ML20217M472
| ML20217M472 | |
| Person / Time | |
|---|---|
| Site: | Rhode Island Atomic Energy Commission |
| Issue date: | 10/14/1999 |
| From: | RHODE ISLAND, STATE OF |
| To: | |
| Shared Package | |
| ML20217M459 | List: |
| References | |
| NUDOCS 9910270228 | |
| Download: ML20217M472 (55) | |
Text
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l TABLE OF CONTENTS f
1.0 '
DE FIN ITIO N S.............................................................................
1.1 -
Certi fied Ope rator................................................................................ 1
- 1.2 Co n fi n e me n t........................................................................
............I j
I 1.3 -
Experiment................................................................................
.. 1 1.4 '
Ex plos i ve M ate rial........................................................................... 2 1.5 -
Instrumentation Cha' nnel.................................................................... 2 1.6 :. Limiting Conditions 'of Operation (LCO).............................................. 2 1.7 Limiting Safety System Setting (LSSS).................................................. 2 1.8 Measured Channel.............................................................................. 3 -
1.9 - : M e as u re d V al ue.....................................................................
1.10 Operable..............................................................................................3 1.11
-Operating.............................................................................................3 1.12 Operational Reactor Core...................................................................... 3 1.13-Protec tive Ac tion............................................................................... 3 -
1.14 Reac tivity Exce s s.................................................................................... 3 1.15 Reac ti vi ty Li mits................................................................................... 3 1.16. Reactivity Wonh of an Experiment........................................................ 4 i
1.17 : Reactor Operating.............................................................................. 4 1.18 Re actor S a fe ty S vs te ms.......................................................................... 4 1.19 Reactor Securc ~..
............................................................................4 1.20 R eactor S hutdown............................................................................... 5
' l.21.
Readily Available on Call.................................................................. 5 '
l.22 -- Refere nce Core Condition.................................................................... 5
)
1.23-Regu latin g B l ade................................................................................... 5 J
'l.24 Re movable Ex peri ment....................................................................... 5 -
1.25 Re portable Occurre nce........................................................................ 5 1.26. Researc h Reac t or. ;................................................................................... 6 1.27 Rundown.................................................................................................6 i
1.28 S afe ty C h an ne l....................................................................................... 6
'l.29 S afe ty Li mi ts........................................................................................... 6 l'.30.
ScramTime............................................................................................6 1.31 S him S afe ty B l ade................................................................................... 6 1.32-S hal l, S houl d an d M ay....................................................................... 7 i
i 9910270228 991014 PDR ADOCK 0S000193 P
.n
. r.
1.33 S h u tdown M argi n.................................................................. 7 1.34 S ecu re d Ex pe-i me nt............................................................................. 7 1.35 S tatic Reacti vi ty. Wort h..................................................................... 7 1.36 S t an d ard R eact or Core........................................................................ 7 1.37-S urveillance Activities....................................................................... 7 1.38-S urveillance Intervals................................................................ 8 1.39: Tru e V al u e...................................................................,.......... 8
.1.40. Un sched uled S h u td own................................................................. 8 2.0.
SAFET / LIMITS AND LIMITING SAFETY SYSTEM SETTINGS.................................................................................................9
'2.1.
Safety Limits........
.............................................................................9 2.2 Limiting Safety System Settings (LSSS)...........................................10 3.0 LIMITING CONDITIONS FOR OPERATION......................................... 13 3.1 ' Reacti vity Li mi t s..................................................................................... I 3 l
3.2 Reac to r S a fe ty S yste m............................................................................... I 5
]
-Table 3.1..................................................................................................17 Table 3.2...............................................................................................I9 1
3.3 Cool an t W a t e r............................................................................
1 3.4, 3.5, 3.6 Confinement and Emergency Exhaust System
. and Emergency Power..................................................
............22
]
3.7 Radiation Monitoring Systems and Effluents....................................... 24 3.8 Limitations on Experiments............................................................. 26 3.9. Reactor Core Components................................................................ 2 8 4.0 S URVEILLANCE REQUIREMENTS......................................................... 30 l
1
'4.1 ' R e ac tivi ty Li mits......................................................................... 3 0 4.2. Reac t or S afe ty S yste m......................................................................... 31 l
4.3: Wat e r Cool an t S yste m '..................................................................... 3 2 4.4, 4.5, 4.6 R1NSC Conf' ement B uitding................................................... 34 m
l L 4.7 Radiation Monitoring Systems and Effluents............................................ 36
. 4.8. S urveillance of Experiments.............................................................. 37 4.9 Reactor ' Core Compone nts..................................................................... 3 8
5.0 DESIGN FEATURES..
.. 40 l
5.1 Description.......
........ 40 5.2 R eac t or Fuel...................................
...40 5.3 Reactor Core.....
......40 i
5.4 Reactor Building.....
.40
{
5.5 Fuel Storage..
.............41 j
6.0 ADMINISTRATIVE CONTROLS..
.... 4,.
6.1 Organization and Management...............................................42 Figure 6.1 - Organizational Chart...................................... 43 6.2 Qualifications of Personnel............................................. 44 6.3 Responsibilities of Personnel................................................. 44 6.4 Review and Audit....................
.......46 6.5 Operating Procedures........
.....................47 6.6 Action to be Taken in the Event of a Reportable Occurrenee.......................................................................48 6.7 Action to be Taken in the Event a Safety Limit is Exceeded..............................................................................48 6.8 Reporting Requirements......
..................................................49 6.9 Plant Operating Records.................
....................51 l
TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-191, License R-95 L
Revisi; ; 1 l
1.0 DEFINITIONS l
1.1 Certified Operator An individual authorized by the U. S. Nuclear Regulatory Commission to carry out the responsibilities as.ociated with the position requiring the certification.
1.1.1 Senior Reacto, Operator An individual who is licensed to direct the activities of reactor operators.
Such an individual may be referred to as a class A operator.
1.1.2 Reactor Operator An individual who is licensed to manipulate the controls of a reactor. Such an individual may be referred to as a class B operator.
1.2 Confmement Confmement means an enclosure on the overall facility which controls the movement of air into it and out through a controlled path.
1.3 Experiment Any operation, component, or target (excluding devices such as detectors, foils, etc.), which is designed to investigate non-routine reactor characteristics or which is intended for irradiation within the pool, on or in a beam tube or irradiation facility and which is not rigidly secured to a core or shield stmeture so as to be part of their design.
1.3.1 Experiment, Moveable A moveable 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.
1.3.2 Experiment, Secured A secured experiment is any experiment, experimens Scility, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means. The restraining force must be substantially greater than I
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 which can arise as a result of credible conditions.
i Page1 Amendment 26 l
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-1 m
4 TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License R-95 Revision 1 1.3.3 Experimental Facilities An experimental facility is any structure or device which is intended to guide, orient, position, manipulate, or otherwise facilitate a multiplicity of experiments of similar character.
1.4 Explosive Material Explosive material is any solid or liquid which is categorized as a severe, dangerous, or very dangerous explosion hazard ~in DANGEROUS PROPERTIES OF INDUSTRIAL MATERIALS by N.I. Sax, third Ed. (1968), or is given an identification.of Reactivity (Stability) Index of 2, 3, or 4 by the National Fire Protection Association in its publication 704-M,1966.
1.5 Instrumentation Channel A channel is the combination of sensor, line, amplifier, and output device which are connected for the purpose of measuring the value of a parameter.
1.5.1 Channel Test Channel test is Qe introduction of a signal into the channel for verification that it is operable.
1.5.2 Channel Check Channel check is a qualitative verification of acceptable performance by observation of channel behavior. This verification, where possible, shall include comparison of the channel with other independent channels or systems measuring the same variable.
1.5.3 Channel Calibration Channel calibration is an adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the parameter which the channel measures.
Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip and sha'l be deemed to include a channel test.
1.6 Limiting Conditions of Operation (LCO)
Lowest functional capability or performance levels of equipment required for safe operation of the reactor (10CFR50.36).
1.7 Limiting Safety System Setting (LSSS).
q Page 2 Amendment 26 l
1
TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License.R-95 Revision 1 Settings for automatic protective devices related to those variables having significant safety functions, and chosen so that automatic protective action will correct an abnormal situation before a safety limit is exceeded (10CFR50.36).
1.8 Measured Channel j
A measured channel is the combination of sensor, amplifiers, and output devices which are used for the purpose of measuring the value of a parameter.
1.9 Measured Value The measured value of a parameter is the value of the variable as indicated by a measuring channel.
1.10 Operable i
. Operable means that a component or system is capable of performing its intended function.
l
' 11 Operating Operating means that a component or system is performing its intended function.
1.12 Operational Reactor Core An operational core is a standard core for which the core parameters of excess reactivity, shutdown margin, fuel temperature, power calibration, and reactivity worths of control rods and experiments have been determined to satisfy the requirements set forth in the Technical Specifications.
1.13 Protective Action 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.
1.14 Reactivity Excess Excess reactivity is that amount of reactivity that would exist if all the control rods were moved to the maximum reactive condition from the point where the reactor is exactly critical.
