ML20094E763
| ML20094E763 | |
| Person / Time | |
|---|---|
| Site: | Vallecitos Nuclear Center |
| Issue date: | 07/31/1984 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20094E761 | List: |
| References | |
| 80NEDO26, NEDO-12725A, NUDOCS 8408090337 | |
| Download: ML20094E763 (51) | |
Text
,
n MEDO-12725A MGd d NEDO-1272$A- 4 CZOOOOGO M30GV HGd ,
. 5080tB LCCO6000t8 .80NED026'
~ Class-I' July 1984-oci Cg_ 'y PROPOSED TECHNICAL SPECIFICATIONS FOR THE GENERAL' ELECTRIC NUCLEAR TEST REACTOR P
._.c-x- .v l
i FACILITY LICENSE R-33 DOCKET NO. 50-73 4
l
NED0-12725A 1
FOREWORD These proposed Technical Specifications were derived from'the General' Elect'ric.
Nuclear Test Reactor Safety Analysis Report, NEDO-12727, dated-April, 1981~, and represent those parametero and requirements which define the boundaries of; e licensed activities.
These proposed Technical Specifications supercede and nullify NEDO-12725, dated April, 1981. These proposed Technical Specifications revise those proposed.
earlier to recognize Technical Specification changes-recently approved by_the NRC;s NRC staff review comments in the letter C. O. Thomas, Division of Licensing, to.
R. W. Darmitzel, General Electric Company' dated June 1, 1984; and the more recent Standard for the Development of Technical Specifications.for Research Reactors, ANS-15.1 (1984). Permitted exceptions, deletions and additions-appear because of
~
the unique nature of the NTR. Technical Specification 2.0,-Safety Limits (ANS-15.11 numbering system) is not included since there are no postulated or anticipated.
events which could cause fuel melt. Technical Specifications 3.4,. confinement.or.
Containment, and 3.5, Ventilation System-(ANS-15.1 numbering system) have been, cochined in NTR Technical Specification 3.4, Reactor Cell-and Ventilation System.
This was done because there is only one Technical Specification:for the Reactort Cell (confinement) and it is logical to include it with the ventilation system.1 Technical Specification 3.6, Emergency Power (ANS-15.1 numbering system) is not included since this does not apply to NTR. Technical Specification 3.7, Radiation Honitoring Systems and Effluents (ANS-15.1 numbering system)-is included'in NTR
~
Technical Specification 3.2.3.6 (including Table-3-2) and TechnicaliSpecificationJ :F': FN1 3.4 because this appears to be the more logical grouping for NTR. : Technical Specification 6.4, Experiment Review and Approval (ANS-15.1 numbering [ system).is: -
included in NTR Technical Specification 6.2.3-i
,?
- ~Q
~
=
~
]c LD
.iI',
NED0-12725A 80NED026 Class I July 1984 J4 , 47"i * '.p 1s>A d PROPOSED TECHNICAL SPECIFICATIONS FOR THE GENERAL ELECTRIC NUCLEAR TEST REACTOR t
n FACILITY LICENSE R-33 DOCKET NO. 50-73 8408090337 840B02 PDR ADOCK 05000073 P PDR 4
=
. , . . . . - - - , , ,- , , . , , , . .w.- ,. -,,-. . , , ~ , , , + . , . , , , . . - - - , , . ~ , , . . .
NEDO-l?725A 1
ya FOREWORD
' These proposed Technical Specifications were_ derived from the General Electric Nuclear. Test Reactor Safety Analysis Report, NEDO-12727, dated April, 1981, and represent those parameters and requirements which define the boundaries of liccused activities.
These proposed Technical Specifications supercede and nullify NEDO-12725, dated
. April,~1981. These proposed Technical Specifications revise those proposed earlier to recognize Technical' Specification changes recently approved by the NRC; g.,_ . .
NRC staff review comments in the letter C. O. Thomas, Division of Licensing, to R. W. Darmitzel, General Electric Company dated June 1, 1984; and the more recent
- Standard for the Development of Technical Specifications for Research Reactors.
ANS-15.1 (1984). Permitted exceptions, deletions and additions appear because of
. the unique nature of the NTR. Technical Specification 2.0, Safety Limits (ANS-15.1 numbering' system) is not included since there are no postulated or anticipated
-events which could cause fuel melt. Technical Specifications 3.4, Confinement or Containment, and 3.5, Ventilation System (ANS-15.1 numbering system) have been combined in NTR Technical Specification 3.4, Reactor Cell and Ventilation System.
This was-done because there is only one Technical Specification for the Reactor Cell (confinement) and it is logical to include it with the ventilation system.
