ML20209C923
| ML20209C923 | |
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|---|---|
| Site: | MIT Nuclear Research Reactor |
| Issue date: | 07/08/1999 |
| From: | NUCLEAR REACTOR LABORATORY |
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Text
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O MASSACHUSETTS INSTITUTE OF TECHNOLOGY NUCLEAR REACTOR LABORATORY 1
(MITR-III) l TECHNICAL SPECIFICATIONS (Rev. 0)
July 8,1999 l l
Massachusetts Institute of Technology l Cambridge, Massachusetts l
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i Technical Specifications l
for The~ Massachusetts Institute of Technology Research Reactor I
MITR III i
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- 1. Introduction - > '1-1 '
.1.1 Scope 1 '
1.2 Application 1-1 1.2.1, Purpose -
1-1 1.2.2 Format 1-l' 1.3 Definitions 1-2 1.3.1 Channel .1-2 .
1.3.2 ChannelCalibration 1-2 1.3.3 Channel Check j ' 'l-2 1.3.4 ChannelTest 1-2; 1.3.5 Containment 1-2
' 1-3 l.3.6 ~ Containment Integrity -
1.3.7 Damaged Fuel 1 1.3.8 Excess Reactivity 1-3 1.3.9 Experiment 1-3 1.3.10 Experimental Facility 1 '
1.3.11 Frequency 1-4 1.3.12 Immediate 1-4 1.3.13 Inadmissible Sample Materials 'l-4 1.3.14 Independent Experiments ' l-4 '
l.3.15 Irradiation ' '14 1.3.16- Irradiation Series 1-5 1.3.17 Measured Valuei 1-5
' 1.3.18 Movable Experiment 1-5 l
1.3.19- 'Non-Secured Experiment i 1-5
,1.3.20 Operable ~ 1 <
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1.3.21 Opercing 1-5 1.3.22 Potential Reactivity Wonh of an Experiment 1-5 1.3.23 Protective Action 1-6 1.3.24 Reactivity Wonh of an Experiment 1-7 1.3.25 Reactor Oi crating 1-7 1.3.26 Reactor Operator 1-7 1.3.27 Reactor Safety System 1-7 1.3.28 Reactor Secured 1-7
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1.3.29 Reactor Shutdown 1 1.3.30 Reference Core Condition 1-8 1.3.31. Regulating Rod 1-8 1.3.32 Reponable Occurrence 1-8 1.3.33 Research Reactor 1-9 1.3.34 Review and Approve 1-9 1.3.35 Safety Analysis Report 1-10 1.3.36 Safety Channel 1-10 1.3.37 Scram Time 1-10 1.3.38 Secured Experiment 1-10 1.3.39 Secured Shutdown 1-10 1.3.40 Senior Reactor Operator 1-10 1.3.41 Shall, Should, and May. 1-11 1.3.42 Shim Blade 1-11 1.3.43 Shutdown Margin 1-11 1.3.44 Shutdown Reactivity 1-11 1.3.45 True Value 1-12 1.3.46 Unscheduled Shutdown 1-12 n
I,_)l 2. = Safety Limits and Limiting Safety System Settings 2-1
Y i2.1 . Safety Limits 2-1:
2.2 Limiting Safety System Settings (LSSS) 2-5
- 3. Limiting Conditions for Operation 3-1.
3.1 Reactor Core Parameters :3-2 3.1.1 Excess Reactivity ' 3-2 --
3.1.2 Shutdown Margin 3-4.
Maximum Safe Step Reactivity Addition 3 a 3.1.3' 3.1.4 Core Configurations 3-7 3.1.5 Reactivity Coefficients 3-12 3.1.6 Fuel Parameters 3-13 3.2 Reactor Control and Safety System' .3-16 ll Operable Control Devices '3-16
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3.2.1 3.2.2 Reactivity Insertion Rates and Automatic Contml 3-17.
3.2.3 Reactor Protection System 3-20 3.2.4 Control System Interlocks. 3-24 3.2.5 Backup Shutdown Mechanisms !3-26 3.2.6 Bypassing Channels 3-27 3.2.7 Control Systems and Instrumentation Requirements for Operation 3-28 3.3 Coolant Systems 3-30 3.3.1 Natural Convection and Anti-Siphon Valves 3-30 '
1 3.3.2 .H2 Concentration Limit 3-32 3.3.3 D 2Concentration Limit and Recombiner Operation 3-34 3.3.4 Emergency Cooling Requirements '3-37
'3.3.5 Coolant Radioactivity Limits '3-39' j 3.3.6 . Primary Coolant Quality Requirements :-
"3-42 .
- / 3.4 . Reastor Containment Integrity and Pressure Relief System. 441 3.5 Ventilation System 3-47--
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l 3.6' ' Emergency Power. 3-48 3.7 Radiation Monitoring Systems and Effluent ^3-50
.3.7.1 Monitoring Systems- 3-50 3.7.2 ' Effluents 3-56 3.7.3 Reactor Floor Hot Cells L3-59 , -
- 4. . Surveillance Requirements 4-1 4.1 Reactor Core Parameters .l 4-1.
4.2 Reactor Control and Safety Systems 4-4:
4.3 Coolant Systems 4-9 4.4 Containment Surveillance- ~4-12 4.5 Ventilation Systems' 4-15 4.6 Emergency Electrical Power Systems 4 4.7 Radiation Monitoring Systems and Effluents ~4-19 4.7.1 Radiation Monitoring Systems 4-19'
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4.7.2 Effluents . 4-21
- 5. Design Features .5-1 5.1 Site and Facility Description 5 5.2 Primary Coolant System 5-3 ,
i 5.3 Reactor Core and Fuel 5-5
- 6. Experiments 6-1 6.1 GeneralExperiment Criteria- 6-1 1 h !
6.2 - Vacant -
i 6.3 Vacant 1
6.4 ._ Closed-Loop Control Systems - 6 !
6.5 ; Generation ~of Medical Therapy Facility Beams for.
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! - 6.5A L Quality Management Program - 6-26 4 N._
- 7. - ..' Administrative Controls; .7-1
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7-1
, / 7.1 Organization Structure 7-2 7.1.1 7.1.2 Responsibility 7-2 7.1.3 Staffing 7-4 )
7.1.4 Selection of Personnel 7-5 7.1.5 Training of Personnel 7 17-8 7.2 Review and Audit 7.2.1 MIT Committee on Reactor Safeguards 7-8 7.2.2 Composition and Qualifications 7-8 7.2.3 Charter and Rules 7-8 7.2.4 Review Function 7-9 7.2.5 Audit Function 7-10 7.3 Radiation Safety 7-12 Procedures 7-13 7.4 7.4.1 Review Process 7-13 7.4.2 Approval Process 7-13 7.4.3 Scope of Procedures 7-14 f
7.5 Experiment Review And Approval 7-15 ;
l 7.5.1 Review Process 7-15 l ;
l L 7.5.2 Approval Process 7-15 l 7.6 Required Action 7-17 l 7.6.1 Action to be Taken in Case of Safety Limit Violation 7-17 7.6.2 Action to be Taken in the Event of a Reportable Occurrence 7-17 7.7 . . Reports 7-19 ~~
-7.7.1 Annual Report ' .7-19 7 7.2 Reportable Occurrence Reports- 7-21
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7.7.3 i Special Reports'. 7-'22 i
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7-23 7.8 Records Retention Five-Year Record Retention 7-23 7.8.1 7-23 7.8.2 Six-Year Record Retention Life of Facility 7-24 7.8.3 c
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1.- INTRODUCTION These technical specifications apply to the MIT Research Reactor, which is designated as the MITR-III, and to its associated experimental facilit es.
1.1 Scope The following areas are addressed: Definitions, Safety Limits (SL), Limiting Safety System Settings (LSSS), Limiting Conditions for Operation (LCO), Surveillance Requirements, Design Features, and Administrative Controls.
1.2 Application i
1.2.1 Purpose These specifications are derived from the MITR's Safety Analysis Report (SAR). They -
consist of specific limitations and equipment requirements for the safe operation of the reactor and for dealing with abnormal situations. These specifications represent a comprehensive envelope d safe operation. Only those operational parameters and equipment requirements directly related to preserving this safety envelope are listed.
1.2.2 Format The format of these specifications is as indicated in Section -1.2.2 of ANSI /ANS-15.1-1990.
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l' l 1.3' Definitions
.vf 1.3.1. Channn] -
A channel is the combination of sensor, line, amplifier, and output devices which arel connected for the purpose'of measuring the value of a parameter.'
l 3.2 ' Channel Calibration A channel calibration is an adjustment of the channel such that its output cornspondsI with acceptable accuracy to known values of the parameter which the channel measures.
Calibration shall encompass the entim channel, including equipment actuation, alarm, er trip, and shall be deemed to include a channel test.
1.3.3 Channel Check
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A channel check is a qualitative verification of acceptable performance by observation -
of channel behavior, or by comparison of the channel with other independent chanc.els or systems measuring the same variable.
.1.3.4 Channel Test
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A channel test is the introduction of a signal into the channel for verificatica that it is operable.
1.3.5 Containnent Containment means a testable enclosure on the overall facility that can support a defined
. pressure differential for functional purposes and which is equipped with isolation system equipment, i
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, r~w 1.3.6 Containment Integrity
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Integrity of the containment enclosure (building) is said to be maintained when all isolation system equipment is either operable or secuud in an isolating position.
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l 1.3.7 Damaged Fuel i
The term " damaged fuel" means: 1) a failure to meet fuel fabrication specifications l
unless waived by MIT after evaluation, or 2) a deterioration of the clad is present that results in fission product levels associated with an element that are elevated by a factor of five or more above the average background level for the core as a whole.
1.3.8 Excess Reactivity Excess reactivity is the 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 O
/ critical (k-effective = 1.0).
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1.3.9 Experiment An experiment is 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 core tank, or in a beampon or irradiation facility, and that is not rigidly secured to a core or shield structure so as to be e part of their design.
1.3.10 Experimental Facility An experimental facility is an appurtenance to the reactor that is generally used to contain and orient an experiment, as in the case of an irradiation thimble, or to provide a desired flux distribution, as in the case of a fihered beam.
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Each required surveillance test or other function shall be performed within the specified l time interval with:
- 1. A maximum allowable extension not to exceed 25% of the specified surveillance interval, unless otherwise stated in these Technical Specifications. !
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- 2. A total maximum combined interval time for any three consecutive surveillance intervals not to exceed 3.25 times the specified surveillance interval.
Surveillance tests may be waived when an instrument, component, or system is not required to be operable, but any such instrument, component, or system shall be tested prior to being used as a required operable instrument, component, or system.
1.3.12 Immediate I Immediate means that the required action will be initiated without delay in an orderly V
manner by using written procedures when applicable.
l 1.3.13 Inadmissible Sample Materials Those materials defined by the MIT Committee on Reactor Safeguards (MITRSC) as either not allowable within the MITR or restricted from the reactor containment building.
Examples include unapproved amounts of combustible, conosive, or explosive materials.
1.3.14 Indenendent Experiments Experiments that are not connected by a mechanical, chemical, or electrical link, i
i 1.3.15 Irradiation Use of reactor experimental facilities where the primary purpose is the production of i
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activated material such as samples for neutron activation analysis, or materials that exhibit radiation damage effects, or radioactive isotopes, or other similar activities. An irradiation may also refer to 1-4
I use of a medical therapy irradiation room for human therapy or other activities such as
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k/ rad.iography.
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1.3.16 Irradiation Series A series ofirradiations reviewed simultaneously on the basis of the maximum material, l
l flux, and irradiation time of any sample of the series.
l l 1.3.'17 Measured Value 1
The measured value is the value of a parameter as it appears on the output of a channel.
1.3.18 Movable Exneriment A movable experiment is one where it is intended that all or part of the experiment may be moved in or near the core or into and out of the reactor while the reactor is operating.
.n f i 1.3.19 Non-Secured Exneriment G
A non-secured experiment is one where it is intended that the experiment should not move while the reactor is operating, but the experiment is held in place with less restraint than a secured experiment.
1.3.20 Onerable Operable means a component or system is capable of performing its in*-Wi ftmetion.
1.3.21 Onerathig Operating means a component or system is performing its intended function.
1.3.22 Potential Reactivity Worth of an Experiment The potential reactivity worth of an experiment is the maximum absolute value of the
,7~s reactivity change that would occur as a result of intended or anticipated changes or credible
- i b malfunctions that alter experiment position or configuration.
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. The evaluation must consider possible trajectories of the experiment in motion relative to the reactor, its orientation along each trajectory, and circumstances which can cause internal changes such as creating or filling of void spaces or motion of mechanical components. For-removable experiments, the potential reactivity worth is equal to or greater than the static reactivity worth.
1.3.23 Protective Action A protective action is the initiation of a signal or the operation of equipment within the
! reactor safety system in response to a variable or condition of the reactor facility having reached a specified limit and the response of system equipment to control the variable and ameliorate the condition:
- 1. Channel Level: At the protective instrument channel level, protective action is the generation and transmission of a trip signal indicating that a reactor variable has reached the specified limit.
- 2. Subsystem Level: At the protective instrument subsystem level, protective action is the generation and transmission of a trip signal indicating that a specified limit has been reached. (Nolc: Protective action at this level would lead to the operation of the safety shutdown equipment.)
- 3. Instrument System Level: At the protective instmment system level, protective action is the generation and transmission of the command signal for the safety shutdown equipment to operate.
- 4. Safety System Level: At the reactor safety system level, protective action is the c,xration of sufficient equipment to shut down the reactor immediately.
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l 1.3.24 Reactivity Wonh of an Exneriment l ((~)
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The reactivity worth of an experiment is the value of the reactivity change that results from the experiment being inserted into or removed from its intended position.
1.3.25 Reactor Operating The MITR is operating whenever it is not in either a secured or a shutdown condition.
1.3.26 Reactor Operator A reactor operator is an individual who is licensed by the U. S. Nuclear Regulatory Commission to manipulate the controls of the MIT Research Reactor.
1.3.27 Reactor Safety System The MITR's safety system consists of those systems, including their associated input
/3 4 channels, which are designed to initiate automatic reactor protection or to provide information for l
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initiation of manual protective action. The MITR reactor safety system is also referred to as the reactor protection system.
1.3.28 Reactor Secured The MITR is secured when:
- 1. Either there is 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) The minimum number of neutron absorbing control devices are fully inserted or other safety devices are in a shutdown position, as required by technical specifications, b) The console key switch is in the off position and the key is removed from the lock, i
s'1 c) No work is in progress involving core fuel, core structure, installed
, control devices, or control device drives unless they are physically decoupled from the control devices, i
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.d)L No in-core experiments are being moved or serviced, and c) No work is in progress involving fuel in the fission convener tank. -
1.3.29 Reactor Shutdown The MITR is shut down when all control devices (shim blades and regulating rod) are fully insened or a reactivity condition exists that is equivalent to one where all control devices are fully insened.
1.3.30 Reference Core Condition The reference core condition is the reactivity condition of the core when the primary..
coolant and D 2O reflector coolant temperatures are at 10* C and the reactivity worth of xenon is zero (i.e., cold, clean, and critical).
O 1.3.31 Regn1ating Rod The MITR's regulating rod is a low worth control device that is used primarily to maintain an intended power level. It does not have a scram capability. Its position may be varied manually or by an automatic controller.
1.3.32 Reponable Occurrence A reponable occurrence i:: any of the following:
- 1. Any actual safety system setting less conservative than specified in the .MITR Technical Specifications except during periods of instrument maintenance with the reactor shut down,
- 2. Operation in violation of a limiting condition for operation, O u
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- (3/ 3. Safety system component malfunction or other component or system i \'# malfunction which renders, or which threatens to render, the safety system incapable of performing its intended function,
- 4. Release of flaion products from a fuel element in a quantity that would indicare a fuel element cladding failure,
- 5. An uncontrolled or unanticipated change in reactivity greater than 1.0% AK/K,
- 6. An observed inadequacy in the implementation of either administrative or procedural controls, such that the inadequacy could have caused the existence or development of an unsafe condition in connection with the operation of the reactor,
- 7. Conditions arising from natural or offsite manmade events that affect or threaten to affect the safe operation of the facility, 1.3.33 Research Regim The term research reactor as used in these Technical Specifications refers to the Massachusetts Institute of Technology's Research Reactor which is licensed by the U.S. Nuclear Regulatory Commission under license #R-37. It supports a self-sustaining neutron chain reaction for research, developmental, educational, training, medical, and experimental purposes. It is also used for the production of non-fissile radionuclides for use in medical treatments and other purposes and for the medical treatment of humans using neutron beams. ]
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1.3.34 Review and Anorove~ i The terminology "shall review and a: Prove" is t o be interpreted as requiring that the reviewing group or person shall carry out a revie e of the ns. 4ter in question and may then either
,, approve or disapprove it. Before it can be implemented, the matter in question must receive an )
approval from the reviewing group or person. I l-9 I
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Safety Annivsis Reoort p 1.3.35.
d The Safety ' Analysis Repon (SAR) is the document submitted to the U.S. Nuclear '
' Regulatory Commission on July 8,1999, entitled, " Safety Analysis Report fo'r the MIT Research Reactor (MITR-III)," and subsequent revisions thereof.
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l 1.3.36 Safety Channel , .,,
A safety channel is a channel in the reactor safety system.
L 1.3.37 Scram Time The scram time, or shim blade insertion time, is the time elapsed between the initiation of a scram signal and movement of the shim blade from its current position to its 80% inserted-position.
1.3.38 Secured Exoeriment A secured experiment is any experiment, experimental facility, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means or by .j gravity. The restraining forces shall 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 credibic malfunctions.
1.3.39 Secured Shutdown Secured shutdown is achieved when the reactor meets the requirements of the definition :
of." reactor secured" and the facility administrative requirements for leaving the facility with no -
licensed reactor operators pmsent.
1.3.40 Senior Reactor Operator O
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A senior reactor operator is an individual who is licensed by the U. S. Nuclcar
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Regulatory Comrr.ission to direct the activities of reactor operators at the MIT Research Reactor.
Such an individual is also a reactor operator.
1.3.41 Shall. Should. and May The word "shall" is used to denote a requirement; the word "should" to denote a recommendation; and the work "may" to denote permission, neither a requirement nor a recommendation.
1.3.42 Shim Blade MITR shim blades (also called control rods) are devices fabricated from neutron-absorbing materials that are used to establish neutron flux changes and to compensate for routine reactivity losses. The MITR shim blades are coupled to their drives by electromagnets and they perform a safety function when the electromagnet is de-energized. Shim blade position may be (O
LJ varied manually or by an automatic controller.
1.3.43 Shutdown Margin Shutdown margin is the minimum shutdown reactivity necessary to provide confidence i
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 tha' the most reactive shim blade and the regulating rod are in their most reactive psitions and tha*. the reactor will remain suberitical l
without further operator action. For the MITR, the minimum shutdown reactivity is 1% AK/K and the most restrictive operating condition is cold (10' C), xenon-free, with all movable samples I
in their most reactive state.
I 1.3.44 ShutdownReactivity Shutdown reactivity is the value of the reactivity of the reactor with all control devices n
Q in their least reactive positions (e.g., inserted). The value of shutdown reactivity includes the 1-11
a reactivity value of #1 i'istalled experiments and is determined with the reactor at ambient conditions.
1.3.45- True Value The true value is the actual value of a parameter. <
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l.3.46 Unscheduled Shutdown An unscheduled shutdown is defined as any unplanned shutdown of the reactor caused by actuation of the reactor safety system, operator error, equipment malfunction, or a manual-shutdown in response to conditions that could adversely affect safe operation, not including shutdowns that occur during testing or check-out operations.
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( 2. SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Lirnits Apolicability This specification applies to the interrelafed variables associated with core thermal and hydraulic performance. These variables are:
P = total reactor power, Wp = reactor primary coolant total flow rate, Tout = reactor primary coolant outlet temperature, and H = height of coolant above top of fuel plates.
Obiective O
V To establish limits within which the integrity of the fuel clad is maintained.
Specification
- 1. For forced convection, the point determined by the true values of P, Wp, and T out shall not be above the line given in Figure 2.1-1 corresponding to the coolant height, H.
- 2. For natural convection, the true values for P and H shall be as follows:
Variable Safety Limit Power, P 350 kW (maximum)
Coolant Height, H 6 feet above top of fuel plates (minimum) m 2-1
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Coolant Height, H ,
50
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710 fee (3 meters)-
6 feet (1.8 meters)
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Fcore (RFrdr j \^ .
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i Reactor Outlet Temperature, Tout ( C) q
, Figure 2.1-1 MITR-III Safety Limits for Forced Convection Operation 1
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'I The basis of this specification is given in Sectionl4.6.5 of the SAR'where it is noted that critical heat flux'(CHF) is normally used as the criterion 'of fuel overheating. . However,;
because the coolant flow path in the MITR core is a multichannel design, there exists the possibihty -
that flow instabilities could occur before reaching CHF limitations. . If onset of flow instability:
(OFI) did occur first, it would have the effect of lowering the flo'w rate _to the hot channel _
l significantly and thus lowering the critical heat flux. . In the safety limit calculations, both CHF and OFI are calculated and the one that would occur first is used to determine the safety limits.
Also, in the SAR, a relationship between the reactor operating parameters and OFI is derived with the assumption that the hot channel receives the minimum flow among all the coolant channels.
The derivation, which uses the energy conservation equation for the hot channel and the channel subcooling ratio for onset of flow instability, yielded the following nlation:
P _r(T,-T cp out g Wp p"* Fr % _3 g RFrd t ,
.J where l Fcore is the fraction of the total power deposited in the core region, Fr is the nuclear hot channel factor,
' Fu is the engineering hot channel factor for enthalpy rise, R is the channel outlet subcooling ratio, Fr is the fraction of primary flow cooling the fuel,
'd{ is the flow' disparity, which is the ratio of the minimum expected flow in the hot channel to the average chennel flow, cr p is the specific heat of the fluid, T,.t is the water saturation temperature at the outlet end of the core ( C), and q >
T out , is the average core outlet temperature (*C),
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jq The safety limits are calculated assuming R=0.86 (as derived in Section 4.6.2.2 of the SAR),
'- Fr =2.0, Frd =f 0.8 , Fu=1.173, and Fcore=1.0. Hence, p** FrFit _ [ , g
( RFrd t j Figure 2.1-1 shows the safety limits for coolant heights of 10 and 6 feet. The coolant height is the t elevation from the top of the fuel plates to the air / water interface at the top of the core tank. A coolant height of 10 feet corresponds to 4 inches below the overflow level. A coolant height of 6 feet corresponds to a point several inches below the anti-siphon valves.
The safety limit factor, which is defined as h
f ,,r F g p"' r i _3 RFrd t ,
will be calculated before reactor operation above 1 kW as required by Specification 3.1.4.4, to O
CI ensure the validity of the calculated safety limits.
The safety limits for natural-convection operation are calculated using a zero flow critical heat flux correlation. Coolant channels can be cooled by countercurrent flow with a downward d
movement of water and an upward flow of bubbles or steam generated in the channel. This is referred to as a flooding condition because the flow channels are submerged in a pool of coolant.
