ML19345G656
| ML19345G656 | |
| Person / Time | |
|---|---|
| Site: | University of Buffalo |
| Issue date: | 04/03/1981 |
| From: | NEW YORK, STATE UNIV. OF, BUFFALO, NY |
| To: | |
| Shared Package | |
| ML19345G655 | List: |
| References | |
| NUDOCS 8104080579 | |
| Download: ML19345G656 (83) | |
Text
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O Appendix A For Facility Lic 2nse No. R-77 Technical Specifications For the Nuclear Science and Technology F cility Reactor of the State University of New York at Buffalo Docket No. 50-57 As Revised and Approved in 1981 l
1810 4 0 8 0M 4 ~
i
INDEX SECTION CONTENTS 1.0 Definitions 2.0 Safety Limits and Limiting Safety System Settings 2.1 Safety Limits 2.1.1 Safety Limits in the Forced Convection Mode 2.1.2 Safety Limits in the Natural Convection Mode 2.2 Limiting Safety System Settings 2.2.1 Limiting Safety S stem Settings in the Forced Convection Mode f
2.2.2 Limiting Safety System Settings in the Natural Convection Mode 3.0 Limiting Conditions for Operation 3.1 Reactivity Limits 3.2 Reactor Safety Systems 2.3 Radiation Monitoring Systems 3.3.1 Fixed Area Monitors 3.3.2 Effluent Monitors and Primary Coolant Monitor 3.4 Engineered Safety Features 3.4.1 Reactor Containment Building 3.5 Primary Coolant Conditions 3.6 Airborn Effluents 3.7 Liquid Effluents 3.8 Limitations on Experiments 3.9 Fuel Limitations d
- 4.0 Surveillance Requirements 4.1 Reactivity Limit Measurements 4.2 Reactor Safety System Tests 4.3 Radiation Monitoring Systems Tests 4.4 Engineered Safety Feature Tests (Containment) 4.5 Primary Coolant Condition Measurements 4.6 Liquid Effluent Monitoring 5.0 Design Features 5.1 Site Description 5.2 containment Building 5.3 Ventilation 5.4 Primary Coolant System 5.5 Fuel and Reflectors 5.6 Reactivity Control 5.7 Fuel Storage-
~ 5.7.1 Cold Fura Storage 5.7.1 Irradiated Fuel Storage 6.0 Administration 6.1.1 Management 6.1.2 Facility Staff Requirements 6.2 Review Functions 6.2.1 Nuclear Safety Committee 6.2.2.
Operating Committee 6.3 Facility Audit
e-0 INDEX SECTION CONTENTS 6.4 Actions to be Taken in the Event of a Reportabla Occurrence 6.5 Operating Procedures 6.6 Operating Records 6.7 Reporting Requirements 6.7.1 Financial Report 6.7.2 Annual Operating Report 6.7.3 Reportable Occurrence Reports 6.7.4 Unusual Event Report 6.7.5 Special Nuclear Materials Report I
L
o 1.0 Definitions 1.1 Safety Limit (SL) - Safety Limits are limits upon important process variables which are found to be necessary to reasonably protect the integrity of certrin physical barriers which guard against the uncontrolled release of radioactivity.
(10 CFR 50.36) 3 1.2 Limiting Safety System Settings (LSSS or LS ) - Limiting Saf ety System Settings are for automatic protective devices related t o those variables having significant safety functions. Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting shall be chosen such that automatic protective action will correct the abnormal situration before a safety limit is exceeded.
(10 CFR 50.36) 1.3 Limiting condition of Operation (LCO) - Limiting Conditions for Operation are the lowest functionalcapability or performance levels of equipment required for safe operation of the facility.
(10 CFR 50.36) 1.4 Measuring Channel - A Measuring Channel is the combination of sensor, amplifiers, and output devices which are used for the purpose of measuring the value of a process variable.
1.5 Safety Channel - A Safety Channel is a measuring channel in the reactor safety system.
1.6 Reactor Safety System - The-Reactor Safety System is that combination of safety channels and asscciated circuitry which forms the automatic protective system for the reactor or provides information which requires manual protective action to be initiated.
1.7 Operable - Operable means that a component or system is capable of performing its. intended function in its normal manner.
1.8 Operating - Operating means that a component or system is performing its intended function in its normal manner.
1.9 Channel Check - A 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 sate variable.
1.10 Channel Test - A Channel Test is the introduction of a signal into the channel to ';erify that it is operating.
1.11 Chauuel Calibration - A Channel Calibration is an adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the parameter which the chanvel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip and sht11 be deemed to include the Channel Test.
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4 1.12 True Value "he True Value of a process variable is its actual value at any instant.
1.13 Measured Value - The Measured Value of a process variable is the value of the variable as indicated by a measuring channel, j
1.14 Reactor Operations - Reactor Operation means that the control blades installed in the core are not fully inserted, that the console key is in the keyswitch, or manipulations are being aonducted in the pool that could affect core reactivity.
i 1.15 Reactor Shutdown - The reactor is considered shutdown if all control-safety blades are fully inserted, the console key is removed, and no menipulations are being conducted in the pool that could affect core reactivity. When the reactor is shutdown, an operator must be in the facility but act necessarily in the control room.
' 1.16 Unscheduled Shutdown - An Unscheduled Shutdown is defined as any i
unplanned shutdown of the reactor caused by actuation of the reactor
' safety system,. operator error, equipment malfunction, or a manual shutdcwn in response to conditions which could adversely affect safe operation, not to include shutdowns which occur during testing ar check-out operations.
- 1.17 Reactor Secured - The reactor in secured when a shutdown checklist has bean completed.
1.18 Reportable Occurrence - A Reportable Occurrence is any of the-P following:
a.
Operation in excess of a safety limit as set forth in section 2.1;
~
b.
Discovery of a safety system setting less conservativF than tt >
. limiting ' setting' established in ' the Technical Specifications; a
Operation in violation of. a limiting condition for operation c.
established.in the TechnicalESpecifications; o
.d.
A safety system component. malfunction'or other component or:
system malfunction which could,.or threatens to, render the
. safety. system incapable.of ' performing ita intended. saf ety l
. functions; e.
Release of fission products from a failed fuel. element;
- f. - An uncontrolled or. unplanned' release of. radioactive material from the' restricted. area.of the facility in excess of applicable limits; r
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. g.
An uncontrolled or unplanned release of radioactive material which results in concentrations of radioactive materials within the restricted area in excers of the limits specified in Appendix B, Table 1 of 10 CFR 20; h.
Conditions arising from natural or man-made events that affect or threatens to affect the safe operation of the facility; or 1.
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.
1.19 Experiment - An Experiment, as used herein, is any of the following:
.a.
An activity utilizing the reactor system or its components or the neutrons or radiation generated therein; b.
An evaluation or test of a reactor syst em operational, sur-veillance, or maintenance technique; c.
An experimental or testing activity which is conducted within the confinement or containment system of the reoctor; or d.
The material content of any of-the. foregoing, includ'_ag
-structural components, encapsulation or confining boundaries, and contained fluids or solids.
1.20-Experimental Facility - 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.
l1. 21-Explosive' Material - Explosive Material.is any solid-or liquid which is - categorized as a Severe,- Dangerous, or Very' Dangerous Explosive 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, "Identifi-cation System for Fire-Hazards of Materials," also enumerate! in
.the " Handbook for Laboratory Safety" 2nd Ed. -(1071) published by
'the Ch'emical Rubber Co; 1.22
- Potential Reactivity Worth of an Expeciment - The Potential
~
Reactivity Worth of an' Experiment.is the maximum absolute-
.value of the reactivity change that would occur as a result of-intended 'orf anticipated changes or credible malfunctions that alter equipment position or' configuration.
1.23 Static! Reactivity-Worth -LAs used herein,.the Static Reactivity
-Worth of an experiment is ' the absolute value of the reactivity change which is. measurable by calibrated control'or' regulating rod comparison methods between:two definedLterminal positions or-configurations of:the experiment. For removable experiments, the ~ terminal, positions are fully ' removed from the reactor and
~
. fully inserted or installed in the normal functioning or intended positicn.
1.24 Reactivity Limits - The Reactivity Limits are those limits imposed on reactor core excess reactivity. Quantities are referenced specifically to a cold core (8G - 1000F) with the effect of xenon poisoning on core activity accounted for if greater than or equal to 0.05% ak/k.
1.ie reactivity worth of samarium in the core will not be included in excess reactivity limits. The reference core condition will be known as the cold, xenon-free critical condition.
1.25 Permanent Experimental Facility - Those experimental f acilities that would require considercble effort and planning to remove or alter such as bean tubes, thermal column, etc.
1.26 Movable Experiment - A Movable Experiment is one which may be inserted, removed, or manipulated while the reactor is critical.
1.27 Removable Experiment - A Removable Experiment is any experiment, experimental facility, or component of an experiment, other than a permanently attached appurtenance to the reactor system, which can reasonably be anticipated to be moved one or more times during the life of the reactor.
1.28 Secured Experiment - Any experiment, experimental facility, or component of an experiment is deemed 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 exert sufficient. force on the experiment to overcome the expected ef fects of hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the experiment, or by forces which might arise as a result of credible malfunctions.
1.29 Unsecured Experiment - Any experiment, experimental facility, or component of an experiment is deemed to be Unsecured if it is not and when it is not secured as defined in 1.28 above.
1.30 Control-Safety Blede - A neutron absorbing blade used to control the reactivity of the core. A Control-Safety Blade is magnetically coupled to its drive unit allowing it to perform the function of a safety device when the magnet is denergized.
1.31 Control Blade - A neutron absorbing blade used to control core reactivity but is not magnetically roupled tc its drive unit.
1.32 Fast Scram - Fast Scram is a rapid reduction of the magnet holding current of.the Control Safety Blades until the blades fall by gravity into the reactor core.
1.33 Slow Scram - Slow Scram is the shutoff of electrical power to the units providing the magnet holding current with subsequent decay of the agnet holding-current until the bladcs fall by gravity into tb eactor core.
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I 1.34 Rundown - Rundown is the autocatic insertion of the Control-Safety Blades.
1.35 Readily Available on Call "Readily Available on Call" Shall cean the licensed senior operator shall insure that he is within a reasonable driving time (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) frca the reactor building. The licensed senior operator shall always keep the licensed operator informed of where he may be contacted.
T 1.36 Fuel Element - The smallest structurally discrete part of a reactor which has fuel as its principal constituent. Sane as fuel pin.
--l.37 Fuel Asse=bly - A grouping of fuel ele =ents which is not taken apart during the charging and discharging of a reactor cor e.
L k.G_.
. 2.0 Safety Limits and Limiting Safety System Settings 2.1 Safety Limits 2.1.1 Safety Limits in Forced Convection Mode Applicability This specification applies to pertinent pre;ess variables, the variation of which may potentially comprom 3e fuel or cladding integrity, during fe ced convection operation.
Objective The objective is to protect the reactor core from the loss of fuel or cladding integrity.
Specification a.
Core thermal power shall not exceed 3.3 FNr.
b.
The minimum depth of water from surface to top of core shall be 5.2 meters (17.0 feet).
c.
The minimum coolant flow rate shall be 63 LPS (1000 gal./ min.).
d.
The maximum bulk pool temperature shall be 60 C (1400F).
Basis The above specifications shall ensure fuel and cladding integrity.
2.1.2-Safety Limits in Natural Convection Mode Applicability This specification applies to core power in natural convection mode.
Objective The objective is to protect fuel and cladding integrity.
'Specificction a.
Cere power shall not exceed 500 KW in:the natural convection mode.
Basis The above specifications shall ensure fuel and cladding integrity.
2.2 Limiting Safer.y System Settings 2.2.1 Limiting Safety System _ Settings in Forced Convection Made Applicability This specification shall provide worst unper bounds on set pol'nts of safety systems, when the reactor is operating-in forced -
convection mode.-
. Objective The_ objective is to provide appropriate safety system actions, so that safety limits shall not be exceeded.
Specification a.
Control blade reverse shall occur at 2.2 MW.
b.
Control blade scram shall occur at 2.4 MW.
c.
Reactor scram shall occur if the depth of water above the core falls below 6.13 meters (20 feet).
d.
Reactor scram shall occur if coolant flow falls below 68 1ps (1080 GPM).
e.
Reactor scram snall occur if the bulk pool temperature reaches'52 C (126 F).
Basis By initiating corrective action based upon the above set _ point limits, safety limits will not be exceeded.
2.2.2 Limiting Safety System Settings in Natural Convection Mode Applicability This specification shall establish the maximum set point f7r overpower scram in natural convection mode.
Objective The objective'is to limit reactor power in natural convection mode.
Specification a.
Reactor scram shall occur if reactor power exceeds 250 KW while in-the natural convection mode.
Basis Reactor scram at 250 KW will prevent the reactor from exceeding the safety limit of 500 KW.
. 3.0 Limiting Conditions for Operation 3.1 Reactivity Limits Applicability This specification applies to the reactivity of the reacter core, the control rods, and experiments.
Objective The objective is to assure that the reactor can be controlled and shut down at all times and that the safety limits will not be exc eed ed.
Specification a.
The shutdown margin relative to the cold, Xenon-free critical condition shall be > the worth of the most reactive safety-control blade + 1% LK/K.
b.
