ML20030D423
| ML20030D423 | |
| Person / Time | |
|---|---|
| Site: | 05000087 |
| Issue date: | 01/28/1972 |
| From: | US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML20030D417 | List: |
| References | |
| NUDOCS 8109010371 | |
| Download: ML20030D423 (31) | |
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APPENDIX A FACILI2T LICENSE NO. R-Il9 TECHNICAL SPECIFICATIONS TCk THE, WESTINGHOUSE NUCLF>R TRAINING REACTOR DOCKET No. 50-87 Date of Issuance: January 28, 1972 l
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O 8109010371 810825 PDR ADOCK 05C10387 P
TABLE OF CONTENTS P,,, age, 1.0 DEFINIT 10NS 1
2.6 SAFETY LI!!1TS AND LIMITING SAFETY SYSTEM SETTINGS 2
2.1 Safety Limits 2
2.2 Limiting Safety Systen Settings 3
3.0 LIMITING CONDITIONS FOR OPERATION 5
3.1 Reactor Control and Safety Systems 5
3.2 Reactor Farameters 8
3.3 Radiation Monitoring 9
4.0 EXPERIMENTS 10 5.0 SURVEILLANCE REQUIREMENTS 12 5.1 Reactor Control and Safety 12 5.2 Reactor Parameters 13 5.3 Radiation Monitoring 14 6.0 DESIGN FEATURES 15 6.1 Site 15 6.2 Facility 16 6.3 Reactor Room 16 6.4 Reactor 16 6.5 Water !!andling System 18 6.6 Fuel Storage 19 7.0 ADMINISTRATIVE CONTROLS 19 7.1 Organization 19 72 Personnel Requirements 19 7.3 Review and Audit 20 7.4 Operating Procedures 21 7.5 Action to be Taken in the Event of an Abnormal Occurrence 22 7.6 Action to be Taken in the Event a Safety Limit is Exceeded 23 7.7 Reporting Requirements 23 i
7.8 Records 25 (aT
, 1.0 DEFINITIONS The following terms are defined to aid in the uniform interpretation of the specifications.
1.1 Reactor Safety System - that combination of measuring channels that forms the autc matic protection system for the reactor or that provides information which requires manual protective action to be initiated.
1.2 Measuring Channel - an arrangement of components and modules as required to measure the value of a process variable. The output.of tne measuring channel is the measured value of the process variable and may be considered the true value within the accuracy of the measuring channel.
l.3 Saf ety Channel - an arrangement of components and modules as required to generate a single protective action signal when required by a facility condition.
1.4 Operable - when a system or component is capable of performing its required function in a normal manner (Operating means it is performing its function).
1.5 Channel Ched - a check to verify by intercomparison of channel outputs whether or not a measuring channel is operable.
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1.6 Channel Test - a test to verify, by introducing externally generated l
signals, that a measuring channel is operable.
l.7 Channel Calibration - adjusting a channel output such that it responds, I
with acceptable range and accuracy, to known values of the par 6uneter which the channe; measures. Calibration shall encompass the entire channel, including equipment actuation, alarm or trip.
consists of a gravity drop of control rods caused by the I.H Re.w tor Trip, of the electrical power to the magnet carriages. The trip interruptica can be initiated automatically by the safety system, manually by the manual reactor trip and manually by disconnecting the facility electrical power.
1.9 Aux 111ary Reactor Trip - consists of the dumping of the moderator-shield water through the ten-inch dump valve. The trip is initiated manually.
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i 1.10 Excess Reactivity - that reactivity above critical which may be added i
by manipulation of remote controls.
1.11 Reactor Secured - means that all control rods are in their down positions and the key is removed from the console lock, or when there is no fuel in the core.
1.12 Reactor Operational - means the reat;or is not secure'd.
1.13 Neutron Source - any neutron-emitting radioactive material, other than the reactor fuel, which is positioned in or near the reactor core to provide an external source of neutrons.
1.14 Experiment - any installed apparatus, device, or material within the confines of the reactor tank which is not a normal part of the assembly, or any core loading which is not the normal core loading.
3.15 Moderator-Shield Water - the water that is placed in the reactor tank.
i 1.16 Administrative Controls - the provisions related to organization and l
management, personnel requirements, procedures, record keeping, review and audit, and reporting that are considered necessary to assure oper-atlen of the facility in a safe manner.
1.17 Revi + and Approve - means that the reviewing group or person shall carry out a review of the matter in question and may then either approve nr 'lisapprove it.
Before it can be implemented, the matter in question i
must receive un approval from the reviewing group or person.
l.lH Rearti ly Available on Call - normally means within a 20-mile radius of the facility and that the operator-on-duty knows the location and t elephone number of the senior operator on duty.
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limits Applicability This specification applies to the reactor power level limitation and the annual integrated thermal power limitation.
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O Objective The purpose of this specification is to establish the upper safety limit on power level for the reactor and the integrated thermal power produced annually.
Specifications 1.
The maximum power level shall not exceed 20 kilowatts (thermsl).
2.
The integrated thermal power (thermal energy) for any calendar year shall not exceed 200 kilowatt hours.
liases A maximum power level of 20 kWt and an integrated thermal power of 200 kW-hr per year provide adequate flexibility for the performance of training and irradiation operations and at the same time appropriately limit the quantity of radioactive material available for release.
