ML19340B420
| ML19340B420 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 06/09/1961 |
| From: | Kirk R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML19340B416 | List: |
| References | |
| NUDOCS 8010230732 | |
| Download: ML19340B420 (27) | |
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SNM1 ATOMIC ENERGY COMMISSION WASHINGTON 25. D. C.
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%ne COMM0tWEALTH EDISON COMPANY DOCKET NO. 50-10 A?ENDMDIT TO FACILITT LICENSE License No. DPR-2, as amended Appendix #A", which constitutes the technical specifications for License No. DPR-2, as amended, is revised in its entirety as set forth below. Tne license authorizes Com:nonwealth Edison Con:pany to possess and operate the dual cycle, boiling water reactor designated as the "Dresden Nuclear Power Station" and located in Gntndy County, Illinois. The amended license shall expire on May h, 1996.
FOR THE ATOMIC E E T COMMISSION 6
t R. L 10rt
'ge g Director Division of Licensing and Regulation Attach.ent:
Appendix #A" Dated at Ge:=.antown, FAmyland this day of
, 1961 Ah 9 1961 80102s o 7p g
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Appendix "A2 to DPR-2, as amended A.
INTROD'CTION J
The following are the principal design and performance specifications and operating limits and procedures of the Dresden Nuclear Power Station pertaining to safety.
Sections B and C set forth the design and performance specifications and operating limits and principles.
Sections D, E, and F specify the basis for the initial loading and critical testing operations, and the limitations to be observed during start-up, power operation, and refueling and maintenance operations. In these sections, as well as~ in Section B, where maximm or minimm limits are not given specifically, the values given are " design" values which are subject to normal manufacturing and other tolerances.
Sections G and H provide certain additional operating and testing pro-cedures applicable to the control rod drive mechanisms.Section I provides wini m m requirements for certain inspections of the control rod drives, poison blades and core grid structure.
B.
DESIGN FEATUPIS 1.
Reactor Vessel J
The reactor vessel is a vertical cylindrical pressure vessel, with dished top and bottom heads, made of 1cw alloy steel and clad inside with stainless steel. The vessel was 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 h0 ft 9-5/8 in Inside diameter 12 ft 2 in Design pressure 1250 psig 0
Design temperature 650 F 2
Nuclear Core I
a.
Dresden Core I Maximum core diameter (circum-scribed circle) 129 in Maximm active cold fuel lengths 107-1/8in Maxi =um number of fuel assemblies h68 1
b.
Dresden Core I-Modified Maximm core diameter (circum-scribed circle) 129 in
May4=m active fuel length - cold 111-15/16$n Maximm number of fuel assemblies by types Type I h88 Type II 2
Type FF-1 through FF-12 (one each) 12 For fuel type number definitions, see Table II Maximm number of all fuel types hCC hel assemblies of Type II or Type FF-1 through FF-12 may be located in any position in the reactor, provided each such assembly is separated from any other such assembly by at least four Type I fuel assemblies.
Fuel assemblies of Type I will occupy the remaining positicas to corglete the co.a configuration.
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 maximm number and location are within the limits specified above.
3.
Fuel a.
D esden Core 2 tuel Each fuel assembly consists of 36 vertical fuel rods (except for cecasional special assemblies such as instrument bearing assemblies which will have 3h or 35 fuel rods plus the thimbles of approximately the same dimension of the clad fuel rods),
each of which is made up of solid cylindrical pellets of uranium j
dicxide, enriched in uranium-235 to a maximum of 1.5%, and clad in Zircaloy-2 Each fuel rod is composed of four separately clad segments. Pertinent fuel design parameters in the cold condition are:
Fuel pellet diameter 0.h98 in Fuel pellet length Regular fuel pellets 0.625 in Segment end pellets 0.500 in Outside diameter of cladding 0.567 in Cladding wall thickness 0.030 in Fuel pellet density averaged over a fuel segment Minimm 9h% of theoretical Crcss-sectional center-to-center distance between fuel rods 0.710 in.
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L b.