The sum of the absolute values of all non-permanent experiments shall not exceed 3% LK/K.
c.
The worth of individual experiments shall be limited as follows:
EXPERIMENT MAX. WORTH Movable 1 0.3% LK/K Unsecured 1 0.6 Secured i 1.5 d.
The reactor shall be subcritical by at least 3% LK/K during any fuel manipulations or experiment manipulations involving reactivities >.3% LK/K.
Control - safety blades shall not be removed from the core e.
unless all four adjacent fuel assemblies have been removed from the grid plate.
Basis The shutdown margin required by specification 3.la assures that the reactor can be shutdown from any operating condition and will remain subcritical after cool down and xenon decay even if the control rod of the highest reactivity worth should be in the fully withdrawn position.
Specification 3.1.b limits the total worth of experiments to a value that can be compensated for by the control blades while maintaining the required shutdown margin.
Specification 3.1.c limits the worth of individual experiments.
The.3% LK/K for moveable experiments was chosen because it covers the needs of isotope production and experience has shown that a combination of temperature coefficient and operator action provide easy control of the change. The value of_.6% for unsecured experiments was chosen to-be slightly less than the value of S.
The value of 1.5% for secured experiments was chosen because this value of a step change was demonstrated to be safe during the pulse
' test program.
_9-Specification 3.1.d assures that the reactor will remain sub-critical during any manipulation of fuel or experiments.
Specification 3.1.e assures that the reactor will remain sub-critical even if any or all control blades are removed from the grid plate.
3.2 Reactor Safety Systems Applicability These specifications apply to the reactor safety system and other safety related instrumentation.
Obj ec tive The objective of this specification is to specify the lowest acceptable level of performance or the minimum number of operable components for the reactor safety system and other safety related instrumentation.
Specification The reactor shall not be operated unless:
a.
The reactor safety systems and safety related instrumentation are operating in accordance with Table 3.1 including the minimum number of channels and the indicated maximum or minimum setpoints.
b.
All control-safety blades are operable.
c.
The time from the initiation of a scram signal until the control-safety blades are fully inserted shall not exceed 650 msec.
Basis Specification 3.2.a assures that the reactor operator will have sufficient information at his disposal to operate the reactor within the safety limits. The specification also provides for sufficient limiting safety system settings to insure that saf ety limits will not be exceeded even under severe combinations of accident conditions, system failure, and operator error.
Specification 3.2.b assures that the primary saf ety and control devices are capable of performing this function when called upon to do so.
Specification 3.2.c assures that the magnet control circuits are functioning properly and that the control-safety blades can move freely in their guide.
Table 3.1 Required Instrumentation Modes in which Min. No.
- Instrument Channel Operating Function Set Point Required l
Log Count Rate
+*
1 Indication / Inhibit
< 2 cps;s9800 CPS Start-Up Linear Pwr.
+*
1 a
Indication All Log Pwr
+*
1 Indication All Period
+*
l' Indication All Pwr. Safety
+*
2' Indication / Scram 120%
All Pwr Safety
+*
1 Reverse 110%
All Manual Scram
+*
5 Scram All D.C. Door Open
+*
1 Scram DR < Full Closed All Flow
+*
1 Indication / Scram 68 1ps Forced Conv.
Flapper.Open
+*
1 Scram (250 kw Forced Conv.
i Water Level Low 1
Scram 6.13m Over Fuel All 1
Annunciation 6.43m All 1
Annunciation 6.74m All High Pool Temp.
+
1 Scram 520C Forced Conv.
Core Out. Temp.
+
1 Annunciation 52 C + AT Forced Conv.
8 3
Inhibit Start-Up g
Recorders Inoperative
.High Excess K
+
1 Annunciation P>100kw/CB<30%
All Conductivity
+
0 Annunciation 200 K Ohms None EPP Valve Open
+
0 Annunciation Valve Open None Demin. Temp.
+
0 Annunciation 41 C None Suc. Valve Closed
+
1 Disables Pri. Pemp V< Full Open Forced Cony.
Servo Deviation
+
1 Annunciation /XFer to
-+ 10%
Servo Control l
Manual Blade Pos. - Analog
+
1 of 2 Indication All Blade Pos. - Digital 1 of 2:
Indication All Nitrogen - 16
+
2 of 3 Indication Forced Conv.
Primary Temps
+
2 o f 3.-
Indication All Core Delta'T
+
2 of 3J Indication Forced Conv.
l a - Linear power channel and any recorder may be inoperable for short periods while operating.
- - Operability check required prior to operation.
+ - Test and/or calibration required four times / year.
l I
t l
l 3.3 Radiation Monitoring Systens 3.3.1 Fixed Area Monitors Applicability These specifications apply to the pernanently counted radiation monitors in the contain=ent building that have readouts and alar =s in the control roon.
Objective t purpose of this specification is to set a 21nicu= level of formance for the facility area nonitor system when the reactor p(rating.
Specification a.
The normal contingent of operating =onitors when the reactor is operating shall be as follows:
Neutron deck #1, 2, and 3 Feactor bridge Hot cell b.
The alar = set points for each sonitor shall be clearly stated in a facility operating procedure.
Monitors cay be removed from service fcr repair, replace =ent, c.
or calibration in accord with the following restrictions -
One of the three mocitors on the neutron deck nay be out of service at a given ti=e.
The hot cell monitor =ay be out of service for extended periods if the key to the cell is administratively restricted. The bridge monitor =ay be out of service for short periods only.
Any monitor =ay be te=porarily replaced by a portable unit that provides equivalent functions.
Basis Specification:3.3.1.a' assures that each monitor vill have a clearl, define' alana point - that cannot be altered without for=al reviev.
Specification 3.3.1.c provides for normal repair and calibratica of the area monitor system without shutting down the reactor.
.3.3.2 Effluent Monitors and Primary Coolant Monitor Applicability These specifications apply to the per=anently installed systens that monitor: the airborne radioactivity leaving the facility and also monitor the activity. of the pri=ary coolant.
Objective The. purpose of - this specification is co ' establish a =inict=2 operability level for the effluent and, primary coolant.sanitor syste=.
t '1
- 2
. Specification a.
The normal contingent of operating monitors when the reactor is operating shall be as follows:
Building air continuous monitor Stack air continuous monitor Stack particulate continuous monitor Primary water monitor b.
The alarm points for the two gaseous effluent monitors shall be clearly stated in a facility operating procedure.
The alarm points for the primary water menitor and stack particulate monitor shall be posted on the radiation monitor panel in the control room.
The outputs of the two air monitors shall be recorded on a strip c.
- chart, d.
Both air monitor systems shall provide fixed filters for evaluat-ing particulate releases.
One of the two gaseous effluent monitors may be out of service e.
for up to four hours while the reactor is operating provided that no unusual experiments are being conducted and no radioactive chemical processing is being done in the hoods. The primary water monitor may be inoperative for up to eight hours while the reactor is operating providing both air effluent monitors are operating. The stack particulate monitor may be out of service for extended periods provided that the fixed filter is evaluated daily. The recorder may b2 out of service for extended periods of time as long as the effluent monitor values are logged at nominal 15 minute intervals while the reactor is operating. The primary water monitor need not be operative in the natural convection mode.
Basis Specification 3.3.2.a provides assurance that adequate operating instrumentation exists to monitor airborne effluents and primary coolant activity when the reactor is operating.
Specification 3.3.2.b provides for the unambiguous establishment of alarm settings for the effluent and coolant monitors.
Specification 3.3.2.c assures that a permanent record of effluent releases shall exist.
Specification 3.3.2.d assures that a means shall be available to evaluate particulate releases from the facility.
Specification 3.3.2.e provides for limited inoperability of the monitor equipment in order to permit maintenance and calibration of ~ the systens while the reactor is operating.
3.4
' Engineered Safety Features 3.4.1-
. Reactor Containment Building.
. Applicability This specification applies to the f acility containment vessel and its associated airlocks and ventilation system.
Obj ec tive The objective of this specification is to control the release of airborn radioactivity f rom the f acility under accident conditions.
Specification The reactor will not be operated unless the f ollowing conditions are satisfied:
a.
The truck door shall be closed and sealed, one door in each of two airlocks shall be closed and sealed, and all other penetrations other than ventilation ducts shall be sealed.
b.
The stack fan in the University steam plant shall be operating and the fan in the building air duct shall be operating.
The pressure in the containment vessel shall be negative c.
with respect to the outside atmosphere.
d.
The dampers in the containment ver.tilation ducts (4) must be operating and be capable of closing in five seconds or less.
They must close automatically in response to coincident alarms from the building air monitor and the bridge monitor or a manual signal.
A charcoal filter shall be maintained in the energency exhaust e.
duct and the modulating damper in the exhaust duct and its controller shall be operable.
f.
The leak rate of the containment vessel shall not exceed 3.32/sec (7 cfm) standard air et a negative diff erential of 1.27 cm of water ( " water).
Easis Specification 3.4.1.a assures that all openings in the containment vessel other than air ducts will be closed during operation.
Specification 3.4.1.b assures that fans capable of maintaining a negative pressure in the building under both normal and emergency ccnditions are operating.
Specification 3.4.1.c assures that any leakage in the containment vessel will be inward.
Specification 3.4.1.d assures operability of the emergeny dampers in the event of an accidental release of activity.
Specification 3.4.1.e assures that, under accident conditions, the containment will be vented in a controlled manner through an activated charcoal filter to the stack.
Specification -3.4.1.f assures that the leak rate of the containment-building will not be greater than the 5% free air volume par day as used in the analysis of the design basis accident.
3.5 Primary Coolant Conditions Applicability This specification applies to the primary coolant purity, radioactivity, and flow distribution.
Objective-The purpose of this specification is to assure that coolant conditions are such that corrosion to the fuel and pool components is minimized, and to also assure that coolant does not by-pass tha fuel.
Specification a.
The primary coolant pH shall be maintained between 5.0 and 7.5.
b.
The resistivity of the primary coolant when the reactor is operating shall normally be greater than 200,000 ohns per centimeter. Operation with resistivity less than 200,000 ohms per cm, but greater than 75,000 ohms per cm. is permissible for time periods not to exceed 15 days.
c.
During forced convection operating, grid plate holes not occupied by fuel must be blocked by plugs, reflector elements, or experiments.
. Basis Experience accumulated at many research reactors has shown that water of.a quality specified in specification 3.5.a and 3.5.b. will not corrode aluminum or zircalloy.
Specification 3.5.c. assures that all primary coolant flow is through
- the fuel when in the forced convection mode.
'3.6 Airborne Effluents Applicability This specification applies to levels of radioactivity discharged to
-the environment through the building air duct and the power plant stack.
Objective This specification assures that persons in the facility or any
- location outside the facility will not be exposed to concentrations of radioactive gases in excess _of those limits stated in 10 CFR 20 appendix B.
l Specifications a.
For the power plant stack, radioactive naterial release rates shall be limited such that:
Yearly Average:
Qi/MPC I 8.1 x 103 i=1 Instantaneously:
Qi/MPC I 2.4 x 10
i 1
where Qi
= concentration (Ci/s) of isotope (1).
(u Ci/cc).
= Unrestricted MPC of the ith isotope.
N
= number of isotopes being released.
b.
For the building air duct, effluent concentrations shall be limited such that:
Yearly Average:
N[
(C MPC N1 g
i=1 N
Instantaneously: b (C':- FTC
)< 2 i
i=1 where C
= concentration (uci/cc)of the i isotope 1
th '
MPC P '
R i
Basis Compliance with the above limits will ensure that neither the exposure limits for the public, or occupational workers, as specified in 10 CFR:20, shall be exceeded.
. L.1 uid Effluents 3.7 3
Applicability _
This specification shall apply to radioactive effluents released to the sanitary sewer.
Obj ec t ive The purpose of the specification is to prevent unraonitored releases to the sanitary sewer, or taleases in excess of limits specified in 10CFR20.
Specification a.
All liquid effluent shall be retained by a hold tank system.
b.
Before releasing effluent from a hold tank to the sanitary sewer, the solution in the tank shall be mixed, and a repre-sentative sample shull be analyzed.
c.
Tie effluent shall not be released to the sanitary sewer unless tae sample analysis has demonatrated that:
N (C : MFCR)
+C
' MPC
<1 i
U RU -
i=1 where C
= concentration of the ith isotope (pCi/cc)
= restricted MFC of the ith isotope (for water)
Rg N
= number of isotopes in the solution C
= concentradon of unMentmed 8 eMtters U
= MPC for unidentified S emitters as established U
in 10CFR20.
For the purposes of establishing the above: The most recently established nominal release rate of water for the Winspear Avenue sewer trunk, shall be included when calculating "C ".
Basis The above specifications shall ensure compliance with NRC limits for releases of radioactive materials to sanitary sewers.
3.8 Limitations on Experiments _
Applicability This specification applies to all experiments installed in the NSTF reactor.
. Objective The purpose of these specifications is to prevent damage to the reactor or excessive releases of radioactivity in the event of an experiment failure.
Specification Experiments shall be designed and operated to prevent local a.
boiling of the moderator-coolant or melting of any structural component of the experiment.
b.
All samples or experiments shall be doubly encapsulated if release of the contained material could cause excessive corrosive attack of the reactor or experimental facility, excessive contamination or production of airborne radioactivity, or a violent chemical reaction.