Calculations and measurements have shown that during steady-statt oper-ation at 20 kWt the average moderator temperature increase is less than 10*F, and therefore, the clad and fuel temperatures remain signifi-cantly below their failure point.
With the moderator-shield water level at a height of about 5 feet above the top of the core and the power level at 20 kWt, estimates based on shielding calculations and extrapolated experimental data show that the radiation level immediately outside the reactor room is below 30 mrem /hr l
and the radiation level at the controlled area boundary is less than 2 mrem /hr.
Access into the reactor room during reactor operation is l
l prohlhited administrative 1y and by a security interlock system.
l The maximum operating power level of the reactor is 10 kWt.
In general, the operating power level is kept as low as possible, consistent with the training operation requirements. and normally is less than one hundred l
watts.
1 2.2 Limiting Safety System Settings l
l Applicability This specification applies to the settings for instruments monitoring parameters associated with reactor safety limits.
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Objertive The purpose of this specification is to assure protective action before safety limits are exceeded.
Specifications The limiting safety system settings shall be as follows:
Maximum Power Level 12 kWt 2
Minimum Flux Level 2.5 neutron /cm -s Minimum Period 3 seconds Maximum Gamma-ray Exposure R/hr value experimentally
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Rate (above the water shield) determined for 12 kWt power level operation with normal moderator-shield water beight Bases g ).
The maximum power Icvel trip setting of 12 kWt is established by
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estimating an error of 20 percent in the absolute neutron flux measure-ment by activation methods and a maximum error of 35 percent in the non-1incarity of monitoring instruments. The safety margin prevailing between the safety limit and the limiting safety system setting is adequate to allow fer these errors.
The minimum flux level has been established to prevent a source-out startup.
The interlock set point on the source level chsanel is 2* cps.
l The specified minimum flux level will' assure that this interlock is I
satisfied.
The minimum 3-second period is.,
ied so that there is sufficient time for the automatic safety system to respond before the power level safety limit is exceeded. The transient would be terminated in less than 200 milliseconds after reactor trip.
The maximum gamma-ray exposure rate setting is established to correspond with the exposure rate above the top of the reactor shield that would occur during reactor operation at 12 kWt with the moderator-shield water level at the normal (5 foot) water height. This setting assures that increasing radiation levels in the vicinity of the reactor room will be detected before they become excessive when the reactor is operated at moderator-shield water heights other than the normal level.
An approximate value for this setting is estimated to be 500 mR/hr.
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O LU l.1MiTING CONDITIONS F_OR OPERATION 3.1 Reactor Control and Safety Systems Applicability These specifications apply to all methods of changing core reactivity available to the reactor operator.
Objective The purpose of these specifications is to assure that an adequate shut-down method is available and that positive reactivity insertion rates are within those analyzed in the Safety Analysis Report.
a Specifications 1.
There shall ae a minimum of three operable control rods. The maximum excess reactivity that can be loaded shall be such th.t the reactor shall be subcritical by a margin greater than l$ with the control rod having the largest reactivity worth fully witiidrawn.
2.
The maximum control rod and moderator-shield water reactivity insertion rate shall be less than 0.10$/s when k,gg is less than 0.99 and less than 0.035S/s wher k,gg is greater than 0.99.
3.
The total control rod drop time for each control rod from its full-out position to its full-in position shall be less than or equal l
to 1.2 seconds.
This time shall include a maximum magnet carriage release time of 0.125 second.
4.
Hegative reactivity shall be available in operable cocked control i
rods prior to adding the moderator-shield water to the reactor. At least l$ of negative reactivity shall be available when core loadings, l
capable of becoming critical, are to be filled with the moderator-shleid water.
5.
The auxiliary reactor trip (moderator-shield water dump) shall add negative reactivity within one minute of its activation. Remote auxiliary reactor trip controls shall be available at the console and in the reactor room.
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6.
The normal moderator-shield water level shall be established at a minimum of 5 feet above the top of the core. Reactor operations water levels below this normal level shall be permitted only at when the operating power is lowered accordingly.
(Refer to Specificat ion 2.2 " Maximum Gamma-ray Exposure Rate".)
Moderator level adjustments near criticality shall be made only af ter first establishing criticality by control rod manipulation.
7.
A manual reactor trip shall be included in the reactor safety system.
Controls for the reactor trip shall be available at the console ana in the reactor room.
8.
A mannal electrical switch shall be provided in the facility for the purpose of disconnecting the electrical power of the facility and causing a reactor trip.
9.
The minimum safety system channels that shall be operating during reactor operation are listed in Table 1.
10.
The interlocks that shall be operable during reactor operations are listed in Table 2.
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Hases A minimum number of three control rods and a maximum excess reactivity are npi cif!.d to assure that there is adequate shutdown capability even for Luc stuck control rod condition. The minimum critical fuel loading in the reactor, in the best right cylindrical configuration. consists of twenty-two fuel elements and three control rods or twenty-one fuel elements and four control rods.
In either case, with control rods withdrawn, there are ef fectively twenty-five fuel elements (considering control rod fuel followers as fuel elements) in the minimum critical loading.
