ML20070M990

Proposed Tech Specs,Supporting Renewal of License R-97

ML20070M990
Person / Time
Site:	Neely Research Reactor
Issue date:	04/19/1994
From:	Neely Research Reactor, ATLANTA, GA
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ML20070M558	List: ML20070M556 ML20070M970 ML20070M986 ML20070M990
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NUDOCS 9405040275
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6 ATTACHMENT 4 TECHNICAL SPECIFICATIONS k

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TECHNICAL SPECIFICATIONS +

GEORGIA TECH RESEARCH REACTOR I Docket 50-160 License R-97 ,

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Table of Contents Page 1.0 Definitions . . . . . . . . . . . . . . . . . . . . . 1 2.0 Safety Limits and Limiting Safety System Settings . . 6 2.1 Safety Limits . . . . . . . . . . . . . . . . . . 6 2.2 Limiting Safety System Settings . . . . . . . . . 7 l

3.0 Limiting Conditions for Operation. . . . . . . . . . . 11 l 3.1 Reactivity Limits . . . . . . . . . . . . . . . . 11 3.2 Reactor Safety System . . . . . . . . . . . . . . 12 3.3 Containment Building. . . . . . . . . . . . . . . 16 1 3.4 Limitations of Experiments. . . . . . . . . . . . 19 1 3.5 Radioactive Effluents . . . . . . . . . . . . . . 22 3.6 Primary Coolant System. . . . . . . . . . . . . . 25 3.7 Emergency Cooling System. . . . . . . . . . . . . 26 3.8 Fuel Handling and Storage . . . . . . . . . . . . 27 3.9 Fast Shutdown System Experiments . . . . . . . . 29 4.0 Surveillance Requirements. . . . . . . . . . . . . . . 31 .

4.1 Reactivity Limits . . . . . . . . . . . . . . . . 31 4.2 Reactor Safety System Surveillance. . . . . . . . 31 4.3 Containment Building. . . . . . . . . . . . . . . 36 4.4 Primary Coolant System. . . . . . . . . . . . . . 36 4.5 Emergency Cooling System. . . . . . . . . . . . . 37 5.0 Site Description. . . . . . . . . . . . . . . . . . . 39 5.1 Specification . . . . . . . . . . . . . . . . . 39 5.2 Fuel Elements . . . . . . . . . . . . . . . . . . 39 6.0 Administrative Controls. . . . . . . . . . . . . . . . 40 6.1 Organization. . . . . . . . . . . . . . . . . . . 40 6.2 Nuclear Safeguards Committee. . . . . . . . . . . 42 6.3 Administrative Controls of Experiments. . . . . . 43 6.4 Procedures. . . . . . . . . . . . . . . . . . . . 45 6.5 Operating Records . . . . . . . . . . . . . . . . 45 6.6 Action to be Taken in the Event of a Reportable f

Occurrence. . . . . . . . . . . . . . . . . . . . 46 6.7 Reporting Requirements. . . . . . . . . . . . . . 47

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1 1.0 DEFINITIONS 1.1 Safety Limits - Safety Limits (SL) are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers against the uncontrolled release of radioactivity.

1.2 Limitina Safety System Settina -

Limiting Safety System Settings (LSSS) are settings for automatic protective devices related to those variables having significant safety runctions. Where a LSSS is specified for a variable on which a SL has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a SL is exceeded.

1.3 Limitina Conditions for Operation - Limiting Conditions for Operation (LCO) are the lowest functional capability or performance level of equipment required for safe operation of the facility. When a LCO for operation is not met, the reactor shall be shutdown.

1.4 Surveillance Recuirement - Surveillance requirements are requirements relating to tests, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within the safety limits, and that limiting conditions for operation will be -

met.

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 associated circuitry which forms the automatic protective system for the reactor or provides information which requires manual protective action to be initiated.

~1.7 Operable - Opeiable means a component or system is capable of performing its intended function in its normal manner.

1.8 Operatina -

Operating means 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 shall include comparison of the channel with other independent channels or methods measuring the same variable.

1.10 Channel Test - A channel test is the introduction of a signal into the channel to verify that it is operable.

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1.11 Channel Calibration - A channel calibration is an adjustment of the channel such that its output responds, within  !

acceptable range and accuracy, to known values of the '

parameter which the channel measures. Calibration shall  :

encompass the entire channel, including equipment actuation, alarm, or trip.

1.12 Unscheduled Shutdown - An unscheduled shutdown is defined as any unplanned shutdown of the reactor, after startup has been initiated, caused by actuation of the reactor safety system, operator error, equipment malfunctions, or a manual shutdown in response to conditions which could adversely affect safe operation.

1.13 Reactor Shutdown - Reactor shutdown means that the shim-safety blades are fully inserted and the control rod power is off.

The reactor is considered to be operating whenever this condition is not met and there are six or more fuel elements loaded in the core.

1.14 Reactor Secured - Reactor secured is defined as follows:

a. The reactor is shutdown as defined in Definition 1.13.

b. Subcriticality of the cold xenon free core by at least one dollar has been confirmed. ,

c. No operation is in progress which involves moving fuel elements within the reactor vessel or during control rod maintenance.

1.15 True value - The true value of a process variable is its actual value at any instant.

1.16 Measured value - The measured value of a process variable is the value of the variable as it appears on the output of a measuring channel.

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1.17 Measurino Channel - A measuring channel is the combination of sensor, lines, amplifiers, and output devices which are connected for the purpose of measuring the value of a process variable.

1.18 Reportable Occurrence - A reportable occurrence is any of the following:

a. A safety system setting less conservative than the limiting setting established in the Technical Specifications.

b. Operation in violation of a limiting condition for operation established in the Technical Specifications.

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c. A safety system component malfunction or other l component or system malfunction which could, or l threatens to, render the safety system or the )

engineered safeguard systems incapable of performing j their intended safety functions. '

d. Release of fission products from a failed fuel element.

e. An uncontrolled or unplanned release of radioactive material from the restricted area of the facility.

f. An uncontrolled or unplanned release of radioactive material yielding concentrations of radioactive materials within the restricted area which could result in personnel exposures in excess of the limits specified in 10 CFR 20.

g. An uncontrolled or unanticipated change in reactivity in excess of 0.005 Ak/k.

h. 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 plant. -

1.19 Experiment - An experiment, es used herein, is defined as -

a. Any activity that introduces material into the reactor or removes material from the reactor which may effect the reactivity of the system; or

b. Changes in the structural configuration of the reactor such that the reactivity or the shielding characteristics of the system may be changed.

1.20 Experimental Facility - An experimental facility is any structure or device which is part.of the reactor structure intended to guide, orient, position, manipulate, or otherwise facilitate positioning of the experiment in the reactor.

1.21 Explosive Material -

Explosive material is any solid or liquid which is categorized as a Severe, Dangerous, or Very Dangerous Explosion Hazard in " Dangerous Properties of Industrial Materials" by N.I. Sax, Third Ed. (1968), or is

. given an Identification of Reactivity (Stability) index of 2, 3, or 4 by the National Fire Protection Association in j its publication 704-M,1966, " Identification System for Fire i Hazards of Materials," also enumerated in the " Handbook for l Laboratory Safety" 2nd Ed. (1971) published by The Chemical Rubber Co.

4 1.22 Movable Experiment - A movable experiment is one which may be inserted, removed, or manipulated while the reactor is critical.

1.23 Potential Reactivity Worth - The potential reactivity worth of an experiment is the maximum absolute value of the reactivity change that would occur as a result of intended or anticipated changes or credible malfunctions that alter experiment position or configuration.

The evaluation must consider possible trajectories of the experiment in motion relative to the reactor, its orientation along each trajectory, and circumstances which can cause internal changes such as creating or filling of void spaces or motion of mechanical components. For removable experiments, the potential reactivity worth is equal to or greater than the static reactivity worth.

1.24 Removable Experiment - A removable experiment is one which may be inserted, removed, or manipulated while the reactor is shutdown.

1.25 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 effects of hydraulic, buoyant, pneumatic, or other forces which are normal to the operating environment of the experiment, or which might arise as a result of credible malfunctions.

1.26 Unsecured Experiment - Any experiment, experimental facility or component of an experiment is deemed to be unsecured whenever it is not secured as defined in 1.25 above.

1.27 Static Reactivity Worth - As used herein, the static reactivity worth of an experiment is the absolute value of the reactivity change which is measured by calibrated control or regulating rod.

1.28 Fast Scram - A fast scram is the spring assisted gravity insertion of all shim-safety blades that begins within 100 milliseconds after introduction of a scram signal into the safety system.

1.29 Delav Scram - A delay scram is the spring assisted gravity insertion of all shim-safety blades that begins within 10 seconds after introduction of a delay scram signal into the safety system.

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5 1.30 Containment Intearity - Containment integrity exists when all of the following conditions are met:

a. One door on each personnel airlock is closed and sealed.

b. The truck door is closed and sealed,

c. Controls, equipment and interlocks for isolation of the containment building are operable.

1.31 Surveillance Frecuency - Unless otherwise stated in these specifications, periodic surveillance tests, checks, calibrations, and examinations shall be performed within the specified surveillance intervals. These intervals may be adjusted plus or minus 25%. In cases where the elapsed interval has exceeded 100% of the specified interval, the next surveillance interval shall commence at the end of the original specified interval.