1.15 Reactivity Limits
. Page 3 Amendment 26
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TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193. License R-95 Revision 1 l
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The reactivity limits are those limits imposed on the reactor core excess reactivity.
Quantities are referenced to a reference core condition.
1.16 Reactivity Worth of an Experiment The reactivity _ worth of an experiment is the maximum absolute value of the reactivity change that would occur as a result of intended or anticipated changes or credible malfunctions that alter equipment position or configuration.
1.17 Reactor Operating The reactor is operating whenever it is not secured or shut down. The reactor has two modes of operation: natural circulation - not to exceed 0.1 MW and forced circulation - not to exceed 2 MW.
1.18 Reactor Safety Systems Reactor saf ty systems are those systems, including their associated input channels, w'
-h are designed to initiate automatic reactor protection or to provide informatio r initiation of manual protective action.
1.19 Reactor Secure The reactor is secure when:
1.19.1 Suberitical:
1 There is insufficient fissile material or moderator present in the reactor, l
control rods or adjacent experiments, to attain criticality under optimum
~
available conditions of moderation and reflection, or the following conditions l
exist:
The minimum number of neutra absorbing control rods are fully a.
inserted in shutdown porta, as required by technical specifications.
b.
The master switch is in the off position and the key is removed from the lock.
c.
No work is in progress involving core fuel, core structure, installed control rods, or control rod drives unless they are j
physically decoupled from the control rods.
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d.
No experiments are being moved or serviced.
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1 l
e TECHNICAL SPECIFICATIONS l
Rhode Island Nuclear Science Center Docket 50-193, License R-95 Revision 1 j
i j
l.20 Reactor Shutdown The reactor is shut down if it is suberitical by at least the shutdown margin in the reference core condition with the reactivity of all installed experiments included.
1.2 i Readily Available on Call Readily available on call shall mean a licensed senior operator shall insure that he can be contacted within ten minutes and is within a 30 minute driving time from the reactor building when the reactor is being operated by a licensed operator.
1.22 Reference Core Condition i
j The condition of the core when it is at ambient temperature (cold) and the reactivity worth of xenon is negligible < 0.05% Ak/k.
l.23 Regulating Blade
{
The regulating blade is a control blade of low reactivity worth fabricated from stainless steel and used to control reactor power. The blade may be controlled by the operator with a manual switch or by an automatic controller.
1.24 Removable Experiment A removable experiment is any experiment, experimental facility, or component of an experiment, other than a permanently attached appurtenance to the reactor j
system, which can reasonably be anticipated to be moved one or more times 4
during the life of the reactor.
l 1.25 Reportable Occurrence i
A reportable occurrence is any of the following:
1 1.
A safety system setting less conservative than the limiting setting established in the Technical Specifications; 2.
Operation in violation of a limiting condition for operation established in the Technical Specifications; 3.
A safety system component malfunction or other component or system malfunction which could, or threaten to, render the safety system incapable of performing its intended safety functions; 4.
Release of fission products from a failed fuel element; Page 5 Amendment 26
TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193. License R-95 Revision 1 5.
An uncontrolled or unplanned release of radioactive material which results in concentrations of radioactive materials inside or outside the restricted area in excess of the limits specified in Appendix B of 10CFR20; 6.
An uncontrolled or unanticipated change in reactivity in excess of 0.5
%AK/K; 7.
Conditions arising from natural or man-made events that affect or threaten to affect the safe operation of the facility; i
8.
An observed inadequacy in the implementation of administrative or procedural controls such that the inadequacy causes or threatens to cause the existence or development of an unsafe condition in connection with the operation of the facility.
1.26 Research Reactor A research reactor is defined as a device designed to support a self-sustaining neutron chain reaction for research, development, educational training, or experimental purposes, and which may have provisions for the production of radioisotopes.
1.27 Rundown A rundown is the automatic insertion of the shim safety blades.
1.28 Safety Channel A safety channel is a measuring channel in the reactor safety system.
1.29 Safety Limits Safety limits are limits on important process variables which are found to be necessary to reasonably protect the integrity of the principal barriers which guard against the uncontrolled release of radioactivity. The principal barrier is the fuel element cladding.
1.30 Scram Time Scram time is the elapsed time between reaching a limiting safety system set point and specified control rod movement.
1.31 Shim Safety Blade Page 6 Amendment 26
TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License R-95 Revision 1
. A shini safety Llade is a control blade fabricated from borated aluminum which is used.to. compensate for fuel burnup, temperature, and poison effects. A shim -
safety blade is magnetically coupled to its drive unit allowing it to perform the function of a safety blade when the magnet is deenergized.
1.32.'
Shall, Should and May _
The word "shall" is used to denote a requirement. The word "should" is used to denote a recommendation, The word "may" is used to denote permission, neither a requirement nor a recommendation.
l.33 Shutdown Margin-Shutdown margin shall mean the minimum shutdown reactivity necessary to provide. confidence that the reactor can be made subcritical by means of the-control and safety systems starting from any permissible operating condition and with the most reactive rod in its most reactive position, and that the reactor will remain subcritical without funher operator action.
'l.34 Secured Experiment Any experiment, experimental fa^ ility, or component of an experiment is deemed c
to be secured, or in a secured position, if it is held in a stationary position relative
. to the reactor by mechanical means. The restraint shall exen sufficient force on the~ experiment to overcome the expected effects of hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the experiment, or of forces which might arise as a result of credible malfunctions.
1.35 Static Reactivity Worth The static reactivity worth of an experiment is the absolute value of the reactivity i
change which is measurable by calibrated control rod comparison methods.
1.36-Standard Reactor Core
'A standard core is an arrangement of (14) 22-plate LEU fuel elements in the reactor grid plate and may include installed experiments.
1.37.
Surveillance Activities
. Surveillance activities (except those specifically required for safety when the reactor is shutdown). may be deferred during reactor shutdown, however, they must be completed voor to reactor startup unless reactor operation is necessary for performance of tne activity. Surveillance activities scheduled to occur during an Page 7 Amendment 26 l,.
TECl!NICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License R-95 Revision I operating cycle which cannot be performed with the reactor operating may be deferred to the end of the cycle.
1.38 Surveillance Intervals Maximum intervals are to provide operational flexibility and to reduce frequency.
Established frequencies shall be maintained over the long term. Allowable surveillance intervals shall not exceed the following:
- 1. 5 years (interval not to exceed 6 years).
- 2. 2 years (interval not to exceed 21/2 years).
- 3. Annual (interval not to exceed 15 months).
- 4. Semiannual (interval not to exceed 71/2 months).
- 5. Quarterly (interval not to exceed 4 months).
- 6. Monthly (interval not to exceed 6 weeks).
- 7. Weekly (interval not to exceed 10 days).
- 8. Daily (must be done during the calendar day).
1,39 True Value The true value is the actual value of a parameter.
1.40 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 which could adversely affect safe operation, not including shutdowns which occur during testing or check-out operations.
l Page 8 Amendment 26 l
TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193. License R-95 Revision I 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS l
' 2.1 Safety Limits 2.1.1 Safety Limits in the Forced Convection Mode l
Applicability:
This specification applies to the interrelated variables associated with core thermal and -hydraulic performance in the steady state with forced convection flow. These variables are:
~
Reactor Thermal Power, P Reactor Coolant Flow through the Core, m Reactor Coolant Outlet Temperature, To Height of Water Above the Top of the Core, H Objective:
To assure that the integrity of the fuel clad is maintained.
Specifications:
1.
The true value of reactor power (P) shall not exceed 2.4 MW.
2.
The trae value of reactor coolant flow (m) shall not be less than 1580 gpm.
3.
The true value of the reactor coolant outlet temperature (To) shall not exceed 125 oF.
4.
The true value of water height above the active core (H) shall not be less than 23.54 feet while the reactor is operating at any power level.
Bases:
1 The basis for forced. convection safety limits is the ac calculated maximum cladding temperature in the hot ch-nd of the most compact core will not be exceeded. The tneanal hydraulic analysis (Part B, of the SAR) shows that if the safety limits are not exceeded the coolant will not reach the onset of nucleate boiling even at the safety limit of 2.4 MW.
Additionally, the limit on coolant outlet temperature will prevent I
exceeding the temperature limit for the cleanup system resin.
1 1
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Page 9 Amendment 26
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. TECHNICAL SPECIFICATIONS Rhode Island Nucle r Science Center Docket 50-193. License R-95 Revision 1 2.1.2 L ' Safety Limits in the Natural Convection Mode Applicability:
This specification applies to the interrelated variables associated with core
- thermal and hydraulic performance in the natural convection mode' of i-operation. These variables are:
I Reactor Thermal Power, P.-
Height of Water Above the Top of the Core, H -
Pool Temperature, Tp i
. Objective:
l 1.
To assure that the integrity of the fuel clad is maintained l'
2.
To assure consistency with other ' defined safety system parameters.
l Specification:
-1.
The true value of the reactor thermal power shall not exceed 217 L
kw.
l
. 2.
The height of pool water above the core shall not be less than 23.54 feet.
3.
