ML19340A902
| ML19340A902 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 08/06/1962 |
| From: | Lowenstein R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Joslin M COMMONWEALTH EDISON CO. |
| Shared Package | |
| ML19340A903 | List: |
| References | |
| NUDOCS 8009080568 | |
| Download: ML19340A902 (27) | |
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- "' '81 UNITED STATES
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ATOMIC ENERGY COMMISSION g
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X+D' WASHINGTON 25, D.C.
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%,n e,.m Docket No. 50-10 Mld C E2 Commonwealth Edison Company 72 West Adams Street j
Chicago 90, Illinois t'
Attention:
Mr. Murray Joslin i
Vice President Gentlemen:
By application dated January 5, 1962, the Commonwealth Edison Company has requested authorization pursuant to paragraph 3a(4) of License No. DPR-2, as amended, to load up to 108 Type II fuel bundles in the Dresden reactor.
In addition, several changes to Appendix "A" (Technical Specifications) have been proposed to accommodate the proposed refueling and to delete certain sections of the Technical Specifications which pertained to initial load-ing and critical testing of the reactor.
Additional information concerning this application was submitted with covering letters dated March 27, April 25, June 8, and July 16, 1962 j
i We have reviewed the information submitted and have determined that the loading of up to 108 Type II fuel bundles in the Dresden core and the proposed changes in the Technical Specifications do not involve significant hazards considerations not described or implicit in the hazards summary report as amended and that there is reasonable assurance that the health and safety of the public will not be. endangered.
j Accordingly, you are hereby authorized to load up to 108 Type II fuel bundles into the Dresden core as described in your appli-cation of January 5,1962 as supplemented by your letters of March 27, April 25, June 8, and July 16, 1962.
Pursuant to your application as supplemented, we have revised Appendix "A" in its entirety.
In addition, the Commission has revised Appendix "B",
which lists schedules of special nuclear material transfers to Commonwealth Edison Company and returns to the Commission in accordance with your application of February 20, 1962.
E 56W 8009080 qs
AU3 6 EE2 A copy of our hazards analysis concerning the fuel relcac together with copies of Appendices "A" and "B" to License No. DPR-2, as amended, are enclosed.
Sincerely yours, Cr:g:nal s'ined tt R. Lcwenstein.
Director Division of Licensing and Regulation j
Enclosures:
1.
Hazards Analysis 2.
Appendix "A" 3.
Appendix "B" i
Distribution H. J. McAlduff E. G. Case R. Lowenstein C. K. Beck r
H. Shapar P. Travelstead P. A. Morris-2
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E. E. Hall D. C. Clark E. Trennel l
H. Steele i.
R. Huard R. Cunningham Doc. Roon Formal Suppl.
LB-LSR readings ND:4ason For concurrences, see attached sheet
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2-Copies of our hazards analysis concerning the fuel reload and Amendment No. 7'to License No. DPR-2 are enclosed.
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Director ty R. Lcv.enstein.
Division of Licensing and Regulation Enclosurec:
Ha:ards Analysis Amendment No. 7 Distribution H. J. "cAlduff E. G. Case R.
lAuenstein C. 1:. Beck
- 11. Shapar P. Travelstead P. A. " orris 2 E. E. liall D. C. Clark E. Tremmel H. Steele R. !!uard R. Cunningham Doc. Room Formal Suppl.
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COMMONWEALTH EDISON COMPAh"I r
DRESDEN NUCLEAR POWER STATION
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p DOCKET NO. 50-10 I.
4 APPENDICES TO FACILITY LICENSE i
License No. DFR-2, i
as amended 4
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i APPENDIX "A" Technical Specifications I.
l.
I' APPENDIX "B" Estimated Schedule cf Transfers cf Special Nuclear Material
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Appendix "A" to DPR-2 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.
Sections B and C set forth the design and performance speci-fications and operating limits and principles.
Sections D and E specify the limitations to be observed l
during start-up, power operation, and refueling and maintenance ope rations.
In these sections, as well as in Section B, where maximum or minimum limits are not given specifically, the values given are " design" values which are subject to normal manufacturing and other tolerances.
Sections F and G provide certain additional operating and testing procedures applicable to the control rod drive mechanisms.
Section H provides minimum requirements for certain inspections of the control rod drives, poison blades end core grid structure.
i B.
DESIGN FEATURES 1.
Reactor Vessel The reactor vessel is a vertical cylindrical pressure vessel, j
with dished top and bottom heads, made of low alloy steel and clad inside with stainless steel.
