ML20037B294
| ML20037B294 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 09/19/1962 |
| From: | Lowenstein R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML20037B292 | List: |
| References | |
| NUDOCS 8009100967 | |
| Download: ML20037B294 (25) | |
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UNITED STATES ATOMIC ENERGY COMMISSION WASHINGTON 25, D.C.
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COMMONWEAL ~ni EDISON COMPANY DOCKET NO. 50-10 AMENDMENT TO FACILITY LICENSE License No. DPR-2 Amendment No. 7 Facility License No. DPR-2, as amended, which authorizes Common-wealth Edison Company to operate the dual cycle, boiling water type nuclear reactor designated as the "Dresden Nuclear Power Station" located in Grundy County, Illinois, is further amended in the following respect:
1.
Item 3.a. (1) of License DPR-2 is amended to read as follows:
"3. a. (1)
Commonwealth Edison Company, Chicago, Illinois, is I-authorized to operate the facility at steady state t
power levels not in excess of 700 megawatts j
(the rmal). "
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This amendment became effective on September 4,1962.
i FOR 711E ATOMIC ENERGY CORfISSION E. L:,.cnS Director Division of Licensing and Regulation Dated at Germantown, Maryland this day of
, 1962.
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l Appendix "A" to DPR-2
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I A.
INTRODUCTION The following are the principal design and performance speci-fications and operating limits and procedures of the Dresden Nuclear Power Station pertaining to safety.
l Sections B and C set forth the design and performance speci-fications and operating limits and principles.
i Sections D and E specify the limitations to be observed during start-up, power operation, and refueling and maintenance l
cperations.
In these sections, as well as in Section B, where l
maximum or minimun limits are not given specifically, the values given are " design" values which are subject to normal manufacturing and cther tolerances.
Sections F and G provide certain additional operating and testing procedures applicable to the ecntrol rod drive mechanisms. 'Section H provider minimum requirements for certain inspections of the control rod drives, poisen blades and core g:id structure.
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B.
DESIGN FEATURES
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1.
Reactor Vessel
- ?ne reactor vessel is a vertical cylindrica'- pressure vessel, with dished tcp and bottom heads, made of low alicy steel and clad inside with stainless steel.
The vessel was cesigned., built, and tested in ac:crdance with the ASME Boile: and Pressure Vessel Code,Section I.
Design parameters for the vessel include:
j Inside height, including heads 40 ft 9-S/B in
- nside diameter 12 ft 2 in i
Design pressure 1250 psig Design temperature 650' F 2
Nuclear Ccre i
l Maximum ccre diameter (ciretm-l scribed circle 129 in F
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Maximum active fuel length - cold 112 in Maximum number of fuel assemblies by types Type I 488 Type II 108 Type PF-1 through PF-12 (one each) 12
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Maximum total number of fuel assemblies 488 The verious fuel assemblies may be located in any position p
of the reactor, provided over-all core symmetry is preserved and provided that fuel assemblies, Type PF-1 through PF-12 are each separated from any other such assembly by at least four Type I or Type II fuel assemblies.
The reactor may be operated at any power up to and including rated power with any number of the various types of fuel assemblies installed provided the maximum number and location are within the limits specified above.
3.
Fuel Each fuel assembly consists of vertically-positioned, rod-type fuel elements.
The physical properties of each assembly are given in Table II.
The number of fuel rods are given for a regular assembly.
In several assemblies, fuel rods have been replaced with instrumentation tubes.
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The minimum fuel pellet density averaged over a fuel segment is 94% of theoretical for all fuel assemblies except PF-7
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PF-8, and PF-9 which is 90% of theoretical, j.
4 Control Rods and Drives I
The control blades, 80 in number, consist of small vertical stainless-steel tubes filled with compacted boron carbide (B C) powder.
The boron carbide powder is separated j
4 Iongitudinally into several independent compartments.
The tube walls are designed to withstand the maximum cal-
't culated internal gas pressure.
The tubes are held in a t
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6.5-inch (over-all width) cruciform array by flat stainless-steel plates.
These plates, extending the full length of the control blades, provide a smooth, flat outside surface.
The control blades travel between fuel channels. The center-to-center distance between blades is 9.96 inches, and all 80 control blades are located within an 8-foot, 3-inch diameter cylindrical space in the central region of the core.
l The drive mechanism for both normal operation and scram is an all-hydraulic system.
Two independent sources of hydraulic pressure are available for scramming the control rods.
These are:
l a.
The accumulator pressure, wh ch will be available and effective under all conditions of reactor operation except " Shutdown" ( at which time inter-locxs prevent any rod withdrawal), and b.
The reactor pressure, if this is greater than about 700 psig and accumulator pressure falls below l
reacto: pressure.
The drives are mounted on the bottom of the reactor vessel, and withdraw the control rods below the core.
Upward l
movement of the control rods, into the core, decreases l
reactivity.
The inserted position of the control rods is determined by a locking device which provides 12 discrete r
approximately equidistant positions for all rods.
Only j'
l one rod at a time can be withdrawn from the core.
Rods
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l may be inserted into the core singly or all may be scrammed i
together.
I Active length of control blades 8 ft 6 in Velocity for normal insertion I
or withdrawal 6 in/sec l
Maximum time from receipt of l
scram signal to:
10% control rod travel 0.6 sec 90% control rod travel 2.5 sec l
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Maximum number of control rods per accumulator 3 non-adjacent rods Frequent and thorough periodic checks will be made to assure the proper functioning of the control rod drive system.