1.32 Surveillance Interval -

The surveillance interval is the calendar time between surveillance tests, checks, calibrations, and examinations to be performed upon an instrument or component when it is required to be operable.

These tests may be waived when the instrument, component, or system is not required to be operable, but the instrument, component, or system shall' be tested pri'or to being declared -

operable. ,

1.33 Mode 1 Operation -

Mode 1 operation is deemed to be in offect whenever the reactor is operating at a thermal power level which is less than or equal to one megawatt and which meet the setpoints specified for Mode 1 in Table 3.1.

1.34 Mode 2 Operation - Mode 2 operation is deemed to be in effect whenever the reactor is operating at a thermal power ,

not to exceed five megawatts and which meet the setpoints l specified for Mode 2 in Table 3.1. a - -

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6 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS 2.1.1 SAFETY LIMITS IN THE FORCED CONVECTION MODE APPLICABILITY This specification applies to the interrelated variables associated with core thermal and hydraulic performance in 1.he steady state with forced convection flow. The variables are reactor thermal power, reactor coolant flow, reactor coolant inlet temperature, and the moderator level in the reactor vessel.

OBJECTIVE To maintain the integrity of the fuel element cladding and prevent the release of significant amounts of fission products.

SPECIFICATION

a. The reactor power safety limit is specified in Figure II-l corresponding to values of reactor coolant flow.

b. The reactor coolant inlet temperature safety limit is .

123 F.

c. The moderator level shall be within 12 inches of overflow.

BASIS Gross fuel element fallare and concomitant fission product release will not occur unless there is departure from nucleate boiling. The integrity of the fuel element cladding can be' ass'ured' by' control of the reactor power,'the

• reactor coolant flow rate, and reactor coolant outlet (or inlet) temperature.

The basis for establishing the safety limits on reactor power, coolant flow, and outlet temperature is a thermal hydraulic analysis calculating the values of these parameters at which flow instability and departure from nucleate boiling occurs.

This analysis (2>2'l establishes that flow instability will not occur at power levels up to 10.6 MW and departure from nucleate boiling will not occur at power levels up to 10.8 MW. These results were obtained with the coolant outlet temperature and coolant flow at their respective safety

7 limits. The analysis is not extended below 760 GPM because the orifices are not designed for extremely low flow.

REFERENCE

1. Letter, R.S. Kirkland to USAEC, June 23, 1972, Enclosure 5.

la. Letter, R.A. Karam to Director, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Jan. 21, 1993, Attachment I, " Analyses for Conversion of the Georgia Tech Research Reactor from HEU to LED Fuel.", J. M. Matos, S.C. Mo, and W. L. Woodruff, Argonne National Lab., Sept. 1992.

2.1.2 SAFETY LIMITS IN THE NATURAL CONVECTION MODE APPLICABILITY This specification applies to the interrelated variables associated with the core thermal and hydraulic performance in the natural convection mode of operation. ,

SPECIFICATION .

The reactor thermal power safety limit in the natural .

convection mode is two (2) kW.

BASIS Experience with the GTRR has shown that no damage to the core and no boiling occurs without forced convection coolant flow at power levels up to two kW.

2.2 LIMITING SAFETY SYSTEM SETTINGS

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2.2.1 LIMITING SAFETY SYSTEM SETTINGS IN THE FORCED CONVECTION MODE APPLICABILITY l

Applies to the settings of those instruments monitoring the safety limits.

OBJECTIVE To assure automatic protective action is initiated before a safety limit is exceeded.

8 SPECIFICATION The limiting safety system settings are graphically illustrated in Figure II-1 for flow and thermal power under  ;

specific inlet temperatures. The GTRR trip settings for power levels greater than one MW (Mode 2) are consertively set as follows:

Thermal Power 5.5 MW Reactor Coolant Flow 1625 GPM Reactor Outlet Temperature 139' F The GTRR trip settings for power levels less than or equal to one MW shall be as follows:

Thermal Power 1.25 MW Reactor Coolant Flow 1000 GPM Reactor Outlet Temperature 125' F 4 BASIS The limiting trip settings are chosen so that the reactor is operated with no incipient boiling. Analyses incorporating all the engineering uncertainty factors were made at 1800 gallons per minute total coolant flow, five MW thermal power ,

and an inlet reactor coolant temperature of 114*F. The results showed that a maximum fuel surface temperature 11'F .

less than the local DO 2 saturation temperature, would be obtained O'2*l .

Operation during the period 1964 to 1973 has demonstrated that a 1000 GPM flow trip setting provides for safe operation of the reactor at power levels less than or equal -

The 1. 25 MW limiting power trip setting was to one MW.

chosen to ensure that no incipient boiling occurs with the  ;

reduced coolant flow. l 0 - . .

REFERENCE l Letter, R.S. Kirkland to USAEC, October 22, 1971, 1.

Response No. 10. i la. Letter, R. A. Karam to Director,_ Of fice of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Jan. 21, 1993, Attachment I, " Analyses for Conversion of the Georgia Tech Research Reactor from HEU to LEU Fuel . ", J.M. Matos , S .C. Mo, and W.L. Woodruf f , ' Argonne National Lab., Sept. 1992, Attachment 1. l l

9 2.2.2 LIMITING SAFETY SYSTEM SETTINGS IN NATURAL CONVECTION MODE APPLICABILITY l

Applies to the limits of safety system settings when i operating in the natural convection mode.

OBJECTIVE i To assure the reactor is not operated in the natural convection mode at a power level sufficient to cause fuel damage.

SPECIFICATION The reactor thermal power limiting safety system setting shall not exceed 1.1 kW when operating in the natural convection mode.

BASIS In the natural convection mode of reactor operation the main coolant pumps are not operating. The reactor isolation valves may be closed so that only internal, natural convection is available for cooling. Experience with the GTRR has shown that the reactor can be operated at one kW indefinitely without exceeding a bulk reactor temperature of .

123*F.

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i Fig.II-l GTRR Safety Limit for Forced Convection 16 .

BASES: Moderator Within 12 Inches of Overflow Tin - 123*F Max When the Fow is Minimized 14 & Power is Maximized: Applicable for Mode 2 Only

- 12 GTRR HEU 3:

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Depanure from ,...................

10 Nucleate Boiling d Une:GTRR HEU [,, \ Fl ow instability

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[^,*' SAFE OPERATING REGION E f* ',

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- NominalOperating Conditions, Tin .114'F 0

500 1000 1500 2000 2500 Reactor Coolant Flow (GPM)

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11 3.0 LIMITING CONDITIONS FOR OPERATION 3.1 REACTIVITY LIMITS APPLICABILITY This specification applies to the reactivity condition of the reactor and the reactivity worths of control blades and experiments.

OBJECTIVE To assure that the reactor can be shut down at all times and that the safety limits will not be exceeded.

SPECIFICATION

a. The shutdown reactivity margin relative to the cold xenon free critical condition shall be at least 0.01 Ak/k with the most reactive shim-safety blade and the regulating rod fully withdrawn.

b. The reactor shall be subcritical by more than 0.0275 Ak/k during fuel or experiment loading changes.

c. No shim-safety blade shall be removed from the reactor if the shutdown reactivity margin is less than 0.01 -

Ak/k with the most reactive shim-safety blade and the regulating rod fully withdrawn.

d. Prior to criticality each shim-safety blade which is withdrawn above full insertion shall be positioned so that a free fall of the blade towards its full inserted position will result in a reactor scram activated by a negative period scram.

e. The exces's reactivity oi the core shall be limited to 11.9% Ak/k.

BASIS The shutdown reactivity margin required by Specification 3.1.a assures that the roactor can be shut down from any operating condition and will remain shutdown after cool down and xenon decay even if the control blade of the highest reactivity worth and the non-scrammable regulating rod-should be in the fully withdrawn position.

Specifications 3.1.b, 3.1.c, and 3.1.e provides assurance that the core will remain suberitical during fuel or experiment loading changes and shim-safety blade maintenance or inspection. l l

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12 The restriction on shim blade position for criticality purposes assures that the negative period generated by the insertion of the blade will cause the three remaining shim safety blades to scram. The reason for this restriction is that in the event of failure, the shtm blade will pass beyond its insertion limit to a position which results in a positive reactivity addition.

3.2 REACTOR SAFETY SYSTEM APPLICABILITY These specifications apply to the reactor safety system and other safety related instrumentation. .

OBJECTIVE 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 made critical unless:

a. The reactor safety systems and safety related instrumentation are operable in accordance with Tables -

3.1 and 3.2 including the minimum number of channels and the indicated maximum or minimum set points.

b. All shim-safety blades are operable.

c. The most recent drop time of each shim-safety blade is less than 0.5 second from fully withdrawn to 90% worth inserted.

d. The delay time from the introduction of a fast scram signal into the safety system to the release of the shim-safety blades is less than 100 milliseconds.

e. At power levels greater than 50 kW the shim-safety blades shall be banked within 5 of each other.

BASIS The rod withdrawal interlock on the Log Count Rate Channel

, assures that the operator has a measuring channel operating and indicating neutron flux levels during the approach to ,

criticality.