Technical Specification 3.6, Emergency Power (ANS-15.1 numbering system) is not included since this does not apply to NTR. Technical Specification 3.7, Radiation
. Monitoring Systems and Effluents (ANS-15.1 numbering system) is included in NTR
' Technical Specification 3.2.3.6 (including Table 3-2) and Technical Specification
. 3.4-because this appears to be the more logical grouping for NTR. Technical Specification 6.4,. Experiment Review and Approval (ANS-15.1 numbering system) is included in NTR Technical Specification 6.2.3 i
O 9
_ _ - - - - . , . _ . , . - , _ , . . . _ _ . , _ , _ . . . . . , , - _ _ _ _ _ . _ _ . _ ,_ . . . . - , . . - m.
NEDO-12725A CONTENTS P,, age
-1.0 DEFINITIONS 1 2.0 LIMITING SAFETY SYSTEM SETTING (LSSS) 7 3.0 LIMITING CONDITIONS FOR OPERATION 9 3.1 Reactor Core Parameters 9 3.2 Reactor Control and Safety System 11 3.3 Reactor Coolant System 16
- 3. 4 ~ Reactor Cell and Ventilation System 18 3.5 Experiments 21
.4.0 SURVEILLANCE REQUIREMENTS 25 4.1 Reactivity Limits 25 4.2 Reactor Control and Safety System 27 4.3 Reactor Coolant System 31 4.4 Reactor Cell and Ventilation System 31 4.5 Experiments 32 5.0 DESIGN FEATURES 35 5.1 Site and Facility Description 35 5.2 Reactor Coolant System 36 5.3 Reactor Core and Fuel 36 5.4 Fissionable Material Storage 36 6.0 ADMINISTRATIVE CONTROLS 37 6.1 Organization and Staffing 37 6.2 Independent Review 41 6.3 Procedures 42 6.4 Required Actions 43 6.5 Reports 44 6.6 Records 47 it
.~ .
_ \
l TABLES l
1 Table Title Pjgte 43-1 Reactor. Safety System - Scram 12 3-2 Reector Safety System - Information 13 3-3 Stack Release Rate Limits 19 4-l' Surveillance Requirements of Reactor Safety System Scram Instruments 28 4-2 Surveillance Requirements of Reactor Safety c;-- = ;j;;
e- System Ia'a--=*daa Ta-+-a-a-** 30 ILLUSTRATIONS r: Figure Title, M 6-1 Facility Organization 38 iii
NEDO-12725A 1.0 DEFINITIONS 1.1. Channel
.The combination of sensors, lines, amplifiers and output devices which are con-nected for the purpose of measuring the value of a parameter.
.1.2 Channel Calibration A comparison and/or an adjustment of the channel so that its output corresponds with acceptable accuracy to known values of the parameter which the channel
-: ;, mamaures. Calibration shall-encompass the entire channel if reasonabic, including equipment actuation, alarm, or trip test and shall include the Channel Test.
1.3 Channel Check A qualitative verification of acceptable performance by observation of channel behavior. This verification where possible shall include comparison of the channel with other independent channels or systems measuring the same variable.
1.4 Channel Test The introduction or interruption of a signal into the channel to verify that it is operable.
1.5 Experiment
' An experiment (also see Secured Experiment) is any of the following:
- s. An activity utilizing the reactors experimental facilities or its
, components or the neutrons or radiation generated therein.
- b. An evaluation or test of a reactor system operationni, surveillance, or maintenance technique.
, l.
1
at-k NEDO-12725A
- c. 'The material content of any of the preceding, including structural components, encapsulation or confining boundaries, and contained fluids or solids.
1.6 Experimental Facility Any. location for experiments which is on or against the external surfaces of the reactor main graphite pack, thermal column, or within any penetration thereof.
-1.7 Explosive Material
'Any chemical compound or mixture, the primary or common purpose of which 18 to function by explosion, i.e., with substantially instantaneous release of gas and heat.
1.8 Facility That portion of the building and adjacent outside areas occupied by the reactor, reactor control room, and associated support areas.
1.9 Flammable A flammable liquid is any liquid having a flash point under 100*F. A flammabJe solid is any solid material, other than one classified as an explosive, which is liable to cause fires through friction or which can be ignited readily and when ignited burns so vigorously and persistently as to create a serious hazard.
< Flammable solids include spontaneously combustible and water-reactive materials.
1.10 Licensed Operator A ' person who is licensed as a reactor operator (RO) or senior reactor operator (SRO) pursuant to 10CFR55 to operate the controls of the Nuclear Test Reactor 2
p
~
11.11 Limitina Conditions of Operation (LCO)
The lowest functional capability or performance levels of equipment required for safe operation of the facility.
l.12 Limitina Safety System Settinas (LSSS)
Settings for automatic protective devices related to those variables having significant reactor safety functions.
1.13 Measured Value The measured value of a parameter is the value as it appears at the output of a channel. i
- . ~
1.14' Operable A system or component is operable when it is capable of performing its intended
, function.
1.15 Potential Excess Reactivity l
That excess reactivity which can be added by the remote manipulation of control rods plus the maximum credible reactivity addition from primary coolant tempera-ture change plus the reactivity worth of all installed experiments.
l 1.16 Reactivity Worth (Experiment) l l
The reactivity worth of an experiment is the maximum absolute value of the reac-tivity change that would occur as a result of intended or anticipated change or credible malfunctions that alter experiment position or configuration.