5.
Liquid Poison System
' System actuation Manual er Maximum time after actuation until poison begins to enter core 20 sec.
q Minimum weight of boron in system g
(present as sodium pentaborate k
in solution) 400 lbs.
Minimum poison worth for operations with reactor vessel head on 0.154k Minimum poison worth for operations with reactor vessel head off 0.04Ak 6.
Steam Supply System Besides the reactor itself, the stean supply system comprises a steam separating drum, four secondary steam generators and recirculating pumps, an emergency condenser, unloading heat exchangers,and the necessary piping and accessories for these components.
The plant may on occasion operate with one or two of the secondary steam generator loops bypassed, as long as the operation is stable and meets any other specifications imposed.
The emergency condenser consists of two separate tube bundles of equal capacity in a common shell.
Each tube bundle can be j
started automatically by the appropriate reactor safety system i
controls (discussed in item B.9 below), and may be also started or stopped manually by remotely controlled valves, Pertinent limits placed on these steam supply system components include the following:
Design pressure of steam drum, secondary steam generators, and primary side of unloading and emergency heat exchangers 1250 psig Design pressure of primary system piping 1150 psig Minimum capacity of emergency condenser 37.8 Mw(t)
R Minimum cooling water stored in emergency condenser 30,000 gal.
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t Minimum emergency condenser cooling period (after scram) without operator attention 8 hrs.
The reactor recirculating pumps will be tripped whenever the level in the steam drum falls to approximately 21 inches below the drum R
center line.
The pump trip is not a safety circuit function, but 3,;
rather is provided for protection of the recirculating pumps against loss of inlet suction.
7.
Main Condenser The condenser is capable of handling the normal steam flow from the.
turbine plus the heater drains or bypassed extraction steam.
As a heat sink for the reactor, the condenser will 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 will reach a maximum of about 300*F.
l In addition to the condenser vacuum scram and isolation tripa indicate later in item 9 " Safety System", a mechanical trip is provided to close the turbine bypass valves, as well as all other hydraulically-opened turbine valves, should the condenser vacuum l
fall to or below 7 inches of Hg.
This vacuum trip is not a part of j
the reactor safety system, but becomes operative whenever the l
condenser vacuum has been increased to above the trip set point.
I 8.
Waste Disposal Systems a.
Solid Wastes Solid wastes containing radioactive materials include filters, defective equipment, and other niscellaneous trash.
Such l
material will be stored in accordance with AEC regulations l
(10 CFR-Part 20) which may involve storage in an underground l
concrete storage vault at the site.
Nhen feasible such material p'
may be compacted before storage, Spent contaminated resins are sluiced to an underground tank for indefinite storage.
Sluice water is decanted after the resins have settled, b.
Liouid Wastes Equipment is provided to treat radioactive liquid wastes by decay in storage, long-term underground storage, filtration, neutralization, demineralization, or evaporation. The treat-ment will most often result in water which can be re-used in the plant or released to the river.
Batch sampling will be.
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used to determine whether permissible limits can be met for any liqaid process wastes to be released. No disposal of these wastes to grcund will be made.
The large quantities of river water used for ecuipment cooling will be monitored before return to the river to detect any process leakage into the cooling flow. This flow may be used to dilute process liglid wastes to below permissible li=its for release to the river. The release of liqaid wastes shall conform to the provisions of 10 CFR, Par.t 20.
c.
Airborne Wastes
-Radioactive airbone wastes are discharged from a 300-foot-high stack. There will be continacts monitoring of the total stack flow and air-ejector flow. A holdap time is provided in both the gland seal exhaust systen and the air-ejector exhaust system.
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The air-cjector m:nitor autcmatically initiates closing of the dischargevalveentheagajectorholdapgystemifthemeasured rate of discharge of Ic exceeds 2 x 100 uc per second, the measurement being made after two minutes decay (travel time h1 the system from the reactor core to point of measurement). The reactor will be ranually shatdown why the stack discharge rate of noble fission gases exceeds'T x.lC-uc per sec.
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d.
General Commonwealth Edisen shall net release into air any concentration of radioactive ms.terial sinich will result in exposure to con-centrations at grc=.d levels in any unrestricted area as that term is defined in 10 CFR Part 20 in excess of the limits speci-fied -in Appendix 3, Table II of 10 CFR Pcrt 20.
For purposes of' this limitation concentratiens may be aversged over periods not greater than one year.
9.
Safety System The reactor safety system vill includs monitoring devices external and internr1 to the ructer,, Th5 cct-of-core system utilices two paralleled safety thznnele, each che el with its separate power n
supply and sensing elements. Both arc of f ail-cafe design through-cut (ths.t 1s de-2nergicing will ca2se a stram), and both I:Ist de-3 energize to cause a scra.. Table I, below, lists the external safety circait ser.:crs, thei.r r/ixirum cr minin= trip settings, the number of sensors of en:;h type in each channel, ths coincidence reading featurs, snd the anterrctic functiens perfc med in addition to a scram. In addition to the t-ips listed in Table I, an auto-matic scran is prcvided in the event of power supply f ailure, by virtue of the fail-safe derign used. A manual scram control 1-
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available to the cperater also, and an additional nanual conel
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scrana the reactcr a.d cic,ses the sphere isclation valves.
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a.