A detailed description of the zero flow critical heat flux correlation is given in Section 4.6.6.3 of 1
l the SAR. For the geometry of the MITR fuel elements, the calculated zero flow critical heat flux is 2.353 x104W/m s, 2 which corresponds to a reacto: power of 468 kW with a radial peaking factor of 2.0. Upon taking into account the engineering hot cham.el factor for enthalpy rise (Fn), the reactor power corresponding to a dryout condition becomes 399 kW. A reactor power of 350 kW is conservatively adopted as the safety limit. The core outlet temperature and the coolant height do not affect the dryout limit, as long as the core is covered with coolant. The coolant height is conservatively set at 6 feet above top of the fuel plates to ensure an adequate coolant inventory.
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_y 2.2 I imitine Safety Svstem Settings (LSSSF Annrnhility'
- This specification applies to the setpoints for the safety channels that monitor reactor <
power, primary coolant flow, reactor outlet temperature, and coolant height above the top of the fuel plates.
Obiective
.To assure that automatic protective actions will prevent incipient boiling in the reactor core and will prevent operating conditions from exceeding a safety limit.
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Specification ,
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- 1. The measured values of the limiting safety system settings on reactor thermal .
power, P, reactor primary coolant flow rate, Wp, the height of water above the top of the fuel plates, H, and the reactor outlet temperature, Tout, shall be as follows:
Table 2.2-1 Limiting Safety System Settings Parameter LSSS (2 pumps) LSSS (1 pump) LSSS (0 pump)
Power 7.4 MW (max) 3.2 MW (max)- 100 kW (max) <
Primary Coolant Flow I800 gpm(min) 900 gpm (min) N/A' Steady-State Core Outlet 60' C (max) 60* C (max) 60' C (max)
Temperature -
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n 4" below overflow - 4" below overflow 4". below overflow -
or 10 feet above top or 10 feet above top or 10 feet above top Coolant Height of fuel plates (min)' of fuel plates (min)! of fuelplates (min)
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Basis N[.
Q The basis of this specification is given in' Section 4.6.7 of the SAR. Onset of nucleate
- boiling (ONB), which is also called incipient boiling, defines the condition where bubbles first start to form on a heated surface. Because most of the liquid it still subcooled, the bubbles do not -
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establish reactor
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detach but grow and collapse while attached to the wall. :It is desirabic
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operating' conditions that will prevent onset of nucleate boiling because doing so will assure that a safety limit is not exceeded. In Section 4.6.7 of the SAR, an expression for the ONB limits was derived based on a coolant height corresponding to a pool level 4 inches below the overflow or 10
' feet from the top of the fuel plates. This coolant height contsponds to a saturation temperature of .
- 107' C. The engineering hot channel factors for enthalpy rise (Fg) and film temperature rise (FAT) are included to account for uncertainties because of measurement, calculation, and possibic
deviations from nominal design specifications that may affect the thermal hydraulic calculation results.
Forced Convection The following equation is derived for a coolant channel with an arbitrary axial power -
deposition distribution q"(z) where z is the axial distance along a fuel plate:
. F - q"(z)
Tout < 107 + CP F "' +0.0177(q (z))0.466 - - . ..H gpHg (z)dz-F AT (2.2-1)
Wetpp mcpr h A comparison of three power distribution profiles (uniform, sine / cosine, and bottom peak) was reported in Section 4.6.6.1 of the SAR. This comparison indicated that the uniform ,
power distribution would lead to a maximum clad temperature at the channel outlet. Therefore, .
Equation (2.2-1) can be simplified to:
H@ - FA avs -
. Tout < 107pp + W c r ,C"' + 0.0177q',yg ^(2.2-2).
theprNe h k/
'2 ,
A' ( f -
,I'
/ where
)
. p FfuejFcoreFr gavs = NcAi j and WP dn= Fgdr Ne where Tout is the bulk outlet temperature in 'C, P is the reactor power, Pii is the heated perimeter of a coolant channel, Fcore is the fraction of the total power deposited in the reactor core, Wp is the primary flow rate, cr p is the heat capacity of the primary coolant, Fii is the engineering hot channel factor, Psr is the engineering hot channel factor for film temperature rise,
)
h is the heat transfer coefficient, Fr is the radial power peaking factor, which is the ratio of the power produced in the fuel plate to the power produced in the average fuel plate, Ne is the number of coolant channels, Aii is the effective heat transfer area of a fuel plate, Fruel is the fraction of the core power deposited in the fuel plates, dr is the flow disparity in the hot channel, and i
Fr is the fraction of primary flow cooling the fuel elements. !
The LSSS core outlet temperature as a function of reactor power is calculated using f Equation (2.2-2) subject to the assumption of a primary flow rate (Wp ) of 1800 gpm (two-pump) and 900 gpm (one-pump) and a coolant height at 4 inches below overflow. Other parameters, such I
as Fr, Fr, and dr are the same as specified in the basis of Specification 2.1. Figures 2.2-1 and
/ 2.2-2 show the LSSS reactor power as a function of the core outict coolant temperature for two- l i
2-7
pump and one-pump operation, respectively. For a core outlet coolant temperature of 60' C, the LSSS power is 7.4 MW for two-pump operation and 3.2 MW for one-pump operation.
.]
Natural Convection :
Natural convection calculations were performed, as described in Section 4.6.7.2 of the SAR, on the assumption that the natural convection valves were open.- A power level of 100 kW and a coolant height of 10 feet above the top of the fuel plate (4 inches below overficw) were .
assumed. The maximum fuel clad temperature was estimated to be below incipient boiling if the pool temperature is maintained below the normal outlet temperature scram point of 60* C.
O n
a 2-8 L:
i.-
1 O ,
9 Coolant Height, H =10 feet ?
8 -ik -- Primary Flow Rate, Wp=1800 gpm .' ,.
7-N
-t 1
o.
--- j 6 - ~ ,
F,Fp = 2.0 Frdr = 0.8 .
5 .-----
pg 3,373 l FSr = 1.275 4 1 50 55 60 65 70 75 80
]
'l Reactor Outlet Temperature, Tout (*C) !
~
Figure 2.2-1 . MITR-III Limiting Safety System Settings for Forced Convection Operation -
(Two Pumps).
l
-1 i
O q 2 ,
_l '< '
.:r.
t '
y >
w v.
' ..) :. -
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.s r
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3.5 -
A' --- Coolant Height H =10 feet -._ -- ,
Primary Flow Rate, Wp=900 gpm
~
~
3 -
k \
t$ 2.5 -
2 .
F,Fp = 2.0 Ff dr = 0.8 1.5 FH " l173 . _ _ _
FEr = 1.275 1
a.
50 55 60 65 70 75 80 Reactor Outlet Temperature, Tout ('C)
Figure 2.2-2 MITR-III Limiting Safety System Settings for Forced Flow Operation .
(One Pump).
u
, .. c.
', <2-10 I
r a 4 k
-3. LIMITING CONDITIONS FOR OPERATION This sectica of the MITR Technical Specifications contains limiting conditions for -
operation (LCOs),: These LCOs are derived from the safety analyses in the SAR, which provide the bases for the LCOs. LCOs are implemented administratively or by control and monitoring .
circuitry to ensure that the reactor is not damaged, that the reactor is capable of performing its -
intended function, and that no one suffers undue radiological exposures because of reactor -
operation.
The LCOs conform to the intent of ANSI /ANS-15.1-1990 as amplified by NUREG-1537, Part I, Rev. O,2/96.
- i O
l 0
3-1 1
{ - 4 y)? f 3.1 - ~ Reactor Core Parameters
' 3.L1 Excess Reactivity -
Applicability This specification applies to the subcritical interlock and thereby to the allowed excess--
reactivity.
Obiective _
To limit the excess reactivity, i
Sperfication
- 1. The reactor shall not be made critical unless the infinite-period, critical shim
't bank height under cold (10' C), xenor.-free conditions shall be greater than 5.0 inches with the regulating rod also at or above 5.0 inches and all movable samples in their mo:;t reactive state.
Basis The basis for the specification is contained in Sections 4.5.3.2 and 4.5.3.3 of the SAR.
Observance of the subcritical limit interlock restricts the allowed critical shim bank height to 5.0 inches or greater. This places an upper limit on the excess reactivity that could be inserted-upon withdrawal of the blades past the critical position. This use of a physical interlock is preferable to reliance'on an administrative limit. - As discussed in Section 4.5.3.2 of the SAR,-
i' observance of this requirement limits the total excess reactivity in the core to about 9.4 beta. This
. equates to about 1.6 beta per blade which is the maximum that could realistically be inserted because only one blade can be physically withdrawn at a time.
N,] a 3 1 V
y
73 The excess reactivity allows for the effects of coolant and D2O reflector temperature N~) (10' C - 55' C), equilibrium xenon, peak xenon, in-core samples, and fuel burnup. The temperature effect is worth about 0.26 beta. Equilibrium and peak xenon are wonh about 4.2 and 6.4 beta, respectively. The number ofin-core samples varies. Two ICSAs totaling about 0.8 beta is a reasonable figure. This leaves about 1.9 beta to allow for fuel burnup. Given a reactivity loss j i
of 0.25 mbeta/MWH, this means that a refueling is required about every 7600 MWH or about 50 l operating days at full power. A shoner refueling interval would increase radiation exposures and )
would not be in keeping with ALARA considerations. f n i
,.m 3-3
( f
-f. ,
- 3.1.2 LShutdown Margin a
' ~Annlienhility
'li ':.
L This specification applies to the shutdown margin requirement. .
- Obiective To assure that the reactor can be safely shut down at any time.
Specification
- 1. The reactor shall not be made critical unless the reactor can be made 'subcritical using shim blades by at least 1% AK/K from the cold (10' C), xenon-frec.-
critical' condition with the most reactive operable blade and the regulating rod fully withdrawn and with all movable experiments in their most reactive state.
O Basis The shutdown margin requirement incorporates the following general philosophy concerning reactor safety:
- It should be impossible for a reactor to be made critical in its most reactive -
situation on the withdrawal of a single rod. Conversely,it should always be
- possible to shut 'down the reactor with one rod stuck in its outermost position. .
Ifit is possible that rods'or mechanisms might interact so that several could be stuck in the out position, then the number of rods inchtded in the stuck rod criterion should be increased accordingly [3.1.2-1].
The value selected for the shutdown margin (1% AK/K) is substantial, and it can be readily determined. l
, I References j
~
3.1.2.-l Thompson,- T.J. and J. G. Beckerley (Eds.), The Technology of Nuclear Reactor l Pt , Safety,1Vol. I, the MIT Press, Cambridge, MA (1964), p. 677. j (f .
'3-4 y * <
-c-9 -
, j 4 , ,
n .]
v
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- -(
3.1.3 ' Maximum Safe Step F, tivity Addition:
F Applicabihty This specification limits the' amount of reactivity.which may be added to the reactor in a :
single amount.
+
Obiective To assure that the integrity of the fuel is maintained during any credible' step reactivity -
excursion.
I Specification
~
- 1. The maximum amount of reactivity that may be added to the critical reactor by the credible failuir or malfunction of any experiment or component or any set of; circumstances which 'could credibly couple.two or more components or
(
experiments shall be less than 1.8% AK/K if forced convection cooling is established prior to taking the reactor critical. This figure shall be less than
.;. 1.2% AK/K if natural convection cooling is used prior to taking the reactor.-
critical.
(.'
- 2. If a.. in-core sample is loaded using the 1.8% AK/K criteria as its reactivity-limit, then the reactor shall not be made critical under conditions of natural:
convection cooling unless the sample has first been removed.
Basis The basis for this specification is contained in Section 13.2.2.1 of the SAR. Analyses ^
~
l performed .with the PARET code shond that the power transient.that would resuit from a step
~
()i ! insertion of the amount of reactivity would be terminated by negative reactivity feedback ' effects ;
,3-5 . .
4 q and that the fuel and clad temperatures would not exceed 531' C, which'is below the melting point.-
These limits are conservative relative to ones obtained from correlations with SPERT data. The step reactivity insertion limit is governed by both feedback effects and heat removal considerations.
The former are not significantly affected by the flow rate. The latter are less efficient at lower flow rates. Hence, higher fuel clad temperatures result from natural convection cooling and a lower -
limit is necessary.
O n}
\
3-6--
, , +
i 3.1.4 - ' core Confieurations
~
~ Apolicability
.i This specification applies to conditions pertinent to reactor cora design and operation.
Obiective
- To assure that reactor core cchngurations are maint'ained within the envelope of - H conditions that were used to establish the themial hydraulic limits.
Specification a
- 1. The reactor shall not be made critical unless all fuel elements and other core .
I components are secured in position and the hold-down plate.is latched in I
-I position. 1 O
V
- 2. The reactor shall not be made critical unless at least five shim blades are-operable and any inoperable blade is at the operating position or higher.
- 3. Pu-Be r.eutron sources shall not be used in the reactor above a reactor power of 500 watts.
.\
I
- 4. The reactor shall not be operated above a power level of 1.0 kW unlessi a) The safety limit factor, FF rH_j p** '
.i RFrdr ,
is predicted to be less than 2.4. Definitions of the' parameters that .
- r. 1 comprise the : safety . limit . factor are given in the _ basis 1 of -
b.
~
Specification 2.1.
v .
'3-7"
, y y [
a:
J
( ,) b) The core is predicted to operate below incipient boiling at every point in the core by the use of Equation (2.2-1).
c) Before each refueling or change in core loading which might increase the parameters listed in 4(a) and 4(b) above over a previous operating condition, an evaluation s 5.. it be made to ensure that items 4(a) and 4(b) above are satisfied. A secord of these evaluations shall be completed and approved by two Senior Reactor Operators.
d) All positions in the core tank are filled with either a fuel element or another approved unit. The value of Fr used in the calculation required by 4(a) and 4(b) above shall correspond to the number of non-fueled positions in the core.
- 5. The reactor shall not be operated at power levels of greater than 100 kW unless:
a) Primary coolant flow is established.
< b) At least five operable shim blades are within 2.0 inches of the average shirn blade height and any inoperable blade is at the average height or above, except that greater imbalance may exist when one or more shim j,m blades is being inserted to make the reactor suberitical.
c) The reactor top shield lid is in position.
- 6. Fuel elements shall neither be inserted nor removed from the core unless the reactor is in a shutdown condition.
- 7. A shim blade shall not be removed from the core unless the heavy-water reflector is dumped or the shutdown margin criteria will be met with the most reactive blade and the regulating rod fully withdrawn after the control blade has been removed.
Basis
- 1. Hold-Down Grid Plate All fuel elements and core components must be secured in position to prevent
\
_j mechanical damage of the components, to preclude reactivity changes that might 3-8
result from inadvertent movement, and to assure proper flow distribution and '
' cooling.
- 2. . Ooerable Blades Section 13.2.9.1 of the SAR discusses operation with a blade below the average bank position. Such operation will not cause a significant change in the core power distribution. Nevertheless,it should be avoided.
- 3. Pu-Be Neutron Sources The use of the two one-curie Pu-Be neutron sources has been analyzed and determined to be safe up to a reactor power of 500 watts. This analysis was submitted to the NRC by letter dated March 28,1975.
- 4. Safety Calenlations a) Safety Urnit Factor The safety limits, as shown in Figure 2.1-1, are calculated using a value of 2.4 for the safety limit factor. The safety limit factor contains measured thermal hydraulic parameters that are not amenable to-continuous monitoring. The safety limits are calculated assuming R=0.86, Fr=2.0, Ffdr=0.8, F H=1.173, and Fcore=1.0. Hence, r H -1 =2.4 F core (RFrdt j It is shown in Section 4.6.2.2 of the SAR that a value of 0.86 for R, which is the' channel outlet subcooling ratio, is a conserva;ive-assumption. The value of 2.4 is selected because it is considerably higher than the actual expected value for the operating core and therefore the resulting safety limits are sufficiently distant from the operating range. Specifically, for a safety limit factor of 2.4, the calculated safety
-limit on the power level is 9.0 MW .with tie .
l flow and outlet temperatures at their limiting safety system settings (LSSS). An adequate margin is thereby provided between the safety limits and the limiting safety system settings.
b)' Incinient Boiline I imit 3-9
g C The overall operating limits to prevent incipient boihng are interrelated by Equation 2.2-1. Evaluation of core operating conditions to prevent incipient boiling should be performed and approved before each refueling operation.
c) Evaluation of the Core Factors and Axial Power Distribution (i) The values of the flow disparity factor, dr, were determined for ,
all fuel element positions by a combinaticn of fuel channel tolerance and relative flow measurements during the preoperational testing of the MITR-II. These values are subsequently used for evaluation of the safety limit factor and the incipient boiling limit.
(ii) The nuclear hot channel factor, Fr , which is used to establish the safety limits and the limiting safety system settings, is determined either by experiment or by calculation. In all cases, the nuclear hot channel factor shall be used with a conservative estimate of its uncertainty as described in Section 4.6.4 of the SAR. By definition Fr is the ratio of the power deposited in the hottest channel to that in the average channel. The relative plate power can be derived by experiment, such as gamma scanning of n presentative fuel plates.
(iii) When experiments or experimental facilities are placed in the V(3
^
core and use part of the primary coolant, the flow rate through them will be determined either by measurement or by calculation such that a conservative value of Fris obtained.
(iv) The axial power distribution shall be evaluated either by measurement or by calculation. The evaluation of the power distribution involves the determination of the hot spot, which has the highest clad temperature,in the core. The hot spot may or may not be at the point of highest power density, which is usually at the core bottom or coolant inlet end of the channel, because a smaher power density peak that occurs further up the channel where the bulk coolant temperature is higher could lead to the highest clad temperature. In all calculations made for the SAR, it is conservatively assumed that the hot charmel has the highest power (maximum Fr) and the lowest flow rate (minimum d f). Therefore, the limiting safety system settings derived in Section 4.6.7 of the SAR should provide a lower bound of the safe operating conditions. However, the incipient boiling limit will be calculated using Equation (2.2-1) before refueling because the power distributions (axial and radial) are affected by factors such as core configuration, fuel burnup, and shim bank height.
d) The safety limits in Specification 2.1 are derived on the basis of no (1 excessive bypass flow among the fuel elements. All positions in the V core must be filled in order to ensure such a flow. The value of Fr depends on the number of non-fueled positions.
3-10
.j
, l h
O 5. ~ Reactor Ooeration at 100 kW or Higher a) For operation at 100 kW or higher, forced convection is required.
~
b) The established safety limits assume a banked shim blade height. An imbalance ofi 2.0 inches will not appreciably affect the core power :
. distribution.
c) The reactor top shield is part of the biological shield. To facilitate the performance of various experiments placed in the core, the reactor may be operated at power levels below 100 kW with this shield removed.
The total dose rate at 100 kW on' the surface of' the coolant is .
approximately 1.3 rem / hour. This dose rate is not in excess of those 1 occasionally encountered during certain maintenance operations, and it {
. has been demonstrated that administrative controls can provide adequate I control under such conditions. Adequat,e controls will be instituted 'l during such experiments to prevent excessive personnel exposure.
- 6. Insertion / Removal of Fuel Insertion or removal of fuel is done while shut down to assure safety.
O 7. Shim Blade Removal The shutdown margin is modified so that it is met with the two most reactive blades in their full-out positions. This assures safety. Alternatively, the D2O reflector may be dumped.
I i
l 1
l w.-
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' (l 3-11 q
~-
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- 'l.
3.1.5 R~tivity CoefficientC '
.t -
Applicability This specification applies to core reactivity coefficients.
Obiective ,
To assure that reactivity coefficients are negative.
Specification
- 1. Reactivity coefficients (fuel temperature, primary and D 2O temperature, and void) shall be negative over the normal operating range (10* C - 60*, C). .Any y significant observed change'in the magnitude of the coefficients in excess of i20% from the values measured during the MITR-II startup testing'shall be evaluated.
Basis MITR reactivity coefficients are discussed in Section 4.5.2.2 of the SAR.. All magnitudes are negative for operation above 10' C. (Nok: The coolant temperature coefficient is slightly positive below 10' C with the total effect being a few millibeta of reactivity.)..
Measurements of the coefficients over the life of the MITR-II (1975-1999) showed no significanti change in magnitude from the values measured during the startup testing.
,j k
n -l.
- 1. :-l , 1 4
j 12 '
4 s .. g .
m.
p - 3.1.6 Fuel Parameters d
1 Applicabihty This specification applies to fuel parameters.
\
Obiective.
To assure that the fuel cladding is periodically inspected.
Specification
- 1. The reactor shall not be operated with damaged fuel except as may be necessay .
to identify the location of the damaged fuel.
~
- 2. Visual inspections of the fuel in the core shall be performed to detect possible deterioration of the clad. The following shall satisfy this requirementi a) A visual examination of the core using normal lighting.
b) A visual examination of the core with normal lighting secured.
c) Visual examination of the visible surfaces of each element whenever the -
1 elementis moved during a refueling. .
- 3. Peak fuel burnup shall not exceed 1.3 x 1021 fissions /cm3 if the fuel is uranium-aluminum (UAl) alloy; or 2.3x102t i ssions/cm3 if the fuel is intermetallic UAlx with 4 to 7% voids.
l M .
'l 1.' Damnoed Fuel l
~
The MITR uses the fuel sipping method to identify incipient fuel element clad q
q ; failures [3.1.6-1]. This method requires operation of.the~ core at power for !
V sufficient time to generate detectable activity.-
m-
'3-13 -
3- q r- l I u
~~~
s
~
3
- V ..
~'
i
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P
- L ,
e
~
- 2. FuelInspections The program described in Specification 3.116.2 above has been in place for >
more than a decade and is effective as noted in Section 4.2.1 of the SAR. The' -
visual inspection with normal lighting secured makes use of Cherenkov:
radiation to backlight the fuel. This mak' es visible any defect, such as a small:
blister, that protrudes into a coolant channel.~ All fuel inspections are done in l the core tank under water.
- 3. Fission Density The peak fission density limit is based on information developed and tested as part of the fuel designs for the Engineering Test Reactor (ETR) and the Advanced Test Reactor (ATR) located at the Idaho Testing Station [3.1.6-2, 3.1.6.3). Studies show that the fission density could be at least 2.3 'x1021:
fissions /cm3 based on the irradiation performance iesting of the UAlx dispersion fuel [3.1.6-4, 3.1.6-5]. Fuel clad performance associated with higher burnup -
has also been evaluated. Crud formation on the fuel clad would present a significant heat transfer resistance because of its relatively small thermal conductivity compared to that of aluminum. Oxidation of aluminum has been-identified as the major contributor to crud buildup on fuel clad.' The reaction rate of the oxidation process depends on the thermal (temperature and/or heat flux), hydraulic (flow velocity), and chemical (pH) conditions' [3.1.6-6, 3.1.6-7]. A crud thickness of 2 mils or less is recommended which is based on the limit to prevent f.pallation that could lead to the release of radioactive gas
[3.1.6-8]. Calculations performed using-.the MITR operating conditions ,
7 indicate that the crud thick r e on the MITR fuel wiD not exceed 2 mils before
~
3
_ ' the burnup limit of 2'.3 x'1021 fissions /cm is reached [3.1.6-9].
