing the testing progrs:: intended to be conducted during ths.t rerpective phaea tad the genersl testing I:ethods to be usad. Within thiny days of co pletion j
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of each operating phase, Comonwealth Edison 'shall s6bmit a' report of the results of'the tests and operation conducted' pertinent to safety, including
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a descri' tion 'of changes made in the facility design, performance characteristics and operati:s; procedures.
The operating phases referred to above are (a) initial loading and criticality, (b) atmospheric pressure tests, (c) low' power (up to 60 MW thermal power) tests, and (d) rates power tests.
(3) Comonwealth Edison shall submit to the Comission a quarterly report for each quarter during the first year after completion of the' rated power tests. Such quarterly reports shall be filed within 30 days after the end of the quarter covered by the report. Thereafter Comon-wealth Edison shall file'an annual' report. The first such annual report shall be filed'within thirteen months after~the filing of the fourth quarterly report referred to above. Each report filed under this paragraph shall include a description of' operating experience pertinent to safety and changes in facility design, performance
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characteristics and operating procedures during the reporting period.
4.
Pursuant to'Section 50.60 of the regulations'in Title 10, Chapter 1, CFR,'Part' 50, the Comission has' allocated to Comonwealth Edison for Use'in'the operation of the facility 9388 kilograms of uranium 235 c~ontained in uranium enriched to'approximately 1 5% and 1 7%
in the' isotope uranium 235 ' Estimated schedules of special nuclear matsri'al transfers to Commonwealth Edison'and returns to the Com-mission ~are~ contained in Appendix "B"' which is atcached hereto.
Shipments by(the'Comi~ssion to Comonwealth Edison in accordance2) in Appendix will be condicioned upon Co:: mon-with colum wealth'Edis6n's return to the Comission of material substantially
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in accordance with column (3) of Appendix "B".
This license is effective as of the date of' issuance and shall expire 'at the date'to be detemined after c6nsideration of the operations at the 315 megawatt (thermal) power level.
6.
"Inis license is subject to the limitations set forth in the ~
" Intermediate Decision and Order for' Limited Power Operatior;s" and the " Order Extending Period of Time for Limited Power Operations"' issued by the Presiding Officer on September 26, 1959 and November 5,1959, respectively, which provided for operations to the extent of, but not in excess of, a power level of one megawatt (themal) for a period of time which will expire on December lo, 1959 1
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In asco'rdafice with the Supplemental Interb-iate Decision issued on Novsmber"12, ~1959,^In the Mattei- 'of'Commavealth Edison Company, vben Commonwealth Edison has attained 6r'r~eached a power level of
-1 315 megawatts or 50% of rated ~ power, Con:menwealth Edison shall file with 'the Commission a report of such"315-megawatt power level, ~in-bluding'among' other pertinent ~ aspects a description of all existini; and prior operating conditions and characteristics of the facility.
FOR THE ATOMIC ENEIGY COMMISSION R. L. Kirk Acting Director Division of licensing and Regulation Attachikots:
Apbend.x "A" Appendx "B" Date of Issuance: November 16, 1959 l
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m Appendix "A" A.
INIRCDUCTION The followin6 cre the principal design and performnce specifications and operating limits and procedures of the Dresden Nuclear Power Station A-pertaining to safety.
Sections B and C set forth the design and performance specifications and operatin6 limits and principles.
Sections D, E, and F specify the basis for the initial loading and critical testing operations, and the limitations to be observed during start-up, power operation, and refuelin5 and maintenance operations. In these sections, as well as in Section B, where maximum or minimm limits are not given specifically, the values given are " design" values which are subject to normal manufacturing and other tolerances.
B.
DESIGN FEAIURES 1.
Reactor Vessel The reactor vessel is a vertical cylindrical pressure vessel, with dished top and bottom heads, made of low alloy steel and clad inside with stainless steel. The vessel was designed, built, and tested in accordance with the ASIG Boiler and Pressure Vessel Code,Section I.
Design parameters for the vessel include:
Inside height, incMing heads 40ft9-5/8in Inside diameter 12 ft 2 in Desi n pressure 1250 psig 5
Design temperature 650 F 2.