The in-core monitoring systen is to provide in required locations indication of local power, autem 1 tic scram at not more than 1255 of rated local power, and alarm at a selected level below the scram se'ttilng. The automatic scram may be actuated by coincidence of signals from two er more monitors, provided that the ccincidence arrangement does not have the effect of leaving unmonitored a core
< region exceeding in size the limits specified below.
Whenever the reacter is operating at a high power level (as used in this paragraph 9. "high power level" chail mean a thermal power level exceeding 350 F4 or 50% of rated local power)3 there shall be a sufficient number of operating in-core local power monitors to P
meet the following conditions:
- i. There will be at least three monitored horizontal layers reasonably evenly spaced in the region of the active core botaded by planes 1 ft below its top and i ft above.its bottcm.
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- 11. Within the central 8.5 foot diameter vertical cylindrical core volume, no two adjacent hericontal icyers may be withent an opercting local power monitor in any vertical cylindrical core volum3 that exceeds h ft. in diameter.
iii. The in-core local power moniters will be so located that when the r.eut_mn flux vithin the cere is perpmly dis-torted by withdrawal of adjacent centrol rods from any region of the core, this dis ortion shall be detected by
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at_least two operating in-ccre local power monitors.
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Commonwealth Edison shall conduct experiments to demon-j stre.te compliance with this requirement.
iv, fhere will be at least 32 operating in-core local power monitors present in the core.
An operating in-ecre local power monitor is defined as one which has a response time of less than one second, indicates approx-imately linear response to changes in local pcwer and does not display erratic changes in calibration. Periodic tests will be condacted. tc demonstrate the operating condition of the in-ccre local power noniters.
Fcr operatica as a power level less than high power level, ar defined above the in-ccre 1ccal pcwer monitoring system is not r
r required presided that at least five of the six external power range neutren fitx monitors are in operation and are so con-nected that indication of reactor. thermal power exceeding 62.5%
of the authoriced power level by any one monitor will scram the reactor.
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'l The mar.imum local heat fluz is given in Item 3 of Section D, entitled Power Operation. The in-core local power monitors will be used to corroborate the calculated shape and absolute j
vaine of the core power distribution + This power distribstion' will be used, together with appropriate analytical technigaes, to deterr.ine the limiting thermal characteristics of each type of fuel assembly, b.
A four-position sa.fety system selector switch is provided to bypass those scram trips which are not reqaired er desirable under some conditions. The bypasses for.the external system are indicated in the " Remarks" colunn of Table I.
The in-core local
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power monitoring system may be bypassed in the " shutdown" position and the "refhel" positien. The four positions for the selector switch are:
i " Start" position, to allow startup before full condenser vacuum is established; ef.c ii " Plan" position, for normal plant operation with period trip bypassed; iii " Refuel" position, to allow some control rods to be with-draws for safety during refueling if the' high neutron finx sensore have been set to trip at a low value; iv " Shutdown" position, to allow ventilation af the reactor j.
encicrure while testing or maintaining the safety system j
with the reactor shutdown. Centrol rods cennet be with-drawn in this position.
- 10. Radiation-Type Recctor Instrumentation In addition to the external and internal core neutron flux channels and the period channels that are a part of the reactor safety j
system, there are provided two indicating startup channels and a battery-operated neutrca flux channel, t
- 11. Reactor h elorure i
The enclosure hearing the reacter and tha eteam.? pply systen is e spherical stesi ehell,. 190 feet in dienoter, with the eqaater apprcrimtely 56 feet a.bove g:wnd level. Fron the enciccare, the minir.n off-site distance is one-half mile to:
a.
Skinner Island in the Ennkakse Eiver; b.
The navigation channel in the Illinois Riter; and c.
The land boondaries of the site.
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The enclorare was designed, bailt, and tested in accordance with the ASME Boiler and Prerrare Vessel Code. Pertinent design parameters for this vessel are:
Dasigr. Pressare 20.5 psic Design Ter.perautre (cofmcident with design pressure) 3:50 F Marix.m Wind Velocity HO mph Horicontal Acceleratien 0.3% gravity Pari =:.m Leakage Rate-at 37 psig 0.5%/ day All normally open lines penetrating the enclosure shell through which leakage could credibly ocrar in the event of the "maximam credible accident" ars provided with check valves or isolation valves which close withcrat operater attention. Table I indicates the sewt signals which initiate closure of the automatic isolation valves. The isolation valves on the primary steam lines, which are not clesed from any scran eignal, will cloco automatically from low reactor prescare -- a ecndition that would be encountered in any system rupture apprcaching the severity of the postclated "rM=nn credible accident." All the prinary 1sclatien valves are backed up by additional valves which can be operated from positions that are tenable after the accident. Normally closed lines penetrating the enclosure are protected against being opened duri.g operation, or in p/or oport. ting rulee.ctentially hazardots ron~:p7 rating situatione, by interlockr ar.d The enclora. e is provided with t. post-ir.eident cooling system, for uee in the event of a =eriota p-imary system re.pture, to aid in j
the redaction of enclorare pressare and leakage. This e-designed for a hect renoval capacity cf t.t least 30 x 10kstem isBtu/hr I
at an enc'lorcro internal terperature of 2560F.