1.17 Reactor Operatina (Reactor Operation)
The reactor is considered to be operating when all of the following conditions are satisfied:
3
.- .. -,-..,,-.,..,,,,,..,-,m,-,-n,,_,,.,,,._, ,n,.,,_,_.-,,n, , , , , , _ , ._,__,n,, . _ _ _ , , , , , , ,n_-,.,,,_
e, -
NEDO-12725A v" .
- s. Console key is in the console key switch and the console is energized.
~
- b. More than one of the installed safety and/or control rods are withdrawn,
- c. -Fotential excess reactivity is greater than, or equal to, aero.
1.18 Reactor Thermal Power
~ The reactor thermal power, as determined by a primary coolant system heat balance.
1.19 Reactor Safety Systems Reactor safety systems are those systems, including their associated input chan-nels, which are designed to initiate automatic reactor protection or to provide information for initiation of manual protective action.
1.20 Reactor Secured
, The reactor is considered secured under either of the following two conditions:
(1) The core contains insufficient fissile material to attain criticality under optimum conditions of moderation and reflection.
(2) That overall condition where all of the following conditions are satisfied
- a. Reactor is shut down.
- b. - Console keylock switch is OFF and the console key is in proper custody.
- c. No work is in progress involving in-core components, installed rod drives, or
- experiments.
4 e
4
..y y._m %- .e -
y--y,ww ew----+v
t
.i .
NEDO-12725A 11.21 Reactor Shutdown That suberitical condition of the reactor where the negative reactivity of the i Eenon-free core would be equal to or greater than the minimum shutdown margin. f 1.22 Readily Available on Call (Senior Reactor Operator) l T
u ,'A. senior reactor operator is readily available on call when all of the following
, conditions are satisfied:
s t-
.x1 __ ci a. _ Is .within a reasonable driving time (1/2 hour) from the reactor facility.
5,
': ,b. '.Can be promptly contacted by telephone;-and h c. Has . informed the reactor operator on duty where he may be contacted. t i.
4-r v :1.23 Secured Experiment L
'Any experiment, experimental facility, or component of an experiment that is held
. Lin's stationary position relative to the reactor by mechanical means. The re-i' straining' forces must be substantially greater than those to which the experiment '
.might.be subjected by hydraulic, pneumatic, buoyant, or other forces which are 7 snormal to,the operating environment of the experieent, or by forces which can
, arise as a result of credible natural phenomena or malfunctions.
o tl.24 Shutdown' Martin
- Shutdown margin shall mean the shutdown reactivity necessary to provide confidence that the reactor can be ma'e d suberitical by means of the control and safety
. . systema. starting from any permissible operating condition, although the most *
- reactive rod is stuck in its most reactive position, and that the reactor will
- + remain subcritical without further operator action.
1.25 Site The area'(approximately 1600 acres) within the confines of the Vallecitos Nuclear Center'(VNC) owned and operated by General Electric.
5 I? ._,.______..,___
- f. . . '
'I.26 True Value The true value for s' parameter is its exact value at any Instant.
1.27'Unacheduled Shutdown Any unplanned shutdown of the reactor caused by actuation of the scram channels, operator error, equipment malfunction, or a manuy1 shutdown in responce to con-ditions which could adversely affect safe operat)en excluding shutdowns which occur during planned equipment testing or check-out operations.
. ,u .
l2 '
4 0
6 -
h NEDO-12725A v:
2.0 LIMITING SAFETY SYSTEM SETTING (LSSS) 2.1 Applicability This specification applies to the scram set point for the linear neutron channels which monitor reactor power level.
2.2 Objective The objective of this specification is to insure that automatic action will prevent the most severe postulated or anticipated transient from causing fuel damage.
2.3 Specification The linear neutron power monitor channel set point shall not exceed the measured value of 125 kW except for time periods of less than four (4) hours during instrument and channel calibrations when the set point may be increased not to exceed 138 kW.
2.4 Bases The limiting safety system setting (LSSS) has been chosen to ensure that reactor scram is initiated in time to prevent significant fuel degradation resulting from excessive reactor thermal power during normal operation, anticipated operational occurrences, and postulated transients. The safety margin (the difference between the safety limit and the LSSS) includes systematic and random types of instrument uncertainties including instrument drif t and calibration uncertainty and, for transient events, also includes system delay times. The value of 138 kW is derived from the trip point of 150 kW used in the analysis of postulated accidents
. (GAR, Sec. 11). The 125 kW setting is used because this is the currently preferred value for research reactors.
}
4 7 (Next page is p. 9) s
i NEDO-12725A 3.0 LIMITING CONDITIONS FOR OPERATION (LCO) 3.1 REACTOR CORE PARAMETERS 3.1.1 Applicability This specification applies to the reactivity condition of the reactor and to the reactivity worths of control rods, safety rods, manual poison sheets, experiments and the coolant temperature coefficient of reactivity.