Ndelear Core Wrimm core diameter (circum-scribedcircle) 129 in Mnvim m active cold fuel lengths 107-1/8-in Mnvimn number of fuel assemblies 488 3
Pael Each fuel assembly consists of 36 vertical fuel rods (except fer occasir..a1 special assemblies such as iastrument bearing asse blies which
.11 have 34 or 35 fuel rods plus the thimbles of approximately the same di=ension of the clad fuel rods), each of which is made up of solid cylindrical pellets of uranium dioxide, enriched in uranium-235 to a maxi =um of 15%, and clad in Zircalcy-b Each fuel rod is co= posed of four separately clad seg:ents. Pertinent fuel design para eters in the cold condition are:
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Nel pellet diam 6ter O.493 in Fuel pellet length Regular fuel pellets 0.625 in Segnent end pellets 0 500 in
.Outside diameter of cladding 0 5Ci in Cladding vall thickness 0.030 in z
Fuel pellet density averaged
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over a fuel segment Minimum 94% of theoretical Cross-sectional center-to-center 1
distance between fuel rods 0 710 in
'. Control Rods and Drives The control rods, 80 in nuroer, are made of bcron (initially about
- 25) stainless stehl, have a 6 5-inch cruedform cross section, and travel between fuel channels. The center-to-center distance between control rods is 9 96 inches, and all 30 control rods are located within an 8 foot 3 inch diameter cylindrical space in the central region of the core.
The drive me+tnism for both normal operation and scrcm is an all-
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hydraulic system. Two independent nourecs of hydmulic pressure are available for serreing the control rods. These are:
a.
The accumniator pressure, which vill be available and effective under all conditions of reactor operation except " Shutdown" (at which time interlocks prevent any rod withdru 31), and b.
The reactor pressure, if this is greater than about 700 psig and accumulator pressure falls below reactor pressure.
The drives are mounted on the bottom of the reactor vessel, and withdraw the control rods below the core. Upward movement of the control rods, into the core, decreases reactivity. The inserted position of the control rods is determined by a locking dcvice which provides 12 discrete equidistant positions for all rods except those in the outer rin5 The last withdrawal step for these is slightly shorter than the others. Only one rod at a time can be withdrawn from the core. Rods may be inserted into the core singly or all may be scrnm~d together.
Active length of control blades 8 ft 6 in Velocity for non:a1 insertion or withdraval 6in/see Mwh-nm time from receipt of scra=
signal to:
10% control rod travel 0.6 sec 90% control rod travel 2 5 sec Maxi = m number of control rods per accu =ulator 3 non-adjacent rods I
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Frequent and thorough periodic checks vill be made to assure
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the proper functionin6 of the control rod drive system.
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5 Licuid Poison System System actuation-Manual W riam time after actuation until poison begins to enter core 20 sec
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wei ht of boron in system Fi nimm 6
(present as sodium pentabortte in solution) h00 lbs.
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Mini = = poi ~on worth 0.20 k s
3 6.
Steam Supply System Besides the reactor itself, the steam supply system comprises a steam separating drum, four seconircy ster.'._. Generators and recirculating ptu:.ps, an emer6cncy comensar, ualoading heat exchangers, and the nccessary-piping urd accessories for these corpenents. 'Ihe plant may on occasico operate with one or two of the secondary steam generatet loops bypassed, as long as the
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operation is stable and meets any other specifications imposed.
The e=ergency condenser consists of tve separate tubo httndles of equal caps. city in a common shell. Each tico band'.e crn be s+md automatically by the appropriate rencter cr.fety system controls (discussed in item B.9 below), and may be also started or stopped manually by remotely controlled val-es. ?crtinent limits placed on these steam supply system co=ponents include the following:
Design pressure of steam drum, secondary steam generators, and primary side of unloading and emergency heat ex+nngers 1250 psig Design pressure of primary system piping 1150 psis-Mini
- m capacity of emergency condenser 6fo of rated reactor therr_.1 power Mini =um cooling water stored in energency condenser 30,000 gal' Mini =um emergency condenser cooling period (after scram) without operator -
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attention 6 hrs D T l0 lD U UYM 1
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7 Main condenser The condenser is capable cf handling the nor=al steam flow from the turbine plus the heater drains or bypassed extraction steam.
As a heat sink for the reactor, the condenser vill handle a flow of 1,900,000 lbs/hr of bypassed primary turbine steam.
The condenser can handle this steam flow without desuperheating spray water, in which case the steam temperature entering the condenser vill reach a maximum of about 300'F.
The followinE condenser vacuum trips vill be operative except during "Startup" or " Shutdown" conditions:
Reactor scram if vacuum falls to or b lov 22 in. Hg Turbine bypass valves close t vacuu=
7 in.Hg falls to er belov 8.
h ate Disposal Systems a.
Solid Wastes Solid vastes containing radioactive materials include filters, defective equipment, and other miscellaneous trash. Such material vill be stored in accordance with AEC regulations (10CE-Part 20) which may involve storage in an underground concrete storage vault at the site. When feasible such caterial may be conpacted before storage. Spent contaminated resins are sluiced to an underground tank for indefinite storage.
Sluice water is decanted after the resins have settled.
b.
Liquid Wastes Equipment is provided to treat radioactive liquid vastes by decay in storage, long-tem underground storage, filtration, neutralization, decineralization, or evaporation. The treat-ment vill most often result in water which can be re-uced in the plant or released to the river. Batch sanpling vill be used to determine whether permissible limits can be cet for any liquid process vastes to be released. I!o disposal of these vastes to ground will be made.