Operation of thic cooling systen win b2 automaticany initiated by the scram Signal irom high epht*3 pr3SSare after a time delay not i
to e:teeed ten mi utes. "Ye reacter c; err.ter may also manually initiate operation of thic syctem or cura-ide the autor.atic
' initiation signal. The system will te maintained in operable con-dition at an times the primary syrten is pressurized. Periodic te-te u111 be conde.cted te de..act ste the proper functiening of the pc.It-ire 2 r9nt r.coling rytten.,
At or bsfon; the time of ths. first in0pection of the nodified centroi 706 drives, but not lawe than 0:,tobsr 15.19ol, c leakage test of the encim.re win be con t:cted at a lew pressure (less than 10 psi) tc, deurnine the rdia%11ty Of renetratione. This test shan include nectartnent of the integrd. leak rate of the conted.n-ment as t>ccarately r.3 poacible and. as part thereof provision t
chall be irrde to check the leah rate thrc;gh the em eteam line clrateff valvve.
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After a study of th'e results of the above leakage testing and dis-cussion of these results with the AEC, the future program of testing will be agreed to.
12.
Control Room The control roon is shielded from all power plant equipment and from the reactor enclosure so as to be tenable even in the event of the " maximum credible accident".
The minimum thickness of concrete shielding between the control room and enclosure for direct-line radiation is the equivalent of five feet.
C.
OPERATING PRINCIPLES The basic operating principles that will be adhered to in the operation of the Dresden plant are as follows:
a.
Operatien and control of most of the power plant equipment will be centralized in the control room, b.
The control room will be manned at all times by at least two operators, one of which shall be licensed except when the safety system selector switch is in the shutdown position and locked.
During this time one licensed operator will man the control room, c.
While mcst cperating and centeci functions are initiated in the contrci room, operators may perform scme functions at remote operating panels and valve racks -- at the direction of the i
cont cl room staff or with their prior knowledge.
i d.
Startup, normal shutdown, and all other repetitive operations will be performed in ac:ctdance with specific check lists.
e.
Maintenance of nuch of the equipment cutside the reactor shielding may be undertaken by ccntact methods and without overal} plant shutdown.
Plant shutdown and semi-remote methods will be employed as necessary, f.
All tests End rcutine maintenante of protective devices and power plant equipmsnt wi.11 be dcne in at:crear.;c with pres:ribed s che cu;e r,.
g.
Radiatict :renitcring by fired or portable instrumentation will be provided f cr entry to a.. radiation :enes, h.
All personnel 1 caving a contaminated radiation zcne, and equipment being removed frem such zcnes, will be appropriately surveyed to assure control of contamination. ;
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Irradiated fuel is to be moved from the reactor to storage; under water, by semi-remote methods.
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- j. Enclosure isolation provisions (i.e., air locks and equipment hatches closed, and instrumentation operating to close the automatic isolation valves if it should beccme necessary) will be in effect during all periods of reactor operation, including startup and shutdown operation, and during any operation involving
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insertion or removal of fuel assemblies in the core or withdrawal of control rods when the reactor head is off.
I k.
Qperation of the radioactive waste-handling system will be done in such a manner that it will be unlikely that the disposal of radio-active materials will result in the exposure of any persons on or off the plant to radiation in excess of the permissible limits.
1.
The plant is so protected at.all times by automatic safety devices that no single operator error or reasonably conceivable combination of operator errors could cause a severe accident.
m.
All significant unexpected incidents, unsafe acts, or incidents of excessive exposure to radiation will be investigated to effect procedures to prevent recurrence.
n.
In the event of any situation which may ecmpromise the safety of continued cperation, it will be the required procedure to shut the plant down as quickly as the situation ctils for, and to take other planned emergency attions to protect persons and property.
D.
P0h'ER OPERATION 1.
Approach to Rated Power i-t After any shutdown the approach to rated power shall be accomplished in a gradual stepwise fashien; and reactivity, power distribution, ar.d stability shall be carefully observed at all times.
2.
Safety System Scram Settings Operat;ve stram ser. sors and their settir.gs with the safety system seletter sw;tch in the " Start" and "Run" positiens are given in iten B.9.
The centro! rods cannot be withdrawn in the " Start" position until ct leas t.10 inches of Hg condenscr vacuum have been cbts;ned: and the switch to the
'Rtm" position cannot be nade without scramming until at least 21 inches of Hg condenser vacuum have'been obtained. The reactor will also scram if a. reactor pressure of 200 psig is exceeded prior to obtaining at least 23 inches of Hg condenser vacuum. D"}D D TQ
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3.
Determination of Maximum Reactor Power The rated power of the reactor shall be limited to a naximum steady state value of 700 }tw(t).
The maximum allowable steady state heat flux limits expressed 2
in units of Btu /(hr)(ft ), shall never exceed the following values:
Fuel Type I 370,000 Fuel Type II 410,000 Fuel Type II I
(maximum of 2 in core) 500,000 Fuel Type PF-1 through PF-4 480,000 Fuel Type PF-5 through PF-9 470,000 Fuel Type PF-10 through PF-12 510,000 ne
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a minleur. Lurnout ratio, o: at Acast 2.9, evat t a t.-
.t 125
.r-cent of rated power, will be maintained in each type of fuel closest to burnout in the hottest channel in the core based on a uniform steam quality over the cross section of the channel.
This burnout ratio shall be based upon the correlation in Edison's
" Recommended Curves of Burnout Limit for Design and Operation of Boiling Water Reactors", dated January 5,1962.
The reactor shall be operated always well within the bounds of stability, as evidenced by the operation itself and any experimental data produced.
4, Pressure Limits i
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!!aximum normal reactor operating pressure 1000 psig a.
h.
Maximum pressure setting for automatic reactor shutdown 1050 psig
'taximum pressure setting for opening of c.