Dresden Core I-Modified - Types II and PF-1 through PF-12 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 i
assembly. Several assemblies, however, have been designed such that rods may be replaced with instrumentation.
H The minimim fuel pellet density averaged over a fuel segment is 2
9h% of theoretical for all fuel assemblies except PF-7, PF-8, and PF-9 which is 90% of theoretical.
- h. Control Rods and Drives The control blades, 80 in number, consist of small vertical stainless-steel tubes filled with compacted toren carbide (B C) powder. The h
boron carbide powder is separated longitur' inn 11_y into several independent compartments. The tube walls are designed to withstand the mav4==
calculated internal gas pressure. The tubes are held in a 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.
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:
a.
The accumlator pressure, which will be available and effective.
under~ all' conditien's of ' reactor ope' rat' ion except " Shutdown" (at which time interlocks prevent any rod withdrawal), and
-b.
The reactor' pressure, if this is greater than about 700 psig and accumlator pressure falls below reactor pressure.
The drives are mounted on the bottom of the. reactor vessel, and.
withdraw the control rods below the core. Upward =cvement of the control rods, into the core, decreases reactivity. The inserted position of the control rods is determined by a locking device which provides 12 discrete approximately ecuidistant 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 6in/see Maximum time from receipt of scram r,ignal to:
10% control rod travel 0.6 see 90% control rod travel 2.5 see l
Maxima numb;r of control rods per accumulator 3 non-adjacent rods Frequent and thorough periodic checks will be made to assure the j
proper functioning of the control rod drive system.
a
- 5. lie 2id Poisen System System actuation Manual Mar 4== time after actuation until poison begins to enter core 20 sec.
Mini = = weight of boron in system (present as sodium pentaborate in solution) h00 Jbs.
Fini== poison worth for operations with reactor vessel head on 0.15 k 4
Minimm poison worth for operations with reactor vessel head off 0.0hjk 6
Steam Surely System Besides the reactor itself, the steam supply system comprises a
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steam separating drum, four secondary steam generators and recirculating pumps, an emergency condenser, unloading heat exchangers, and the necessary pip hg and accesnories for these components. The plant =ay 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 ircosed.
The emergency condenser consists of two separate tube bundles of eqac1 capacity in a co: mon shell. Each tube bundle can be started automatically by the appropriate reactor safety system 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 F4 ni - m capacity of emergency condenser 6% of rated reactor thermal power
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Fini= m cooling water stored in emergency condenser 30,000 gal.
Minimun emergency coadenser 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 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.
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 0
a marim m of about 300 F.
In addition to the condenser vacuun scram and isolation trips indicated 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 fall to or below 7 inches of Eg. This vacuum trip is not a part of the reactor safety system, tut becomes cperative whenever the condenser vacuum has been increased to above the trip set point.
- 8. Waste Disnosal Systems a.
Solid Wastes i
Solid wastes containing radioactive materials include filters, defective equipment, and other miscellaneous trash. Such material will be stored in accordance with AEC regulations (10 CFR-Part 20) which may involve storage in an underground concrete storage vault at the site. Wnen feasible such material may be compacted before storage. Spent contaminated resins I
are sluiced to an underground tank for indefinite storage.
Sluice water is decanted after the resins have settled.
b.
Licuid Wastes Equipment is provided to treat radioactive liquid wastes by decay in storage, long-term underg ound 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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uccd to d;termin) whether permissible limits can be met for
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any liqaid process wastes to be released. No disposal of these wastes to ground will be made.
The large quantities of river water used for equipment 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 liquid wastes to below permissible limits for release to the river. The release of liquid wastes shall conform to the provisions of 10 CFR, 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 monitor automatically initiates closing of the dischargevalveontheagejectorholdupgystemifthemeasured rate of discharge of Ie exceeds 2 x 10 uc per second, the measurement being made after two minutes decay (travel tine in the system from the reactor core to point of measurement). The reactor will be manually shutdown whpn the stack discharge rate of noble fission gases exceeds 7 x los uc per sec.
d.