The core loading consisting of twenty-eight fuel elements and nine control rods in a hexagonal configuration centered in the core structure represents the normal loading. The normal core leading is used to establish a maximum excess reactivity of 15$ for the reactor.
The maximum reactivity insertion rates, far from and near criticality, are specified to assure that the reactivity addition rate is less than that analyzed in the maximum credible accident (MCA). The maximum control rod withdrawal rate and the moderator-shield water addition rate are controlled by these limitations.
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O Tl'e insertion time of less than 1.2 seconds for each control rod frcm its fally withdrawn position is specified to assure that the ir.sertion time does not exceed that assumed when establishing the minimum period specified in specification 2.2 as a limiting safety system setting.
required control rod withdrawal prior to adding the moderator-shield The water is specifLed to assure that reactor trip will have the capability of adding negative reactivity during reactor startup.
The auxiliary reactor trip is specified to assure that there is a secondary mode of shutdown available during reactor operations. The requirement that negative reactivity be introduced in less the.n one minute following artivation ol the trip lu established to limit the consequences of a potential power transient. By including a remote auxiliary reactor trip control in the reactor room, the trip may be activated readily by Individuals in the room under emergency conditions.
The normal moderator-shield water level of the reactor is established at a minimum of five f eet above the top of the core to assure that an adequate shield is provided at the maximum power level of the reactor.
When reactor operations require a lower moderator-shield water height (down to a one foot level), the operating power must be lowered accordingly so that the gamma-ray exposure rate limit setting is not exceeded over the When moderator water lexel reactivity control is to be utilized core.
(water level below one foot from the top of core), the gamma-ray exposure rate limit setting is J owered accordingly - to approximately 1/10 of its max i nnun permissible value - to further reduce the possibility of oper-at in;; wit h a high neut ron und gamma radiation field in the vicinity of the reactor room.
Controlling the reactor by moderator level near critirality v.,>ermittel only when the reactor is first made critical by control rod movements. This assures that the control rod is the primary mode of reactivity control in a critical reactor.
A manual reactor trip assures that a reactor trip can be readily initiated by an operator at his demand. Including a manual reactor trip control in the reactor room assures that the trip may be activated readily by individuals in the reactor room under emergency conditions.
liy providing a method of disconnecting the facility electrical power, an additional mode is established to manually cause a reactor trip.
The safety system channels listed in Table 1 provide a high degree of redundancy to assure that human or mechanical f ailures will not endanger the reactor facility or the general public.
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The interlock system listed in Table 2 assures that only authorized per-sonnel can operate the reactor, that the proper sequence of operations is performed, that no one can accidentally enter the reactor room when the reactor is operating, and that the reactor room is entered with proper conditions prevailing when the master consola key is on.
3.2 Reactor Parameters Applicability These specifications apply to core nuclear parameters and moderator-shield water physical parameters.
Objective The purpose of the specifications on reactivity coefficients is to assure that the inherent reactivity feedback mechanisms of the water moderator are safe. The purpose of the specification on purity of the water moder-ator is to assure adequate corrosion control in a room temperature, open air aluminum-water system.
Specifications 1.
At temperatures greater than 80 F, the temperature coefficient of reactivity shall be negative and shall have a minimum absolute value of 1 x 10-3 $/ F.
2.
The void coefficient of reactivity shall be negative and shall have a minimum ebsolute value of 1 x 10-1 $/% void fraction.
3.
The moderator-shield water shall have a pH within the range of 4.5-8, inclusive, and a resistivity greater than 200,000 ohm-cm when averaged over a two-month period.
Bases The minimum absolute value of the temperature coefficient of reactivity is specified to assure that an adequate inherent negative reactivity effect takes place when the reactor temperature increases above the value where the coefficient becomes negative. At lower temperatures, where the coef-ficient together with the slow rate at which controlled temperature in-creases may be effected provide assurance that positive reactivity inser-I tions due to controlled temperature increases will be small enough and slow enough as to be safely controllable. Uncontrolled heatups will quickly raise the moderator temperature to the range where the coefficient is negative while providing only a negligible positive reactivity effect.
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The minimum absolute value of the void coefficient of reectivity is
- pacified to assure that the negative reactivity insertion due to veld formation is greater than that which was calculated to occtr in the SAR.
The moderator-shield water quality is specified to assure aiequate corrosion control in a room temperature, open air, aluzinum-water system. This ccr-rosion is a long-term reaction. Based on years of experience, a two-month period for averaging has proven adequate to avoid quality degradation which would result in apprec*able corrosion.
3.3 Radiation Monitoring Applicability These specifications apply to the minimum radiation monitoring requirements for reactor operations.
Objective The purpose of these specifications is to assure that adequate monitoring is available to preclude undetected radiation hazards or uncontrolled releases of radioactive material.
Specifications 1.
The minimum radiation monitoring systems for reactor operation shall include:
a.
A criticality detector system which monitors the main fuel storage area and also functions as an area monitor. This system shall have an audible alarm in the reactor room.
b.
An area gamma monitor in the console room with an audible alarm.
2.
Instruments to permit the periodic sampling and measurement of radio-l activity in the air and the moderator-shield water shall be provided.