1 electromatic relief valves 1085 nsir d.
"aximum pressure setting for opening of first main safety valve 1250 psig e.
"aximum safety valve pressure setting 1250 psig 6
f.
Combined capacity of safety valves 2.1 x 10 lb steam /hr.,
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5.
Raactivity Limits 4
a.
The average reactivity addition rate from withdrawal of the control rod with the maximum reactivity worth in the most adverse withdrawal pattern will not exceed 0.0029a k/sec.
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b.
With the reactor in any condition, the following shutdown criterion shall be met:
~~;g "St ack nod" Criterion:
At every stage during loading and in the fully loaded configuration, the c
control rods must provide a shutdown control margin
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of at least 0.01a k with any rod wholly out of the
.I core and completely unavailable.
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During core alterations after the first fuel cell is loaded, the following " cocked rod" criterion shall be met:
" Cocked Rod" Criterion:
The reactor must be sub-i critical by at least 0.01Ak with at least one control rod fully withdrawn in the region of the alteration and availabic for rapid scram insertion.
1 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 otherwisc 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 least 0.01 A k, the liquid poison will be introduced prior i
to reaching the " cold" condition where this criterion
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could not be satisfied.
d.
Void coefficient (s k/% change in voids).
The void coefficient averaged over the interior of a fuel channel will always be negative when the core is critical, e.
The moderator temperature coefficient of reactivity during uniform heating of the core shall be limited so that the potential maximum reactivity addition available from nuclear heating of the moderator shall always he less than one dollar.
t The maximum moderator temperature at which the coefficient is positive shall be less than 550*F.
Measurements shall'be made to confirm that these conditions are met, 7..
6.
Naste Disposal The disposal of wastes resulting from power operations is dis-cussed in item B.S.
Disposal of all waste off site will be in a D #
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L manntr such th t it is unlikoly any person will receive radiation '
exposures in excess of the approximate permissible limits.
E.
REFUELING AND MAINTENANCE 1
Operating Principles 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 pertinent to these operations are those. lettered b, d, e, f, g, h, i, j, and 1.
2.
Safety System Scram Sensors Operative scram sensors and their setting with the selector switch in the " Refuel" position are given in item B.9.
l Since some maintenance can be carried out under any of the possible reactor conditions, the safety sensors in operation 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 discussed in items D.5.b_and D.S.c.
During movement of fuel in the core, or centrol rod maintenance, the minimum shutdown margin will satisfy the " cocked rod" criterion given in item D.5.b.
4.
Liquid Poison System i
The liquid poison system will be operative during refueling and maintenance operations in the reactor as well as during normal power operations.
5.
Minimum Critical Testing A critical core may be constructed for testing purposes, using any combination of fuel within the limits of these Technical Specifications, subject. to the following restrictions:
a.
The minimum critical shall be located within the control rod patte rn.
b.
A minimum of three neutron sensitive instruments which are connected to the safety cir:uit shall be located inside the
vsssel in th) vicinity of'tha srall tcre.
Tha circuitry shal1
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be arranged so that any one of the three can actuate a scram.
Two of these sensors will scrza on high neutron level, and y
one will scram on short period.
At least one Icw level neutron monitoring instrument shall be operating and located in the vicinity of the small core.
c.
The shutdown criterien in item D.5.b will be met.
d.
Except for the final fuel increment, the size of each fuel increment will never exceed one half the estimated critical increment er cne assembiy, whichever is larger.
This estimate will be based on neutron cultiplication measurements made between fuel additions.
The final fuel increment will not I
exceed one fuel assembly.
6.
Refueling Instrumentation At every stage of refueling or maintenance involving replacement of fuel, the following minimum instrumentation requi:ements shall be met:
a.
All in-vessel low level neutron monitors designated in this section shall be operable as required herein and be sufficiently sensitive to detect and register changes in neutron level that would result from the insertion of fuels or movement of control
- rods, b.
A minimum of twc centinuously cperating low level neutron nonitors is required in the vicinity of refueling or maintenance activity which involves fuel additicn whenever there are more than four empty fuel cells in the core.
Exception to this*
requirement shall be as noted in items
- C" and "D" below, No in-vessel neutron sensitive devices will be required during c.
fuel changes which do not involve fuel additions or replace-ments; none will be required during refueling or maintenance involving removal and replacement cf one assembly at a time or when there are four c: less vacant fuel cells in the control zone of the core, d.
A minimum number Of I ve neu rcn sensitive devices: two of which sha_1 De cperat:ng Icw itve. neutron monitors, shall be ic:ated ins dc ne reatic: vessel whenever more than 16 empty Edj acent fue. celis are to be refueled.
Three of these devices shall be ::nne::cd te the safety system and arranged so that any cne of the three will actuate a scram.
Two channels will provife fo: s::am en high neut:cn level and one channel will provide for scram on short pericd, The Icw level neutron m:nitoring devices will be located in the vicinity of the immediate alteration.
These requ rements shall not redete the total safety system requ.Iements as specified under 3
E.2.
-ne numbcr of ex:erna. safcTv system : Icuits may be redu:cd by tne ncrse: of :nternc_ monitors pla:cd in the safety systen. DMD 4]Juulk DWyR d
1 s
TESiS AND 33PECIIG.iS OF COh"fROL RCD IEi.IVES Tests and inspections of centrol drive mechanisms shall be made according to the follwing plan while the reacter is shnt dcwn. Records shall be maintain:o ef the da',e :htained by each test or inspection. Proper test cond51c.ne shall be utablished in a nanner consistent with the nature of the observati r.s to be made. These tests represent a minisa:.m reqairement.