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TABLE 3.1 i

REOUIRED SAFETY CHANNELS '

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Mode 1 Mode 2 Minimum No.

. Channel Setooint Setpoint Recuired Function I

Start up (cps) 2 2 1(*3 Minimum count rate permissive rod -

withdrawal interlock Period trip (sec +) _ . 310 110 2 (*) Scram. 7 Power trip (MW) l'.25 5.5 2(*3 Scram Low D 2 0 flow (gpm). 1000 33 1625 2 *H"3 Scram High D 2 O Temperature ( F) 125 139 2( 3 Scram

- Low D 2 O Level (inches below $12 512 2 t*3 Isolate reactor vessel overflow) Scram Initiate ECCS '

No D 2 0 Overflow - - . 1 Scram Manual scram ,

- - 1 Scram Reflector drain - - - 1 Backup scram Containment doors open - -

1/ airlock- Scram

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i Reactor isolation-valves closed - -

2(*3/ valve Scram

("3 Required during startup and for operation with less than 1 decade overlap between the startup channel and the picoammeter. channel.

• 3 Not required for natural: convection operation

(*3One of the twelve required safety channels may be bypassed for a period not to exceed 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for test, repair, or calibration l 13 i-m m--mmu__m.m_m -_- n wn - ^ne*- ww- em eus'weea=se- & =v swe-ew=--v = w e * -w*eew =' 7 --rwr-w < -we'ew->g-e'* --w-'- w- r- v 'i s e c tied +------=-1 ..e ~3e-F- 5-w.-- . pee e- -w-mee--* +--v,- wwe ,--- w* -e,---m- , w -m --+ e- s-- -

14 The neutron flux level scrams provide redundant automatic protective action to prevent exceeding the safety limit on reactor power. The period scram conservatively limits the rate of rise of the reactor power to periods which are manually controllable without reaching excessive power levels or fuel temperatures.

The primary coolant flow rate scram setpoints provides redundant channels to assure that an automatic shutdown of the reactor will occur if sufficient flow is not maintained.

Two D2 0 low level indicators actuate a scram at 66-inches of D2 0 in the reactor tank and also provide a signal to close the reactor tank isolation valve and to initiate the Emergency Core Cooling System.

In addition to the automatic protective systems, reactor shutckwn by operator action can be initiated by pushing the manual scram button and/or dumping the reflector. The reflector provides a shutdown capability of -2.75% Ak/k.

The containment doors are interlocked to scram the reactor unless one of two of the personnel airlock doors, one of two of the emergency airlock doors and the truck door are closed.

When the reactor isolation valves are closed, two -

independent switches on the valves are interlocked to scram the reactor.

The picoammeter channel provides an accurate indication of power level over a wide range of reactor operation.

Area radiation monitors through their actuation, alert personnel to abnormal radiation levels that might exist in the facility due to improper , sample handling, equipment or shielding movements.

The gas monitor, filter assembly monitor, Kanne chamber, and the particulate monitor provide redundant channels which measure particulate and gaseous releases from the reactor building. These systems provide readout in the control room and initiate a containment isolation in the event an alarm set point is exceeded. In addition, the D 2 0 leak detector system senses small leaks in the primary coolant system and provides building isolation to prevent the release of tritium.

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15 TABLE 3.2 SAFETY RELATED INSTRUMENTATION REOUIRED FOR OPERATION Minimum No. Reauired Instrumentation Setooint Mode 1 Mode 2 Function Picoammeter channel - 1 1 Linear power level measurement

' and input for the automatic control mode i Area radiation monitors <10 mr/hr 5("3 5(***3 Alarm and prevents startup Gas monitor 1(b3 1(b3 Initiates containment isolation Filter assembly monitor 1(b3 1(b3 Initiates containment isolation Kanne chamber -

1(b3 1(b3 Initiates containment isolation D2 0 Leak detection system 1(b3 1(b3 Initiates containment isolation Particulate monitor lib 3 lib) Initiates containment isolation-  !

Emergency Core' Cooling System <280. gal - 2(b3 Alarm and prevents startup tank level No D 2 O Overflow No overflow 1 1 Alarm and prevents startup

(*3 Area monitors shall be located on the experimental level (main floor), the reactor top, in the. reactor basement, and in an area that will allow changes in reactor. coolant radioactivity to be detected. _

(b3 Either channel may be' bypassed for a. period not to exceed 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for test, repair or calibration.

("3 Or 2x background for a particular power level.

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16 The Emergency Core Cooling tank level alarm provides redundant channels which alarm if there is an insufficient D 2 0 available in the emergency cooling system tank.

The requirement that all shim-safety blades are to be banked within 5* of each other at power levels greater than 50 kW assures that peaking factors less conservative than those assumed in Safety Analyses Report do not occur due to uneven shim blade placement.

Specifications 3.2.b through 3.2.d assure that the safety system response will be consistent with the assumptions used in evaluating the reactor's capability to withstand the design basis accident in the Safety Analysis Report.

3.3 CONTAINMENT BUILDING APPLICABILITY This specification applies to the GTRR containment building requirements.

OBJECTIVE To minimize the release of airborne radioactive materials from the GTRR.

SPECIFICATION Containment integrity shall be maintained when any of the following conditions exist:

a. The reactor is operating.

b. Maintenance or operational activities which could change core reactivity are in progress.

c. Movement of" irradiated fuel is in progress, except movement of irradiated fuel contained in the fuel transfer cask to or from the containment building. 1 1

d. The reactor has been shutdown from a power level greater than 1 MW for less than eight hours.

BASIS  ;

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Building containment is a major engineered safety feature which serves as the final physical barrier to contain

- radioactive particles and gases following an accident.

Containment integrity is therefore required during all operations which could result in significant radioactive releases.

17 Any maintenance or operational activities that change the core reactivity could reduce the shutdown margin of the reactor. Therefore, containment integrity is required during such activities.

The movement of irradiated fuel containing significant fission product inventories poses a potential hazard should the fuel clad fail. Therefore, containment integrity is required except when the transfer cask is being moved to or from the containment building.

Fuel melting and the subsequent release of fission products could result from a loss-of-coolant accident following reactor shutdown if sufficient decay heat is present.

Containment integrity is therefore required until such time l as the decay heat generation rate is less than that required to melt the fuel plates. A limit of 450 C was set as the upper value for a fuel element plate temperature to preclude melting of the plates. The decay ti2ne needed to assure that this temperature would not be reached was calculated based on the analysis described below.

The calculation was based on a method developed and tested at MIT.m The analysis for the GTRR considered heat retention of a fuel element with a standard operating element and with only natural convective cooling. Two situations were evaluated: 1) heat retention of a fuel ,

element which remains in the reactor vessel with its large convective volume, but which is also subject to gamma ray heating from adjacent fuel element; and 2) heat retention of a fuel element maintained in a transfer cask which has a restricted heat transfer volume, but at the same time, does not receive additional gamma ray heat from other elements.

The standard operating history is shown below and relies upon experimentally proven values for the product of convection heat transfer coefficient, h, and fuel element

' heat transfer area, A. Assuming a flux peaking factor.of 1.5 during operation, the fuel element was subsequently either removed after some cooling period, t, 3 or else was l subjected to a loss of coolant in the reactor vessel.

a 18 Standard Operating History P = Plate Power -

Removal from D0 2 4.33 2.66 4.33 2.66 4.33 0

4pm time

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< ta ->

b The results for loss of coolant from the reactor vessel after eight hours showed a maximum plate temperature of 426 C when the element was retained.in the reactor. The plate maximum temperature occurred 45-minutes after loss of coolant.

For the more confined heat transfer situation, 1. e.,

removal of the element to a fuel cask, to restrict the heat -

transfer volume and the gamma ray heating, the maximum ,

temperature after a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> cooldown was determined to be 361*C. The temperature maximum occurred approximately one hour after removal. Consequently, a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> cooldown was- 1 proposed for limiting transfers of irradiated fuel elements -

from the GTRR reactor.

It can be concluded from the above that a fuel element in-the GTRR reactor will not melt if it has been cooled for eight hours following operation at reactor power levels greater than one MW.* If.-the reactor has not operated at- t power levels in excess of one MW,-the eight hou'r cooling ,

time will not be required. This is because the - fuel elements will not reach temperatures high enough to melt the cladding should a loss-of-coolant accident occur.(2)

Therefore containment integrity is not necessary.

L REFERENCES >

U3 MITR Operations Memo No. 98 dtd. December 25, 1965.

1 (23 Final Safeguards Report for the GTRR, February 1963, Section 8.3, pp. 106-111.

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3.4 LIMITATIONS OF EXPERIMENTS l APPLICABILITY j These specifications apply to experiments performed at the  ;

GTRR. For the Fast Shutdown System (FSS) Experiments only, j the reactivity worth limitations in Specifications 3.4.a.,

3.4.b., 3.4.c., and 3.4.d., are governed by Specification ,

3.9.

OBJECTIVE I

To prevent damage to the reactor and to limit radiation dose to .M ility personnel and the public in the event of '

expemnent f ailure.

