ML20235N870
| ML20235N870 | |
| Person / Time | |
|---|---|
| Site: | Rensselaer Polytechnic Institute |
| Issue date: | 07/07/1987 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20235N799 | List: |
| References | |
| CX-22-A-007-ERR, CX-22-A-7-ERR, NUDOCS 8707200256 | |
| Download: ML20235N870 (21) | |
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t ATTACHMENTTOLICENSEAMENbMENTNO.7 FACILITY' OPERATING LICENSE NO. CX-2e.
1 DOCKET NO. 50-225
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Revised Appendix A. Technical Specifications is.as follows.
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APPENDIX A
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FACILITY LICENSE NO. CX-22
' l, TECHNICAL SPECIFICATIONS j
1 AND BASES FOR THE I
RENSSELAER POLYTECHNIC INSTITUTE j
REACTOR CRITICAL FACILITY l
SCHENECTADY, NEW YORK
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l DOCKET NO. 50-225 Dated:
May 1987
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-e TABLE OF CONTENTS Page 1.0 DEFINITIONS................................................
1-1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS..........
2-1 2.1 Safety Limit - Fuel Pell et Tempera ture................
2-1 2.2 Limiting Safety System Settings - Reactor Power.......
2-2 3.0 LIMITING CONDITIONS FOR OPERATION..........................
3-1 3.1 Reactor Control and Safety Systems....................
3-1 3.2 Reactor Parameters....................................
3-4 3.3 Radiation Monitoring..................................
3-5 3.4 Experiments...........................................
3-6 4.0 SURVEILLANCE REQUIREMENTS..................................
4-1 4.1 Reactor Control and Safety............................
4-1 4.2 Reactor Parameters....................................
4-2 4.3 Radiation Monitoring..................................
4-3 c5 5.0 DESIGN FEATURES............................................
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5.1 Site..................................................
5-1 5.2 Facility..............................................
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5.3 Reactor Room..........................................
5-1 5.4 Reactor...............................................
5-1 5.4.1 Reacter Tank...................................
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5.4.2 Reactor Core...................................
5-2 5.4.3 Fu e l P i n s......................................
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- l 5.4.4 Control Rod Assemblies.........................
5-2
' l 5.5 Water Handling System.................................
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5.6 Fuel Storage and Transfer.............................
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TABLE OF CONTENTS (Continued) d
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6.0 ADMINISTRATIVE CONTROLS....................................
6-1 6.1 Organization..........................................
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6.1.1 Structure......................................
6 6.1.2 Responsibility ~.................................
6-1 6.1.3 Staffing.......<..............................
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6.1.4 Selection and Training of Personnel............
.6-2 6.1.5 Review and Audit...............................
6-2 6.1.5.1 Composition.and Qualification......
6-2 6.1.5.2. Charter and Rules..................
6-3 6.1.5.3 Review and Approval Function.......
6-3 6.1.5.4 Audit Function.....................
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6.2-Procedures............................................
6-4
~i 6.3 Experiment Review and Approva1'........................
6-4 --
1 6.4 Required Actions.......................................
6-5 6.4;1 Action to Be Taken.in Case of Safety Limit Violation......................................
6-5 6.4.2 ~ Action to Be Taken in the Event of an Occurrence of the Type Identified in Section 1.0Q..........
6-6 6.5 Reports...............................................
6-6 6.5.1 Operating Reports..............................
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6.5.2 Non-Routine Reports............................
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.Z-l 6.6 Operating Records.....................................
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1.0 DEFINITIONS The terms Safety Limit (SL). Limitina Safety System Setting (LSSS), and l
Limiting tondition for Operation (LCO), and Surveillance Requirements are as defined in 50.36 of 10 CFR Part 50.
A.
Channel Calibratiori.- The correlation of channel outputs to known' input signals and other known parameters.
Calibration shall encom-pass the entire channel, including equipment actuation, alarm, or
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trip.
B.
Channel Check - Qualitative determination of acceptabie operability by observation of instrument behavior during operation.
This deter-
-j mination shall include, where possible, comparison of the instrument with other independent instruments measuring the same variable.
C.
Channel Test - The injection of a simulated signal into the instru-ment primary sensor to verify the proper instrument response alarm and/or initiating action.
D.
Control Rod Assembly - A control mechanism consisting of a stainless steel basket that houses two absorber sections, one above the other.