General Co=monwealth 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 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-out (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 maximm 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 scram. 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 available to the operator also, and an additional manual control scrams the reactor and closes the sphere isolation valves.,
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Th; -4== loc 1 hest flux is given in Item 3 of Section E, 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 regaired or desirable under some conditions. The bypasses for the external system are indicated in the " Remarks" colu m o# Table I.
The in-core local power monitoring system may be bypassed in the " shutdown" position and the " refuel" position. The four positions for the selector switch are:
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 evclosure while. testing or maintaining the safety system with the reactor shutdown. Control rods cannot be with-drawn in this position.
- 10. Radiation-Type Reacter Instrumentation In addition to the external and internal core neutron flux channels and the period channels that are a part of the reactor safety system, there are provided two indicating startup channels and a battery-operated neutron flux channel.
- 11. Reactor % closure The enclosure housing the reactor and +".e steam supply system is a spherical steel shell,190 feet in diameter: with the egaator approximately $6 feet above ground level. Irem the encicsure, the minimum off-site distance is one-half mile tor a.
Skinner Island in the Kankakee River; b.
The navigation channel in the Illinois River; and c.
The land boundaries of the site..
i 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 c'esign pressure) 3250 F Maximm Wind Velocity 110 mph Horizontal Acceleration 3.3% gravity Maximm Ieakage Rate at 37 psig 0.5%/ day All nomally open lines penetrating the enclosure shell through which leakage could credibly occur in the event of the "maximm j
cre.lible 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 e t closed from any scram signal, will close automatically from low reactor pressure -- a condition that would be enccuntered in any system zupture approaching the severity of the postulated "Mmm credible accident." All the primary isciation 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 during operation, or in p/or operating rules.otentially hazardous non-eperating situations, by interlocks and The enclosure is provided with a post-incident cooling system, for
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use in the event of a serious primary system rapture, to aid in the reduction of enclosure pressare and leakage. This s designedforaheatremovalcapacitycfatleast30x10gstemisBtu/hr at an enclosure internal tegerature of 256 F.
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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 manually initiate operation of this system or override the autcmatic 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 functioning of the post-incident cooling system.
At er before the time of the first inspection of the mcdified control rod drives, but not later than October 15, 1961, a leakage test of the enclosure will be conda ted at a low pressure (less than 10 psi) to determine the reliability of penetrations. This test shall include 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 throt.gh the main steam line shutoff valves.
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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 uhielded from all power plant equipment and from the reactor enclosurt so as to be tenable even in the event of the "maximm credible accident". The minimm 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 followsr a.
Before being placed into regular service, the plant will be proved out in a comprehensive controlled test and initial operation program, as indicated in Sections D and E.1, below.
b.
Operation and control of most of the power plant equipment will be centralized in the control room.
c.
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.
Daring this time one licensed operator will man the control room, d.
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, e.
Startup, normal shutdown, and all other repetitive operations will be performed in accordance with specific check lists.
f.
Maintenance of much of the equipment outside the reactor shielding may be undertaken by contact methods and without overall plant shutdown. Plant shutdown and semi-remote methods will be employed as necessary, g.
All tests and routine maintenance of protective devices and power plant ecuipment will be done in accordance with prescribed schedules.
h.
Radiation monitoring by fixed or portable instrumentation will be provided for entry to all radiation zones.
- i. All personnel leaving a contaminated radiation zone, and equipment being removed from such zones, will be appropriately surveyed to assure control of contamination.,
- j. Irradiated fuel is to be moved from the reactor to storage,-under water, by semi-remote methods.
I k.
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 insertion or removal of fuel assemblies in the core or withdrawal of control rods when the reactor head is off.
I 1.
Operation of the radioactive waste-handling system will be done in such a manner that it will be unHkely 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 per:M ssible limits.
i m.
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.
n.
All significant unexpected incidents, unsafe acts, or incidents of excessive exposure to radiation will be investigated to effect procedures to prevent recurrence.
o.