The vessel was i
designed, built, and tested in accordance with the ASME Boiler and Pressure Vessel Code,Section I.
Design parameters for the vessel include:
Inside height, including heads 40 ft 9-5/8 in Inside diameter 12 ft 2 in Design pressure 1250 psig Design temperature 650' F 2.
Nuclear Core Maximum core diameter (circum-scribed circle 129 in 1
i
4
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l 4
"aximum active fuel length - cold 112 in
?!aximum number of fuel assemblies by types i
Type 1 488 Type II 108 i
Type PF-1 through PF-12 (one each) 12
!!aximum total number of fuel assemblies 488 The various fuel assemblies may be located in any position of the reactor, provided over-all core symmetry is preserved j
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.
4 The reactor may be operated at any power up to and including rated power with any number of the various types of fuel
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assemblies installed provided the maximum number and 1
location are within the limits specified above, i
4 1
3.
Fuel Each fuel assembly consists of vertically-positioned, rod-f 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.
i The minimum fuel pellet density averaged over a fuel segment
[
is 94% of theoretical for all fuel assemblies except PF-7 PF-8, and PF-9 which is 90% of theoretical.
i 4
Control Rods and Drives t
The control blades, 30 in number, consist of small vertical stainless-steel tubes filled with compacted boron carbide l
(B c) powder.. The boron carbide powder is separated a
longitudinally into several independent compartments. The tube walls are designed to withstand the maximum cal-culated internal gas pressure.
The tubes are held in a F
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l 6.5-inch (over-all width) cruciform array by flat l
stainless-steel plates.
These plates, extending the full l
1ength of the control blades, provide a smooth, flat i
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 S-foot, 3-inch diameter cylindrical space in the central region of the core.
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, which will be available
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and effective under all conditions of reactor operation except " Shutdown" ( at which time inter-locks prevent any rod withdrawal), and b.
The reactor pressure, if this is greater than about 700 psig and accumulator pressure falls below reactor pressure.
l 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 i
reactivity.
The inserted position of the control rods is determined by a locking device which provides 12 discrete approximately equidistant positions for all rods.
Only one rod at a time can be withdrawn from the core.
Rods may be inserted into the core singly or all may be scrammed together.
Active length of control blades 8 ft 6 in Velocity for normal insertion or withdrawal 6 in/sec "aximum time from receipt of scram signal to:
10% control rod travel 0.6 sec 90% control rod travel 2.5 sec 1
l l
Maximm number of control rods i
per accumlator 3 non-adjacent rods Frequent and thorough periodic checks will be made to assure the proper functioning of the centrol rod drive system.
- 5. Liquid Poisen System system actuation Manual Maximm time after actuation until poison begins to enter core 20 sec.
Minimm. weight of boron in system (present as sodium pentaborate in solution) h00 lbs.
s Minimm poison worth' for operations with reactor vessel. head on 0.15a k Minimm poison worth for. operations with reactor vessel nead off 0.0L k 3
6.
Steam supply system Besides the reactor itself, the steam supply system ec= prises 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 i= posed.
The emergency condenser consists of two separate tube bundles of equal capacity in a common shell. Each tube bundle can be staz;ted automatically by the appropriate reactor safety system controls (discussed in item B.9 below), and may be also started cr stopped manally by remotely centro 11ed 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 emargency heat exchangers 1250 psig Design pressure of prirary system piping 1150 psig Minimm capacity of emergency condenser
@ of rated reactor thermal power h-D"D
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& &l]h 0
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i Ab s
' Minim 1m cooling water stored in emergency condenser 30,000 gal.
Mh M m emergency 'cohdenser cooling period (after scram) w 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 drun 1
center line. The pump trip is not a safety circuit function, but rather is provided for protection of the recirculating pumps against loss of inlet suction.
7.
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/n 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 0
a maximim of about 300 F.
j In addition to the condenser vacuum scram and isolation trips indicated later in item 9
- Safety System", a mechanical trip is i
provided to close the turbine bypass valves, as well as all other hydraulically-opened turbine valves, should the condenser ' vacuum j
fall to or below 7 inches of Hg. This vacuum trip is not a part' of the reacter' safety system, but becomes operative whenever the j
condenser vacuun has been increased to above the trip set point.
8 Waste Disposal Systens i
a.
Solid Wastes i
Solid wastes containing radioactive materials include filters, j
defective ecuipment, and other miscellaneous trash. Sdch j
material will be stored in accordance with AEC regulations (10 CFR-Part 20) which may involve storage in an underground _
cencrete storage vault at the site. When feasible such material r.ay be compacted bef ore storage. Spent contaminated resins are sluiced to an underground tank for indefinite storage.