The lar6e quantities of river water-used for equip =ent cooling vill be monitored before return to the river to detect any a 2...
process leakaEe into the cooling flow. This flov ny be used to dilute process liquid vastes to below permissible limits for relense to the river. The release of liquid vastes shall confom to the provisions of 10 CB, Part 20.'
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Airborne Wastes Radioactive airborne vastes are discharSed from a 300-foot-high stack. 'Ibere vill be continuous conitoring of the total stack flow and air-ejector flow. A holdup time is provided in both the gland seal exhaust system and the air-ejector exhaust system. The air-ejector monitor auto-matically icitiates closin5 of the discharge valve on the air-ejector holdup system if the measured rate of discharge of Xe 13d exceeds 2 x lo uc per second, the measurement being ude after two minutes decay (travel time in the system from the reactor core to point of measurement).
The reactor vill be manually shutdown when the stack discharge 5
rate of noble fission Bases exceeds 7 x 10 uc per sec.
d.
General a
Co=monwealth Edison shall not release into air any con-
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centration of radioactive material which will result in exposure to concentrations at ground levels in any unre-stricted area as that term is defined in 10 CFR Part 20 in excess of the limits specified in Appendix B, Table II of 10 CFR Part 20.
For purposes of this limitation concentrations may be averaged over periods not greater than one year.
9 Safety System The reactor safety system vill include monitoring devices external and internal to the reactor. The out-of-core system utilizes two paralleled safety channels, each channel with its separate power supply and sensing elements. Both are of fail-safe desi6n throu6hout (that is, de-energizing vill cause a scram), and both must de-energize to cause a scram. Table I, below, lists the external safety circuit sensors, their mvie or minimum trip settings, the number of sensors of each type in each ehnnnel, the coincidence reading feature, and the automatic functions performed in addition to a scra=.
In addition to the trips listed in Table I, an autc=atic scram is provided in the event of power supply failure, by virtue of the fail-safe desi n used.
6 A manual scram control is available to the operator also, and an additional ::nnual control scrans the reactor and closes the sphere isolation valves.
The in-core monitoring system is to provide in required loca-tions indication of local power, automatic scram at not more than 12$of rated local power, and alam at a selected level.
below the scram setting. The automatic scram may be actuated by coincidence of signals from two or more monitors, provided that the coincidence arrangement does not have the effect of leavinc unaonitored a core regicn exceeding in size the limits specified belov, i
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Whenever the reactor is operating at a high power level (as used in this paragraph 9."high power level" shall mean a therm 1 power level exceeding 315 la or 50% of rated local power), there shall be a sufficient nu=ber of operating in-core local power monitors to meet the following conditions:
There will be at least three monitored horizontal layers a.
reasonably evenly spaced in the region of the active core bounded by planes 1 ft below its top and 1 ft above its bottom.
b.
Within the central 8.5 foot diameter vertical cyliMrical core volume, no two adjacent horizontal layers may be without an opemting local power monitor in any vertical cyliMrical core volume that exceeds 4 ft. in diameter.
c.
The in-core local power monitors will be so located that vben the neutron flux within the core is purposely dis-torted by withdrawal of adjacent control rods from any
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region of the core, this distortion shall be detected by at least two operating in-core local power monitors.
Commonwealth Edison shall conduct experi=ents to demon-strate compliance with this requirement.
d.
There will be at least 32 operating in-core local power monitors present in the core.
An operating in-core local power monitor is defined as one which has a response time of less than one second, indicates approximately linear response to changes in local power and does not display erratic changes in calibration. Periodic tests vill be conducted to demonstrate the operating condition of the in-core local power monitors.
For operations at a power level less than high power level, as defined above, the in-core local power monitoring system is not required pro-vided that at least five of the six external power range neutron flux monitors are in operation and are so connected that indicatien cf reactor thermal power exceeding 62 5% of the authcrized power level by any cna =cuitor will scram the reactor.
Rated local power is defined as the condition corresponding to a peak heat flux of 280,000 Btu /(hr)(sq.ft.) in the vicinity of the point monitored.
A four-pasition safety system selector switch is provided to bypass those semm trips which are not required or desirable under some conditions. The bypasses for the external systen are indicated in the "F.e arbs" colunn of Table I.
The in-core local power icnitoting system ray be bypassed in the " shutdown" position t.nd the " refuel" position. The four positions for the selector cwitch are:
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" Start" position, to allow startup before full condenser vacuum is established; b.
"Run" position, for normal plant operation with period trip bypassed; c.
" Refuel" position, to allow some control rods to be withdrawn for safety during refueling if the high neutron flux sensors have been ret to trip at a lov value; d.
" Shutdown" position, to allow ventilation of the reactor enclosure while testing or maintaining the safety system with the reactor shutdown. Control rods cannot be with-drawn in this position.