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These absorber sections may contain either enriched boron in iron, Eu0@er-cadmium-indium.in a stainless steel cermet, stainless steel, or an alloy o sil All absorber sections except the one con-taining silver-cadmium-indium are clad in stainless steel. All are of the same dimensions, nominally 2.6 inches square, with their poisons uniformly distributed.
The absorbers, when fully inserted, i
shall extend above the top and to within one inch of the bottom of 3
the active core.
E.
Excess Reactivity - The available reactivity above a cold, clean, critical ~ configuration which may be added by manipulation of controls.
F.
Experiment - (1) An apparatus, device, or material placed in the re-actor vessel, and/or (2) any operatior. designed to measure reactor characteristics.
G.
Measuring Channel - The combination of sensor, lines, amplifiers, and output devices which are connected for the purpose of measuring 1
the value of a process variable.
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H.
Measured Value - The vtlue of the process variable as it appears on the output of a measuring chanr.e1.
I.
Movable Experiment - A movable experiment is one in which material may be inserted, removed, or manipulated while the reactor is critical.
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Operable - A system or component is capable of performing its in-
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tended. function in its required. manner.
K.
Operating - A system or component is performing its intended function
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L.
Reactor Safety System - Combination of safety channels and asso-l ciated circuitry which forms the automatic protective system for the I
reactor.or provides -information which requires manual protective l
action to be initiated.
i M.
. Reactor Scram - A gravity drop of the control rods accompanied by the opening of the moderator dump valve.
The scram can be initiated either manually or automatica1Jy by. the safety system.
N.
Reactor Secured - (1) The full insertion of all. control rods has been verified, (2) the console key is removed, and (3) no operation is in progress which involves moving fuel pins. in the reactor vessel, the l
i insertion or removal of experiments from the reactor vessel, or con-
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trol rod maintenance.
0.
Reacter Shutdown - The control rods are fully inserted and the re-actor is shutdown by at least 1.00$.
The reactor is considered'to be operating whenever this condition is not met and more than 60% of the total number of fuel pins re configuration (Core A or Core B) quired for criticality in a given have been loaded in the core.
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Readily Available on Call - The Licensed Senior Operator (LS0) on 2
duty shall remain within a 15 mile radius or 30 minutes travel time X
of the facility, whichever is closer, ard the operator-on-duty shall know the exact location and telephone number of the LSO on duty.
Q.
Reportable Occurrence - The occurrence of any facility condition that:
1.
Causes a Limiting Safety System Setting to exceed the setting established in Section 2 of the Technical Specifications; l
2.
Exceeds a Limiting Condition for Operations as established in Section 3 of the Technical Specifications; i
3.
Causes any uncontrolled or unplanned release of radioactive-material from the restricted area of the facility; 4.
Results in safety system component failures which could, or l
threaten to, render the system incapable of performing its l
intended safety function as defined in the Technical Specifi-cations or SAR; 5.
Results in abnormal degradation of one of the several boundaries which are designed to contain the radioactive materials result-ing from the fission processes; 4
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Results in uncontrolled or unanticipated changes in reactivity l
of greater than.60$;
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Causes conditions arising from natural or offsite manmade events l
that affect or threaten to affect safe operation of the facility, or; I
8.
Results in observed inadequacies 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.
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R.
Review and Approve - The reviewing group or person shall carry out a
- ]4 review of the matter in question and may either approve or disap-prove it; before'it can be implemented, the matter in question must receive approval from the reviewing group or person.
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S.
Safety Channel - A measuring channel in the reactor safety system.
T.
Secured Experiment - Any experiment, experimental facility, or com-ponent 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. The restraining forces must be equal to or greater than those that hold the fuel pins themselves in the reactor core.
l U.
Source - A neutron-emitting radioactive material, other than reactor Tuel, which is positioned in or near the assembly to provide an ex-ternal source of neutrons.
V.
Surveillance Frequency - Unless otherwise stated in these specifica-T tions, periodic surveillance tests, checks, calibrations, and C,!
examinations shall be performed within the specified surveillance l
intervals.
In cases where the elapsed interval has exceeded 100% of the specified interval, the next surveillance interval shall com-mence at the end of the original specified interval. Allowable sur-1 veillance intervals, as defined in ANSI /ANS 15.1 (1982) shall not ex-ceed the following:
1.
Five-year (interval not to exceed six years).
