ML20207J872
| ML20207J872 | |
| Person / Time | |
|---|---|
| Site: | 05000131 |
| Issue date: | 03/12/1999 |
| From: | DEPT. OF VETERANS AFFAIRS MEDICAL CENTER, OMAHA |
| To: | |
| Shared Package | |
| ML20207J794 | List: |
| References | |
| NUDOCS 9903160423 | |
| Download: ML20207J872 (40) | |
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TECHNICAL S?ECIFICATIONS FACILITY LICENSE NO. R-57 i l
l FOR THE OMAHA DEPARTMENT OF VETERANS AFFAIRS l A.J. BLOTCKY REACTOR FACILITY l
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1 1-9903160413 990312 PDR ADOCK 05000131 P PDR _
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\g TABLE OF CONTENTS 1
l 1.0 D E FI N I TI O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 l 1 1 1
2.0 SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTINGS... ............. ... . .... 5 2.1 Temperature of the reactor fuel..... .. .... .... ......... .. . ...... ... . ....... 5 2.1.1 Fuel Element Temperature. ......... ............................ ....... .... . ..... 5 2.1.2 Thermal Power Level .. ... . ...... . .. . . ...... .. .. . . . .. .. ... .. . . . .. ... ... . .. ....5 2.2 Limiting Safety System Settings.... ... .. .............................. . . . . . . . . .. . . 5 2.2.1 Bulk Pool Tem perature .... .. ... . ... . . ..... . ... ... .. .. .. . .. . . . . . . .. .. .... .. . . ... .. 5 l
2.2.2 Thermal Power Level ... . . .... . . .. .. .. .. ... .... .. . . . . .. ... . . .. . . .. .. .. . .... 5 2.2. 3 Pool Wate r Lev el . . . . . . . . . . . . . . . . . .. . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 !
3.0 LIMITING CONDITIONS FOR OPERATION ...... ................................6 3.1 Reactor Core Parameters ......... . .. ..................................................6 3.1.1 Ex ce ss Rea ctivity. . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 3.1.2 Shutdown Margin........ . ....... .. . . ............:..........................6 3.1.3 Core Config uration .. ... .. . . ....... ..... .. . . . . ..... .. . . . . . ... . . . . .. . . . . . ... .. . . . 6 3.1.4 Temperature of the reactor fuel ..... .......... ........... . ... ........... ....... 7 y 3.1. 5 Fuel Pa ra m eters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . 7 3.2 Reactor Control and Safety Systems.... ..................... . .. .. ... ... .. ...... 8 3.2.1 Operable Contro: Rods ................. .... ................. . . ,....... . . . 8 3.2.2 Reactivity Insertion Rates ................... .................... .. ................ 9 3.2.3 Scram Ch annels ... . . . . ... ... .. ... .... ..... . ... . . . ... .. . .. .... . ............9
- 3. 2.4 I nt e rl o cks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 3.2.5 Control Systems and instrumentation Requirements of Operation 11 3.3 Coolant System .. . . .......... .........................................................12 3.4 C onfineme nt System.. . . . . . . . .. . . . ...... . .. ... ...... .. ..... . ...... . . .... ... . . . ...... 13 3.5 Ventilation Systems ... ...................... ................................................14 3.6 RadiaM Monitoring Systems and Effluents ... ........... . .... ..... ... ........ 15 3.6. i Monitoring Systems ... ... . . .. . . ... ... .. . .. . . ... .. . ...... . . . . . . .... .. . . ... . 15 3.7 Experiments.. .. ..... . .... . ................ .. . . . . . . . . . . . . . . . . . . . . . . ....................16 3.7.1 Reactivity Limits... ..... ... ...................................................16 3.7.2 M aterials . . . . . . . . . . . . . .. . . . . . . . . . ... . .. . . . . . . . .. . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 3.7.3 Failure and Malfunction.. . .... . .............................. . . .. ........... ... . 18 4.0 SURVEILLANC E REQUIREM ENTS ............. .............................. . ......... ........... 19 4.1 Rea ctor core pa rameters ..... ... . . .. . .............. . .. ...... ... . ...... . . . .... . . . ... . .. . . 19 4.1.1 Exce s s Re a ctivity.. .. . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . 19 l O
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i i 4.1.2 Shutdown Margin.. ............ ... . . . . ..................................20 h\
4.1.3 Core Configuration..... ................. .. .
4.1.5 Fuel Element inspection .. ............. ......
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. . . . . . . . . . . . . . . . . . . . . . 20 4.2 Reactor Control and Safety System.......... . .. ... . ..... . ..... ...... .......... 21 4.2.1 Reactivity Worth of Control Rods . ... . ..... . .... .................... ... ... 21 4.2.2 Rod Maximum Reactivity insertion Rate. . . .... ............. .............. 21 4.2.3 Scram Times of Control Rods............ .. . . .... ................. ......... 21 4.2.4 Scram and Measuring Channels....... ..... .. . .... ............................. 22
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4.2.5 Thermal Power Calibration ....... ...... .. .... . . .........................22
- 4. 2. 6 Rod I n spe ction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . .. . . . . . . . . . . . . . 22 4.3 Coolant Systems ......... ..... ................... .. . . ....................................23 4.3.1 Analysis of Coolants for Radioactivity. ....................................23 4.3.2 Conductivity and pH......... .............. ......................................23 4.4 C o n fi ne m e n t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 4
- 4. 5 Ve n tilation S ystem s . . . . . . . . . . . . . . . . . .. . . . . . . . . .. . . . . . . . .. . . . . ... . . ... . .. . . . . . . . . . . . . . . . . . .. 24 4.6 Radiation Monitoring Systems........... .. ...... ..................................24 4.6.1 M onitoring Systems .. . .. .... . ..... ... . .. ... . . . ...... ... . .. ....... .. . . . .. . .. . 24 5.0 iSIGN FEATURES... ................... .......................................................25 5.1 Site and Facility Description ..... . . .. . .... .... . .. ........ . ............... ........ 25 5.2 Reactor C oola nt System ......... ... . ... .......... .... . . .. .. . . ... .. ... .... .. . . ... . . . ... .... 25
- 5. 3 R ea cto r C ore a n d F uel . . . . . . . . . . . .. . . . . . . . . . . . . . . . .. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26
- 5. 3.1 Rea ctor C ore . . . . . . . . . . . . . . . . .. .. . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . 26
- 5. 3. 2 R e a ctor Fuel . .. . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . .. . . . . . . .. . . . . . . . . . . . . . . . . . . 26 5.3.3 Control Rods... .. .. ....... ....... ... . . .....................................26 5.3.4 Fissionable Material Storage . .. . . . ................................26 6.0 ADMINISTRATIVE CONTROLS ...... .. ..... . . .. . . . . . .......................27 6.1 O rga n izatio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ................................27 6.1.1 St ru ct u re . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 6.1.2 Responsibility....... ... . .... . ... .......... .. .. ................................27 6.1. 3 S t a ffi n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 9 6.1.4 Selection and Training of Personnel .. ....... . ................... . .... .. . 29 6.2 Review and Audit ........ ....... .......... .. ......... . .. ...............................30 ,
6.2.1 Composition and Qualifications... ...... ...... ... ... ..... ... ........ ... .. 30 ;
6.2.2 Charter and Rules...... . ... . ....... .. .. .. . . . . . .. .. . . . .. ......... .. . . . . . . . 30 6.2. 3 Review Fu n ctio n .. . . . . . . . . . . . . .... .. . . .. . .. . . . . .. . . . . . . . . . .. . . . ... . . . . . . .. . 30
- 6. 2.4 Aud it Fu nctio n . . . . .. . . . . . . . . . . .. . .. . . . . . . . . . .. . ... . . . . . . .. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 6.3 R adiatio n S a fety . . . . . . . . . . . . . . . . . . . . ... ... . . . . . . . .. . . . . . . . . . .. . . . . .. . . . . .. . . . . . 32 6.4 P roced u res . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 32 6.5 Experiment Review and Approval.... ............ . ..... ............ .......... .. ... ... 33 6.6 Req u i red A ctio n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . .. . . . . .. . 33 6.6.1 Required Action to be Taken in Case of Safety Limit Violation.... ... 33 6.6.2 Action to be Taken in the Event of an Occurrence of the Type identified in 6.7.2(1)b, and 6.7.2(1)c. ............. .......... . ........ ...... .. 34 6.7 Reports................................................................................................34 6.7.1 Operating Reports . ... ... ...... .. . . .. ... . ..... . ... ...... ............ .. .. . . .. . .. 34 t
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6.7. 2 S pecial R eports . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 6.8 Records..................................................................................... ... 36 6.8.1 Records to be Retained for a Period of at Least Five Years or for the Life of the Component involved if Less than Five Years..... 36 6.8.2 Records to be Retained for at Least One Certification...... .. .. ....... 37 6.8.3 Records to be Retained for the Lifetime of the Reactor Facility ..... 37
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1.0 DEFINITIONS channel. A channel is the combination of sensor, line, amplifier, and output devices which are connected for the purpose of measuring the value of a parameter.