SPECIFICATIONS j i

a. The potential reactivity worth of each secured '!

removable experiment shall be limited to 0.015 Ak/k, except a5 indicated for FSS experiments in '

Specification 3.9.

b. The magnitude of the potential reactivity of each unsecured experiment shall be limited to 0.004'Ak/k, .  ;

except as indicated for FSS experiments in Specification 3.9. '. 1 9

c. The rate of change of reactivity of any unsecured experiment, any movable experiment, or any combination >

of such experiments having a total reactivity worth in excess of 0.0025 Ak/k introduced by intentionally setting the experiment (s) in motion relative to the  !

reactor shall not exceed 0.0025 Ak/k-sec, except ' as i indicated for FSS experiments in Specification 3.9. l

d. The sum of the magnitudes of the static reactivity i worths of all unsecured experiments which coexist shall <

not exceed 0.015 Ak/k, except as indicated for. FSS experiments in Specification 3.9.

i

e. The surf ace temperature of the material which bounds or supports any experiment shall not exceed the lowest of l the following,~where applicables l (1) The saturation temperature of liquid reactor coolant at any point of mutual contact.

(2) A temperature conservatively below that at'which the corrosion rato of the boundary material at any surface would lead to its failure, or, i

i 20 (3) A temperature conservatively below that at which the strength of the boundary material would be reduced to a point predictably leading to failure,

f. Construction, fabrication and assembly materials utilized in experiments shall be so specified and used that assurance is provided that no stress failure can occur at stresses twice those anticipated in the manipulation and conduct of the experiment or twice those which could occur es a result of unintended but credible changes of, or within the experiment.

g. The radioactive material content, including fission products, of any singly encapsulated experiment shall be limited so that the complete release of all gaseous, particulate, or volatile components from the encapsulation will not result in doses in excess of 25%

of the equivalent annual doses stated in 10 CFR Part

20. This dose limit applies to persons occupying (1) unrestricted areas continuously for two hours starting at time of release or (2) restricted areas during the length of time required to evacuate the restricted area.

h. Explosive materials in excess of 25 milligrams of TNT equivalent shall not be J.rradiated in the GTRR. .

i. Explosive materials in amounts up to 25 milligrams TNT equivalent may be irradiated or stored within the containment only if they are encapsulated in such a manner as to assure compliance with Specification 3.4.f. in the event of detonation of the explosive material.

j. Explosive materials in excess of 25 milligrams, but not to exceed 300 milligrams TNT equivalent, may be stored within the reactor containment building provided that the explosive is encapsulated in such a manner as to assure compliance with Specification 3.4.f.

k. Experiments which could increase reactivity by flooding, shall not remain in or adjacent to the core unless measurements are made to assure that the shut down margin required in Specification 3.1.a would be satisfied after flooding. l l

~ '

BASIS Limiting the potential reactivity worth of secured, removable experiments to 0.015 Ak/k assures that any l

21 transient arising from the instantaneous removal of such experiments will not result in cladding failure and concomitant release of radioactive material which could lead to doses in excess of the limits set forth in 10 CFR Part 20.

A positive step change caused by the removal or insertion of unsecured experiments worth less than 0.004 Ak/k would not result in a transient behavior exceeding the Safety Limits established in Section 3.4.b of these Specifications.

Manipulations of movable experiments within the limits established in Specification 3.4.c will result in asymptotic periods longer than 20 seconds. Periods of this magnitude are easily accommodated by automatic response of the reactor safety system or by operator action. Prior to the manipulation of movable experiment the reactor power level will be reduced as needed to accommodate the calculated prompt jump associated with the step insertion of the ,

potential reactivity worth of the experiment.

Conformance with Specification 3.4.d assures that common mode failures resulting in the insertion of the total reactivity worth of all unsecured experiments will not result in accident consequences more severe than those evaluated for the failure of a single secured experiment.

Specifications 3.4.e and 3.4.f provide assurance that experiments will not fail due to the pressure or temperature ef fects of operation under anticipated operating conditions.

For the purposes of this specification the reactor shall be assumed to be operating at the Limiting Safety System Settings established in section 2 of these Technical Specifications.

Specifications 3.4.g will assure that the quantities of radioactive materials contained in experiments will be limited such that their failure will not result in personnel .

receiving doses which exceed the maximum annual exposures stated in 10 CFR 20 in either restricted or unrestricted areas.

Adherence to Specification 3.4.h, 3.4.1 and 3.4.k will ,

prevent large quantities of explosives from being present within the reactor containment building and thereby preclude damage to the safety system and safety related equipment.

Small quantities of explosive material may be safely used and stored as long as the encapsulation used has been shown to withstand the detonation of twice the quantity of explosive to be used.

22 Specification 3.4.1 assures that the shutdown margin required by 3.1.a will be met in the event of a positive reactivity insertion caused by flooding of an experiment.

3.5 RADIOACTIVE EFFLUENTS APPLICABILITY This specification applies to the controlled release of radioactive liquids and gases from the reactor site.

OBJECTIVE To define the 112 nits and conditions for the release- of radioactive effluents to the environs to assure that any releases of radioactive effluents are as low as reasonably achievable and would not result in radiation exposures in excess of limits of 10 CFR Part 20.

SPECIFICATIONS

a. Liould Effluents (1) The concentration of gross-radioactivity, above background, in liquid effluents discharged from ,

the Reactor Building to the sanitary sewer shall not exceed 3 x 10-6 microcuries/ml,- excluding' '.

tritium, unless the discharge is controlled on a

, radionuclide basis in accordance with Appendix B, Table 3,-10 CFR Part 20.

(2) The quantity of tritium released to the sewer in one month divided by the average monthly volume of water released to the sewer shall not exceed 1 x 10 2 microcuries/ml.

(3) If any of the limits of Specification a(1) or (2) are exceeded, normal orderly shutdown of the liquid waste system shall be initiated and liquid -

discharge from the facility shall not be resumed until the cause of the. excessive discharge rate is identified and corrected..

(4) The annual total quantity of gross radioactivity, excluding' tritium, to be released in liquid effluents from the reactor facility shall . not

, exceed one curie to the sanitary sewerage system.

The quantities of tritium released into the sanitary sewerage shall not exceed 5 curies / year.

23 (5) Before discharging any liquid waste from any of the holdup tanks, the following shall be performed:

1. Isolate the tank to be emptied so that no liquid waste can be added during discharge.

2. Circulate the suspect waste i.ank liquid for 30 minutes prior to sampling.

3. Obtain a sample and analyze for content of radioactivity.

If the radioactivity in that sample is within the limits of specifications 3.5.a.1 and 3.5.a.2, release to sanitary sewerage system may begin. The process of discharging the liquid waste from that tank shall be stopped at a point past the discharge halfway mark, but before the tank is 75% discharged, to analyze another sample. If results of the second analysis are within specified limits, release of the rest of the liquid waste may resume. The time for completing the second discharge shall be 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from the start.

(6) Equipment installed for the control and treatment

• of liquid effluents shall be maintained and operated, as appropriate.

b. Gaseous Effluents (1) The maximum release rates of gross radioactivity in gaseous effluents shall not exceed 585 pCi per second of Ar-41 equivalent.

(2) The maximum release rate of I-131 shall not exceed 8.3 x 10-' pCi/sec.

(3) The maximum release rate of particulates with half lives longer chan eight days in gaseous '

effluents shall be limited in accordance with the effluent concentration in air as defined in Appendix B, Table 2, Column 1 and Notes thereto of 10 CFR Part 20.

(4) If the maximum release rate for any of the above gaseous effluents is exceeded, normal orderly shutdown of the gaseous waste system shall be initiated and gaseous discharge from the facility shall not be resumed until the cause of the excessive discharge is identified and corrected.

(5) During release of gaseous radioactive effluents, the following conditions shall be met:

24 (a) One of the gross radioactivity monitors, the charcoal filter cartridge and particulate monitor shall be operable.

(b) Two gross radioactivity monitors shall be set to alarm and automatically isolate the gaseous waste releases prior to exceeding the release rates in Specification 3.5.b.(1).

(6) When the containment building is not isolated, at least one exhaust effluent monitoring channel with readout in the control- room shall be operable and capable of initiating building isolation. The time from initiation of closure to isolation valve closure shall not exceed five seconds.

BASIS

a. Licuid Effluents The liquid waste handling system is described in the Safety Analysis Report. Radioactive effluents released to the sewage on the basis of gross radioactivity are assumed not to contain I-129 and radium. The maximum amount of tritium in the discharge is limited to the value given in 10 CFR 20.

The total quantity of radioactivity limit is in accord with 10 CFR 20 for the disposal to a sewage system. The '

independent samples taken prior to and during liquid ,

effluent release shall determine the radioactivity concentration in the liquid released from the tanks and the radioactivity concentration in the discharge line to the sanitary sewers. The equipment installed for control and treatment of liquid effluents shall be maintained and operated as required by 10 CFR 50.

b. Gaseous Effluents The release-rate limit for gross radioactivity takes into account onsite meteorological data developed by the licensee and diffusion assumptions appropriate to the site. The AEC Staff determined the annual average diffusion parameters (X/Q) to be 5.2 x 10-4 sec/m3 in the most critical sector, ENE at 40 meters. The method utilized by the AEC staff is described in Sections 7.4 and 7.5 of " Meteorology and Atomic Energy - 1968," equation 7.21 being used for the beta dose and equation 7.63 being used for the gamma dose (whole body). Based on these calculations using Ar-41 as the primary dose contributor, the skin dose due to the gamma plus beta was determined to be controlling rather than the whole body dose from the gamma. A maximum release rate limit of gross radioactivity in the amount of 585 pCi/sec based upon actual perimeter surveys and validated with the COMPLY (EPA NESHAPS) code will not result in annual doses to unrestricted areas in excess of the limits specified in 10 CFR 20.