? 3 ~
p , ,
r
- r5 References V
3.1.6-1 Clark, L. Jr., Bernard, J.A., and E. Karaian, " Fuel Cladding Failure at the MIT
- Research Reactor," Transactions of the American Nuclear Society, Vol. 38, Suppl.1, , -1
' Aug.1981,'pp 25-26.
3.1.6-2 J.M. Beeston, R.R. Hobbins, G.W. Gibson, and W.C. Francis, " Development and Irradiation Performance of Uranium Aluminide Fuels in Test Reactors," Nuclear Technology 49, pp.136-149,~ June 1980.
3.1.6-3 G.W. Gibson, The Development of Powdered Uranium-Aluminide Compoundsfor .
Use as Nuclear Reactor Fuels, IN-1133, Idaho Nuclear Corporation, Idaho Falls, Idaho, December 1967.
3.1.6-4 J.L. Snelgrove and G.L. Hofman, Evaluation of Existing Technology Base for Candidate Fuels for the HWR-NPR, Argonne National Laboratory ANIJNPR-93/002, .
Feb.1993.
3.1.6-5 R.W. Cahn, P. Haasen, and E.J. Kramer, Materials Science and Technology - A .
Comprehensive Treatment, Published by VCH, Germany, 3.1.6-6 S.J. Pawel, D.K. Felde, R.E. Pawel, " Influence of Coolant pH on Corrosion of 6061 Aluminum Under Reactor Heat Transfer Conditions," ORNL/TM-13083, Union Carbide Corp., Oak Ridge National Lab. Oct.1995. ,
1 O 3.1.6-7 R.E. Pawel, G.L. Yoder, D.K. Felde, B.H. Montgommery, M.T. McFee, "The V Corrosion of 6061 Aluminum Under Heat Transfer Conditions in the ANS Corrosion .
Test Loop," Oxidation ofMetals, Vol. 36,1991.
3.1.6-8 J.C. Griess et. al., "Effect of Heat Flux on the Corrosion of Aluminum by Water: Part ;
III Final Report on Tests Relative to the High Flux Isotope Reactor", ORNL-3230, Union Carbide Corp., Oak Ri& National Laboratory, Dec.1961.
3.1.6-9 File Memo (MITR Fuel Oxide Layer Buildup). l l
i u -
, .. 1 3-15 t ,
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' t, " '
' 3.2 : P~ tor Control and Safety System O .
~
3.2.1 Ooerable Control Devices -
Applicability t.
This specification applies to the reactor control and protection systems.-
Obiective To specify the number and type of operable control and safety devices as well as the kilowed scram times.
Specification
A minimum of five shim blades shall be operable - as stipulated in Specification 3.1.4.2.
- 2. The time from initiation of a scram signal and movement of each operable blade from its current position to its 80% inserted position is less than one second for each blade.
i Basis
. The MITR is equipped with six shim blades and one regulating rod. The blades are connected to their drives by electromagnets and hence can be dropped into'the core upon initiation L. ,.
L of.a scram signal. The basis of the second part of the specification is industry practice.' A scram
- time of 1.0 second is the standard. Analyses in the SAR show diat a 1.0 secon'd scram time will' .
.4
.I
, not result in any damage to the fuel. (Npd: For some analyses,'a 2.0 second delay ~was assumed -
(n) -
~
as a conservative measure.)- ;
f- j 3 [-
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'3.2.2 Reactivity insertion Rates andAutomatic Control .
_Anolicability This specification applies to the reactivity control system.
Obiective To assure that the integrity of the fuel is maintained during any credible ramp reactivity .
excursion.
- Soecification
~
i
- 1. The maximum controlled reactivity addition rate is no more than 5 x 10 AK/K/s.
(N k 2. Only one shim blade shall be withdrawn at a time.
- 3. Shim blades and/or the regulating rod may be connected to automatic controllers within the limitation of Specification 3.2.2.4 below.
- 4. The total available positive reactivity of any control device connected to an automatic controller, other than those covered by Specification 6.4, shall be less than 0.5% AK/K.
.5. The nuclear safety system shall be separate from any automatic controller. -;
-l 1
d
, l lp ,
V .l
,. : .]. 1 X 10 3-17 .i t a ,
9
- r. ;
O '
a'
)- Definitions
. The following facility-specific definitions are provided:
- 1. Total Available Positive Reactivity The " total available positive reactivity" of any control device connected to an automatic conthol system is the positive reactivity beyond the critical condition; that could be inserted if the control device were fully withdrawn.-
- 2. Separate The word " separate" means that the output of an instrument used in the safety 1 system will not be influenced by interaction with the control system. For-example, a signal derived from aninstrument that forms part of the safety-system would not be transmitted to the control system unless first passed through an isolation device.
. The basis of Specification 3.2.2.1 is discussed in Section 13.2.2.2 of the SAR whereL ramp reactivity insertions are analyzed. The control blade and regulating rod speeds are designed to limit the reactivity addition rate to less than 5 x 10-4 AK/K/s. This value is conservative within =
the range of reactivity insertion rates normally accepted for reactor operation. Control systems in -
this range give ample margin for proper human response during approach to criticality and power operati6n. In the event of an accidental continuous insertion of reactivity at this maximum rate, the response of the reactor safety system period and level trips will adequately protect the reactor.' The MITR's control system is constructed so that only one shim blade can be withdrawn at a time.
The basis of Specification 3.2.2.2 is that the simultaneous. withdrawal of two or more
.t blades is not physically possible.
~ The basis of Specifications 3.2.2.3 and 3.2.2.4 is discussed in Section 10.3.2.8 of tim
~ SAR.L There are two methods for assuring the safety of automatic controllers, either analog and J idigitalZ One is to limit the reactivity worth of the control device. The other is to design the controller to' incorporate the property of feasibility of control." The first'of these options is - ,
t '
3' y.
_l'.
addressed here.' The second is addressed in Specification 6.4. Controllers designed under this Specification 3.2.2 are governed by (1) a limitation of the rate of reactivity insertion, (2) a
- limitation on the reactivity worth of the associated absorber, (3) a requirement that the capability of the safety system to perform its function is not impaired. The value chosen for the reactivity worth limitation is 0.5% AK/K which is the limit for a non-secured experiment. This is appropriate because it is not intended that the controller insert the full amount of the available reactivity..
However,ifit did so, the safety system would protect the fuel.
O .
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1 3-19
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('j 3.2.3 J Reactor Protection System -
b ,
9il Applicabihty This specification applies to the reactor protection system. 1 Obiective To assure that automatic protection action is provided as required by the reactor protection -
- system.
Specifications
- 1. The reactor shall not be made critical unless the reactor protection system is operable in accordance with Table 3.2.3-1.
- 2. Fuel shall not be moved in the core unless the peri 6d and neutron flux level channels are set to alarm'within the zero primary pump limits of Table 3.2.3-1.
7 In addition, the manual major scram is operable for building isolation and the D 2O dump valve selector switch is operable unless the D2O reflector is already -
dumped.
BAEi1 The neutron flux level channels and period channels provide protection on power level q and change in power level. These instmments are therefore required at all power levels including l subcritical operation. At power levels above 100 kW protection is also required on primary, D2 0,-
and shield coolant flows, u 8 /.Q U
L 3-20 .
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i O The parameters listed in Table 3.2.3-1 are monitored by the reactor protection system.
This system automatically initiates action to assure that appropriate limiting safety system settings ;
and limiting conditions of operation are not violated.
In practice, low power physics tests are usually performed at power levels of less than 10 kW and in the absence of forced convection flow. The upper limit of 100 kW for this type of operation was established on the basis of adequate natural convection cooling. The maximum plate s temperature at 100 kW with natural convection cooling is estimated to be below incipient boiling, if the coolant outlet temperature is maintained below the normal scram point of 60* C. Therefore, the reactor outlet temperature channel is specified in Table 3.2.3-1 as 60' C for zero pump operation.
The reflector tank low D 20 level scram must be bypassed during low power operation if calibration of the reactivity effect of the D 20 reflector dump safety system is to be performed.
For refuelings, the reactor is in a shutdown condition, primary flow is secured, and the D 2O reflector is normally dumped. Therefore, the period and level channels are set to alarm within the zero primary pump limits. The capability to isolate the building is required. This is provided by the major scram. Finally,it should be possible to dump the D2 O reflector,ifit is not already dumped.
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3-23
i 3b.4 . Control System interIncke
. konlicability This specification applies to the tractor control system.
.ObNtive To assure that interlocks that prevent control device withdrawal are incorporated in the control system as appropriate.
Specification
- 1. The following interlocks shall be operable, except' as noted in Specification 3.2.4.2 and 3.2.4.3 below, before the reactor is made critical:
a) Withdraw Permit (Startup) Interlock - Shim blade electromagnets car.not be energized unless all required safety channels in the reactor protecti on system and all stanup interlocks are satisfied.
b) Suberitical I imit - Shim Blades Interlock - Shim blade cannot be -
withdrawn above 5.0 0.5 inches unless each blade is first raised to that height.
2 The "Suberitical Limit - Shim Blade Interlock" may be bypassed during critical operation for the purpose of measuring the reactivity worths of the shim blades and the regulating rod.
- 3. The "No Overflow Reflector Etartup" interlock and the " Low Level D2O Reflector" scram may be bypassed at power levels of less than 100 kW for the' purpose of reflector reactivity measurement.
- 4. An interlock shall be operable that allows reactor startup only when the containment pressure is below atmospheric pressure by at least 0.1' inch of
[v '
water.1 ,
t 3-24 oa
,rq Basis
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The basis for the specification is given in Section 7.3.1 of the SAR. Closure of the withdraw permit circuit means that both the reactor protection system scrams and the startup i
interlocks (all absorbers full in, no overflow core tank and no overflow reflector tank clear, and ;
building AP) are satisfied.
Satisfaction of the suberitical interlock assists the operator in establishing a uniform bank height prior to the final approach to criticality. Reactivity worth measurements requiw that I this interlock be bypassed because the blades cannot be kept at an even bank height. Also the "No .
Overflow Reflector Startup" interlock and " Low Level D O 2 Reflector" scram must be bypassed 1
during reflector reactivity measurements because it is necessary to vary the level of the reflector.
It is intended to operate the reactor with a negative containment AP. An interlock must be satisfied requiring a minimum 0.1 inch water pressure differential (negative) in the building j before a reactor stanup can be conducted. This condition ensures that the building containment AP is not grossly violated at the time of startup. This interlock forms part of the withdraw permit
(
circuit.
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3-25' 1
- m. . . . .
- 3.2.5 - Backun Shuidawn Mechanisma -
. Annlicahility This specification applies to the heavy . water reflector system. '
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Obiective To assure a backup means of shutdown for the reacto' r.
Specification
- 1. The reactivity worth of the D2O reflector dump is greater than the reactivity .
worth of both the most reactive blade and the regulating rod.
- 2. The levelin the D2O reflector dump tank shall not exceed 20 inches whenever A
g the reactoris critical.
B.iulis The basis is given in Section 4.5.3.1 of the SAR. The capability to dump the D2O reflector serves as a diverse means of reactor shutdown and hence the reactivity worth of that action should exceed the worth ~of both the most reactive shim blade and the regulating rod because
- the definition of shutdown margin assumes that these two devices are stuck in their full-out position on shutdown. Nowhere in the SAR is it assumed that dumping of the reflector provides
-I
~
protection against a transient condition. Hence, there is no limit on the time for a D2O reflector .
dump. The limit on level of the dump tank assures that there is sufficient free volume in the tank to
' accommodate the reflector dump.
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r 3.2.6 Bvnnuine Channels-Aplicability ( ,
This specification applies to the reactor safety system. - !
Obiective To assure that required safety channels are not bypassed during reactor operation.
Specifications l ~. The safety channels and interlocks listed in Specifications '3.2.3 and '3.2.4 as required for critical operation shall not be bypassed during critical operation of the mactor except as noted in those Specifications.
Basis Refer to the bases of Specifications of 3.2.3 and 3.2.4. .
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/m 3.2.7 Control Systems and Instrumentation Reauirements for Operation
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v Applicability This specification applies to the reactor control system and to the control console display instmmentation.
Obiective To assure that the console operator has sufficient indication of power level, reactor period, primary coolant flow, primary coolant outlet temperature, core tank level, and control device position.
Specification
- 1. Indication from the instrumentation listed in Table 3.2.7-1 shall be provided to 9 the reactor console operator prior to reactor startup and during reactor opertion.
- (V D21is The basis of this specification is given in Section 7.4 of the SAR. The limiting safety system settings are a function of the reactor power, coolant flow, coolant temperature, and core tank level. These parameters, together with reactor period, are important to safe operation.
Indication of shim blade position is also important. The operator requires continuous indication of reactor power, reactor period, and control device position in order to perform power manipulations. Ilence, these parameters are displayed on the reactor console. The operator requires knowledge of whether or not flow, temperature, and level are within their normal ranges.
IIence, they are displayed in the control room but not necessarily on console, w
3-28
Table 3.2.7-1 iO Required Instrumentauon for Delay Parameter Minimum Number Location
- 1. Period 1 Console
- 2. Neutron Flux level a) Startup - 1 Console b) Linear Power 1 Console
- 3. Core Tanklevel 1 Control Room
- 4. . Primary Coolant Flow- 1 ' Control Room 5, Coolant Outlet Temperature 1 Control Room-
- 6. Shim Blade Position (1) 5: Console
- 7. Regulating Rod Position (2) 1 Console (1) Indication required for all operable shim blades. Indication may be either numeric or analog .
meter or both.
(2) Required only if reactor is to be operated in analog automatic control mode.'
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3-29 '-
<- 2
'** .s.
- 3.3 ' < Coolant Systems 3.3.1 Natural Convection and Anti-Sinhon Valves Annlicability This specification applies to the primary coolant system.
Ob_iective :
To assure that core bypass flow is with' in design limits and to assure a safe transition from forced circuladon cooling to natural convection cooling for decay heat removal.
Specificauon
- 1. The natural and anti-siphon convection valves shall be verified closed prior to operation above 100 kW if the primary coolant pumps have been off. 'This .
check may be made visually or by use of stethoscope or other similar device capable of detecting valve closure.
- 2. The reactor shall not be made critical unless:
a) The natural convection valves are capable of automatically opening to establish natural convection cooling of the core in the event that forced cirrulation of the coolant stops.
b) The anti-siphon valves are capable of opening automatically to disrupt a -
siphon that might form if the primary coolant inlet pipe were to break.
~
- 3. The coolant ' level in the core tank shall be maintained as stated in.
- Specification 2.2 whenever the reactor is critical. At other times, the coolant level shall be maintained at or above the level of the anti-siphon values, unlessi W . there is no fuil within the core tank. >
- d. .
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, i 3-30
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Rasis Operation of the natural convection and anti-siphon valves is described in Sections 6.2 and 6.3 of the SAR. Closure of these valves assures that primary coolant flow does not bypass the core in amounts greater than anticipated in the derivation of the limiting safety system settings. -
Both sets of valves promote the establishment of natural convection cooling upon loss of forced convection. In addition, the anti-s,iphon valves prevent siphoning of the coolant should there be a .
break in the primary coolant inlet pipe.
The coolant level in the core tank should be maintained at overflow whenever possible-for the reasons stated in Section 5.2.1.I'of the SAR. However, for some maintenance operations,~,
such as blade drive changes,'it is necessary to lower the level. This is done in accordance'with -
written, approved procedures that include a requirement to monitor radiatiod levels.'
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.' 3-31 '
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3.3.2 ~ ~- H2 Cancentration Limit Applicability
' This specification applies to the H 2gas concentration in the air space above the core.-
Ob_iective To prevent a flammable concentration of H2 gas.
l Specification l 1. The-H2 concentration in the air space above the core shall not ' exceed:
3.5 volume percent.
- 2. In the event ofisolation of the air space above the core for more than one hour,~-
H2 analysis shall be performed every hour. Reactor power shall be reduced to less than 100 kW if the H2 concentration in the isolated air space exceeds 1%.
Basis The basis of this specification is given in Section 5.2.1.12 of the S'AR. The minimum explosive concentration for mixtures of H2 in the air is given as 4.1% by Ref. 3.3.2-1.* A' limit of 3.5% is therefore conservative.
In normal operation, with a continuous purge of the air space, no buildup of radiolytic -
- gases will occur. However, if abnormal radioactivity is detected in the purge stmam, the air space
- above the pool will be isolate:1 automatically. The msponse to 'such an occurrence would be for the '
~
? operator to investig' ate the cause.'of the isolation and to'open the air space and resume the -
continuous purge as soon as possible. Hence, a buildup of H2 gas is a very unlikelyl occurrence.
Nevertheless, provision ~ is made for monitoring H2 levels should the space remain isolated. The q A- l
- h.
- Water vapor (up td saturation) and pressure up to 200 atniospheres'do not affect limits. Temperatures up to 400* C do not change lower limit." [3.3.2-2]
- )
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1 3-32 j:
i n , j
q build-up rate of the H2 gas at a power level of 6 MW is 2.25% per hour as given in the basis of . l t i V
Specification 3.3.3. This is based on a measured rate from the MITR-I. Its use here is conservative because a lower fraction of the core radiation is absorbed in the H2 O coolant of the MITR-III than was absorbed in the D20 coolant of the MITR-I. Thus,if one assumes an initial level of the permissible 1% concentration of H2, continued operation for 1.0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> at 6 MW would result in a concentration less than 3.5% which has been specified as the lim: ting value. I Below 100 kW, the rate of H2O decomposition is insignificant and operation of the reactor at such power could continue with the air space isolated.
References 3.3.2-1 Lewis, R.J. Jr., " Sax's Dangerous Propenies of Industrial Materials," 8th edition, Van Nostrand Reinhold Publishing Company, New York,1994.
3.3.2-2 Weintraub, A.A., " Control of Liquid Hydrogen Hazards at Experimental Facilities,"
HASL-160. Health and Safety Laboratory, New York Operations Office, AEC, May CJ 1965. !
/
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3-33
q 3.3.3 D2Concentration Limit and Recombiner Oneration
\s' Applicability This specification applies to the D2 gas concentration in the helium blanket above the D 2O in the reflector system and to the operation of the recombiner system in the D2 O reflector system.
1 Objective To prevent a flammable concentration of D2gas in the helium blanket.
t l i
Snecification
- 1. The D2 concentration in the helium blanket shall not exceed 6 volume percent.
l
- 2. Before increasing reactor power above 100 kW and at least every two hours during operation above 100 kW the tempercture in the middle of the recombiner shall be verified to be 350' C. In addition, the recombiner flow rate shall be verified >1.5 and <8 cfm prior to reactor startup if the reactor has been shutdown for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- 3. If either of the parameters listed in Specification 3.3.3.2 above falls outside the above limits and cannot be corrected within a one hour period, D2 analyses shall be performed every hour. Reactor power shall be reduced to <100 kW if the D2 i
concentration in the helium gas cover blanket exceeds 1% while the recombiner is not operating.
-)
3-34
I i
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l7'T Basis The basis of this specification is given in Section 5.3.1.12 of the SAR. Recombination ,
of the disassociated D2 and O2is accomplished by continuously circulating the helium from above the reflector through a catalytic recombiner. The flow through the recombiner is held at l approximately two cubic feet per minute, and the recombiner operates at a temperature above 50* C as measured at the middle of the reaction chamber. A rise in temperature in the recombiner is a result of the recombination process and is positive indication that the recombiner is operating i
properly. i in a report, " Flammability of Deuterium in Oxygen-Helium Mixtures," issued by the Explosives Research Center of the Bureau of Mines [3.3.3-1), it is shown that the volume percent )
of D2 needed for flammability is independent of the volume percent of O 2 from 4 to 30%. The data in this report give the flammable concentration of D2as 7.8 volume percent at 25' C and 7.5 volume percent at 80' C. Extrapolation of these two points by a straight line approximation indicates a flammable concentration of 6.8 volume percent at a temperature of 200* C. These results are conservative because ignition in the tests was initiated at the base of the combustion tube.
The maximum temperature in the helium system will be less than 200* C under all foreseeable circumstances. Hence,it can be concluded that combustion will not occur if the D2 concentration is kept less than 6 volume percent.
Disassociation of the heavy water in the MITR-III should be considerably less than that in the MITR-I where larger inventories of D 2O were subjected to higher fast neutron fluxes. A study of the recombiner was done during operation of the MITR-I [3.3.3-2]. It was shown that the recombination efficiency of the recombiner is 100% at 2 MW and it should be 100% at 6 MW.
In the experiments at 2 MW, the recombiner was shut off for varying times and the change in D2 concentration measured. This data gives a D2 production rate of 0.75% per hour at 2 MW or 2.25% per hour at 6 MW [3.3.3-3]. Hence, the D2 concentration would not exceed 6% following
,3
(,/ a two-hour shutdown of the recombiner.
3-35
p l
The one hour fruluency for performing D2 analysis and the 1% D2 concentration limit
- specified for periods of recombiner shutdown are conservative. Specifically,if one assumes an i
I initial level of the permissible 1% concentration of D 2, continued operation for one hour at 6 MW would result in a D2 concentration less than 6% which has been established as the limiting value.
Continued reactor operation at <100 kW with the recombiner out of service is considered acceptable because the rate of D2O decomposition at this low power level is insignificant.
References 3.3.3-1 " Flammability of Deuterium in Oxygen-Helium Mixtures," TID-208998, Explosives Research Center, Bureau of Mines, June 15,1964.
3.3.3-2 J. N. Hanson, " Efficiency Study of the MITR Catalytic Recombiner," S.B. Thesis, MIT, June 1964.
3.3.3-3 File Memo (H2/D2Concentration) i 1
O 3-36
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c 3.3.4 Emernency cooline Raads=:s -j iO Annlicahilitv .. ,
' This specification applies to the emergency cooling system including pipes, valves, and .
spray nozzles.
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t
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To ensure that adequate core cooling is provided to prevent fuel overheating in the' event
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o-
- of a complete loss of coolant from the main core tank.
Specification
- 1. The reactor shall not be operated at power levels in excess of 100 kW unless the emergency core cooling system is operable and capable of providing the reactor core with a minimum total emergency cooling flow rate of 10 gal / min within 5 minutes after a low level core tank scram.
- 2. The emergency core cooling spray nozzles shall be positioned so that each fuel }
element will receive at least 20% of the average flow.
Basis The basis of this specification is given in Section 6.5 of the SAR. >The complete loss-3
- of-coolant accident requires a simultaneous massive rupture of the core tank and the ' reflector tank. l
. . .. 1
!A second type of loss-of-coolant accident would be from the rupture of the inlet pipe below the L l core level, together with a failure of the anti-siphon valves to open.