Sinice water is decanted after the resins ~ have settled, b.
Licuid Wastes Eqaipment is provided to treat radioactive liqaid wastes by decay in storage, long-term underground storage, filtration, nectrf.:.ization, demineralization, or evaporaticn. The treat-nent will nos; of ten reruit in water which can be re -used in ths plant cr released to the river. Batch sampling vill be
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used to determine wheth-r permissible limits can be :nct for any liqaid process vastes to be released. No disposal of these wastes to ground will be nade.
The large gaantities of river water used for eqaipment 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 liqaid wastes to below permissible limits for release to the river. The release of liqaid wastes shall conform to the provisiens of 10 CFF., Part 20.
c.
Airborne Wastes Radioactive airborne wastes are discharged from a 300-foot-high stack. There will be continuous monitoring of the total stack flow and air-ejector flow. A holdap time is provided in both the gland seal exhaust system and the air-ejector exhaust system.
The air-ejector nonitor automatically initiates closing of the discharge valve en the agejector holdup gystem if the neasured rate cf discharge of Ie exceeds 2 x 10 uc per second, the measurement being rade after two minutes decay (travel time in the system from the reactor core to point of measurement). The reactor will be manually shutdown when the stack discharge rate of noble fission gases exceeds 7 x.10> uc per sec.
d.
General Commonwealth Edison shall not release into air any concentration of radioactive material which will result in exposure to con-centrations at ground levels in any unrestricted area as that term is defined in 10 CFR Part 20 in excess of the limits speci-fled 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 t
The reactor safety system will 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 design through-cut (that is, de-energizing will cause a scram), and both mst de-energize to cause a scram. Table I, below, lists the external safety circuit sensors, their raxi: rum or minimm trip settings, the number of sensors of each type in each channel, the coincidence reading feature, and the automatic functions performed in addition to a scran. In addition to the trips listed in Table I, an auto-matic scram is provided in the event of power supply failure, by virtue of the fail-safe design used. A manual scram centrol is A
available te the operater also, and an additional manual control
- i scra.s the reactor and c1cses the sphere isolation valves. D**]D
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I a.-
The in-core monitoring system is to provide in required locations
-indication of local power, automatic scram at not more than 125%
of rated local power, and alarm at a selected level belcw the scram setting. The automatic scram may be actuated by coincidence of i
signals from two or nore =cnitors, provided that the coincidence arrangement does not have the effect of leaving unmonitored a core region exceeding in size the limits specified below.
Whenever the reactcr is operating at a high power level (as used in this paragraph 9. "high power level" shall mean a thermal power level exceeding 315 M4 or 50% of rated local power), there shall be a sufficient number of operating in-core local power monitors to meet the fol]owing conditions:
i 1.
There will be at least three monitored horizontal layers reasonably evenly spaced in the regicn of the active core i
bounded by planes 1 ft below its top and 1 ft above its 4
bottom.
I i
- 11. Within the central 6.5 foot diameter vertical cylindrical core volume, no two adjacent horizontal layers may be without an operating local power monitor in any vertical l
cylindrical core volume that exceeds h ft. in dianeter.
iii. The in-core local power monitors will be so located that when the neutren flux within the core is purposely dis-torted by withdrawal of adjacent control rods from any region of the core, this distortion shall be detected by j
at least two operating in-core local power monitors.
Commonwealth Edison shall conduct experiments to demen-strate compliance with this recuirement.
iv.
There will be at least 32 operat'ng in-core local power monitors present in the core.
An opeisting in-core local power monitor is defined as one which has a response time of less than one second, indicates approx-i br.ately linear response to changes in local power and does not i
display erratic changes in calibration. Periodic tests will be conducted to demonstrate the operating conditien of the in-core local power nenitors.
For operation as a power level less than high pcwer level, as defined above, the in-core local power monitoring syste.- is not required provided that at least five of the six external power range neutron flux monitors are in operation and are so con-
~~
nected that indication of reactor thermal power exceeding 62.5%
of the authorized power level by any one monitor will scram the reactor.
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. - ~ -_.
The muie m local heat flux 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 value of the core power distribution. This power distribution will be used, together with appropriate analytical techniqaes, to determine the limiting thermal characteristics of each type of fuel assembly.
b.