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Portable detection and survey instruments shall be provided.
I Bases The continuous monitoring of radiation levels in the reactor room and console room assures the warning of the existence of any abnormally high radiation levels. The availability of instruments to measure the amount
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of radioactivity in the air and moderator-shield water will assist in l
monitosing fuel clad integrity and assures continued compliance with the j
requirements of 10 CFR Part 20.
The availability of the required portable monitors provides a.ceurance that personnel will be able to monitor potential radiation fields before an area is entered.
4.0 EXPERIMENTS Applicability These_ specifications apply to all experiments placed in the reactor tank.
Objective The objective of these specifications is to define a set of criteria for experiments to assure the saf ety of the reactor and personnel.
I Specifications
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y No experiment shall be performed until a written program, which has been developed in sufficient detail to permit good understanding of the safety aspects, is reviewed by the Reactor Safeguards Consmittee (RSC) and approved by the Facility Manager.
2.
No experiments shall be conducted if the associated experimental equipment could interfere with the control rod functions or could adversely affect the nuclear instrumentation.
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The maximum reactivity change for withdrawal and insertion of experimental samples and devices by remote means shall be 0.25$.
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Ec mcximum positive step insertion of reactivity, which can be l
caused by an experimental accident or experimental equipment failure, shall not exceed 0.80$.
5.
Experiments shall not contain a material which may produce a violent chemical reaction and/or air'orne radioactivity.
Bases The experiments to be performed in the reactor programs are discussed in the Safety Analysis Report (SAR). H e present programs are oriented almost exclusively toward fundamental reactor technology training. Other O
. upccial programs may involve the use of the reactor as an irradiation facility. To assure that experiments are well planned and evaluated Prior to being performed, detailed written procedures for all new experiments must be prepared, reviewed and approved by the Facility Manager and RSC.
Since the control rods enter the core by gravity and are required by other Technical Specifications to be operable, no experiment should be allowed to interfere with their functions. To assure that specified power limits are not exceeded, the nuclear instrumentation must be capable of accurately monitoring core parameters.
All reactor experiments are reviewed and approved prior to their per-formance to assure that the experimental techniques and procedures are safe and proper, and the hazards from possible accidente are minimal.
A maximum reactivity change is established for the rer.ote positioning of experimental samples and devices during reactor operations to assure that the reactor controls are readily capable of controlling the g
reactor.
All experimental apparatus placed in the reactor must be properly fabricated and made physically secure in the reactor. In consideration of potential accidents, the reactivity effect must be limited to the maximum accidental step reactivity incertion analyzed in the SAR.
This value of reactivity would cause the reactor to rise on a stable period of I second and allow the safety system to trip the tsactor before an excessive power level is reached. In actual practice, no single unit of experimental apparatus will be placed in the reactor I
which has a reactivity worth greater than 0.80S which is less than the reactivity addition encident analyzed in the SAR.
Restrictions on irradiations of explosives and highly flammable materials are imposed to minimize the possibility of explosions or fires in the vicinity of the reactor.
l To minimize the possibility of exposing f acility personnel or the public l
to radioactive materials, no experiments will be performed with materials that could result in a violent chemical reaction and/or produce airborne l
radioactivity, i
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~ "s. O SURVEIL. LANCE REQUIRD4ENTS 5.;
Reactor Control and Safety Applicability These specifications apply to the surveillance of the safecy and control apparatus and instrumentation of the facility.
Objective the purpose of these specifications is to assure that the safety and control equipment is operable and meets the criteria established in the design bases.
g.71fIcations 1.
The total control rod drop time and magnet release time shall h measured at intervals not exceeding six months to verify that the requirement of specification 3.1, item 3,is set.
2.
The moderator-s.'
- eld water dump time shall be naasured at intervals not exceeding six : anths to verify that the requirement of specifi-cation 3.1, item 5,is met.
3.
The maximum control rod and moderator-shield water reactivity insertion rates shall be validated at intervals not exceeding twelve months to verify that the requirements of specification
- 3. L, item 2, are met.
4.
The following shall be performed each day prior to initial reac. tor operation, except when continuous reactor operations are scheduled, then they shall ba performed once each day, a visual inspection of reactor components, a.
b.
an operability check or test of safety system channels, an operability check of the interlock system, and c.
d.
an cperability check of the radiation monitors and alarm set points.
. 5.
The safety system channels shall be calibrated at intervals not exceeding six months.
P.as e s Past performance of conttol rods and control rod drives, and the woderator-shield water fill-and-dump valve system have demonstrated that testing at intervals of six months is adequate tc. assure compliance with specifi-cation 3.1, items 3 and 5, and validation at intervals of twelve months compliance with item 2 of specification 3.1.
assures Visual inspection of the reactor components, including the control rods, prior to operation is to assure that the comptnents have not been damaged and that the core is in the proper condition. Since redurdancy of all safety channels is provided, random f ailures should not jeopardize the ability of the overall system to perform its required functions. The nterlock system for the reactor is designed so that its failure places t he system in a saf e or non-operating condition. However, to assure that f ailures in the safety channels and interlock system are detected soon as possible, frequent surveillance is desirable and thus specified.
as The frequent checks of the area radiation monitors and their alarm points assures the ability of the system to perform its required function.