The pool temperature does not exceed 130 oF.
i Bases:
l The basis 'for natural convection safety limits is that the calculated maximum cladding temperature in the hot channel of the most compact core will not reach nucleate boiling of the water coolant at a pool depth of i
23.54 feet.
1
.-2.2 Limiting Safety System Settings (LSSS)
..2.2.1 Limiting Safety System Setting in the Forced Convection Mode
- Applicability
I-LEU Fuel Temperature - Forced Convection Mode
. Objectivei l
l l
Page 10 Amendment 26 Lu-
TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50193. License R 95 Revision I This specification applies to the setpoint for the safety channels monitoring reactor power, primary coolant flow, pool level and core outlet j
temperature to assure that the maximum fuel temperature permitted is such that no damage to the fuel cladding will result in the forced convection rnode.
Specification:
The limiting safety system settings for reactor thermal power (P), primary coolant flow through the core (m), height of water above the top of the i
core (H), and reactor coolant outlet temperature (To) shall be as follows:
Parameter LSSS P
(Max) 2.30 MW m
(Min) 1600 gpm H
(Min) 23.7 ft To (Max) 121 oF Bases:
These specifications were determined to prevent fuel temperatures from exceeding NRC temperature limits of 530 oC. This temperature was found acceptable as a result of NUREG-1313, " Safety Evaluation Report related to the Evaluation of Low-Enriched Uranium Silicide-aluminum Dispersion Fuel for Use in Nonpower Rea-tors" The SAR (Part B) provides the analyses showing fuel cladding tenmcratures well below the NUREG limit at normal operation. Flow and ter erature limits were chosen to prevent incipient boiling even if transient power rises to the 2 MW trip limit of 2.4 MW. Variables used in the SAR were analyzed using uncertainties in flow measurement (3%) and temperature measurement (3%).
These uncertainties were incorporated in the hot channel factors (1) used in the SAR thermal hydraulic studies. These same uncertainties were applied to the inlet and outlet temperature measurements.
The LSSS for the pool level is set for a scram upon a 2 inch drop in water level. The reference height of 23.7 feet (16 inches below suspension frame base plate elevation) is the depth of water above the top of the active fuel sitting in the existing reactor grid box. This depth was used in the SAR Loss of Coolant Analysis (Part B of the SAR). The safety limit settings chosen provide acceptable safety margins to the maximum fuel cladding temperature. The startup accident transient analysis (Part A,Section XI of the SAR) aise provides results showing that the cladding temperature limit is not exac!.ed. The LOCA analysis (Design Basis Accident, Part A,Section IX and Part B,Section X and Appendix D of the S AR) shows that the fuel cladding limit is not exceeded, Page11 Amendment 26
TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License R-95 Re vision 1 The LSSS for the pool level results in a higher number since the pool level scrams upon a 2 inch drop in water level.
(1)
Reference:
Report on the Determination of Hot Spot Factors for the RINSC Research Reactor, August,1989 2.2.2 Limiting Safety System Settings in the Natural Convection Flow Mode Applicability:
These specifications apply to the setpoint for the safety channels monitoring reactor thermal power level (P), monitors for pool level (H),
and pool water temperature (T ) in the natural convection mode.
p Objective:
To assure that automatic protective action is initiated to prevent a safety limit from being exceeded.
Specification:
1.
The limiting safety system setting for reactor thermal power (P),
height of water above the top of the core (H), and pool water temperature (T ) shall be as follows:
p Parameter LSSS P
(Max) 115 kw H
(Min) 23.7 ft.
T (Max) 126 oF p
Bases:
The SAR has determined that up to 217 kw can be removed by natural convection, however, the existing license requirement of 100 kw operation will be maintained and with a 15% overpower trip,115 kw will be the LSSS. The pool level scram (2 inch drop) is the same as the forced convection mode. The pool temperature 130 oF safety limit, having a 3%
error, results in a LSSS of 126 oF. The LSSS for natural convection assures that automatic protective action will prevent a safety limit from being exceeded.
Page 12 Amendment 26
TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193. License R-95 Revision 1 i
3.0 LIMITING CONDITIONS FOR OPERATION 3.1 -
Reactivity Limits Applicability:
This specification upplies to the reactivity of the reactor core and to the reactivity wonhs of control rods and experiments.
Objective:
To assure that the reactor can be controlled and shut down at all times and that the safety limits will not be exceeded.
Specification:
1.
The shutdown margin relative to the reference core condition shall be at least 1.0 ScAK/K with the most reactive shim safety rod and the regulating rod fully withdrawn.
2.
The overall core excess reactivity including movable experiments shall not exceed 4.7 %AK/K.
3.
The total reactivity wonh of all experiments shall not exceed 0.6
%AK/K.
4.
The reactivity worth of each experiment shall be limited as follows:
Experiment Maximum Reactivity Worth Moveable 0.08 %AK/K Secured 0.60 %AK/K 5.
The r: ctor shall be suberitical by at least 3.0 %AK/K during fuel loading changes.
6.
The reactivity worth of the regulating rod shall not exceed 0.6 %AK/K.
7.
Experiments which could increase reactivity by flooding, shall not remain in or adjacent to the core unless the shutdown margin required in l
Specification 3.1.1 would be satisfied after flooding.
8.~
The temperature coefficient will be negative and surveillance will be conducted at initial startup and change in fuel type.
Page 13 Amendment 26
l TECHNICAL SPECIFICATIONS l
Rhode Island Nuclear Science Center l
Docket 50-193. License R-95 Revision 1 9.
For operation at power levels in excess of 0.1 MW in the forced convection mode, all grid positions shall contain fuel elements, baskets, reflector elements, grid plugs or experimental facilities.
10.
For operation at powers in excess of 0.1 MW, the pool gate must be in its storage location.
l Bases:
l Specification 3.1.1 assures that the reactor can he shutdown from any operating condition and will remain subcritical after cool down and xenon decay even if the rod of the highest reactivity worth should be in the fully withdrawn position. The SAR (Part A,Section V) demonstrates that the shutdown margin conservatively j
exceeds the 1% in Specification 3.1.1.
Specification 3.1.2 limits the allowable excess reactivity to the value necessary to i
i overcome the combined negative reactivity effects of: (1) an increase in primary coolant temperature; (2) fission product xenon and samarium buildup in a clean core; (3) power defect due to increasing from a zero power, cold core to a 2 MW, hot core; (4)- fuel burnup during sustained operation; and (5) moveable experiments.
Specification 3.1.3 limits the reactivity worth of experiments to values of reactivity which, if introduced as positive step changes, will not cause fuel melting.
Specification 3.14 limits the individual reactivity worth of an experiment to a value that will not produce a stable period of less than 30 seconds and which can be compensated for by the action of the control and safety system without exceeding any safety limits.
Specifications 3.1.5 provide assurance that the core will remain subcritical during fuel loading changes.
Specification 3.1.6 assures that failure of the automatic control system will not introduce sufficient excess reactivity to produce a prompt critical condition.
l Specification 3.1.7 assures that the shutdown margin required by Specification 3.1.1 will be met in the event of a positive reactivity insenion caused by the flooding of an experiment.
Specification 3.1.8 assures that the power increase is self limiting.
Specification 3.1.9 will prevent the degradation of flow rates due to flow bypassing the active fueled icgion through an unoccupied grid plate position.
Page 14 Amendment 26
1 TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50193, License R-95 Revision 1 Specification 3.1.10 assures that the full volume of the pool water is available to provide cooling of the core during normal operation and in the event of a loss of coolant accident.
3.2 Reactor Safety System Applicability:.
These specifications apply to the reactor safety system and other safety related instmmentation.
Objective:
To specify the lowest acceptable level of performance or the minimum number of acceptable components for the reactor safety system and other safety related instrumentation.
Specification:
The reactor shall not be made critical unless:
1.
The reactor safety systems and safety related instrumentation are operable in accordance with Tables 3.1 and 3.2 including the minimum number of channels and the indicated maximum or minimura setpoint; 2.
All shim safety blades are operable' in accordance with Technicai Specification 4.1.1 and 4.1.2.
3.
The time from initiation of a scram condition until the cor. trol element is fully inserted shall not exceed I second in accordance with Technical Specification 4.2.5 and 4.2.6.
4.
The reactivity insertion rates of individual control and regulating blades will not exceed 0.02 %AK/K per second.
Bases:
Ne"'.ron flux level scrams provide redundant automatic protective action to prevent exceeding the safety limit on reactor power. The period scram limits the rate of rise of the reactor power to periods which are manually controllable without reaching excessive power levels or fuel temperatures.
)
Page 15 Amendment 26
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i TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License R-95 l
Revision i I
The loss of flow scram assures that an automatic loss of flow scram will occur in the event of a loss of flow when the reactor is operating at power levels above 0.1 MW.