, 3.1.2 Objective The objective of this specification is to ensure the reactor can be safely con-trolled at all times and shut down when required.
3.1.3 Specifications 3.1.3.1 The reactor configuration shall be controlled to ensure that the potential excess reactivity shall be 50.76$. If it is determined that the potential excess reac-tivity is >0.76$. the reactor shall be shut down immediately. Corrective action shall be taken as required to ensure the potential excess reactivity is 5 0.76$.
3.1.3.2 The reactor shall be suberitical whenever the four safety rods are withdrawn from the core and the three control rods are fully inserted.
3.'1.3.3 The minimum shutdown margin shall be 1$.
9 i .
+ +-
NEDO-12725A 3.1.3.4 Each manual poison sheet used to satisfy the requirements of Specification 3.1.3.1 shall be restrained in its respective graphite reflector slot in a manner which will prevent movement by more than 1/2 inch relative to the reactor core.
3.1.3.5
. The temperature coefficient of reactivity of the reactor primary coolant shall be negative above a primary coolant temperature measured value of 124*F.
3 '.' 1. 4 . gases s
. Operation in compliance with Specification 3.1.3.1 ensures that there would not be any mechanism for addition of reactivity greater than 0.76$. Detailed analyses have been made of reactivity insertions in NEDO-12727, the NTR Safety Analyses Report (SAR). The analyses show that a reactivity step addition of 0.76$ will not
' cause significant fuel degradation.
- Operation in accordance with Specift :ation 3.1.3.2 ensures that criticality will not be achieved during safety rod withdrawal. Adherence to the 0.760 limit also ensures that the reactor will not go critical during safety rod withdrawal.
. 0peration in accordance with Specif' cation 3.1.3.3 ensures that the reactor can be
, . brought and maintained suberitical without further operator action under any
. permissible operating condition even wtth the most reactive safety rod stuck in its most reactive position.
Operation in accordance with Srecification 3.1.3.4 ensures that the manual poison sheets will not be recoved from the reactor core during the maximum postulated seismic event.
Operation in accordance with specification 3.1.3.5 ensures there is no significant positive reactivity feedback from coolant temperature change during reactor power transients.
10 m-
-97 p1 NEDO-12725A 3.2 REACTOR CONTROL AND SAFETY SYSTEM 3.2.1 Applicability L
This specification applies to the reactor safety rod, control rod and reactor safety systems.
3.2.2 Objective The objective of this specification is to specify the lowest acceptable level of performance to reasocebly ensure proper operation of the reactor safety rod, l control rod and reactor safety systems. !
3.2.3 Specifications r
s 3.2.3.1 Reactor operation shall be permitted only when all safety and control rods are operable. The reactor shall be shut down immediately if it is known that a safety or control rod is not operable.
3.2.3.2 No more than one safety rod shall be allowed to be moved in an cutward direction at a time.
3.'.3.3 e
i t
The rate of withdrawal of each safety rod drive diring reactor operation shall be less than 1-1/4 inches per second.
t i 3.2.3.4 The rate of withdrawal of each control rod drive during reactor operation shall be less than 1/6 inch per second.
I i
,-._m.___,-..._-_--_,--_,,_..,-,_.m . , . . _ - . _ . - _ . , _ , . . .,__,,,,,--_,,-.,,x.__--,~ . - - - - , .-,_r__~
c:
NEDO-12725A REACTOR SAFETY SYSTEM-SCRAM Table 3-1 Itenf _ system- Condition Trip Point Function (measured values)
- 1. Linear High reactor power No higher than 125kW* Scram (2 out of 3 or 1 out of 2)
Loss of positive high No less than 90% of Scram (2 voltage to ion chambers operating voltage out of
=2 -
(if-fon chambers are usad) 3 or 1 out of 2)
- 2. Los N Fast reactor period No less than +5 see Scram
' Amplifier Mode switch N/A Scram not in operate Loss of positive high No lsss than 90% of Scram voltage to ion chambers operating voltage (if ion chambers are used) 3.- Primary Coolant Nigh core outlet No greater than 222'T Scram i Temperature temperature
- 4. Primary Coolant Low Flow No less than 15 sps when Scram Flow reactor power is >0.1 kW S. Manual' ' Console button depressed N/A Scram
- 6. glectrical Power Reactor console key in N/A Scram off. position (loss of ac power to the. console)
- 7. ,
seismic Vibratory ground motion Modified Mercalli IV Scrum
- The Trip Point may be adjusted up to 138 kW for time periods of less than four (4) hours during instrument and channel calibrations.
12 W
p e
NEDO-12725A REACTOR SAFETY SYSTEM-INFORMATION Table 3-2 Item No. System Condition Set Point
- Function lI"~ Reactor cell Low Differential >0.5 in, water AP Visible & audible alarm; Pressure pressure audible alarm may be by-passed after recognition.