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Past experience has indicated that, in conjunction with the daily check, l
calibration of the safety channels at intervals of six months assures that proper accuracy is maintained.
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Reactor Parameters Applicability These specifications apply to the verification of control rod reactivity worths, temperature and void coefficients of reactivity, and reactor power level, which are pertinent to the reactor control and transient analysis and to water quality.
Objective The purpose of these specifications is to assure that the analytical basec are and remain valid and that the reactor is safely operated.
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U1 Npecificallons The following parameters shall be determined during the initial J.
physics measurements of each new reactor core configuration or composition and validated periodically with the stated frequency:
a.
individual control rod reactivity worths (at intervals not exceeding twelve months),
b.
temperature and void coefficients of reactivity (at inter-vals not exceeding twenty-four months), and reactor power calibration (at intervals not exceeding six c.
months).
2.
The reactor moderator-shield water quality shall be determined at least once cach month.
15as es Measurements of the above core parameters are made when a new reactor
- unfiguration or composition is assembled. Whenever the core configuration or compoultion is altered, the core parameters are evaluated to assure that they are within the limits of these specifications and the values analyzed in the SAR.
During the initial startup test period of the reactor, measurements and determinations of the core parameters will be made for all standard assemblies which are to be utilized in the reactor's operational programs. Verification of these parameters are made periodically to assure that changes have not taken place.
Past experience indicates that monthly measurements of the moderator-shield water quality are adequate to comply with the requirement of specification 3.2, item 3.
5.3 Radiation Monitoring Applicability These specifications apply to the surveillance of the radiation monitoring equipment and activities of the facility.
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t ill>I object ive The purpose of these specifications is to assure the continued validity of ra.istion protection standards in the facility.
Specifications 1.
The area radiation mo.nitors and the portable radiation survey instru-ments shall be calibrated at intervals not exceeding three months.
Inc air in the reat' r room shall be sampled and measured for partic-ulate radioactivity at Icast once each month.
J.
The water in the reactor
- ank and dump tank shall be sampled and rc.rasured for radioactive contaminants at least once each month.
4.
The reactor f acility shall be surveyed for radioactive contamination at intervals not exceeding six months.
Experience has demonstrated that calibration of the area radiation monitors and the portable survey instruments at three-monti. intervals is adequate to assure that significant deterioration in accuracy does not occur.
The specified frequencies for monitoring radioactive contamination in the air md water in the reactor room as well as in the overall reactor based on previous experience.
Iariii y 4.i 111.'; I GN FlWIURES 8,. I il t e The NTR site shall be located on Commonwealth Edison property adjacent to the cy:lusion area for the Zion Station. An exclusion redius of 4 3t1 feet shall define the exclusion area for the site.
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- 6.7 Facility The reactor facility shall be housed in the Nuclear Training Center building. The main security actrance into the f acility shall be an interior access through the Training Center. The controlled area of the f acility shall be bcunded by a personnel security fence and interior building walls. The distance from the centerline of the NTR to the edge of the controlled area shall not be less than 40 feet.
6.3 Reactor Rocm The reactor room shall be an eight inch concrete block enclosure with approximate floor dimensions of 27 x 27 feet. The height from the ground floor to the ceiling shall be about 27 fact. One exterior side wall shall have an equipment-emergency door and one interior wall shall have a main access door. A 12-foot d'.ameter, aluminum dump cank shall be centered in the reactor room with its top below the flocr lesel. The dump tank shall be surrounded by 18 inches of reinforced concrete and shall rest on a 30-inch thick reinforced concrete pad. A steel structure approximately
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10 feet above the floor level and supported on beams at the side walls shall be used to support the control rod drive platform.
i 6.4 Reactor 6.4.1 Reactor Tank l
The aluminum reactor tank shall have a capacity of approximately 7000 gallons of water with no core components in place. The tank nominal dimensions shall be 8 feet in diameter and 19 feet high.
The reactor tank shall fit inside the dump tank in an eccentric f ashion and shall hang on a floor level steel structure which lies over the dump tank.
,A horizontal support shall be provided near the bottom of the tank. The tank has openings for the inlet water line of the water sill system, the drain line at the bottom of the tank, and the water dump line. An annular support elate shall be provided to support the core structure.
6.4.2 Reactor Core l
The aluminum reactor core structure shall be comprised of upper and lower grid places, core shroud tubes for fuel elements and control rod locations, and tie-down rods. The two hcrizontal grid places are separated by approximately 44 inches. The Lottom O')
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- pl. te 13..ipproulmately 48 inchen in diamet t and varies in thick-ne :. from 2 Inthes at its edge to a nominal 6 inches in the center portion.
The upper grid plate is an elongated hexagon, 1 inch thick. The distance between the flats along the two short axes is approximately 31 inches nd along the long axis 34 inches. The center-to-center spacing of the position holes is 3.125 inches.
There.is a total of ninety-three position holes; nine of which contain control rod shroud tubes, seven are experimental hole positions normally containing fuel element insert adapters and the rentin4r contain fuel element shroud tubes. The core structure s radially cente.ed in the reactor tank and positioned vertically so tha*. the top af the reactor core is approximately 11.5 feet from the floor level.