Additicssl tuting shs"._1 be pe..-fered as nay be necessary to gather signincar.t dsta ec.r.cerning the activity being investigated.
1.
liomi Terive Over.tice f
These c,ects sh.tll be nado with centrol blades properly attached to their respective drive mechanicms.
Each drive sha"1 he exercised through the full length of the drive stroke withoc.t stopping. Time elapsed in movement of each blade betwec., ths extrene p.u".tien+., shal?, be measured for movement in both din cricr.r.,
Esch drive :,hs11 be eccrcised up and down, stopping at each latch p ultic.. P cper er fr:lty latchir.g, unlatching, position switch cperation, pcsitien indic.ater cperations, and movement of drive shall be observed.
.d :.. w f.r.a ai.6.As.: thi.ll ba tested in the foregoing manner
.;e ;f 3, r, v. y.. ar*./. it cf irt;..'.an which is expscted to exceed h6
~~
_ e... <. en d.% c
,r sont r.:e 100 freqJency than once every gaarter.
.C.1 ufu.Gru shall be r;;.bscted to a. frictica test no less 2.q:.-r'0 7 t'.r. :.rw arery c:crim A test p: essure which permits i
o ve.e.<1; ire r. ora. enent, ci iricticr.ai fereed in the drive shall be 2
.: a r:.
- 7. e.>.i.f r ri'.y of =tio: t) cagh a full upward strcke and 7:his:ti,.w 1.. pretstro sha2. bv chee red.
2 Seven O e:ct': Test, t
Cniitior.: fcze ".hv.e tants i:4.1 he as for tuta in 1 above o
&./Mm t s of tha cf trr.vol sht.11 be cede as follcus:
c.
v.*. r."- vr. :-.
of noti::. to b;.iferr
. h 2.:. b d i rr Z.m., te m $ :i be n:de inring crery period of sl:tdown which is 2
erp4.cr. sa n c:r.u.G ';.8 c.w.rs., c.d in any cTent ne less fregaently
- &:.. :+ r.:.q q u:q vr,
u h
IdD lD'3~3h
$" ^'r *J A X R i
4
- 3. Tests Mer,to Rcirstansdion of Ke.actg Hez.d Whensys the rse.cter hand has been removed for any cruee during any sbntdann, each contrel red shall be given a ;tl~ ts:t to demonstrate attach =ent of tha blade to its' driTc prior to reinetal'.ation ei the, head en the react.,r vessel.
L OPERATIlFJ TOCEUJRF(
Opers. ting pr: eedires and limitations shun to in ceco-dtnce with the following r+quiromentst 1
In7satigation of Ancar.1:n.s Centrol P.ed Ibive 3ehe.vior Records chall be mintcined reflecting the oerarrence, investigation, es::ss, effsets, and safety significcnces of anoms.lics and any resulting corrective er remed.ial mecsuras, licensee sht11 pro:rptly report in c iting te the Connissien the incidence of cay appsrent drive mal-fert.ctiet. whbi; :etcirso sucptavien of racetcr cparazion in order to carq cut the previsions of this se: tion.
In cacc of cny observatien of ano.w?cas behavier of cny d-ive, there shall be prcmpt and thorough investigation to deter: tins the cause, effecte and safety significan:s of the oc7;rrence. Dr.e sts;.derd c.? cro :alace ber.vio.r for a 6.-ive ohr.1 be dsvintien from perferu.,.as iptci'icezio. establishad for tha preopers.ticnal testing pregrs.r, it iddent: No.1 to the Rew.--t on *.>resden Control Rod D 17s Ldifier tion.s, fA.tsd Fabsrxry D, 2M1 Operation ray be cont". wf, or'rmme:S after chtdova e517 if it his b en deterrined thst tne e. nom]ots behr.viev observed in c pr.rticalar mechanism does acb ir ai-the ability to control the reacter c:' indicate that i.g.si: vat of tz:e perfurnance cf othar mic;x.irus nay be iminent.
Cperation :ory he continued, or resumsd after s.mr:c m, with a defective centeel mehri.nt which hst: been deactivatsd so as to le.ck its con-t-cl'bicio Le place, providsd that (1) the e tuck r:d* criteria of s
t% ?.icensi csr.be met (2) it has bser d.ste nired that the inactive 3
drive dwt ato i:nmir the chility to contr:1 the reactor except for rnwnilability of the inactive v. rive in shtivn, srd (3) it has bued dne:ri.w? tM the conditien cf the inactive d;ive does not indicatt W.0.'.ptirrsnc of t:.e psrfe..re:e er ether.eehanisms i
. sty be hd.- en.
2
- tre.kr
- Xtib:n.to:st; of Co. trol Lcde Iteition T'.o Terrie:.? cb1?.1 tc p:-nidad.r.'/2 e definc.d 72d withdrawal se?cer.ce sad a p: edicted critical red cer.figura. tion f ar each start upo 'Ihe operator ch'i'.1 follow this uithdtc al seqtance.
b.
Il.rin;; stcrt-up the notion of peicen blades shall be verified, iniefsr en patrible, by cbserving the reg: ree of the external i-#:nt entstice.
.9 u
1
-d c.