channel calibration. A channel calibration is an adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the parameter which the channel measures. Calibration should encompass the entire channel, including equipment actuation, alarm, or trip and shall be deemed to include a channel test. <
channel check. A channel check is a qualitative verification of channel behavior, or by comparison of the channel with other independent channels or systems measuring the same variable.
channel test. A channel test is the introduction of a signal into the channel for verification that it is operable.
senior reactor operator. A senior reactor operator is an individual who is licensed to direct the activities of reactor operators. Such an individualis also a reactor operator, reactor operator. A reactor operator is an individual who is licensed to manipulate the controls of a reactor.
confinement. Confinement means a closure on the overall facility that controls the movement of air into it and out through a controlled path.
j I containment. Containment means a testable enclosure on the overall facility (for example, a '
reactor room) that is in the normally closed configuration and can support a defined pressure differential for functional purpose.
excess reactivity. Excess reactivity is that amount of reactivity that would exist if all reactivity control devices were moved to the maximum reactive condition from the point where the reactor is exactly critical (k.n=1) at cold, clean conditions without xenon.
l experiment. Any operation, hardware, or target (excluding devices such as detectors, foils, '
etc.), that is designed to investigate'non-routine reactor characteristics or that is intended for irradiation within the pool, on or in a beamport or irradiation facility, and that is not rigidly i secured to a core or shield structure so c: to be a part of their design.
measured value. The measured value is the value of a parameter as it appears on the output of a channel. !
moveable experiment. A moveable experiment is one where it is intended that all or part of 4 the experiment may be moved in or near the core or into and out of the reactor while the reactor is operating.
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l l operable. Operable means a component or system is capable of performing its intended function.
reactivity worth of an experiment. The reactivity worth of an experiment is the value of the reactivity change that results from the experiment being inserted or removed from its intended posinon.
reactor operating. The reactor is operating whenever it is not secured or shutdown.
reactor safety system. Reactor safety systems are those systems, including their associated ,
input channels, which are designed to initiate automatic reactor protection or to provide l information for initiation of manual protective action.
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reactor secured. A reactor is secured when- I (1) Either there is insufficient moderator available in the reactor to attain criticality or there is l insufficient fissile material present in the reactor to attain criticality under optimum available conditions of moderation and reflection, or (2) The following conditions exist:
- a. Three neutron absorbing control devices are fully inserted or other safety devices are in shutdown position, as required by technical specifications, and
- b. The console key is in the off position and the key is removed from the lock, and
- c. No work is in progress involving core fuel, core structure, installed control rods, or control rod d.-ives unless they are physically decoupled from the control rods, and
- d. No expedments are being moved or serviced that have, on movement, a reactivity worth O exceeding one dollar.
l reactor shutdown. The reactor is shutdown if it is subcritical by at least 1 dollar both in the reference core condition and for all allowed ambient conditions with the reactivity worth of all experiments installed.
reference core condition. The reactivity condition of the core when it is at 20 oC and the l reactivity worth of xenon is zero (i.e. cold, clean, and critical). '
research reactor. A research reactor is defined as a device designed to support a self- !
sustaining neutron chain reaction for research, development, educational training or i experimental purposes, and that may have provisions for the production of radioisotopes. I rod-control. A control rod is a device fabricated from neutron absorbing material or fuel, or both, that is used to establish neutron flux changes and to compensate for routine reactivity losses. A control rod can be coupled to its drive unit allowing it to perform a safety function when the coupling is disengaged.
rod-regulating. The regulating rod is a low-worth control rod used primarily to maintain an intended power level that need not have scram capability and may have a fueled follower. Its position may be varied manually or by the servo-controller.
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i scram time. Scram time is the elapsed time between the initiation of a scram signal full insertion of a control or safety device.
secured experiment. A secured experiment is any experiment that is held in a stationary position relative to the reactor by mechanical means. The restraining forces must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the experiment, or by forces that can arise as a result of credible malfunctione, secured shutdown. Secured shutdown is achieved when the reactor meets the requirements j of the definition of " reactor secured" and the facility administrative requirements for leaving the facility with no licensed reactor operator present.
j shall, should and may. The word "shall"is used to denote a requirement; the word "should" to l denote a recommendation; and the word "may" to denote permission, neither a requirement nor a recommendation.
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shutdown margin. Shutdown margin is the minimum shutdown reactivity necessary to provide l confidence that the reactor can be made subcritical by means of the control and safety systems (
starting from any permissible operating condition. It should be assumed that the most reactive I scrammable rod and all non-scrammable rods are in the!r most reactive position and that the j reactor will remain subcritical without further operator action. i shutdown reactivity. Shutdown reactivity is the value of the reactivity of the reactor with all
, control rods in their least reactive positions (e.g., inserted). The value of shutdown reactivity ;
includes the reactivity of allinstalled exp -iments and is determined with the reactor at ambient
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l 2.0 SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTINGS l
( 2.1 Temperature of the reactor fuel.
l Obiective: The objective of this specification is to define the maximum fuel temperature that can l be permitted with confidence that no damage to the fuel element will result.
Soecifications: The temperature in any fuel element in the OVAMC TRIGA reactor shall not exceed 500 oC under any conditions of operation.
l B331g A loss in the integrity of the fuel element cladding could arise from a buildup of excessive pressure between the fuel and the cladding if the fuel temperature exceeds the ;
safety limit. The pressure is caused by the heating of air, fission product gases, and hydrogen from the dissociation of the fuel-moderator. The magnitude of this pressure is determined by the temperature of the fuel element and by the hydrogen content. Experience with operation of TRIGA-fueled reactors at power levels up to 1500 kW shows no damage to the fuel due to :
thermally induced pressures, i l
The thermal characteristics of aluminum-clad TRIGA fuel elements using ZrH1.1 moderator l have been analyzed (S. C. Hawley and R. L. Kathren, NUREG/CR-2387, PNL4028, Credible
! Accident Analyses for TRIGA and TRIGA-fueled Reactors,1982). A loss-of-coolant analysis l showed that in a typical graphite-reflected Mark l TRIGA reactor fueled with 60 aluminum-clad i fuel elements (Reed College) the maximum fuel temperature would be less than 150 oC following infinite operation at 250 kilowatts terminated by the instantaneous loss of water.
l These temperatures are well below the region where thea+ p + y to a + 6 phase change occurs l in ZrH1.1 (560 oC).
2.2 Limiting Safety system Settings (LSSS)
Aoolicability: This specification applies to the reactor scram setting which prevents the reactor fuel temperature from reaching the aafety limit.
l Obiective: The objective of this specification is to provide a reactor scram to prevent the safety l
limit from being reached.
Soecification: The LSSS shall not exceed 20 kW as measured by the calibrated power ,
channels. The LSSS, which does not exceed 20 kW, provides a considerable safety margin. l One TRIGA reactor (General Atomics, Torrey Pines) showed a maximum fuel temperature of 350 oC during operation at 1500 kilowatts, while a 250-kilowatt TRIGA reactor (Reed College) showed a maximum fuel temperature of less than 150 oC (reported by S C Hawley, R. L.
- Kathren, NUREG/CR-2387, PNL4028 (1982), Credible Accident Analyses for.TRl(GA and i
! TRIGA-Fueled Reactors). A portion of the safety margin could be used to account for variations i of flux level (and thus the power density) at various parts of the core. The safety margin should be ample to compensate for other uncertainties, including power transients during otherwise :
steady-state operation, and should be adequate to protect aluminum-clad fuel elements from
! . cladding failure due to temperature and pressure effects.
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( ,/ 3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Reactor Core Parameters 3.1.1 Excess Reactivity Aoolicability: This specification applies to the reactivity of the reactor core in terms of the available excess above cold xenon free, critical condition.
Obiective: The objective is to prevent the fusi element temperature safety limit from being reached by limiting the potential reactivity available in the reactor for any condition of operation.
Specifications: The reactor shall not be operated unless the maximum available excess reactivity with secured experiments in place and the reactor in the cold, xenon free condition ,
shall be limited to $ 1.00.