A four-position safety system selector switch is provided to bypass those scram trips which are not recuired or desirable under some conditions. The bypasses for.the external system are indicated in the " Remarks" column of Table I.
The in-core local power monitoring system may be bypassed in the " shutdown" position j
and the " refuel" position. The four positions for the selector switch are:
I i " Start" position, to allow startup before full condenser vacuum is established; ii "Run" position, for normal plant operation with period trip bypassed; iii " Refuel" position, to allow some control rods to be with-drawn for safety during refueling if the high neutron flux sensors have been set to trip at a low value; iv " Shutdown" position, to allow ventilation of the reactor l
en:losure while testing or maintaining the safety system with the reactor shutdown. Control rods cannot be with-drawn in this position.
~
- 10. Radiation-Type Reactor Instrumentation
- f In addition to the external and internal core neutron flux channels 2
and the perios channels that are a part of the reactor safety system, there 2re provided two indicating 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 dia:teter, with the egaator approximately $6 feet above ground level. From the enclosure, the minimm off-site distance is one-half mile tot a.
Skinner Island in the Kankakee River; b._ The navigation channel in the Illinois River; and c.
The land boundaries of the site.
w The enclosure was designed, built, and tested in accordance with the ASME Boiler and Pressure Vessel Code. Pertinent design parameters for this vessel are:
Design Pressure 29.5 psig Design Temperature (coincident with design pressure) 325 F Maximm Wind Velocity 110 mph Horizontal Acceleration 3.3% gravity Maximm Leakage Rate at 37 psig 0.5%/ day All normally open lines penetrating the enclosure shell through which leakage could credibly occur in the event of the "maximm credible accident" are provided with check valves or isolation valves which close without operator attention. Table I indicates the scram 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, will close automatically from low reactor pressure -- a condition that would be encountered in any system rupture approaching the severity of the postulated "maximm credible accident." All the primary isolation valves are backed up by additional valves which can be operated from positions that are tenable after the accident. Normally closed lines penetratin,the enclosure are protected against being opened during operation, or in p/or operating rules.ctentially hazardous non-operating situations, by interlocks and The enclosure is provided with a post-incident cooling system, for use in the event of a serious primary system rupture, to aid in the reduction of enclosure pressure and leakage. This system is designed for a heat removal capacity of at least 30 x 10" Btu /hr at an enclosure internal terperature of 256 F.
Operation of this cooling system will be automatically initiated by the scram signal from high sphere pressure after a time delay not to exceed ten minutes. The reactor operator may also nanually initiate operation of this system or override the automatic initiation signal. The system will be maintained in operable con-dition at all times the primary system is pressurized. Periodic tests will be conducted to demonstrate the proper functiening of the post-incident cooling system.
At or before the time of the first inspection of the.cdified control rod drives, but not later than October 15, 1961, a leakage test of the enclosure will be conducted at a low pressure (less than 10 psi) to determine the reliability of penetrations. This test shall 'aclude measurement of the integral leak rate of the contain-ment as accurately as possible and, as part thereof, provision shall be made to check the leak rate through the nain steam line shutoff valves.
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 will be agreed to.
12.
Control Room The control room is shicided 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 concret ; 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.
Operation 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 most operating and control functions are initiated in the control room, operators may perform some functions at remote operating panels and valve racks -- at the direction of the control room staff or with their prior knowledge, d.
Startup, normal shutdown, and all other repetitive operations will be performed in accordance with specific check lists.
c.
Maintenance of much of the equipment outside the reactor shielding may be undertaken by contact methods and without overall plant i
shutdown.
Plant shutdown and semi-remote methods will be employed j
as necessary, f.
All tests and routine maintenance of protective devices and power plant equipment will be done in accordance with prescribed schedules, g.
Radiation monitoring by fixed or portable instrumentation will be provided for entry to all radiation zones, h.
All personnel leaving a contaminated radiation zone, and equipment being removed from such :ones, will be appropriately surveyed to assure control of contamination. l
~ -
1.
Irradiated fuel is to be moved from the reactor to storage, under water, by semi-remote. methods.
i
- j. Enclosure isolation provisions (i.e., air locks and equipment hatches closed, and instrumentation operating to close the
-automatic isolation valves if it should become necessary) will 2
be in effect during all periods of reactor operation, including startup and shutdown operation, and during any operation involving insertion or removal of fuel assemblies in the core or withdrawal of control rods when the reactor head is off.
k.
Operation of the radioactive waste-handling system will be done in such a manner that it will be unlikely that the disposal of radio--
I active materials will result in the exposure of any persons on or j
off the plant to radiation in excess of the permissible limits.