- 2. Puel Loading Low Level <3-ft below the Visible & audible alarm; Tank Water overflow audible alarm may be Level bypassed after recognition
- 3. Primary Coolant High core outlet <200*F Visible & audible alarm; Temperature temperature audible alarm may be by-passed after recognition.
- 4. Primary Coolant Core Delta N/A Provide information for A Temperature temperature the heat balance determination.
- 5. Reactor Cell High Level 5Five times Visible & audible alarm; Radiation reactor full Audible alarm may be by-Monitor power back- passed after recognition.
ground level
, 6. Stack High Level At a level to Visible & audible alarm; Radioactivity ensure com- Audible alarm may be re-pliance with set after recognition.
Specs.3.4.3.3 &
3.4.3.4
- 7. Linear Power Low Power k2% on any scale Safety or control rods indication cannot be withdrawn (2 out of 3 or 1 out of 2).
- 8. Log Power Nigh Power 5 140 kW Audible & visual alarm.
Indication
- 9. Control or Rods not in N/A Safety rod magnets cannot Safety Rod be reenergized.
- 10. Safety Rod Rods not out N/A Control rods cannot be withdrawn; safety rods must be withdrawn in sequence; may be by-passed to allow with-drawal of one control rod, or one safety rod (drive) out of sequence for purposes of inspec-tion, maintenance and testing.
- Setpoint values are the nominal measured values and need not take into account the uncertainty of the channel.
13 I
77 . --
NEDO-12723A i 4 .., .
i The average scram time (inflight time) of the four safety rods shall not exceed
'270 meec.
3.2.3.6 Reactor operation shall be permitted only when the reactor safety system is opera-ble in accoadance with Tables 3-1 and 3-2.
The reactor shall be shut down immediately if any portion of the reactor safety system malfunctions except as provided for in Tables 3-1 and 3-2.
3.2.4 Bases
' Operation in accordance with Specification 3.2.3.1 ensures that during normal operation adequate shutdown margin is provided.
.0peration in accordance with specification 3.2.3.7 and specification 3.2.3.3 ILmits the rate of reactivity addition during safety rod withdrawal to that from one safety' rod. This value is easily controlled by the operator.
Operation in accordance with specification 3.2.3.4 limits the rate of reactivity addition during control rod withdrawal. Experience has shown that this is a value which is easily manually controlled by the operator. This rate is also less than the value analysed in the rod withdrawal accident in the SAR.
I '
Operation in secordance with Specification 3.2.3.5 ensures that the safety rod system performs satisfactorily. The specified time is the inflight time originally established for this type reactor when higher potential excess reestivities were permitted. With the current limit on potential excess reestivity (see Technical specificaton 3.1.3.1), a scram is not required during postulated events to prevent significant fuel degradation (see SAR, Section
'11.4.3). Maintaining the safety rod system, then, is conservative.
14 a
c ,
MEDO-12725A Operation in accordance with Specifications 3.2.3.6 ensures that the reactor safety system is adequate to control operation of the facility, measure operating parameters, warn of abnormal conditions, and scraa the reactor automatically if required.
The bases for items listed in Table 3-1 are as follows:
The linear high reactor power scram will be set no higher than the LSSS. Scram action as a result of a predetermined decrease of positive high voltage to ion chambers for the linear channels provides assurance that the high voltage power supply is functioning and the ion chsabers are operating on a flat portion of the 1-V curve.
The fast period scram limits the rate of rise of the reactor power to periods which are manually controllable. The Log N amplifier mode switch scram ensures that the Log N amplifier is in the Operate Mode. Screa action as a result of loss of positive high voltage to the ion chamber for the Log N channel provides assur-ance that the high voltage power supply is functioning and the ion chamber is operating on a flat portion of the I-V curve.
The primary coolant high core outlet temperature scram provides assurance that the reactor will be shut down if the primary coolant outlet temperature is high.
The primary coolant low-flow scram provides diversification in the safety system to ensure, when the reactor is at power levels which require forced cooling, that the reactor will be shut down if sufficient primary coolant flow is not main-tained.
The manual console scram button provides a method for the reactor operator to manually shut down the reactor if an unsafe or abnormal condition should occur and th's automatic reactor protection action as appropriate does not function.
The loss of electrical power with the reactor console key in the off position (loss of se power to the console) means that the reactor cannot be operated because ac power is no longer provided to the reactor safety system.
The bases for items listed in Table 3-2 are as follows:
15
~.
-The reactor cell low differential pressure alarm gives adequate assurance that operation of the reactor will be in compliance with specification 3.4.3.1.
The fuel loading tank low water level alarm gives adequate assurance tha: opera-tion of the reactor will be in compliance with specification 3.3.3.1.
The primary coolant high core outlet temperature alarm gives adequate assurance that warning vill be given to the operator of high primary coolant core outlet temperature.