Each such capsule shall be capable of containing the reaction and resulting pressure.
The container for experiments shall be designed to contain c.
the maximum pressure which could he produced by the sum of gamma heating, fission heating, exrernal heating, and radiolytic decomposition.
Doubic encapsulation shall be provided where appropriate, with the outer encapsulation capable of containing the stored energy of the inner container.
Pressure relief devices shall be provided where appropriate, d.
Experiments shall not interfere with the normal operation of the control blades, nor should they adversely affect any other safety system or required instrument channel, Experiment variables that could affect the safety of the facility e.
shall be monitored, and limits established.
Alarms and reactor trips shall be provided as appropriate, f.
No single e.cperiment shall occupy mare than two adjacent fuel spaces on the grid plate.
g.
Explosive materials shall not be irradiated in the reactor tank, nor inside 1_ beam tube.
Irradiations in the thermal column, dry chamber, or outside a beam tube must be described in detail in an experiment plan, and approved by the Nuclear Safety Committee. The aggregate mass of explosive material that may be irradiated at any one time shall noe exceed 175 grams equivalent TNT, and the maximum stored inside the reactor containment building shall not eaceed two pounds equivalent TNT.
h.
A fission plate containing up to 1 kg of highly enriched U-235 may be used at-the outside face of the thermal column only.
i.
Except for the fission plate, no special nuclear material or source material other than trace quantities shall be irradiated unless a detailed experiment plan is proposed and approved by the Nuclear Safety Committee.
. Basis Specifications a. b, c, d, and e are logical design considerations which, if followed, minimize the probability of experimert failure or adverse ef fects on reactor operation.
Specification f assures that no in-core experiment with a cross sectional area greater than 16 in.2 (103.2 cm ) will be considered, 2
and thus require facility re-licensing as a test reactor.
A double hole has a cross section of 17.26 in.2 (111.4 cm ); thus, 2
any experiment designed for such a hole must be slightly smaller.
Specification h assures that the facility owned fission plate will be used only in a very low neutron flux.
Explosives and fissile material present unusual hazards.
Since the various irradiation facilities vary greatly in both flux and physical characteristics, no minimum amount of these materials can be unequivocally stated as being safe to irradiate.
Specifications g and i assure that these mate;ials are irradiated only under very specific conditions that have been thoroughly evaluated 4
3.9 Fuel Limitations Applicability This specification pertains to the fuel used in the NSTF reactor.
It does not pertain to experiments that contain fissile material.
Objective The objective of this specification is to assure that only "PULSTAR" type fuel is used in the NSTF reactor.
Specification a.
Only "PULSTAR" type fuel using nominal 6 per cent enriched UO2 shall be used in the NSTF reactor.
b.
The reactor will not be operated with fuci in the core that is known to be leaking fission products except for auch operation that may be necessary to find the leaking fuel pin.
c.
Fuel will be limited to a maximum burnup of 20,000 Mw-days / metric ton.
The first fuel assembly to reach 15,000 Mw-days / tonne will be removed from the core. After six months decay,;this fuel assembly will be disassembled and a representative sampling of pins will be inspected and measured. The results will be reported to the NRC.
If significant degradation of the cladding is found, all fuel in the core with >15,000 Mw-days / tonne will be removed and no additional fuel will be allowed to exceed 15,000 Mw-day / tonne without NRC approval.
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Basis Specification 3.9.a ensures that only fuel that is of a type that has been thoroughly tested and characterized will be used. The NSTF Safety Analysis and the specifications in this text were based on "PULSTAR" fuel.
Specification 3.9.b assuret that only non-leaking fuel will be used for normal operation.
v,* ration with a leaking fuel pin would lead to unnecessary levels of gaseous effluents and pool contamination.
Specification 3.9c assures that the burnup will be limited to the region where operating experience exists for zircalloy clad power reactor fuel. Reference license amendment #10 dated 5/29/75.
4.0 Surveillance Requirements 4.1 Reactivity Limits Measurements Applicability This specification applies to the surveillance requirements for
. reactivity limits.
Obj ective The objective of these -specifications is to ensure that the specifications of section 3.1 are satisfied.
Specification The shutdown margin as specified in section 3.1.a shc11 be a.
verified each time the core fuel configuration is changed and also at any time that an experiment worth > +.3% aK/K is loaded.
b.
The worth of individual samples or experiments shall be measured when they are initially loaded into the reactor.
Samples of very small worth or samples and experiments of a repeated nature need not be measured each time they are loaded.
At least one control blade shall be removed from the core each c.
year for inspection.
Inspection shall be on a rotating basis, d.
The worth of the control blades shall be measured at least once per year.
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t Basis Specification 4.1.a assures that the minicum shutdown cargin is not lost due to fuel or experiment =anipulations.
i Specification 4.1.F assures that the worth of cach experiment /
sample or class of experiments / samples is known.
Specification 4.L.c assures that the coat ol blades will be examined on a fixed schedule so that 2echanical defects can be spotted before gross failure occurs.
i Specification 4.1.d assures that valid blade worth data is r
available for measuring experiment worth, Kex, shutdown cargin, etc.
4.2 Reactor Safety System Tests t
l Applicability This specification applies to the surveillance of the reactor safety systems. It applies at any time that a critical mass is assembled on the grid plate.
i Objective The objective of these specifications is to assure operability of the reactor safety systems as described in section 3.2.
4 Specification a.
A channel test for each channel. labeled
- in table 3.1 shall be performed prior to each start-up following a period when-i the-reactor was secured.
4 1
b.
A channel calibration for each channel labeled + in table 3.1 shall be performed in each calendar quarter.
- c..Any safety related instrument requiring maintenance or repair shall be operability checked and calibrated before being returned to service.
d.
Ion chamber channels shall be calibrated as frequently as
- required to maintain true power indication as determined by i
either heat balance measurements or 'the N-16 channel. The linear and log channel may deviate by _+10 and -30% _ f rom nominal ~
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These limits may be exceeded for short periods of time immediately following a startup, a sample loading, or other event that alter rod shadowing of the ion chambers. This entire specification is null for powers < 1 Mwatt.
e.
Control blade drop times shall be measured in each calendar quarter or whenever repairs are made that could effect their performance.
Basis Sp::sfication 9.2.a assures that all saf ety related systems are on and functioning each time the reactor is to be operated.
Specification 4.2.b assures that all safety related systems are tested on a periodic basis to insure proper operability, trip point settings, and proper relationship between measured variable and output signals.
Specification 4.2.c insures that all instruments will perf orm as intended when installed in the reactor systems.
Specification 4.2.d assures that the ion chamber power indicating channels will be calibrated frequently when operating near full power. The neutron flux incident to the ion chambers is significantly affected by coctrol rod positions and hence xenon concentration, sample loadings, etc.
Two temperature channels and an N-16 based power channel are available for power determination and are not af fected by control blade positions. The ion chamber positions are easily adjustible to permit frequent cross calibration of the power channels. The limits on deviation from true power are somewhat arbitrary and are based on past practice.
Specifications 4.2.e assures that control blade operation will be verified on a routine basis or as warranted by maintenance.
4.3 Radiation Monitoring Systems Tests Applicability These specifications pertain to the surveillance requirements for the permanently installed area monitors, effluent monitors, and primary water monitor.
It applies at all times that radioactive material is being used, reored, or generated within the containment building.
Obj ec tive The purpose of these specifications is to assure that all radiation monitoring channels are tested frequently for operability and satisfactory performance.
Specification a.
The area monitors, effluent monitors, and primary water monitor shall be tested for operability on o monthly basis.
b.
A complete calibration of the area monitors, effluent monitors, and primary coolant monitor shall be performed each calendar quarter.
The hot cell monitor may be excluded from ' specifications a and c.
. b above if conditions in the cell make entry inpractical.
d.
Effluent monitor sensitivities shall be determined experimentally once per year.
Basis Specifications 4.3.a and 4.3.b assure that the various radiation monitors are tested and calibrated on a routine basis.
Specifications 4.3.c allows the hot cell conitor be be exempted f rom normal test and calibration schedules when activity levels in the cell prohibit entry. During such periods, the monitor will be maintained to the extent possible.
Specification 4.3.d assures that accurate sensitivity data is s.vailable for calculating effluent releases.
4.4 Eng,ineered Safety Features Tests (Containment)
Applicability This specification applies to the surveillance requirements for the reactor containment vessel.
Objective The purpose of these specifications is to assure that the containrent and ventilation systems are tested on a routine basis.
Specification a.
Prior to each startup that follows a period when the reactor was secured, it will be confirmed that the air locks and truck door are sealed, that the containment vessel pressure is negative with respect to the outside atmosphere, and that hydraul'.c pressure is available to the ventilation dampers.
b.
The following items will be tested in each calendar genrter -
1.
All dampers close in less than five seconds in response to both normal trip and radiation signals.
2.
The control syste=s maintain negative pressure in the building under both normal and emergency conditions.
3.
The condition of the emergency duct chercoal filter will be visually inspected.
Thevolumetric leak rate of the containment vessel vill be measured c.
annually.
Basis Specification 4.4.a assures that basic containme.nt and ventilation system components are operable when the reactor is in use.
Specification 4.4.b assures that ventilation components required for emergency conditions will function properly if called upon'to do so.
. Specificetion 4.4.c assures that all penetration seals are intact and that there has been no general deterintion of containment integrity.
4.5 P*timary Coolant Conditions Measurements Applicability This specification applies to the surveillance of pool water conditions, It is in effect as long as fuel is present in the reactor tank.
Obj ective The purpose of this specir'. cation is to insure that water quality is measured of ten enough to prevent corrosion of sytem components and also to detect higher than normal activity levels in the coolant.
Specification Pool water pH, conductivity, and grass beta activity will be a.
measured on a weekly basis.
b.
A detailed isotopic analysis will be made of pool water semi-annually.
c.
A record will be maintained 0? all pool water additions.
Basis Specification 4.5.a assures that pccr pool water quality could not exist for long without being detected.
Specification 4.5.b provides for detailed water analysis that can be useful in detecting small fuel leaks, experiment failure, etc.
Specification 4.5.c provides a means for detecting prtmary system leaks.
4.6-Liquid Effluents Monitoring Applicabil!ty This specification applies to the surveillance of liquid wastes discharged to the sanitary sewer system.
Objective The purpose of this specification is to insure that no waste is discharged to the sewer without first being assayed.
Specification No potentially active liquid wastes will be discharged to the sanitary sewer without first being assayed. Records of the assay data' and volume of liquids discharged must be maintained. All such discharges must be within applicable state and federal limits.
Basis The basis for this specification.is the federal and state requirements that all radioactive material released to the environment be monitored.
l f ;
I 5.0 Design Features 5.1 Site Description The site of the NSTF reactor is the south center edge of the Main Street Ca= pus of the State University cf New York at Euffalc.
The ca= pus is in the triangle bounded by Bailey Avenue, running f
almost due north and south, Winspear Avenue, running roughly east and west, and Main Street, running northeast and scathvest -
The reacter site is about 152 =eters (500 f eet) due nort5 c i Winspear Avenue. The nearest buildings are Acheson Hall, Have Eldg., the stea= plant and Clark Gy=.
The reactor is about 30 neters (100 f eet) north of Acheson, 46 ceters(150 feet) west of the stea:
plant, and 91 =eters (300 feet) south of Clark Gy=.
A large Veterans Administration Hospital is situated about 610 =eters 5
(2000 f eed east of the reactor. The nearest residential aren is on the south side of Winspear Avemue.
i The reactor restricted access area consists of the centain=ent building and the attached laboratory and office ving.
5.2 Contain=ent Building The contain=ent building is a flat roof ed, right cylinder nc=inally
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21.34 =eters (70 feet) in dia=eter and 15.34 meters (52 f eet) high.
The vessel is constructed of normal density reinforced concrete.
The walls are a no=inal 61 c= (2 f eet) thick and the roof is 10.2 c= (4 inches) supported by steel and concrete bea=s.
The floor is 1.07 =eter (3 feet) thick and the entire building rests on bedrock.
The total free air volu=e of the building is 5267 cubic =eters 3
(166,000 f eet ).
.The building contains two personnel air locks and a single barrier truck door. All electrical and piping penetrations are sealed Drain lines are provided with 61 c= (24 inches) dip legs to =aintair a seal.
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5.3 ventilation Under nor=al conditions, the contain=ent building is ventilsted by a single pass type syste=.
Filtered, conditiened air is supplied to r,he vessel through two 75 c= (30 inch) diaseter ducts. Inhabitec areas and sc=e inter =ediate level fu=e hoods are exhausted through t
the roof of the contain=ent via a 91 c= (36 inch) duct. The blower in the contain=ent exhaust duct has a variable da=per on its suction side which is used' to control the negative pressure.of the building.
So=e reactor experi= ental facilities and several high level fu=e hoods are. exhausted.into a 46 c= (18 inch) duct that discharges at' the ' top of the stea= plant stack which is 50.9 =eters (167 f eet) above ground level. The stack-duct is driven by.tvo blowers one in the facility base =ent. and another in the stea= plant. All air leaving the' containment passes. through' absolute filters.