In the event of any situation which may compromise the safety of g
continued operation, it will be,he required procedure to shut the plant dcwn as quickly as the situation calls for, and to take other planned energency actions to protect persons and property.
D.
INITIAL IDADim AND CRITICAL TESTING 1.
General On successful completion of all pre-cperational checks necessary for safe loading, the reactor will be loaded to initial critical i
and then to the proper loading for power operation. Between these two loading steps a number of tests will be m n and measurements made to check the calculated nuclear parameters. This phase will be conducted at atmospheric pressure with the reactor vessel lid removed, and at approximately ambient temperature. Items which will receive particular. attention during this phase of operation, or that will be carried cut in a ranner different from norral refueling, are discussed below.
2 Instrumentation A minimum of one neutron-sensitive chamber per grid quadrant will be installed in the reactor before the initial loading starts. A neutron source of sufficient strength to provide appreciable readings j
at all times on the neutron-sensitive chambers will be installed, and H
the chamber-source geometric relationships will be maintained as the l l
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loading progresses so that the chamber sensitivity will always be sufficient to measure the neutron mitiplication. A rinimum of four neutron-sensitive chambers will be tied into the reactor ~ safety system in such a way that tripping of any channel will cause a scram.
k ring this period with the head removed the reactor safety system in-core monitors will not be in place.
The initial loading and critical testing is to be done with the safety system selector switch in the " refuel" position (see item B.9).
This indicates that the following scram sensors will be in operation:
i a.
High sphere pressure b.
Iow water level in reactor (This sensor may be bypassed in the
" refuel" position until initial power operation.)
c.
High reactor pressure (This sensor is in service in the " refuel" position because there is no reason to byr ss it. With the reactor head off, of course, it cannot be effective.)
d.
High level in scram dump tank e.
High neutron flur (For the initial loading operations, the normal out-of-core chambers and dual-channel system are replaced by the special in-core chamber system described above i
in this item 2 As with the normal system, however, with the selector switch in the " refuel" position, the neutron flux trip setting mast be reduced to a :rn4mm setting of 10-3 rated power.)
f.
Short period.
3.
Reactor Water Ievel Prior to the initial loading, the reactor vessel will be filled with water to a minimm of three feet above the active height of the core. As indicated in D. 2, the low water level sensor may be bypassed during the initial loading and test period when the water l
1s required primarily as a moderator.
h.
" Stuck Rod" Criterien: At every stage daring the initial loading, and in the fully loaded configuration, the control rods must provide a shutdown control margin of at least 0.033 with any rod wholly out of the core and completely unavailable.
b.
" Cocked Rod" Criterion: During core alterations after the first fuel cell is loaded, the reactor mast be suberitical by at least 0.01Ak with at least one centrol rod fully withdrawn in the regien of the alteration and available for rapid scram 4
insertien. i
- 5. Minimm critical core The following criteria will be used in loading to the minimm critical size core:
a.
The selected core area vill be within the control rod pattern.
b.
Except for the final fuel increment, the size of each fuel addition will never exceed one-half the estimated amount to reach criticality. This estimate will be made using neutron mitiplication measurements determined between fuel increment additions. The final fuel increment will not exceed 4
a reactivity worth of 0.005ak.
c.
The shut down margin criteria given in item D.h, will be applied.
The following tests will be conducted, as a minimm, with the minimm critical core:
d.
Measurements of the strength of some of the control rods encompassed by or adjacent to the fuel loading; Temperature coefficient (gk/oF change in moderator I
e.
temperature);
f.
Voidcoefficient(ak/ !o change in voids).
The void coefficient averaged over the interior of a fuel channel will always' be negative when the core is critical.
Sufficient measurements will be made to conclude that the void coefficient, measured as indicated in the sentence above, is negative in all fuel channels.
6 Twice Minimm Critical Core On successful completion of item D.5, the core will be loaded to approximately twice the minimm critical size, but not beyond the control rod outer periphery. The following measurements will be made, as a minimm, with this core:
a.
Selected control rod calibrations; b.
Rod configurations for critical; c.
Temperature coefficient; d.