The reactivity insertion rate limit was determined in the SAR,Section XI and predicts a safe fuel clad temperature.
l Page 16 Amendment 26 l
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TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License R-95 Revision i TABLE 3.1
. REQUIRED SAFETY CHANNELS Reactor Reactor Safety Minimum Operating Mode System Component / Channel Required Function in Which Reauired Reactor Power Level 2
Automatic Both Modes scram when 2115% of range scale with 2.3 MW max Coolant Flow Rate 1
Automatic Forced Convection scram at above 0.1 MW 51600 gpm Seismic Disturbance 1
Automatic Both Modes l
scram at Modified Mercalli Scale IV Bridge Misalignment 1
Automatic Forced Convection scram above 0.1 MW Pool Water Level 1
Automatic Both Modes scram at 16" below suspension frame base plate elevation Coolant Outle! Temperature 1
Automatic Forced Convection scram above 0.1 MW L
2121oF l
I l
Page 17 Amendment 26
l TECHNICAL SPECIFICATIONS l
Rhode Island Nuclear Science Center Docket 50-193. License R-95 Revision 1 Reactor Reactor Safety Minimum Operating Mode l
System Component / Channel Reauired Function in Which Reauired l
Bridge Movement 1
Automatic Both Modes scram Coolant Gates Open 1
Automatic Forced Convection l
either the l
coolant riser or Coolant downcomer gates open Detector High Voltage Failure 3
Automatic Both Modes scram if Voltage decreases 50V max Log N Period 1
Automatic Both Modes scram if period 5 4 sec No Flow fhermal Column 1
Automatic Forced Convection scram above 0.1 MW Manual Scram 2
Manual Both Modes Switch (console, bridge) scram k
Pool Temperature 1
Automatic Natural Convection
{
Scram at l
2126 F l
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l TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License R-95 Revision i TABLE 3.2 Required Safety Related Instrumentation Minimum Reactor Number Operating Mode Instrumentation Setpoint Reauired Function In Which Read.
1.
Reactor Coolant s 11loF 1
Alarm FC 2 0.1MW Inlet Temperature 2.
Reactor Coolant s119oF 1
Alarm FC 2 0.1MW Outlet Temperature 3.
Log Count Rate
< 3 cps 1
Rod with-Both Modes drawal interlock 4.
Servo Control 2 30 sec 1
Auto Control Both Modes Interlock (fullout)
Interlock Facility Radiation (a)
Monitoring System 5.
Building AirGaseous 2.5x 1
Alarm Both Modes Exhaust (Stack) normal particulate 2 x normal 6.
Reactor Bridge 2 x normal Alarm Both Modes 7.
Fuel Safe 2 x normal Alarm Both Modes or 5mR/hr,
)
which ever is t
h:3 er h
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TECHNICA.L SPECIFICATIONS Rhode Island Nucler.r Science Center d
(
Docket 50-193, License R-95 Revision 1 i
1 Minimum Reactor 1
Number Operating Mode l
Instrumentation Setpoint Reauired Function In Which Read.,
8.
Thermal Column 2 x normal Alarm Both Modes i
or 2mR/hr, which ever is higher 9.
Heat Exchanger 2 x normal Alarm Both Modes 10.
Primary 2 x normal Alarm Both Modes Demineralizer (Hot DI) 11.
Continuous Air 2 x normal 1
Alarm Both Modes Monitoring Unit NOTES (a)
The facility radiation monitoring system consists of 8 radiation detectors which alarm and readout in the control room except for #1I which has a local alarm and readout only. The normal setpoints for this system are shown in Table 3.2. Use of higher than normal setpoints will require approval of the Director or the Assistant Director. Any senior operator member may adjust a setpoint lower than the normal value.
(b)
- The reactor shall not be continuously operated without a minimum of one radiation monitor en the experimental level of the reactor building and one monitor over the reactor pool operating and capable of warning personnel of high radiation levels.
l l
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TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193. License R-95 Revision 1 3.3 Coolant Water (a)
Primary Coolant Water Applicability:
This specification applies to the limiting conditions for primary coolant pH, resistivity, available pool water volume and radioactivity.
Objective:
To maintain the primary coolant in a condition to minimize the corrosion of the primary coolant system, fuel cladding, and other reactor components, and to assure proper conditions of coolant for normal and emergency requirements.
Specification:
1.
The primary coolant pH shall be maintained between 5.5 and 7.5.
2.
The primary coolant resistivity shall be maintained at a value greater than 500K ohms /cm (conductivity 2 micrombos/cm).
3.
The primary coolant shall be analyzed for radioactivity.
Bases:
Experience at this end other facilities has shown that the maintenance of primary coolat., system water quality in the ranges specified in specification 3.3.1 and 3.3.2 will control the corrosion of the aluminum components of the primary coolant system and the fuel element cladding.
Conductivity Specification 3.3.2 also insures adequate water purity to control activation of coolant water impurities.
The requirement in specification 3.3.3 ensures that the presence of unusual impurities or corrosion products is detected.
(b)
Secondary Coolant Water Applicability:
This specification applies to the limiting conditions for secondary coolant pH, cycles of chloride, resistivity and radioactivity.
Page 21 Amendment 26
TECHNICAL SPECIFICATIONS I
Rhode Island Nuclear Science Center Docket 50-193, License R-95 Revision 1 Objective:
To maintain the secondary coolant in such a condition as to minimize corrosion and/or scale buildup on the heat exchanger tubes and to detect a primary to secondary system leak.
Specification:
1.
The secondary coolant water pH shall be maintained between 5.5 and 9.0.
2.
The sample will be analyzed for the presence of sodium-24.
. Bases:
The facility has maintained the above coolant water conditions for many years based on consultant recommendations and have good results in maintaining heat exchanger tube and shell cleanliness.
)
Radioactivity in the secondary system would indicate a leak and therefore samples are analyzed for detectable concentrations of sodium-24.
3.4, 3.5, 3.6 Confinement and Emergency Exht.ust System and Emergency Power Applicability:
This specification applies to the operation of the reactor confinement and emergency exhaust system which must be operable during reactor operation, fuel handling and any operation that could cause the spread of airbome radioactivity in the confinement area.
Objective:
To assure that the confinement and emergency exhaust system is capable of operation to mitigate the consequences of possible release of radioactive materials resulting from reactor operation, fuel movement and handling of radioactive material.
Specification:
The reactor shall not be operated unless the following equipment is operable and/or conditions met:
Page 22 Amendment 26
t TECilNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License R-95 Revision i Eauipment/ Condition Function Personnel access doors To maintain to reactor closed (except confinement for entrance and egress).
system integrity Roof hatch closed.
Truck door closed; Reactor Room fresh air To maintain intake valve and exhaust confm' ement ventilation valve to the system integrity stack are open; Initiation system for To initiate confinement isolation, system i.e. evacuation buttons operation and and alann horns; alert personnel Emergency cleanup exhaust To maintain a system; negative building pressure without unloading any large fraction of possible airborne activity Emergency generator.
To insure power source to clean up system and other designated systems
- Bases:
The confinement system will be activated by depressing any one of five emergency evacuation buttons when an unsafe radiological situation develops as defined in facility operating and emergency procedures. In the unlikely event of a release of fission products or other airborne radioactivity, the confinement isolation initiation system will secure the normal ventilation exhaust fan, will bypass the normal ventilation supply up the stack, and will close the normal inlet and exhaust valves. In confinement, the emergency exhaust system will tend to maintain a negative building pressure with a combination of controls intended to prevent unloading any large fraction of airborne activity. The emergency exhaust purges the building air through charcoal and absolute filters and controls the discharge which is diluted by supply air through a 115 foot stack.
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TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License R-95
. Revision 1 3.7 Radiation Monitoring Systems and Effluena 3.7.1 Radiation Monitoring Systems Applicability:
This specification applies to the availability of radiation monitoring equipment which must be operable during reactor operation, fuel movement and handling of radioactive materials in I
the reactor building.
Objective:
To assure that radiation monitoring equipment is available for evaluation of radiation conditions and that the release of airborne radioactive material is rnaintained below the lir-;ts established in 10CFR20.
Specification:
1.
When the reactor is operating, gaseous and particulate sampling of the stack effluent shall be monitored by a stack monitor with a readout in the control room.
The particulate activity monitor and the gaseous activity monitor for the facility exhaust stack shall be operating. If either unit is out of service for more than one shift (6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />), either the reactor shall be shut down or the unit shall be replaced by one of comparable monitoring capability.
2.
When the reactor is operating, at least one constant air monitoring unit (Table 3.2.11) located in the confinement building shall be operating. Temporary shutdown of this unit shall be limited as in 3.7.1 above.
3.
Che reactor shall not be continuously
- operated without a minimum of one area radiation monitor (Table 3.2.8) on the experimental level of the reactor building and one area monitor (Table 3.2.6) over the reactor pool (reactor bridge) operating and capable of warning personnel of high radiation levels.
- In order to continue operation of the reactor, replacement of an j
inoperative monitor must be made within 15 minutes of recognition of failure, except that the reactor may be operated in Page 24 Amendment 26
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f TECliNICAL SPECIFICATIGNS Rhode Island Nuclear Science Center Docket 50-193, License R-95 Revision 1 a steady-state power mode if the moaitor is replaced wi h ponable gamma-sensitive instruments having their own alarm.
1 Bases:
A continuing evaluation of the radiation levels within the reactor building I
will be made to assure the safety of personnel. This is accomplished by the monitoring systems described in Table 3.2.