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_ _ _ _ Under emergency conditions, all fans in the conttinment tailding are automatically turned of f and dam, ers in the normal ventilation ducts are automatically closed. Only the fan in the steam plant remains operating. This fan draws sir from the containment through a 15\\ cm ( 6 inch) duct containing an absciute filter and an activated charcoal filter and exhausts it out the steam plant stack duct.
5.4 Primary "oolant System The primary coolant system consists of the renetor tank, a nitrogen-16 decay tank, a cump, Seat exchanger, and various support syatems.
The reactor tank is 8.0 meters (29 feet) deep, aluminum lined and holds approximately 55,266 liters (14,600 gallons) of water.
The remairder of the system contains approximately 21,803 liters (5,760 gallons) of water and is f abricated of either aluminum or stainless steel.
Valves are located at points which permit isolation of the reactor tank or each of the other major components.
Two demineralizer systems are available; one provides pure water for make-up to the system, and the other is in a continuously circulating clean-up leap.
An emergency pool fill system is available for adding city water to the pool should this be desired. In the event of a gross leak from the reactor tank, a manual valve can be opened to supply water directly from the municipal water supply.
5.5 Fuel and Reflectors Fuel assemblies are.8 cm (3.15 inches) x 6.96 cm (2.74 inches) in cross section and 96.5 cm (38 inches) long. Each assembly contains 25 pins in a-5 x 5 array. The pins are positioned by aluminum grids at each end. The end grids each contain 25 holes of.63 cm (k inch)-
diameter for. coolant passage. The lower end of the assembly consists of an aluminum nosepiece that mates with the grid plate. The top e#
the assembly contains a bail for handling purposes. The center bo:.
portion is made of.152 cm (.060 inches) zircaloy.
Each fuel pin is made up of a zircaloy tube 1.194 cm ( 47 inches) 0.D. and 0.047 cm (.0185 inches) minimum wall, containin, a stack 60.96 cm ( 24 inches) long of sintered uranium dioxide pellets.
Welded caps form the closure for che tube ends. Pellets are 1.067 cm (.42 inches) diameter and han s a minimum density of 10.2 g/cc.
Enrichment is to 6% in U-235.
Typical composition of a f,.i assembly is as follows -
Uranium-235 =.768 Kg (1.69 lb).
= 12.83 Kg (28,29.1h)
Uranium dioxide = 14.56 Kg (32.1 lb)
Complete Fuel Ar.sembly = 20.37 Kg (45 lb)
. ihe reactor may be reflected by ncr=al water, graphite, lead, aluminum or voids.
5.6 Reactivity Control Reactivity control ic provided by six neutron absorbing blades.
Each is composed of d"% silver, 15% indium, and 5% cadmiu=.
The Blades are 12.32 cm (4.85 inches) wide by.457 cm (.180 inches) thick by 73.7 cm (29 inches) long. The blades are plated with
.0076 cm (.003 inches) nick 1.
The blades are located in alumictm guides which are positioned between and supported by adjacent fuel assemblies. Five of the blades have extensions which extend above the surf ace of the pool where they mate with out of water magnets. The sixth blade (formerly the pulse blade) connects directly with its drive. The drives have a stroke of 66 en (26 inches) and a nominal drive speed of.127 cm/sec (3"/ min.).
The five blades supported by magnets provide scram capability for the functions listed in Table 3.1.
The non-scra= ming blade is used f or flux distribution control only. Its worth is net considered in computations of shutdown margin. It may be left at its fully with-drawn position and unused as was the cace when it was intended for pulsing. The non-scramming blade may be converted to a scram-safety blade at any futere time as long as its resulting perfor=ance characteristics are the same as the other five blades. The NRC will be notified prior to initiation of any such modification.
One of the control-safety blades may be used for automatic servo-g control of reactor power. When in use, it maintains a constant power level as indicated by the linear power chant el.
5.7 Fuel Storage.
5.7.1 Cold Fuel Storage Cold fuel pins and asse=blies shall be stored within the containceuc building in secure vaults constructed of non-fla=nable materials.
Fuel will be stored in rows of metal cylinders. The cylinders will have a minimum center to center spacing of 12.7 cm (5 inches) and the rows will have a minimum center line spacing 'f 40.6 c2 (16 inches).
Each cylinder will contain no more than one fuel assembly or 25 individual pins. Cold fuel is defined as any fuel that can be stored in the vault without creating a radiation field at the outer vault boundary in excess of 5 mr/hr. The vaults are specifically exempted from the criticality alarm requirements of 10 CFR part 70.24.
The fission plate may be stored in a fuel vault, a locked cask within the containment building or in the dry chamber or thermal coluun.
A minimum distance of 40.6 cm (16 inches) will be maintained between the fission plate and any other fissile material.
l
. 5.7.2 Irradiated Fuel Stcelyg Normally all irrediated fuel will be stored in the reactor tank either on the grid plate or in storage racks. The pool storage racks will be linear with a minimum center to center spacing of 15.5 cm (6.1 inches).
In order to perform repairs in the lower tank, ir adiated fuel may be stored out of the pool. Storage may be in the hot cell or in a specially fabricated shielded f acility within the containment building.
In ',1ther case, storage must be in accord with the following st'.pulations -
The cal ulated Keff of the storage array must be <.85 for a.
the flooded conditico.
b.
The storage configuretion must be approved by the Nuclear Safety Committee.
Transfer of fuel to the facility must be done according to c.
detailed written procedures approved by the Operating Committee.
d.
Shielding must be adequate to reduce the radiation level to
< 100 nR/hr at the outer boundry of the storage facility.
Any facility containing > 15 fuel assemblies or an equivalent e.
number of pins shall be equipped with a neutron sensing criticality alarm.
f.
The facility will be vented to the stack exhaust system.
6.0 Administration 6.1.1 Manag ement The Nuclear Science and Technology Facility is owned and operated a.
by the State University of New York. Its position in the University structure is shown in Figure 6.1.1.
b.
The University will provide whatever resources are required to maintain the facility in a condition that poses no hazard to the general population.
6.1.2 Facility Staff Requirements Figure 6.1.2 shows the minimum staffing require =ents and their a.
functional relationship.
b.
The Facility Director shall have overall responsibility for the safe operation of the plant.
c.
A Radiation Protection Lept. organizationally independent of the Operations Dept. shall be responsible for radiological safety at the Facility, d.
A licensed reactor operator (RO) or a licensed senior reactor operator (SRO) shall be present in the control room whenever the reactor is operating.
A minimum of two persons must be present in the facility when e.
the reactor is operating; the operator in the control roes and a second person that can be reached from the control room by telephone or intercom.
f.
The follawing operations must be supervised by a senior reactor operator -
.- _ FIGURE 6.1.1 Regents of the State University of New York Chancellcr or the State University of New York President of SUNY at Buffalo Vice President Research & Graduate Studies Director Campus Nuclear Science and Radiation Safety Technology Facility Officer
- 29 FIGURE 6.1.2 l
Nuclear Campus t
Director Safety
,___}
Radiation Safety j
Committee Officer I
1 i
Operations Manager Reactor Manager Engineer Radiation Protection Dept.
Senior Reactor Opera tors Reactor Operators Note: The position of Reactor Engineer and Operations Manager may be held by the same individual.
The position of Reactor Engineer may be filled by the director or a qualified member of the faculty.
Positions need not be filled by full time employees.
During periods of absence, supervisory personnel shall delegate their authority to the next best qualified person available.
i
_. i.
1.
Fuel manipulations in the core when there are a 15 fuel assemblies on the grid plate.
2.
When experiments are being manipulated in the core that have an estimated worth >.6% AK/K.
3.
Removal of one or more control blades if there are 1 10 fuel assemblies on the grid plate.
4.
Removal of shielding plugs from a beam tube.
5.
Resumption of operation following an unsc heduled shutdown.
This requirement is waived if the shutdown is due to an interruption of electrical power to the plant.
g.
A licensed senior reactor operator must be present or readily available on call at any time the reactor is in operation.
h.
The reactor must be secured at times when no reactor operator or senior reactor operator is present in the facility.
- i. All licensed operators at the facility shall participate in an approved requalification program as a condition of their continued assignment of operator dutica.
6.2 Review Functions 6.2.1 Nuclear Safety Committee a.
A Nuclear Safety Committee shall exist for the purpose of reviewing matters relating to the health and safety of th 7ublic in accordance with the constitution and by-laws of that Comm_ttee.
b.
The NSC shall consist of a minimum of six persons with expertise in the physical sciences and preferably some nuclear experience.
Permanent members of the committee are the Facility Director, the Campus Radiation Safety Officer, the Radiation Protection Department Manager, and the Operations Department Manager.
c.
The NSC shall meet at least twice per year, or as of ten as
-required.
d.
A quorum of the NSC must have at least five (5) members present, and must have a majoricy of non-NSTF members present..
c.
The Committee Chairperson shall notify the members of the Agenda not later than forty-eight (48) hours before a regular meeting.
For emergency meetings, the Agenda shall be included with the meeting notice.
.f.
-Questions before the Committee shall be decided by vote of the members present, the concurrence of a majority of those present being required for approval, except that no question aquiring specialized knowledge shall be decided unless a member or con-sultant who is qualified by training and experience in that field is present.
g.-
Minutes of all_ meetings will be retained in'a file, and also distributed to all NSC members.
_ = _.
. 4 1
h.
The NSC shall review the following:
1.
Experiments referred co it by the Operating Committee because of the degrca of hazarf involved or the unusual nature of che experiment.
2.
Reportable occurrences.
3.
Proposed changes to the facility license, changes to technical specifications, and experiments or changes made pursuant to 10CFR 50.59.
6.2.2 Operating Committee a.
An Operating Committee shall exist as a sub-group of the NSC.
b.
The Operating Committee shall consist of the Facility Director, The Radiation Protection Department Manager, and the Operations Department Manager. The Committee shall seek advice and 7
comment from other qualified individuals as appropriate.
c.
The Committee shall meet as frequently as needed. Minutes will be maintained for all formal meetings, d.
The Operating Committee shall perform the following:
1.
Review experiments which present no significant safety problems.
2.
Approve additions to or revisions of any Operating Procedures.
3.
Review abnormal occurrences.
4.
Perform facility inspections.
Facility Audit a.
A consultant will be retained by the University to perform an' annual audit of reactor operations. The consultant shall be selected by the Director of the NSTF and shall be an individual with expertise in the nuclear field. The consultant shall submit a report on his findings to tne Director of the NSTF.
b.
The consultant shall audit the following:
'1.
Reactor operators and operational records for compliance with internal rules, procedures, and regulations, and with license provisions.
2.
Existing Operating Procedures for adequacy and accuracy.
3.
Plant equipment performance and its surveillance requirements.
v 6.4 Action To Be Taken In The Event Of A Reportable Occurrence In the event of a reportable occurrence, as defined in these Technical Specifications, the following actions will be taken:
a.
Immediate action will be taken to correct the situation and to mitigate the consequences of the occurrence.
b.
The Operating Committee wi.1 investigate the causes of :he occurrence, and review actions taken in response to the occurrence. The Operating Committee will report its findings to the NRC, the Nuclear Safety Conmittee, and to the Vice President for Retcarch and Graduate Studies.
The report shall include an analysis of the causes of the occurrence, the ef fectiveness of corrective actions taken, and recommendations of measures to prevent or reduce the probability or consequences of recurrence.
6.5 Operating Procedures Written procedures will exist which define how and when various aspects of facility operation will be performed. These procedures will be reviewed and updated as frequently as needed, but the review will be nc less frequent than once per year. All new or revised procedures will be reviewed and approved by the Operating Commit tee.
Procedures will address tne following areas:
a.
Normal reactor operation.
b.
Use, surveillance, and matatenancc of auxiliary systems, c.
Use of experimental facilities.
d.
Abnormal and emergency situations - health and safety, e.
Reactor electrical and mechanical surveillance and maintenance.
Temporary changes to the procedures that do not change their original intent may be made with the approval of a senior reactor operator.
All such changes shall be documented.
. 6.6 Operating Records The following records and logs shall be retained by the a.
licensee for at laast five years:
1.
JIormal facility operation and maintcaance.
2.
Reportebic occurrences.
3.
Tests, checks, and measurements documenting compliance with surveillance requirements.
4.
Records of experiments performed.
5.
Operator requalification program records.
6.
Facility radiation and contamination surveys.
7.
Minutes of the Operating Committee meetings.
b.
The following records shall be retained by the licensee for the life of the facility:
1.
Gaseous and liquid waste released to the environs.
2.
Radiation exposure records for all facility personnel.
3 Fuel inventories and ttansfers.
4.
Updated, corrected, and as-built facility drawings.
5.
Minutes of the Nuclear Safety Committee meetings.
6.7 Reporting Requirements 6.7.1 Financial Report A copy of the Facility annual financia? report shall be filed with the NRC as required by 10 CFR 50.71 (b).
6.7.2 Annual Operating Report A report summarizing facility operations will be prepared for each calendar year. A copy of this report shall be submitted to the NRC Region 1 office of Inspection and Enforcement by lbrch 31st of each year.
The report shall include the following:
a.
A brief narrative summary of:
1.
Changes in facility design or performance that reiate to reactor safety.
2.
Results of surveillance tests and inspections, b.
A tabulation showing the energy generated for each month and for the year in MW-hrs.