The use of the area radiation monitor is to detect a release of radioactive material to the primary coolant water.
The stack radioactivity high level alarm gives adequate assurance that operation of the reactor will be in compliance with specification 3.4.3.2.
The low power indication on the linear channel ensures that the operator has a linear power channel operating and indicating neutron flux levels during rod withdrawal.
The high power indication on the log power channel provides a backup high power warning on a diverse type instrument.
The control rods "not in" interlock ensures that the reactor will be started up by withdrawing the four safety rods prior to withdrawing the control rods.
The safety rode "not-out" interlock ensures that the method of reactivity centrol is with the control rods during reactor operation.
3.3 REACTOR C001. ANT SYSTDf 3.3.1 Applicability This specification applies to the reactor primary coolant system.
16
_- _ . _ . . . . _ . _. . _ . - . - _ _ _ . _ _ _ _ _ _ _ . - _ _ _ _ ~ . . _ _ _ _ _ . . - _ - _ _ _ _ _ _ . _ _
r i
3.3.2 Obj ective The objective of this specification is to minimize the adverse effects on reactor components and to ensure the proper conditions of the coolant system for reactor operation.
3.3.3 Specification 3.3.3.1 Reactor operation shall not be permitted unless the core tank is filled with water. It during operation of the reactor it is determined or suspected that the core tank is not filled with water, the reactor will be shut down immediately and l corrective action will be taken as required.
3.3.3.2 The specific conductivity of the primary coolant water shall be maintained less than 10 pahos/cm except for time periods not exceeding 7 consecutive days when the specific conductivity may exceed 10 pahos/cm but shall rea:in less than 20 pmhos/ca. If the specific conductivity exceeds 10 unhos/ca, steps shall be taken to assure the specific conductivity is reduced to less than 10 pahos/cm.
3.3.4 Bases operation in accordance with Specification 3.3.3.1 encures that there will be no reactivity insertions due to the removal of voids or the sudden addition of water into the core tank during reactor operation.
-The minimum corrosion rate for aluminum in water (< 50'C) occurs at a pH of 6.5.
Maintaining water purity below 10 pahos/cm will maintain the pH between 4.8 and O.7. These values are acceptable for NTR operation. High specific conductivity can be tolerated for shorter durations during unusual circumstances. Oparation in accordance with Specification 3.3.3.2 ensures aluminum corrosion is within accept-able levels and that activation of impurities in the primary water remain below hasardous levels.
17
- W'
. l NEDO-12725A i 3.4 REACTOR CELL AND VENTILATION SYSTEM 3.4.1 Applicability l
~
This specification applies to the reactor cell sad ventilation system.
3.4.2 Obj ective I
The objective of this specification is to ensure the release of airborne radioac-tive materials is below authorized limits.
3.4.3 Specifications 3.4.3.1 Reactor power shall nct be increased above 0.1 kW unless the reactor cell is I maintained at a negative pressure of not less than 0.5 in. of water with respect '
to the control room. If during operation of the reactor above 0.1 kW the nega-tive pressure with respect to the control room is not maintained, the reactor power shall be lowered to 5 0.1 kW immediately and corrective action shall be teken as required.
3.4.3.2 The limits for radioactive material discharged through the reactor ventilation ;
system shall be as specified in Table 3-3.
I t
18
NED0-12725A Table 3-3 STACK RELEASE RATE LIMITS
-l
- 1setose Group Anniual Averase (9C1/sae) _Short Term (uct/see) taloge.1, > 34 T 1/2 s (2.0 x 10 ) x (MFC,) (1.4 x 10') x (MPC,)
Partievieta,'> Sd 7'/2
~
' 1 7
,, tota Camma (2.0 x 10 ) x (MFC,) (1.4 x 10') x (MPC,)
'(,[,Alrha '
7 (2.0 x 10 ) x (MFC,) (1.4 x 10') x (MPC,)
U M1othor(includias. (1.4 x 1010) x (MPC,) (1.0 x 10 II) x (MPC,)
r
' Noble Cas) 7 '
/
where MPC = the concentratf.on in WLi/ml shown in Table II, Appendix de 10 CFR, l 2 Part 20.
~ '
'3.4.3.3
~
Alans points for particulate and noble gas' continuous monitors shall not exceed a value corresponding to the annual average release rate limit for the most restric-tive isotope in the category (except as specified in 3.4.3.4):
c i.
s Me, niton kalease Rate L
geta-Camma perticulate s Acnual Average of Sr-90 Limit Noble Cen 5 Annetal Average of Kr-87 limit 4 ..
.) 3.4.3.4 Alarm pef"nts may be raised to a value corresponding to the short tern release rate, not to exceed the short tern release limit, provided grab samples are taken and analysed for isotopic content. The short term release rate shall be based on the race of release of the seat restrictive isotope. The releases shall be monitored to ensure that the annual release limit is not exceeded.
w- 19 N ~
NEDO-12725A 3.4.3.5 t
During operation of the reactor above 0.1 kW or the performance of activities that seuld release radforetivity to the ventilation system, the stack particulate activity monitor and the gaseous activity monitor shall be operating.