Bases The maximum power excursion that could occur in the reactor would be one resulting from inadvertent rapid insertion of the total available excess reactivity. As demonstrated in Section 3.2.13 of the SAR, limiting the fuelloading of the reactor to $ 1.00 excess reactivity under cold clean, xenon free critical conditions will assure that the fuel temperature will not reach the critical temperature of 500 oC. ;
l 3.1.2 Shutdown Margin Acolicability: This specification applies to the reactivity minimum negativity by which the reactor core will be shutdown with the maximum worth control rod fully withdrawn.
ObjectivA The objective is to ensure that the reactor can be shutdown safely by a margin that is sufficient to compensate for the fai'ure of a control rod or the movement of an experiment.
Specification: The reactor shall be shutdown by more than $ 0.51 with the:
(1) the highest non-secured experiment in its most reactive state.
(2) the highest worth rod fully withdrawn (3) the reactor in the cold clean critical condition without xenon j 3.1.3 Core Configuration 3.1.3.1 Restrictions on removal of fuel element or control rod for inspection.
Acolicability: This soecification applies to the removal of a fuel element or control rod from the reactor core for the purpose of inspection.
Obiective: To establish reactivity limits to ensure that changing the core configuration by removing a fuel element or control rod would not lead to loss of control of the reactor, loss of 6
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fuel integrity, uncontrolled release of radioactivity or potential exposures exceeding 10 CFR 4
b Part 20.
l Soecifications: i (1) No fuel element shall be inserted or removed from the core unless the reactor is subcritical by more than the worth of the most reactive fuel element.
The shutdown margin shall be observed.
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(3) All control rods will be fully inserted into the core ;
Basis. These reactivity limits will ensure inadvertent criticality is not reached and that the shutdown margin condition is maintained.
I 3.1.4 Pool Water Level Aoolicability: This specification applies to the requirement for maintaining a minimum height of j water above the reactor core.
Obiective: To ensure that sufficient water covers the core to provide necessary shielding and acceptable limits of cooling temperature.
I Soecifications: A float alarm switch shall be operable so that if the water level drops to less
. than 12 feet above the core a visual and audible alarm will sound at the Medical Center
\ switchboard which is manned 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day, and a visual alarm will indicate on the reactor i
console.
Basis This specification is to ensure that if there was a large leak or fracture of the reactor i tank; attention would be called to the occurrence so that immediate action could be taken The effect of a complete loss of cooling accident is analyzed in Appendix C of the SAr Mee there I is a siphon break from the water skimmer 6" below the water surface, a drop fr. vel of 6" would cause the water circulation system to lose its prime and cease to operate. Tru are is L cooled by natural thermal convective flow.
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e 3.1.5 Fuel Parameters
- Acolicability
- This specification applies to conditions set forth for dealing wnh damaged fuel.
Obiective: The objective is to set limiting conditions for identifying and handling damaged fuel.
Specifications:
Individual fuel elements will be visually analyzed remotely to detect cladding (1) i deterioration that results from erosion, corrosion, or other damage.
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l (2) A clad defect will be indicated by release of fission products or visual observation. Any indication of the release of fission products by the facility
(/ monitoring instruments will be considered a clad defect and damaged fuel will be assumed to be in the core.
(3) No operation of the reactor will be performed with damaged fuel except to locate such fuel.
(4) A fuel element shall be removed from operation if the burnup of uranium-235 in the UZrH fuel matrix exceeds 50 percent of the initial concentration.
Basis; Visual inspection of fuel elements since 1963 and comparison with previous examinations of each element indicates that the method of examination is adequate to detect any defects. SAR, Section 3.2.13 discusses thermal cycling tests that have been performed to verify fuel matrix stability with resect to swelling or elongation. Simnad [GA-4314] has described these tests with temperatures in the range 500 oC to 725 oC. He has explained why there are no important changes in length or diameter of the test samples even though a small phase transition did occur at 653 oC (orthorhombic to tetragonal). For a TRIGA fuel with fuel temperatures 5 200 oC, there is no phase change or other transition to produce elongation or swelling in the fuel matrix.
3.2 Reactor Control and Safety Systems 3.2.1 Operable Control Rods Acolicability: This specification applies to the control rods used in the reactor core.
Objective: The objective is to determine that the control rods are operable by specification of apparent physical conditions, and to set limits on the scram times of the control rods.
Specifications:
(1) The reactor will contain 3 control rods (shim, safety and regulating) all of which wi!I have scram capabilities.
(2) The control rods will be composed of B4C powder or boron and its compounds contained in a suitable cladding material, such as aluminum or stainless steel to ensure mechanical stability during movement and to isolate the poison from the pool water environment.
(3) The maximum scram time for a fully withdrawn rod shall be 2 sec from the time of initiation of the scram signal to full insertion of the rod.
Basis The apparent condition and presence of the control rod assemblies will provide assurance that the rods will continue to perform reliably and as designed. The specification for l rod scram time assures that the reactor will shut down promptly when a scram signal is I
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initiated. However, as described in Section 3.2.13 of the SAR a rapid insertion of as much as l $2.00 of excess would terminate with no safety problem even without backup insertion of l
l 3.2.2 Reactivity insertion Rates
- Anolicability: This specification applies to the maximum rate of positive reactivity to the reactor.
Obiective: To assure that the reactor will start up on a controlled rate when the control rods are withdrawn.
l Soecifications:
(1) The maximum reactivity insertion rate of a standard control rod shall be less than ;
$0.10 per second.
l (2) Gang or multiple withdrawal of control rods is not allowed.
Basis: As described in Section 3.2.13 of the SAR a rapid insertion of as much as 5 2.00 of excess would terminate with no safety problem even without backup insertion of control rods, however, the above limit will assure that the reactor can be started up, or power changed at a controlled rate.
3.2.3 Scram Channels i
, Acolicability: This specification applies to required scram channels and set points.
Obiective: To determine the minimum safety system scrams that must be operable for the operation of the reactor.
Soecifications:
(1) The reactor shall not be operable unless the minimum safety channels are operable.
The following scram channels shall be operable:
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Table 1 Reauired Scrams Safety Channel jet Point Minimum number U reauired '
Percent oower' 100% licensed oower 1 Linear oower~ 100% licensed oower 1 lon chamber oower suoolv' Loss of hiah voltaae 1 Fission counter oower sucolv~ Loss of hiah voltaae 1 l Console scram button Not Acolicable 1 1
Maanet current key switch Not Acolicable 1 Watchdoc timer Not Acolicable 1 !
- lon chamber analog system
" Fission counter diaital system
- 1. Basis: The redundancy of scrams from the analog and digital systems provides protection from single-system failure and human error. The scrams minimize the probability reaching the safety limits. The power level scrams provide protection to ensure that the reactor can be shut down before the licensed power is exceeded. Experience with this and other !
TRIGA-fueled reactors, some of which are licensed to pulse, indicates a period scram would be extremely conservative and that the TRIGA-fueled reactors may be safely operated at periods considerably less than 7 seconds. This experience is summarized and analyzed in the report " Credible Accident Analyses For TRIGA and TRIGA-fueled Reactors, SC Hawley and RS Kathren, NUREG/CR-2387 (PNL-4028),1982. The major concern with minimum period is that the fuel not be allowed to approach the Safety Limit, which for our facility is
(]/
y 500*C for our mostly aluminum core. With our relatively low power (20KW) and our maximum core excess of $1.00 it is not feasible that temperatures would even approach the Safety Limit.
We routinely operate at 18 kW, which is 10% below the licensed power so the scram set points can be set at the licensed power. The manual scram allows the operator to shut down the system if an unsafe or abnormal condition occurs. In the event of failure of the power supply for the neutron detectors, operation of the reactor without adequate instrumentation is prevented. The NM 1000 through self-checking with a Watchdog Timer detects a failure of its critical task routines and will scram if these routines are not performed within 1.5 seconds.
3.2.4 Interlocks Apolicability: This specification applies to the mechanical limitations installed in the reactor control system.
Obiective: To ensure that there are sufhcient neutrons in the reactor core for the reactor to be started up in a controlled manner and to prevent a rapid insertion of too much excess reactivity.