6.4.3 Standard Fuel Elements A st.indard fuel element shall be composed of 3 concentric tubes of an aluminum-clad fuel alloy that are held together and positioned at each end by aluminum brackets. The fuel alloy (13 w/o U-Al, 93.5% U-235 enriched) has a thickness of 0.053 inch
' the clad thickness is a nominal 0.036 inch. The aluminum an, grade used is #1100 (25) or its equivalent.
Each standard element nominally crsntains 200 gas of U-235 (outer tube - 82.4 gms, middle tube - 66.6 gms and inner tube - 50.9 ges).
The length of the fuel alloy is 36 inches and is centered length-wise in the 42-inch fuel tube.
The total fuel element length, including the top and bottom aluminum brackets, is 47-5/16 inches.
The outside dlareter of the outer, middle and inner tubes is 2.50 inches, 2.06 'actes and 1.62 inches, respectively. The nominal water gap bet.ieen the fuel tubes is 0.094 inch.
When the f uel eleme..ts are disassembled and reassembled la any combination of one or two fuel tubes, the resulting element shall be considered a special, non-standard fuel element.
6.4.4 Control Rods The control rods shall consist of a cadmium neutron adsorbing section, clad with aluminum and an attached standard fuel element follower.
'Ihe length of the poison section of the rod is 36 inches.
The maximum diameter of the rod is 2.5 inches. The aluminum-clad
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cadmium tube and the fuel element follower are linked by an axial stainless steel support rod. This linkage extends to and includes a shock adsorber piston at its lower end, a stop cap above the poison section, and a magnet carriage armature at its upper end.
The control rods are guided through their respective core position holes by aluminum shroud or guide tubes. The shroud tubes extend above and below the core grid plates. The enclosed lower section with appropriate holes serves as a water dash pot and the upper section provides a physical stop to prevent further downward motion of the control rod.
There are nine control rod locations ih the core structure. The control rods in the three locations l
lying closest to the center of the structure are safety rods and those in the six outer locations are shim rods.
6.4.5 Control Rod Drive Assembly The drive assembly for the control rods shall consist of a G\\)
magne? carriage, a verti~ cal drive, a position indicator, and suitable drive motors.
Indications of magnet carriage "up",
magnet carriage "down", and control rod "down" are. provided.
i The position indicator gives an indication of the control rod or its magnet carriage position throughout its distance of travel (maximum 41 inches).
The drive assemblies shall be mounted on an overhead platform located approximately ten feet above the floor level and centered over the reactor tank.
f e. 's Winter flandling System The water handling system shall allow remote filling and emptying of the reactor tank.
It shall provide for a water dump in the event that an auxiliary reactor trip is necessary. The filling system shall be controlled by the operator who must satisfy the sequential interlock system before adding water to the reactor tank. Two sump pumps shall be provided to add the moderator-shield water from the storage-dump tank into the reactor tank. Slow and fast fill rates of about 90 gpm and 180 gpa shall be possible. A nominal ten-inch valve shall be installed in the dump line and have the capability of emptying the reactor tank on demand of the operator. A valve shall be installed in the bottom drain line of the reactor tank to provide for completely emptying the reactor tank.
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. 6.6 Fuel Storare When not in use, the fuel elements and tubes shall be stored in the reactor room in facilities which will provide adequate isolation and radiation safety. No more than six fuel elements shall be out of the core or the approved storage areas at one time. The following facilities shall be considered approved storage areas:
1.
Reactor tank storage tubes, a maximum number of sixteen, shall be located inside the reactor tank and each shall be capable of holding not more than one fuel element. These t'$es shall be located a r.inimum of five feet below the reactor tank top and next to the reactor tank wall in a single curved slab array.
2.
The main fuel storage area shall consist of an annular ring of storage tubes each of which is capable of holding not more than one fuel element.
The resulting annulus shall be lined along the inside perimeter with a 20-mil thick cadmium liner. An alternate main fuel storage area may consist of a single row of single storage locations, forming a si=ple slab.
3.
When the control rods are unloaded from the core and the fuel element followers are not removed, the rods chall be hung in the control rnd storage rack in the reactor room. This rack shall have storage locatiot.s with spacing equal to or greater than 5-3/4" in a single line and shall have no = ore than 9 positions.
7.0 ADMINISTRATIVE CONTE 0LS 7.1 Orvanizction l
The Manager of the Westinghouse Nuclear Training Reactor Facility shall be j
responsible for the safe operation of the reactor. The Manager reports j
through the nor=al line of management as indicated in Table 3 to higher levels of management. The facility organization shall consist of the Manager, a Training Systems Coordinator, a Reactor Lead Engineer and a i
l staff of reactor operators. All members of the facility operating staff shall be licensed operators or operators in training.
7.2 Personnel Recuirements 7.2.1 When the reactor is not secured, the reactor console shall be under l
the surveillance of a licensed operator.
- Revision #2 to R-119, 5/4/76 l
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. 7.2.2 A licensed senior operator shall be "readily available on call" at all times during reactor operations. A licensed senior operator shall be present at the reactor facility during initial loading and approach to critical and power following each configuration change, recovery from an unplanned shutdown or significant reduction in power, and fuel handling and refueling. However, a licensed senior operator's presence will not be required during recovery from an unplanned shutdown or significant reduction in power when the cause has been clearly established and corrected. The identity of and a method for rapidly contacting the senior operator on duty shall be known to the operator on duty.