25 The release rate limit for iodines and particulates with half lives longer than eight days takes annual average atmospheric dilution into account and ensures that at any point on or beyond the restricted area fence the requirements of 10 CFR 20 will be met. The limit is based on the annual average diffusion value of X/Q which is 4.1 x 10-' sec/m2 , for the 22.5 sector having the least diffusion on an annual average. The release rate for I-131 was determined on the basis of the method given in Regulatory Guide 1.42, the nearest worst real cow as permitted by this guide was not utilized but may be applied if appropriate data is collected for determining the nearest real cow.

Isolation of the exhaust effluent stack is initiated by high radiation in the of f-gas system. Such isolation is required for abnormally high gross radioactivity releases either due to abnormal reactor operation or reactor accident.

3.6 PRIMARY COOLANT SYSTEM APPLICABILITY This specification applies to the limiting conditions for the primary coolant system pH, resistivity, flow distribution, level and D2 concentration.

OBJECTIVE To assure adequate reactor core cooling and to protect the integrity of the primary coolant system.

SPECIFICATIONS The reactor shall not be critical unless:

a. The primary coolant pH is between 4.5 and 7.5.

b. The primary coolant resistivity is at a value greater than 200,000 ohm-cm except for periods of thne not to exceed seven days when the resistivity may fall to 70,000 ohm-cm.

c. All grid positions contain fuel elements, grid plugs or experimental facilities for operation in the forced convection mode.

d. The reactor vessel coolant level is within 12 inches of the overflow standpipe level.

e. The D 2 concentration in the cover gas sweep system is

, less than 2% by volume. The cover gas shall be helium or nitrogen,

f. The concentrations of radioactive materials in the secondary coolant system are less than the values listed in 10 CFR 20, Appendix B, Table 2, Column 2.

l 26 j BASIS l

Experience at GTRR and other facilities has shown that the maintenance of primary coolant system water quality in the ranges specified in Specification 3.6.a and 3.6.b will minimize the amount and severity of corrosion of the aluminum components of the primary coolant system and the fuel element cladding.

The requirement that all grid positions be occupied will prevent the degradation of calculated flow rates due to flow bypassing the active fueled region through an unoccupied grid plate position.

The limiting value for reactor vessel coolant level is somewhat arbitrary since the core is in no danger of melting so long as it is covered. However, a drop of vessel coolant level indicates a malfunction of the reactor system and possible approach to uncovering the core. Therefore, a measurable value, well above the core is chosen.

Determination of the worth of the top reflector drain requires operation of the reactor at low power levels without a restriction on reactor vessel coolant level.

Because of radiolytic disassociation of D 20, deuterium and oxygen gas will build up in the cover gas sweep system. If .

this D2 gas were to reach an explosive concentration (about 7.8% by volume at 25 C in helium),!" damage to the primary system could occur. To assure a substantial margin of safety, a limit of 2% is set.

REFERENCES (U " Flammability of Deuterium in Oxygen-Helium Mixtures, "

USAEC Report No. TID 20898, Explosives Research Center, Bureau of Mines, June 1964.

3.7 EMERGENCY COOLING SYSTEM APPLICABILITY These specifications apply to the emergency core cooling system.

OBJECTIVE To assure that the fuel elements are adequately cooled to prevent fission product release in the event of loss of primary coolant from the reactor vessel.

SPECIFICATIONS The reactor shall not be operated at power levels in excess of one MW unless:

27

a. The D 2 0 Emergency Core Cooling System is operable.

b. A source of make-up water to the Emergency Coolant Tank is available.

c. The water level in the irradiated fuel storage pool is within 12 inches or less of normal overflow.

BASIS In the event of a loss of coolant accident on this reactor, the emergency core cooling system provides a means of retaining suf ficient cooling water in the reactor vessel or, if necessary, a means of supplying additional cooling water to prevent melting of the reactor core and the associated release of fission products.(2H') Therefore, operability of this system is required for reactor operation. The Emergency Core Cooling System Tank provides supply of D O to 2

the core spray nozzles for 30 minutes. It is then necessary to supply makeup water to the tank from the city water supply to the Nuclear Research Center building. Therefore, this source of water is also required for reactor operation.

Should city water not be available, cooling water from the irradiated fuel storage pool will be used. For this reason, a minimum level in the pool is required. .

When the reactor is operated at one MW or less, the -

emergency core cooling system is not required because the fuel can dissipate the decay heat with only natural convection cooling in air.(2na>

REFERENCES (2)Saf ety Analysis Report for the 5 MW GTRR, GT-NE-7, December 1967, Section 4.4.8.3, pp. 79-81.

423 Final Safeguards Report for the GTRR, February 1963, Section 8.3, pp. 106-111.

(3) Thompson, T.J. and Beckerly, (eds.), The Technoloov of Nuclear Reactor Safety, Volume I, p. 692, The MIT Press, Cambridge, Massachusetts, 1964.

(') Response to Question D.1, Letter to USAEC, Docket 50-160, dated 6/23/72.

3.8 FUEL HANDLING AND STORAGE APPLICABILITY Applies to the handling and storage of fuel elements.

28 Q_BJECTIVE To prevent inadvertent criticality outside of the reactor vessel and to prevent overheating of irradiated fuel elements.

SPECIFICATIONS

a. All fuel elements outside of the reactor shall be stored and handled such that the calculated k tr is less than 0.85 under optimum conditions of water moderation and reflection.

b. No more than four unirradiated fuel elements shall be together in any one room outside of the reactor, shipping containers or fuel storage racks.

c. An irradiated fuel element shall not be removed from the reactor within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of a reactor shutdown from a power level in excess of one MW.

BASIS Criticality of stored or handled fuel elements outside of the reactor can be prevented if the fuel elements are maintained in a geometry that assures an adequate margin below criticality. This margin is estabiI.shed as a k.tr of -

0.85.

The irradiated fuel storage racks in the storage pool will accommodate up to 40 fuel elements stored in a linear array along the pool walls. Experiments at ORNL have demonstrated the sub-criticality of such an array of s12nilar elements.

Fresh fuel elements will be stored in the fuel storage vault. calculations have indicated that the k.tr of an array of fresh elements in a flooded condition to be less than 0.85.

Calculations have shown that four unirradiated fuel elements cannot achieve criticality.m Therefore, grouping of fresh fuel outside of the reactor, shipping containers or normal storage will be limited to this number.

An analysis was made to determine the time-fuel temperature relationship that occurs following removal of a fuel element

~

from the core into the fuel transfer cask. These results, detailed in the basis for specification of 3.3 indicate that the maximum fuel plate temperature reached following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of cooling before removal into the cask will be 361 C.

This provides a sufficient margin to assure that no fuel plate melting will occur.

l 29 REFERENCES U1 Final Safeguards Report for the GTRR, February 1963, Section 7.7, pp. 96-7.

2See correspondence relative to Change No. 10 to the Technical Specification of Operating License No. R-97, Docket 50-160.

I') Safety Analysis Report for the 5 MW GTRR, GT-NE-7, December 1967, Section 4.4.10.

3.9 FAST SHUTDOWN SYSTEM EXPERIMENTS APPLICABILITY This specification applies to the conduct of the Fast Shutdown System Experiments performed in the GTRR.

OBJECTIVE To define the limits and conditions for the performance of the Fast Shutdown System Experiments in the GTRR. These specifications are pertinent to the performance of the Fast Shutdown System Experiments only and do not supersede any .

other specifications except as noted in those specifications.

SPECIFICATION

a. The potential reactivity worth of the experiment shall be limited to 1.75% delta-k/k.

b. The rate of change of potential reactivity worth for the experiment shall not exceed 50% delta-k/k-sec.

c. The gas shall not be removed from the experimental facility unless all shim-safety blades are fully inserted.

d. The negative period scram time delay circuit shall be activated only when the Fast Shutdown System Experiments are in progress.

e. The negative period scram' time delay circuit shall be tested before each use with calibrated instruments whose calibrations are traceable to National Institute of Standards and Technology standards. The delay time shall be one second 1 5%.

f. During the conduct of these experiments, no other experiments will be conducted.

I

30 BASIS Specification 3.9.a. and 3.9.b. restrict the conduct of the experiments to reactivity quantities and insertion rates consistent with the safety analyses for these experiments.

Specification 3.9.c. assures that the reactor will remain shutdown when the gas is released.

Specification 3.9.d. restricts the use of the negative period scram time delay circuit to only when the Fast Shutdown System Experiments are in progress, for which credible scenarios are considered in the safety analyses for these experiments.

Specification 3.9.e. limits the negative period scram delay time to one second, thereby limiting the amount of positive reactivity addition due to reactor cooling following the FSS experiment. This will limit the amount of power overshoot to a small percentage of the initial power in the event of a release of the gas during the one second delay time, as described in the FSS experiments' safety analyses.