A list of the unplanned shutdowns including the reasons therefor, c.
and corrective action taken, if any, d.
Discussion of the major maintenance operations performed during the period, including the effects, if any, on safe operation of the reactor, and the reason for any corrective maintenance required.
c.
A brief description of:
1.
Each change to the facility to the extent that it changes a description of the facility in the Safety Analysis Report.
2.
Reviews of changes, tests, and experiments made pursuant to 10 CFR 50.59.
f.
A summary of the nature and amount of radioactive ef fluents released or discharged to the environment.
g.
A description of any environmeraal surveys performed outside the facility.
. h.
A summary of radiation exposures received by facility personnel and visitors, including details of unusual exposures, and a brief summary of the results of radiation and contamination surveys performed within the f acility.
- i. Any changes in facility organization.
6.7.3 Reportable Occurrence Reports Notification shall be made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone or telegraph to the Director of the Regional Inspection and Enforcement Office followed by a written report within 14 days in the event of a reportable occurrence, as defined in Section 1.0.
The written report and, to the extent possible, the preliminary telephone or telegraph notification shall:
a.
Describe, analyze, and evaluate safety implications.
b.
Outline the measures taken to assure that the cause of the condition is determined.
c.
Indicate the corrective action taken to prevent repetition of the occurrence including changes to procedures.
d.
Evaluate the safety implications of the incident in light of the cumulative experience obtained from the record of previous f ailure and malfunction of similar systems and components.
6.7.4 Unusual Event Report A written report shall be forwarded within 30 days to the Director of the Regional Inspection and Enforcement Office in the event of:
a.
Discovery of any substantial errors in the transient or accident analysis or in the methods used for such analysis as described in the Safety Analysis Report or in the basis for the Technical Specifications.
b.
Discovery of any substantial variance from performance specificaticus contained in the Technical Specifications or Safety Analysis Report.
c.
Discove:y of any condition involving a possible single failure whic!., for a system designed against assumed failure, could result in a loss of the capability of the system to perform ito safety function.
6.7.5 Special Nuclear Materials Reports Material status reports and nuclear material transfer reports for special nuclear materials shall be made in accordance with Sections 70.53 and 70.54 of 10 CFR.
1 JUSTIFICATION FOR CHA';GES TO TECHNICAL SPECIFICA110NS The most sign if icant change in the proposed Technical Specifications is the elimination of pulsing.
Pulsing was never utilized to any great ex t en t a nd in recent years the demand for it has been zero.
On the other hand, maintain!,g pilse capability has a number of disadvantages such as the f ollowing -
1.
Maintenance of unused instrumentation and mechanical sy stes.
2.
Maintaining operator proficiency is a probl(a since pulsing is too cumbersome to do f or practice.
Operatora licensed in the past several years have not been tested on pulsing and are qualified for steady-state operation only.
3.
The major objection to further pulsing is due to potentic.1 damage to the recently procured fuel.
Pulsing does produce transient thermal stress in the cladding and f ractures the UO pellets.
With the high cost of fuel and mir commitnents 2
to three shift steady-state operation, these added risks are not considered worthwhile.
Our decision to discontinue pulsing should not be laterpreted as a f ailure of the original concept.
The reactor Ld meet most of the original design obj ec tiv es.
If nothing else the pulse test program did show that "pulstar" fuel can survive extrane power tran.ients and is thus an extraordinarily saf e reactor.
The decision not to pulse raises the question of what to do with the pu l se blade.
There are three alternatives - 1) Convert it to a control-saf aty blade identical to the others, 2) Leave it as it is and maintain it fully withdrawn in the steady-state as has been required for the past 17 years, oc ?) Def eat the pilse mechanisms, af fix the blade extension rigidly to its drive, and use it as a control blade with no safety capability.
The first option is the most desirable but would be very expensive to accomplish since it would require a new magnet, magnet control amplifier, and many mechanical components.
All of these thinta would have to be custom fabricated since there is no supplier of such p rts.
i-second option is the easiest since it requires no changes at all to the existing system.
It is undesirable because, with the pulse rod full out, the flux peaks in that corner of the core where flux peaking is generally not needed.
The third option is a good canpromise and is the one we have chosen. All that is need ed is to plug the inlet ports of the pneumatic cylinder to render it inoperative end to provide a rigid coupling between the contral blade extension and the rod drive.
Saf ety would not be compromised since the worth of this rod would not be considered when computing shutdown margin.
In order to provide 2 axis flux distribution spanetry, an even number of control blades is needed. Use of the pulse blade would provide six control blades. Control blades are discussed in section 5.6 of T.S.
I will proceed through the revised T.S. sequentially. Any specification tha t is significantly dif f erent from a previously approved specification will be discussel and justified.
All of the limits in section two pertalaing to power, flow, and coolant temperatures are justified by the thenno-hydraulic analysis performed by Mr.
Louis Henry. A copy of this document is contained in the appendix of this letter.
1 Spec if icat ion 3.1. b ra ises the total wor t h allowed f or ex periment s f ren 2 to 3% AK/K.
Experiments presently conducted at NSTF approach the Z limit.
Future work loads could require p eater than 2%.
The control-safety blades ar e capable of controlling this uuch K-excess.
The blade gang (not counting the pulse blade) is typically wcrth between 9 and 12% AK/K.
About 2 is needed to overcome temperature and xenon eff ects.
Add to this the 3% f or experiments and the to tal K-exc es s ne ed ea is 5%.
On the other hand, the shu tdown margin requirement is 1% plus the worth of the most reac tive blad e.
Typically this is 1% + 3% or 4%.
The 4% shs tdown margin plus the 5% V-(
<ss adds up to 9% which is available in the blad. gang.
Even if some future core loading resulted in a gang worth less than 97, there would be no problem.
The shutdown margin requirement is absolute and snuld take precedence over the permissive 3% limit.
Specification 3.1.c sets a worth Ibnit of. 61 AK/K f or unsecured e< p e r im en t s.
Unsecured experiments are not discussed in the old T.S.s and so this is a new LCO (Limiting Cor.dition for Operation).
The talue of.6 was chosen to be less than the delayed neutron fraction. The pulse test progran demonstrated that a step change of this magnitude creates no problens.
Specification 3.1.d is also a new LCO.
The old T.S. do not specif y where the rods should be during fuel manipulations. The value of 31 was chosen because it la a reasonable value and a precedent exists in the University of 511chigan T. S.
The worth of a fuel assembly in a central core position is approximately
+1%.
Past practice has been to load fuel and experiments with the rods at 30%
wit hd rawn.
For most loadings, this provides considerably more margin than 3%.
Spec ification 3.1. e is new.
The old T.S. had no precautions pertaining to control blade re= oval.
The removal of all adjacent fuel (4 assemblies) from each blade removed insures that the reactor will rc=ain well subcritical.
Inc id en tally, it is physically impossible to remove a control blade unless the adjacent fuel is first r emov ed.
Specification 3.3.1.b allows radiation nonitor alarm points to be determined locally.
The old T.S.
have specific values stated.
Frequently, experi ents, waste storage, etc. can cause an area monitor to be in alarm for extended periods of tbse even though the area is barricaded and posted.
Control of radiation areas is well defined in 10CFR20 and needs no elaboration in T.S.
The primary function of area monitors is to warn personnel of unanticipated radiation increases.
This function is romewhat abrogated if there is no local control of alnra settings.
Specification 3.6 airborne effluents; historically there has been considerabla confusion, with regard to the interpretation of the airborne ef fluent ltnits as sta ted in the old T. S.
It was decided to start again from scratch and derive new limit s.
A copy of t's analysis which supports the stated specifications is appended to the enclosed SAR update. The new limits are greatly reduc ed f rom the oid limits.
Socinal release rates have historically been considerably below the st ated limits.
Specificatian 3.7 have been added to formalize past practices to insure compliance with applicabie state and federal regulations.
M
Specification 3.3.f does not a ppea r in the old T. S. bu t wa s a d d ed t c insu re that the reactor is not operated as a test reacter as defined in ICCFR50.2(r).
Specification 3.8.g is dif f erent f roc the old 1.S.
The old LCO f or irradiatict of explosives was written to fit a particular radiography f acility.
This facility is no longer installed in the reactor and probably never vill be.
We vculd like the new T.S.
to be flexible enough to allow future irradiations of explosives under very limited conditions. We belicie the verding of the new T.S. vill insure that no 52zard to the reactor vill be created by explosives.
The pcssessien linits a re the sa=e as the old T.S.
Specification 3.3.1 was =odif ied.
The old T.S. alleved specific arcunts cf fissile material to be irradiated. This did not see: to be a 1cgical way to write the specification since s=all amounts of fissile =aterial in a high flux f acilit; and a poorly designed expert =ent could cause large probless while large quantities of fissile material in a low flux facility and a properly designed experiment could be very saf e.
We believe the wording of the new I.S.
is a reasonable solution to the probles.
Specification 3.9.c pertaining to fuel burnup is worded a lot diff erently than it was in the old T.S.
Obviously the intent of the specification was to provide inspection of the first fuel asse:bly to recch 15,C00 Mv-days / tonne burnup.
This presents a proble= in fuel nanage=ent since a fuel asse=bly cust be alleved to decay at least six conths before it is saf e to disasse:ble with existing techniques and equip:ent.
The new specification per=its other high turnap assemblies to remain in the core for up to six onths pending the results of the inspection. Assuming 3 shif t 5 day /veek operation, this cculd result in burnups of up to 15,800 Mv-day / tonne by inspecticn thne.
Since there has never been a de onstrated threshold for damage at 15,000 Mv-days / tonne, this is a practical approach to the problem.
Ref erence license a= end:ent No. 10 dated May 29, 1975.
Section 4.0 of the new T.S.
(Surveillance Requirements) is considerably dif f erent f rom the old T.S.
This is due to the f act that the old T.S. vere weak in this area and many additions were required in order to confor= to th. ASS standard 15.1.
The new specifications provide considerably nore depth and clarity in the area of surveillance.
The last paragraph of section 5.4 describes an e:ergency pool fill syste=
that can be manually activated to add untreated city water to the pool.
The oli T.S. describes this system as an emergency core cooling systea that operates automatically upon a loss of pool head pressure.
This systet is a holdover frc=
the original MTR fueled reactor with a design capability of 5 Mv.
The loss of coolant accident evaluated in the Pulstar Saf ety Analysis Report clearly states that there vould be no fuel or clad da= age even if all coolant were lost i==ediately af ter an infinite period of full power operation.
Clearly, the system is not needed to protect the general public.
However; the system as it ncv exists, does pose a significant hazard to the f acility, its personnel, and possibly to persons outside the contai=:ent building.
e As it s tand s, the systes is a 7.62 cm (3") diameter pipe conducting water directly to the pool f rom the city water =ain with enly one connally energized solenoid valve preventing flow.
The valve and control systen is designed to fail saic (water flow) if there is a loss of power or systet
=alfunction.
if this happens, water is added to the poc1 at approxinately 9.46 liters /sec (150 gpo) and will overflow the pool in about 5 minutes.
The system is connected to an emergency generater; however, if there is a short circuit, the generator fails to start, the solenoid ceil burns out, etc.,
the water comes on.
If this were to happen on a weekend or holiday, it cculd exist for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> before a security patrol would find it and even longer before someone qualified could reach the scene to turn it ef f.
The result (assuming 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />) would be 85,170 liters (22,500 gallons) of conta=inated water in the f acility and a very large a=ount of egaipment da age.
We, therefore, propose to =aintain most of the systems functions and eliminate its shorteccings by re=oving the solenoid valve and controlling the syste= with a manual valve.
Specification 5.7.2 pertaining to irradiated fuel stcrage has been codified.
The old T.S. section on this subject was written to allow storage of only 24 fuel assenblies in the hot cell.
(Ref erence license maendment #14 dated June 24, 1977).
At that tbse, authority to store fuel in the hot cell was urgently needed so that leaking primary pipes could be replaced.
Since we were shipping spent fuel at the time, only 24 storage cells were needed and asked for.
If we ever had problems in the lower tank in the future, eur present lic ense does not allow suf ficient out of pool fuel storage to re=ove all fuel f rem the tank.
Since we know from experience that we have the capability for such stcrage.
it seems prudent that our license should allow m2ch storage so that future repair operations would not require rush license enenanents.
Our experience in this area during the winter of 77-78 was most favorable.
We transferred the 24 fuel assemblies in and out of the hot cell several times with no difficulty.
The radiation levels outside the cell were barely ceasurable and there were no temperature or contamination probicas. We possess a heavily shielded fuel transf er cask that can be used to transfer fuel f rom the ocol to a storage facility other than the ' hot cell if that option should ever be desiruvie.
We plan to develop analysis, equipnent, and procedures that would per=it us to resove all fuel from the pool on short notice in case that would ever be needed.
The wording in the new T.S. would allow us to do this without additional license a= endm ents.
Section 6 (Administration) has little rese:blance to anything in the old I. S.
There is nothing in the new T.S. that conflicts with the old; however, in order to satisfy the requirements of ANS Standards 15.1 and 15.18, a great deal of new material had to be added. Most of the new material pertains to audits, reporting requirements, and record keeping.
A lot of material in the old T.S. has been deleted from the new.
For the
.most part, this was descriptive material of systems and f acilities that have little or no bearing on reactor saf ety.