After K effective is equal to or greater than 0.995 (determined on the basis of either predictions or observations), no unveri-fled blade having a worth of more than l's K shall be withdrawn or remain in a withdrawn position.
l d.
Up to one rod or 1.0% K, whichever is more restrictive, may be withdrawn in addition to the predicted critical control blade patte rn.
If criticality is not attained during this withdrawal, l
an attempt shall be made to verify all unverified blades by observing response of nuclear instrumentation to movement of control rod drives.
Any blades which are not verified during this operation shall be inserted.
Thereafter, all blades withdrawn must be verified.
l e.
All blades which were not verified for following during the start-up rod withdrawals will be verified as soon as possible after criticality is achieved.
Any blades which cannot be l
verified for following shall be inserted.
In the event that the verification tests during this operation do not show blade following or separation, then additional verification tests shall be conducted at operating power levels when the in-core monitors are effectife, f.
The provisions of the previous paragraph (c) will apply to all blades withdrawn during critical operation.
1
(
g.
During periods of sustained operation, the following of all i
poison blades will be verified at least once each week.
h.
Records shall be maintained to show for each operation:
l 1)
Predicted and actual control blade patterns for I
criticality; t
l
- 2) The identity of all unverified blades and their eventual i
disposition, including circumstances under which they l
were verified; and I
- 3) The worth of each unverified blade involved in operations l
with K effective greater than 0.995, j
H.
PROJECTED INSPECTIONS OF CONTROL ROD DRIVE MECHANISMS, POISON BLADES l
AND CORE GRID STRUCTURE The inspections described in
'1" and "2" shall be performed (a) after 1500 and before 3000 hours0.0347 days <br />0.833 hours <br />0.00496 weeks <br />0.00114 months <br /> of reactor operation at pressure exceeding i
100 psig following the resumption of operation after shutdown in November 1960; and (b) after 4500 and before 9000 further hours of.
e a :r-reactor operation at such pressure subsequent to the first inspection.
Of the drives.and blades selected for the second series of inspections,
'one or two shall have been selected for the first inspection.
,The inspection describedin '3' shall be performed at the time specified as
- (a)*above.
1.
Control Rod Drive Mechanisms A MM=im of six drive mechanisms shall be removed from the reactor, disassembled and inspected. Four drives shall be selected which have experienced the most severe conditions of high water temperature and mechanical service. Two drives with comparatively light service shall be inspected. Any additional drives that have during operation -
given indication of possible defects shall be removed and inspected to the extent necessary to determine whether or not such defects exist.
The drives shall be disassembled to permit (a) visual inspection aided by hand lens magnification (5 - 10I), (b) fluid penetrant inspection, and (c) ultrasonic inspection of the following parts:
Index tube and piston head assembly, i
Conet lock assembly (guide sleeve, retainer, and conet finger).
Guide Ping.
4
{
Shuttle piston.
1 The rener mount-sssembly shall be inspected by methods (a) and (b) abo've. Roller operation shall be checked by rotation of each r' oller A.
~
on its shaft.
The remaining' assembly, consisting of.the flange, piston tubs, and cylinder tube, and the graphiter seals shall be visually inspected.
The procedures to be followed in ultrasonic inspections shall be 4
essentially as described in Section D.2. of # Addendum No. 1 to the Report on Dresden Control Rod Drive Modification" dated February 20, 1961 The fluid penetrant inspections shall be performed so as to detect surface defects. Indicated defects shall be examined visually with the aid of magnification to determine their nature.
2 Poison Blades A minimm of six poison blades shall be removed from the reactor azid inspected. Inspection shall consist of thercugh visual exacination for structural defects and measurenent for changes in shape or dir.ension.
'.i-t
a 3.
Core Grid Structure The inspection will consist of examination of all previously located cracks; examination of fillet welds on the reverse side of beams opposite cracks; examination of a sufficiently.large sample of pre-viously examined welds to give reliable indication of the integrity
=
of the core structure as a whole, including a minimum of 15 of the beam to ring welds previously examined.
Date of Issuance:
SEP 19 EB2 i
1 h i.
T/mLE K SAFETY SYSTDi Exterviot Sensors Scram Scum W Mer Monatic Number Joincidence Type in each in each Trip Setting Trip Point Functions Remarks Channel Channel Perf6rmed Ifigh Sphere Closes isolation No bypasses Pressure 2
1 out of 2 Maximum Pressure Setting [
valves & ventile of 2.0 psig 0.2 psig.
tion ducts I.ow Water 2
1 out of 2 '
At: level;which31s" Setting [
Closes isolation Bypassed in " Shutdown Level in a minimuC of'.43".
1" valvee & ventila-position prior to initial Reactor above the top of
}
tion ducts power operation only.
Vesnel the active fuel s
Bypass is to be removed.
a Closure of 2
1 out of 2 When oil pressure Settingf'5-Closes ventila-Bypassed in " Shutdown" or Turbine Stop controlling these psig tion ducts.
" Refuel" positions.
h Bypass valves drops to a Vnives minimum of 50 psig Closure of 2(b) 2 out of 2 Closure of both Setting [
Closes ventila-Bypassed in " Shutdown" or Primary Stean (one from valves beyond a 5% stroke tion ducts and "Refuct" positions.