Void coefficient. This m st meet the limitations indicated in item D.5.f. i
7.. Core Ioaded to Feriphery e* Control Rod Pattern On successful completion of item D.6, the core will be loaded ~to encompass all control rod cells (i.e., one row of fuel assemblies outside of control rod periphery). Criticality checks will be made i
periodically during this loading. Calculations and extrapolations of measurements made in accordance with items D.5 and D.6 win be used, prior to the actual fuel loading, to indicate whether the shuttown margin and void coefficient criteria can be met. Measurements will be made with this core, as a minimm, of the critical rod con-figurations with non-uniform distributien of withdrawn centrol rods.
8 completed Core Loading Calculations and test data from item D.7 win be used to determine the = ri= = nu ser of fuel assemblies permitted in the area outside the control rod periphery in the fully loaded core. Following this determination, approximately half the number of fuel elements win
>e added uniformly around the core. The determination will then be
'epeated and followed by another " half loading." This sequence will
>e repeated until the last fuel assembly is inserted. The completed
- ore will meet, as a minimum, the shutdown margin requirements indicated in item D.h.
The test program for the completed core win include, as a minsium, the following items:
a.
Critical rod configurations with non-uniform distribution of withdrawn control rods; b.
Temperature coefficient; c.
Void coefficient. This must meet the limitations indicated in item D.5.f.
E.
PNER OPE *JLTION 1.
Power Test Program a.
Dresden Core I Following successful completion of the initial loading and critical testing program, the power test program will be carried out. This program will include the following:
1 A series of tests at low power, and low to rated pressure, with studies of stability, reactivity, and poner dis-trilution as they are affected by variations in reactor pressure, te=perature, voids, control' rod positions, and transientpoisons;
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11 A series of tests at various power levels up to rated power, at low to rated pressure, with studies similar to those indicated in item E.1.a.i; iii Studies of power transient effects, simulated eqaipnent failures, and process mishaps; iv Radiation surveys.
Tnis testing program will,be conducted at stepwise increasing power levels. At each power level, experiments will be conducted to predict the power level or other operating condition of the reactor at which maximam heat flux, reactor instability, or any other reactor design or operational limitation would be reached.
The reactor power level or other limiting condition will not be increased in the next step more than halfway from the test conditions to the predicted limitation, provided that for reactor power levels in excess of 50% of rated power no step increase in power shall exceed 60 ri, b.
Dresden Core I-Modified The power testing for the first core as modified by the sub-stitution of 2 Type II fuel assemblies and 12 experimental fuel assemblies will be limited to careful observation of reactivity, power distribution, and stability as indicated by the reactor instiamentation daring the normal gradual approach to power.
2 Safety System Scram Settings i
Operative scram sensors and their settings with the safety system i
selecter switch in the " Start" and "Run" positions are given in item B.9.
The control rods cannot be withdrawn in the " Start" position until at least 10 inches of Hg condenser vacuum have been 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. Tne reactor will also scram if a reactor pressure of 200 psig is exceeded prior to obtaining at least 23 inches of Eg condenser vacaum.
3.
Determinaticn of Maxi:nm Reactor Power a.
Dresden Core I The " rated" er cperational thermal power of the reactor.is not, at least during initial pcwer operation, a fixed value. Lower than egailibrium fuel cycle power production may be necessary initially, due to the possibilities oft 1 Less than full lead of h88 fuel assemblies in order to meet shutdewn margin criteria; and i
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O 11 Higher peaking factors initially than at egallibrium fuel
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cycle.
The er4*m reactor power is, conseqaently, defined as that therral power at which the rWmm heat flux for any fuel rod is rea*hed. This maximam heat flux will never exceed 350,000 I
Btu /ft -hr, based on calculations and experimental data. The rated heat flux and resulting rated reactor power are then set to 80% of their maximam values and, as indicated in items B.9, the high neutron flux scram setting will be no higher than an indicated 120% of the rated reactor power. However, in no case will the high neutron flux setting be allowed to exceed an indicated reactor thermal power of 782 W (125% of the planned operational power of the fully loaded core).