2.
Two-year (interval not to exceed two and one-half years).
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Annual (interval not to exceed 15 months).
4.
Semiannual (interval not to exceed seven and one-half months).
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Quarterly (interval not to exceed four months).
6.
Monthly (interval not to exceed six weeks).
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Weekly (interval not to exceed ten days).
8.
Daily (must be done during the calendar day).
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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limits - Fuel Pellet Temperature i
Applicability 1
Applies to the maximum temperature reached in any incore fuel pellet 1
as a result of either normal operation or transient effects.
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Objective To identify the maximum temperature beyond which material degradation of the fuel and/or its cladding is expected.
Specification i
Fuel pellet temperature at any point in the core, resulting from normal operation or transient effects, shall be limited to no more than 2000*C.
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Bases l
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Specific determination of the melting point of the SPERT fuel has not j
been reported. A safety limit of 2000*C is below the listed melting l
point of U0 under a wide variety of conditions.
The chosen value is p
conservative in view of variations that might result due to the presence of small quantities o' impurities and the comparatively high vapor pressure of U0 at elevated temperatures. The safety limit
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specified is about 7v0*C below the measured melting point of UD, in a helium atmosphere.*
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Reference:
W. A. Duckwcrth, ed., " Physical Properties of Uranium Dioxide,"
i Uranium Dioxide:
Properties and Nuclear Applications (Washington, D.C.:
Naval. Reactors, Division of Reactor Development), 1961, pp. 173-228.
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2.2 Limiting Safety System Settings - Reactor Power
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Aoolicability Applies to the settings to initiate protective action for instruments monitoring parameters associated with the reactor power limits.
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To assure protective action before safety limits are exceeded.
Specification The limiting safety system settings on reactor power shall be as follows:
ll Maximum Power Level 135 watts Minimum Flux Level 2.0 counts /see Minimum Period 5 seconds Bases The maximum power level trip setting of 135 watts on Log Power and Period Channel 2(PP2) correlates with a reading of not greater than 90% on the highest scale of either of the two Linear Power Channels (LP1, LP2) as established by activation techniques. These scram setpoints ensure reactor shutdown and prevent significant energy 1
deposition or enthalpy rise in the core in the event of any credible accident scenario.
_i The minimum flux level has been established at 2 cps to prevent a source-out startup and provide a positive indication of proper instrument functicn before any reactor startup.
The minimum 5-second period is specified so that the automatic safety system channels have sufficient time to respond in the event of a very rapid positive reactivity insertien.
Power increase and energy depo-sition subsequent to scram initiation are thereby, limited to well below the identified safety limit.
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l 3-1 3.0 LIMITING CONDITIONS FOR OPERATION
-' j 3.1 Reactor Control and Safety Systems l
Applicability Applies to all methods of changing core reactivity available to the reactor operator.
l Objective To assure that available shutdcwn method is adequate and that positive reactivity insertion rates are within those analyzed in the Hazards Summary Report (hereinafter safety analysis report).
Specifications l
1.
The excess reactivity of the reactor core above cold, clean I
critical shall not be greater than 0.60$. The maximum reactiv-1 ity worth of any clean fuel pin shall be 0.20$.
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2.
There shall be a minimum of four operable control rods.
The re-I actor shall be suberitical by more than 0.70$ with the most reactive control rod fully withdrawn.
3.
The maximum control rod reactivity rate shall be less than 0.125/sec up to 10 times source level and 0.05$/sec at all higher'
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levels.
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The total centrol rod drop time for each control rod from its C
fully withdrawn position to its fully inserted position shall
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be less than or equal to 900 milliseconds.
This time shall include a maximum magnet release time of 50 milliseconds.
5.
The auxiliary reactor scram (moderator-reflector water dump) shall add negative reactivity within one minute of its activa-tion.
6.
The normal moderator-reflector water level shall be established not greater than 10 inches above the top grid of the core.
7.
The minimum safety channels that shall be operating during the l
reactor operation are listed in Table 1.
8.
Af ter a scram, the moder ator dump valve may be reclosed by a senior reactor operator if the cause of the scram is known, all centrol rods are verified to have scrammed and it is deemed wise to retain the moderator shielding in the reactor tank.
9.
The interlocks that shall be operable during reactor operations are listed in Table 2.