If either the gas or particulate monitor is not operable, the reactor shall be shut down, or the activity involving releases shall be terminated, or the unit shall be promptly repaired or replaced with one of comparable monitoring capabil-ity. During this period, any indication of abnormal reactor operation shall be cause to shut down the reactor immediately.
3.4.4 Moses Operation in accordance with Specification 3.4.3.1 ensures that potentially contaminated reactor cell air due to reactor operation, is released through the ventilation system.
, The ventilation system release limits in Specification 3.4.3.2 are based un the fellowing:
The annual average dilution factor from the NTR stack to the site boundary based on 1976 and 1977 meteorological conditions and a stack flow rate of 3000 cu ft/ min equals approximately 20.000. That is, the concentration at the site boundary of any release from the NTR stack will be 51/20.000 of the concentration at the stack when averaged over 1 year.
The above listed annual average limit contains a reduction factor of 2 to account for discharges from other VNC stacks. Additionally, the release limits for all but the "all other" category contain a reduction factor of 700 to account for reconcentration in the environs.
The alarm points in specification 3.4.3.3 ate set for the release rate of the most restrictive isotope in the category which is equivalent to the annual average release limit.
20
y w- .
/ ji' ,'S .s ,
(,- s yx , NEDO-12725A h
The alarm points in Specification 3.4.3.4 will be set for the release rate of the )
most1 restrictive / isotope in the category which is equivalent to the short term .
i releas's limit. In no care will the annual release limit be exceeded. !
='. , l 3.5 EXPERIMENTS l
\
- 3. 5.1 ' Applicability I This specification applies to reactor experiments.
I 3.5.2 objective The objective of th'is specification is to prevent an experiment from resulting in a hazard to the operating personnel or the general public or damage to the reac-tor. ,
.g 3.5.3 Specifications s
, 3.5.3.1 '
e The reactivity worth of'all cr.periments shall be limited so that the sum of the rea,ctivity worths off,sil experimenta performed at any one time shall be limited to comply-with spscification 3.1.3.l.
/
~3.5.3.2 -
-~
s
<y:'s,, '
Ths a:aximum amount ,of' explosive ratterial permitted 'in the NT2 facilities is:
I-
- a. South Cell, W 5 (D/2)2~ with W 5 9 lbs and D 2 3 ft.
-b.. No th' room (withgut Edular Stone Monument) . W 5 D with W 5 16 lbs and I:21 It. ,
- c , ,
(' )
/t # f
- c. 'dortKRoom-(with Modulir Stone Honume'nt). W 5 2 lbs.
> =- .
d,- Setup Room.- W 5 23 lbs.'. ~
~ . ,
w3 g ,- )
.
- s dx
- where:
W-= Total weight of explosives in pounds of equivalent TNT D = Distance in feet from the South Cell blast shield or the north face of
.the North Room wall.
. 3.5.3.3 Experimental ~ objects shall not be allowed inside the core tank when the reactor is at a power. greater than.0.1 kW.
3.5.3.4 4.
Experimental objects located in the fuel loading chute shall be secured to prevent their' entry into the core region during reactor operation.
3.5.3.5:
A maximun'of 10 Ci of radioactive material and up to 50 g of uranium may be in storage in a neutron radiography area where explosive devices are present (i.e.,
cin:the South Cell or North Room).' The storage-locations must be at least 5 ft.
from any explosive device. Radioactive materials, other than those produced by
'the neutron radiography of the explosive devices and imaging systems, are not
~ permitted in the Setup Room if explosive material is present.
3.5.3.6' c
Unshielded high frequency generating equipment shall not be operated within 50 m.
Ifeet of any explosive devices.
- 3. 5. 3. 7 -
Experimental capsules to be utilized'in the experimental facilities shall be designed or tested to ensure that-any pressure transient produced by chemical
'22 fi ' . . . . _ . _ _ . _
-reaction of their contents and/or leakage of corrosive or flammable materials will not damage the reactor.
3.5.3.8 l ENperimental fuel elements containing plutonium to be utilized in the experimental facilities shall be clad and other experimental devices containing plutonium shall be encapsulated.
3.5.3.9-The maximum possible chemical energy release from the combustion of flammable substances contained in any experimental facility shall not exceed 1000 kW/sec.
The total possible energy release from chemical combination or decomposition of substances contained in any experimental capsule shall be limited to 5 kW/sec, if
.the rate of the reaction in the capsule could exceed 1 W. Experimental facilities containing flammable materials shall be vented external to the reactor graphite pack.
3.5.2.10 A~ written description and analysis of the possible hazards involved for each type of experiment shall be evaluated and approved by the facility manager, or his designated alternate, before the experiment may be conducted.
3.5.3.11 No irradiation shall be performed which could credibly interfere with the scram action of the safety rods at any time during reactor operation.