Within the limitations on reactor power and heat flux set forth in the paragraph above, the reactor will be operated in such fashion as to:
i Maintain at all times a burnout heat flux margin of at least 2 at the point closest to burnout in the hottest channel in the core based on a uniform steam quality over the cross-section of the channel; and 11 Be always well within the bounds of stability, as evidenced by the operation itself and any experimental data produced daring the " Power Test Programa phase of operation.
b.
Dresden Core I-Modified The =rimm reactor power for Dresden Core I-Modified is defined therefore as that thermal power at which the maxirum heat flur 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 hh5,000 Fuel Type PF-1 through PF-h h25,000 Fuel Type PF-5 through 9 h15,000 Fael Type PF-10, 11 and 12 h75,000 The peak rated heat flux and resulting rated reactor power are then set to 80% of their maxirum values and, as indicated in item B.9, the high neutror. flux scram setting will be no higher than an indicated 120% of the rated reactor power. However, in no case will the high neutron flux setting be allowed to exceed an indicated reactor thermal power of 782 W (125% of the planned operational power of the fully loaded core.) l i
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I The reactor win b2 cperated within tha cbova 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.
h.
Pressure Limits a.
Wrimm normal reactor operating pressure 1000 psig b.
Maximm pressure setting for automatic reactor shutdown 1050 psig c.
W ri m m pressure setting for opening of electromatic relief valves 1085 psig d.
Marimm pressure setting for opening of first main safety valve 1205 psig e.
Marimm safety valve pressure setting 1250 psig f.
Combined capacity of safety valves At least 150%
rated primary steam flow
- 5. Reactivity Limits a.
The average reactivity addition rate from withdrawal of the control rod with the marim m reactivity worth in the most adverse withdrawal pattern will not exceed 0.0029Ak/sec.
b.
With the reactor in any condition, the shutdown margin criterion in item D.h.a must be met.
c.
With the reactor in the hot operating conditon, operation will not be continued when the reactivity worth cf the control rods known to be stuck out of the core, or otherwise unavailable for control, exceed s half the value at which hot shutdown could not be accomplishec. (It is expected that this would require shut-down 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.Oljk, the liquid poison will be introduced prior to reaching the " cold" condition where this criterion could not be satisfied.
d.
The void coefficient of reactivity will meet the limitations indicated in item D.5.f. !
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The moderator temperature coefficient of reactivity will change from positive to negative at a temperature below the mwhm reactor temperature that can be attained under atmospheric conditions. Measurements will be made to confirm this.
- 6. Waste Discosal 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 manner such that it is unlikely any person will receive radiation exposures in excess of the approximate permissible lini.ts.
F.
REFTIELING 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 c, e, f, g, h, i, j, k, and m.
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.
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 dene. Even with the reactor in the shutdown condition, however, any maintenance work involving the removal of control rods fro = 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 satisiv '.he " stuck rod" criteria discussed in items E.5.b and E.5.c.
During movement of fuel in the core, or control rod maintenar e, the minimm shutdown margin will satisfy the " cocked rod" criterion given in iten D.h.b.
h.
Licuid Pcison System The liquid poison system will,e operative during refueling and maintenance operations in the reactor as well as during normal power operations. :
G.
TESTS AND INSPECTIONS OF CONTROL RCD IRIVES 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 recuirement.
Additional testing shan be performed as may be necessary to gather significant data concerning the activity being investigated.
1.
Normal Drive Oceration These tests shall be made with control blades properly attached to
'their respective drive mechanisms.
Each drive shall be exercised through the fu n length of the drive ctroke 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 ranner during every period of shutdown which is expected to exceed h6 hours, and in any event no less frecuently than once every quarter.
All mechanisms shall be subject,ed to a friction test no less frequently than once every quarter. A test pressure which permits a sensitive measurement of fricticnal forces in the drive shall be used. The uniformity of motion +hrough a full upward strcke and fluctuations in pressure shall observed.