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- 10. The thermal power level shall be controlled so as not to exceed 100 watts, and the integrated thermal power for any consecutive 365 days shall not exceed 200 kilowatt-hours.
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TABLE 1 Minimea Safety System Channels Minimum Reactor Conditions - Ranges Channels Number Functions Log Count Rate (a) 1 Minimum Flux 4
Start - 2 cps - 10 cps Up Level Power - 10~4 - 150%
Linear Power 2
High Neutron l
Level Scram
- 10'3 - 300%
Log-N; Period (b) 1 High Neutron
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Level and Period Scram
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Reactor Scram Building Power 1
Loss of Power Reactor Door 1
Reactor Scram (d)
Scram (a)Maybebypassedwhenlinearpow;rchannelsarereadinggreaterthan
_3 3 x 10 ' amps.
(b)During steady-state operation, this safety channel may be bypassed with the permission of the Operations Supervisor.
(c)The nenual scram shall consist of a regular manual scram at the console and a manual electric switch which shall aisconnect the electrical power of the facility from the reactor, causing a loss of power scram.
(d)The reactor door scram may be bypassed during maintenance checks and radiation surveys with the specific permission of the Operations Supervisor provided that no other scram channels are bypassed.
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3-4 Basy The minimun number of four control rods is specified to ensere that there is adequate shutdown capability even for the stuck control rod condition.
The insertion time of less than 900 milliseconds for each control rod from its fully withdrawn position is sp3cified to assure that the insertion time does not exceed that assumed when establishing the minimum period of Specific 6tien 2.2 as a limiting safety system setting.
The auxiliary reactor scram is specified to assure that there is a secondary mode of shutdown available during reactor operations. The requirement that negative reactivity be introduced in less than one minute following activation of the scram is established to minimize the consequences of any potential power transients.
The normal moderate'r-reflector water level of the reactor is established at not greater than 10 inches above the top grid of the core to assure l
that the moderator-reflector water dump, back-up scram will introduce nega-tive reactivity within the time assumed in the safety analysis by loss of 1
reflector at the top of the core.
The safety system channels listed in Table 1 p'tovide a high degree of re-dundancy to assure that human or mechanical failures will not endanger the reactor facility or the general public.
The interlock system listed in Table 2 assures that only authorized person-nel can operate the reactor and the proper sequence of operations is per-15; formed.
It also limits the actions that an operator can take, and assists 11 him in safely operating the reactor.
Limitations imposed on core reactivity, control rod worth, and reactor power preclude conditions that could allow the development of a potentially damaging accident.
The limitations are conservative in view of core energy deposition, yet permit adequate flexibility in the research and instruction l
1 for which the facility is intended.
3.2 Reactor parameters i
Applicability 1
These specifications apply to core parameters and reactivity l
coefficients.
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-l The purpose of these specifications is to assure that the reactor is operated within the range of parameters that have been analyzed.
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3-5 Specifications 1.
Above 100*F the isothermal temperature coefficient of reactivity shall l
be negativa.
The net positive reactivity insertion from the minimum operating temperature to the temperature at which the. coefficient becomes negative shall be less than 0.15$.
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The void coefficient of reactivity shall be negative, when the modera-tor temperature is above 100*F, within all standard fuel assemblies l
and have a minimum average negative value of 0.00043$/cc within the boundaries of the active fuel region.
3.-
The minimum operating temperature shall be 50*F.
Bases The minimum absolute value of the temperature coefficient of reactivity is specified to ensure that negative reactivity is inserted when reactor temperature increases above 100 F.
It is of note that even in the worst postulated accident scenarios, such as considered in Section 4 of the SAR (1964), reactivity insertion because of temperature change would be g
negligible. The minimum average negativd'value of the void coefficient is specified to ensure that the negative reactivity inserted because of. void formation is greater than that which was calculated in the SAR.
The minimum operating temperature of 50*F establishes the temperature range for which the net positive reactivity limit can be applied.
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3.3 Radiation Monitoring Applicability These specifications apply to the minimum radiation monitoring requirements for reactor operations.
Objective The purpose of these specifications is to assure that adequate monitoring is available to preclude undetected radiation hazards or uncontrolled release of radioactive material.
Specifications 1.
The minimum complement of radiation monitoring equipment required to be opersting for reactor operation shall include:
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i 3-8 Bases The basic experiments to be performed in the reactor programs are described in the Safety Analysis Report (SAR).