10.
Pnaintion-Type Reactor Instrumentation In addition to the external and' internal core neutron flux channels
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and the period channels that are a part of the reactor safety system, there are provided two indicatin6 startup channels and a battery-operated neutron flux channel.
11.
Reactor Enclosure The enclosure housing the reactor and the steam supply system is a spherical steel shell, 190 feet in diameter, with the equator approximately 56 feet above ground level. From the enclosure, the mini e off-site distance is one-half mile to:
a.
Skinner Isinna in the Fankakee River; b.
The navigation channel in the Illinois River; and c.
The land bonnanries of the site.
The enclosure was desi5ned, built, and tested in accordance with the-ASME Boiler and Pressure Vessel Code. Pertinent design parameters for this vessel are:
Desi n Pressure 29 5 psig 5
Desi n Te=perature 5
(coincident with design pressure) 305 F
!@vimm Wind Velocity 110 mph Horizontal Acceleration 3 3% gravity leximum LeakaEe Rate at 37 psig o.5fc/ day All normally open lines penetrating the enclosure shell through which leakage could credibly occur in the event of the ":::animum credible accident" are provided with check valvec or isolation valves which close without operator attention. Table I indicates
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the seras signals which initiate closure of the automatic isolation
-valves. The isolation valves on the primary steam lines, which are not closed from any scram signal, vill close automatically from low reactor pressure -- a condition that would be encountered in any system rupture approaching the severity of the postulated " maximum credible accident." All the primary isolation valves are backed up by additional valves which can be operated from positions that are tecable after the accident. Normally closed lines penetrating the enclosure are protected against being opened during operation, or in potentially hazardous non-operating situations, by interlocks and/oroperatingrules.
5 The enclosure is provided with a post-incident cooling system, for use in the event of a serious primary system rupture, to aih in the reduction of enclosure pressure and leakage. Thissgstemis desi ned for_a heat removal capacity of at least 30 x lo stu/hr 6
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at an enclosure internal temperature of 256 F.
Operation of this cooling system vill be automatically initiated by
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the scram signal from high sphere pressure after a time delay not to exceed ten minutes. The react,or operator may also manually initiate operation of this system or override the automatic initiation si nal. The system vill be maintained in operable 6
condition at all times the primary system is pressurized. Periodic tests will be conducted to demonstrate the proper functioning of the post-incident cooling system.
About one-year after initial power operation, a leakage test of the enclosure vill be coaducted at a low pressure (less than 10 psi) to detemine the reliability of penetrations.
(It is expected that this test will probably be made durin6 the time the plant is down for the first major inspection of the turbine which is also expected to be done after about a year's operation.)
After a study of the results of the above leakage testing and dis-cussion of these results with the AEC, the future program of testing vill be agreed to.
Control Room The control room is shielded from all power plant equipment and frca the reactor enclosure so as to'be tenable even in the event of the "say4-- credible accident." The minimum thickness of concrete shielding betvec che control room and enclosure for direct-line radiation is the quivalent of five feet.
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C.
OPERATIIC PRINCIPIES The basic operating principles that vill be adhered to in the operation of the Dresden plant are as follows:
a.
Before being placed into regular service, the plant vill be proved out in a comprehensive controlled test and initial operation program, as indicated in Sections D and E.1, below.
b.
Operation and control of most of the power plant equipment vill be centralized in the control room.
c.
The control room vill be mnnned at all times by at least two operators, one of which shall be licensed.
d.
While most operating and control functions are initiated in the control room, operators may perfom some functions at remote operating panels and valve racks -- at the direction of the control room staff or with their prior knowledge.
e.
Startup, normal shutdown, and all other repetitive operations vill be performed in accordance with specific check lists.
f.
Paintenanca of much of the equipment outside the reactor shielding may be undertaken by contact rethods and without overall plant shutdevn. Plant shutdown and semi-re=ote methods vill be emphyod as necessary.
g.
All tests and routine maintenance of protective devices and power plant equipment will be done in accordance with pre-scribed schedules.
h.
Radiation monitoring by fixed or portable instrumentation vill be provided for entry to all radiation zones.
1.
All personnel leaving a contaminated radiation zone, and equipment being removed from such zones, vill be appro-priately surveyed to assure control of contamination.
j.
Irradiated fuel is to be moved fro = the reactor to storage, under water, by se=i-remote methods.
h.
Enclosure isolation provisions (i.e., air locks and equipment batches closed, and instrumentation operating to close the automatic isolation valves if it should become necessary) vill be in effect during all periods of reactor operation, including startup and shutdown operation, and during any operation involving insertion or renoval of fuel assemblies in the core or withdrawal of control rods when the reactor head is off.
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Operation of the radioactive vaste-handling system vill be done in such a manner that it vill be unlikely that the disposal of radioactive materials will result in the exposure of any persons on or off the plant to radiation in excess of the per=issible limits.