1 3.7.2 Effluents a.
Airborne Effluents l
Applicability:
This specification applies to the monitoring or airborne effluents from the Rhode Island Nuclear Science Center (RINSC).
Objective:
To assure that containment integrity is maintained during reactor operation and that the release of airborne radioactive material from the RINSC is maintained below the limits established in 10CFR20.
Specification:
1.
The concentration of radioactive mate-ials in the effluent l
released from the facility exhaust stacks shall not exceed 105 times the concentrations e ecified in 10CFR20, Appendix B, Table II, when averaged over time periods permitted by 10CFR20.
Bases:
The limits established in specification 3.7.2 incorporate a dilution factor of 4x104 for effluent concentrations released through the exhaust stacks. This dilution factor is based on a dispersion factor (X/Q = 10-5 sec/M3) calculated from actual meteorological data which is determined using the highest frequency of wind in any sector. Because of the use of the most conservative measured values of wind directional frequency and dispersion factors, this dilution factor will assure that concentrations of radioactive material in unrestricted areas around the Rhode Island Nuclear Science Center will be far below the limits of 10CFR20. (Refer to letter dated April 16, 1963 sent to the NRC in connection with license questions.) This dilution factor is used for calculating maximum ground Page 25 Amendment 26
r-f TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License R-95 Revision I concentration of noble gases down wind vs. exhaust stack effluent concentrations. The SAR contains calculations for doses from the iodine at the 48 meter distance.
b.
Liquid Effluents Applicability:
This specification appiks to the monitoring of radioactive liquid effluents from the Rhode Island Nuclear Science Center.
Objectives:
The objective is to assure that exposure to the public resulting from the release of liquid effluents will be within the regulatory limits and consistent with as low as reasonably achievable requirements.
Specification:
The liquid waste retention tank discharge shall be batch sampled and the l
gross activity per unit volume determined before release. All off-site releases shall be directed into the municipal sewer system.
Bas s:
All radioactive liquid and solid wastes disposed of off-site shall be within the limits established by 10CFR20 or shall be removed from the site by a ccmmerciallicensed organization.
3.8 Limitations on Experiments Applicability:
]
This specification applies to experiments to be installed in the reactor and associated experimental facilities.
Objectives:
To prevent damage to the reactor or release of radioactive materials in excess of 10CFR20.
Specification:
The reactor shall not be operated unless the following conditions goveming experiments exist; Page 26 Amendment 26
l TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193. License R-95 Revision i 1.
All materials to be irradiated shall be either corrosion resistant or encapsulated within corrosion resistant containers to prevent interaction with reactor components or pool water. Corrosive materials shall be doubly encapsulated.
2.
Irradiation containers to be used in the reactor, in which a static pressure i
will exist or in which a pressure buildup is predicted, shall be designed and tested for a pressure exceeding the maximum expected by a factor of 2.
3.
Fissionable materials shall have total iodine and strontium inventory less than that allowed by the facility by-product license.
4.
Explosive materials, in any quantity, shall not be allowed in the reactor pool or experimental facilities.
5.
All experiments shall be designed against failure from internal and external heating at the true values associated with the LSSS for reactor power level and other process parameters.
6.
Experimental apparatus, material or equipment to be irradiated shall be positioned so as not to cause shadowing of the nuclear instrumentation, interference with control blades, or other perturbations which may interfere with safe operation of the reactor.
7.
Cryogenic liquids shall not be used in any experiment within the reactor pool without approval from the Nuclear Regulatory Commission.
8.
No highly water reactive materials shall be used in an experiment in the reactor pool.
9.
No experiment should be performed unless the material content (with the exception of trace constituents)is known.
10.
If a capsule fails and releases material which could damage the reactor fuel or structure by corrosion or other means, removal and physical inspection shall be performed to determine the consequences and need for corrective action. The results of the inspection and any corrective action taken shall be reviewed by the Director, or his designated alternate, and determined to be sar.isfactory before operation of the reactor is resumed.
Experimental materials, except fuel materials, which could off-gas, sublime, volatilize, or produce aerosols under: (1) normal operating Page 27 Amendment 26
TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License R 95 Revision I conditions of the experiment or reactor, (2) credible accident conditions in the reactor, and (3) possible accident conditions in the experiment shall be limited in activity such that: if 100% of the gaseous activity or radioactive aerosols produced escaped to the reactor room or the atmosphere, the airbome concentration of radioactivity averaged over a year.would not exceed the occupationel limits for maximum permissible concentration.
In calculations pursuant to the above, the following assumptions shall be used: (1) If the effluent from an experimental facility exhausts through ductwork which closes automatically on high radiation level, at least 10% of the gaseous activity or aerosols produced will escape. (2) If the effluent from an experimental facility exhausts through a filter installatior. designed for greater than 99% efficiency for 0.3 micron particles, at least 10% of these vapors can escape. (3) For materials whose boiling point is above 550C and where vapors formed by boiling this material can escape only through an undisturbed column of water above the core, at least 10% of these vapors can escape. (4) Limits for maximum permissible concentrations are specified in the appropriate section of 10CFR20.
Bases:
Specifications 1 through 5, 8 and 9 are intended to reduce the likelihood of damage to reactor components and/or radioactivity releases resulting from experiment failure and, along with the reactivity restriction of peninent specification in 3.1, serve as a guide for the review and approval of new and untried experiments by the operations staff as well as the Nuclear and Radiation Safety Subcommittee.
Specifications 3 and 4 are self explanatory, i
Specification 6 assures that no physical or nuclear interferences compromise the safe operation of the reactor by, for example, tilting the flux in a way that could effect the peaking factor used in the Safety Analysis.
Specification 7 insures NRC review of experiments containing or using cryogenic materials. Cryogenic liquids present structural and explosive problems which enhance the potential of an experiment failure.
l Specification 10 is self explanatory.
3.9.
Reactor Core Components l
a.
Beryllium Reflectors Page 28 Amendment 26
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TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License R-95 Revision 1 Applicability:
l This specification applies to neutron flux damage to the standard and plug type beryllium reflectors.
l Objective:
' To prevent physical damage to the beryllium reflectors in the core from
. accumulated neutron flux exposure.
l
~ Specification:
l-1.
The maximum accumulated neutron flux shall be Ix1022 neutrons /cm2, l
l Bases:
L The RINSC.SAR (Part A Section VIU) has addressed this limit as a conservative limit.
b.
LEU Fuel 1
Applicability:
This specification applies to the physical condition of the fuel elements.
i Objective:
To prevent operation with damaged fuel elements.
Specification:
Fuel elements to be inspected for physical defects and reactor core box fit in accordance with manufactured specifications.
{
j Bases:
1 The RINSC inspects and tests each fuel element for reactor core box fit in accordance with written procedures to assure operation with fuel elements that are not damaged and meet specifications.
Page 29 Amendment 26
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TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193. License R-95 Revision 1 4.0 SURVEILLANCE REQUIREMENTS Surveillance tests for Reactivity Limits (4.1), Reactor Safety System (4.2),
Surveillance of Experiments (4.8) and Reactor Components (4.9) may be deferred for periods of reactor shutdown providing they are perfonned prior to restart (ANS 15.1,4.1). Surveillance tests for the following will be performed as stated in the appropriate sections:
Water Coolant System (4.3)
Confinement and Emergency Exhaust System (4.4,4.5,4.6)
Radiation Monitoring System and Effluents (4.7) 4.1 Reactivity Limits Applicability:
This specification applies to the surveillance requirements for reactivity limits.
Objective-h To assure that the reactivity limits of Specification 3.1 are not exceeded.
Specification:
l.
Shim safety blade reactivity worths and insertion rates shall be measured:
a.
annually; b.
whenever the core is changed from the startup core to the three other cores as analyzed and specified in the SAR (Part A,Section V).
2.
Shim safety blades shall be visually inspected and checked for swelling at least annually.
3.
The reactivity worth of all experiments shall be measured prior to the experiment's initial use.
Bases:
Specification 4.1.1 will assure that shim safety blade reactivity worths are not degraded or changed by core arrangements.
Page 30 Amendment 26
TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193. License R-95 Revision 1 Shim safety rod inspections are the single, largest source of radiation exposure to facility personnel. In order to minimize personnel radiation exposure and provide an inspection frequency that will detect early evidence of swelling and cracking, an annual inspection interval was selected for Specification 4.1.2.
The specified surveillance relating to the reactivity worth of experiments will assure that the reactor is not operated for extended periods before determining the j
reactivity worth of experiments. This specification also provides assurance that experiment reactivity worths do not increase beyond the established limits due to core configuration change.s.
4.2 Reactor Safety System Applicability:
This specification applies to the surveillance of the reactor safety system.
Objective:
To assure that the reactor safety system is operable as required by Specification
3.2. Specification
1.
A channel test of the neutron flux level safety channels and period safety channel shall be performed:
Prior to each reactor startup following a period when the reactor i
a.
was secured; b.
After a channel has been repaired or deenergized.
2.
A channel calibration of the _ safety channels listed in Table 3.1, which can be calibrated, shall be performed annually.
3.