7.3 Review and Audit 7.3.1 The Reactor Safeguards Committee (RSC) shall include at least four scientists or engineers who are not in the Itae ctganization responsible for reactor operations (i.e., Nuclear Training Oper-ations Group - NID) and shall represent at Icust one half of the g)
Committee membership. The rainimum qualifications of the RSC members with regard to nuclear experience shall be:
1.
Each member must have a minimum of five years industrial experience in nuclear and related fields and must have a minimum of three years of active participation in his nuclear orientated discipline.
2.
The experience and knowledge of each member must be applicable to or pertain to the Committee's responsibility to properly review the facility and its operation from the standpoint of safety.
3.
Each member must be papable and willing to exercise his individual judgment in regard to all Connaittee reviews and decisions.
4.
The " nuclear orientated discipline" of a minimum of two Committee members must lie in the areas of reactor physics and reactor operations.
7.3.2 The RSC shall meet at least once each six months. A quorum of the I
RSC shall consist of at least four members and at least half of those l
present shall be from organizations outside the line organization I
responsible for reactor operations.
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- 7.3.3 The RSC shall review activities and advise the Manager of NTR and/or whatever echelon of management it feels appropriate on all matters pertaining to the safe operation of the NTR. The reviews shall cover:
1.
Proposed experiments, tests and operations not described in the Safety Analysis Report.
2.
Proposed changes or modifications to tt.e facility not described in the Safety Analysis Report.
3.
Proposed changes to the Technical Specifications.
4.
Proposed normal operating procedures and emergency procedures, and proposed changes thereto.
i 5.
Facility operation for compliance with internal rules,tro-cedures and regulations, and with license provisions.
6.
Performance of facility apparatus and equipment.
7.3.4 Recording and reporting requirements for the RSC shall include:
1.
Minutes of each meeting.
2.
Special reports on experiments reviewed and facility inspections, including the Committee's findings.
3.
Special reports on facility radiation safety practices and records made semiannually.
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4.
All Committee reports and meeting minutes shall be transmitted through the line management up to and including the Manager.
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7.4 Operating Procedures 7.4.1 Personnel entry into the reactor room shall be subject to the
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No person shall be allowed in the reactor room unless the reactor is subcritical by at least 35.
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4 2.
No person shall be allowed in the reactor room if remote changes are being made to the reactor which may produce positive reactivity effects.
3.
When personnel enter the reactor room to perform activities which may affect the reactor's condition in any way, the following requirements shall be met:
a.
An inter-communication system shall be operational providing voice communications between the reactor room and the console room.
b.
An audible signal of the reactor scurce multiplication shr.11 be heard in the reactor room.
7.4.2 Approved written operating procedures shall be followed for the following items:
1.
Facility security rules.
2.
Routine reactor operations.
3.
Radiation safety practices.
4.
Preventive or corrective maintenance operations which could have an ef fect on the safety of the reactor.
S.
Non-routine operations and emergency situations.
6.
Fuel handling, storage, and changes in the core.
7.4.3 New procedures and changes in the operating procedures shall require review by the RSC and the approval of the Facility Manager.
7.4.4 Temporary changes in the operating procedures which de c9t change the intent of the original procedures may be made by the Facility Manager. Such changes shall be recorded in the operating records and reported to the RSC.
7.5 Action to be Taken in the Event of an Abnormal Occurrence 7.5.1 Abnormal occurrences shall include but not necessarily be limited to the following:
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A violation ut a limiting safety system setting.
2.
A violation of a limiting condition for operation.
3.
An engineered safety system component malfunction or other component or system malfunction which could render the reactor safety system incapable of performing its intended safety function.
4.
An unctitrolled and unanticipated change in reactivity.
5.
A personnel action that may cause an unsafe condition in connection with the operation of the reactor.
7.5.2 In the event of an abnormal occurrence, reactor operation aball not be resumed until the cause is known and appropriate corrective measures are taken.
7 The occurrence shall be reported to the AEC in accordance with j
Section 7.7.1 of the specifications.
4 7.5.4 A report shall be prepared which shall include an analysis of the causes of the occurrence and recommendations for action to prevent ce reduce the probability of recurrence. The report shall be submitted to the RSC for review and shall be maintained as part of the f acility records.
7.6 Action to be Taken in the Event a Safety Limit is Exceeded 7.6.1 If a safety limit is exceeded, the reactor shall be secured or otherwise placed in a safe condition and reactor operation shall not be resumed until authorized by the AEC.
7.6.2 An immediate report of th'e occurrence shall be made to the AEC in accordance with Scction 7.7.1 of the specifications.
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7.6.3 A complete analysis of the incident * : ether with recommendations for preventing or reducing the probability of recurrence shall be prepared and submitted to the RSC and to the AEC when authorization to resume operation is sought.