The present programs are oriented toward reactor operator training, the instruction of students, and with such research and development as is permitted under the terms of the facil-ity license. To assure that all experiments are well planned and evaluated prior to being performed, detailed written procedures for all new experi-ments must be reviewed by the NSRB and approved by the Operations Supervisor.
Since the control rods enter the core by gravity and are required by other technical specifications to be operable, no equipment should be allowed to interfere with their functions. To assure that specified power limits are not exceeded, the nuclear instrumentation must be capable of accurately monitoring core parameters.
All new reactor experiments are reviewed and apprcved prior to their per-formance to assure that the experimental techniques and procedures are safe and proper and that the hazards from possible accidents are minimal.
A maximum reactivity change is established for the remote positioning of experimental samples and devices during reactor operations to assure that i
i the reactor controls are readily capable of controlling the reactor.
All experimental apparatus placed in the reactor must be properly secured.
In consideration of potential accidents, the reactivity effect of movable apparatus must be limited to the maximum accidental step reactivity inser-tion analyzed. This corresponds to a 0.60$ positive step while operating 1
at full power followed by one failure in the reactor safety system.
5.
Restrictions on irradiations of explosives and highly flammable materials are imposed to minimize the possibility of explosion of fires in the vicinity of the reactor.
l To minimize the possibility of exposing facility personnel or the public to radioactive materials, no experiment will be performed with materials that could result in a violent chemical reaction, produce airborne activity, or cause a corrosive attack on the fuel cladding or primary i
I coolant system.
Specifications 8 and 9 will ensure that the quantities of radioactive materials contained in experiments will be so limited that their failure will not result in exposures to individuals in restricted or unrestricted areas to exceed the maximum allowable exposures stated in 10 CFR 20. The restricted area maximum is defined in 10 CFR 20.101 and 10 CFR 20.103. The unrestricted arca maximum is defined in l
10 CFR 20.105 and 10 CFR 20.106.
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5-2 5.4.2 Reactor Core
'The reactor core shall consist of uranium fuel provided in the form of 4.8 weight percent enriched UO pellets in stainless steel-cladding, arranged in roughly a c,ylindrical fashion with
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four control rods placed symmetrically about the core periph-ery.
Fuel pins, with an effective length of 91.44 cm are set j
on a square pitcF of 1.49 cm to yield an effective core radius of approximately 35 cm. Two fuel pin arrangements are provided for. The first, referred to as " Core A," is the
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solid array of pins shown in Figure 4.3 of the SAR. The second, referred to as "Cere B," is the annular array of pins shown in Figure 4.4 of the SAR.
The pins themselves are '
supported and positioned on a fuel pin support plate, drilled e,
with 1/4-inch-diameter holes to accept tips on the end of each I
pin. The support plate rests on a carrier plate which forms the base of a three-tiered overall core support structure. An i
l upper fuel lattice plate rests on the top plate and both are q
drilled through with 1/2 inch diameter holes ori the prescribed
_i pitch to accommodate the upper ends of the fuel pins'.
The-lower fuel pin support plate, a middle plate, and the upper fuel pin lattice plate are secured with tie rods and bolts.
The entire core structure is supported vertically and anchored by four posts set in the floor of the reactor tank.
- Finally, in the event the fuel pins are bowed but still satisfactory for use in the core, a plastic spacer plate may be installed on the middle plate.
Figure 4.6 of the SAR depicts the total.
core assembly.
- .s 5.4.3 Fuel Pins Fuel pins to be utilized are 4.8 weight percent enriched SPERT j
(F-1) fuel rods.
Each fuel rod is made up of sintered UO pellets,encasedinastainlesssteeltube,cappedonbotb ends with a stainless steel cap and held in place with a chromium nickel spring. An aluminum oxide (Al 0 insulator between the fuel pellets and stainless steel c$p$)on each end of the rod is installed.
Gas gaps to accommodate fuel expan-sion are also provided at both the upper end and around the fuel pellets.
Figure 4.7 of the SAR depicts a single fuel pin and its pertinent dimensions.
5.4.4 Q ntrol Rod Assemblies
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Four control rod assemblies are installed, spaced 90 degrees apart at the core periphery.