The radiation monitoring system required in Table 3.2 shall be operable prior to every reactor startup for which safety system channel tests are required as in 4.2.1.
If the system has been repaired, the system shall be operable prior to use.
4.
Shim safety blade release-drop time shall be measured annually.
5.
Shim safety rod release-drop time shall be measured whenever the shim safety rod's core location is changed or whenever maintenance is performed which could effect the rod's drop time. (Specification 3.2.3)
Page 31 Amendment 26
TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License R-95 Revision 1 j
6.
Shutdown Margin (Specification 3.1.1) i The shutdown margin shall be determined annually.
It shall be determined when a new core is configured as described in the SAR (Part A,Section V). The determination will be made in accordance with operating procedures.
7.
Excess Reactivity (Specification 3.1.2)
The excess reactivity shall be determined annually.
It shall be determined when a new core is configured as described in the SAR (Part A,Section V). The determination will be made in accordance with operating procedures.
8.
Reactivity Insertion Rate (Specification 3.2.4)
The reactivity insertion rate shall be measured annually. It shall be determined when a new core is configured as described in the SAR (Part A,Section V). The determination will be made in accordance with written procedures.
Bases:
Prestartup tests of the safety system channels assure their operability. Annual calibration detects any long term drift that is not detected by normal
'intercomparison of channels. The channel operability check of the neutron flux level channels assures that the detectors are properly adjusted to accurately monitor the parameter they are measuring.
Radiation monitors are checked for proper operation in Specification 4.2.3.
Calibration and setpoint verification involve use of a calibration source and significant personnel radiation exposure. It is determined that annual calibration of radiation monitors is adequate since they displayed excellent stability over many years of operation.
The measured release-drop times of the shim safety blades have been consistent over many years. Annual check of these parameters is considered adequate to detect any deterioration which could change the release-drop time. Binding or rubbing caused by rod misalignment could result from maintenance; therefore, I
release drop times will be checked after such maintenance.
4.3 Water Coolant System a.
Primary Coolant System Page 32 Amendment 26 i
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TECHNICAL SPECIFICATIONS 4
Rhode Island Nuclear Science Center l
Docket 50-193. License R 95 i
Revision I Applicability:
This specification applies to the surveillance of the primary coolant system.
Objective:
To assure high quality pool water and to detect the deterioration of components in the primary coolant loop.
Specification:
1.
The pH of the primary coolant shall be measured weekly, i
2.
The resistivity of the primary coolant shall be measured weekly.
3.
The radioactivity of the primary coolant shall be analyzed weekly for gross activity and quarterly for isotopic activity.
4.
Pool water level scram switch shall be checked for operation monthly.
5.
Pool inspections shall be made annually in accordance with operating procedures.
6.
Pool level shall be visually inspected daily in accordance with operating procedures.
Bases:
Regular surveillance of pool water quality and radioactivity provides assurance that pH and resistivity changes that could accelerate the corrosion of the primary system components would be detected before significant damage would occur, and that the presence of leaking fuel elements in the reactor is detected.
The low pool level switch is checked for operation monthly'. Upon a one inch pool level drop, the automatic fill begins; upon a two inch drop, the i
reactor scrams (if operating) and a local and remote alarm sounds. The i
remote alarm is continuously monitored offsite.
Annual pool system inspections are made to provide assurance that other cooling system components (eg. gate valves, gasketing etc.) are functioning properly.
Page 33 Amendrnent 26 l
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TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License R-95 i
I Revision I b.
Secondary Coolant System l
Applicability:
1 This specification applies to the surveillance of the secondary coolant water.
Objective:
To assure the conditions of the coolant meet specification 3.3.(b) and to detect a primary to secondary water leak.
i i
Specification:
1.
The pH shall be measured weekly during reactor operation.
2.
A sample shall be drawn weekly and analyzed for sodium-24 activity, during reactor operation.
Bases:
Proper secondary coolant conditions are obtained by blowdown and makeup water systems which maintain the proper water quality pH.
Radioactive concentrations are measured in accordance with written procedures.
4.4, 4.5, 4.6 RINSC Confinement and Emergency Exhaust System 1
Applicability:
This specification applies to the surveillance of the fncility openings and dampers.
Objective:
To assure that the condition of the closure devices for the building openings are in satisfactory condition ard to assure their ability to provide adequate confinement of any airborne radioactivity released into the building.
l l
Specification:
l 1.
The confinement and emerg:ncy exhaust system described in Specification 3.4 shall be tested weekly for operability and after any maintenance that could affect gsystem operability.
The system operation is as described in the operating procedures and as herein l
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s TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License R-95 Revision I discussed. The building cleanup system shall be activated by pressing an evacuation button, then automatically:
a.
the evacuation horn sounds 1
b.
the building ventilation blowers deenergize (air conditioner, exhaust blower, off gas blower, rabbit system blower, heating system blowers);
c.
the building ventilation dampers close (air intake and exhaust system);
d.
the cleanup system blower (through the scrubber filter) and air dilution blower (chem lab) are energized e.
the negative differential pressure between the inside and outside of the building is at least 0.5 inches of water. This is determined by reading the pressure gauge in the control room; f.
the exhaust rate through the emergency cleanup system shall not be more than 1500 CFM coming from the reactor building and passing through the scrubber filters.
Dilution air will be provided by a separate blower from an uncontaminated source.
2.
The condition of the following equipment shall be inspected in accordance with written operating procedures every 6 months.
Building ventilation blowers and dampers (including solenoid a.
valves, pressure switches, piping, etc.);
2 b.
Personnel access and reactor room overhead doors.
3.
The testing and maintenance of the emergency generator will be performed in accordance with the RINSC operating procedures and manufacturer recommendation.
4.
The efficiency test for the charcoal filter shall be tested annually as specified in the operating procedures.
. Bases:
The weekly check of the confinement system provides assurance that the automatic function will be actuated when confinement isolation is required. The semiannual inspection of valves and doors will provide assurance that the closures will perform their function of limiting leakage through these Page 35 Amendment 26
TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License R-95 Revision I openings in the event of a release of airborne activity into the building.
The testing of the emergency generator assures reliable response and operation.
The load testing assures proper handling of expected system loads.
The emergency generator system has maintenance performed in accordance with manufacturer recommendations.
4.7 Radiation Monitoring Systems and Effluents a.
Airborne Effluents Applicability:
This specification applies to the surveillance of the monitoring equipment used to measure airborne radioactivity.
Objective:
The objective is to assure that accurate assessment of airborne effluents can be made.
Specification:
1.
The particulate air monitors Nll be calibrated annually.
2.
The gaseous activity monitor shall be calibrated annually.
3.
A channel check of the stack monitor and the main floor monitor shall be performed daily when the reactor is in operation.
Bases:
Experience with the electronic reliability and calibration stability of the units used by the Rhode Island Nuclear Science Center Reactor demonstrates that the above periods are reasonable surveillance frequencies.
b.
Liquid Effluents Applicability:
This specification applies to the surveillance of the monitoring equipment used to measure the radioactivity in liquid effluents.
Page 36 Amendment 26 i
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TECHNICAL SPECIFICATIONS L
Rhode Island Nuclear Science Center i
Docket 50-193, License R-95 Revision 1 Objective:
l The objective is to assure that accurate assessment ofliquid effluents can l
be made.
Specification:
1.
The monitoring equipment used to measure the radioactive concentrations in the waste retention tanks shall be calibrated annually.
2.
. The contents of every tank released shall be sampled and evaluated for radioactive concentrations 'and pH prior to its release.
Bases:
l Experience with the electronic reliability and calibration stability of the units used by the Rhode Island Nuclear Science Center Reactor demonstrates that the above periods are reasonable surveillance frequencies.
4.8 Surveillance of Experiments Applicability:
This specification applies to the surveillance of experiments and the limitations on experiments as described in Technical Specification 3.8.
Objective:
To assure that the experiments and their limitations are reviewed with respect to 10CFR50.59 for reactor operation and personnel safety and prevent release of radioactive materials in excess of 10CFR20.
Specification:
Experiments shall be reviewed, approved and properly installed and operational in accordance with written operating procedures.
I Experiments in progress shall undergo a review annually.
Bases.
Page 37 Amendment 26
F TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License R.95 Revision 1 Review of the experiments using the appropriate'LCO's and the Administrative Controls assures that.- the insertion of experiments will not negate the consideration implicit in the Safety Limits.
14.9 Reactor Core Components
' Applicability;
- This specification. applies to the surveillance requirements for reactor core components affecting reactor power.
a.
Beryllium Reflectors Applicability:
i This specification applies to the surveillance of beryllium lifetime for the l
standard and plug type beryllium reflectors.
. Objective:
To' prevent physical damage to the beryllium reflectors in the core from accumulated neutron flux exposure.
Specification:
- The maximum accumulated neutron flux shall be lx1022 neutrons /cm2, The exposure shall be' determined annually in accordance with the operating procedures.
Inspections and core fit shall be conducted annually.
Bases:
' The RINSC SAR'(Part A Section VIII) has addressed this limit as a
- conservative limit.