Specifications: The reactor shall not be operate unless the following interlocks are operable:
O
Table 2 Reauired Interlocks I interlock Minimum Number Function N Remired Neutron count rate 1 Prevent rod withdrawal l (startup) (startup inhibit)if count- rate
< 2 cos Simultaneous manual 1 Prevent withdrawal i withdrawal of two rods Withdrawal of shim or reg i Prevent withdrawal rod with safety rod not all the way out or seated
- Withdrawal of safety rod with 1 Prevent withdrawal shim or rea rod not seated'
- May be defeated for contret rod calibration Haain; The startup countrate interlock requirement is to ensure that during startup there are -
adequate neutrons available to ensure that there is a continuous indication of the presence of neutrons in the neutron monitoring channels. The restriction on the removal of more than one l control rod is to limit the maximum positive reactivity insertion rate available for steady state :
operation As was shown in Section 3.2.13 of the SAR of as much as $ 2.00 of excess reactivity i would terminate with no safety problem even without backup insertion of control rods, however ,
this restriction will ensure that a rapid neutron multiplication does not occur. The bypassing of l the control rod interlocks is necessary in order to calibrate each individual rod. Since the O individual rods will be moved in small increments in order for them to be calibrated, and since as described in the basis to 3.2.4 above a $ 2.00 step reactivity insertion will not result in a hazard, it is concluded that temporally bypassing the rod interlocks for calibration purposes will i
be safe. -
3.2.5 Control Systems and Instrumentation Requirements of Operation Aoolicability: These specifications apply to measurements of reactor operating parameters.
Obiective: The objective is to determine the minimum instrument measurement and control systems to be operable for continued operation of the reactor Soecifications: The reactor shall not be operable unless the entire measuring channels as noted in Table 3 are operable, this includes readout meters, recorders and the protective functions they perform.
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) Table 3 Required Minimum Measuring Channels Measurina Channel Number Ooerable Function Startup 1 Monitor subcritical
- multiplication for j- startuo t Power Level (NM1000, 1 Input for safety power level j fission chamber) scram and to digital display
! unit and recorder Log N (NM1000, fission 1 Wide range power level and chamber) display on digital unit and on recorder e Period (NM1000, fission - 1 Input for period display on chamber) diaital unit and oeriod scram Per Cent Power (lon 1 Input for power level scram j chamber) and disclav on analoa meter Pool water 1 Display on analog meter *
- Temoerature j Pool Level 1 Alarms when waterlevelis
!- less than 12 ft above top of i core 1
l Basis: The minimum measuring chann. .s are sufficient to provide signals for automatic safety
} operation. Signals from the measuring ystem provide information to the control and safety j system for a protective action. The digital NM1000 instrumentation utilizing wide range fission
' and the analog percent power instrument provide diversity and redundancy for the measuring channels. Both the digital and analog channels are calibrated separately for power level. The
- pool water temperature is measured by means of a thermistor installed in cooling system loop
. as shown in Fig 3.8 of the SAR and allows the manual monitoring of bulk pool water. Since the output of the NM1000 is indicated on the digital d; splay unit, the signal to the recorders is 1
merely an analog signal in parallel to the digital signal. Failure of the recorder will not effect the safety of the reactor.
i l 3.3' Coolant System ,
Acolicability: These specifications apply to the quality of the coolant in contact with the fuel
- cladding, to the level of the coolant in the pool, and to the bulk temperature of the coolant.
] Obiective: The objectives of this specification are:
r (1) To minimize corrosion of the cladding of the fuel elements and minimize neutrons activation of dissolved materials.
For purposes of maintenance the inline thermistor may be replaced by thermistor placed in l the reactor tank and read on a separate meter.
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(2) To detect releases of radioactive materials to the coolant before such releases O become significant.
(3) To ensure the presence of an adequate quantity of cooling and shielding water in the pool, and goecifications:
(1) The conductivity of the pool water shall no exceed 5 mhos/cm averaged over one month.
(2) The pool water pH shall be in the range of 4 to 7.5.
(3) The amount of radioactivity in the pool water shall not exceed 0.1 pCi/mL.
(4) The water must cover the core of the reactor to a minimum depth of 15 feet during operation of the reactor.
(5) The bulk temperature of the coolant shall no exceed 35 oC during operation of the reactor.
Basis:
Increased levels of conductivity in aqueous systems indicate the presence of corrosion products and promote more corrosion. Experience with water quality control at many reactor facilities, including operation of the AJBRF since 1959, has shown that maintenance within the specified limit provides acceptable control. Maintaining low level of dissolved electrolytes in the pool water also reduces the amount of induced radioactivity, in turn decreasing the exposure of personnel to ionizing radiation during operation and maintenance.
Monitoring the pH of the pool water provides early detection of extreme values of pH, O which could enhance corrosion. In the event that the pH should exceed its specified limits the h reactor shall remain shutdown until the pH is within its acceptable range; typically this is remedied by changing the ion-exchange resin.
Monitoring the radioactivity in the pool water levels to provide early detection of nossible cladding failures. Limitation of the radioactivity according to this specification decreases the exposure of personnel to ionizing radiation during operation and maintenance in accordance with ALARA procedures.
Maintaining the specified depth of water in the pool provides shielding of the radioactive core which reduces the exposure of personnel to ionizing radiation in accordance with the ALARA procedures.
Maintaining the bulk temperature of the coolant below the specified limit assures minimal thermal degradation of the ion exchange resin.
3.4 Confinement System Acolicability: This specification applies to the confinement system of the reactor facility.
Objective,, To control confines of the Nuclear Reactor.
.. - -_ - - - . - . - - ~ _ _ _ . . . _ . - - -
I
,, Soecification:
(
(1) During operation of the reactor, movement of irradiated fuel, core or control rod work or any other manipulation involving the reactor, the exhaust fan and the fume hood that handles the"Ar exhaust from the pneumatic tube shall be in operation.
(2) The ventilation system shall keep the reactor laboratory at a slightly negative pressure.
Basis:
This specification ensures that the confinement is configured to control any releases of radioactive material during fuel handling, reactor operation, or the handling of possible airbome radioactive material in the reactor room.
3.5 Ventilation Systems l
Apolicability: This specification applies to the air ventilation conditions in the reactor area during reactor operation.
Obiective: The objective is to ensure that the ventilation system is in operation to mitigate the consequence of the possible release of radioactive materials resulting from reactor operation.
~Goecifications:
O Q (1) An exhaust fan with a flow rate of 2970 CFM together with two laboratory fume hood fans with a combined flow rate of at least 500 CFM shall be operable to exhaust the laboratory air to the environs during reactor operation.
(2) The output of the pneumatic tube shall be exhausted into one of the fume hoods and then to the roof of the Medical Center with a flow rate of nominally 250 CFM during reactor operation.
(3) The exhaust fan shall exhaust into the below ground water treatment pit and ultimately to ground level, while the pneumatic tube will exhaust on the roof of the Medical Center which is at least 371 feet above ground.
(4) The fume hood exhaust system for the pneumatic tube shall have an flow switch with an audible alarm that will indicate if the exhaust fan stops.
Basis: The specifications for exhaust ventilation and isolation of the reactor room provide for releases for both rou'ine and non routine operation conditions. Analyses in Appendix A and B of the SAR show that complying with the above specifications ensure that the doses to the public are well below regulatory limits for unrestricted areas and the potential doses to facility staff are within regulatory limits.
O 14 0
3.6 Radiation Monitoring Systems I
/m Apolicability: This specification applies to the radiation monitoring systems and effluent release limits in the reactor area during reactor operation.
l Obiective: To monitor the radiation and radioactivity conditions in the reactor area in order to control exposures or releases.
3.6.1 Monitoring Systems ,
Specifications: Radiation monitoring while the reactor is operating requires the minimum !
conditions shown in Table 4 below:
Table 4. Reauired Radiation Measurina Channels Channel Minimum number Function Set point equal or !
reauired less than i Area radiation 1 Alarm 2 mrem /hr I monitor-poollevel (aamma) !
Continuous Air 1 Alarm and recorded 2000 pCl/mi Monitor outout (1) When a specified monitor becomes inoperable operations may continue only if the monitor is replaced by a substitute or portable monitor. The replacement monitor i must perform essentially the same function until the origins: me:iitor is replaced j (generally not to exceed 1 week)
(2) The continuous air monitor shall sample the reactor room air within 5 meters of the pool at the pool access level. The alarm set point shall be equal to or less than a measurement concentration of 2000 pCi/ml.
P Basis:
(1) Monitoring Systems - The radiation monitoring provides information to operating personnel of impending or existing hazards from radiation so that there will be sufficient time to take he necessary steps to control the exposure of personnel and release of radioactivity or evacuate the facility. Alarm set points do not include measurement uncertainty. These values are measured values and not true values.
(2) Area radiation monitor- The gamma radiation monitor is part of the permanent installation and is installed at the edge of the pool approximately 18 inches above the top of the reactor tank grid cover. The alarm set point of 2 mr/hr enables the operator to monitor the radiation at the top of the reactor tank both from experiments being removed and from the reactor. The set point of 2 mr/hr is designed to allow the operator and radiation safety personnel ample waming to ensure that ALARA principles are complied with.