The plant in co protected at all times by automatic safety m.
devices that no single operator error or reasonably con-ceivable comb'.aation of operator errors could cause a severe accid':nu.
All significant unexpected incidents, unsufe acts, or n.
incidents of excessive exposure to radiation vill be in-vestigLted to effect procedures to prevent recurrence.
In the event of any situation which may cocpromise the o.
safety of continued operation, it vill be the required procedure to shut the plant down as quickly as the situation calls for, and to take other pinnned emergency actions to protect persons and property.
D.
INITIAL ICADING AND CRITICAL TESTING 1.
General On successful completion of all pre-operational checks necessary for safe loading, the reactor vill be loaded to initial critical ani then to the proper lesMng for power operation. Between these two loading steps a number of tests will be run and measurements made to check the calculated nuclear parameters. This phase vill be con-ducted at atmospheric pressure with the reactor vessel lid removed, and at approximately ambient temperature. Items which vill receive particular attention during this phase of operation, or that vill be carried out in a mnner different from normal refueling, are discussed below.
2.
Instrumentation A m1M-um of one neutron-sensitive chamher per grid quadrant vill be installed in the reactor before the initial loading starts. A neutron source of sufficient strength to provide appreciable readings at all times on the neutron-sensitive cha=bers vill be installed, and the chamber-source geonetric relationships vill be maintained as the loading progresses so that the chnrher sensitivity vill always be sufficient to measure the neutron multiplication. A mini =um of four neutron-sensitive chambers vill be tied into the reactor safety system g
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in such a way that trippin6 of any chnnnel vill cause a scram. During this period with the head removed the reactor safety syste= in-core monitors vill not be in place.
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The initial loading and critical testing is to be done with the safety system selector switch in the "rafuel" position (seeitenB.9). This indi utes that the following scram sensors vill be in operation:
a.
High sphere pressure b.
Iov vater level in reactor (This sensor ray be bypassed in the " refuel" position until initial power operation.)
High reactor pressure (Thf s sensor is in service in the c.
" refuel" position because there is no reason to bypass it.
With the reactor head off, of course, it ennnnt be effective.)
Hi h level in scram dump tank d.
6 E1 h neutron flux (For the initial loading operation.
e.
5 the normal out-of-core cha +ers and dual-channel syste are replaced by the special in-core chamber system described above in this ' item 2.
As with the nor=al system, however, with the selector switch in the " refuel" position, the neutron flux trip setting cast be reduced to a mavimm setting of 10-3 rated power.)
f.
Short period.
3 Reactor water level Prior to the initial loading, the reactor vessel vill be filled with water to a minimm of three feet abov3 the active hei ht 5
of tba ire.
As indicated in D. 2, the low water level sensor, ray be bypassed durin6 the initial loading and test period when the water is required pri=arily as a moderator.
.=
4.
" Stuck Rod" Criterion: At every stage diring the initial loam ng, and in the fully loade. confiEuration, a
the control rods must provide a shutdown control cargin of at least 0.01 k with any rod wholly out of the core and completely unavailable.
b.
" Cocked Bod" Criterion: During core alterations after the first fuel cell is loaded, the reactor must be sub-critical by at least 0.01,,k with at least one centrol rod fully withdrawn in th'e' region of the alteratio'n and available for rapid scra= insertion.
1-
w H
_g.
5, Minimum Critical Core The following criteria vill be used in loading to the minim", critical size core:
a.
The selected core area vill be within the control rod pattern.
b.
Except for the finai fuel increment, the size of each fuel addition vill never exceed one-half the estimated amount to reach criticality. This estimte vill be made using neutren multiplication measurements determined between fuel increment additions. The final fuel in-crement will not exceed a reactivity worth of 0.005 k.
3 The shut down mar 6 n criteria given in item D.4, vill i
c.
be applied.
The following tests vill be conducted, as a minimum, with the minirm critical core:
d.
Measurements of the strength of some of the control rods encompassed by or adjacent to the fuel loading; Temperature coefficient (gk/
change in moderator e.
temperature);
f.
Voidcoefficient(gk/ !o change in voids).
The void coefficient averaged over the interior c a fuel channel vill always be negative when the core 3.s critical.
~
Sufficient measurements vill be made to conclude that the void coefficient, measured es indicated in the sentence above, is neSative in all fual channels.
6.
Twice Mini =um Critical Core On successful completion of item D.5, the core vill be loaded to approximately twice the minimum critical size, but not beyond the control rod outer periphery. The following measurements vill be made, as a minimum, with this core:
a.
Selected control rod calibrations; b.
Rod confi6urations for critical; c.
Temperature coefficient; d.
Void coefficient. Thit rast meet the limitations indicated in item D.5.f.
i
1 i
.c
/1
.g.