^
The primary safeguard against fuel overheating bec'ause of these extirme failures is a provision to spray cool the core with an independent continuous water supply until such time as '!
air-cooling is adequate or until recirculation can be initiated ' through od er means. The emergency
, 1
'3-37 d
.- " , u
1 spray system is a redundant system. Two independent connections to the city water, including
('N spray nozzles, piping, and valves are available, either of which will satisfy the emergency cooling requirement. This sys:em is described in Section 6.4 of the SAR, where it is shown that 9.5 gal / min is adequate flow to remove the heat generated at any time after shutdown. As discussed in the S AR, the nozzles are positioned so that each element receives at least 20% of the average spray flow. The system does not depend on the availability of normal electrical power.
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3-38
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n o I Coalant RadMvitv i imits Q 3.3.5' .
O ;
i Annlienhility"
-- t. n The specification is ' applicable to the primary, D20, shield, and secondary coolants.
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To assme detection of deterioration of components in the reactor coolant systems and to ;
identify leakage in heat exchangers.
Specification
- 1. The primary coolant shall be sampled at least quarterly. Analyses shall be '
performed for gross alpha-beta activity, gamma spectrum, and tritium. The i
radioactivity in the primary coolant should not exceed 3 times the nominal' fission product activity. If this guideline activity is exceeded, action shall be initiated in accordance with Specification 3,1.6 to determine if any in-core fuel element is damaged and Specification 3.3.6 to determine if water chemistry requirements are being met. Reactor operation may continue if the cause of the' problem is known and if such operation is allowed pursuant to Specifhations 3.1.6 and 3.3.6.
- 2. The D2O reflector coolant shall be sampled quarterly. Analysis shall be' .
performed for gross activity and tritium. The tsdioactivity of the tritium in the : -
\
D20' coolant should not exceed 5 mci /ml. If the radioactivity of the tritium in .
the D20 coolant approaches this guideline limit, preparations shall be initiated to replace the D20 coolant. Reactor operation may continue while these preparations are in progress.-
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3-39
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- 3. The shield coolant shall be sampled quarterly for gross activity. The
(]
V radioactivity of the shield coolant should not exceed 3 times the nominal activity level. If this guideline activity is exceeded, the resin,in the shield ion column should be replaced at the next scheduled maintenance ontage. Reactor operation may continue in the interim.
- 4. The secondary coolant shall be sampled daily any day that the reactor is operating or that secondary flow is supplied to a D20 exchanger. Analysis shall be made for tritium. The tritium activity shall be in accordance with Specification 3.7.2.l(a). If not, action shall be taken as required by the specification. In addition, the presence of other detectable activity in excess of 10 CFR 20 limits in the secondary coolant shall require isolation of the affected heat exchanger. Reactor operation may be continued only as necessary to I identify the affected heat exchanger.
Basis The basis of this specification is given in Sections 5.2.1.11, 5.3.1.11, 5.4.1.11 and 5.5.1.5 of the SAR. Core performance is monitored continuously by the core purge detector ]
which is addressed by Specification 3.7.1.3. The primary coolant sample analysis serves as a i backup to the indication provided by that monitor. Also,it provides a means for the detection of trends. An elevated activity level could be the result of a damaged fuel or it could be the result of an activated impurity. The former would require a reactor shutdown. The latter would require verification that the primary cleanup system was functional. Accordingly,if the guideline limit is exceeded, the actions required by Specifications 3.1.6 and 3.3.6 will be initiated. Also, the primary ion column will be evaluated for operability.
The principal concern of the D2 0 reflector coolant system is tritium which builds up slowly over many years. The system is closed and hence the tritium that is contained within it does
('M C) not pose a hazard during normal operation. However, the radiological controls needed to perform 3-40
h (
e fy maintenance activities become greater as the activity levels increases. Accordingly, a guideline U' tritium activity limit of 5 mci /ml is established along with the requirement to initiate preparations for changing the reflector coolant if the tritium activity approaches this guideline limit. Reactor operation may continue while preparations, which can take many months, are progressing.
The shield coclant is not usually radioactive although it may exhibit some activity as the - ,
result of the occasional exfoliation of system surfaces. This activity presents no hazard and the required action is to verify that the shield system's clean-up system is operable.
The secondary coolant is continuously monitored for activity by on-line detectors -
(Specification 3.7.1.6). These detectors do not sense tritium because it emits a very low energy l
beta particle. Accordingly, a daily analysis is done for tritium if secondary coolant is supplied to a D20 heat exchanger (main or cleanup). If tritium levels exceed that permitted by Specification 3.7.1.6, corrective action is to be in accordance with that specification. In addition, if other detectable activity in excess of 10 CFR 20 limits is identified, the affected heat :xchanger will be identified and removed from service.
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. l 3.3.6 ' Primary Coolant Onnlity Renyirements O
Anolicability
!This specification applies to the pH and conductivity, and activity of the primary h a
coolant.
- l Obiective -
l To control corrosion of the fuel, core components, and primary coolant loop stmeture ,
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- as well as to minimize activation of impurities and to maintain visual clarity of the coolant.-
Specification
- 1. The pH of the primary coolant shall be kept between 5.5 and 7.5, except as noted in Specification 3.3.6.3 below.
O 2. The conductivity of the primary coolant shall be kept less than 5 pmho/cm at 20' C, except as noted in Specification 3.3.6.3 below.
l
- 3. Operation of the reactor with the pH or conductivity outside the limits given in '
Specifications 3.3.6.1 and 3.3.6.2 above is permitted provided:
a) The pH is between 5.0 and 8.0,
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b)' The increase in conductivity is not the result of a chloride ion- .;
concentration in excess of 6 ppm, . .l c) Sampling of the coolant is done at least once every eight hours, and ' i d) The pH band specified in Specification 3.3.6.1 is re-established within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. .
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- 4. Specifications 3.3.6.1 and 3.3.6.2 shall apply to the coolant it. the fuel storage
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pool. Specification 3.3.6.3 shall also apply except sampling need only be done weekly and the time to reestablish the pli band shall be one nionth.
Basis The basis of this specification is given in Section 5.2.2.11 of the SAR. The purpose of pH monitoring is to ensure corrosion of the fuel, core coc.ponents, and tlic primary coolant loop stmeture is maintained within an acceptable limit. The fuel cladding and the core tank are made of aluminum alloy. A portion of the primary coolant loop is constructed of stainless steel. Lower pH will reduce aluminum alloy corrosion and oxide film formation on the fuel surface and higher pH is favored to control stainless steel corrosion. Thus, a pH range between 5.5 and 7.5 is selected for the primary coolant.
Electrical conductivity is also monitored to control purity of the primary coolant. A o
conductivity limit of 5 pmho/cm has been traditionally adopted by research reactors.
v Operation with out-of-specification chemistry is acceptable for shor' intervals. The imponant factors are plI and the absence of a high chloride ion concentration. A high conductivity by itself is not of concern. A wider pH band is acceptable for the fuel storage pool for a longer interval (one month) because the fuel is not subject to a heat flux.
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-l 3.4 j lj Reactor Cm =inment Intrerity and Pressure Relief System-4 .i
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Anol'uahility This specification ~ applies to the reactor containment building and to its pressure relief systems.
Obiective To minimize the consequence of a possible release of airborne radioactive ~ effluent from the containment building and to protect its integrity.-
Specification
- 1. Containment integrity shall be maintained, except as noted in Specification 3.4.2 below, when any of the following conditions exist:
a) The reactor is not secured, or b) Movement ofirradiated fuel is being performed, except when the fuel is in a properly sealed and approved shipping container, or c) Work involving radioactive samples is in progress in one of the reactor floor hot cells.
- 2. The requirement for containment integrity may be omitted during the performance of tests to assure the operability of the airlock' gasket deflated. j scrams. These tests shall be done with the reactor in a shutdown condition and . j l
Specifications 3.4.l(b) and 3.4.1(c) above shall be observed.
- 3. The leakage rate of the. building containment shall be less than 1%; of the' .q containment volume per day per psig overpressure when the ventilation ,
dampers are closed and the pressure relief system valves are sealed.
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.(m 4. The efficiency of the charcoal absorber in the pressure relief system shall be at
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least 95% for the removal ofiodine.
- 5. There shall be a vacuum relief system to relieve pressure when the atmospheric pressure exceeds reactor building pressure by greater than 0.1 psig.
Easis The basis of this specification is given in Section 6.5.1 of the SAR. The reactor containment is an engineered safety feature that serves as the final physical barrier to the release of radioactive particulates and gases. Proper operation of the containment, therefore, is required during all operations that could result in radioactive releases. These include reactor operation, fuel movements, and work with radioactive samples in the reactor floor hot cells.
In order to test the airlock gasket deflated scrams, it is necessary for the reactor to be in a shutdown instead of a secured condition. This activity requires about fifteen minutes per test.
The specified containment leak rate has been used in Chapter 13 of the S AR to evaluate the consequences of the maximum hypothetical accident (MHA).
The building pressure relief system provides a method for controlled depressurization of the reactor containment after initial assessment of any contained radioactivity. The specified efficiency of 95% for the charcoal filter together with ti. cified containment leak rate were used in the MHA analysis to show that the use of the pressure relief system under accident conditions will not lead to significant increased doses at the site boundary.
The vacuum relief system protects the containment building against underpressure.
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3.5 : Ventilatian S0ctam - ,
l l Apnlicahilitv1 This specification applies to the containment building ventilation system. . -
! . Obiective ..
To assure that the release of airborne radioactivity is within 10 CFR requimments.
Specification
- 1. Except as stipulated in Specification 3.5.2 below, the exhaust ventilation flow rate through the containment building stack shall be at least 7500 cfm whenever the reactor is operating at power levels in excess of 250 kW.
- 2. If the reactor is operating at power levels in excess of 250 kW and ventilation is lost, operation may continue for a maximum of five minutes. Reactor power.
shall then be reduced to less than 250 kW unless ventilation has been'restomd.
- 3. Except as noted in Specification 3.5.2 above, the containment building pmssure will be maintained below atmospheric pressure by at least 0.1 inch of water.
Basih
' Section 9.1 of the SAR describes the containment building ventilation system. A.-
minimum exhaust flow rate is specified to preclude the buildup of radioactive gases, primarily -
' Ar41. The ventilation system is interlocked so that if the building differential pressure (relative to
- the atmosphere)b' ecomes excessive, a trip will occur. . Such ' trips may occur as the result of' iweather fronts. ' Accordingly, operation without ventilation is permitted for. short intervals in order to allow for the adjustment of the stack exhaust damper and the mstoration of ventilation.
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The containment building is maintained at a negative differential pressure (relative to the atmosphere) so that any leakage is into the building. Satisfaction of this condition is one of the startup interlocks.
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O 3-47
/ 3.6 Emergency Power Y)T Anolicability This specification applies to the emergency electrical power system. -
Obiective To assure that loads imponant to safety are supplied by emergency electrical power.
upon loss of normal offsite electrical power.
Specification
- 1. Emergency electrical power with the capacity to operate the equipment listed in Table 3.6-1 shall be available when the reactor is operating and shall be capable of operation for at least one hour following a loss of normal electrical power to the facility,
[a]
- 2. Startup of the emergency power system and the transfer of all loads, except the primary coolant auxiliary pump, to the emergency power system shall be automatic.
Basis The basis of this specification is given in Section 8.2 of the SAR. The use of emergency power is probably not necessary for the MITR because loss of normal electrical power automatically scrams the reactor and because natural convection is sufficient to remove decay heat.
Nevertheless, provisions are made for emergency power to supply both a minimum set of instruments and the auxiliary pump. The information supplied to the operator will assure an orderly procedure in all such cases. The availability of the auxiliary pump will facilitate decay heat removal. The choice of a minimum one hour is based on providing information to the operator for r~}
a sufficiently long period following the scram to assure that the core is receiving adequate cooling.
3-48
Table 3.6-1 (3
(/ Minimum Equipment to be Supplied by Emergency Electrical Power
- 1. One neutron flux level channel.
- 2. Core tank coolant levelindicator.
- 3. Primary coolant outlet temperature.
- 4. Radiation monitors required by Specification 3.7.
- 5. Containment intercom system.(1)
- 6. Primary coolant auxiliary pump.
- 7. Lighting as required for personnel safety.(2)
N_nts: (1) Alternately, telephones (which are all on MITs emergency power) may be substituted.
(2) Ahernatively, self-contained battery operated lights may be substituted.
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v 3-49
F 3.7 Radiation Monitorine Systems and Effluent f~3 O
3.7.1 Monitoring Svstems I 1
I Applicability l i
This specification applies to the radiation monitoring system. j Obiective ,
To assure that facility personnel are alerted to the presence of radioactivity, both on site and in effluent paths, and to assure that the engineered safeguard features that preclude the discharge of radioactivity are operable.
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l Snecifications
- 1. Whenever the scactor containment building is occupied, there shall be an (v) operable continuous air monitor that has an audible alarm and a means of recording data.
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- 2. Whenever the containment is not isolated and containment integrity is required, the following shall be provided:
a) The main and exhaust ventilation dampers shall be interlocked with a plenum effluent monitor so that the dampers close upon detection of abnormal levels of radioactivity. This monitor shall indicate and alarm in the control room. The time for this closure to occur shall be less than the transit time of the exhaust air to traverse the exhaust plenum.
b) In the event that the main dampers fail to close as stipulated in Specification 3.7.1.2(a), the auxiliary intake and exhaust dampers shall close within ten seconds.
c) A radiation monitor that samples the stack effluent shall be operating.
This monitor shall indicate and alarm in the reactor control room.
7 i Gl 3-50
f-I,A The tritium concentration in the stack effluent shall be measured so as to j d)
'() provide the information required for reponing pursuant to Specification 7.7.1.8.
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- 3. An installed instrument capable of detecting fission products shall be used to monitor the effluent in the purge gas that is drawn through the space between the reactor top lid and the surface of the primary coolant. Portable instmments, surveys, or analyses may be substituted for the installed monitor for periods of one week or until the next scheduled outage in cases where the MITR is scheduled to continuously operate.
- 4. Whenever the reactor floor is occupied, there shall be at least one area radiation monitor capable of warning personnel on the reactor floor of ganuna radiation levels. If any area monitor is inoperable and work is to be done in that area, ,
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g portable instruments shall be used to survey radiation in that area.
- 5. Whenever secondary coolant is flowing through the D20 heat exchangers to the cooling tower the following shall be provided: l a) The secondary water shall be sampled daily for tritium content, and b) The level of the primary storage and dump tanks shall be monitored, either by low level alarms in the control room, or by hourly readings of the tank sight glasses.
- 6. Whenever the reactor is operating with secondary coolant circulating between i the containment building and the cooling tower, a secondary water monitor which indicates and alarms in the control room shall be operating.
(3 7. At least one environmental monitor at the site and one within approximately one C/
quarter mile of the site shall be used to verify compliance with environmental 3-51
l p dose limits. These devices may b' e real-time monitoring or passive-monitoring.
They shall be capable of detecting the radiation from the facility or radiation from effluent releases from the facility.
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- 8. Setpoints for the required radiation monitors shall be as listed in Table 3.7.1-1.
Basis i
The radiation monitoring system is described in Section 7.7 of the SAR. The MITR j has three continuous air monitors, one on every level of the building. Any one of the three, or a 3
portable device, can perform the required function.
If the containment is not isolated, an engineered safety feature in the form of an j interlock between any one of the four plenum effluent monitors (two gas and two particulate) will cause the main ventilation isolation dampers to close. If these fail to close, the auxiliary ones will
( do so. The main ones will close before the release can leave the building. The auxiliary will close within 10 secsnds. The effluents that would be released during even the maximum hypothetical accident during ten seconds are insignificant.
There are five possible stack monitors (2 gaseous,2 particult.te, and I area). Any one of the five fulfills the required function.
Tritium may be sampled by means of bubblers (one in the stack base and one in the exhaust plenum) or by means of special samples as described in Section 11.1.4.2 of the SAR.
The air purge that is drawn through the space above the primary coolant and below the reactor top lid is continuously monitored for fission product activity as described in Section 9.1.5.2 of the SAR. This detection method provides notice of any incipient fuel clad failures.
Area monitors are located throughout the containment building. At a minimum, one is required to be operable on the main reactor floor when it is occupied by experimenters. However, if any one of these monitors is inoperable, portable instruments will be substituted.
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3-52
,o Monitoring for heavy water leakage into the secondary coolant is based on three V, .
mdependent measurements. These are:
a) The secondary water monitor is a gamma-sensitive scintillation detector. It cannot detect tritium but is sensitive to N16 and F17 which are also present in heavy water when the reactor is operating.
b) Daily sampling of the secondary water will allow detection of very small leaks.
c) Because of the nature of the primary and reflector systems, any loss of coolant inventory will be reflected by a decrease in the level in either the primary storage tank or the D20 dump tank.
The secondary water monitors will also detect primary leakage.
The environmental monitors are used to verify that the potential maximum dose, annual or other, in the unrestricted environment is within the values analyzed in the SAR.
Action guidelines are chosen to alert operator that attention to the condition is warranted. Guidelines are not absolute limits and operation in excess of these guidelines will not r)
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U necessarily result in exceeding 10 CFR 20 limits which are based on annual averaged quantities.
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4 3-k L3.7.2. Effluents A}nnliemhility ' ,
This specification applies'to the radioactive effluents that are released from the reactor - d
, site.
.3 Obiectivity
~
To assure that the release of radioactive effluents to the environment is within the limits .
of 10 CFR Part 20.
Specification
- 1. The release of radioactive effluents from the reactor site will comply'with all provisions of Part 20, Title 10, Code of Federal Regulations with the following :
exemptions:
a) A dilution factor of 50,000 shall be applicable to the concentration of-
. airborne effluents released from the stack. For particulates and iodines with half lives greater than eight days in each case, an effective dose scaling factor of 1,200 is applied to the dilution, b) Given that periodic leakage can occur between the D2O system and the cooling tower water, tritium from the reactor secondary coolant system is exempt from 10 CFR 20 limits with the requirement that on indication of I pCi/ liter of tritium in the secondary coolant water, the cooling tower spray shall be shut down, the secondary system water discharge shall be stop xd, and the D 2O reflector heat exchangers'shall be isolated '
until tritium . eakage into the secondary has been controlied.
Basis Concentrations of radioactive gases from the MITR stack will be maintained as low as-seasonably achievable. Because of atmospheric diffusion and variation in wind direction, the
- yearly average concentration ~of radioactive materials in all onrestricted areas which could be-3-56i m ,s q >
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occupied by individuals will be much less than the maximum permissible concentrations-established in 'IO CFR 20.'
The use of a maximum dilution factor of 50,000 for radioactive effluents that arei discharged from the'MITR ventilation stack.will still maintain all possible yearly averaged -
concentrations to a small fraction of the 10 CFR 20 limits in any occupied area. This' dilution g
- factor is applicable to all gaseous effluents with the exception of radioactive isotopes that are" subject to environmental reconcentration such as Iodine-131. The dilution factor was ' calculated using the CAP 88-PC code for straight-line Gaussian diffusion [3.7.2-1]. This gave a.value of -
>lx106 at 900 m which was the point of maximum dose to an individual. Calculations made in .
.c onjunction with the MITR-II startup allowed for the effect of nearby buildings. These and subsequent calculations gave a dilution factor of at least 50,000. Accordingly, a factor of 50,000 is selected. An effective dose scaling factor of 1,200 is applied for radioactive iodines and particulates with half-lives greater than eight days to account for differences in dose pathways and dose conversion ratios. This factor is determined by estimating the dose from all pathways for' Iodine-131 compared to noble gases for a unit release, based upon values generated'using the CAP 88-PC code. This factor assures that, if the MITR stack releases ofiodines and particulates -
with half-lives greater than eight days are kept within the 10 CFR 20 limits at the nearest point of public occupancy, the potential for radiation doses after dilution will be a small fraction of the 10 CFR 20 limits.
Liquid waste is discharged to the municipal sanitary sewer systems from two waste -
s torage tanks and from the cooling tower basin. Radioactive nuclide concentrations that are less than the limits set on the monitoring and sampling systems are such that conformity with the-limitations specified in 10 CFR 20 is assured, with the po_ssible exception of tritium. In this case, the average concentrations released in both water and air are _well below the permissible-concentrations, but the total amount of tritium released to the sewer in a year might exceed the one .
curie permitted by' 10 CFR 20 because of leakage of reflector D2 O (maximum' tritium concentration expected to be 5 mci /cc) into the secondary coolant.
L3 .
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The average discharge of water from the secondary coolant system of the reactor to the sanitary sewer system is approximately 13,000 gallons per operating day. This is diluted by an approximately equal volume of water from the Nuclear Engineering Building. The sewerage from the site enters the Cambridge sewer system where it is further diluted by discharge by the rest of MIT (at about 1 MGD) and by an unmeasured amount of storm drainage. The Cambridge sewerage enters a Metropolitan District Commission tmnk sewer line. The estimated discharge rate from this line, based upon the permitted discharge rate, is 436 MGD [3.7.2-2].
Thus, the reactor effluent is diluted by a factor of about 4.3 x 108/1.3 x 104 = 30,000 at the ultimate point of discharge from the sewer system. The proposed limit of I Ci/ liter in the cooling tower water assures concentrations at the point of discharge from the sewer system will be well below the limit of 10 CFR 20.
References
' 3.7.2-1 Parks, B.. " Mathematical Models in CAP 88-PC," U.S. Department of Energy, June !
1997.
3.7.2-2 Massachusetts Water Resources Authority, " Monthly Compliance and Progress Report," June 1999.
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L 3.7.3 Reactor Floor Hot Cells Apolicability This specification applies to the reactor floor hot cells.
Objective To assure that the hot cells are utilized as described in the SAR.
Specifications
- 1. Whenever the hot cells are used the following shall be provided:
a) No experiments shall be conducted in the hot cells and on-going experimer ts shall be terminated when the containment building and/or hot cell ventilation systems are shut down.
b) The hot cell blower shall be interlocked with the containment building exhaust system blower. The hot cell blower shall automatically shut off O v when the main ventilation system blower or dampers are shut down.
c) In event of a fire in the hot cell, the hot cell blower shall be automatically shut off and an alarm shall sound in the hot cell area and in the control room. l d) A visual alarm in the hot cell area shall activate when the containment l building exhaust system is shutdown. !
1 Basis The basis of this specification is given in Sections 10.2.8 and 10.3.2.7 of the SAR.
When the containment ventilation system is shutdown, the exhaust for the ventilation gases and i
particulates from the hot cell is lost. Accordingly, experiments in the hot cells should not be initiated and those in progress should be terminated until the containment ventilation system is l
again operable.
In event of a fire in the hot cells, the hot cell blower that exhausts the atmosphere from r] the hot cells shall be shutdown. When the containment building exhaust system is shutdown, it is
- 1 necessary to alert any operators involved in the hot cell operations that those operations must be 3-591
.i terminated until the containment blowers are once more operable Accordingly, visual alarms are I located in the hot cell area.
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3-60
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i C 3.7.4 ' Bvoroduct Material .
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? Annlicabilhv This spec'ification applies to the receipt of byproduct material on the Reactor Operating j License.