15
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(3) Air particulate radioactivity accumulates on the filter of a continuous monitor that records the radiation levels. An alarm set point informs the operator when the set point is s exceeded and the monitor meter and recorder are within view of the operator. The alarm set point causes the monitor to annunciate when it detects particulate activity below the occupational DAC values of Appendix B of 10 CFR 20.1001-20.2401 for the relevant isotopes in the range 84-105 and 124-149 as described in Section 4.3.2 and 7.4.1 of the SAR.
3.7 Experiments Acolicability: These specifications apply to experiments installed in the reactor and its experimental facilities.
Obiective: The objective is to prevent damage to the reactor or excessive release of radioactive materials in the event of an experimental failure.
Specifications: The reactor shall not be operated unless the following conditions goveming experiments exist:
3.7.1 Reactivity Limits 1
(1) The reactivity worth of any individual experiment shall not exceed $ 1.00. This includes both secured and moveable experiments as defined in Section 1 of these Tech Specs.
(2) The sum of the absolute reactivity worths of all experiments in the reactor and in the associated experimental facilities at one time shall not exceed $ 2.00.
(3) The reactor shall be shut down during the changing or moving on any secured experiment.
(4) The actual experiment worth shall be measured and recorded at the time of initial insertion of the experiment if the estimated worth is greater than $0.40. l Basis:
The intent of these specifications is to limit the reactiviy of the system so thea the Safety Limit would not be exceede even if the contribution to the total reactivity by the experimtn reactiviyt should be suddenly removed. In addition, these specifalcations limit power excusions ,
which might be induced by the changes in reactiviy due to inadvertent motion of an unsecured !
experiment. The safety sugnificance of significant reactivity insertions has been analyzed in SAR, section 3.2.12.
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3.7.2 Materials Aoolicability: These specifications apply to experiments installed in the reactor and its experimental facilities.
1 Obiective: The objective is to prevent the release of radioactive materialin the event of an l
experiment failure, either by failure of the experiment or subsequent damage to the reactor components Soecifications: The reactor shall not be operated unless the following conditions goveming experiment material exists:
(1) Experiments containing liquid, gas and potentially corrosive will be doubly encapsulated.
(2) Compounds highly reactio with water, potentially explosive materials, and liquid fissionable materials shall nv be irradiated in the reactor.
(3) Guidance for classification of materials shall be " Dangerous Properties of Industrial Materials" by N.I. Sax (Reinhold Publishing) or equivalent.
(4) The radioactive material content, including fission products of any experiment shall be limited so that the complete release of all gaseous, particulate, or volatile components from the encapsulation will not result in doses in excess of the annual limits stated in 10 CFR 20.
4 (5) if a capsule fails and releases material which could damage the reactor fuel or j
, structure by corrosion or other means, removal and physical inspection shall be performed to determine the consequences and need for corrective action. The i results of the inspection and any correctiv(action taken shall be reviewed by the i Reactor Director, or his designated attemate, and determined to be satisfactory !
before operation of the reactor is resumed. !
(6) No experiment should be performed unless the material content, with the exception ,
of trace constituents, is known. I (7) Fueled experiments shall not be irradiated in the reactor except for the activation of uranium foils for calibration or other purposes.
Basis:
(1) Double encapsulation requirements lessen the leakage hazards of experiment materials (2) The restriction on materials that are highly reactive with water, potentially explosive or liquid fissionable is intended to prevent damage to the reactor and the release of fission products at a level that has not been analyzed in the SAR.
l 17
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- j (3) The specification of a source for classifying materials assures that adequate information is available to assess the potential danger of materials.
(4) The specification on restriction of material content above is intended to reduce the likelihood that airbome activity in excess of the limits in Appendix B of 10 CFR 20 will be released to the reactor room or the atmosphere outside of the facility boundary. Radiation hazards are analyzed in Section 7.5 of the SAR.
(5) Operation of the reactor with the reactor fuel or structure damaged is prohibited to ,
avoid release of fission products.
(6) Knowing the material content of each experiment is necessary so as to determine the potential danger of irradiating the sample both within the reactor and upon removal.
(7) The restriction on fueled experiments is intended to reduce the severity of a possible
- release of these fission products, which pose a hazard to workers and the general 3
public.
3.7.3 Failure and Malfunction j Aoolicability: This specification addresses the failure and malfunction of experiments.
l l Obiective: To design experiments so that they will not contribute to the failure of other
- experiments, core components, or principal physical barriers, or to the uncontrolled release of j ( radioactivity.
4 l Soecifications: j (1) In calculations pursuant to 3.8.2(4) above, il an experiment fails and releases i radioactive gases to the reactor room or atmosphere, the following is assumed:
! a.100% of the radioactive gases or aerosols escape. i a
1
- b. If the effluent exhausts through a filter with 99% efficiency for 0.3-micron l particles, at least 10% of these vapors escape.
i
- c. For materials whose boiling point is above 130[F (540) the vapors of at least 10 il % of the materials escape through an undisturbed column of water above the core.
(2) Experiments shall be designed such that they will not contribute to the failure of other experiments, core components, or principal physical barriers or to the uncontrolled release of radioactivity.
333 1g; These specifications establish as.i mptions for calculating the activity that could be released under normal operating conditions, accident conditions in the reactor, and accident 18 4
L.
conditions in the experiment. It also specifies that experiments must be designed so as to t prevent damage to the reactor or other experiments. Restrictions on experiments are provided
( in detail in SAR, Sections 7.2.3 and 7.5.
4.0 SURVEILLANCE REQUIREMENTS ]
The allowable surveillance intervals shall not exceed the following:
(a) Five-year: interval not to exceed six years (b) Biennial: interval not to exceed two and one-half years -
(c) Annual: interval not to exceed 15 months l (d) Semiannual: interval not to exceed seven and one-half months j (e) Quarterly: interval not to exceed four months I (f) Monthly: interval not to exceed six weeks (g) Weekly: interval not to exceed ten days (h) Daily: must be done during the calendar day.
l Established frequencies shall be maintained over the long term. For example, any monthly surveillance shall be performed at least 12 times during a calenaar year of normal operation. If the reactor is not operated for a period of time exceeding any required surveillance interval, that surveillance task shall be performed before the next operation of the reactor. Any surveillance tasks, which are, missed more than once during such a shutdown interval need be performed only once before operation of the reactor. Surveillance tasks scheduled daily or weekly which cannot be performed while the reactor is operating may be postponed during continuous l O operation of the reactor over extended times. Such postponed tasks shall be performed following shutdown after the extended period of continuous operation before any further operation, where each task shall be performed only once no matter how many times that task has been postponed.
I 4.1 Reactor core parameters 4.1.1 Excess Reactivity Acolicability: This specification applies to the measurement of reactor excess reactivity.
Objective: The objective is to periodically determine the changes in core excess reactivity to assure compliance with Section 3.1.1 of the Tech Specs.
Specifications: Excess reactivity shall be determined on an average annually (at intervals not to exceed 15 months) and after changes in either the core, in-core experiments, or control rods for which the predicted change in reactivity exceeds the absolute value of the required shutdown margin.
Basis: Annual determination of excess reactivity and measurements after reactor core or control rod changes are sufficient to monitor significant changes in the core excess reactivity, v
1
i 4.1.2 Shutdown Margin Ar-:4c+bintv: This specification applies to the measurement of reactor shutdown margin.
Obiehtive: The objective is to periodically determine the core shutdown reactivity available for reactor shutdown.
1 Soecifications: The shutdown margin should be determined on an average annually (at intervals not to exceed 15 months) and after changes in either the core, in-core experiments, or control rods.
Basis: Annual determination of shutdown margin and measurements after changes in reactor core, in-core experiments, or control rods are sufficient to monitor significant changes in the l core shut down margin.
l 4.1.3 Core Configuration Aoolicability: This specification applies to the configuration of the core.
Obiective: The objective is to assure that the core config'uration remains as specified in the SAR and in Section 5 of the Technical Specifications.
Specifications: A visual observation of the reactor core will be made before each initial daily ,
startup of the reactor and proof of the observation recorded in the daily startup checklist. The !
observation will assure that the fuel elements, control rods, detectors and experimental facilities are in place as specified in the SAR and the core is free of any extraneous material.
Basis: Daily visual surveillance of the reactor tank is sufficient to assure that there have been i no change in the core configuration and to verify compliance with all applicable specifications.
4.1.4 Fuel Element inspection Ano!!cibility: This specification applies to the inspection requirements for the fuel elements.