O v

ab w

i 31 l 4.0 SURVEILLANCE REOUIREMENTS l

4.1 REACTIVITY LIMITS APPLICABILITY This specification applies to the surveillance requirements for reactivity limits.

OBJECTIVE To assure that the reactivity limits of Specification 3.1 are not exceeded.

SPECIFICATIONS

a. Shim-safety blade reactivity worths and the shutdown margin shall be measured annually and whenever a core configuration is loaded for which shim-safety blade worths have not been measured. Prior to shim-safety blade calibration, the reactor shall be confirmed to be subcritical in the cold xenon free conditions with the most reactive blade and the regulating rod withdrawn and all other shim-safety blades fully inserted.

b. The reactivity worth of experiments inserted in the GTRR shall be measured during the first startup .

subsequent to the experiments insertion. The reactivity change associated with any experiment in any core configuration shall not exceed those values specified in Specification 3.4.

BASIS Specification 4.1.a will assure that shim-safety rod reactivity worths are not degraded or changed by core manipulations which cause these rods to operate in regions where their effectiveness is reduced.

The specified surveillance relating to the reactivity worth of experiments will assure that the reactor is not operated for extended periods before determining the reactivity worth of experiments. This specification will also provide assurance that experiment reactivity worths do not increase beyond the established limits due to core configuration changes.

I 4.2 REACTOR SAFETY SYSTEM SURVEILLANCE APPLICABILITY l

These specifications apply to the surveillance of the reactor safety system and other safety related instrumentation.

32 OBJECTIVE To assure that the reactor safety system is operable as required in Specification 3.2.

SPECIFICATIONS

a. The channels listed in Tables 4.1 and 4.2 shall be tested and calibrated as indicated.

b. A channel check of the power trip channels and the picoammeter channels, comparing the channel outputs to a heat balance, shall be made weekly when the reactor is operated at a power level at or above one MW and after the installation of a new core configuration.

c. The drop time and withdrawal rate of each shim-safety blade shall be measured monthly.

d. The withdrawal and insertion rate of the regulating rod shall be measured monthly.

e. The charcoal cartridge sampler on the containment building exhaust shall have a radioisotopic analysis performed biweekly. ,

f. Grab samples of the exhaust stack effluent shall be .

obtained and have a radioisotopic analysis performed monthly.

g. The delay time from the introduction of a fast scram signal into the safety system to the release of the Shim-Safety Blades shall be measured annually.

BASIS Calibration of the safety system and other safety related instrumentation means to measure the performance as guided by vendors instructions and performance specifications of the instrument in its response to accurately prescribed input signals. Past experience and maintenance- records attest to the reliability of these channels, and justifies the calibration frequency in Tables 4.1 and 4.2.

Instrument channel tests are made to obtain assurance that the critical parameter trip channelsm are operating correctly. These tests involve the observation of trip operation in both the electronic and the electro-mechanical circuits of the trip channels. Loss of control blade magnetic clutch current shall be the indicator of correct trip operation. The testing frequency as prescribed in Tables 4.1 and 4.2 has been shown to be adequate.

J I

33 i

i Table 4.1 1

SURVEILLANCE REOUIREMENTS FOR REACTOR CONTROL AND SAFETY SYSTEMS

]

Surveillance Reauirements Test Prior. Test. -Calibrate to Startuo(*) Weekly Semi-annually 1 Channel j 1, Power trip X X X t

2. Period trip X X X  !

3. Start up channel X X X. ,

4. Logic and magnet i amplifier channel X X X .;

5. Picoammeter channels X

6. Reactor D 0' level channels 2 X X

7. D2 0 temperature channels X X

8. D2 0 flow rate channels X. X ,

f

(*1 Performed if reactor has been shutdown for eight or more hours or j any listed system has been de-energized for one or more hours. '!

+

.I l

1 l

9 t

i

l 34 Table 4.2 j l

SURVEILLANCE REOUIREMENTS FOR SAFETY RELATED INSTRUMENTATION )

l Surveillance Recuirement i I

Source Known Parameter j Daily Weekly Calibration Source Calibration i Channel Check (*) Tes t(*) Monthly Annually Kanne exhaust gas x x x GM gas monitor x x x x Moving filter particulate x x NA Cooling water gamma monitor x x x NA Area radiation monitors x x x NA ,

(*) Applicable only when the reactor is operating. ,

NA = Not Applicable j i

1 l

l

l l

l 35 In addition, an operating cycle begins in accordance with written startup procedures including a check-list to establish that all instrumentation channels are operable and scrams and l important interlocks have been tested prior to startup. The eight-hour time limit in Table 4.1 permits reactor restart following a brief shutdown for maintenance or because of a spurious scram without retesting instrumentation unrelated to the scram or maintenance. Instrument drift is not significant during this period of time if the instruments remain energized.

If an instrument is de-energized for an hour or more, or is repaired, it is considered prudent to test it before it is I returned to service. Furthermore, redundant instrumentation is i provided and channel performance can readily be checked by inter-comparison.

A check of the readings of the neutron channels can readily be accomplished by comparing them to the results of a primary system heat balance. Because the primary coolant temperature rise across the core is small, these measurements will be performed with the reactor power level at or above one MW to improve the accuracy of the measurements. Based on past experience weekly thermal power calibration is justified.

The shim-safety blades are provided to control large amounts of reactivity and to insure adequate shutdown margin. The .

reactivity worth of the shim-safety blade can vary due to-changes in core configuration, changes in the number and type of experiments, or burnup of the neutron absorber. The absolute reactivity worth of all installed unsecured experiments is ,

limited to 0.015 Ak/k (see Section 3.4) and could significantly I affect control rod reactivity worth. Burnup of the cadmium neuMon absorber is a long-term effect. Consequently, annual verification of shim-safety blade reactivity worths la considered adequate.

The shim-safety blade drives are constant speed mechanical devices. Withdrawal and insertion rates should not vary except due to mechanical wear. The surveillance frequency was chosen to provide a significant margin over the expected failure or  ;

wear rates of these devices. Because of its importance, scram j times from the full out blade position are tested as often as the driven withdrawal times.

It is not expected that any radioactive iodine or particulate radioactive material will be released through the exhaust gas ,

system. Grab samples of the stack exhaust will be analyzed to verify that the controlling radio-nuclide is Ar-41.

REFERENCES mLetter, R.S. Kirkland to USAEC,0ctober 22, 1972, Question No.2.

l

36 4.3 CONTAINMENT BUILDING APPLICABILITY This specification applies to the surveillance of the containment building.

OBJECTIVE To verify containment building integrity and to determine and record the building leakage rate under test conditions.

SPECIFICATION

a. The containment building isolation initiating system shall be tested twice a year at approximately six month intervals.

b. An integrated leakage rate test of the containment building shall be performed annually at a pressure of at least 2.0 psig. Leakage from the building shall not exceed 1.0% of the building air volume in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 2.0 psig over-pressure.

c. All additions, modifications, or maintenance of the containment building or its penetrations shall be tested to verify that the building can maintain its required .

leak tightness.

BASIS Containment building isolation is initiated by a signal from either the gaseous waste monitor, Kanne chamber, the moving-air-particulate monitor or a manual push button on the reactor console. Annual operability and semi-annual trip point checks are considered standard frequency for isolation initiating systems in nuclear facilities.

The containment has been leak tested annually since 1963 and the permissible leak rate has never been exceeded. No trend has developed which would indicate a gradual deterioration of the containment building. An annual leak rate test frequency is therefore consistent with past experience.

Any additions, modifications, or maintenance to the building or its penetrations will be tested to verify that such work i has not adversely affected the leak tightness of the building.

4.4 PRIMARY COOLANT SYSTEM l

APPLICABILITY This specification applies to the surveillance of the primary coolant system.

37 OBJECTIVE To assure high quality primary water, to detect the release of fission products from fuel elements and to detect leakage from the primary coolant system to the secondary coolant  ;

system.

SPECIFICATIONS

a. The pH of the primary coolant shall be measured weekly prior to reactor startup.

b. The resistivity of the primary coolant shall be measured weekly prior to reactor startup.

c. A radio-nuclide analysis of the primary coolant system shall be performed monthly.

d. Samples of the secondary coolant system shall be analyzed for tritium on a monthly basis.

BASIS Weekly surveillance of primary water quality and radio-activity provides assurance that pH and conductivity changes that would cause significant corrosive damage in the primary coolant system would be detected, and that leaking fuel .

elements are not being used in the reactor.

Samples of the primary coolant water will be analyzed to detect the possible presence of fission products as well as miscellaneous corrosion products. The presence of tritium in the secondary coolant water would indicate a leak in the D 2 0 system. Therefore, a routine check of the H 2 O system will be performed.

4.5 EMERGENCY COOLING SYSTEM APPLICABILITY This specification applies to the surveillance of the emergency core cooling system.

OBJECTIVE To assure that the emergency core cooling system will function properly if required.