(Annual inspections and core box fit as well as calculated total exposure serve as a method to monitor the beryllium lifetime.)
b.-
' LEU Fuel Elements
' Applicability:
This specification applies to surveillance of LEU fuel elements.
Objective:
Page 38 Amendment 26
TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License R-95 Revision i To prevent operation with damaged fuel elements and verify the physical
. condition of the fuel element.
Specification:
The fuel elements shall be visually examined'and functionally fit into the core grid box annually.
Bases:
Fuel' elements are initially inspected for manufactured specifications and then inserted into the grid box in-accordance with QA/QC program requirements for functional fit. Core reloading is performed in accordance 3
- with operating procedures. Routine fuel movements are logged and visual inspections are conducted during fuel movements. ~ Pool sampling also is used to detect a ruptured element (Tech. Spec. 4.3.3). The fission density limit for this reactor cannot be exceeded (reference SAR, Part A,Section VI). Burnup calculations are made quarterly (4.9.1).
1 Page 39 Amendment 26
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TECHNICAL SPECIRCATIONS Rhode Island Nuclear Science Center Docket 50-193, License R-95 Revision 1 5.0 DESIGN FEATURES The basic design features of the facility are described in " General Electric's Operation and Maintenance Manual-GEI-77793", Oct.1962, also in the " Safety Analysis Report for the Low Enriched Fuel Conversion of the Rhode Island Nuclear Science Center Research Reactor", Rev.1,1992. These documents are on file at the Science Center. A general description of the imponant components is included in the following sections.
5.1 Description The reactor is located at the Rhode Island Nuclear Science Center on 3 acres of a 27-acre former military reservation, originally called Fort Kearney and now called the Narragansett. Bay Campus of the University of Rhode Island. The 27-acre reservation is controlled by the State of Rhode Island through the University of Rhode Island. The reservation is in the Town of Narragansett, Rhode Island, on the west shore of Narragansett Bay, approximately 22 miles south of Providence, Rhode Island, approximately six miles north of the entrance of the Bay from the Atlantic Ocean. The Rhode Island Nuclear Science Center and various buildings used for research, edu.ation and training purposes are located on this 27-acre campus.
5.2 Reactor Fuel The fuel assemblies shall be of the MTR type, consisting of plates containing uranium silicide fuel enriched to less than 20% in the isotope U-235 clad with aluminum. Each fuel element will contain 22 plates for a total of 275 grams of U-235 per element.
5.3 Reactor Core The reactor core consists of a 9 x 7 array of 3 inch square modules with the 4 corners occupied by posts. The reference core for these technical specifications consists of 14 standard LEU fuel elements arranged symmetrically within 4 safety control blades as shown in Figure 4 of the SAR (Revision 1,Section V, Dec.1992) as approved by the NRC in the conversion order (le;ter of March 17,1992).
5.4 Reactor Building The reactor shall be housed in a building capable of meeting the following functional requirements:
Page 40 Amendment 26
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TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License R-95
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Revision 1 In the event of an accident which could involve the release of
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radioactive material, the confinement building air shall be exhausted through a clean-up system and stack creating a flow of air into the building with a negative differential pressure between thc cuilding and the outside atmosphere. The building shall be gas tight in the sense that a negative differential pressure can be maintained dynamically with all gas leaks occurring inward. The confinement and cleanup systems shall become operative when a building evacuation button is
. pressed. This action shall: (1) turn off all ventilation fans and the air 4
conditioner system and (2) close the dampers on the ventilation intake
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and exhaust, other than those which are a part of the clean-up systein.
No further action shall be required to establish copinement and place
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the clean-up system in operation.
suxiliary elenrical power system shall be provided at the site to are the avali bility of power to operate the clean-up system.
The reactor building exhaust blower operates in conjunction with additional I
exhaust blower (s) which provide dilution air from non-reactor building sources.
Upon activation, the clean-up system shall exhaust air from the reactor building through a filter and a 115 foot high stack, creating a pressure less than atmospheric pressure. The clean-up filter shall contain a roughing filter, an absolute particulate filter, a charcoal filter for removing radioiodine and an absolute filter for removing charcoal dust which may be contaminated 'vith radiciodine. Each absolute filter cartridge shall be individually tested and certified by the manufacturer to have an efficiency of not less than 99.97% when tested with 0.3 micron diameter dioctylphthalate smoke. The minimum removal efficiency el the charcoal filters shall be 99%, based or. ORNL data and measurements perforrrad j
locally.
i i
Gases from the beam ports, thermal column, pneumatic system, and all other radioactive gas exhaust points shall be exhausted to the stack through a roughing and absolute filter system.
5.5 Fttei Storage All reactor fuel element storage facilities shall be designed in geometrical configuration where kerr is less than 0.8 under flooding with water. A maximum of four fuel elements will be stored in the fuel sale with no two elements in adjacent positions in the storage box. The adjacent row will be an empty box. Irradiated fuel is stored in the underwater storage racks as described in the SAR (Part A,Section XII).
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TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193. License R-95 Revision 1 6.0 ADMINISTRATIVE CONTROLS 6.1 Organization and Management 1.
The Rhode Island Atomic Energy Commission (RIACC) shall have the responsibility for the safe operation of the reactor.
The organization of RIAEC is shown in Figure 6-1. The RIAEC shall appoint a Director and a Nuclear and Radiation Safety Committee (NRSC) consisting of a minimum of seven members, as follows:
a.
The Director, RIAEC b.
The Assistant Director for Reactor Operauons c.
The Radiation Safety Officer d.
A qualified representative from the faculty of Brown University A qualified representative from the faculty of Providence e.
College f.
Two qualified representatives from the faculty of the University of Rhode Island i
A qualified alternate may serve in lieu of one of the above. The Director, Assistant Director and Radiation Safety Officer are not j
eligible for chairmanship of the Committee.
j 2.
^.. operator or senior operator licensed pursuant to 10CFR55 shall be present in the control room unless the reactor is secured as defined in these specifications. The minimum operating crew shall i
be two individuals.
3.
A licensed senior operator shall be on duty or readily available on call whenever the reactor is in operation.
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f TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License R-95 Revision 1 4.
In accordance with the emergency plan, a list of emergency personnel, management and offsite agencies is posted in the control room.
6.2 Qualifications of Personnel 1.
At the time of appointment to the position, the Director shall have a minimum of six years of nuclear experience. The Director shall have an advanced degree in one of the physical sciences or engineering, and be a licensed senior operator. The degree will fulfill four years of the six-year requirement.
2.
The Radiation Safety Officer shall have a master's degree in health physics or radiological health and three years of applied health i
physics experience in a program with radiation safety problems similar to those in the program to be managed.
3.
The reactor operators and senior operators shall be licensed in accordance with the provisions of 10CFR55.
4.
In the event of temporary vacancy in the position of Director or the Radiation Safety Officer, the functions of that position shall be assumed by qualified alternates appointed by the RIAEC.
6.3 Responsibilities of Personnel 1.
Director The Director shall have responsibility for all activities in the a.
reactor facility which may affect reactor operations or involve radiation hazards, including controlling the admission of personnel to the building.
I This responsibility shall encompass administrative control of all experiments being performed in. the facility including those of outside agen6es.
b.
It shall be the responsibility of the Director to insure that all proposed experiments, design modifications, or changes in operating and emergency procedures are performed in accordance with the license. Where uncertainty exists. the Director shall refer the decision to the NRSC.
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TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License R-95 Revision 1 2.
Senior Rutor Operators A licensed senior reactor operator pursuant to 10CFR55 shall a.
be assigned each shift and -be responsible for all activities during his shift which may affect reactor operation or involve radiation hazards. The reactor operators on duty shall be responsibis directly to the senior operator.
b.
The identity of and method for rapidly contacting the on-call senior reactor operator shall be known to the reactor operator on duty. The on-call senior reactor operator must be capable of being contacted by the duty reactor operator within ten minutes.
The senior reactor operator shall be present at the facility during initial rtartup and approach to power, recovery from an unplanned or unscheduled shutdown or significant reduction in power, and refueling. The name of the person serving as senior reactor operator as well as the time he assumes the duty sha'l I
be entered in the reactor log. When the senior operator is relieved, he shall tum the operation duties over to another licensed senior operator.
In such instances, the change of duty shall be logged and shall be definite, clear, and explicit. The senior reactor operator being relieved of his duty shall insure that all pertinent information is logged. The senior reactor operator assuming duty shall check the log for information or instructions.
3.
Reactor Operators The responsible senior reactor operator shall pursuant to a.
10CFR55 designate for his shift a licensed operator (hereafter called " operator") who shall have primary responsibility under the senior reactor operator for the operation of the reactor and all associated control and safety devices, the proper functioning of which is essential tw the safety of the reactor or personnel in the facility. The opes stor shall be responsible directly to the senior reactor operator.
I b.
Only one operator shall have the above duty at any given time.
i Each operator shall enter in the reactor log the date and time he assumed duty, t
I When operations are performed which may affect core c.
reactivity, a licensed operator shall be stationed in the control 1
room. When it is necessary for him to leave the control room l
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TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License R.95 Revision 1 during such an operation, he shall turn the reactor and the reactor controls over to a designated relief, who shall also be a licensed operator. La such instances, the change of duty shall be definite, clear, and explicit. The relief shall acknowledge his entry on duty by proper notation in the reactor log.
d.