I 1.
The plant is so protected at all times by automatic safety devices J
that no single operator error or reasonably conceivable combination of operator errors could cause a severe accident.
j i
n.
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 compromise the safety of continued operation, it will be the required procedure to shut i
the plant down as quickly as the situation calls for, and to take i
j other planned emergency actions to protect persons and property.
1 D.
P0h'ER OPERATION i
1.
Approach to Rated Power l
After any shutdown the approach to rated power shall be accomplished y
in a gradual stepwise fashion; and reactivity, power distribution, and stability shall be carefully observed at all times.
4 2.
Safety ' System Scram Settings Operative scram sensors and their settings with the safety system selector switch in the " Start" and "Run" positions are given in j
item B.9.
The control rods cannot be withdrawn in the " Start" j
position until at least 10 inches of lig condenser vacuum have been I
obtained, and the switch to the "Run" position cannot be made without scramming until at least 23 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 j
inches of lig condenser vacuum. -
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e 3.
Determination of Maximum Reactor Power The maximum reactor power is defined as that therma) power at which the maximum heat flux for any fuel rod is reached.
This maximum heat flux, based on calculations and experimental data, will never exceed the following values in Btu /(hr)(sq ft):
Fuel Type I 350,000 Fuel Type II 445,000 Fuel Type PF-1 through PF-4 425,000 Fuel Type PF-5 through PF-9 415,000 Fuel Type PF-10 through PF-12 475,000 The peak rated heat flux and resulting rated reactor power are then set to 80% of their maximum values and, as indicated in item B.9, the high neutron flux scram setting will be no higher than an indicated 120% of the rated reactor power.
Ilowever, in no case will the high neutron flux setting be allowed to exceed an indicated reactor thermal power of 782 ?M (125% of the planned operational power of the fully loaded core).
The reactor will be operated within the above limits such that a burnout margin of at least 2.0 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.
The reactor shall be operated always well within the bounds of stability, as evidenced by the r;cration itself and any experi-mental data produced.
4 Pressure Limits a.
Maximum normal reactor operating pressure 1000 psig b.
Maximum pressure setting for automatic reactor shutdown 1050 psig c.
"ax'imum pressure setting for opening of electromatic relief valves 1055 psig d.
"aximum pressure setting for opening of first main safety valve 1205 osig e.
'faximum safety valve pressure setting 1250 nsig f.
Combined capacity of safety valves At least 150%
rated primary steam flow i 1
i i
5..
Reactivity Limits i
a.
The average reactivity addition rate from withdrawal of the control rod with the maximum reactivity worth in the most adverse withdrawal pcttern will not exceed 0.0029 A k/sec, b.
hith the reactor in any condition, the following shutdown 7
criterion shall be met:
" Stuck. Rod" Criterion : At every stage during loading and in the fully loaded configuration, the
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s j
control rods must provide a shutdown control margin of at least 0.01 a k with any rod wholly out of the j'
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core and completely unavailable.
During core alterations after the first fuel cell is loaded, j
the following " cocked rod" criterion shall be met:
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" Cocked Rod" Criterion: The reactor must be sub-critical by at least 0.01 Ak with at least one i-control rod fully withdrawn in the region' of the
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alteration and available for rapid scram insertion.
h'ith the reactor in the hot operating condition, operation 3
c.
will not be continued when the reactivity worth of the control rods known to be stuck out of the core, or otherwise j
unavailable for control, exceeds half the value at which hot shutdown could not be accomplished.
(It is expected I
that this would require' shutdown for repair of the inoperable i
j rods if three adjacent, or the equivalent in reactivity. worth of nonadjacent, rods ne known to be unavailable.)
If, in l
such case, it is calculated that following shutdown and j-cooling of the reactor the shutdown margin will not be at l
1 east 0.01 A k, the liquid poison will be introduced prior j
to reaching the " cold" condition where this criterion i
could not be satisfied.
Il d.
Void' coefficient (Ak/'s change in voids).
The void i
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 will change from positive to negative at a temperature below the maximum reactor temperature that can be attained under atmospheric conditions. Measurements will be made to con-firm this.
6.
Waste Disposal The disposal of wastes resulting from power operations is dis-cussed in item B.8 Disposal of all waste off site will be in a
, l-
l manner such that it is unlikely 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 i
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 t.