Obiective: The objectiveJs to verify the continuing integrity of the fuel element cladding.
Soecifications: The reactor fuel elements shall be examined for physical damage by a visual inspection at least once each five years, with at least 20 percent of the fuel elements examined each year. Observation will include inspection for swelling, cracks, corrosion and pitting.
The reactor shall not be operated with damaged fuel except to detect and identify damaged fuel for removal. A fuel element shall be removed from the core if a clod defect exists as indicated by the release of fission products.
l Ragig The frequency of examinaticn allows each element to be inspected ever 5 years.
l Previous inspection experience has shown that this frequency ofinspection is adequate and
- thus reduces the risk of accident or damage fuel to handling.
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l 4.2 Reactor Control and Safety System 4.2.1 Reactivity Worth of Control Rods Aoolicability: These specifications apply to the surveillance requirements for reactivity control of experiments and systems. j Obiective: The ob>ctive is to measure and verify the worth of the control rods.
Soecifications: The integral worth of all control rod and safety rods will be determined on an )
average annually (at intervals not to exceed 15 months) and after changes of the core or l control rods. ;
Basis The reactivity worth of the control rodo is measured to ensure that the required shutdown margin is available and to provide an accurate means for determining the reactivity worth of experiments inserted in the core. Past experience with TRIGA reactors gives assurance that measurement of the reactivity worth on an annual basis is adequate to ensure no significant changes in the shutdown margin.
4.2.2 Rod Maximum Reactivity insertion Rate Apolicability: This specification applies to surveillance of the control rods.
[
\c Obiective: To establish operable conditions of the control rods by periodic measurement of the rod withdrawal and insertion speeds.
Soecifications: Withdrawal time of the Safety rod will be measured daily prior to the first start
. up of the reactor and withdrawal times of the other two control rods will be measured on and average annually (at intervals not to exceed 15 months). Insertion speeds of all control rods will be measured annually. Deviations that are significant from acceptable standards will be promptly corrected.
Basis Measurement of withdrawal times will give an indication if there is any malformation of the control rods due to operation and assure operable performance of the rods. Measurement of insertion speeds will assure compliance with Section 3.2.2 of the Technical Specifications. ;
4.2.3 Scram Times of Control Rods Aoolicability: This specification applies to the requirements for measurement of scram times of
. the control rods. 1 Obiectivt To establish the operable conditions of the control rods by periodic measurement of th@ ocram times. ,
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Soccifications: Scram times, of all control and safety rods shall be measured on an average 4 annually (et intervals not to exceed 15 months) or whenever work is done on the rods or rod drive systems.
Elagia; Measurement of scram times on an annual basis is a check not only of the scram system electronics, but also is an indication of the capability of the control rods to perform properly. The measurement of scram times will also ensure compliance with Section 3.2.1 of the Technical Specifications.
4.2.4 Scram and Measuring Channels Aoolicability: This specification applies to the logic of the reactor control system.
Obiective: To set surveillance requirements for performing channel checks of all scram and power measuring channels required by the Technical Specifications.
Soecifications: Channel tests of all scram channels and interlocks required by Sections 3.2.3 and 3.2.4 of the Technical Sper*hcafons shall be required before each reactor startup at the beginning of each opersting day after a secured shutdown. The fission counter power supply, the watch dog timer and the pool level alarm which shall be tested monthly.
Basis The above tests will ensure that scram systems and interlocks are operable on a daily basis or prior to an extended run. Previous operating experience has shown that the monthly surveillance of the fission counter power supply, the watch d 3 timer and the poollevel alarm is sufficient to ensure operation. The extended surveillance interval for the power supp!y and watchdog timer is recommended to prevent undue stress on the systems.
4.2.5 Thermal Power Calibration Am!imbHily.; This specification applies to determining the thermal power of the reactor.
Obiective: Thc V .se is to rti r.f/ the interval and method used to thermally calibrate the reactor. i l
, Soecification: Calibration of the power measuring channels shall be done by the calorimetric method annually or after any modification or repair of the measuring system or change in the fuel configuration.
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Basis: The power level channel calibration will be done by the calorimetric method. Calibration will ensure that the reactor will be at the proper power level, e.g. Indicated measurement will match actual thermal value, and that the requirements nf the fecility license will be complied with.
4.2.6 Rod Inspection
. Am!!mbility: This spec'fication applies to the physical inspection of the poison section of the control rods, and to the rod drive and scram mechanisms
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m Obiective: To assuro that the integrity of the control rod cladding is maintained and to look for any degradation of the rod drive and scram mechanisms.
Specifications: The control rods shall be visually inspected for deterioration on a average annually (at intervals not to exceed 15 months). During this inspection the rod drive and scram mechanisms shall also be inspected.
Basis The visual inspection of the control rods is made to evaluate any corrosion or wear caused by operation of the reactor.
inspection of the rod drive and scram mechanism for excessive wear and deterioration will ensure that a malfunction is not forthcoming. )
4.3 Coolant Systems 4.3.1 Analysis of Coolants for Radioactivity Applicabi!!!y.,. This specification applies to the monitoring of the coolant for radioactivity.
Obiective: To establish trends to quickly identify fuel or heat exchanger failure.
Specifications: The reactor water shall be sampled for gross activity on an average monthly (at intervals not to exceed 6 weeks) and for isotope idenbfication on an average quarterly at' (intervals not to exceed four months).
Basis Periodic analysis of the reactor water will establish a trend to quickly identify fuel or heat ;
exchanger failure. '
4.3.2 Condu::tivity and pH' Aoolicability: This specification applies to the surveillance of primary water quality.
Obiective: To ensure that the water quality does not deteriorate thereby causing corrosion of the reactor components.
Soecifications:
(1) Conductivity shall be measured weekly.
(2) pH shall be measured at least once every month.
Basis: Surveillance of the condition of the primary reactor coolant on a regular basis will ensure
. that the water quality is sufficient to control the corrosion of such components as the reactor
. fuel cladeing, structure and pool and to maintain clarity of the reactor water. Monitoring of the pH of the water over many years has shown that the value varies very slightly and consequently analysis at the intervals specified above is sufficient to detect a change.
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4.4 Confinement l
AndicaN!ity:~ This specification applies to the surveillance requirements of the confinement system of the reactor facility.
l Obiective: To assure that the confinement system as specified in the SAR and Section 3.4 of l the Technical Specifications are complied with.
Specifications:
(1) A check that the doors to the reactor facility will be closed when the reactor is in l operation except for normal entry.
(2) A daily check of the ventilation system operability shall be made prior to reactor eperation.
Basis: !n accordance with the statements made in the SAR the only confinement necessary for the reactor facility is proper ventilation and a secure facility. The above surveillance requirements will assure compliance.
' 4.5 Ventilation Systems Acolicability: This specification applies to the ventilation system within the reactor facility.
Obiective: To assure that the ventilation system is operating as specified in Section 3.1 of the SAR.
Soecifications:
(1) The fume hood exhaust system audible alarm for the pneumatic tube, as specified in TS Section 3.5 (4), shall be tested on an overage weekly (at intervals not to exceed ten days) and following repair and maintenance.
(2) The automatic absolute damper, as specified in Section 3.5 (6), shall be tested on an average m'onthly (at intervals not to exceed 6 weeks) and following repair or maintenance.
l Spain; Testing of the above items will assure that the radioactive concentration in the reactor room and that exhausted to the environs are as specified in the SAR.
4.6 Radiation Monitoring Systems and Effluents l 4.6.1 Monitoring Systems ADNk+Niity: This specification applies to the surveillance conditions of the radiation monitoring channels. l
. Qhjective To assure that the radiation monitons are functional.
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Specifications:
b V (1) A channel check shall be performed daily, before reactor startup, of the area and inline fission product monitor.
(2) All radiation monitos !ncluding the CAM will be calibrated annually (at intervals not to exceed 15 months,) and after maintenance, according to the manufacturer's recommendations.
Basis: Periodic calibration and frequent check are specified to maintain reliable performance of the radiation monitoring instruments. Calibration and check frequencies follow the general recommendation of guidance documents from the manufacture.
5.0 DESIGN F5ATURES 5.1 Site and Facility Description Specifications:
(1) The reactor shall be housed in a room in the basement of the Omaha Department of Veterans Affairs Medical Center. The room shall be considered a restricted area with locked doors and entrence controlled by reactor laboratory personnel.
(2) The TRIGA reactor is assembled in a below ground shie!d and pool structure with only vertical access to the core.