The operator, under the senior reactor operator on duty, shall be responsible for the operation of the reactor according to the approved operating procedures.
The operator shall be authorized at any time to reduce tiie e.
power of the reactor or to scram the reactor without reference to higher authority, when in his judgment such action appears advisable or necessary for the safety of the reactor, related eqaipment, or personnel. Any person working on the reactor bridge shall be similarly authorized to scram the reactor by pressing a scram button located on the bridge.
l 4.
Radiation Safety Officer i
The Radiation Safety Officer shall be responsible for assuring that adequate radiation monitoring and control are in effect to prevent undue exposure of individuals to radiation.
6.4 Review and Audit 1.
The NRSC shall review reactor operations to assure that the facility is operated in a manner consistent with public safety and within the terms of the facility license.
2.
The responsibilities of the NRSC include, but are not limited to, the following:
i Audit of operating, and emergency procedures and records, a.
b.
Review and audit of proposed tests and experiments utilizing the reactor facilities.
Review and audit of proposed changes to the facility systems or c.
equipa.ent, procedures, and operations.
d.
Determination of whether a proposed change, test, or experiment would constitute an unreviewed safety question or which may require a change to the Technical Specifications or facility license.
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TEtnNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193. License R-95 Revision i Review of all violations of the Technical Specifications and e.
Nuclear Regulatory Commission Regulations, and significent violations of internal rules or procedures, with recommendations for corrective action to prevent recurrence.
f.
Review of the qudifications and competency of the operating organization to assme retention of staff quality.
g.
Review changes to the NRSC charter.
h.
Review, at least annually, the radiation sa ',ty aspects of the r
facility.
3.
The NRSC shall have a written charter defining su::h matters as the authority of the Committee, the subjects within its purview, and other such administrative provisions as are required for effective l
functioning of the Committee. Minutes of all meetings of the Committee shall be kept.
All minutes of the previous Reactor Utilization Committee shall be retained for the life of the facility.
4.
A quorum of the NRSC shall consist of not less than four (4) members and shall include the Radiation Safety Officer or designee, the Director or the Assistant Director for Operations and the Chairman or designee.
5.
The NRSC shall meet at least annually.
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6.5 Operating Procedures l
Written procedures, reviewed and approved by the NRSC, shall be used for items 1-9 listed below. The procedures shall be adequate to assure the i
safe operation of the reactor, but should not preclude the use of independent judgment and action should 'Se situation require such.
1 1.
Startup, operation and shutdown of the reactor; j
l 2.
Installation and removal of fuel elements, control blades and incere devices where necessary i
3
' Ma'.ntenance procedures which could have an effect on reactor
~
safety; 4.
Periodic surveillance of reactor insuunematMn and safety systems, I
area monitors, and continuous air moniter:.;
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TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License R-9$
Revision 1 5.
Implementation of the physical Security Plan and Emergency Plan; 6.
Radiation control procedures; 7.
Receipt, inspection, and storage of new fuel elements;
-8.
Storage and shipment ofirradiated fuel elements.
9.
Experiment review on a case-by-case basis assuring that section 3.8.3(2) of ANSI /ANS 15.1 is satisfied. Operational approval shall be by written approval by a licensed senior operator.
Written
]
procedures should be established and supervision of the installation of such experiments shall be defined and exercised.
Substantive changes to the above procedures shall be made only with the approval of the NRSC. Temporary changes to the procedures that do not j
change their original intent may be made by a Senior Operator. Temporary changes to procedures shall be documented and subsequently reviewed by the NRSC Subconumttee.
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6.6 Action to be Taken in the Event of a Reportable Occurrence In the event of a reportable occurrence:
1.
The Senior Reactor Operator shall be notified promptly and corrective action shall be taken immediately to place the facility in a safe condition until the cause of the reportable occurrence is determined and corrected 2.
The Director shall report the occurrence to the NRSC. The report shall include an analysis of the cause of the occurrence, corrective actions taken, and recommendations for appropriate action to prevent or reduce the probability of a repetition of the occurrence.
3.
The NRSC shall review the report and the corrective actions taken.
4.
Notification shall be made to the NRC in accordance with Paragraph 6.8 of these specifications.
6.7 Action to be Taken in the Event a Safety Limit is Exceeded In the event a Safety Limit has been exceeded:
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TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193. License R-95 Revision i 1.
The reactor will be shut down and reactor operations will not be resumed until authorization is obtained from the NRC.
2.
Immediate notification shall be made to the NRC in accordance with paragraph 6.8 of these specifications and to the Director.
3.
A prompt report shall be prepared by the Senior Reactor Operator.
The report shall include a cc splete analysis of the causes of the event and the extent o; possible damage together with recommendations to prevent or reduce the probability of recurrence.
This report shall be submitted to the NRSC for review and appropriate action, and a suitable similar report shall be submitted to the NRC in accordance'"ith Paragraph 6.8 of these specifications and.in support of a request for authorization for resumption of operations.
6.8 Reporting Requirements In addition to the requirements 'of applicable regulations, all written reports sha!! te sent to the U. S. Nuclear Regulatory Commission, Attn:
Document Contrcl Desk, Washington, DC 20555, with a copy to the Region I Administrator. The written reports include the following:
1.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, a report by telephone through the NRC Operations Center, Washington, DC and the NRC Region 1:
Any accidental release of radioactivity to unrestricted areas a.
above permissible limits, whether or not the release resulted in property damage, personal injury or exposure.
b.
Any significant variation of measured values from a corresponding predicted or peviously measured value of safety related operating characteristics occurring during operation of the reactor.
c.
Any reportable occurrences as defined in Paragraph 1.25 of these specifications.
d.
Any violation of a Safety Limit.
Discovery of any substantial variance from performance c.
specifications contained in the technical specifications and safety analysis.
2.
The written report shall be sent within 14 days. The report shall:
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TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193, License R-95 Revision 1
~. Describe, analyze, and evahiate safety implications;
- a. -
' b.. ~ Outline the measures taken to assure that the cause of the condition is determined; Indicate the corrective action taken, including any changes c.
made to the procedures and to the quality assurance program, :o prevent repetition of the occurrence'and of similar occurrences involving similar components or systems; d.
Evaluate the safety. implication of the incident in light of the cumulative experience obtained from the record of previous
~ failure and malfunctions of similar systems and components.
3.
- Unusual Events k written report ' hall be forwarded within thirty (3'0) days in the s
event of:
a.
Discovery of any substantial errors in the transient or accident analyses or in the methods used for such analyses, as described in the safety analysis or in the bases for the technical
_ specifications; Discovery of any condition involving a possible single failure which, for a system designed against assumed failure, ccr Id result in a loss of the capability of the system to perform its safety function;
' b.
Permanent changes in the facility organization involving the Director or Assistant Director.
4.
- An. annual report shall be submitted in writing within 60 days ifollowing the 30th of June of each year. The report shall include the following information:
. a.
Tabulation showing the energy generated' by the reactor (in megawatt days), the number of hours the reactor was critical, and the cumulative total energy output since initial criticality.
b.
The number of emergency shutdowns and inadvenent scrams, including the reasons.
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TECHNICAL SPECIFICATIONS Rhode Island Nuclear Science Center Docket 50-193. License R-95 Revision I
. Discussion of the major maintenance operations performed c.
during the period, including the effect, if any, on the safe operation of the reactor, and the reasons for any corrective-maintenance required.
d.
A description of each change to the facility or procedures, tests, and experiments carried out under the conditions of Section 50.59 of 10CFR50 including a summary of the safety evaluation of each.
A description of any environmental surveys performed outside e.
the facility, f.
A summary of annual radiation exposures in excess of 500 mrem received by facility personnel, including the dates and times of significant exposures.
g.
A summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the licensee as measured at or prior to the point of such release or ' charge.
6.9 Plant Operating RecorIs In addition to the requirements of apphcable regulations and in no way substituting therefore, records and logs of the following items, as minimum, shall be kept in a manner convenient for review and shall be retained as indicted:
1.
Records to be retained for a period of at least five years:
a.
Reactor operations; b.
Principal maintenance activities; Experiments performed including aspects of the experiments c.
which could affect the safety of reactor operation or have radiological safety implications; d.
Reportable occurrences; e.
Equipment and component surveillance activities; f.
Facility radiation monitoring surveys; Page51 Amendment 26 I
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TECHNICAL SPECIFICATIONS Rhode bland Nuclear Science Center Docket 50-193, License R-95 Revision I g.
Fuelinventories and transfers; and h.
Changes to procedures systems, components, and equipment.
- 2.
' Records to be retained for the life of the facility:
a.
. Gaseous and liquid radioactive ' effluents released to the environs; b.
Off-site' environmental monitoring surveys; c.
Personnel radiation exposures; d.
' Updated, "as-built" drawings of the facility; and e.
' Minutes of the NRSC- (and previous Reactor Utilization Committee) meetings.
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