Each rod consists of a 6.99-cm
-j square stainless steel tube which passes through the d
core and rests on a hydraulic buffer on the bottom carrier l
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t 5-3 plate of the support structure, lioused in each of.these
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" baskets" are two neutron-absorber sections, one positioned above the other as depicted in Figure 4.8 of the SAR. The combination of the four rods must meet the values given in Table 5.2 of the SAR, with regard to reactivity with one stuck rod and shutdown margin.
l 5.5 Water Handling System The water handling system allows remote filling and emptying of the reactor tank.
It provides for a water dump by means of a fail safe butterfly-typr gate valve when a reactor scram is initiated. The filling system shall.be controlled by the operator who must satisfy the sequential interlock system before adding water to the tank. A pump is provided to add the moderator-reflector water from the storage dump tank into the' reactor tank.
Slow and fast fill rates of about 10 gpm and 50 gpm, respectively, are provided. A nominal six inch L
valve is installed in the dump line and has the capability.of emptying the reactor tank on demand of the operator or when a reactor scram is
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initiated, unless bypassed with the approval of the licensed senior operator on duty. A valve is installed in the bottom drain line~of 1
the reactor tant. to provide for completely emptying the reactor tank,
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5.E Fuel Storage and Transfer 1
When not in use, the fuel shall be stored within the storage vault J
l located in the reactor room. The vault shall be closed by a locked t
door and shall be provided with a criticality monitor near the vault door. The fuel shall be stored in cadmi"m clad steel tubes with no d
more than 1 kg fuel per tube mounted o" a steel wall rack. A storage tube in the storage vault cannot contain more than 15 SPERT (F-1) fuel pins at any time. The center-to-center spacing of the storage tubes together with the cadmium clad steel tubes assures that the infinite multiplication factor is less then 0.9 when flooded with water.
All fuel transiers shall be conducted under the direction of a licensed senior operator.
Operating personnel shall be familiar with health physics procedures and monitoring techniques and shall monitor the operation with appropriate radiation instrumentation.
For a completely unknown or untested system, fuel loading shall follow the inverse multiplication approach to criticality and, thereafter, meet Specification 4.2.
Should any interruption of the loading occur (more than four days), all fuel elements except the initial loading l
l step shall be removed from the core in reverse sequence and'the
-l operation repeated.
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For a known system, up to a quadrant of fuel pins may be removed from the core or a single stationary fuel pin be replaced with another stationary pin only under the following conditions:
1.
The net change in reactivity has been previously determined by measurement or calculation to be negative or less than 0.205.
l 2.
The reactor is subcritical by at least 1.005 in reactivity.
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There is initially only one vacant position within the active fuel lattice.
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4.
The nuclear instrumentation is one scale and the dirmp valve is l
not bypassed.
5.
The critical rod bank position is checked after the operation I
is conplete.
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6-4 b)
Existing operating procedures for adequacy and to assure that they achieve their intended purpose in light of any changes since their implementation; c)
Plant equipment performance with~particular attention to operating anomalies, abnormal occurrences, and the steps taken to identify and c?rrect their use.
6.2 Procedures Written procedures shall be prepared, reviewed and approved prior to.
initiating any of the activities listed in this section, The proce-dures, including applicable check lists, shall be reviewed by the NSRB and followed for the following operations:
l 1)
Start up, operation and shutdown of the reactor.
2)
Installation and removal of fuel. pins, control rods,
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experiments, and experimental facilities.
3)
Corrective actions to be taken.to correct specific and foreseen
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malfunctions such as for power failures, reactor scrams, radia-tion emergency, responses to alanns, moderator leaks and abnormal reactivity changes.
4)
Periodical surveillance of reactor instrumentation and safety systems, area monitors, and continuous air monitors.
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5)
Implementation of the facility security plan.
6)
Implementation of facility emergency plan in accordance with 10 CFR 50, Appendix E.
7)
Maintenance procedures which could have an effect on reactor j
safety.
j Substantive changes to the above procedures shall be made only with the prior approval of the NSRB.
Temporary changes to the procedures that do not change their original intent may be made with the approval of the Operations Supervisor. All such temporary changes to the pro-j cedures shall be documented and subsequently reviewed by the Nuclear l
Safety Review Board.
l 6.3 Experiment Review and Approval 1)
All new experiments or classes of experiments that might involve an unreviewed safety question shall be reviewed by the
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Nuclear Safety Review Board. NSRB approval shall assure that compliance
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o t,
a.