In the past, the detail in the T.S. have been a frustration to'both the f acility and the NRC because they o' ten conflicted with ef forts to perf oon maintenance, replacement, and modernization of equipcent.
. s SAFETY ANALYSIS REPORT UPDATE This document is a review and commentary on our Saf ety Analysis Repo r t.
It pertains to the following two documents -
Hazards Summary Report - Revision II - September 23, 1963 Additions to Hazards Summary Report - Revision II - February,1964.
The information that follows is intended to provide a detailed critique of the Facility SAR.
This information may point out errors, changes, more recent data, or simply information in support of the original document. Any portions of the SAR not specifically discussed in this document should be assumed to be valid as it is.
II B Site The site description is essentially unchanged. Only minor demographic shif ts have occured due to the expansion of suburban communities. The Main Street Campus and its surrounding residential neighborhood is very much as it was when the reactor was built. The State University at Buffalo has grown considerably since 1960; however, this growth has taken place on a new campus located approximately 4 miles north of the facility.
II C Meteorology The meteorological data in the SAR was quite old and so more current data was obtained. Appendix A contains plots and tables of this current data.
No significant changes in weather paterns for the area were discernable.
II D Hydrology II E Seismology Dr. Chester Langway, who is Chairman of the University Dept. of Geologi al Sciences, was asked to review the SAR sections on hydrology and seismology.
He assured us that the original information is correct and currently still valid. His only additional comment was that this area has never recorded an earthquake greater than 5 on the richter scale (Attica 1929).
III A Building - General The medical facility mentioned on page 34 was never constructed.
In its
- place was built a large volume irradiation facility.
(Ref. license amendment
- 7 date 6/19/64)
III B Building - Neutron Deck The N-16 hold tank that was originally buried in the carth had a concrete vault built around it.
(Ref. license change #34 dated 12/19/69)
e
- 111 F
Monitoring of Air Discharge The description of the effluent system to the stack is not entirely current.
In 1964 at the time of PULSTAR conversion, four more fume hoods and the dry irradiation f acility were added to the stack exhaust systen.
In order to accommodate the increased load, a booster fan was installed in the facility basement.
It is in series with the fan in the power hcuse.
Both fans draw at their maximum capacity and no attempt is made to control the negative pressure in the underground exhaust duct.
The effluent monitor sensitivity quoted in the SAR is that quoted by the equipment manufacturer. Actual measurements are at variance with this value. A graduate student has experimentally characterized our ef fluent system and the results are well documented. (Ref. " Reactor Air Monitor Calibration" by F. A. Tibold,1975)
Monitor sensitivities will change from time to time as G.M. tubes age or are replaced.
In order to justify the effluent concentrations in the proposed new technical specifications, the calculations of dispersion from the stack has been upgraded.
This information appears as appendix B to this document.
The general building air duct is as described except that the monitor detectors are not located in the' 36" exhaust duct. When in this location, they were significantly influenced by direct radiation from the reactor.
They are located in a 30 liter shielded container through which a sidestream of effluent is drawn. Again the measured sensitivity is not as stated in the SAR.
IV A
Reactor - General Description The dry irradiation f acility is.not mentioned. The 12" square beam tube and one of the 6" round beam tubes were eliminated to make possible the replacement of primary piping.
(Ref. license amendment #16 dated 2/16/78)
IV B
Reactor Tank The list of tank penetrations should read as follows -
1.
Five, six-inch round beau tube ports radiate from the core around the lower tank section.
2.
One. Pneumatic conveyor system enters near the top of the upper tank and terminates at the upper edge of the core.
3.
The primary coolant exits the tank through what was formerly a 12" square beam port.
4.
The primary coolant returns to the pcol through what was formerly a 6" round beam port.
5.
. A pass-through canal provides a passage between the upper portion of the tank and the hot cell.
. 6.
Eight emergency pool fill nozzles are located in the lower tank section, just below the step.
(Ref. license change 24 dated 5/31/65, amendment #12 dated 7/27/76, and amendment #16 dated 2/16/78)
IV C Core and Support Structure The core is no longer supported by the ef fluent pipe but by four legs.
(Ref. License amendment #16 dated 2/16/78) The lead thertml shield mentioned has not been used since the PULSTAR conversion and the building of the dry irradiation facility.
IV E Cooling and Purification The core support & N-16 vault previously discussed. Since the primary pipe replacement in 1978, the primary flow rate is 1150 gpm and the core differential temperature is 11.70 F.
The clean-up demineralizer piping is entirely within the pumproom.
(Ref. license change #24 and amendment #12)
IV F Shim Safety Rods & Drives The stroke of the driv (s is 26 inches not 24.
(Ref. license change #2 dated 7/24/64)
The average rod worth in a typical core as measured is more like 2-1/2%
delta K/K as opposed to the stated 6%.
Rod drive speed has always been 3" per minute in the PULSTAR core and is so stipulated in the proposed Tech-Specs.
The limit on the transient rod stroke should be associated with 44 Mw-sec pulse, not 90.
(ref, old Tech-Specs.).
IV G I ns t rumen tat ion Table XXIII should be made to read the same as Table 3.1 in the proposed new Tech-Specs.
The control instrumentation is essentially as described in the SAR; however, approximately one half of the original tube type equipment has been replaced by newer solid-state type instruments.
The new instruments perform the same functions as the old at a performance and reliability level superior to the original.
. IV 11 Experimental Facilities The loss of 2 beam tubes was previously discussed.
The medical f acility should be deleted and the dry chamber characteristics inserted in its place.
V A Temperature Coefficient The moderator and reflector temperature coefficients were measured independently during the initial low power testing program. The results indicated a moderator coefficient of
.008% delta K/K/oF and a reflector coefficient of +.004% delta K/K/oF.
The bulk coef ficient would then be
.004% delta K/K/of thus requiring approximately 0.15% delta K/K to compensate for the rise in temperature from ambient to normal operating.
(Ref. Report WNY-017 dated 10/9/64)
V C Beam tubes, pneumatic conveyors and dry chamber nosepiece. Change tabulation as follows -
Five 6 inch beam tubes 0.12% delta K/K each Tuo 2 inch rabbit tubes 0.03% delta K/K each One dry chamber nosepiece 0.50% delta K/K TOTAL 1.167 delta K/K V E Equilibrium Xenon Measurements indicate an equilibrium xenon reactivity effect of 1.6%
delta K/K after 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> of operation. The post operation peak has a value of 1.7% delta K/K four hours af ter shutdown.
V H Thermal Column The lead shield for the thermal column has not been used since PULSTAR conversion. That side of the core is water reflected.
V I Pulse Reactivity Requirements The measured reactivity inputs necessary to produce the routine pulse of 35 Mw-secs, and the maximum pulse of 44 Mw-secs, are 1.5 and 1.7% delta K/K respectively.
(Ref. NSTF pulse log book)
V J Long Term Operation Experience has shown that in the steady-state mode, burnups of 15,000 Mw-days / tonne or more can be tolerated. For pulsing,10,000 Mw-days / tonne is practical limit with a 4 x 5 core.
(Ref, license amendment #10 dated 5/29/75)
-S-V K Typical Core Loading and Control Rod Effects The typical core loadings generally contain between 22 and 27 fuel assemblies depending on burn-up and experimental needs.
Typical measured rod worths are as follows -
Rod # 1 1.4% delta K/K
- 2 1.5
- 3 3.0
- 4 3.4
'# 5 1.5 Pulse 1.8 Sum of 1 to 5 12.6 Typical shutdown margins are in the order of 7 or 8% delta K/K.
Typical excess reactivity needs are as follows -
Xenon overide 1.70% delta K/K Power defect (0-2Mw) 0.35 Burn-up 0.45 Experiments 1.50 TOTAL 4.00 V1 Personnel and Organization Omit the position of general manager and citange the Nuclear Hazards Committee to Nuclear Safety Committee.
VIII B1 Loss of Ventilation Due to the loss of two beam tubes, the release of all Argon-41 to the reactor room would result in a concentration of 4.6 x 1010 C1/cc instead of 5.7 x Ig10 C1/cc.
Ar-41 f rom the dry chamber is not considered. While production of Ar-41 in the facility is calculated to be 1.25 mci / min when in use, the facility is well ventilated and so no significant inventory would be expected to accumulate.
(Ref. Letter to NRC dated 4/15/64)
VIII B-4 Loss of Pool Water The gross loss of all pool water should be considered the worst credible accid en t.
As previously stated, this would not result in loss of fuel-integrity, but would result in e<tremely high radiation levels within the building.
The hazard would be shot t term because the core could be recovered by flooding the lower level of s containment vessel if necessary.
a
~VIII B-5 Maximum Start-up Accident, A re-evaluation of..iis 4 ident is enclosed in the appendix C.
The original analysis was bad.d on erroneous rod speeds and rod worths. Also, the assumption that the reactivity is inserted as a step does not result in a worst case analysis.
s VIII C
, Maximum Credible Accident This accident analysis should be renamed - design basis accident.
It is considered well beyond the real= of credibility for several reasons.
However, it does serve to_ demonstrate the engineered safety syste='s capacity to deal with gross loss of fuel integrity, no =atter what the
.cause may be.
Appendix l-D-1 Steady State Heat Transfer Analysis in order to determine " limiting conditions for operation" and
" limiting safety system setting" for the revised tech-specs, it was necessary to further evaluate the thereal-hydraulic perfor=ance of the NSTF core. This evaluation is contained in Appendix D.
This analysis includes a justification for established safety limits, and limi*.ing safety system settings, as well as an examination of fuel perfor=ance during a loss of flow accident.
APPENDIX A
Meteorology B
Justification of Airborne Effluent Limits C
Start-up Accident D
Thermo-Hydraulic Analysis
i Appendix A Meteorology Examination of weather history up to the end of 1978 reveals that there has been no significant climatological change for the 58TF site since the last Hazards Summary Report was filed for this facility in 1963.
Tables I through IV illustrate pertinent data summaries.
Figures I and II present a detailed view of wind conditions in Buff alo during 1977 and 1978. Wind calms are rare phenomena in Buf f alo, being recorded in less than 0.3% of all observationc.
Data used were supplied by the National Oceanic and Atmospheric Administration. Observations were recorded at the Buffalo International Airport, located s' miles from the site.
I Table I MONTHLY WINDS, 1969 THROUGH 1978 Normal Mean Average Max.
Prevailing MONTH M. P. H. ([3 M. P. H.
M. P. H. dd Direction J
14.6 14.4 60 W
F 14.1 13.1 47 WSW H
13.7 12.7 56 WSW A
13.0 12.3 51 WSW M
11.8 10.8 47 SW J
-11.2 10.4 37 SW J
10.6 10.2 39 SW A
10.0 9.0 41 SW S
10.6 9.8 41 SW 0
11.4 10.5 38 SW N
12.9 11.7 47 SWW D
13.5 12.4 49 WSW
([) 1941 - 1979 period (D >1 minute duration
Table II PRECIPITATION, 1970 THROUGH 1978 INCHES WATER EQUIVALENT Inches Snow MONTH Normal (l)
Max. Monthly Min. Monthly Max. in 24 hrs.
Max. Monthly J
2.90 6.47 1.03 2.46 68.3 F.
2.55 5.80 0.81 2.31 54.2 M
2.85 5.59 1.20 2.14 29.2 A
3.15 5.90 1.27 1.17 15.0 M
2.97-6.39 1.21 2.03 2.0 2
J
. 23 6.06 0.11 3.04 0.0 J
2.93 6.43 0.99 3.38 0.0 A
'3.53 10.67 1.10 3.88 0.0 S
3.25 8.99 0.77 3.63 T
0 3.01 9.13 0.30 3.49 3.1 N
3.74 6.37 1.44 2.51 31.3 D'
3.00 8.02 0.69 2.16 60.7-([)Il941through1970
Table III TEMPERATURES, F, 1941 THROUGH 1970 Normal Normal Record Record MONTH Daily Max.
Daily Min.
Max.
Min.
J 29.8 17.6 72
-12 F
31.0 17.7 64
-20 M
39.0 25.2 81
-4 A
53.3 36.4 87 12 M
a64.3 45.9 90 26 J
75.1 56.3 95 35 J
79.5 60.7 94 43 A-77.6 59.1 99 38 S
70.8 52.3 98 32 0
60.2 42.7 87 20 N
46.l' 33.5 80 9
D 33.6 22.2.
66
-4
r l
Table IV SKY COVER, 1943-1978 MEAN NUMBER OF DAYS MONTH Clear Partly Cloudy Cloudy J
1 7
23 F
2 5
21
-M 4
7 20 A.
5 8
17 M
6 9
16 J
6 12 12 J
7 13 11 A
7 12 12 S-7
'9 14
-0 7
8 16 N
2' 5
23
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APPENDIX B Justification of Airborne Effluent Release Limits for the State University of New York at Buffalo Nuclear Science and Technology Facility 1.0 Introduction There are two airborne effluent release points for the NSTF.
The first point is a 50.9 meter stack which resides inside the coal boiler plant release duct, immediately behind the facility. This duct is co=monly referred to as the
" stack".
It is used to exhaust air from high level fume hoods and various experimental facilities.
(For details, see Figure 1. ) The second release point is commonly referred to as " building air".
It is a 36-inch duct which exhausts air through the containment building roof.