7 Core Loaded to Periphery of Control Eod Pattern On successful co=pletion of iten D.6, the core vill beloadedtoencorpassallcontrolrodcells(i.e.,one rev of fuel assemblies outside of control rod periphery).
Criticality checks vill be =ade periodically during this loading. Calculations and extrapolations of teasurements made in accordance with itecs D.5 and D.6 vill be used, prior to the actual fuel loeding, to indicate whether the shutdown cargin and void ccefficient criteria can be cet.
Measurements vill be cade with this core, as a nininum, of the critical rod configurations with non-uniform distribution of withdrawn control rods.
8.
Ce=oleted Ccre Leadira Calculations and test data from item D.7 vill be used te determine the maximum number of fuel assecblies permitted in the area outside the centrol rod periphery in the fully loaded core. Following this deternination, approximately half the number of fuel elements vill be added unifornly around the core. The determination vill then be repeated and followed by another " half loading."
This sequence vill be repeated until the last fuel assembly is inserted. The completed core vill teet, as a mini =un, the shutdown =argin reqt. ire =ents indicated in ites D.h.
'Ibe test progren for the completed core vill include, as a mininum, the following itens:
a.
Critical rod configurations with non-uniform distribution of withdrawn enntrol rods; b.
Te=perature coefficient; c.
Void coefficient.
This =ust meet the 11=itations indicated in ite= D.S.f.
E.
PCWER OPERATICN 1.
Power Test Progre Folleving successful co pletion of the initial loading and critical testing progra=, the power test progran vill be carried out.
This program vill include the folleving:
f
i n
l}
.s1 a.
A series of tests at low power, and lov to rated pressure, with studies cf stability, rcactivity, and power distribution as they are riffected by variations in reactor pressure, te.cperature, voids, control rod positions, and transient poisens; b.
A series of tests at various power levels up to rated power, at lov to rated pressure, with studies similar to those indicated in iten E.1.a; c.
Studies of power transient effects, siculated equipment failures, and process nishaps; d.
Radiation surveys.
This testing progrs= vill be conducted at stepwise in-creas!ng pcver levels. At each power level, experinents vill be conducted to predict the power level or other operating condition of the reactor at which taximum heat flux, reactor instability, or any other reacter design or operational limitation vould be reached.
The reactor power level or other limiting condition vill not be increased in the next step more than half-vay from the test conditions to the predicted li=ita-tion, provided that for reactor power levels in excess of 507,of rated power no step increase in pcVer shall exceed 60 IG.
2.
Safety System Scran Settings Operative scram sensors and their settings with the safety systen selector switch in the " Start" and "Run" positiens are given in iteo B.9 The control rods
'~
eannot be withdrawn in the " Start" position until at least 15 inches of Hg condenser vacut= have been obtained, and the switch to the "Run" position cannot be made without sera = ming until at least 23 inches of Hg condenser vacuu have been obtained. The reactor vill also scram if a reactor pressure of 200 psig is exceeded prior to
)
cbtaining at least 23 inches of Hs condenser vacuun.
3 Detezzination of Maxicu= Reactor rever The " rated" er operational thermal power cf the reactc'r is not, at least during initial power operation, a fixed i
value. Lower than equilibrium fuel cycle power pro-O duction may be necessary initially, due to the possi-
.j bilities of:
i-
_c
/
^
lY G.
Less than fuli load of 488 fuel assemblies in order to a.
meet shutdown margin criteria; and
- b. Higher peaking factors iti.ially than at equilibrium fuel cycle.
The maximum reactor power is, consequently, defined as that thermal power at which the maximum heat flux for any fuel rod This maximum heat flux will never exceed 350,000 isreaghed.
Btu /ft -hr, based on calculations and experimental data.
The rated heat flux and resulting rated reactor power are then set to 80% of their maximum values and, as indicated in items B.9, the high neutron flux scram setting will be no higher than an indicated 120! of the rated reactor power.
However, in no case will the high neutron flux setting be allowed to exceed an indicated reactor thermal power of 782 MH (125% of the planned operational power of the fully loaded core).
Within the limitations on reactor pouer and heat flux set forth in the paragraph above, the reactor will be operated in such fashion as to:
a.
Maintain at all times a burnout heat flux margin of at least 2 at the point closest to burnout in the hottest channel in the core based on a uniform steam quality over the cross-section of the channel; and b.
Be always well within the bounds of stability, as evidenced by the operation itself and any experimental data produced during the " Power Test Program" phase of operation.
4.
Pressure Limits a.
Maximum normal reactor operating pressure 1000 psig b.
Maximum pressure setting for automatic reactor shutdown 1050 psig c.
Maximum pressure setting for opening of electromatic relief valves 1085 psig d.
!!aximum pressure setting for opening of first main safety valve 1205 psig e.- Maximum safety valve pressure setting 1250 psig f.