Sphere Isola-each valve) maximum of 25% of starts emergency
- t. i on Valves stroke cooling -
1.ow water 2
1 out of 2 At level which is a Settinghl" Closes ventila-Bypassed in " Shutdown" or 1.evel in maximum of 12" below Primary the drum center line tion ducts
" Refuel" positions.
g Stenm Drum 4
Im Conden-2(c) 'l At minimum condenser SettingY Closes ventila-Bypassed in " Shutdown" or ner Vacuum
' vacuum of 22" Hg 0.25" j tion ducts
" Refuel" positions.
By-
-l passed in " Start" position-if reactor pressure is below 1
At minimum condenser.
Setting ![
200,psig & condenser vacuun vacuum of 23" Irg 0.25"
}.
is greater than 10" of Hg.
)
i l
I Ifigh Roset.or
-2 1 out of 2 At maximum reactor; Setting f f Closes verit'ila-Bypass'ed in " Shutdown" h t-7eure' presstire of 1050 peig 10 psig.
t.ien ducts' & :
. position.
Maff!,***rauncy
.s a09
TABLE I SAFETY SYSTEM (Continued)
Ex7ernal Sensors
~
Scram Scram (a)
Other Autamatic Typee Nuraber Coincidence in each in each Trip Setting Trip Point Functions Remarks Performed Channel Channel
.r ~
liir,b Level 2
1 out of 2 At tank level Setting f 1"
Closes ventila-Bypassed iri " Shutdown" in Scram which is 4'41"'
tion ducts position Manual bypass 3
Dump Tank above the base also prevents contro1 rod line of the lower withdrawal. At scram tangent of the' point, there is suf-tank ficient free volume remaining in the scram dump tank to accommoda:
the water from 2.7 scrams.
!!s r,h Neu tron 3
1 out of 3 When leakage Setting f 37.
Closes ventila.
Bypassed in'"Sh6tdo'un'.' pon
~
riox or 1 out of flux indicates of iated power tion ducts ition. Interlocked in 2 if 1 is a maximum of
" Refuel" position so as tM '.
bypassed 1207. rated power require reduction in trip setting to a
- +
maximum setting of 163 rated power. At any time other than in the " shut-down" position only one of the six flux trips can be bypassed.
iho r t Period 3
2 out of 3 At minimum Period Setting d 0.5 Closes ventila.
Bypassed,in " Shutdown" of 4 seconds acconds tion ducts and "Run" positions.
(a) The point at which a scram is actually initiated nay be different from the ' Scram trip setting" by th e emount of the tolerance f or instrument inaccuracies. The amount of this tolcrnnce is indicated by the values given in this column.
(h)
Each valve-position switch (of which there is one per valve) has two contacts in each safety channel.
(c) There are two vacuum sensors.
Each sensor has one contact in each of the two safety channel.
(d).Thore.are three period sensing chambers.
Each chamber has two contacts in each of the tuo safety channels.
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AFFE CIX "B"
.T,p COMMONWEALTH EDISON COMPANY FACILITY LICENSE Estimated Schedule of Transfers of Special Nuclear Materf al from the Comission to Comonwealth Edison and to the Comission from Commonwealth Edison Returns by Co=on-Date of Transfers from wealth Edison to Transfer AEC to Common-AE0 Net Yearly Cumlative (Fiscal wealth Edison Cold Irradiated Distribution Distribution f
Year)
Kilograms U-235 Kilograms U-235 Kilograms U-235 Kilograms U-235 Thru 1962 1,8h2.1 362.0
-o-1,hBo.1 1,hBo.1 1963
-o-30.8 219.2 (250.0) 1,230.1 196h 292.0 90 7 201.3 1,h31.h 1965 365.0 30.8 77.2 257.0 1,688.h 1966 365.0 38.5 111.2 215.3 1,903.7 1967
-o-38.5 138.3 (176.8) 1,726.9 1968 292.0 292.0 2,018.9 1969 292.0 30.8
.120.5 1ho.7 2,159.6 1970
-o-
-o-
-o-2,159.6 1971 292.0 30.8 109.5 151.7 2,h63.0 311.3 1972 292.0 30.8 109.5 151.7 2,
1973 292.0 30.8 109.5 151.7 2,61h.7 197h 292.0 30.8 109.5 151.7 2,766.h 1975 292.0 30.8 109.5 151.7 2,918.1 1976 292.0 30.8 109.5 151.7 3,069.8 1977 -o-
-o-3,069.8 1978 292.0 30.8 109.5 151.7 3,221.5 1979 292.0 30.8 109.5 151.7 3,373.2 1980 292.0 30.8 109.5 151.7 3,52h.9 1981 292.0 30.8 109.5 151.7 3,676.6 1982 292.0 30.8 109.5 151.7 3,828.3 1983 292.0 30.8 109.5 151.7 3,980.0 198h 292.0 30.8 109.5 151.7 hh,,131.7 1985 292.0 30.8 109.5 151.7 283.h 1986 -o- -o-h,283.h 1987 292.0 30.8 109.5 151.7 h,h35.1 1988 292,0 30.8 109.5 151.7 h,586.8 1989 292.0 30.8 109.5 151.7 h,738.5 1990 292,o 30.8 109.5 151.7 h,890.2 1991 292.0 30.8 109.5 151.7 5,0hl.9 1992 292.0 30.8 109.5 151.7 5,193.6 1993 292.0 30.8 109.5 151.7 5,3h5.3 199h 292.0 30.8 109.5 151.7 5,h97 0 1995 -oo
-o-5,h97.0 1996 292.0 30.8 109.5 151.7 5 6h8.7 Total lo,16h.1 1,? ? ? 4 3;275,6 5,6h8.7
._.