Oblective To establish criteria for the receipt of byproduct material on the Reactor Operating License, q Specification
- 1. Byproduct material will be limited to the following:
a) Atomic numbers 3 through 83 in solid form for the purpose of materials 4 studies not to exceed 1,000,000 Curies total inventory,10,000 Curies O per specimen and not to exceed 1,000 rem /h at one meter from an unshielded source, b) Atomic numbers 3 through 83 in any form for the purpose of-calibration, characterization, and detection for radiation protection purposes not to exceed 1000 Curies.
This authorization allows receipt of byproduct material for use in radiation damage j studies and for radiation protection purposes.- Materials would be used in the reactor floor hot cells.
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N 4. SURVEILLANCE REQUIREMENTS
]
4.1 Reactor Core Parameters -
Applicability This specification applies to surveillance of reactor core parameters.
Obiective To assure that reactor core parameters are as specified in the analyses cantained in the SAR.
Soecification
- 1. Excess Reactivity: The operability of the subcritical limit interlock shall be O verified at least annually.
U
- 2. Shutdown Margin: The shutdown margin shall be determined at least annually and after changes in core configuration, refuelings, the insertion / removal of in-core experimental assemblies, and control device (shim blade and/or regulating rod) changeouts.
- 3. Core Configuration: Refer to Specification 3.1.4.4(c).
- 4. Reactivity Coefficients: Refer to Specification 3.1.5.1. In addition, reactivity coefficients will be measured if analysis of a proposed change of core
' configuration or fuel type indicates that there could be a significant change in the magnitude of the coefficient in question.
(%)),
4-1
G '5. Fuel Parameters: ,
a) Fuel elements' shall be' inspected quarterly;in accordanbe with
' Specification 3.1.6.2(a) and 3.11.2(b).
b) ~ The fission density limit (Specification 3.1.6.3) shall be calculated for all in-core elements whenever the reactor is refueled.
c) Self-protection of fuel in the spent fuel storage pool shall be verified,by '
measurement at least biennially if any element has been discharged from the core for more than 10 years. Measurements which show that the '
worst-case elements (those with the least burnup and longest time since discharge) are self-protecting shall satisfy this requirement.
d) The measurement frequency required by Specification 4.1.5(c) shall be changed to annual if any element is within a factor of two of the self ..
protectionlimit. s.
Basis The subcritical limit interlock limits the amount of excess reactivity by requiring that the reactor not attain criticality unless all shim blades are withdrawn by at least four inches. This" ,
physical interlock is preferable to an administrative limit.' Shutdown margin is determined and' documented in writing any time there is a change of core configuration, refueling, insertion / removal of an in-core sample assembly, or control device changeout.
Compliance with thermal-hydraulic limits is verified for every change of core configuration. This includes refuelings.
Reactivity coefficients were measured during the initial startup of the MITR-II.
Coefficients have been periodically remeasured either as teaching exercise or as part of the program for the evaluation of digital control strategies. No significant changes have been noted.- The commitment is therefore to investigate any change should one be observed and to remeasure if a
- change in core design or fuel could cause a significant change.
The , procedure for performing . fuel inspections is ' noted 'in the basis of
- Specification 3.1.6.-
The fission density limit is calculated as part of the MITR-II fuel management program for every element. - It is updated at every refueling for those elements that are in-core.
42-h
- 'I,
tN Self-protection of the fuel is assured for at least a year by operating an element in-core U for about a week. Elements are in-core typically for a three to five year period. Hence, there is no need to measure dose rates. After sixty days of cooling, radiation levels exceed 10,000 rem / hour unshielded at one meter. Measurements of dose rates on elements in the spent fuel storage pool show that they remain self-protecting for at least 10 years [4.1-1]. No MITR-II element has ever dropped below the self-protection limit during the time that elapsed between its discharge from the core and its return to the U. S. Department of Energy.
References 4.1-1. File Memo, " Fuel Self-Protection," February 1997.
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4.2 Reactor Control and Safety Systems Annlicability ,
This specification applies to the surveillance of reactor control and safety systems.
l 1
Obiective To assure the reliability of the reactor control and safety systems.
Snecification Reactivity Wonh of Control Devices: The integral and differential worths of the j 1.
six shim blades and of the regulating rod shall be measured at least annually.
Either calculations of the expected change or measurements shall be made upon l
changeout of an absorber and upon significant changes in core configuration.
O 2. Rod Withdrawal and Insertion Speed: The withdrawal and insertion speed of (D each shim blade and the regulating rod shall be verified annually.
- 3. Scram Times: The scram time of each shim blade shall be verified annually or whenever any work has been done on either the shim blade, its electromagnet, or its associated drive. For purposes of this check, the scram time shall be measured from the full-out position to the 80% inserted position of the shim blade.
- 4. Scram and Power Measuring Channels: The instruments or channels listed in Table 4.2-1 shall be tested at least quarterly and each time before startup of the reactor if the reactor has been shut down more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or if the instrument or channel has been repaired or de-energized. Calibration of these J.
4-4 1
instruments or channels (except those such as scram pushbuttons that do not .
requhe calibration) shall be done at lest annually.
- 5. Onerability Tests: Operability tests of the instruments or channels lis'ted in Table 4.2-1 shall be performed if the channel or instrument has been modified or repaired.
- 6. The following instruments shall be calibrated and trip points verified when initially installed, any time a significant change in indication is noted, and at
. least annually:
a) Period b) . Neutron Flux Level c) Primary Coolant OutletTemperature d) Core Tank Level e) ReflectorTanklevel f) Primary Coolant Flow g) D 2O Reflector Flow h) Shield Coolant Flow i) Period Channel Level Signal Off-Scale 7, Thermal Power Indicator: The thermal power indicator shall be calibrated at I least annually.-
- 8. Heat Balance: The signal from the linear power channel shall be checked against ,
a heat balance calculation at least monthly, for any month that the reactor is i
- ( operated above 1 MW continuously for at least 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. ;
., 4-5 :
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- 9. ' Control Device Insoection: Control devices shall be inspected annually as follows:
a) Shim blade absorbers shall be checked visuallp.
~
b) Shim blade electromagnets shall be checked both visually and by:
measuring the resistivity of the coils.
c) Shim blade and regulating rod drives shall be monitored for proper operation.
Basili The MITR has observed the criteria given in Specification 4.2.1 for determination of -
control device reactivity worths and found it to be adequate. Measurements of the integral and differential worths are required annually. Measurements following changeouts of absorbers and '
change of core configuration are desirable. However, such measurements are very time ,
consuming. Moreover, sufficient experience exists with such changes that their effect on integral and differential reactivity worths can be predicted with reasonable accuracy. Accordingly, normal MITR practice is to do a complete set of measurements following replacement of all absorber sections rather than to do measurements as each is replaced. (Note: It requires several days to l replace one absorber and the entire process is usually done over an interval of several months.)
Estimates of the change of worth are used pending the measurement. Estimates, not measurements, are normally used for changes of core configuration.
The insertion and withdrawal speed of the control devices is fixed by the motor and drive design as discussed in Section 4.2.2 of the SAR. These speeds are verified annually.
Scram time is as defined by Specifications 1.3.37 and 3.2.1.2. It is verified at least annually and whenever maintenance has been performed that could affect it.
s The instruments and channels listed in Table 4.2-1 correspond to those in a
.i Table 3.2.31, 1 " Required Safetp Channels" with the exception that surveillance of the' building
.g overpressure and gasket deflated scrams is addressed elsewhere (Specification 4.4).
4-6 i
-(
m Table 43-1 Surveillance of Scram and Power Measurine Channels Instrument or Channel Surveillance 1 Period Scram Test
- 2. Neutron Flux Level Scram Test
- 3. Primary Coolant Outlet Temperature , Scram Test
- 4. Core Tank Level Scram Test
- 5. ReflectorTank Level , Scram Test
- 6. D 2O Dump Valve Switch Scram Test
- 7. Containment Isolation and Reactor Major Scram Test Shutdown
- 8. Magnet Currents Cut-Off on Scram Signal OperationalTest
- 9. Experiment Shutdown As Specified in Experiment Approval 4
- 10. Primary Coolant Flow
- Scram Test 11 D20 Reflector Flow
- Scram Test
- 12. Shield Coolant Flow
- Scram Test
- 13. Fission Converter As specified in Fission Convener 4 SER
- 14. Period Channel Level Signal Off Scale Scram Test
- 15. Hold-Down Orid Unlatched Scram Test
- 16. Reactor Remote Shutdown (s) Scram Test from Medical Facilities l
- Not requin:d for stanup in natural convection cooling mode.
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4-7
,q The thermal power indication is calibrated at least annually and the signal from the O linear power channel is compared against a heat balance at least monthly for any month that the reactor is operated above 1 MW. These actions are done under conditions of thermal equilibrium which, because of the MITR's heat capacity (especially that of the graphite reflector), occurs after 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of steady-state operation.
Control devices are inspected at least annually. The inspection focuses on those i
components that are important to safety. Those include the absorber sections (Section 16.3.1.5 of j the SAR) and electromagnets (Section 16.3.1.4(d) of the SAR). The status of the shim blade and regulating rod drives can be deduced from external observations such as the measurement of blade and regulating rod insertion / withdrawal speeds (Specification 4.2.2). Internal inspections require lowering of the core tank level and removal of the drive. These are usually done whenever an l l
absorber is changed out. As described in Section 16.3.1.5 of the SAR, this is normally done l cvery 125,000 MWH. A prespecified frequency for an internal inspection would involve serious f)
V ALARA issues.
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4-8
j 4.3 Coolant Systems ' j
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LJ' Applicability j This specification applies to the surveillance of the reactor coolant systems.
Objective To assure that reactor coolant systems are maintained as specified in the analyses in the j
SAR.
Sr>ccifications
- 1. Each emergency core cooling system (ECCS) shall be tested at least annually.
The minimum flow through each ECCS spray nozzle system shall be 10 gpm. l The test shall include verification of operability of the manual valves to the city
( ~) water supply lines and the core spray nozzles. l s.s
- 2. If an In-core Sample Assembly (ICSA) of a type not previously evaluated is to be installed, verification that the ECCS spray nozzles are positioned so that each fuel element receives at least 20% of the average flow shall be determined by :
l' measurement using an ex-core mock-up if the ICSA has the potential to obstmet ECCS spray.
- 3. In-service inspections of core components shall be performed quarterly. j 1
l
- 4. Analysis of coolant for radioactivity shall be as required by Specification 3.3.5. 1 l
A l a
4-9
(3 5. Sampling for H2 in the core purge gas and D2 in the helium blanket above the V D2O reDector shall be as required by Specifications 3.3.2 and 3.3.3 respectively and at least annually.
- 6. The pH and conductivity of the primary coolant and fuel storage pool shall be '
measured at least quarterly. Whenever the reactor is operating, primary conductivity shall be verified at least weekly and ifit is out of speci5 cation, pH shall be measured.
Basis The test of each emergency cooling system will consist of opening the manual valve in the city water supply line to ensure proper operation of the valve and of spraying water into the core through the core spray nozzle. During the test, the normal flow path will not be used in order to prevent adding city water to the primary reactor coolant. The city water will be diverted to the liquid effluent system without discharging through the spray nozzle while primary coolant water will be used to test the spray nozzle. I In-core sample assemblies may contain extensions that serve as conduits for the experiment coolant, sensors, or heater cables. These conduits may obstruct ECCS nozzle spray.
If an ICS A of a type not previously evaluated is to be installed and if it has the potential to obstruct ,
1 l
ECCS nozzle spray, then an ex-core mock-up is used to verify that each fuel element will receive at least 20% of the average ECCS spray.
In-service inspections of core components are performed at the same time as the fuel-inspection required by Specification 3.1.6. l 1
The frequencies for analysis of coolant radioactivities, H2 and D2 gas are each i l
- governed by their respective technical specifications.
The minimum frequency for pH and conductivity sampling is quanerly. However, !
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V' primary samples are taken prior to most startups and conductivity is monitored on-line. For pure 4-10
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water, pH and conductivity vary linearly with each other. Therefore, the reading of the conductivity instrument is checked at least weekly and,if the conductivity is out of specification, the pH is measured. If the conductivity is in specification, the pH will also be in specification.:
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- Co'ntainment Surveillance O .4.4.-
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Applicability This' specification applies to the surveillance of the reactor building containment and its associated systems including the vacuum breakers, pressure relief system, and associated scrams;.
Obiective To assure that building integrity is as specified in the SAR.
- j. Specification
- 1. An integral air leak rate tes't of the reactor building containment shall be performed biennially. Each test shall be performed at a test pressure greater than 1 psig and less than 2 psig with the blow-offleg set at below 2 psig.
t 2, Leak tests of individual penetrations shall be performed between integral tests when either new penetrations or repairs of existing penetrations are made. 'The sum of the results of the last integral building leak rate test and any increase in ~'
the penetration leakage since the integral test must satisfy Specification 3.4.3.-
- 3. The main and auxiliary intake and exhaust ventilation isolation dampers shall be inspected annually.
E 4. A test of the proper functioning of the independent vacuum relief breakers shall be performed annually.
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A test of the charcoal filters in the pressure relief system shall be performed
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V annually to determine their efficiency for the removal of the elemental iodine.
The filters shall be replaced if the efficiency is 95% or less.
- 6. Flow shall be established through the charcoal filters in the pressure relief system for 15 minutes or longer, at least monthly and each time before startup of the reactor if the reactor has been shut down more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- 7. The following channels shall be tested at least quarterly or if the channel has been modified or repaired.
Channel Igs_t Building Overpressure Scram Test Main Personnel Lock Gaskets Deflates Scram Test
(_.) Basement Personnel Lock Baskets Deflates Scram Test Basis The reactor containment building has been tested annually for leakage from 1958 to 1999. On only one occasion has the facility failed to pass the leak rate test. In that one case, the margin by which the facility failed was very small and was caused by deterioration of the rubber gaskets on the intake and exhaust ventilation isolation dampers. Because each of these two I dampers is redundant (main and auxiliary), the building could still have been maintainea at the test pressure within the permissible leak rate in the event of an emergency. Inspeccion of the l l
containment design indicates that the most probable points of excessive leakage are the isolation dampers, which has been confirmed by the containment testing history as stated above. Other seals on the reactor building are either welded or duplicated. Failure of these seals is therefore much less probable than damper gasket failure.
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4-13
In view of this conclusion, the venti'lation isomion dampers will be visually inspected (3-N) annually. An integrated leak rate will be conducted biennially. The results of that test are conected for temperature, pressure, and humidity changes, and an error analysis is also made to estimate the limits of uncenainty in the measurements.
New or repaired penetrations are tested as necessary using special procedures that are prepared for the particular penetration.
The vacuum breakers are tested under pressure as an integral part of the containment leakage test. It is necessary also to test them under vacuum to ensure that they will open properly.
The building pressure relief charcoal filters will be tested by methods described in Section 7.5.2 of Reference 4.4-1 or the equivalent. Successful experience with the system over the past twenty-five years justifies the annual frequency of testing.
Air flow is established periodically through the charcoal filters in order to maintain the charcoal activated so that it will remove iodine. This is done in accordance with the vendor's n
recommendations.
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Scram channels associated with the containment building are tested quarterly, i Experience over the past twenty-five yeart (1974-1999) has shown this frequency to be more than adequate.
l Reference 4.4-1 Burschsted, C. A. and A. B. Fuller, " Design, Construction, and Testing of High-Efficiency Air Filtration systems for Nuclear Application," ORNL-NSIC-65, January, 1970.
4-14
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. 4.5.- Ventitation Systems ! ,
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Applicability This specification is applicable to surveillance of the containment building ventilation 1 system.
Obiective To assure that the containment building ventilation system functions as described in the
-I
. S AR.
Specification
- 1. The flow rate through the ventilation systems exhaust stack shall be measured at least annually.
O 2. The following interlocks shall be tested at least quarterly and each time before startup of the reactor if the reactor has been shut down for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or if the equipment / circuitry involved has been repaimd.
Interlock Surveillance Building AP - Reactor Start - . Operation Test ;
Plenum Effluent Monitor - Main Damper Trip - Operation Test .
, Main - Auxiliary DamperInterlock Operation Test
- 3. - The Building AP - Reactor Start interlock shall be calibrated annually.
4.- The AP across the filters in the ventilation exhaust shall be monitored and the
', filters replaced when the AP exceeds the manufactumr's recommendations."
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- 4-15.
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O Experience from both the MITR-I (1958-1973) and MITR-II (1974-1999) shows that1 4
an annual measurement of ventilation flow rate is sufficient to detect trends. The interlocks listed -
under Specincation 4.5.2 of the specification are tested in the same manner as scram channels.- 1 The AP across the filters is monitored by permanently installed manometers. These allow the AP to be monitored for trends.
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' 4.6' - Emergency Electrical Power Systems Applicabihty This specification applies to the surveillance of the emergency electrical power supply.
- Obiective -
To assure that the emergency electrical power supply is maintained and tested in accordance with accepted standards.
Specification
- 1. The voltage and specific gravity of the pilot cell shall be measured quarterly.
- 2. The specific gravity of all batteries shall be measured at any time that a significant change is noted in the pilot cell and at least every two years.'
- 3. A discharge test shall be performed once every two years.
- 4. Operability of the inverter motor-generator set and associated switches shall be verified annually. Performance of a discharge test satisfies this requirement. -l J
Basis 1 I
The emergency electrical power system consists of batteries, an inverter motor. j generator set, and'the switches necessary to tie into the nonnal electrical distribution; system.
Specific gravity and voltage measurements of individual battery cells are the' accepted method ofJ ensuring that the batteries are in satisfactory condition.;In addition, periodic discharge tests are
' performed to detect deterioration of cells. To ensure the operability of the inverter motor-generator Q;/m
' set, the generator and associated switches _will be operationally tested.
J
/^N The frequency of these component tests are based on experience to date with the MITR U and on standard practice as recommended in ANSI / ANSI 5.1- 1990. Where appropriate, the latter has been modified by the manufacturer's recommendations. Thus, the discharge test frequency is biennial rather than every five years.
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. 4_,7 c Radiation Monitoring Svstems' and Effluents 4.7.1- Radiation Monitoring Systems Applicabihty
.This specification applies to the surveillance of the radiation monitoring systems.
1 Obiective To assure that radiation monitoring systems are maintained as specified by the SAR.' q i
1 i
Specifications q
- 1. A channel check shall be made of the area and effluent radiation monitors on .
any day that the reactor is operating. In addition, this check shall be done priorf l
- to any reactor stanup, if the reactor has been shutdown for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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- 2. The following radiation monitors shall be tested at least monthly and each time before startup if the reactor has been shut down for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or if the -
instrument has been mpaired or de-energized. j l
j Monitor Test !
a) Ama Radiation Channel Test Using a Source i
b) Plenum Gas and Particulate ChannelTest Using a Pulse - l c) Stack Gas and Paniculate ChannelTest Using a Pulse d) Secondary Coolant ChannelTest Using'a Pulse f9 4 m l
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p E .3. ; The following radiation monitors shall be tested quarterly: , h V
, 1 Monitor = Test '
a) Plenum Gas and Particulate Channel Test Using a Source b) Stack Gas and Particulate ChannelTest Using a Source c) Secondary Coolant Channel Test Using a Source .
- 4. The radiation monitors listed in Specification 4.7.1.2 shall be calibrated and the trip points verified when initially installed and annually thereafter. -
a Basis i The channel check provides a qualitative verification of the performance of the radiation i monitors. The channel tests verify operability by the introduction of a test signal. The calibration provides a complete verification of the performance of the instrument.
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4.7.2-O - Effluents -
Annlicability This specification applies to the surveillance of the quantities of radioactive effluents -
released td the environment.
- Obiective -
To assure that releases to the environment are in compliance with the requirements of:
.10 CFR 20.'
y Specification
- 1. Continuous monitoring of effluents and collection' of information 'shall be performed as required for reporting as specified within Specification 7.7.1.8.
O 2. Comparison of the information collected to 'other methods such 'as -
environmental monitoring to verify the data (s) validity should be used as necessary.
Has This information is documented in the facility's annual repon.
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51 DESIGN FEATURES O
5.1 Site and Facility Description Applicability This specification applies to the reactor site and facility.
Obiective To assure that features of the site and facility which, if altered, would significantly affect safety are specified.
Specification
- 1. The reactor facility is located at 138 Albany Street on the MIT Campus in the City of Cambridge, Massachusetts.
O 2. The distance to the nearest point of normal public occupan:y is at least 68 feet.
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- 3. The restricted area shall include the reactor containment building, the adjoining -
one-story building, and the fenced-in area in which the cooling tower ventilation exhaust stack and liquid waste storage tanks are located.
- 4. The height of the ventilation exhaust stack is at least 150 feet.
- 5. . The volume of the reactor containment building is approximately 200,000 cubic feet.
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h The site location is specified to ensure ownership of the' site by the licensee; The '
. reactor site is described in Section 2.1.1 of the~ SAR. The closest. point of public access is' a
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railroad track that runs _ parallel to the south side of the restricted area. However, that location is
.normally not' occupied. The nearest point of normal public occupancy is tiie sidewalk on the nonherly side of the site.
' Access to the restricted area is limited to authorized personnel only.
The height of the ventilation exhaust stack and the free-air volume of the containment:
building are design features that bear on radiological safety. The figures specified are the ones used in the SAR for effluent calculations.
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77 5.2) ' Primary Coolant System--
Annlicability -
This specification applies to the design of the primary coolant system. :
l:
L Obiective
~
To assure compatibility of the primary coolant system with the safety analysisc l
l Specification
- 1. ~ The reactor coolant system shall consist of a reactor vessel, a single cooling '-
. loop containing two or more heat exchangers, and appropriate pumps and.
valves. All materials, including those of the reactor vessel, which are in contact with the primary coolant (H2O), shall be aluminum alloys or stainless steel, except for small non-corrosive components such as gaskets l filters, and valve -
diaphragms. The reactor vessel shall be designed in accordance with the ASME Code for Unfired Pressure Vessels. It shall be designed for a working pressure of 24 psig and 150' F. Heat exchangers shall be designed for 75 psig and a temperature of 150' F. The ccnnecting piping shall be designed to withstand a-1 60 psig hydro test.
Basia The reactor coolant system originally consisted of a single loop .that contained two h' eat ~
exchangers. It was subsequently modified to add a third heat exchanger. It is normally operated J with only two heat exchangers on line. Core safety is unaffected by the number of heat exchangers j
1 provided that the required heat transfer surface ama is available for heat removal and that primary coolant flow remains as required by Specification 3.2.3.
The materials of construction are primarily aluminum alloy and stainless steel and are -
~
- chemically compatible with the H2O coolant.' The design, temperature, and pressure of the reactor
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vessel and other. primary system components provide adequate margins over operating temperatures and pressures. The reactor vessel was designed to Section VIII,1968 edition, of the -
ASME Code for Unfired Pressure Vessels. Subsequent design changes will be made in accordance with the most recent edition of this code.
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'5-4
' 5.3 ~ Wactor Core and Fuel Apnlicability This specification applies to the design of the reactor core and the fuel.'
! Obiective q l.
l To assure compatibility of the reactor core and fuel with the present safety analysis._ .