Combined capacity of safety valves At least 150%
rated primary steam flow 5.
Reactivity Limits a.
The averagc reactivity addition rate from withdrawal of the control rod with the canimum reactivity worth in the most i
l
{
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//
$d-adverse withdrawal pattern will not exceed 0.0029fk/sec.
b.
With the reactor in any condition, the shutdown margin criterion in item D.4.a must be met.
c.
With the reactor in the hot operating condition, operation will not be continued when the reactivity worth of the control rods known to be stuck out of the core, or other-wise unavailable for control, exceeds half the value at which hot shutdown could not be accomplished. (It is expected that this would require shutdown for repair of the inoperable rods if three adjacent, or the equivalent in reactivity worth of nonadjacent, rods are known to be unavailable.)
If, in such case, it is calculated that following shutdown and cooling of the reactor the shutdown margin will not be at 1 cast 0.01 k, the liquid poison will be introduced prior to 4
reaching the " cold" condition where this criterion could not be satisfied.
d.
The void coefficient of reactivity will meet the limitations
~
indicated in item D.5.f.
e.
The moderator temperature coefficient cf reactivity will change from positive to negative at a teriperature below the maximum reactor terperature that canbe attained under atmospheric conditions. licasurcmeats will be made to con-firm this.
6.
Vaste Disposal The dispecal of wastes resulting from power operations is discussed in iten B.8.
Dicposal of all waste off site will be in a manner such that it is unlikely any person will re-ceive radiation exposures in excess of the approximate per-missible Ibnits.
F.
REFUELING AND MAINTENANCE 1.
Operatine Princinles All refueling and maintenance operations will be carried out in accordance with all the applicable operating principles given in Section C.
Items in Section C which are particularly per-tinent to these operations are those lettered c, e,
'f, g, h, i, j, k, and m.
2.
Safety System Scram Sensors Opcrative scram sensors and their setting with the selector switch in the " Refuel" position are given in itcm B.9.
7
- t. "
g 44 Since some maintenance can b'e carried out under any of the possible reactor conditions, the safety sensors in opera-tion will depend upon the particular job to be done.
Even with the reactor in the shutdown condition, however, any maintenance work involving the removal of control rods from the core will be done with the safety system selector switch in the " Refuel" position.
3.
Shutdown Margin At every stage of refueling or maintenance, the minimum shutdown margin will satisfy the " stuck rod" criteria dis-cussed in items E.5.b and E.5.c.
During movement of fuel in the core, or control rod maintenance, the minimum shutdown margin will satisfy the " cocked rod" criterion given in item D.4.b.
4.
Liquid Poison System The liquid poison system will be operctive during refueling and maintenance operations in the teactor nr. well as during normal power operations.
e
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TABLE I SAFETY SYS'ITM Paterial Sen. ors Nurber Coincidence Scram Scmm Other Automatic Tyr.2 in each in each Trip Setting Trip Ibint ")
Functions Eccarks I
Channel Chr'nnel Thrfor=ed FJch Sphere 2
1 out of 2 Pnximum Pressure Setting [0.2 Closes isolation T.o byp.sses Pressure of 2.0 psig psig valves and ventile.-
tion ducts Icv Water 2
1 out of 2 At level which is Setting [1" Clcscs isolction Bypassed in " Shutdown"
,. m level in a minimum of h3" valves and ventilc-position prior to initial "
PO.ctor above the tcp of tion ducts, power o p ration only.
Vessel the active fuel Bypass is to be removed.
Closure of 2
1 out of 2 When oil pressure Setting [5 Closes ventilation Bypassed in "Shutdovn" or Turbine Stop controlling these psig ducts
" Refuel" positions.
j
& Bypa.sc valves dm ps to a
(
Vcives minimum of 50 psig Closure of 2(b) 2 out of 2 Closure of both Setting [5%
Closes ventilation Bypassed in " Shutdown" or Irir.ary Steam (One fzum valves beyond a stmke ducts and starts "Ecruel" positions.
Sphere Isola-each valve) maximum of 25% of c=lergency cooling tion Valves stroke Icv vnter 2
1 out of 2 At level which is Setting [1" Closes ventilation Bypassed in " Shutdown" or Level in a maximum of 5" ducts
" Refuel" rositions.
Irir2ry below the drum Stenn Drum center line Q
Ic. Conden-2(c) 1 At minimum con-Setting [0.25" Closes ventilation Bypssed in "Shutdo a" or i
cer Vacuum denser vacuum of ducts
" Refuel" positions.
By-22" E8 passed in " Start" position if reactor pressure below j
1 At minimum con-Setting [O.25" 200 psig & corZenser 5
denser vacuum of vacuum greater 'than 15"Eg i
=
23" Hg b
"J Eigh Re.ctor '
2 1 out of 2 At maximum reactor Setting [10 Closes ventilation Bypassed in "Shutiovn" pressure of 1050 psig ducts & starts position.