2 Scram Ooeration 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 during eve y period of sh2tdown which is expected to exceed h6 hours, and in any event no less frequently than once every quarter. i
. M.'
~
3.
T::sts Prior to R; installation of Rector Hrd 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.
H.
OPERATING FROCEDURES AND IlMITATIONS Operating procedures and limitations shall be in a.cordance with the following requirements:
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 writing to the Commission the incidence of any apparent drive mal-function which requires suspension of reactor operation in order to carry out the provisions of this cection.
In case of any observation of anomalous behavior of any drive, there shall be prc:pt and thorough investigation to determine the 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 contirraed, or resumed after shutdown, only if it has been determined that the anomaleus behavior observed in a particular mechanism does not impair the ability to control the reactor or indicate that i=pairment of the performance of other mechanisms may be imminent.
Operation may be continued, or resumed after shutdown, with a defective control mechanism which has been deactivated so as to lock its con-trol blade in place, provided that (1) the ' stuck rod" criteria of the license can be met, (2) it has been determined that the inactive 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 imminent.
2 Nuclear Indications of Control Blade Position a.
The operator sball be provided with a defined rod withdrawal sequence and e. predicted critical rod configuration for each start-up. The operator shall follow this withdrawal sequence.
b.
During start-up the motion of poison blades shall be verif3ed, insofar as possible, by observing the response of the external instrumentation. !
I
Qf 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 1L K shall be withdrawn or remain in a withdrawn position.
d.
Up to one rod or 1.0$K, whichever is more restrictive, nay be rithdrawn in addition to the predicted critical centrol 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 bo inserted. Thereafter, all blades with-drawn aust be verified.
e.
All blades which were not verified for following daring the start-up rod withdrawals will be veri #ied as soon as possible after criticality is achieved. Any blades which cannot be verified for following shall be inserted. In the event that the verification tests daring this operation do not show blade fellow-ing or separation, then additicnal verification tests shall be conducted 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.
Daring 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 operaticnr
- 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 operations with K effective greater than 0.995.
I.
FROJECTED INSPECTIONS OF CONTROL ROD DRIVE IECFJJISMS, POISON EIADES AND CORE GRID STRUCTURE The inspections described in "1" and $2" shall be perfcrmed (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 resur:ption of operation after shutdown in November M O: and (b) after h500 and before 9000 further hours of
33 reactor operntion et such pr:scura subsecuent 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 =inimim of six drive nechanisms 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 operat. en given indication of possible defects shall be removed and inspected to the extent necessary to determine whether or not sach defects exist.
The drives shall be disassembled to per=it (a) visual inspection aided by han( lens =agnification (5 - 10I), (b) fluid penetrant inspection, and {cy ultrasonic inspection of the following parts:
Index tube and piston head assembly.
Collet lock assembly (guide sleeve, retainer, and collet finger).
Guide Plug.
Shuttle piston.
The roller mountressembly 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 ultrasonic inspections shall be 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 exami ed 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 and inspected. Inspection shall consist of thorough visual examination for structural defects and measurement for changes in shape or dimension.
4*
i
.e S
3.
Core Grid Stmeture 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.stmcture as a whole, including a riniwem of 15 of the beam to ring welds previously examined.
Date of Issuance:
JUN 9 1931 l
TAlill[ t SAFFifY SYSTD1 External Sensors Scram Scram 6)
Other Automatic Number Joincidence Type in each in each Trip Setting Trip Point Functions Remarks Channel Channel Performed liigh Sphere Closes isolation No bypasses Pressure 2
1 out of 2 flaximum Pressure Setting d valves & ventilnL of 2.0 psig 0.2 psig tion ducts Low Water 2
1 out of 2 At level which is Setting [
Closes isolation Bypassed in " Shutdown Level in a minimum of 43" 1"
valves & ventila-position prior to initial Renctor 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, f
& 15ypann valves drops to a y
Valves rrinicum 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
- 57. s t roke tion ducts and
" Refuel" positions.