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.5.c.
i During movement of fuel in the cort, or control rod maintenance, the minimum shutdown margin will satisfy the " cocked rod" criterion given in item D.S.b.
i 4
Liquid Poison System The liquid poison system will be operative during refueling and p
maintenance operations in the reactor as well as during normal j
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 circuit shall be located inside the 14 -
3
?.
vessel in the vicinity of the small core.
The circuitry shall be arranged so that any one of the three scn actuate a scram.
Two of these sensors will scram on high neutron level, and one will scram on short period.
At least one low IcVel neutron monitoring instrument shall be operating and located in the vicinity of the small core.
c.
The shutdown criterion in item D.S.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 or one assembly, whichever is larger.
This estimate will be based on neutron multiplication measurements made between fuel additions. The final fuel increment will not exceed one fuel assembly.
6.
Refueling Instrumentation At every stage of refueling or maintenance involving replacement of fuel, the following minimum instrumentation requirements 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 two continuously operating low level neutron monitors is required in the vicinity of refueling or maintenance activity which involves fuel addition 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 of one assembly at a time or when there are four or less vacant fuel cells in the control
- one of the core, d.
A minimum number of five neutron sensitive devices, two of which shall be operating low level neutron monitors, shall be located inside the reactor vessel whenever more than 16 empty adjacent fuel cells are to be refueled.
Three of these devices shall be connected to the safety system and arranged so that any one of the three will actuate a scram.
Two channels will provide for scram on high neutron level and one channel will provide for scram on snort period.
The low level neutron monitoring devices will be located in the vicinity of the immediate alteration.
These requirements shall not reduce the total safety system requirements as specified under E-2 The number of external safety system circuits may be reduced by the number of internal monitors placed in the safety system.
15 -
_m.
?.
TESTS AND INSPECTIONS OF CONTROL ~ ROD DR1VES Tests and inspections of control drive mechanisms shall be made according to the following plan while the reactor is shut down. Records shall be maintained of the data obtained by each test or inspection. Proper test conditions shall be established in a manner consistent with the nature of the observations to be made. These tests represent a minimm reqairement.
Additional testing shall be performed as may be necessary to gather significant data concerning the activity being investigated.
1.
Normal Drive Ooeration These tests shall be made with control blades properly attached to their respective drive mechanisms.
Each drive shall be exercised through the full length of the drive stroke without stopping. Time elapsed in movement of each blade between the extreme positions shall be measured for movement in both directions.
Each drive shall be exercised up and down, stopping at each latch position. Proper or faulty latching, unlatching, position switch operation, position indicator operations, and movement of drive shall be observed.
All mechanisms and blades shall be tested in the foregoing manner during every period of shutdown which is expected to exceed h6 hours, and in any event no less frecuently than once every qaarter.
All mechanisms shall be subjected to a friction test no less frequently than once every gaarter. A test pressure which permits a sensitive measurement of frictional forces in the drive shall be used. The uniformity of motion through a full upward strcke and fluctuations in pressure shall be observed.
2 Scram Operation Tests Conditions for these tests shall be as for tests in 1 above.
Measurements of times of travel shall be made as follows:
a.
Time from start of motion to buffer; b.
Time in buffer These tests shall be made daring every period of shatdown which is expected to exceed h8 hours, and in any event no less frequently than once every gaarter.
I 7-4 3.
Tests Prior to Reinsta11ation of Reactor Head Whenever the reactor head has been removed for any cause during any shutdown, each control rod shall be given a pull test to demonstrate attachment of the blade to its drive prior to reinstallation of the head on the reactor vessel.
C.
OPERATING PROCEDURES Operating procedures and limitations shall be in accordance with the following recuirements:
1 Investigation of Anomalous Control Rod Drive Behavior Records shall be maintained reflecting the occurrence, investigation, cause, effects, and safety significances of anomalies and any resulting corrective or remedial measures.
Licensee shall promptly report in I
writing to the Comnission the incidence of any apparent drive mal-function which regaires suspension of reactor operation in order to carry out the provisions of this section.
In case of any observation of anomalous behavior of any drive, there shall be prompt and thorough investigation to determine the j
cause, effects and safety significance of the occurrence. One standard of anomalous behavior for a drive shall be deviation from performance specification established for the preoperational testing program, in Addendum No.1 to the Report on Dresden Control Rod Drive Modifications, dated February 20, 1961. Operation may be continued, or resumed after shutdown, only if it has been determined i
that the anomalous behavior observed in a particular mechanism does not impair the ability to control the reacter or indicate that impairment of the performance of other mechanisms may be imminent.
j Operation may be continued, or resumed after shutdown, with a defective i
control mechanism which has been deactivated so as to lock its con-j trol blade in place, provided that (1) the " stuck rod" criteria of l
the license can be met, (2) it has been determined that the inactive i
drive does not impair the ability to control the reactor except for unavailability of the inactive drive in shutdown, and (3) it has been determined that the condition of the inactive drive does not indicate that impairment of the performance of other mechanisms may be 1 =.inent.