SPECIFICATION

a. The emergency core cos_ 'ing system shall be tested for operability following any maintenance or modification of the system which could affect its performance but at ,

least monthly. This test shall include closing of the l reactor isolation valves, initiation of flow on drop of l 1

38 reactor tank coolant level and verification of system flow rate.

b. Flow to each element shall be verified semiannually at approximately six month intervals and to each element when added or moved to another position within the core.

c. The light water supply to the Emergency Core Cooling Tank shall be verified initially and annually thereafter.

d. The light water pump that supplies water from the irradiated fuel storage pool shall be tested monthly.

e. The level instrumentation on the DO 2 Emergency Core Cooling Tank shall be calibrated semi-annually.

BASIS Testing of the system in the manner prescribed will assure that it is operable. The frequency chosen is consistent with the importance of this system.

In addition to establishing closure of the isolation valves and verifying system flow, it is necessary to be certain that each individual element is receiving flow. Since the .

flow distribution system is fixed and generally inaccessible and unalterable, a semi-annual surveillance requirement is .

satisfactory except when a fuel element is moved or added to the core.

Initial testing of the entire system verified light water flow through the entire system from city water supply to the Emergency Core Cooling Tank. Annual tests will be made of the light water system to verify flow capability. A monthly surveillance interval for the pump supplying light water from the irradiated fuel storage pool is considered adequate.

The level instrumentation on the Emergency Core Cooling Tank provides continuous monitoring of the tank D 20 level as well as providing the means for determining system flow during a test. It should, therefore, be calibrated on a semi-annual-basis.

REFERENCES mSafety Analysis Report for the 5 MW GTRR, GT-NE-7, December 1967, Section 7.7.1, p. 133.

l

39 5.0 SITE DESCRIPTION 5.1 SPECIFICATION

a. The reactor facility is located on the Georgia Institute of Technology campus in-the city of Atlanta, Georgia.

b. The restricted area is formed by six-foot fence on the east, south and west of the containment building and-the laboratory building on the north. The closest unrestricted area is in a southern direction from the stack, approximately 55 feet from the reactor stack -

exhaust.

c. The exclusion area is the area inside the circle formed -

by a 100 meter (328 foot) radius centered at the reactor.

d. The low population zone outer boundary is formed by a 400 meter (1312 foot) radius from the containment building,

e. The population center distance for the~ GTRR is established as a radius of 523 meters (1750 feet) from ,

the containment building.

5.2 FUEL' ELEMENTS SPECIFICATION The LEU fuel elements shall be of the MTR type consisting of-18 fuel plates of uranium silicide with an enrichment of 20%. Each fuel. plate will have a nominal loading of 12.5 grams of U-235. The HEU fuel elements shall also be of the MTR type consisting of 16 fuel plates of uranium aluminide with an enrichment of 93%. Each fuel plate will have nominal loading of 11.75 grams of U-235.

40 ,

6.0 ADMINISTRATIVE CONTROLS l

6.1 ORGANIZATION

a. The organization for the management and operation of the reactor shall be as indicated in Figure 6.1. The Director, Nuclear Research Center shall have over all responsibility for direction and operation of the reactor facility, including safeguarding the general public and facility personnel from radiation exposure and adhering to all requirements of the operating i license and Technical Specifications. -

b. The Manager, Office of Radiation Safety, shall advise the Director, Nuclear Research Center in matters pertaining to radiological safety at the GTRR. She/he 1 has access to the Vice President, Interdisciplinary Programs and/or the President of the Institute as needed,

c. The minimum qualifications with regard to education and experience backgrounds of key supervisory personnel in inn the Reactor Operations group shall be as follows: 1 (1) Reactor Supervisor The Reactor Supervisor must have a college degree .

or equivalent in specialized training and applicable experience, and at least five years experience in a responsible position in reactor operations or related fields including at least one year experience in reactor facility management or supervision. He shall hold a Senior Reactor Operator's license for the GTRR.

(2) Reactor Enaineer The Reactor Engineer must have a combined total of at least seven years of college level education and/or nuclear reactor experience with at least three-years experience in . reactor ,

operations or related fields. He shall be qualified to hold a Senior Reactor Operator's license.  ;

d. Senior Reactor Ope'2 eor's License )

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Whenever the reat tor is not secured, the minimum crew complement at the facility shall be two persons, )

including at least one senior operator licensed pursuant to 10 CFR 55.

41 -

office of the 5 President s

Office of the Nuclear Vice President Safeguards '

for Committee Interdisciplinary Programs i

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Reporting (Safety i l (& Safety policy) l '

Office of the g >

NNRC Director .

Supervisor, Admin. Reporting 1

l Manager of Manager of Coordinator of Manager, Office Reactor Gamma Radiation Experimental of Radiation Operations Operations Research Safety i

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Figure-6.1 Georgia Tech Organization for Management and Operation of GTRR.

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42

e. An operator or senior operator licensed pursuant to 10 CFR 55 shall be present at the controls unless the reactor is shutdown as defined in these specifications.

6.2 NUCLEAR SAFEGUARDS COMMI_TTEE

a. A Nuclear Safeg' lards Committee shall be established by the President. of the Institute and shall be responsible for maintaining health and safety standards associated with operation of the reactor and its associated facilities.

b. The Committee shall be composed of five or more senior technical personnel who collectively provide experience in reactor engineering, reactor operations, chemistry and radiochemistry, instrumentation and control systems, radiological safety, radiation protection, and mechanical and electrical systems. No hore than a minority of the Committee members shall be from the GTRR staff.

c. The Committee shall meet quarterly and as circumstances warrant. Written records of the proceedings, including any recommendations or occurrences, shall be distributed to all Committee members and the .

President's Office. ,

d. The quorum shall consist of not less than a majority of the Committee membership and shall include the chairman or his designated alternate. The operating staff may not constitute a majority of those present.

e. The Committee shall:

(1) Review and approve proposed changes in equipment, systems, tests, experiments, or procedures that do not involve an unreviewed safety question pursuant to 10 CFR 50.59(a).

(2) Review reportable occurrences.

(3) Review and approve proposed operating procedures and proposed changes to operating procedures.

Minor modifications to operating procedures which do not change the original intent of the operating procedure may be approved by the Director of NNRC on a temporary basis. The Committee shall consider for approval such minor modifications at its next scheduled meeting.

(4) Review and approve proposed changes to Technical l Specifications and license excluding i

organizational structure. The responsibility and authority for organizational structure resides

, with the President of the Institute.

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43 (5) Review and approve proposed experiments and tests utilizing the reactor facility which are significantly different from tests and experiments previously performed a*: the GTRR.

(6) Review and approve proposed changes to the facility made pursuant to 10 CFR 50.59(c).

(7) Review violations of Technical Specifications, license, or internal procedures or instructions having safety significance.

(8) Review operating abnormalities having safety significance.

(9) Review audit reports.

(10) Audit reactor operations and reactor operation records for compliance with internal rules, procedures, and regulations and with licensed provisions including Technical Specifications at least once per calendar year (interval between audits not to exceed 15 months).

(11) Audit the retraining and re-qualification program ,

for the operatinrj staff, at least once every other calendar yr.ar (interval between audits not .

to exceed 30 r.anthc).

(12) those Auditdeficiencies the results thatof action taken to correct may occur in the reactor facility equipment, systems, structures, or methods of operations that affect reactor safety, at least once per calendar year (interval between audits not to exceed 15 months).

6.3 ADMINISTRATIVE CONTROLS OF EXPERIMENTS

a. Evaluation by Safety Review Group (1) No experiment shall be performed without review and approval by the Nuclear Safeguards Committee.

Repetitive experiments with common safety considerations may be reviewed and approved as a class.

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(2) Criteria for review of an experiment or class of experiments shall include (a) applicable regulatory positions including those in 10 CFR Part 20 and the technical specifications and (b) in-house safety criteria and rules which have been established for facility operations, including those which govern requirements for encapsulation, venting, filtration, shielding,

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i 44 and similar experiment design considerations, as well as those which govern the quality assurance program required under 50.34.

(3) Records shall be kept of the Nuclear Safeguards Committee's review and authorization for each experiment or class of experiments.

b. Operations Approval (1) Every experiment shall have the prior explicit written approval of the Neely Nuclear Research Center management including the Reactor ,

Supervisor.

(2) The execution of any experiment shall be under the active control of licensed operators,

c. Procedures for Active Conduct of Experiments (1) Written procedures shall be provided for the use or operation of experimental facilities.

(2) While the reactor is operating, no experiment shall be moved without permission of the licensed operator at the console.

(3) Each experiment removed from the reactor or reactor system shall be subject to a radiation survey which anticipates exposure rates greater than those predicted. The results of such monitoring should be documented.

d. Procedures Relatino to Personnel Access to Experiments (1) There shall be a documented procedure for the control of visitor access to the reactor area to minimize the likelihood of unnecessary exposure to radiation as a result of ongoing activities.

(2) Experimenters and/or visitors to the Radiation Control Zone shall be under the positive control of Neely Nuclear Research Center personnel.

e. Ouality Assurance Procram There shall be a Quality Assurance Program covering the design, fabrication, and testing of experiments, .

including procedures for verification of kinds and l amounts of their material contents to assure compliance ,

with the technical specifications in Section 3.4. ,

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45 6.4 PROCEDURES j i

a. All procedures and major changes thereto shall be  !

reviewed and approved by the Nuclear Safeguards  :

Committee prior to being effective. Changes which do not alter the original intent of a procedure may be approved by the director of the facility. Such change's shall be recorded and submitted periodically to the Nuclear Safeguards Committee for routine review.

b. Written procedures shall be provided and utilized for the following:

(1) Normal startup, operation and shutdown of the reactor and of all systems and components involving nuclear safety of the system.