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(3) The minimum number of fans necessary for the facility are one nominal 2,970 CFM exhaust fan and two fume hood fans located on the roof of the Medical Center. In addition there is one electrically operated absolute damper that prevents Medical Center forced air from entering the reactor confinement.
(4) The minimum free volume in the reactor area shall be 25.000 ft'.
5.2 Reactor Coolant System Apolicability: This specification applies to the AJBRF.
Obiective: The objective of this specification is to define the characteristics of the cooling system of the reactor.
Specifications: The reactor core shall be cooled by natural convective water flow.
l Basis: Experience has shown that TRIGA reactor operating at power levels up to 1000 kilowatts can be cooled by natural convective water flow without damage of the fuel elements.
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5.3 Reactor Core and Fuel
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5.3.1 Reactor Core Specifications: The reactor core shall consists of a compact array of standard TRIGA fuel elements, graphite dummy elements, 3 boron carbide control rods, control rod guides, a startup neutron source, and irradiation facilities. The fuel elements are spaced so that about 33% of the core volume is occupied by water, yielding a fuel-to-water ratio resulting in a critical mass near the minimum value for 20% enriched uranium fuel. The elements are held in concentric rings by an upper and lower grid plate. The reactor currently requires 56 fuel elements but this number may change depending on the burn-up of the fuel. The balance of the 90 fuel element positions in the grid are occupied by experimental facilities or graphite-reflector elements. The latter are elemer.ts in which the U-Zrh, fuel is replaced by graphite.
i 5.3.2 Reactor Fuel Soecifications: The standard TRIGA fuel element at fabrication shall have the following characteristics:
(1) Uranium content - maximum of 9.0 % uranium enriched to a nominal 20 % uranium-235. _
(2) Hydrogen-to-zirconium atom ratio (in the ZrH,): nominal 1.0 for aluminum clad elements and 1.7 for stainless steel clad elements.
p (3) Cladding: Aluminum, nominal 0.076 cm thick. Stainless steel, nominal 0.05 cm thick. l i
(4) Any bumable poison shall be any .otegral part of the as-manufactured fuel element. l A burnable poison is defined as a material fixed in place in the core for the specific purpose of compensating for fuel bumup and/or other long-term reactivity adjustment.
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' 5.3.3 Control Rods Sgecifications: The reacior shall have three control rods (safety, shim and regulating). All three control rods shall have scram capability and contain borated graphite, B4C powder, or boron and its compounds as a poison.
5.3.4 Fissionable Material Storage Soecifications: Fuel not in the reactor core shall be stored in a geometrical array where L ;
is no greater than 0.9 for all conditions of moderation and reflection using light water, except in ;
j cases where an approved fuel shipping container is used, then the L for the container shall j apply.
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6.0 ADMINISTRATIVE CONTROLS l
6.1 Organization
! 6.1.1 Structure The facility chall be under the control of the Director / Reactor Supervisor. The management for operation of the facility shall consist of the organizational structure as shown in Figure 6.1.
6.1.2 Responsibility i
(1) Responsibility for the safe operation of the reactor facility shall be with the chain of command established in Figure 6.1. The Director / Reactor Supervisor shall be
! responsible to the Medical Center Director and the Associate Chief of Staff for Research for safe operation and maintenance of the reactor and its associated equipment. Individuals at the various management levels, in addition to having responsibility for the policies and operation of the reactor facility, shall be
{ responsible for safeguarding the public and facility personnel from undue radiation J exposure and for adhering to all requirements of the operating license or charter I
and technical specifications.
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t-DIRECTOR OMAHAVA PMDiCAL CENTER 23 c
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- Responsible for Focility Operation
1 O (2) In allinstances, responsibilities of one level may be assumed by desigr.ated altemates or by higher levels, conditional upon appropriate qualifications.
6.1.3 Staffing (1) The minimum staffing when the reactor is not secured shall bc:
- a. A licensed reactor operator in the control room.
- b. A second designated person present at the facility complex able to carry out prescribed written instructions. Unexpected absence for as long as two hours to acccmmodate a personal emergency may be acceptable provided immediate action is taken to obtain a replacement.
- c. A designated Senior Reactor Operator (SRO) shall be readily available on call.
"Readily Available on Call" means an individual who (1) has been specifically designated and the designation known to the operator on duty, (2) keeps the operator on duty informed of where he may be rapidly contacted and the phone number, and (3) is capable of getting to the reactor facility within a reasonable time under normal conditions (e.g. 30 minutes or within a 15-mile radius).
(2) A list of facility personnel by name and telephone number shall be readily available in the control room for use by the operator. The list shall include:
A a. Management Personnel
- b. Radiation Safety personnel
- c. Other operations personnel (3) Events requiring the presence at the facility of a Senior Reactor Operator:
- a. Initial startup and approach tc power
- b. All fuel or control-rod relocations within the reactor core region
- c. Relocation on any in.-core experiraent with a reactivity worth greater than one dollar.
- d. Recove,ry from unplanned or unscheduled shutdown or significant pov. sr reduction (In these instances documented concurrence from a Senior Reactor Operator is required) 6.1.4 Selection and Training of Personnel ;
Training and requalification of personnel shall be in compliance with 10 CFR Part 55. i Additional guidance for selection, training and requalification of operators may be found in ANSl/ANS 15-4-1988, " Selection and Training of Personnel for Research Reactors" l
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6.2 Review and Audit A
j 6.2.1 Composition and Qualificatioris (1) The Reactor Safeguards Committee (RSC) shall function to provide independent review and audit of facility activities and be composed of a minimum of 6 members j- including the following:
Chairman Chief of Staff Member (ex-officio) Radiation Safety Officer Member Reactor Director / Supervisor
, (2) A minimum qualification for persons on the RSC shall be 5 years of professional l work expericace in the discipline or specific field he represents. A baccalaureate l l degree may fulfill 4 years of experience.
l (3) Qualified and approved altemates may serve in the absence of regular members.
No more than two attemates shall participate on a voting bas ls in RSC activities at any one time.
I (4) Members and altemates shall be appointed by and report to Level 1 management.
6.2.2 Charter and Rules
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The operations of the Reactor Safeguards Committee shall be in accordance with an established charter or directive including provisions for:
(1) Meeting frequency - at least on an average semiannually.
(2) Quorums - not less than one-half of the membership where the operating staff does not constitute a majority.
(3) Use of subgroups (4) Dissemination, review, and approval of minutes in a timely manner (within a month following the meeting).
6.2.3 Review Function The review function shall include facility operation related to reactor and radiological !
safety. The following items shall be reviewed: )
1 (1) . Determinations that proposed changes in equipment, systems, tests, experiments, or procedures do not involve an unreviewed safety question as required by 10 CFR
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(2) All new procedures and major revisions thereto having safety significance, proposed changes in reactor facility equipment, or systems having safety significance 1
(3) All new experiments or classes of experiments that could affect reactivity or result in the release of radioactivity (4) Proposed change in technical specifications, license, or charter j (5) Violations of technical specifications, license, or charter. Violations of intemal procedures or instructions having safety significance. ,
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! (6) Operating abnormalities having safety significance.
(7) Reportable occurrences listed in 6.7.2 (8) Audit reports
- A written report of minutes of the findings and recommendations of the review group shall be submitted to Level 1 management and the review and audit group members within a month l after the review has been completed.
6.2.4 Audit Function The audit function shall include selective (but comprehensive) examination of operating l
f records, logs, and other documents. Discussions with cognizant personnel and observation of operations should be used also as appropriate, in no case shall the individual immediately responsible for the area perform an audit in that area. The following items shall be audited:
(1) Facility operations for compliance to the technical specifications and applicable
! license or charter conditions annually.
(2) The retraining and requalification program for the operating staff, on an average of at least once every other calendar year (intervals between audits not to exceed 30 months)
! (3) The results of' action taken to correct those deficiencies that may occur in the reactor facility equipment, systems, structures, or methods of operation that affect reactor safety, on an average of at least once per calendar year (intervals between audits not to exceed 15 months) 1 (4) The reactor facility emergency plan, and implementing procedures on an average of at least once every other calendar year (at intervals between audits not to exceed l 30 months) l Deficiencies uncovered that affect reactor safety shall immediately be reported to Level 1
- management. A written report of the findings of the audit shall be submitted to Level 1 I !
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j management and the review and audit group within three months after the audit has been 4
j completed.
6.3 Radiation Safety (1) The radiation safety program must comply with the requirements of 10 CFR 20.1001 to 20.2401. Additional guidance for the radiation safety program may be found in ANSI.ANS 15.11-1993 " Radiation Protection at Research Reactor 3 Facilities" 1
(2) The Radiation Safety Officer or his designate shall be assigned the responsibility for
, implementing the radiation protection program at the reactor facility using the above guidelines (3) The Radiation Safety Officer shall report to Level i management through the Chief l of Staff.