2 Nuclear Indications of Control Blade Position a.
The operator shall be provided with a defined rod withdrawal segaence and a predicted critical rod configuration for each start-up. The operator shall follow this wC,hdrawal seqaence.
b.
During start-up the motion of poison blades shall be verified, insofar as possible, by observing the response of the external instrumentatien.
- 17
2'
, c.
After K effective is equal to or greater than 0.995 (determined on the basis of either predictions or observations), no unveri-fied blade having a worth of more than 1%;K shall be withdrawn or remain in a withdrawn position.
d.
Up to one rod or 1.0$K, whichever is more restrictive, may be withdrawn in addition to the predicted critical control blade pattern. If criticality is not attained during this withdrawal, 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 daring this operation shall be inserted. Thereafter, all blades with-drawn rust be verified.
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 verified for following shall be inserted. In the event that the verificatien tests during this operation do not show blade follow-ing or separation, then additional verification tests shall be condacted at operating power levels when the in-core monitors are effective.
f.
The provisions of the previous paragraph'(e) will apply to all blades withdrawn during critical operation.
g.
During periods of sustained operation, the following of all poison blades will be verified at least once each week, h.
Records shall be maintained to show for each operation:
- 1) Predicted and actual control blade patterns for criticality;
- 2) The identity of all unverified blades and their eventual disposition, including circumstances under which they were verified; and
- 3) The worth of each unverified blade involved in operatiens with K effective greater than 0.995.
. FROJECTED I' SFICTIONS OF CONTROL ROD DRIVE EECFJLNISFS, POISON BIADES
/dD CORE GRIL $?RUCTURE 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 100 psig following the resumption of operation after shutdown in November 1960: and (b) after h500 and before 9000 further hours of
-IS-
^
.=
reacter operation at such pressure subsequent to the first inspection.
Of the drives and b1Ades selected for the second series of inspections, one or two shall have been selected for the first inspection.
' The inspection desc.-ibedin "3" shall be performed at the time specified as
'(a)" above.
1.
Centrol Rod Drive Mechanisms A HMmm 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 daring operation given indication of possible defects shall be removed and inspected to the extent necessary to determine whether or not such defects exist.
The d-ives shall be disassembled to permit (a) visual inspection aided by band lens magnification (5 - 10I), (b) fluid penetrant inspection, and (c) ultrasonic inspection of the following parts:
Index tube and piston head assembly.
Collet lock assembly (guide sleeve, retainer, and collet finger).
Guide Flug.
Shuttle piston.
The roller mount assembly shall be inspected by methods (a) and (b) above. Roller operation shall be checked by rotation of each roller on its shaft.
The remaining assembly, consisting of,the flange, piston tube, and cylinder tube, and the graphiter seals shall be visually inspected.
The procedures to be followed in ultraronic inspections shall be essentially as described in Section D.2. of "Addendun No.1 to the Report on Dresden Control Rod Drive Modificatien dated February 20, t
- ic61, 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
?cisen Blades A mini:.2m of six poison blades shall be removed frc= the reactor and inspected. Inspection shall censist of thorough visual examinaticn fer st: ctural defects and measurement for changes in shape or dimension.
-1
\\
2 ^
~
3.
Core Grid St,nicture The inspection will consist of==4nntion of all previcusly located cracks; examination of fillet welds on the reverse side of beams opposite cracks; examination of a sufficiently larEe sample of pre-viously examined welds to give reliable indication of the integrity of the core structure as a whole, including a rin4=m of 15 of the beam to ring welds previously examined.
Date of Issuance:
All6 6 2 02 D i l
i l
6.v.t a. 11 SAFEfY SYSTai Externni Sensors Scram Sc ram (2)
Other Automatic fiumber loincidence Type in each in each Trip Setting Trip Point Functions Remarks Channel Channel Performed liigh Sphere Clones isolation fio bypasses P re s r.u re 2
I out of 2 Maximum Pressure Setting d valves & ventila'-
of 2.0 psig 0.2 psig_
tion ducts Iow Water 2
1 out of 2 -
At level which is Setting [
Closes isolation Bypas
' in " Shutdown Level in a minimum of'43"-
1" valves & ventilaj-position prior to initial Reactor above the top of tion ducts power operation only.