T m e.ure i
i
]
psig cmrgency cooling.
t
TAMR I SAFETY SYSTEM (Continued)
External SensArs Number Coincidence Scram Scran Other Automatic TripIbint(8)
Functions Bemarks I
hTe in.cach
-in cach Trip Setting Channel Channel Ibrformed- _- --
~{:.7
(
Ei.h level 2
1 out of 2 At tank level Setting [1" Closes ventilation Bypassed in " Shutdown position t
in Scram whichisk'h)"
l ducts Ihnual bypass also prevents ~'
Dump Tsnk j
above the base.
contrul red withdrawal. At line of the lower scram point, there is sur-tangent of the ficient free volume rainin6 tank in the scrac dump tank to accommcdate the water from-2 7 scrams.
\\
Bypassed in " Shutdown" pos ~
Nting/3%or Closes ventilation ition. interlocked in "Re-
@ Ecutron 3
1 out of 3 when Icakage Flux
.or 1 out of flux ind'. cates rated power ducts
'udi" position so as to re--
2 if 1 is a rn.h of quire reduction in trip set-bypassed 120% rated power ting to a maximum setting of 10-3 rated power. At any time.
other than in the " shutdown" position only one of the six flux trips provided can be M
bypassed.
{3
@_)
" m r t. Itriod 3
2 out of 3 At minimum period Octting /_ 0 5 Oc ec ventilation Bypassed in " Shutdown" and of 4 seconds seconds du :*.a "Run" positions.
g (It)l5e Mint at which n scrum :.s actually initiated rny be different, fr m the "Jeram trip setting" by the er:ount of the tolerence for instrt: ment iroccuricles. Ec amount of Ulis toicrance is indico.ted by the values given in this colt =n.
M 101 rach valve-position evitch (or which there is one per valve) 1:an tvu co:rtacts in each safety cbnvmel.
Each censor has one contact in cach of the two safety channel.
Nd (c) There are two vacuum sensors.
Each chamber has two contacts in cach of the two safety cbnn els.
p (d) Ecre P.re three period sensoring chambers.
11
[.
~
IM ENDIX "B"
{
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q
- T_o, CCMMCh"a'EALTH CDISCN Co!'PANY 1
PROPOSED LICENSt:
Estimated Schedule of Transfers of Special Nuclear Material from the Commissio'n'to Cee.onwealth Edison and to_ the Commission from Cocmonwealth Edison (1)
(2)
(3)
(4)
(5)
Date of Transfere from Returns by Common-Trancfor AEC to Co= mon-wealth Edison to Net Yearly Cu=ulative (Fiscal Year),. Kilograms U-235 Kilorrams U-235 Kilograms U-235 Kilograms U-235 wealth Edison AEC Distribution Distribution,
1958 243 3 243 3 243 3 1959 974.3 198.0 776.3 1,019.6 1960 376.7 84.0 292.7 1,312 3 1961 376.7 185 0 191 7 1,504.0 1962 376.7 185 0 191 7 1.695 7 1963 376.7 185.0 191 7 1,887 4 1964 376.7 185.o 191 7 2,079 1 1965 426.5 185.o 241.5 2,320.6 1966 386.7 185 0 201.7 2,522 3
~
1967 346.8 346.8 2,869.1 l1968 312.9 312.9 3,182,0 1969 272.1 272.1 3,454.1 1970 228.2 228.2 3,682 3 1971 228.2 228.2 3,910 5 1972 223.2 228.2 4,138.7 1973 228.2 228.2 4,366.9 1974^
228.2 228.2 4,595.1
~
1975 228.2 228.2 4,823.3 1976 228.2 228.2 5.051 5 1977 228.2 228.2 5,279 7 1978 228.2 228.2 5,507 9 1979 228.2 228.2 5,736.1 2980 228.2 223.2 5,964.3 1981 22.6.2 228.2 6,192 5 1982 228.2 228.2 6,420 7 1983 228.2 228.2 6,648.9 1984 228.2 228.2 6,877.1 19S5 228.2 228.2 7,105 3 1936 228.2 228.2 7,333.5 1937 228.2 22S.2 7,561.7 1988 228.2 228.2 7,789 9 1989 228.2 228.2 8,018.1 1990 228.2 228.2 8,246,3 1991 228.2 228.2 8,474.5 1992 228.2 228.2 8,702.7 1993 228.2 22d.2 8,930.9 1994 228.2 228.2 9,159 1 1995 223.2 228.2 9,357 3 10,779 3 1,392.0-9,3S7 3
- The fuel being diccharged durinE fiscal yearn 1967 through IFj5 will have te ct.
depleted to cuch an extent in U-235 that it no longer is specal nuclear raterial.
Therefore, no credit for returns has been included for these years, l
.