(2) Installation and removal of fuel elements, control blades, experiments and experimental facilities.

(3) Actions to be taken to correct specific and -

foreseen potential malfunctions of systems or components, including responses to alarms, suspected primary system leaks and abnormal reactivity changes.

(4) Emergency conditions involving potential or '

actual release of radioactivity.

(5) Preventive or corrective maintenance operations which could have an effect on the safety of the reactor.

(6) Radiation and radioactive contamination control.

4 (7) Sarveillance and testing requirements.

(8) A site emergency plan delineating the action to be taken in the event of emergency conditions and accidents which result in or could lead to the release of radioactive materials in quantities that could endanger the health and safety of employees or the public. Periodic evacuation drills for facility personnel shall be conducted to assure that facility personnel are familiar with the emergency plan.

(9) Physical security of the facility and associated special nuclear material.

6.5 OPERATING RECORDS

a. The following records and logs shall be prepared and retained at the facility for at least five years:

i 46 (1) Normal facility operation and maintenance.

(2) Reportable occurrences.

(3) Tests, checks, and measurements documenting compliance with surveillance requirements.

(4) Records of experiments performed.

b. The following records and logs shall be prepared and retained at the facility for the life of the facility (1) Gaseous and liquid waste released to the environs.

(2) Offsite environmental monitoring surveys.

(3) Radiation exposures for all GTRR personnel.

(4) Fuel inventories and transfers.

(5) Facility radiation and contamination surveys.

(6) Updated, corrected, and as-built facility drawings. .

(7) Minutes of Nuclear Safeguards Committee meetings. '.

(8) Records of radioactive shipments.

6.6 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 action shall be taken:

a. Reactor conditions shall be returned to normal or the reactor shall be shutdown. If it is necessary to shut the reactor down to correct the occurrence, operations shall not be resumed unless authorized by the director of the facility.

b. All reportable occurrences shall be promptly reported to the reactor supervisor and the director of the facility.

c. All reportable occurrences shall be reported to the Nuclear Regulatory Commission in accordance with Section 6.7 of these specifications.

d. All reportable occurrences shall be reviewed by the Nuclear Safeguards Committee.

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47 j 6.7 R.EPORTING REOUIREMENTS I

i The following information shall be submitted to the USNRC in {

addition to the reports required by Title 10, Code of  ;

Federal Regulations.

a. Annual Operatina Reports i

A report covering the previous year shall be submitted  !

to the of fice of the Regional Administrator, Region II,  !

with a copy to the Director, Office of Nuclear Reactor  !

Regulation, by Kirch 1 of each year. It shall include i the following:  ;

(1) Operations Summary j A summary of operating experience occurring  ;

during the reporting period including: 1 (a) changes in facility design; ,

l (b) performance characte.ristics (e.g. , equipment l and fuel performance); ,

(c) changes in operating procedures which relate .i to the safety of facility operations; .

(d) results of surveillance tests and -.

inspections required by these technical -

specifications; j (e) a brief summary of these changes, tests, and  ;

experiments which required authorization  !

from the Commission pursuant .to 10 CFR  ;

50.59(a), and l (f) changes in the plant operating staff ' serving  !

in the.following positions:

1. Director, Nuclear Research Center l ,

2. Reactor Supervisor l

2 Reactor Engineer

4. Manager, Office of Radiation Safety

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1 Nuclear Safeguards Committee members j

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l 48 (2) Power Generation A tabulation of the thermal output of the facility during the reporting period.

(3) Shutdowns A listing of unscheduled shutdowns which have occurred during the reporting period, tabulated according to cause, and a brief discussion of the preventive actions taken to prevent recurrence.

(4) Maintenance A discussion of corrective maintenance (excluding preventative maintenance) performed during the reporting period on safety related systems and components.

(5) Chances, Tests and Experiments A brief description and a summary of the safety evaluation for those changes, tests, and experiments which were carried out without prior Commission approval, pursuant to the requirements .

of 10 CFR Part 50.59(b). .

(6) Radioactive Effluent Releases A statement of the quantities of radioactive effluents released from the plant, with data summarized following the general format of USNRC Regulatory Guide 1.21:

(a) Gaseous Effluents

1. Gross Radioactivity Releases

a. Total gross radioactivity (in curies), primarily noble and activation gases.

b. Average concentration of gaseous >

effluents released during normal steady state operation. (Averaged over the period of reactor

, operation.)

c. Maximum instantaneous concentration of gaseous radionuclides released during special operations, tests, or l

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49 experiments, such as beam tube experiments, or pneumatic tube operation.

d. Percent of technical specification limit.

2. Iodine Releases (Required if iodine is identified in primary coolant samples, isotopic analysis required in (a)1. above or if fueled experiments are conducted at the facility.) ,

a. Total iodine radioactivity (in curies) by nuclide released, based on representative isotopic analyses performed.

h. Percent of technical specification limit.

3. Particulate Releases

a. Total gross radioactivity ( ,y) released (in curies) excluding .

background radioactivity.

h. Gross alpha radioactivity released (in curies) excluding background radioactivity. (Required if the operational or experimental program could result in the release of alpha emitters.)

p. . Total gross radioactivity (in curies) of nuclides with half-lives greater than eight days.

d. Percent of Ef fluent Concentration, )

as listed in 10 CFR 20, Appendix  !

B, Table 2, Column 1, for particulate radioactivity with half-lives greater than eight days.

(b) Licuid Effluents

1. Total gross radioactivity (p,y) released (in curies) excluding tritium and average concentration released to

50 the unrestricted area or sanitary sewer (averaged over period of release).

2. The maximum concentration of gross radioactivity (p,y) released to the unrestricted area.

3. Total alpha radioactivity (in curies)  :

released and average concentration i released to the unrestricted area )

(averaged over the period of release).

4_ . Total volume (in ml) of liquid waste l released.  ;

1 Total volume (in ml) of water used to i dilute the liquid waste during the  !

period of release prior to release from the restricted area.

1 Total radioactivity (in curies), and

• concentration (averaged over the period  ;

of release) by nuclide released, based '

on representative isotopic analyses performed for any release which exceeds 1 x 10-7 Ci/ml. ,

2 Percent of technical specification ,

limit for total radioactivity released from the site. '.

(7) Environmental Monitorino For each medium sampled, e.g., air, surface water, soil, fish, vegetation, include:

(a) Number of sampling locations and a.

description .of their location relative to the reactor.

(b) Total number of samples.

(c) Number of locations at which levels are >

found to be significantly- above local; backgrounds. t (d) Highest, lowest, and the annual average ~

concentrations or levels of radiation for the sampling point with the highest average and the location of that point Nith respect to the site.

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51 (e) The maximum cumulative radiation dose which could have been received by an individual continuously present in an unrestricted area during reactor operation from:

1. direct radiation and gaseous ef fluent,

2. liquid effluent.

If levels of radioactive materials in environmental media, as determined by an environmental monitoring program, indicate the likelihood of public intakes in excess of 1% of those that could result from continuous exposure to the concentration values listed in Appendix B, Table 2, 10 CFR i Part 20, estimates of the likely resultant exposure to individuals and to population groups and assumptions upon which estimates are based shall be provided.

(8) Occupational Personnel Radiation Exposure A summary of radiation exposures greater than 500 mrem (50 mrem for persons under 18 years of age) received during the reporting period by facility personnel (faculty, students, or experiments).

b. Non-Routine Reports .

(1) 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 and telegraph to the Office of the regional Administrator, Region II, with a copy to the Director, Office of Nuclear Reactor Regulations followed by a written report within 10 days to the Office of the regional Administrator, Region II, with a copy to the Director, Office of Nuclear Reactor Regulations in the event of the reportable occurrences as defined in section 1.0. The written report on these reportable occurrences, and to the extent possible, the preliminary telephone and telegraph notification shall:

(a) describe, analyze, and evaluate safety implications,

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(b) outline the measures taken to assure that the cause of the condition is determined,

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(c) indicate the corrective action (including l any changes made to the procedures and to the quality assurance program) taken- to prevent repetition of the occurrence and of ,

j similar occurrences involving similar '

components or systems, and (d) evaluate the safety implications of the incident in light of the cumulative .

experience obtained from the record of previous failures and malfunctions of similar systems and components.

(2) Unusual Events A written report shall be forwarded within 30 days to the Office of the Regional Administrator, Region II, with a copy to the Director, Office of Nuclear Reactor Regulations in the event of:

(a) Discovery of any substantial errors in the transient or accident analyses or in the methods used for such analyses, as described ,

in the Safety Analysis Report or in the bases for the Technical Specifications.  ;

(b) Discovery of any substantial variance from .

performance specifications contained-in the .!

Technical Specifications or in the Safety r Analysis Report. '. l (c) Discovery of any condition involving a i possible single failure which, for a system designed against assumed ' single failures, &

could result in a loss of the capability of  ;

the system to perform its safety function. [

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