It is used to exhaust low level hoods and the general " breathing air" from the containment offices and bay areas.
Airborne radioactive material concentration limits are presented for each release point, subject to the limitations imposed by 10CFR.20 and Part 16 of the New York State sanitary code.
Federal regulations allow for averaging of release levels for a period of up to one year. The limits established herein are based on more restrictive criteria, as follows:
1.1 For the stack, limits are established such that the maximum ground level concentration of radioactive materials shall not exceed FTC for unrestricted areas, as specified in Appendix B, Table 2, of 10CFR.20 at any time. A calculational uncertainty factor of three is assumed.
1.2 For building air, limits are established such that the maximum airborne radioactive materials concentration shall not exceed twice the occupational limit at any time. The yearly average concentration shall be less than the occupational limits. Resulting concentrations outside the containment shall be less than the limits for unrestricted area s.
2.0 Calcular
ions to support stack limits 2.1 Calculation of effective stack height An " effective" stack height is generally employed to make allcwance for t
~
upward momentum and for thermal buoyancy effects. Thus the stack heigat (h)
~
is equal to the actual height of the stack (h
) plus a second term (ah).
i.e..
h=h
+ Ah.
act A momentum source is assumed since the exit velocity is appreciable
(>10 m /s) and there is little temperature excess (<50 C above ambient).
Additional momentum and mixing from the releases of coal ' plant gases during winter are conservatively. disregarded.
It-is therefore assumed that:
1.4 3,
Ah =
(ref. 1.)
6 i
l
- TABLE 1 Effective Stack Ave. Wind Ave. Wind Height (m)
Speed (m/s)
Speed (mph) 115.9 0.5 1.12 75.5 1.0 2.23 64.9 1.5 3.36 60.2 2.0 4.47 57.7 2.5 5.59 56.2 3.0 6.71 55.2 3.5 7.83 54.4 4.0 8.95 53.9 4.5 10.07 53.5 5.0 11.18 53.2 5.5 12.30 52.9 6.0 13.42 52.7 6.5 14.54 52.5 7.0 15.66 52.4 7.5 16.78 52.2 8.0 17.90 52.1 8.5 19.01 52.04 9.0 20.13 51.95 9.5 21.25 51.9 10.0 22.37 51.5 15.0 33.55
. where D
stack diameter in meters
=
s=
vertical efflux velocity in m/s 6= mean wind speed in m/s At a flow rate of 6000 CFM, with a stack diameter of 18 inches, the following equation results:
V = 17.245 m/sec.
3 D =.4572 m h
= 50.9 m 1.4 I 245 (1) h = 50.9 +.4572 1
U Equation (1) was used to determine effective stack heights for wind speeds from 0.5 to 15 m/s as presented in L 5le (1).
2.2 Meteorological Conditions Four classes of meteorological conditions are considered:
- stable, neutral, unstable, and very unstable.
(Reference 1).
The primary importance of such classifications is the approximation of crosswind and vertical plume standard deviations (cy and az respectively).
For simplicity it is assumed that the basic " power law" equations apply; b
i.e.
o=ax, where o is the standard deviation, x is the distance downwind from the plume origin, and a and i are constants as specified in Table II.
These equations are generally considered adequate up to a dawnwind distance of 10,000 meters.
TABLE II Meteorological oz oy Condition a
b a
b fVeryUnstable t
0.4 0.91 0.4 j
0.91 l Unstable 0.33 0.86 0.36 j 0.86
' Neutral 0.22 0.78 0.32 ;
0.78 Stable 0.06 0.71 0.31 0.71 i
i
. 2.3 Basic Dispersion Equations A simple form of the Caussian Equation is used (2)
X (X,Y,0)
- eXP (
2
+
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> z no o,u zj
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=
at x meters downwind a.4 y :,eters crosswind.
Q Release Rate (Ci/sec).
=
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= Crosswind and vertical plume standard deviations (m)
U
= mean wind speed (M/sec) h
= effective stack height (m)
For locations directly downwind (y=o), equation (z) reduces to:
(3)
X (x,0,0) exP
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X o
=
=
z Equations (2), (3), and (4) may be divided on both sides by Q.
The resulting left side quantities are normalized concentrations in units of 3
(Ci/m )/(C1/sec), and hence represent the resulting concentration at ground level in Ci/m3 (or g ) per each C1/sec of. material released. The convenience of this ce normalization shall become apparent in subsequent sections of this analysis.
. 2.4 Calculations 2.4.1 Maximum (worst case) Concentration X
Both max and the corresponding X' were calculated for the four Q
meteorological conditions presented, at wind speeds up to 15 m/sec.
Results are presented in tabular and graphic form in Tables IV and V.
The limiting condition was identified at NO.9 m/see wind speed under very unstable meteorological conditions.
max under such conditions is equal to 4.12 x 10~ @ an X'
= 230 meters.
Q To satisfy our previously presented criteria, Xumx is set equal to 3
MFC in Ci/m.
Therefore, Q MPC/4.12 x 10~
max =
4 Q,
{2.43 x 10 } - MPC
=
Thus, for a given isotoge the maximum allowable release rate is equal to its MPC x 2.43 x 10, if ground level concer.trations are not to exceed FTC.
Inserting a conservative safety factor of 3. yields a maximum allowable release rate in (Ci/sec) of:
3 (8.1 x 10 ) x (MPC) 2.4.2-Population Exposures The preceding release limit is based upon worst case conditions.
Using this limit, a maximum credible exposure concentration nay be estima:ed under predominant meteorological conditions.
Meteorological data for the years 1968 to 1978 was obtained from the National Oceanic and Atmospheric Administration, Environmental Data and Information Service. This data was compared to data compiled to support the original NSTF Hazards Summary Report, Revision 2,1964. No significant changes were detected (see Appendix A).
The average wind speed is approximately 5.11 m/s ('1.44 mph).
max is calculated 9 this wind speed for the four meteorological Q
conditions, and presented in Table III.
_. TABLE III I
fnditon Xmax x'(m) max 0 Sax 9
'Very Unstable 1.55 x 10-148
.126 x MPC Unstable 1.47 x 10-250
.119 x MPC I
Neutral i
1.1 x 10' 730
.089 x MPC f
i
-5 stable 3.1 x 10 8,700
.025 x MPC i
The predominant wind direction is SW with a persistence of 22 per cent.
Projected yearly equivalent exposures are therefore between.55 per cent of MPC and 2.77 per cent of MPC, if the NSTF were to continuously release
.@. (x.
,m
- j-2.5 Multiple 1sotope Releases The previous calculations may be easily extended for the release of more than one isotope at a tice.
In this cade, the sum of the ratios 3
of each isotope's release rate to its MPC must be less than P.1 x 10 ; i.e.:
9 I 8.1 x 103 4
MPC.L t=
1 where N = number of isotopes being released.
0i = Release rate of i isotope (Ci/s)
'TC.
.th
(
= FTC (unrestricted) of the t isotope (uc/cc) 3.
Building Air Release Limit For the building air ventilation system, the limiting safety crite l.
for airborne radioactive materials is exposure to workers within the containment vessel, not the resulting exposures to the public. The following limits on building air effluent are therefore established.
3.1 Yearly Average:
N C.1 MPC
>< 1 s
t R.#
i=1 3.2 At any time:
N C i MPC g
R'.
i=1 th where C
= concentration of i isotope (uc/cc) in building air exhaust ch MPC
= Restricted MPC of the i oe %
R.4 cc number of isotopes in effluent.
N
=
. Under routine circumstances, the only measurable isotope in the building air exhaust stream is Ar-41.
The nominal concentration of Ar-41 in the building air is below MPC.
On occasion, however, the concentration does exceed FTC forshorEperiodsoftime. Operating procedures stipulatethatkhereactorshallbeshutdownifbuildingairexceeds 2 x MPC
- R Occupational exposures to airborne radioactive materials are considered acceptable if the release limit as stipulated in 3.1 is met.
Workers at NSTF seldom, if ever, spend a full 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> / week within the containment vessel.
Anticipated exposures are therefore below 10CFR 20 limits.
Because of the geometry of the building air release point cnd because of the influence of the nearby positive-draft cooling tower, standard calculational models for dispersal of radioactive effluent will not yield meaningful results. However, an examination of the magnituies of the concentrations involved suffices to ensure compliance with applicable safety standards.
For Ar-41, a dilution factor of 50 would be required to meet unrestricted MPC at ground icvel, if the concentration at point of release was 2 x MPC
- R The prevailing wind direction is 22 per cent persistent.
Extensive occupancy of areas immediately adjacent to the release point is not anticipated.
If the yearly limit, as in limit 3.1, is met, with a wind persistence of 22 per cent, a dilution ratio of about 5.5'would be required to protect members of.the public. The distances involved, coupled with feasible occupancy factors, will ensure a better than 5.5 dilution factor.
CONCLUSIONS:
The following release limits are herein established:
For the Stack (notation as before) 1.
Yearly Average N
Q /MPC I 8.1 x 103 i=1 2.
Instantaneous
( 3 x) the above.
The duct has a cap to prevent rain from entering the duct.
_ =...
_9_
For Building Air (notation as before) 1.
Yearly Average N
C i MPC g
R7
.L = 1 2.
Instantaneous N
i Rj L=1 It is concluded that compliance with these limits will adequately protect both the public and NSTF occupational workers.
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Wind Speed Very Unstable Unstable Neutral Stable (m/s) 0.2 1.72 1.58 1.18 0.191 0.5 3.49 3.20 2.40 0.387 1.0 4.11 3.76 2.82 0.456 1.5 3.71 3.39 2.55 0.411 2.0 3.23 2.96 2.22 0.358 2.5 2.81 2.57 1.93 0.312 3.0 2.47 2.27 1.70 0.274 3.5 2.20 2.01 1.51 0.244 4.0 1.98 1.81 1.36 0.219 4.5 1.79 1.64 1.23 0.199
.5.0 1.64 1.50 1.13 0.181 6.0 1.39 1.28 0.959 0.154 7.0 1.21 1.11 0.835 0.135 8.0 1.07 0.985 0.739 0.119 9.0 0.96 0.881 0.661 0.107 10.0 0.87 0.797 0.598 0.096 15.0 0.59 0.540 0.405 0.065 P
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TABLE V X@Xmax (*
Average Wind Very Unstable Unstable N tcal Stable Speed (m/s) 0.2 847 1,566 5,600 81,483 0.5 347 609 1,978 25,982 1.0 217 370 1,142 14,200 1.5 183 311 940 11,484 2.0 169 285 854 10,330 2.5 161 271 809 9,731 3.0 157 263 782 9,377 4.0 151 253 750 8,957 5.0 148 248 734 8,749 10.0 144 239 706 8,383 15.0 142 237 699 8,292
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( */5 4
's APPENDIX C UPDATE: Fbximum Startup Accident For the purpose of evaluating the maximum startup accident, it is assumed that all control blades are driven out, and that no protective action is taken until the high power scram is tripped at 2.4 mw.
The maximum control blade with-drawat rate is 3 inches per minute, which corresponds to a maximum reactivity insertion rate c. 0.038% AK/K per second (.76% AK/K/ inch), at the point of maximum differential worth.
The minimum induced period may be estimated by the following correlation:
fn \\ p/N N
L l_
7 2
2a8 7
k where:
N
= Reactor Power at termination of p
ramp reactivity insertion N
= initial reactor power a
= Reactivity insertion rate in ($/sec)
S
= Beta Effective Substituting the following values:
N
= 2.4 x 106 g, p
li
= 1.0 x 10 3 y, y
f
= 3 x 105 sec (a8 )
=.038 % AK/K yields a period of 47.5 msec.
Reference:
Reactor Handbook, Volume III, H. Soodak, 1962, J. Wiley and Sons
- As measured during the pulse test program.
Two pulses, performed during the pulse test program, produced periods which bracketed the above calculated value:
Data for these pulses were:
T(msec) Peak Power Energy Temp.
Pulse 28 57.9 7.05 mw 2.5 mw-s 2620F Pulse 24 37.6 13.9 mw 2.8 mw-s 271 F Thus, the energy released and temperatures associated with the maximum startup accident are modest, and pose no hazard to personnel, the public, or the reactor core.
- Maximum surface temperature of test pins.
4 9
m__
APPENDIX D Analysis to Support Safety Limits and Safety System Settings A.
History A thermal hydraulic analysis of the NSTF PULSTAR core was performed in 1963, and is presented in Revision II of the NSTF Hazards Summary Report.
Since 1963, a pulse test program was carried out at NSTF. Much has been learned about the thermal-hydraulic behavior of the PULSTAR core, as a result of these tests.
In particular, it has been demonstrated that boiling in the coolant outlet channel does not jeopardize fuel or cladding integrity. The analysis presented herein, provides justification for the Limiting Safety System Settings and Safety Limits for the NSTF PULSTAR.
B.
Forced Convection 1.
Limiting Criterion The limiting criterion for safety is the assurance of fuel / cladding integrity.
For the purposes of this analysis, fuel cladding integrity is considered compromised if fuel centerline melting, or departure from r.ucleate boiling (DNB),occure.
2.
Safety Limits The following Safety Limits are herein established:
a.
Core Power 3.3 Mw.
Minimum height of H O above top of core = 17 feet.
b.