Vessel the active fuel
' Bypass is to be removed.
Closure of 2
1 out of 2 When oil pressure Setting [ 5 Closes ventila-Bypassed in " Shutdown" or Turbine Stop controlling these psig tion-ducts.
" Refuel" positions.
& Bypass valves drops to a
,',,(
Valves minimum of 50 psig Closure of 2(b) 2 out of 2 Closure of both Setting [
Closes ve.itila-Bypassed in " Shutdown" or Primary Stean (one from valves beyond a 5% stroke tion ducts and
" Refuel" positions.
Sphere Isola-each valve) maximum of 25% of starts emergency tion Valves stroke cooling Low water 2
1 out of 2 At level which is a Setting 11" Closes ventila-Bypassed in " Shutdown" or Level in maximum of 12" below tion ducts
" Refuel" positions.
Primary Steam Drum the drum center line Low Conden-2(c) 1 At minimum condenser Setting [
Closes ventila-Bypassed in " Shutdown" or ser Vacuum vacuum of 22" lig 0.25" tion ducts
" Refuel" positions.
By-passed in " Start" position if reactor pressure is below 1
At minimum condenser Setting'[
2CO-psig & condenser vacuum vacuum of 23" Ifg 0.25" is greater than 10" of lig.
Ihgh Reactor 2
1 out of 2 At maximum reactor Setting d Closes ventila-Bypassed in " Shutdown" Prensure pressure of 1050 psig 10 psig Lion ducts &
- posttion, coognge emergancy sta m
W "
TABLE 1 SAFETY SYSTEM (Continued)
~
-^-
-~'
~En tu rn a'1' Sen so rs a;
Scram Scram Gi)
Other Automatic
~
Type
' Number Coincidence in each in each Trip Setting Trip Point Functions Remarks Performed Channel Channel High 1.evel 2
1 out of 2 At tank level Setting 7{
1" Closes ventila-Bypassed in " Shutdown" in Scram which is 4'4)"'
tion ducts position Manual bypass 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 scrao ;
dump Lank to accommod 3
the water f rom 2.7 se...ms I
iligh Neutron 3
1 out of 3 When Icakage Setting f 3%
Closes ventil a-Bypassed its " Shutdown.'.' pos Flux or 1 out of flux indicates of rated power tion ducts ition.
Interlocked in 2 if 1 is a maximum of
" Refuel" position so ay, bypassed 120% r 'ed power require reduction in trip setting to a maximum setting of l'fe,'
rated power. At any t 6 c" other than in the " shut down" position only oma of the six flux trips can be bypassed.
Short Period 3(
2 out of 3 At minimum Period Setting d 0.5 Closes ventila-Eypassed in "Shutdos:
,f 3
of 4 seconds seconds tion ducts and "Run" posi tions.
a (a) The point at which a scram is actually initiated may be dif ferent f rom the 'hcram trip setting" by th e amount ot the tolerance for instrument inaccuracies. The amount of this tolerance 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.
g (c) There are two vacuum sensors.
Each sensor has one contact in each of the two safety channel.
(d).Thers.are three period sensing chambers.
Each chamber has two contacts in each of the two safety channels.
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3 APTE' DIX "B" 3
COMMONWEALTH EDISON COMPANY FACILITY LICENSE Estimated Schedule of Transfers of Special Nuclear Material from the Comission to Comonwealth Edison and to the Commission from Commonwealth Edison Returns by Common-Date of Transfers from wealth ErHson to Transfer AEC to Common-AEC Net Yearly Cumilative (Fiscal wealth Edison Cold Irradiated Distribution Distribution Year)
Kilograms U-235 Kilograms U-235 Kilograms U-235 Kilograms U-235 Thru 1962 1,8h2.1 362.0
-o-1,h80.1 1,h80.1 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.o 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
-o-
-o-292.0 2,018.9 1969 292.0 30.8 120.5 1ho.7 2,159.6 1970 -o-2,159.6 1971 292.0 30.8 109.5 151.7 2,311.3 1972 292.0 30.8 109.5 151.7 2,h63.0 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,o69.8 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-
-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 292.0 30.8 109.5 151.7 h,890.2 1991 292.0 30.8 109.5 151.7 5,ohl.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
-o- -o-5,h97.0 1996 292.0 30.8 109.5 151.7 5,6h8.7 Total lo,16h.1 1,239.8 3,275.6 5,6h8.7