6-6 6.4.2 Action to be Taken in the Event of an Occurren:e of the Type Identified in Section 1.0 Q (Reportable Occurrence) a)
Reactor conditions'shall be returned to normal or the reactor shall be shut down.
If it is necessary to shut down the reactor to correct the occurrence, operations shall not be resumed unless authorized by the Facility Director or designated alternate.
b)
Occurrence shall be reported to the Facility Director or designated' alternates and to the Comission as required.
c)
All such conditions, including action taken to prevent or reduce the probability of a recurrence, shall be review d by the NSRB.
6.5 Reports In addition to the requirements of applicable regulations, and in no way substituting therefore, all written reports shall be ~sent to the
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U.S. Nuclear Regulatory Comission, Attn:
Document Control Desk, Washington, D.C. 20555, with a copy to the Region I Administrator.
6.5.1 Operating Reports l
A written report covering the previous year shall be submitted k
by March 1 of each year.
It s1all include the following:
a) Operations Summary. A sumary of operating experience occurring during the reporting period that relate to the safe operation of the facility, including:
1)
Changes in facility design; l
l 2)
Performance characteristics (e.g., equipment and l
fuel performance);
3)
Changes in operating procedures which relate to the safety of facility operationr.
4)
Results of surveillance tests and inspections required by these Technical Specifications; 5)
A' brief sumary of those changes, tests, and experi-ments which require authorization from the Comission pursuant to 10 CFR 50.59(a), and; 6)
Changes in the plant operating staff serving in the folle 8ng positions:
a).
Facility Director; 1
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1_
l n m.
. +.
6-7 1
b)
Operations Supervisor; I
c)
Health Physicist; 1
d)
Nuclear Safety Review Board Members.
b) Power Generation. A tabulation of the integrated thermal power during the reporting period.
l c) Shutdowns. A listing of unscheduled shutdowns which have I
occurred during the reporting period, tabulated according l
to cause, and a brief discussion of the preventive action 1]
taken to prevent recurrence.
l l
d) Maintenance. A tabulation of corrective maintenance (ex-cluding preventative maintenance) performed during the reporting period on safety related systems and components.
l I
I e) Chances, Tests and Experiments. A brief description and a i
summary of the safety evaluation for all changes, tests, and experiments which were carried out without prior Comission approval pursuant to the requirements of 10 CFR Part 50.59(b).
f) A summary of the nature, amount and maximum concentrations
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of radioactive effluents released or discharged to the environs beyond the effective control of the licensee as 2-i measured at or prior to the point of such release or dis-
~
charge.
_.l g) Radioactive Monitoring.
A summary of the TLD dose rates taken at the exclusion area boundary and the site boundary during the reporting period.
i l
h) Occupational Personnel Radiation Exposure.
A surmtary of l
radiation exposures greater tnan 25% of the values allowed l
by 10 CFR 20 received during the reporting period by faci-l lity personnel (faculty, students or experimenters).
6.5.2 Non-Routine Reports l
l a)
Reportable Operational Occurrence Reports.
Notiffre-tion 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 Administrator of Region I followed by
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i a written report within 10 days in the event of a reportable operational occurrence as defined in Section 1.0.
The written report on these reportable operational occurrences, and to the extent possible,
____._____________m____m___
6-8 the preliminary telephone and telegraph
- notification shall:
(1) descrite, analyze, and evaluate safety implications; (2) outline the measures taken to ensure that the cause of the condition is determined; (3) indicate the corrective action (including any changes made to the procedures and to the quality assurance program) taken to prevent repetition of the occurrence and of similar occurrences involving similar components or systems; and (4) 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.
b)
Unusual Events. A written report shall be forwarded within 30 days to the Administrator of Region I in the l
event of:
- 1) 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
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l Specifications.
6.6. Operating Records l
6.6.1 The following records and logs shall be maintained at the Facility or at Rensselaer for at least five years.
a) Normal facility operation and maintenance.
_b b) Reportable operational occurrences.
l c) Tests, check, and measurements documenting compliance with surveillance requirements.
d) Records of experiments performed.
e) Records of radioactive shipments.
6.6.2 The following records and logs shall be maintained at the Facility or at Rensselaer for the life of the Facility.
a) Gaseous and liquid radioactive releases from the facility.
I b) TLD environmental monitoring systems.
- Telegraph notifi ation may be sent on the next working day in the event of a c
reportable operational occurrence during a weekend or holiday period.
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