(4) Management is committed to practice an effective ALARA program that is ainied at making every reasonable effort to maintain radiation exposure as far below the ;
facility staff, facility users, general public and the environment. '
6.4 Proccdures
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i Written procedures shall be reviewed and approved by the Reactor Director / Supervisor j and reviewed by the Reactor Safeguards Committee prior to initiation of the fol!owing activities.
1 1 (1) Startup, operation, and shutdown of the reactor 4
(2) Fuelloading, unloading, and movement within the reactor l i
j (3) Maintenance of major components of systems that could have an effect on reactor safety
- (4) Surveillance c, hecks, calibrations, and inspections required by the technical j specifications or those that may have an effect on reactor safety ;
i l (5) Personnel radiation protection, consistent with applicable regulations or guidelines.
. The procedures shall include managements commitments and programs to maintain exposures and releases as low as reasonably achievable (ALARA) in
, accordance with the guidelines of ANSI /ANS-15.11-1993, " Radiation Protection at e
Research Reactor Facilities".
(6) Administrative controls for operation and maintenance and for the conduct of irradiations and experiments that could affect reactor safety or core reactivity.
(8) Implementation of required plans such as emergency or security plans.
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[ (9) Any additional plans that may be deemed necessary for operation of the facility.
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Substantive changes to the procedures shall be made effective only after documented review by the Reactor Safeguards Committee and approval by the Reactor Director /Supertisor.
Minor modifications that do not change their original intent may be made by the Reactor Director /Gupervisor (Level 3), but the modifications must be approved by the Reactor Safeguards Committee (Level 2) within 14 days if not a unreviewed safety question.
6.5 Experiment Review and Approval Approved experiments shall be carried out in accordance with written procedures properly reviewed and approved.
(1) All new experiments or class of experiments shall be reviewed by the Reactor Safeguards Committee and approved in writing by the Committee and the Reactor Director / Supervisor or designated alternates prior to initiation. The review and approval shall be consistent with the guidelines provided in ANSI /ANS 15.1-1993, Section C.3 of Regulatory Guide 2.2 and Regulatory Guide 2.4 " Review of Experiments for Research Reactors".
(2) Substantive changes to previously approved experiments shall be only after review by the Reactor Safeguards Committee and approved in writing by the Reactor
/' Director / Supervisor or designated alternate. Minor changes that do not significantly
( alter the experiment may be approved by the Reactor Director / Supervisor or a designated shift Senior Reactor Operator.
6.6 Required Actions j 6.6.1 Required Action to be Taken in Case of Safety Limit Violation l i (1) The reactor shall be shutdown, and reactor operation shall not be resumed until
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authorized by the Nuclear Reguictory Commission.
(2) The safety limit violation shall be reported to the Reactor Director / Supervisor or j i designated attemates.
i j (3) The safety limit violation shall be reported to the Nuclear Regulatory Commission.
! (4) A safety limit violation report shall be prepared. The report shall describe the j following:
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! a. Applicable circumstances leading to the violation including, when known, the l cause and contributing factors J
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, b. Effects of the violation upon reactor facility components, systems, or structure
[ and on the health and safety of personnel and the public.
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- c. Corrective action to be taken to prevent recurrence l The report shall be reviewed by the Reactor Safeguards Committee and any follow-up report shall be submitted to the Nuclear Regulatory Commission when authorization is sought to resume operation.
! 6.6.2 Action to be Taken in the Event of an Occurrence of the Type Identified in 6.7.2(1)b, and 6.7.2(1)c.
(1) Reactor conditions shall be retumed to normal or the reactor shall be shut down. A retum to normal event shall only occur when the reactor scrams due to a known cause such as an electric transient. Reactor shut down is defined in Section 1.1 of the Technical Specifications. If it is necessary to shut down the reactor to correct the occurrence, operation shall not be resumed unless authorized by the Reactor Director / Supervisor or designated altemates.
(2) Occuirence shall be reported to the Reactor Director / Supervisor or designated attemates and to the Nuclear Regulatory Commission as required (3) Occurrence shall be reviewed by the Reactor Safeguards Committee at their next scheduled meeting.
6.7 Reports 6.7.1 Operating Reports Routine operating reports covering the operation of the facility during the previous calendar year shall be submitted before March 31 to the Nuclear Regulatory Commission, Attn:
Document Control Desk, Washington, D.C.
(1) A narrative summary of reactor operating experience including the energy produced by the reactor (2) The unschedyled shutdowns including, where, applicable, corrective action taken to preclude recurrence.
(3) Tabulation of major preventive and corrective maintenance operations having safety significance (4) A tabulation, as required by 10 CFR 50.59, of major changes in the reactor facility and procedures, and tabulation of new tests or expe.-iments, or both, that are significantly different from those performed previously and are not described in the Safety Analysis Report, including a summary of safety evaluation leading to the conclusions that no unreviewed safety questions were involved 34
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(5) A summary of the nature and amount of radioactive effluents released or discharged I to environs beyond the effective control of the VA Medical Center as determined at or before the point of such release or discharge. The summary shall include to the extent practicable an estimate of individual radionuclides present in the effluent. If the estimated average release after dilution is less than 25% of the concentration allowed or recommended, a statement to this effect is sufficient.
(6) A summarized result of environmental surveys performed outside the facility (7) A summary of exposures received by facility personnel and visitors where such exposures are greater than 25 % of that allowed or recommended 6.7.2 Special Reports Special reports shall be used to report unplanned events as well as planned major facility and administrative changes.
t (1) There shall be a report not later than the following working day by telephone to the l NRC Operations Center and the Region IV staff, and confirmed in writing FAX to l the NRC Document Control Desk, l
l a. Violation of safety limits
- b. Release of radioactivity from the site above allowed limits l
- c. Any of the following:
(i.) Operation with actual safety system settings for iequired systems ,
l less conservative than the limiting safety-system settings specified in the technical specifications l
(ii.) Operation in violation of limiting conditions for operation established in the technical specifications unless prompt remedial action is taken (iii.) A reactor safety system component malfunction which renders or could render the reactor safety system incapable of performing its intended safety function unless the malfunction or condition is discovered during maintenance tests or periods of reactor shutdown.
l (Note: Where components or systems are provided in addition to those required by the technical specifications, the failure of the extra components or systems is not considered reportable provided that the minimum number of components or systems specified or required perform their intended reactor safety function)
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l (iv.) An unanticipated or uncontrolled change in reactivity greater than f
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one dollar. Reactor trips resulting from a know cause are excluded.
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(v.) Abnormal and significant degradation in reactor fuel or cladding, l or both, coolant boundary, or containment boundary (excluding '
minor leaks) where applicable, which could result in exceeding prescribed radiation exposure limits of personnel or environment, or both (vi.) An observed inadequacy in the implementation of administrative or procedural controls such that the inadequacy causes or could ha e caused the existence or development of an unsafe condition with regard to reactor operations.
(2) A written report within 30 days to the NRC address in the first paragraph of this section is required for:
- a. Permanent changes in facility organization involving Level 1 or 2 personnel as described in Figure 6.1
- b. Significant changes in the Safety Analysis Report 1
6.8 Records
( Records may be in the form of logs, data sheets, or other suitable forms. The required information may be contained in single or multiple re:ords, or a combination thereof.
l 6.8.1 Records to be Retained for a Period of at least Five Years or for the Life of the ;
- Component involved if Less than Five Years.
(1) Normal reactor facility operation (but not including supporting documents such as checklists, log sheets, etc., which shall be maintained for a period of at least one year)
(2) Principal maiqtenance operations (3) Reportable occurrences (4) Surveillance activities required by the Technical Specifications (5) Reactor facility radiation and contamination surveys where required by applicable regulations (6) Experiments performed with the reactor ,
I (7) Fuel Inventories, receipts, and shipments 1
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l l (8) Approved changes in operating procedures (9) Records of meeting and audit reports of the review and audit group.
6.8.2 Records to be Retained for at least One Certification Cycle Records of retraining and requalification of certified operations personnel shall be maintained at all times the individualis employed or until the certification is renewed.
6.8.3 Records to be Retained for the Lifetime of the Reactor Facility Applicable annual reports, if they contain all of the required information, may be used as records in this rection.
(1) Gaseous and liquid radioactive effluents released to the environs (2) Off-site environmental-monitoring surveys required by the Technical Specifications (3) Radiation exposure for all personnel monitored (4) Drawings of the reactor facility 37
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