2 c.
Maximum core inlet temperature = 140 F.
d.
Minimum Flow = 1000 GPM.
3.
Safety System Settings The following Safety System Settings are established:
Power - Control Blade Reverse 1 2.2 Mw a.
b.
Power - Control Blade Scram 5 2.4 Mw c.
Minimum Height above Core 2
20' d.
Minimun Flow 3 1080 GPM 0
126 F.
e.
Pool Temperature Scram
=
a 4.
Parameter Variation Pertinent process variables have been established as:
1.
Depth of H O above core 2
2.
Primary coolant flow rates 3.
Core power level 4.
Coolant inlet (pool) taperature.
It is asstaned that parameters 1.,
2., and 4. are at their safety limits.
The actual flow through coolant channels shall vary with the number of fuel assemblies being operated (hereafter "N").
As N increases, channel flow decreases. This is offset (but not linearly) by a reduction of, the specific power (and hence heat flux) of each assembly. There fore, in this analysis N will be varied from 16 to 35, and the reactor power level will be determined for each N, corresponding to critical (DNB) heat flux.
The coolant channel shall be as indicated in. figure I.
Channel Data:
D,
=.0123 ft. (equivalent hydraulic diameter)
-4 A,
= 4.243 x 10 f t.2 (area of channel) g.
heated perimeter /n D
=.01975 ft.
=
5.
' Correlation The following correlation is used to estimate critical heat flux (q" crit)'.
q". crit
~.h O. crit ~ bulk)
=
crit-v 1.8 57 in(P) - 54 w, crit.
P+15
+ 32 T
=
e 48v
.h 0,8 M D u.6 for D,s: 0.1 ft.
=
crit.
D + D):
t e
L e where-fP system pressure (PSYA)
=
coolant velocity (f t/sec)
V
=
h
- " " * * "" *"' I
- ~
)
erit.
T, g = critical wall t aperature, F
~ D,, D -
W au He A ameter, Mt.)
=
g Bernath, :L., "A-Theory of-Local! Boiling Burnout"
~ Heat Transfer Sympositan, AICHE,1955.
P e
ye
--g-v - us n
,rs em-r--e-r-
~
w
6.
Calculations of Maximum Power 6.1 Flow velocities through the fuel have been calculated at 1000 CPM with N varying from 16 tbtough 35.
(Table 1).
6.2 Flow velocities at assembly inlet, were also calculated (V ) and tt:en used to estimate (P)from the following equation:
=
P in
+
14.7 144 where Ah distance from pool surface to hot spot (% fucl centerline)
=
density (evaluated @ 150 F) p
=
6.3 With P and V calculated, T h
and q" can also be 1
g calculated using a T f 1500F.
Bulk Results are shown in Table 2.
6.4 Once q" crit has been calculated, the maximum core power is cal-culated from:
(N)(q" crit
(*
p 2Fq where F
heat flux hot spot factor
=
f 3.5 (see hazard summary report)
=
(6.52) =
heat transfer area of a fuel assembly.
This assumes that correlational error may be as much as 30%,
and uses a DNB ratio of 2.
6.5 The centerline fuel temperature was then calculated for the hot channel, using the correlations establishcd on pages 138 and 139 of the Hazards Summary Report, Revision 2, at the core powers calculated for P The values of P and T -(Fuel centerline temperature) are presented in table III.
Values of T, T (surface of fuel),
9 0
Ts (inn r clad) and Tp (outer clad) are presented in Table IV.
TABLE 1.
N V(ft/sec)
V (ft/sec)
P(PSIA) f 16 5.685 2.906 21.27 20 4.548 2.325 21.65 25 3.638 1.860 21.91 30 3.032 1.550 22.04 35 2.599 1.329 22.12 TABLE 2.
w, crit ( F)
N T
hcrit hr-ft2 F q"
hr-ft' 16 286 7,999.
1.088 x 106 20 288 7,235.
9.984 x 105 25 289 6,624.
9.207 x 105 30 290 6,216.
8.702 x 105 35 291 5,926.
8.356 x 105 TABLE 3.
nax cax C
16 1.135 x 107 3.31 MW 3794
=
2'O 1.302 x 107
' =
3.81 MW 3322 25 1.501 x 107 4.39 MW 3263
=
30 1.702 x 107 4.98 MW 3096 35
.1.907 x 107 5.57 MW 2976
=
TABLE 4 N
T ( F)
T (OF)
T (OF)
T ( F) 9 g
F 16 3794 1834 285 200 20 3322 1619 272 198 25 3263 1594 274 202 30 3096 1518 271 202 35 2976 1464 268 202 C.
Natural Convection 1.
Limiting Criterion The limiting criterion for safety shall be the same as in forced convection.
2.
Safety Limit.
2.1 Power 1 500 KW 3.
Safety System Settings.
3.1 Power = 1250 KW.
4.
Justification No analysis is needed to justify these limits. Limits are establishem based on actual testing, conducted in 1966. These tests demonstrated that PULSTAR fuel can'be operated in natural convection mode at powers in excess of 1MW without damage. A copy of Technical Note J-435 is appended to this report.
State University of New York at Buffalo
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State University of NewYork at Buffalo NUCLEAR SCIENCE AND TECHNOLDGY FACILITY REV.
ISSUE D BY DATE SHEET OF CHECKEO DATE FILE v
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LOSS OF FLOW ACCIDENT:
L'pon loss of flow, the reactor will scram at a measured flow rate of 1080 GPM.
Flow decay from full flow to 1080 GPM has been measured to be about one second.
Flow will continue to decay until the flapper opens at a nominal zero flow.
This has been measured to occur at about 10 seconds after loss of flow.
Once the flapper opens, reversed natural convection cooling will connence.
For the purposes of this analysis, it is assumed that the flapper fails to open and no natural convection is supplied.
To estimate the amount of heat generated after reactor scram, the followinc correlation is used:*
Power = PM(t
,t
)
(watts) g 200 where P = steady state power (2 mw) t = seconds of constant power operation before shutdown g
seconds since shutdown t =
s s)
M(=,t ) - M(=,t
+t M(to,t )
=
s s
o M(=,t) may be represented in analytical form, with a maximum negative deviation of 9%, as:
M(=,t) = At "
~
where A,a are constants as specified in Table I.
TABLE I Applicable Ti A
a Max. Min deviation Interval (sec) 1<
1 10
- t < 10 12.05 0.069 3% 0 t = 10 1
2 10
-t < l.5 x 10 15.31 0.1807 1% 0 t = 30 2
6 3
1.5 x 10 < t < 4 x 10 26.02 0.2834 5% 0 t = 3 x 10 6
8 E
4 x 10 < t < 2 x 10 53.18 0.3350 9% 0 2 x 10
Reference:
AEC Report WAPD-BT-24, K. Shure
" Fission Product Decay Energy"
a 7
It is assumed that (tg) = 7.884 x 10 sec. (s one calendar quarter).
Decay heat generation rates are then calculated from (0,1) sec. to 600 seconds after scram.
Results are presented in Table II for total core, and per assembly for a 16 assembly core (worst case). The decay heat generation heats may be integrated a determine the total amount of heat released f rom the fuel as a tunction of time.
Pesults of these calculations are presented in Table III.
The total energy released due to the decay of fission products during the first 600 sec. after shutdown is therefore 32 mw-sec.
A second source of heat generation, neutron power, must also be calculated.
It is conservatively assumed that upon scram, the highest worth control blade is full out and remains full out.
The available reactivity for shutdown is assumed to be 3% AK/K. This number is conservative, based on control blade reactivity measurements for cores operated since 1964.
The initial prompt drop in power is calculated from the following equation :
P + S(1 - p)P(tg)
(S-p)
Reference:
Nuclear Reactor Theory, Lamarsh,1966, Addison-Wesley Pub. Co.
TABLE II Time Since Core Decay Assembly Shutdown Heat Decay Heat (sec.)
(watts)
(watts)
.1 1.37 x 105 8.56 x 103
.5 1.23 x 105 7.69 x 103 1.0 1.18 x 105 7.38 x 103 2.0 1.13 x 105 7.06 x 103 3.0 1.06 x 105 6.63 x 103 10.0 9.84 x 10 6.15 x 103 4
15.0 9.13 x 10 5.71 x 103 4
4 5.41 x 103 20.0 8.63 x 10 30.0 8.02 x 10 5.01 x 103 4
50.0 7.29 x 10 4.56 x 103 4
4 4.28 x 103 70.0 6.85 x 10 4
4.00 x 103 100.0 6.40 x 10 120.0 6.19 x 10 3.87 x 103 4
150.0 5.93 x 10 3.71 x 103 4
200.0 5.80 x 10 3.63 x 103 4
4 3.23 x 103 300.0 5.17 x 10 400.0 4.76 x 10 2.98 x 103 4
500.0 4.47 x 10 2.79 x 103 4
4 2.66 x 103 600.0 4.25 x 10
- 16 4'.ssembly Core
4 TABLE III
- +
Time since Core Heat Assembly Heat Shutdown Release Release (sec.)
(w-sec.)
(w-sec.)
1 1.26 x 105 7.88 x 103 2
2.41 x 105 1.51 x 104 5
5.68 x 105 3.55 x 104 10 1.08 x 106 6.75 x 10 4 20 2.00 x 106 1.25 x 105 50 4.35.x 106 2.72 x 105 100 7.74 x 106 4.83 x 105 150 1.08 x 107 6.76 x 105 200 1.37 x 107 8.56 x 105 400 2.36 x 107 1.47 x 106 600 3.20 x 107 2.00 x 106 Due to fission product decay
+16 Assembly core Where P
= Reactor power after prompt change P(tg) = Power before jump (2 Mw)
=(.0069)Sf 8
p
= Reactivity inserted (-3% AK)
K Substituting as specified yields a prompt downward change to 0.385 mw.
This cange will occur in a fraction of a second.
It is conservatively assumed that the reactor power remains at 2 mw for 2 sec. during and immediately after control blade insertion. After 2 sec. it is conservatively assumed that the reactor power decays at a period corresponding to the longest lived delayed neutron precursor (80 sec.).
i.e.,
after 2 sec.
6 P(t) = (.385 x 10 ) expf-t/80}
P(t) is calculated for the first 600 seconds af ter prompt jump.
Results appear in Table IV.
Also presented in Table IV are the estimated integral neutron power' released to the core. This data represeats integration of the exponential decay data plus the above-referenced 4 mw-sec. release during scram.
e TABLE IV Time since Core Neutron Integral Scram Power Power Release (sec.)
(watts)
(w-sec.)
1 3.80 x 105 4.383 x 106 2
3.75 x 105 4.760 x 106 5
3.62 x 105 5.87 x 106 10 3.39 x 105 7.62 x 106 20 2.99 x 105 1,og x 107 50 2.06 x 105 1.83 x 107 100 1.10 x 105 7
2.59 x 10 150 5.90 x 104 3.01 x 107 200 3.16 x 104 3.23 x 107 400 2.59 x 103 3.46 x 107 600 2.12 x 102 3.48 x 107 The total heat generation rates and cumulative releases are calculated by summing the data in Tables IV and III and II.
Results are presented in Table V.
As indicated, the total heat released to a cote in the initial 600 sec, of a loss of flow accident will be less than 67 mw-sec.
(<4.18 mw-sec/ assembly for 16 assembly core.)
TABLE V Time since Total Core Total Heat shutdown Heat Rate Released (sec.)
(watts)
(w-sec.)
1 4.98 x 105 4.509 x in6 2
4.88 x 105 5.001 x los 5
4.18 x 105 6.438 x 106 10 4.37 x 105 8.70 x 106 20 3.86 x 105 1.28 x 107 50 2.79 x 105 2.27 x 107 100 1.74 x 105 3.36 x 107 150 1.18 x 105 4.09 x 107 200 8.96 x 104 4.60 x 107 400 5.02 x 104 5.82 x 107 600 4.27 x 104 6.68 x 107
Converting to BTU's, the maximum heat released $n an average assembly would be 3957 BTU, for a 16 assembly core. Assuming the hottest assembly leads the core average by a factor of 2.5, the hottest assembly would generate 9,892 BTU of heat in 600 cec., or about 321 BTU /lb. of fuel. Obviously, most of this heat will transfer from the fuel even if the flapper remains closed.
- However, assuming that no acat is transferred from the fuel whatsoever, one can estimate the average temperature increase in the fuel by dividing the heat added (BTU /lb.)
by the specific heat of the fuel. The C of UO2 increases with temperature.
p It is assumed that C remains at.56 BTU /lb.oF. Thus the average temperature p
rise in the hot assembly with full heat retention by the fuel would be:
321 (BTU /lb.)
o F.
=
.56 (BTU /lb.
F.)
o Under steady state conditions, the average fuel temperature is 349 F; the maximum fuel temperature is 464 F.
Therefore, subject to the previously presented conservative assumptions, one would expect a maximum fuel temperature of approximately 10400F, and an average fuel temperature of %9220F, in the hottest assembly of a 16 assembly core, if there were no heat transfer of the fuel, since the melting point of UO2 is %5000 F, and the melting point of Zircalloy II is %3300 F.
No loss of fuel or cladding integrity is anticipated upon loss of flow accidents, in which the flapper fails to open.
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- From Hazards Summary Report, Rev. II, 1963.
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