Specification l
(
- 1. The reactor core may consist of up to 27 fuel elements approximately 2-3/8" on -
a side. The fuel shall be plates of uranium in the form of UAl alloy.or UAlx with a maximum of 50 w/o uranium in the fuel matrix clad by a layer of aluminum metal that incorporates fins on the surface to enhance heat transfer.
The fuel plates shall have a nominal clad thickness not less than 0.015 inchss at the base of the groove between the fins, with local areas not less than 0.008 inches.
- 2. Design ofin-core sample assemblies shall conform to the following criteria:-
They shall be positively secured in the core to prevent movement during - j a) reactor operation. .l
- i b) Materials of construction shall be radiation resistant and compatible with j those used in the reactor core and primary coolant system.' ]
c) Sufficient cooling shall be provided to ensure structural integrity of the . 1 assembly and to preclude any boiling of the primary coolant.
4
'd) -~ The size of the irradiation thimble shall be less than'16 square inches in
, cross section.
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Cj The thermal design analysis in the SAR and the power distributions on which the analysis is based assume fuel elements of the type specified in Specification 5.3.1. Any; change in :
this design would require re-evaluation of the heat transfer and flow characteristics of the element.
The nominal clad thickness of 0.015 inches is based upon standard practice for MTR -
type elements with clad of similar thickness. Referen e 5.3-1 states that the release of radioactive fission gas to the primary cooling water appears to be adequately prevented by cladding of uniform L thickness >0.2 mm (>0.008 inches). Fabrication specifications for MITR fuel provide for a a nominal clad thickness of 0.015 inches at the base of the grooves. Fabrication tolerances and 1 minor manufacturing deviations (e.g., scratches, indentations, etc.) may result in the clad thickness .
being reduced to not less than 0.008 inches in local areas. Reference 5.3-2 shows that a thick clad increases the delay time for heat removal in the event of a fast transient. Therefore, the clad should j
! be as thin as possible while still remaining compatible with fission product retention requirements.-
O In-core sample assemblies which satisfy Specification 5.3.2(a)lcannot be credibly' l V
ejected during operation and are therefore considered part of the reactor structure. Specifications 5.3.2(b) and 5.3.2(c) ensure the structural integrity of the assembly and prevent chemical interactions with the core and primary coolant system. Specification 5.3.2(d) limits the size of the irradiation area as required by 10 CFR 50.2.
References i
5.3-1 Beeston, J. M., R. R. Hobbins, G. W. Gibson, and W. F. Francis, "DevQ nent and Irradiation Performance of Uranium Aluminide Fuels in Test Reactors," Nuclear
. Technology, Vol. 49, p.136-149 (June 1980) 5.3 Thompson, T. J. and J. G. Beckerly (eds.) The Technology of Nuclear Reactor Safety, j Vol. I, The MIT Press, Cambridge, Mass. (1964) '
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5.4 . Fissile Material Storage Applicability This specification applies to the storage and handling of fuel elements.
Obiective To assure that fuel elements, which are not currently in the core, will be stored safely.-
Specification
-1. Fuel elements shall be stored in accordance with the requirements of the MITR '
Security Plan.
- 2. Unirradiated fuel elements may be stored in any of the following locations, q subject to the MITR Security Plan and to Specification 4 below:
a) In the reactor core provided that the reactivity is below the shutdown ,
margin given by Specification 3.1.2.
b) In the cadmium-lined fuel storage ring that is attached to the flow shroud, c) In the storage safe in the reactor containment building (fuel plates also).
o
- 3. Irradiated fuel elements can be stored in any of the following location:
a) In the reactor core provided the reactivity is below the shutdown margin given by Specification 3.1.2.1, b) In the cadmium-lined fuel storage ring attached to the flow 'sliroud, -
. c) - In the spent fuel storage tank in the basement of the reactor building,-
d)- In the fuel element transfer cask or other proper shield within the controlled area, and I.i m -
e)- 'In the fission converter tank.- l
.5-7J
- 4. Handling of fuel elements: A maximum of one fuel element shall be moved in or out of the reactor core at a time. Not more than four of the MITR fuel elements or the equivalent of two fuel elements including loose plates (maximum of 15 loose fuel plates) shall be outside of the approved storage areas except during the prccesses of receiving or shipping fuel from the site in approved containers. In all cases of fuel element storage outside of the reactor >
core, the value of kerr shall be less than 0.90. Records of fuel element transfers shall be maintained. Prior to transferring an irradiated element from the reactor vessel to the transfer cask, the element shall not have been operated in the core l at a power level above 100 kW for at least four days.
Basis One of the principal requirements in regard to storage of fuel elements is prevention of ,
I n accidental criticality. The locations given in Specifica' ions 5.4.2 and 5.4.3 provide for complete J
v criticality control. The reactor itself is, of course, shielded and appropriate written procedures assure that it is loaded properly. The fuel elements in the cadmium-lined storage ring in the core tank are neutronically isolated from the reactor core by the cadmium of each individual box and by the six control blades that surround the reactor core. The fresh fuel storage safe and the spent fuel storage pool both have carefully designed geometric arrays to assure that criticality will not occur.
The row of storage positions in the safe are separated by cadmium, and each fuel element box in i
the pool is lined with cadmium. The transfer cask can only hold one element at a time. Finally, the fission converter is desigt 2d as a suberitical facility.
Calculations have been made, by using the methods described in Chapter 4 of the SAR, which indicate that at least 8-l/3 MITR fuel elements with optimum spacing and optimum moderation are required for criticality. Similar calculations show that 72 fuel plates with 27 grarns U-235 are required for criticality, giving a safe handling limit of about 30. The specification of no more than four elements outside of the designated storage areas assures that no criticality will occur 5-8
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~
elsewhere. No calculations have been 'done for 34 gram plates, and so a conservative limit of 15
- unassembled plates is established.
It has been calculated that fuel elements when stored in the locations specified in 2b,2c, .
. 3b,3c,3d, and 3e will have an effective multiplication (ken) factor of less than 0.9 under optimum i
! conditions of water moderation (5.4 1. 5.4-2].
The chief additional problems with spent fuel are those.of shielding personnel from the -
emitted fission product gamma rays and' preventing melting from decay heat. The shielding requirement is met by utilizing a shielded transfer cask (item 3d) for movements and temporary .
storage and more permanent shielding as indicated in 3a, b, c, and e. The requirement to prevent melting is met by specifying that four days elapse between use of the fuel element in a core l
! operating above 100 kW and removal of the element from the core tank. Calculations show that radiant heat transfer is adequate to remove the decay heat thereafter [5.4-3].
References 5.4-1 S. Dinsmore Report, "MITR Criticality Safety Evaluation of Storage Locatio'ns for 510 Gram Fuel Elements," MITR SR# 0-8-13,11 August 1980.
5.4-2 Fission Converter Safety Evaluation Report.
5.4-3 File Memo (Decay Heat Removal Using Radiant Heat Transfer) i j
U: l l .'5-9
.I. _
y Il 6. .
EXPERIMENTS Q
6.1 General Exneriment Criteria j Aonlicability This specification applies to experiments that use the reactor.
Obiective To assure that experiments that use the reactor do not affect the safety of the reactor.
Snecification All experiments that are within the reactor or its surrounding structure shall confonn to the following conditions:
I A
Q l. Reactivity Effs.q.ls The reactivity worth of experiments shall not exceed the values indicated in the following table: .
Sinnie Exneriment Total Worth Worth Movable 0.2% AK/K 0.5% AK/K Non-secured 0.5% AK/K 1.0% AK/K Total of the above N/A 1.5% AK/K Secured 1.8% AK/K N/A i
- 2. Thermal-Hydraulic Effects a) All experimental capsules shall be designed against failure from internal and external heating at the reactor power level or relevant process bd variable that corresponds to the Limiting Safety System Setting associated with that power level or process variable. ;
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b)- ' The outside surface temperature of a submerged experiment or capsule shall not cause nucleate boiling of the reactor coolant during operation of the reactor,
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c) - The insenion of an experiment into the core shall not cause a coolant flow redistribution that could negate the safety considerations implicit in the limiting safety system settings.
l
- 3. Chemical Effects h a) Metastable or other materials that could react to create a rapid pressure rise shall be encapsulated. The capsule shall be prototype-tested under experimental conditions to demonstrate that it can contam without failure an energy release equivalent to at least twice the material to be irradiated '
or at least twice the pressure that could be expected from any reaction of these materials. These tests must also include effects of any fragments which may be generated. If a change in experimental conditions could -
result in a greater potential'for failure than design experimental' conditions, the capsule shall also be tested under these changed conditions. In addition, the quantity of material should be limited such that if the maximum calculated energy release should occur, significant damage to the reactor core will not result, assuming that the material is -
not encapsulated.
b) No explosive materials (defined to include all materials that would constitute Class A, Class B, and Class C explosives as described in Title 49, Pans 172 and 173 of the Code of Federal Regulations) shall be placed in the reactor core or within the primary biological shield, which, if completely detonated, could cause any rearrangement or damage to the reactor core. Proposed quantities of ex?l osive materials greater than the equivalent of 25 milligrams of TNT shall require a documented safety analysis and approval by the MIT -Reactor Safeguards Committee. Capsule designs for explosive materials shall' be prototype-tested to demonstrate that they can contain at least twice the pressure produced inside the capsule as a result of detonation of the material or the pressure produced by the detonation of twice the amount ofmaterial, c) Corrosive materials that could affect or react with another material present in the reactor system shall be doubly encapsulated. If the material can adversely affect the reactor core or any of its component pans or auxiliary systems or the building containment to cause loss of function of the affected component or system, means shall be provided to monitor the integrity of the material container.- .;
~ 4. Radiolytic Decomoosition a) . Compounds subject to radiolyti; decomposition shall be irradiated in -
p '
containers that can withstand the maximum gas pressure produced as .
d result of the deccmposition under irradiation including the effect of any l x
6-2 e,
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temperature rise. This pressure shall be determined by previous
,-O. p experience or by testing as described in Specification 6.1.6.
' b) Consideration shall also be given to any pressure buildup resulting from :
the decomposition of the sample container, such as might occur with a polyethylene vial. ,
l Compounds subject to radiolytic decomposition rnay be irradiated in a {
c) capsule that is vented, provided that the vented release is less than 10.0% - !
of the limits of 10 CFR 20 at any point of possible exposure.
- 5. Exocriment Scrams Experiment scrams may be added for the protection of the experimental equipment and/or reactor components in the event of some malfunction. If malfunction of the experiment can adversely affect the reactor core or any ofits component parts or auxiliary systems or the building containment to cause loss of function of the affected component or system, the experiment scrams shall be redundant.
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- 6. Prototvoe Testing Materials whose properties (composition, heating, radiolytic decomposition, etc.) are uncertain shall be prototype tested. These tests will be designed to give a stepwise approach to final operating conditions. The tests may either be stepwise time or flux irradiations with proper instrumentation to determine temperature, pmssure, and radioactivity for each step as required. -)
i i
- 7. Radioactive Releases a) Experiments shall be designed so that operation or malfunction is predicted not to result in exposures or releases of radioactivity in excess l of the limits of 10 CFR 20 either to onsite or offsite personnel.
b) The total radioactive materials inventory of an experiment or credibly ;
. coupled experiments shall be limited such that the dose in unrestricted f
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areas resulting from release of this inventory at its calculated maximum
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C value shall not exceed the limits of 10 CFR 20.
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1 finns Accidents resulting from the insertion of reactivity are discussed in Section 13.2.2 of ,
J the SAR. The 0.2% AK/K limit for movable experiments corresponds to a 25-second period, one i
which can be easily controlled by the reactor operator with little effect on reactor power. The l' limiting value for a single non-secured experiment,0.5% AK/K is set conservatively below the prompt critical value for reactivity insertion and below the minimum shutdown margin. The sum of the magnitudes of the static reactivity worths of all non-secured experiments,1.0% AK/K, does not exceed the minimum shutdown margin. The total worth of all movatS and non-secured experiments will not reduce the minimum shutdown margin as the shutdown margin is determined with all movable experiments in their most positive reactive states. Finally,it was determined that,
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Q for operation with forced convection, a step increase of 1.8% AK/K would result in fuel plate {
temperatures that are below the clad melting temperature and significant core damage would not result. ;
Specifications 6.1.2 - 6.1.6 are intended to minimize the probability of experiment i
failure. Experiment capsules should be designed to withstand expected temperatures, pressures, chemical and radiochemical effects. The requirement for testing containers at twice the pressure or with twice the amount of explosive or metastable material to be irradiated provides a factor of two safety margin as allowance for experimental uncertainties. Table 6.1-1 gives a summary of the requirements for specimen irradiations for ease of review and classification of the specifications. l j
The radiological consequences of experiment malfunctions must be considered as stated i
in Specification 6.1.7. Consistent with the Commission's regulations, predicted onsite personnel exposures or offsite concentrations resulting from these malfunctions must not be in excess of
,s those permitted by 10 CFR Part 20.
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y 6.4 . Clo -Loop Control Systems l
Anplicability This specification applies to systems for the closed-loop control of the reactor exclusive -
- of those automatic controllers covered by Specification 3.2.2. (Noic: The reactivity' restrictions contained in ' Specification 6.1.1 do not apply 'to experiments -performed under this
' Specification 6.4.)
Obiective l q 3
To assure that the reactor can be safely shutdown at any time.' >
b Specification
- 1. Shim blades and/or the regulating rod may be connected to a closed-loop controller provided that the overall controller is designed so that the control of reactor power will always be feasible at either the desired termination point of p-any transient or at the maximum allowed operating power. Only one shim blade -
shall be withdrawn at a time.
- 2. Each proposed closed-loop controller shall require a documented safety analysis and approval by the MIT Reactor. Safeguards Committee (MITRSC) or, if-authorized by the MITRSC, by its Standing Subcommittee.
- 3. The nuclear safety system shall be separate from any closed-loop controller.
- 4. A period trip set at or longer than 20 seconds shall be operable whenever any .
closed-loop controller is in use. This trip shall transfer control to manual and 1 sound an alarm.
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- 5. The operability of the period trip shall be tested prior to use of any closed-loop controller during any week that a closed-loop controller is to be used.
Epcility specific Definitions
- 1. A reactor together with a specified control device is defined as constituting a system that is " feasible to control" ifit is possible to transfer the system from a given power level and rate of change of power to a desired, steady-state power level without overshoot, or conversely, undershoot. (Mole: If a deviation band is specified about the desired power level, then the term "without overshoot" means that there will be no overshoot beyond the permitted deviation.)
- 2. The word " separate" means that the output of an instrument used in the safety n system will not be influenced by interaction with the control system. For (v) example, a signal derived from an instrument that forms pan of the safety system would not be transmitted to the con.;ol system unless first passed through an isolation device.
i Basis The basis of the specification is given in Section 10.3.2.8 of the SAR. Digital control of the MITR is permitted subject to either of two approaches. The first is to place a lirnitation on the reactivity worth of the control device that is associated with the controller. This is the traditional approach that is commonly associated with analog controllers. It is covered by Specification 3.2.2. The second approach is to design the controller so that it incorporates the concept of feasibility of control. The seguirement that a closed-loop controller be designed so that control will always be feasible at the desired termination point of any transient ensures that there will be no power overshoots, or conversely, undershoots other than those permitted by specified 6-7
(m deviation bands, if any. The basis for the approach is that the reactor period can be made rapidly L.
infinite if the total reactivity, both that added directly by the control devices and that present indirectly from feedback effects, is maintained less than the maximum available rate of change of reactivity divided by the effective, multi-group decay parameter. Physically, if the reactivity is so constrained, then, by reversal of the direct:on of motion of the specified control device, it will be possible to negate the effect of the reactivity present and make the period infinite at any time during the transient. This condition, the absolute reactivity constraint,is unnecessarily restrictive. A less stringent constraint may be written that specifies that there be sufficient time available to eliminate whatever reactivity is present beyond the amount that can be immediately negated by reversal of direction of the designated control mechanism before the desired power level is attained. This condition is the sufficient reactivity constraint. This constraint's function is to review the decision of whatever control law is being used and, if necessary, override that decision. Provided that the net reactivity is always restricted to that permitted by the sufficient constraint,it should always be n
possible to halt a power increase before the desired termination point is attained by merely (v) reversing the direction of travel of the control device. Therefore, adherence to this constraint l means that no automatic control action should ever result in a challenge to the nuclear safety system. Additional information is given in References 6.4-1 and 6.4-2.
The requirement that each proposed closed-loop controller require a documented safety analysis and approval by the MITRSC or, if authorized by the MITRSC, by its Standing i
Subcommittee, ensures that the design of each controller will be carefully reviewed and that !
necessary off-line testing will be performed.
The requirement that the nuclear safety system be separate from that of any closed-loop controller means that the capability of that safety system to perform its intended function will not be compromised. ,
The existence of a trip that will transfer control to manual should the period become equal to or shorter than 20 seconds will provide a safety factor set more conservatively than the g)
C nuclear safety system. The signal used for this trip is separate from the nuclear safety system and i
6-8
i is not processed by the closed-loop controller. This assures that the capability of the trip signal to }
9 perform its intended function will not be compromised.
l 1
References i
- 1. Bernard, J.A., Henry, A.F., and D.D. Lanning, " Application of the ' Reactivity Constraint Approach' to Automatic Reactor Control," Nuclear Science and Engineering, Vol. 98, No. 2, Feb.1988, pp 87-95.
Bernard, J.A. and D.D. Lanning, " Considerations in the Design and Implementation of j 2.
l Control Laws for the Digital Operation of Research Reactors," Nuclear Science and Engineering, Vol.110, No. 4, Apr.1992, pp 425-444.
6-9
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i 6.5' Generation of Medical Therapy Facility Beams for Human Theraov Applicability This specification applies solely to the generation of medical therapy facility beams for the treatment of human patients. It does not apply to any other use of the medical therapy facilities and/or their beams. Surveillances listed in this specification are only required if human therapy is planned for the interval of the surveillance. However,in the event of a hiatus in the scheduled !
l' performance of any given surveillance, that surveillance shall be performed prior to the initiation of human therapy during the interval in question.
Obiective To provide for the protection of the public health and safety by ensuring that patients are treated in accordance with the treatment plan established by the BNCT physician authorized I user and that the ALARA principle is observed for all non-therapeutic radiation exposures.
Soecification l
- 1. Patients accepted for treatment shall have been referred by written directive from i
a BNCT physician authorized user from a medical center with an NRC or Agreement State medical use license that contains BNCT specific conditions and a
commitments for BNCT treatment on humans conducted at the Massachusetts l Institute of Technology Research Reactor's Medical Therapy Facilities. :
1
- 2. All medical treatments, including irradiations and analyses of the neutron -
capture agents in the patients, are the responsibility of the BNCT physician :
authorized user in charge of the therapy and the medical physicists from the p
d NRC-licensed or Agreement State-licensed medical center. The Massachusetts 6-10
Institute of Technology is only responsible for providing current and accurate beam characteristic parameters to the medical use licensee and for delivery of the desired radiation fluence as requested in the written directive. Before the start of a thcrapy, both the certified medical physicist and the Director of the Nuclear Reactor Laboratory, or his designate, must agree that the therapy can be initiated. The BNCT physician authorized user is responsible for monitoring the therapy and for directing its termination. However, a radiation therapy can also be terminated at any time if either the BNCT physician authorized user or the NRL Director, or their designates, judge that the therapy should be temiinated.
- 3. It shall be possible to initiate a minor scram of the reactor from a control panel located in each medical therapy facility area.
- 4. Access to each medical therapy facility shall be controlled by means of the shield doorlocated at its entrance.
- 5. The following features and/or interlocks shall be operable:
(a) An interlock shall prevent opening of the shutters that control beam delivery unless the medical therapy facility's shield door is closed.
(b) The shutters that control beam delivery shall be interlocked to close automatically upon opening of the medical therapy facility's shield door.
(c) The shutters that control beam delivery shall be designed to close automatically,upon failure of either electric power or on low air pressure if the shutter is operated pneumatically. Alternatively, on loss of either electric or pneumatic power an alarm could sound which alerts the reactor operator to the need to lower power when shutter closure is needed.
(d) Shutters that control beam delivery and that are normally pneumatically-
- operated shall, in addition, be designed for atanual closure.
6-11 l 1
i
F (c) : It shall be possible to close the shutters that control beam delivery from .
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within the medical therapy facility.
- 6. Each of the shutters that controls beam delivery shall be equipped with a light '
that indicates the status of the shutter. These lights shall be visible ~at each -
medical therapy facility's local control panel. In the event of a status light .
malfunction, it shall be acceptable 'to use the affected shutter provided that an -
alternate means of verifying position is available. Use of this alternate means of shutter position verification is limited to seven consecutive working days.
- 7. Each medical therapy facility shall be equipped with a monitor that provides a visual indication of the radiation level within the facility, that indicates both within the facility and at the local control panel, and that provides an audible alarm both within the facility and at the local control panel.
A V (a) This radiation monitor shall be _ equipped with a backup power supply such as the reactor emergency power system or a battery.
(b) This radiation monitor shall be checked for proper operation by means of a check source on the calendar day of and pnor to any patient irradiation.
(c) This radiation monitor shall be calibrated quarterly.
i The audible alarm shall be set at or below 50 mrem /hr. This monitor '!
(d) and/or its alarm may be disabled once the medical therapy room has . -l been searched and secured,'such as is done immediately prior to !
i initiation of patient therapy. If this is done, the monitor and/or its alarm shall be interlocked so that they become functional upon opening of the medical therapy facility's shield door.
In the event that this monitor is inoperable, personnel entering the- !
(c) medical therapy facility shall use either portable survey instruments or audible alarm personal dosimeters as a temporary means of satisfying- q this provision. These instruments / dosimeters shall be in calibration as - - !
definedby the MIT Research Reactor's radiation protection ?rogram and shall be source-checked daily prior to use on any day that ticy are used to satisfy this provision. Use of these' instruments / dosimeters as a temporary means of satisfying this provision'is limited to seven ;
consecutive working days.
6-12
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- 8. An intercom or other means of two-way communication shall be operable both
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V between each medical therapy facility control panel and the reactor control room, and also between each medical therapy facility control panel and the interior of the facility. The latter is for the monitoring of patients.
- 9. It shall be possible for personnel monitoring a patient to open ech medical therapy facility's shield door manually.
- 10. It shall be possible to observe the patient through both a viewing port and by means of a closed-circuit TV camera. Both methods of patient visualization shall be operable at the outset of any patient irradiation. Should either fait during the irradiation, the treatment may be continued at the discretion of the BNCT physician authorized user. Adequate lighting to permit such viewing shall be assured by the provision of emergency lighting.
I 1. The total radiation fluence delivered by the medical therapy facility beam as measured by on-line beam monitors shall not exceed that prescribed in the patient treatment plan by more than 20%. The treatment is normally delivered in fractions in accordance with standard practice for human therapy. The 20%
criterion applies to the sum of the radiation fluences associated with all fractions in a given treatment plan. A criterion of 30% applies to the difference between the administered and prescribed fluence for any given week (seven consecutive days). Finally, if the treatment consists of three or fewer fractions, then a criterion of 10% shall apply.