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7.7 Reporting Requirements l
In addition to reports otherwise required by applicable rer.n i n i i nm-O
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7.7.1 The licensee shall inform the Commission of any abnormal occurrence or violation of a safety limit. For each such occurrence, the licensee shall notify, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone or telegraph, the Director of the appropriate Atomic Energy Commission Regional Compliance Of fice listed in Appendix D of 10 CFR 20 and shall sub-mit within ten days a report in writing to the Director, Division of Reactor Licensing (hereinafter, "the Director, DRL").
7.7.2 The licensec shall reert to the Director, DRL, in writing within 30 days of its observeo occurrence any significant change in the trat.aient or accident stalyses as described in the Safety Analysis Report, as amended, or any changes in the facility organization structure.
s.7.3 The licensee shall submit a report within 60 days after criticality of tne reactor, in writing to the Director, DRL, upon receipt of a new facility license or an amendment to the license authorizing an it. crease in reactor power level or the installation of a new core, describing the measured values of the operating conditions or characteristics of the reactor under the new conditions, including:
1.
Total control rod reactivity worth,
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2.
Reactivity worth of the singic control rod of highest reactivity worth, 3.
Total and individual reactivity worths of any experi-ments inserted in the reactor, and 4.
Minimum shutoown matgin both at room and operating ten-peratures.
7.7.4 The licensee shall sulait in writing, to the Director, DRL, m.
annual operating report within 60 days af ter the end of each calendar year, providing the following information:
1.
A narrative summary of operating experienes (including experi-ments performed) and changes in facility design, performance characteristics and operating procedures related to reactor safety.
2.
The energy generated by the reactor and the number of hours the reactor was operational.
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3.
The number of inadvertent reactor trips, including the reasons therefor.
4.
Discussion of the major maintenance operations performed during the reporting period, includi.5.g the effect, if any,
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O wafe operation of the reactor and the reasons for on the any corrective maintenance required.
- i. A summary description of changes in the facility or pro-cedures, and tests and experiments carried out under the conditions of Section 50.59 of 10 CFR 50.
6.
A summary of the nature and amount of radioactive ef fluents released or discharged to the environs beyond the effective control of the licensee as measured at or prior to the point of such release or disci.arge.
- 7. 8 Records Records maintained by the licensee shall include, but not necessarily be limi ted to, the following:
7.M.1 Reactor operating records, including power levels and periods of operation at each power level.
7.H.2 Records of inadvertent reactor trips, including rr.nsons therefor.
7.H.3 Records of c:xperiments, including any unusual events involved in their performance and in their handling.
7.H.4 Records of abnormal occurrences.
to the 7.8.S Records of tests and measurements performed pursuant Technical Specifications.
Records of maintenance operations involving substitution or 1
7.8.6 replacement of reactor equipment or components.
l 7.8.7 Records of fuel inventories and transfers.
Records showing radioactivity released or discharged into the air 7.8.8 or water beyond the ef fective control of the licensee as measured or prior to the point of such release or discharge, at Records of f acility contamination and radiation survey results.
7.8.9 Records of radiation exposures for all facility personnel and 7.8.10 visitors.
7.8.11 Updated, corrected, and as-built drawings of the facility.
Items 1 through 6 shall be retained for at least five years; items 7, 8,
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11 shall be retained for the life of the facility, and item 10 9 and records shall be retained indefinitely or until the Commission authorizes their disposal.
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MINIMUM SAFETY SYSTEM CHANNELS l
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i Reactor Conditions i
and Ranges Channels Min (mum Number Functions I
dource Range Linear or Log Neutron Level 1
High Neutron Level Reactor Trip j
l (kgg < 0.99)
Linear or Log Neutron Level 1
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1 Start-up and Power Linear or Log Neutron Level 2
High Neutron Level Reactor Trip i
Range Period 1
Period Trip l
(kd f > 0.99)
Linear or Log Camma Level 1
High Camma Level Reactor Trip i
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TABLE 2 MINIMUM INTERLOCKS INTERLOCKS ACTION IF INTERLOCK NOT SATISFIED Console Master Key "On" Reactor Trip Reactor Room Door Closed (a)
Reactor Trip and prevents control rod withdrawal and moderator insertion Neutron Flux Up (a)
Prevents control rod withdrawal and moderator insertion Safety Rods Cocked (a)
Prevents control rod withdrawal (b) and moderator insertion I
tj Water Level Up (a)
Prevents control rod withdrawal (b) i Reactor Room Access Key "On" Prevents control rod withdrawal (b)
Count rate cutout (high and' low)
Prevents bank control rod withdrawal (b) and
" fast fill" mode of moderator insertion (a) During maintenance checks, special operations and " console master key on" reactor room entry, these interlocks may be temporarily bypassed using special individual key switches.
(b) With following exception: safety rods are moved to cocked position prior to any othar positive insertion operation O:
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TABLE 3 ORGANIZATION CH.\\RT (W) Nuclear Training Reactor Facility Nuclear Training Center, Zion, Illinois Manager Startup & Training Services Manager Nuclear Training Center Zion, Illinois NTR Reactor Safeguards
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Manager Nuclear Training Reactor
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Facility Coordinator Training Systems Reactor Lead Enoineer Nuclear Training ~ Reactor Facility Operators i
Nuclear Training Reactor Facility
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- Revision #2 to R-119, 5/4/76
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