Sphere isola-each valve) maximum of 257,of starts emergency tion Valves stroke cooling Y
Low water 2
1 out of 2 At Icvel which is a Setting [1" Closes vent la-Bypassed in " Shutdown" or (d
Level in maximum of 12" below tion ducts
" Refuel" positions.
g Primary the drum center line Steam Drum O
Low Conden-2(c) 1 At minimum condenser Setting [
Closes ventila-Bypassed in " Shutdown" or nrr Vacuum vacuum of 22" Ifg 0.25" tion ducts
" Refuel" positions.
By-passed in " Start" position
{c q J if reactor pressure is beloo 1
At minimum condenser Setting [
200 psig & condenser vacuum vacuum of 23" IIg 0.25" is greater than 10" of Ilg.
b liigh Reactor 2
1 out of 2 At maximum reactor Setting [
Closes ventila-Bypassed in " Shutdown'.'
Pressure pressure of 1050 psig 10 psig tion ducts &
position.
gg{gmergiincy A
=
TABLE 1 SAFETY SYSTFJi (Continued)
F.xternal Sensors
~
Scram Scram (a )
Other Au'omatic Type Number Coincidence in each in each Trip Setting Trip Point Functions Remarks Performed Channel Channel Iligh Level 2
1 out of 2 At tank level Setting f 1"
Closes ventila-Bypassed in " Shutdown" in Scram which is 4'4)"
tion ducts position F!anual bypass Doinp 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 accommodate the water from 2.7 scrams, liigh Neutron 3
1 out of 3 When leakage Settir:g f 37.
Closes ventila-Bypassed in " Shutdown" pos t 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 as td ) '
bypassed 1207 rated power require reduction in trip setting to a 3
maximum setting of 1(f rated power. At any time other than in the " shut-down" position only one of the six flux trips can be bypassed.
C 2 out of 3 At minimum Period Setting d 0.5 Closes ventila-Bypassed in " Shutdown" Short Pe ri od 3
of 4 seconds seconds Lion ducts and "Run" positions.
U[) The point at which a scram is actually initiated may be dif ferent from the "sc ram trip setting" by th e amount of the tolerance for instrument inaccuracies. The amount of this tolerance is indicated by the values given in this column.
(b)
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) There are three period sensing chambers.
Each chamber has two contacts in each of the two safety channels.
i l
TABLE II CLADDING REGULAR E005 CORHER RODS Oet M.
Hnba
- Fue, 7.UO2 Pdlet
- Hude, F.I
- 00 Twel Type Mei,elel 17 ell 11 tid ness Configurotten g
p,g,,
Dio.eter Reps,ed Cempositten En,Ichment Oscueen Req. ired Co-postil.a Eadclw.c.:
Die =<en
~
2 2.5 0.379 9
2.1% UO2 93.5 0.399 11 30455 0.441 O.019 7x7 40 Ud 97.9% Tl102 UO 2.4% UO7 PFI 30455 0.430I 0.011 6x6 29 2
2.6 0.457 7
97.25% T!!92 93.5 0.(
0.35% Ery0_3 0.35% Et20_3 PF2 30455 0.."'O O.011' 6
6-29 M
OE8 7
0.35% E 20 3 0
E 3 f
2.3% UOz FF3 30455 0.480 0.011 6x6 29
.U02 2.6 0.458 7
97.7 T1102 g.
2.4 u02 U02 2.6 0,457 7
PF4 30455 0.430 0.011 6x6 29.
97.6 TH02 2.4 U O2 PFS.6 30455 0.439 0.017 6x6 29 UO2 2.6 0.455 7
93.5 0.455 97.6 TH02 P F 7, 8, 9 30455 0.439 0.017 t
6x6 29 U02
..'6 0.455 7
UO2 1.9 0.455
'.35". E 0 PF10 Zr4 0.412
.0'.025 8.x 8 55 M
ED.
I O
2 3 0.35 E 203
~"'
1.8% UO2 2
2.0 0.353 9
98.2% TH02 l
PFil,12 Zr4 0.412 0.025 8x8 55' UO mm O
,.e m e 4
. =.
t
..w=
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