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V 6-13
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- 12. The following interlocks or channels shall be tested at least monthly and prior to treatment'of human' patients if the interlock or channel has been repaired or . .
]1 deenergized:
.i Interic,ck or Cha l! Surveillance a) Medical therapy facility minor scram Scram test b) Skers will not open unless . Operational test shield dooris closed '
c) Shutters close upon both manual and ' Operational test :
automatic opening of shield door d) Shutters close and/or alarm on loss Operational test of electrical power and reduction of 3 pressure in pneumatic operators, ifapplicable e) Manual closure of pneumatic shutters Operational test '
f) Shutters can be closed manually Operationaltest i from within the facility O
(V g) Shutter status lights Operational test h) Radiation monitor alarm Operational test i) Radiation monitor and/or alarm Operational test ,
enabled upon opening of shield door j) Intercoms Operational test 4
In addition to the above, each medical therapy facility minor scram shall be ;
tested prior to reactor startup if the reactor has been shut down for more than siyteen hours.
- 13. Manual operation of each medical therapy facility's shield door in which the .
- door is opened fully shall be verified semi-annually, jN) t 6-14 '
.u ,
- 14. Use of the medical therapy facility beams shall be subject to the following:
a)- ' A functional check'of the beam monitors that are described in provision L 11 of this specification shall be made weekly for any week that the beam
'will be used for human therapy. This check shall be made prior to any patient irradiation for a given week. In addition, a functional check ~ shall q x performed prior to any patient irradiation in the event of a component '
replacement or a design modification. :
b) A calibration check of the beam shall be performed every six months for any six-month interval that the beam will be used for human therapy.
This six-month calibration check shall be made prior to any patient 1 irradiation for a p.ven six-month interval. In addition, a calibration check shall be puformed prior to any patient irradiation in the event of a :
component rep acement or a design modification.
c) A characterization of the beam shall be performed every twelve months '
for any twe'.ve-month interval that the beam will be used for human therapy. This twelve-month characterization shall be made prior to any ,
patient irradiation for a given twelve-month interval. A characterization' shall also beperformed prior to any patient irradiation in the event of a-design modification. As part of the characterization process, the proper response of the beam monitors that are described in provision 11'of this specification shall be established.
d) The instruments (e.g... tissue-equivalent chamber or graphite-r magnesium wall ionization chamber or the equivalent) that are to be used to perform both calibration checks and characterization of the beam shall be calibrated by a secondary calibration laboratory. This calibration shall be xrformed at least once every two years for any two-year interval t iat the beam will be used for human therapy. The two-year calibration shall be made prior to any patient irradiation during any given two-yearinterval. (Noic: If a method (e.g., foil activation) other than these checks is used for the calibration and or the characterization, then the devices (e.g., foils) used in that method shall either be traceable to the National Institute of Standards and Technology or be selected in accordance with the relevant ANSI /ANS standards, e) There shall be a minimum of two neutron-sensitive beam monitors to:
~
initiate a patient irradiation. Once initiated, a patient irradiation may be-continued at the discretion of both the certified medical physicist and the Director of the Nuclear Reactor Laboratory, or his designate, provided that at least one neutron-sensitive beam monitor is operable.
- 15. Maintenance, repair, and modification of the medical therapy facilities shall be performed under the supervision of a senior reactor operator who is licensed by the U.S. Nuclear Regulatory Commission to operate the MIT Research Reactor.
. The ' medical. therapy facility' includes the beam, beam shutters, beam -
.6-15 g
(~T monitoring equipment, medical therapy facility shielding, shield door, and V
patient viewing equipment. All modifications will be reviewed pursuant to the requirements of 10 CFR 50.59. The operating couch, patient positioning equipment, medical instruments, and other equipment used for the direct medical support of the patient are not considered part of the medical therapy facility for purposes of this provision, except insofar as radiation safety (i.e.,
activation and/or contamination) is concemed.
- 16. Personnel who are not licensed to operate the MIT Research Reactor but who are responsible for either the medical therapy or the beam's design including construction and/or modification may operate the controls for the corresponding medical therapy facility beam provided that:
(a) Training has been provided and proficiency satisfactorily demonstrated n on the design of the facility,its controls, and the use of those controls.
t j Proficiency shall be demonstrated annually. j v j (b) Instructions are posted at the medical therapy facility's local control panel that specify the procedure to be followed: ,
I (i) to ensure that only the patient is in the treatment room before !
turning the primary beam of radiation on to begin a treatment; I (ii) if the operator is unable to turn the primary beam of radiation off with controls outside the medical therapy facility, or if any other abnormal condition occurs. A directive shall be included with these instructions to notify the reactor console operator in the event of any abnormality.
(c) In the event that a shutter affects reactivity (e.g., the D2 0 shutter for the ;
medical room below the reactor and the convener control shutter for the i fission converter beam), personnel who are not licensed on the MIT Research Reactor but who have been trained under this provision may operate that shutter provided that verbal permission is requested and received from the reactor console operator immediately prior to such action. Emergency closures are an exception and may be made without first requesting pernussion.
Records of the training provided under subparagraph (a) above shall be retained f in accordance with the MIT Research Reactor's training program or at least for '
6-16
,r'~T three years. A list of personnel so qualified shall be maintained in the reactor Nj' control room.
{
- 17. Events defined as ' recordable' under definition 8 of this specification shall be recorded and the record maintained for five years. Events defined as
'misadministrations' under definition 9 of this specification shall be reported to the U.S. Nuclear Regulatory Commission (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> verbal,15 day written report). The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> verbal reports will be made to the Regional Administrator, Region I, or his designate. The 15 day written reports will be sent to the NRC Document Control Desk with a copy to the Regional Administrator, Region I, or his designate.
- 18. The requirements of the Quality Management Program (QMP) for the I Generation of Medical Therapy Facility Beams for Human Therapy at the Massachusetts Institute of Technology Research Reactor shall be observed for any human therapy. (Note: The presence of this commitment to observe the l
QMP in these specifications does not preclude modifying the QMP as provided in that document. Any such modifications are not considered to be a change to the MITR Technical Specifications.)
- 19. Reactor facilities (e.g., prompt gamma for the determination of boron concentration in blood or tissue) that are used to perform measurements associated with the conduct of medical therapy shall be calibrated every twelve months for any twelve-month interval that the beam will be used for human therapy. This twelve-month calibration shall be made prior to any patient irradiation for a given twelve-month interval. This calibration could be done by n
() measuring a series of standards that span the anticipated range of boron in blood 6-17
, I
^
. or tissue.' In addition,'a single point check, (e.g., verification that a single.
standard is measured i10% of its true value) shall be performed prior to any .
- patient irradiation.
O
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.)
6-18 i, .
1
f i
I
! f~'si Definitions i
a
- 1. The medical therapy facilities are equipped with shutters that are used (i) to control beam delivery and (ii) to adjust the neutron energy spectrum of the l
! beam. The former currently include lead, boral, and light water shutters as described in Reference 6.5-1. The heavy water blister tank, which is also described in Reference 6.5-1,is an example of the latter. It is conceivable that these designations may change should it be found desirable to alter the beam configuration. Accordingly, the phrase " shutters that control beam delivery" refers either to the aforementioned three existing shutters or to any future shutter or group thereof that provides an equivalent or greater reduction in beam intensity. Shutter-effect analyses shall be documented through the standard safety review process including, where appropriate, an SAR revision and submission to NRC under 10 CFR 50.59.
- 2. The term ' calibration check' refers to the process of checking the beam intensity and quality via one or more of the following: foil activation; use of a fission chamber; use of an ion chamber; or an equivalent process. The purpose of a calibration check is to ensure that the beam has not changed in a significant way (e.g., energy spectrum or intensity) from the beam that was characterized.
- 3. The term ' functional check of the beam monitors'shall consist of verifying that system output is consistent (i 10%) with previously measured values upon normalization to a common reactor neutronic power level.
- 4. The term ' characterization' refers to the process of obtaining the dose-versus-
~ depth profile in phantoms as described in Reference 6.5-2 or an equivalent (y)
/
process. The dose-versus-depth profile from the surface of the phantom to a 6-19 4
i
)
depth at least equivalent to the total thickness of the body part to be treated on a r]
L/
central axis is deemed adequate for a characterization. Fast neutron, thermal neutron, and gamma ray components are determined in a characterization and _
monitors are normalized by this characterization.
- 5. The term ' component replacement' means the replacement of a component in the beam with an identical unit or the re-installation of a component in the beam for
(
which a characterization has already been performed. For example, the latter l may be a change of collimators.
- 6. The term ' design modification' as applied to a medical therapy facility beam refers (a) to a change that is shown to alter the dose-versus-depth profile of the !
fast neutrons, thermal neutrons, or gamma rays in the beam as sensed by the O
v calibration check and (b) to a change that has the potential to increase significantly the amount of activation products in the medical therapy facility when the beam is to be used for the treatment of human patients.
i l
- 7. The term ' radiation fluence' means the total fluence of neutrons and gamma radiation that is emitted in a medical therapy facility beam. The determination of the ratios of gamma, fast neutron, and thermal neutron fluences is pan of the beam characterization. Knowledge of these ratios allows the total radiation fluence to be monitored by the on-line detectors, which are neutron-sensitive.
Compliance with the limits specified on radiation fluence by this specification is determined by reference to the fluence monitored by these detectors.
- 8. The term ' recordable event' means the administration of:
- r m
() (a) A radiation treatment without a written directive; or 6-20
fm (b) A radiation treatment where a written directive is required without V reporting to the medical use licensee in writing each fluence given within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the treatment; or (c) A treatment delivery for which the administered radiation fluence for any given fraction is 15% greater than prescribed.
- 9. The term ' misadministration' means the administration of a radiation therapy:
(a) Involving the wrong patient, wrong mode of treatment, or wrong treatment site: or (b) When the treatment delivery is not in accordance with provision 11 of this specification.
- 10. The term ' written directive' means an order in writing for a specific patient, l dated and signed by a BNCT physician authorized user prior to the administration of radiation and which specifies the treatment site, the total l v(")
l radiation fluence, radiation fluence per fraction, and overall treatment period.
L 11. The term ' human therapy' means radiation treatments that are of direct therapeutic benefit to the patient and/or part of investigatory studies that involve humans.
- 12. The term 'BNCT physician authorized user' means a medical physician authorized by the medical use licensee's radiation safety committee to act as an authorized user for BNCT on humans.
l
- 13. The term ' certified medical physicist' means a medical physicist certified in either radiological physics or therapeutic radiation physics by the American Board of Radiology, or in therapeutic radiation physics by the American Board
/m i \_,
6-21
l l
p of Medical Physics and who also has specific training in neutron dosimetry and 1\ 'i
' ' ~
neutron beam capture therapy.
Basis The stipulation that patients only be accepted from a medical use licensee that has an NRC or an Agreement State medical use license that contains BNCT specific conditions and commitments for BNCT treatment of humans conducted at the Massachusetts Institute of Technology Research Reactor's Medical Therapy Facilities ensures that medical criteria imposed by NRC or the Agreement State on such licensees for the use of the MIT Research Reactor's medical therapy facility beams for human therapy will be fulfilled. The second provision delineates the division of responsibilities between the Massachusetts Institute of Technology and the medical licensee that refers the patient. Also, it establishes administrative authority and protocol for initiating and terminating a radiation therapy.
l The requirement that it be possible to initiate a minor scram from control panels located in the medical therapy facility areas assures the attending physician and/or medical physicist of the capability to terminate the treatment immediately should the need arise. The provision that access to each medical therapy facility be limited to a single door ensures that there will be no inadvertent entries. The various interlocks for the shutters that control beam delivery ensure that exposure levels in the medical therapy facility will be minimal prior to entry by personnel who are attending the patient. The shutter-indication lights serve to notify personnel of the beam's status. The provision for a radiation monitor ensures that personnel will have information available on radiation levels in the medical therapy facility prior to entry. The purpose of this monitor's audible alarm is to alert personnel to the presence of elevated radiation levels, such as exist when the l shutters that control beam delivery are open. This monitor and/or its alarm may be disabled once the medical therapy facility has been searched and secured so that it will (1) not disturb a patient l and (2) not distract attending personnel. The monitor and/or its alarm are interlocked with the
() shield door so that they are made functional upon opening that door, and hence prior to any l 1
6-22
i (3 possible entry to the medical therapy facility. One intercom provides a means for the prompt k) exchange of information between medical personnel and the reactor operator (s). The second !
intercom is for monitoring the patient.
The provision for manual operation of each medical therapy facility's shield door ensures l access to any patient in the event of a loss of electrical power. The presence of a viewing window and a closed-circuit TV camera provide the attending BNCT physician authorized user and/or !
medical physicist with the opportunity to monitor the patient visually as well as through the use of various instruments. The viewing window will function even during an electric power failure because of the provision for emergency lighting.
The specification that the total radiation fluence for a therapy (i.e., the radiation fluences for the sum of all fractions specified in a given treatment plan) not exceed that prescribed in the patient I treatment plan by 20% establishes a trigger limit on the delivered fluence above which NRC has to be notified of a misadministration. The 20% criterion is based on the definition of i G misadministration (clause 4(iv)) as given in 10 CFR 35.2. The criterion that the difference between the administered and prescribed fluence for any seven consecutive days is set at 30%.
This is also in accordance with the definition of misadministration (clause 4(iii)) as given in 10 CFR 35.2. Finally,if a treatment involves three or fewer fractions, then a more stringent criterion, 10%, applies to the difference between the total radiation fluence for a therapy and that prescribed in the treatment plan (10 CFR 35.2(4ii)). The surveillance requirements for the functional checks as well as those for the beam calibration checks and characterizations provide a mechanism for ensuring that each medical therapy facility and its beam will perform as originally designed.
Similarly, the surveillance requirements on the instruments used to perform these checks and characterizations ensure that these instruments are calibrated by a means traceable to the National Institute of Standards and Technology. The chambers specified (tissue-equivalent, and graphite or magnesium-wall) were chosen because they measure dose as opposed to fluence. Finally, the requirement on the number of beam monitors is in keeping with standard practice for gamma-ray V sources.
6-23
q The specification on maintenance and repair of the medical therapy facilities ensures that all ~
such activities are performed under the supervision of personnel cognizant of quality assurance and -
other requirements such as radiation safety. The provision on the training and proficiency of non-
~
licensed personnel ensures that all such personnel will receive instruction equivalent to that given to
. licensed reactor operators'as regards use of the medical therapy facility beams. (Note: Licensed reactor operators may, of course, operate the medical therapy facility beams.) Also, this provision provides for the posting of instructions to be followed in the event of an abnormality.
The specification on ' recordable events' and 'misadministrations' provides for the documentation and reporting to the U.S. Nuclear Regulatory Commission of improper events -
z regarding the generation and use of medical therapy facility beams. The requirement that the Quality Management Program (QMP) be observed ensures that radiation treatments provided by a medical therapy facility beam will be administered as directed by the BNCT physician authorized user.
The specification on calibration of reactor facilities that are used to measure the
-j concentration of boron in blood _or tissue ensures that these measurements are accurate.
References i
6.5-1 MITR Staff, " Safety Analysis Report for the MIT Research Reactor (MITR-II)," Report j No. MITNE-115,22 Oct.1970, Section 10.1.3, l 6.5-2 Choi, R.J., " Development and Characterization of an Epithermal Beam for Boron Neutron Capture Therapy at the MITR-II Research Reactor," Ph.D. Thesis, Nuclear Engineering l' Department, Massachusetts Institute of Technology, April 1991.
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Ouality Management Program -
for Generation of MITR-III Medical Therapy Facility Beams for Human Therapy 6-25
- Oiinlity Manngement Program
\
Purpose:
The objective of this quality management program is to ensure that radiation-1.
treatments ?rovided by the MIT Research Reactor's (MITR-II) Medical Therapy Facility
~ beams will sadministered as directed by a BNCT physician authorized user.
- 2. Authodzed Medical Use Licensees: Use of the MIT Research Reactor's Medical Therapy Facility beams, for the treatment of human subjects, is limited to the BNCT physician -
authonzed users from medical centers with an NRC or Agreement State medical use license L that contains BNCT specific conditions and commitments for BNCF treatment on humans ;{
conducted at the Massachusetts Institute of Technology Research Reactot's Medical Therapy Facilities.
J
- 3. Program Reauirements: The following requirements are established as part of this quality management program-(a) A written directive will, except as noted in subparagraph (it) below, be prepared by 1 a BNCT physician authorized user of either the NRC or Agreement State medical use licensee prior to the administration of any radiation therapy. This directive shall be written, signed, and dated by the BNCT physician authorized user and it shall include the following information:
(i) Name and other means ofidentifying the patient.
(ii) Name of the BNCT physician authorized user and certified medical physicist in charge of the therapy. ;
.l (iii) The total radiation fluence to be administered, the radiation fluence per d fraction, the treatment site, and the overall treatment period.
(iv) If, because of the patient's condition, a delay in order to provide a written revision to an existing written directive would jeopardize the patient's ;
health, an oral revision to an existing written directive will be acceptable, provided the oral revision is documented immediately in the patient's record and a revised written directive is signed by a BNCT physician authorized user within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of the oral revision.
Also, a written revision to an existing written directive may be made for any ;
therapeutic procedure provided that the revision is dated and signed by a ;
BNCT physician authorized user prior to the administration of the next fraction.
If, because of the emergency nature of the patient's condition, a delay in order to provide a written directive would jeopardize the patient's health, an oral directive will be acceptable, provided that the information contained in H the oral directive is documented immediately in the patient's record and a written directive is prepared within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the oral directive.
(v) In order to ensure that the Staff of the MIT Research Reactor has the most recent written directive from the medical use licensee and the correct .
directive for the patient in question, a copy of that directive shall be hand-9 delivered to the MITR Staff by the Staff of the medical use licensee who
! accompany the patient to MIT. This copy shall then be checked against the e 6-26 4
/7 most recent previous transmission. Any discrepancy shall be resolved by the medical use licensee prior to the initiation of patient irradiation.
()
(vi) The Director of the MIT Nuclear Reactor Laboratory, or his designate, will date and sign the written directive to verify that current and accurate beam characteristic parameters were provided to the NRC or Agreement State medical use licensee as appropriate and that the radiation fluence desired in the written directive was delivered. A copy of this signed directive shall be provided to the medical use licensee within twenty-four hours of a treatment.
(b) Prior to each administration of any radiation, the patient's identity will be verified by more than one method as the indiv; dual named in the written directive. The MIT Nuclear Reactor Laboratory will use any two or more of the following acceptable methods ofidentification:
(i) Self-identification by patients who are conscious upon arrival at the MIT Research Reactor. Information provided by the patient shall include any two of the following: name, addNss, date of birth, or social security number. The information provided by the patient is to be compared to the corresponding information in the patient's record.
(ii) Hospital wrist band identification with the wrist band information to be compared to the corresponding information in the patient's mcord.
(iii) Visual identification against photographs provided with the written
) directive.
.( G (iv) Other methods as specified in U.S. Nuclear Regulatory Commission ,
Regulatory Guide 8.33, " Quality Management Program." i (c) The plan of treatment is certified by the certified medical physicist to be in accordance with the written directive. In this regard, the Massachusetts Institute of i J
Technology is res},onsible for calibrating the output of the beam monitoring instrumentation versus dose in phantom and for providing a central axis dose j versus depth profile. This information will then be used by personnel at either the NRC or the Agreement State medical use licensee as appropriate to generate a plan l of treatment. Conformance of the beam to its design characteristics is confirmed through the measurements specified in MITR Technical Specification #6.5,
" Generation of Medical Therapy Facility Beams for Human Therapy." Functional checks are made of the beam monitors at least weekly for any week that the beam will be used for human therapy (provision 14(a)). Calibration checks are performed every six months for any six-month interval that the beam will be used for human therapy (provision 14(b)). Each beam is characterized dosimetrically every twelve months (provision 14(c)). The instruments that used to perform calibration checks and characterization of the beams are calibrated every two years by a secondary calibration laboratory (provision 14(d)).
(d) Each administration of radiation is in accordance with the written directive subject to the tolerances established in provision 11 of MITR Technical Specification #6.5,
" Generation of Medical Therapy Beams for Human Therapy."
}g)_
v (e) Any unintended deviations from the written directive shall be identified and evaluated, end appropriate action taken. Such action shall include informing the 6-27
("s medical use licensee of the deviation. These reviews shall be performed monthly for any month in which human therapy was conducted. For each patient case Q! reviewed, it shall be determined whether the administered total fluence, fluence per fraction, treatment site, and overall treatment period were as specified in the written directive. In the event of any deviation from the written directive, the licensee (MIT) shall identify its cause and the action required to prevent recurrence. These actions may include new or revised policies, new or revised procedures, additional training, increased supervisory review of work, or other measures as deemed appropriate. Corrective actions shall be implemented as soon as practicable.
- 4. Program Implementation: The following practices shall be observed in order to ensure )
i proper implementation of the quality management program:
l (a) A review shall be conducted of the quality management program. This review shall l include, since the last review, an evaluation of: j (i) A representative sample of patient administrations, I
(ii) All recordable events, and (iii) All misadministrations. l The objective of this review is to verify compliance with all aspects of the quality management program. For purposes of this review, the term ' representative' in i statement (i) above is defined as 100% sampling up to twenty patients; a sample of l twenty for twenty-one to one hundred patients, and 20% sampling for more than l one hundred patients. In order to eliminate any bias in the sample, the patient cases lk ,J) to be reviewed should be selected randomly.
(b) The procedure for conducting the above review is as follows: !
(i) The review shall be aerformed by the Director of the MIT Radiation Protection Program or ais designate.
(ii) The review shall be performed annually.
(iii) Patient administrations selected for review shall be audited to determine ,
compliance with each of the requirements listed in paragraph (3) above.
(iv) The review shall be written and any items that require further action shall be I so designated. Copies of the review shall be provided to the NRL Director and to the MIT Reactor Safeguards Committee who will evaluate each review and, if required, recommend modifications in this quality management program to meet the requirements of paragraph (3) above. A copy of these reviews will also be provided to each medical use licensee.
(c) Records of each review, including the evaluations and findings of the review, shall be irtained in an auditable form for three years.
(d) The licensee (MIT) shall reevaluate the Quality Management Program's policies and procedures after each annual review to determine whether the program is still
(~] effective or to identify actions required to make the program more effective.
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6-28
Q 5. Response to Recordable Event: Within thirty days after the discovery of a recordable V event, the event shall be evaluated and a response made that includes:
]
(a) Assembling the relevant facts, including the cause, (b) Identifying what, if any, corrective action is required to prevent recurrence; and (c) Retaining a record,in an auditable form, for three years, of the relevant facts and what corrective action,if any, was taken. ,
1 A copy of any recordable event shall be provided to the affected medical use licensee. j i
i
- 6. Records Retention: The following records shall be retained:
(a) Each written directive for three years; and (b) A record of each administered radiation therapy where a written directive is lequired in paragraph (3(a)) above, in an auditable form, for three years after the date of administration.
1 I
- 7. Program Modification: Modifications may be made to this quality management program to increase the program's efficiency provided that the program's effectiveness is not decreased. All medical use licensees shall be notified of any modifications and provided with a copy of the revised program. The licensee (MIT) shall furnish the modification to the NRC (Region I) within 30 days after the modification has been made.
- 8. Report and Surveillance Frecuenev: Any report or other function that is required to be ;
performed in this Quality Management Program at a specified frequency shall be performed i within the specified time interval with:
s (a) a maximum allowable extension not to exceed 25% of the specified surveillance interval, unless otherwise stated in this Quality Management Program; i
(b) a total maximum combined interval time for any three consecutive surveillance intervals not to exceed 3.25 times the specified surveillance interval.
- 9. Definitions:
e !
(a) The term 'BNCT physician authorized user' means a ' medical physician authorized i by the medical use licensee's radiation safety committee to act as an authorized user for BNCT on humans.
(b) The term ' certified medical physicist' means a medical physicist certified in either radiological physics or therapeutic radiation physics by the American Board of Radiology, or in therapeutic radiation physics by the American Board of Medical Physics and who also has specific training in neutron dosimetry and neutron beam capture therapy.
- 10. Applicability: This Quality Management Program applies solely to the generation of medical therapy facility beams for the treatment of human subjects. It does not apply to any other use of the medical therapy facilities and/or their beams. Reports and surveillances O. listed in this specification are only required if human therapy was conducted during the (j referenced interval.
l 6-29
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- 7. ADMINISTRATIVE CONTROLS
[g)
Apolicability Administrative controls are the means by which reactor operations are subject to I
management control. Measures specified in this section provide for the assignment of {
responsibilities, reactor orgau!zation, staffing qualifications and related requirements, review and j audit mechanisms, procedural controls and reporting requirements. Each of the measures are applicable as minimum requirements throughout reactor life.
Objective I To assure that adequate management controls are available for safe facility operation.
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7-7-1
7.1 Oreanization 7.1.1 Structure The organization for the management and operation of tne reactor facility is shown in Figure 7.1-1.
7.1.2 Responsibility
- 1. The Director of Reactor Operations is directly responsible for the safe operation of the facility.
- 2. In all matters pertaining to safe operation of the MIT Reactor (MITR) and to these Technical Specifications, the Director of Reactor Operations shall repon to and be directly responsible to the Director of the Nuclear Reactor Laboratory.
9 V
The management organization is shown in Figure 7.1-1.
- 3. The MIT Reactor Radiation Protection Officer shall be responsible for radiation protection at the MITR. He shall advise the Director of Reactor Operations on all matters pertaining to radiation protection.
- 4. The MIT Reactor Radiation Protection Officer shall report to and be directly responsible to the MIT Radiation Protection Officer.
- 5. In the event of disagreement between the recommendations of the MIT Reactor Radiation Protection Officer and the Director of Reactor Operations or their altemates, on matters pertaining to radiation protection, the course determined by the Director of Reactor Operations or his designated alternate to be the more O
7-2
f President MIT Vice-President . Vice-President for Research for Human Resources i
MIT Director .
Committee on .
Radiation Protection Nuclear Reactor Laboratory Officer T- T- Reactor Safeguards-l Level 1 - I l
I I l l i_ _ _J Director of Reactor Operations I Reactor Radiation --- Audit; Protection Officer T- Review l Level 2
.I 'I O '
I-I Superintendent of
-- Operations and Maintenance Level 3 Operating Staff Level 4 Reporting Lines
- - - - - --- - - - Communication Lines Figure 7.1-1: Management Organization '.
1
, I__'.- i_i-._
conservative shall be followed. Records of the disagreement shall be sent for (3
V review and possible reconsideration to the MIT Radiation Protection Officer, the Director of the Nuclear Reactor Laboratory, and the Chairman of the MIT Reactor Safeguards Committee.
- 6. The responsibilities of any given level in the management organization chart may be assumed by designated alternates conditional upon appropriate qualifications. Such delegation of authority shall be documented in writing by the regularly assigned individual or that person's immediate supervisor.
7.1.3 Staffine The m.nimum reactor staff organization shall be as follows:
- 1. When the reactor is not shut down, the minimum crew complement for a shift shall be two licensed operators including at least one licensed senior reactor
(
operator, one of whom shall be in the control room. In addition, the MITR Radiation Protection Officer or a designated alternate shall be onsite or on call.
If on call, one of the licensed operators will have responsibility for >
implementing radiation protection procedures.
- 2. Whenever the reactor is not secured, two persons shall be onsite, one of whom shall be a licensed senior reactor operator. An operator or senior operator shall be present in the control room. In addition, the MITR Radiation Protection Officer or a designated altemate shall be onsite or on call. If on call, the onsite licensed senior operator will have responsibility for implementing radiation protection procedules.
7-4
p 3. A list of reactor facility personnel by name and telephone number shall be i /
readily available in the control room. This list shall include management, radiation safety, and other operating personnel.
- 7.1.4 Selection of Personnel Minimum educational and/or experience requirements for those individuals who have line responsibility and/or authority for the safe operation of the facility are as follows:
- 1. Director of Reactor Operations - The Director of Reactor Operations shall have a minimum of seven years of nuclear experience. The individual shall have a recognized baccalaureate or higher degree in an engineering or scientific field. I Education or experience that is job-related may be substituted for a degree on a case-by-case basis. The degree may fulfill four years of the seven years of nuclear experience required on a one-for-one time basis. At least three years of
\ experience shall be in a responsible position in reactor operations or a related !
field including at least one year's experience in reactor facility management or l l
supervision. The Director of Reactor Operations shall hold a senior operator's license for the MIT Research Reactor, or have held such a license at the MIT Research Reactor.
- 2. Superintendent of Operations and Maintenanta - The Superintendent of Operations and Maintenance shall have a minimum of five years of responsible reactor experience. A maximum of two years of experience may be fulfilled by academic or related technical training on a one-for-one basis. The Superintendent shall hold a senior operator's license for the MIT Research Reactor.
LJ 7-5
1 l
i l
l Reactor Radiation Protection OfGeer - The Reactor Radiation Protection Officer j 77 3.
1 L) shall have a minimum of five years of experience in radiation protection {
including at least one year of experience at a nuclear reactor facility. A maximum of four years of the five years experience may be fulfilled by related I
technical or academic training.
l l
I
- 4. Shift Supervisor - Shift supervisors shall have a minimum of a high school diploma or equivalent and three years of responsible reactor experience. A maximum of 1-1/2 years experience may be fulfilled by academic or related l
technical training on a one-for-one basis. The requirement for experience may be completely satisfied by acceptance in a program leading to an advanced i J
degree in Nuclear Engineering. Shift supervisors shall hold an NRC senior operator license for the ; IT Research Reactor.
O l U l
- 5. Reactor Operator - Reactor operators shall have a high school diploma or equivalent. They shall hold an NRC reactor operator license for the MIT I i
Research Reactor.
7.1.5 Training of Personnel
- 1. A training program shall be established to maintain the overall proficiency of the Reactor Operations organization. This program shallinclude components for both initial certification and requalification.
- 2. The training program shall be under the direction of the Superintendent of l Operations and Maintenance and/or the Director of Reactor Operations.
<~s 7-6
.(
- 3. Records ofindividual plar.t staff members' qualifications, experience, training,-
and requalification shall be maintained as specified in Specification 7.8.2.
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.7-7--
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% 7.2 l : Review nad Audit -
> 1 7.2.1 MIT Committee on Reactor Safeguards'
. Overall direction 'on matters of reactor safety rests with the MIT Reactor -
l.
Safeguards Committee or MITRSC. Approval of the MITRSC is necessary fob
. all new operating plans and policies and all significant modifications thereto-which may involve questions of nuclear safety. The ,MITRSC is also-responsible for auditing operation of the reactor. The Chairman of the MITRSC -
reports directly to the President of MIT. The MITRSC communicates directly with the Director of the Nuclear Reactor Laboratory and with the Director of-Reactor Operations, both of whom are members of the MITRSC.
7.2.2 Comoosition ed Ounlifications
-1. The MITRSC shall be composed of a minimum of five persons with not more than one-third of the total membership chosen from the reactor staff organization and a minimum of three members from outside MIT. All members and the Chairman shall be selected by the Presiden: J MIT. At least four voting members including participating alternates shall have a minimum of a Bachelor's Degree in Engineering or the Physical Sciences and have a minimum of three years of professional level experience in nuclear services, nuclear plant operation, or nuclear engineering, and the necessary overall nuclear background ;
to determine when to contact consultants for analyses beyond the scope of the '
MITRSC's expertise. The MIT Radiation Protection Officer and the MIT
. Safety Officer shall be members of the MITRSC.
7'.2.3' . ' Charter and Rules :
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. . .1.. - Meetino Frequency: Meetings shall be held no less frequently than annually.
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- 7-8.
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O 2. Ouon e A quorum shall consist of no less than a majority of the Committee's voting membership. In addition, either the Chairman or a designated alternate b shall be present. Finally, no mom than a minority of the quorum shall have line responsibility.for reactor operations.
~
- 3. Subcommittee: The full MITRSC may, after discussion of a' particular topic,
~ designate a subcommittee to' conduct further review. The MITRSC may also choose to delegate approval authority to a subcommittee. The membership of a subcommittee, its chairman, its purpose, and its authority must be stathd in the minutes of a meeting of the full MITRSC. These items must be reaffirmed at least annually. The quorum rule that applies to the membership of ths MITRSC as a whole also applies to the membership of any subcommittee.
O
- 4. Minutes: Minutes am kept of all MITRSC (and subcommittee) meetings. These '
are distributed to all members of the Committee, to the MIT Administration, and to the Reactor Operations /RRPO Staff.
-l
- 5. Access to Reactor Records: The MITRSC has complete access to all records.
pertaining to reactor operation. .j i
7.2.4 Review Function 1
- 1. The MITRSC's review function shall include the following:
'E a) : Determinations th'at proposed changes in' equipment, systems, tests, ;
experiments, or procedures.do not: involve an unreviewed safety.
question. (10 CFR 50.59).
b)' _ DAll new procedures -and major revisions thereto .having' safety-1 significance, proposed changes in reacter facility equipment, or systems -
W[
( ,
having safety significance.
~
L 7-9=
F $
(~ - c) All new experiments or classes of experiments that could affect X_]j reactivity or result in the release of radioactivity.
d) Proposed changes in technical specifications or license.
e) Violations of technical specifications or license as well as violations of internal procedures or instructions having safety significance.
f) Operating abnormalities having safety significance.
g) Reportable occurrences, h) Audit reports.
- 2. The review findings shall be documented in the MITRSC minutes.
7.2.5 Audit Function
- 1. The MITRSC audit function may be performed by members of the MITRSC who do not have line responsibility for the reactor or by a consultant who has
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(s qualifications equivalent to those listed in Specification 7.2.1.2.
- 2. Audits shall be perfc. ed at least annually.
- 3. The scope of the audit shall include, as a minimum, the following:
a) Facility operations for conformance to the technical specifications and license conditions, b) The requalification program for the operating staff.
c) 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.
d) The reactor facility emergency plan and implementing procedures.
e) The physical security plan.
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the Director of the Nuclear Reactor Laboratory. A written report of the findings p
of the audit shall be submitted to the Director of the Nuclear Reactor Laboratory
- and to all MITRSC members within three months after the audit has been completed.
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7.3 Radiation Safety
- 1. The Radiation Protection Program shall be designed to achieve the requirements of 10 CFR 20 with the exceptions given in Specification 3.7.2.
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- 2. An ALARA program shall be designed that applies to facility staff and users as well as to the general public and the environment. )
- 3. The Reactor Radiation Protection Officer has the authority to interdict or l terminate safety-related activities. Disagreements between Reactor Operations and Reactor Radiation Protection shall be resolved pursuant to Specification 7.1.2.5
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1 WY 7.4 Procedures 7.4.1 . Review Process
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Prior to the implementation of a new procedure or a change to an existing one, a written safety review shall be prepared that includes the following:
- 1. A description of the new procedure and/or change to an existing one shall be
summarized.
2 '. A safety evaluation of the new procedure and/or change to an existing one shall be prepared.
- 3. The new procedum and/or change shall be evaluated to determine if:
a) An unreviewed safety question exists, or -
. b) The emergency or security plans are affected, or c) The requalification program is affected, or .
d) The ALARA program is impacted.-
7.4.2 Approval Process 1
No new procedure or a change to an existing one shall be implemented until the material l prepared in Specification 7.4.1 has been reviewed and approved by two licensed senior reactor operators and the Director of Reactor Operations. Where radiation protection considerations are involved, the review'and approval of the MITR Radiation Protection Officer, or a designated.
alternate, shall also be required. Review and approval by the MITRSC is also required for procedures that concern:
1.. The standard operating plan for the reactor including startup and shutdown procedures.
- 2 .' .. f The emergency plan and its implementing procedures.
3; The security plan and its implementing procedures.
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.14;- Ths requalification program.
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-3 In the event that the review required by Specification 7.4.1 identifies an unreviewed safety question or finds that the emergency or security plan is degraded or that the requalification program is weakened or that troLARA program is negatively impacted then the proposal must be referred to the MITRSC (and possibly to the NRC).
7.4.3 Scope of Procedures Written procedures that have been reviewedL and approved pursuant to -
Specification 7.4.1 and 7.4.2 shall be prepareGor all operational activities described in the Safety Analysis Report. As a minimum, these include:
a) Startup, shutdown, and operation of the reactor.
b) Refueling operations, c) Maintenance of components that have nuclear safety significance.
d) Surveillance and testing as required by these technical specifications.
e) Personnel radiation protection consistent _with-applicable regulations or guidelines. These procedures shall include a management commitment and programs to maintain experiments and releases in accordance with the ALARA concept.
f) Administrative controls for operation and maintenance and for the conduct of -
irradiations and experiments that could affect reactor safety or core reactivity.
g) Implementation of required plans such as emergency or security plans.
7.4.4 Equipment Changes
- 1. The review and approval process (Specifications 7.4.1.1 and 7.4.1.2) shall apply to changes of equipment that can affect reactor safety.
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i :7.5' ' Exneriment Review And Anoroval k
I L7.5.1 Review Process -
- 1. - Prior to performing any reactor experiment of a type not previously approved,'
the proposed experiment or irradiation series shall be evaluated in terms 'of its effect on reactor operation and the possibility and consequences ofits failures, including, where significant, consideration of chemical reactions, physical _-
integrity, design life, proper cooling, interaction with core components, and reactivity effects. This evaluation shall normally consist of the following:
a) Preparation of an experiment description and safety evaluation by either-the experimenter or individual (s) appointed by the Director of Reactor .
Operations. ,
b) Preparation and approval of a written safety review as described in ,
Specifications 7.4.1 and 7.4.2.
. . i m c) Preparation of written procedures for the conduct of the experiment. q l
Item (a) above may be done as part of item (b) above at the discretion of the-Director of Reactor Operations.
'I 7.5.2 AnorovalProcess
- 1. No reactor experiment of a type not previously approved shall be performed .
until the materials prepared in Specification 7.5.1 have been reviewed and-approved in writing by two licensed senior reactor operators, the Director of l Reactor Operations, and the MITRSC. .j 4
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- 2. Substantive changes to previously reviewed experiments shall require the. Li j
process described in Specifications 7.5.1 and 7.5.2.1. Minor changes that do 1
not significantly alter the experiment may be implemented upon the review and - l
- .. J approval'of a safety review by two senior reactor operators and the Director of j f);
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Reactor Operations. The safety review shall, as a minimum, describe the change and evaluate its safety.
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- 3. Previously approved experiments may be performed at the discretion of a 1
licensed senior reactor operator without the necessity of any further review or approval.
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- 4. All experiments shall conform to the General Experiment Criteria as given in Specification 6.1.
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.. . 1. . i hj j 7.6 ' Remimd Action V
7.6.1 Action to be Taken in Case of Safety Limit Violation
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- 1. The following action shall be taken in the event of a iafety limit violation:
a) The reactor shall be shut down, and reactor operation shall not be resumed until authorized by the U.S. Nuclear Regulatory Commission.1
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b) The safety limit violation shall be promptly reported to the Superintendent of Operations and Maintenance, the Director of Reactor - o i
Operations, the Director of the Nuclear Reactor Laboratory, and the J Chairman of the MITRSC.
c) The safety limit violation shall be reported to the U.S. Nuclear
. Regulatory Commission.
d). A safety limit violation report shall be prepared. The report shall' q describe the following:
(i) A?plicable circumstances leading to the violation including, wxn known, the cause and contributing factors.
O, (ii) Effect of the violation upon reactor. facility ' components, systems, or structures, and on the health and safety of personnel and the public.
(ii) Conective action to be taken to prevent recurrence.
l This report shall be reviewed by the MITRSC and any follow-up report shall be -
submitted to the U. S; Nuclear Regulatory Commission when authorization is sought to resume operation of the reactor. i 7.6.2 Action to be Taten in the Event of a Reportable Occunence
- 1. The following actions shall be taken in the event of a reportable occurrence:
a) - - The reactor shall be shut down unless the cause for the occurrence has i been identified and rectified immediately upon discovery or unless the -
, occurrence has no immediate safety sigmficance.
'b): The D'irector of Reactor Operations shall be notified.
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c) : Operation of reactor shall not be resumed until so authorized by the' .i
(): , Director of Reactor Operations, or a designated alternate.
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d) The U.S. Nuclear Regulatory Commission shall be notified.
V(~N c) A report shall be prepared that includes the time and date of the occurrence, reactor status at the time of the occurrence, a description of the occurrence, an evaluation of the cause of the occurrence, a record of the corrective action taken, and recommendations for appropriate action to prevent or reduce the probability of recurrence. This report shall be i submitted to the U. S. Nuclear Regulatory Commission and it will be revie,wed by the MITRSC no later than its next regularly scheduled meetmg.
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7.7 i Repods
'- Annual Recon 7.7.1 An annual or operating report will be submitted to the U.S. Nuclear' Regulatory /
Commission within ninety days following the 31st of December of each year. Its content is as c.
follows:
i 1.- A narrative summary of operating experience. (including experiments.
performed) and of changes in facility design, performance characteristics and 3 operating procedures related to reactor safety occurring during the reporting -
period, as well as results of surveillance tests and inspections required by these :
Technical Specifications.
- 2. Tabulation showing the energy generated by the reactor, the number of hours the reactor was critical, and the cumulative total energy output since initial.
criticality.
- 3. The number of emergency shutdowns and inadvertent scrams, including the reasons therefore.
- 4. Discussion of the major maintenance operations performed during the period, including the effect,if any, on the safe operation of the reactor, and the reasons for any corrective maintenance required.
5.. A ' description of each change to the facility or procedures, tests, and experiments carried out under the conditions of 10 CFR 50.59 including a summary of the safety evaluation of each.
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,_A description of any environmental surveys performed outside the facility.
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ai' A' summary of-radiation exposures. received by facility personnel and experimenters, including the dates and time of significant exposures, and a summary of the results of radiation and contamination surveys performed :
within the facility.
- 8. A summary of the nature and amount of radioactive effluents released or-discharged to the environment beyond the effective control of the licensee as -
measured at or prior to the point of such release or discharge.
a) Liauld Waste (summarized on a monthly basis)
- Total gross radioactivity released during the reporting period, excluding tritium.
- Total tritium radioactivity and average concentration released during the reporting period.
- Total radioactivity (beta / gamma) released for specific nuclides, if '
the gross beta radioactivity exceeds 1x10-5 Ci/ml at point of release, during the reporting period.'
- Total volume of effluent water (including diluent) during periods of release.
b) Gaseous Waste
- Radioactivity of principal radionuclides discharged during the reporting period for
- Gases
- Particulates, with halflives greater than eight days.
- The effluent concentration limit used and the estimated activity -
discharged during the reporting period, by nuclide for principal radionuclides, based on representative isotopic analysis.
c) Solid Waste
- The total amount of solid waste packaged.
.. The total activity and type of activity involved.
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- The dates of shipment and disposition (if shipped offsite).
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- 9. 'A' summary of the use of the medical therapy facilities for human therapy..
a) Investigative Studies
- Nature and status of the studies.
. . Number of patients involved.
b) Human Theraov
- Number of patients treated.
- Type of cancer treated.
7.7.2 Renoitable Occurrence Reoorts
- 1. A report shall be made by telephone or similar conveyance to the U.S. Nuclear Regulatory Commission within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of:
I a) Operation in violation of a safety limit.
b) Any release of radioactivity to unrestricted areas above permissible limits, whether or not the release resulted in property damage, personal l Injury, or exposure.
c) Any reportable occurrence as defined in Specification 1.3.32 d) Any significant variation of measured values from a corresponding predicted or previously measured value of safety-related operating characteristics.
- 2. A written report shall be provided as a follow-up to the verbal one within ten working days of the occurrence. This report shall provide the information required by Specification 7.6.2.l(e). The report shall be submitted to the NRC Document Control Desk.
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7.7.3 Stiecial Reports
- 1. Special repons, other than reponable occurrence repons, shall be provided in writing within 30 days to the NRC Document Control Desk in the event of:
a) Permanent change in the facility organization including Level 1 or Level 2 personnel as defined in Figure 7.1-1.
b) Significant changes in the transient or accident analysis as described in this Safety Analysis Repon.
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9-r 7.8 Records Retention 4 7.8.1- Five-Year Record Retention -
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- 1. The following records.will be retained for five years or for the life of the j l
component involved if less than five years:
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a) Records of normal reactor operation including power levels and periods . r,'
of operation at each power level. (Note: -Excludes retention.of supporting documents such as checklists, log sheets, etc4 which shall -
be retained for a period of at least one year.)
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b) Records of principal maintenance activities including inspection, repair, substitution, or replacement of principal items of equipment pertaining -
to nuclear safety.
c) Records of reportable occurrences.
d) Records of surveillance activities that are required by these technical' specifications.
c) Records of reactor facility radiation and contamination surveys.
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f) Records of experiments performed with the reactor.
- b)- g) Records of fuel inventories, receipt, and shipments. (Nnla: Records of individual fuel element usage shall be retained until the element isL returned to the U.S. Department of Energy.)
h) Records of changes made in the operating procedures.
i) Records of audit reports including both internal audits and those performed for or by the MITRSC.
7.8.2 -- Six-Year Record Retention ,
- 1. The following records will be retained for six years:
a) Records of individual licensed staff members indicating qualifications, experience, training, and requalification. (Noic: These are retained at .)
' all times that the individual is employed or until certification is y renewed.) Li i
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F , 3 7.8.3 Life'of Facility -
- 1. The following records will be retained for the life of the facility:
a)- Records of' radioactivity in liquid and gaseous waste released to'the .
environment.
~ b) ' Records of off-site environmental monitoring.
. c) Records of radiation of all plant personnel and others who enter .
radiation control areas.
d) ' Records and drawing changes reflecting plant design modifications made to systems and equipment described in the Safety Analysis Report.
c) Records of radioactive shipments including solid waste disposal.
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