L -
3.5.3.12 I The radioactive material content, including fission products, of any singly encapsulated experiment to be utilized in the experimental facilities shall be i
i limited, so that the complete release of all gaseous, particulate, or volatile l
23 I
m
NEDD-12725A components from the encapsulation could not result in doses in excess of 10% of the equivalent annual deses stated in 10CFR, Part 20. This dose limit applies to persons occupying unrestricted areas continuously for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> starting at time of release or restricted areas during the length of time required to evacuate the restricted area.
3.5.3.13 The radioactive material content, including fission products, of any doubly encapsulated or vented experiment to be utilized in the experimental facilities shall be limited so that the materials at risk from the encapsulation or confining boundary of the experiment could not result in a dose to any person occupying an unrestricted area continuously for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> starting at the time of talease in excess of 0.5 rem to the whole body or 1.5 rem to the thyroid or a dose to any person occupying a restricted area during the length of time required to evacuate the restricted area in excess of 5 rem to the whole body or 30 rem to the thyroid.
3.5.4 Bases Operation in accordance with specification 3.5.3.1 ensures that there would not be any mechanism for addition of reactivity greater than 0.76$, including experi-ments. See the bases for Specification 3.1.3.1.
Specifications 3.5.3.1 through 3.5.3.11 are intended to reduce the likelihood of damage to the reactor components and/or radioactivity releases resulting from experiment failure and serve as a guide for the review and approval of new and untried experiments by the facility personnel.
Specifications 3.5.3.12 and 3.5.3.13 ensure the radiological effects of experiment failures do-not pose a hazard to the general public or to operating personnel.
24 Y _.
4.0 SURVEILLANCE REQUIREMENTS 4.1- REACTIVITY LIMITS 4.1.1- Applicability
'This specification ~ applies to the surveillance requirements for reactivity limits.
4.1.2 Obj ective To ensure that the reactivity limits of Specification 3.1 are not exceeded.
4.1.3 -Specification 4.1.3.1.
^ 7
. Potential excess reactivity will be calculated before each startup. Actual critical rod position shall then be used to verify that the measured value is 50.76$.
4.1.3.2 Neutron multiplication will be observed throughout each startup. Safety rod withdrawal shall be stopped if.it appears critical will be reached befere all safety rods are out.
- 4.1.3.3
, The'ainimum shutdown margin shall be determined by calculation or measurement whenever a decrease in the reactivity worth of a safety rod is suspected.
4.1.3.4 Each manual poison sheet in the core region of the reactor shall be verified to be properly restrained upon insertion.
25
- .- l I
NEDO-12725A 1 4.1.3.5 The temperature coefficient of reactivity of the reactor primary coolant shall be '
verified to be negative above 124*F whenever changes made to the reactor could affect the temperature coefficient.
l 4.1.4 Bases
~
Operation in accordance with Specification 4.1.3.1 will ensure that the reactor is not operated with a potential excess reactivity of > .76$.
Operation in accordcnce with Specification 4.1.3.2 will ensure that the reactor will be subcritical when all the safety rods are in the full-out position and the control rods are inserted.
' Minimum shutdown margin is assured when the potential excess reactivity is limited to 76C and safety rod reactivity worths are unchanged. The shutdown margin, then, should be determined as specified in Specification 4.1.3.3 when changes to the reactor are made which could decrease the reactivity worth of a safety rod.
Verification that the manual poison sheets are properly restrained as specified in Specification 4.1.3.4 ensures that they cannot be ejected during any postulated natural phenomena or operational occurrence.
Compliance with specification 4.1.3.5 ensures that the temperature coefficient is negative above 124*F. It is not affected by reactor configuration and fuel burnup ar.d is therefore not expected to vary significantly with core life (bcr c'ould be affected by fuel, core or moderator design changes).
i l'
I 26
.-, ---,-m. ,+ - - - , , , . , . - . ,- .----,.---n.,w-,,-
-,,-,,.y- - - - -me ,-- 7
~
- 4.2 ' REACTOR CONTROL AND SAFETY SYSTEM' 4.2.1 Applicability '
This specification applies to the surveillance requirements for the reactor control'and safety system.
4.2.2 Objective
- The objective of this specification is to specify the minimum surveillance re-
..quirements to reasonably ensure proper performance of the safety rod, control rod and safety. systems.
14.2.3 Specifications
- 4.2.3.1
_ Each safety rod and control rod drive shall be tested for operability annually.
4.2.3.2 s
'Theinterlockwhichrestrictssafetyrodwithdrawaltoonerodatatimeshallbe
. --tested annually.
it 7
p 4.2.3.3 tj_ g M #-
The rate of withdrawal of each safety rod shall be measured annually.
4.2.3.4
'the rate of withdrawal of each control rod shall be measured annually.
27
~~'
l l
- 4. 2.'3. 5 ,
The safety rod scram time (inflight time) shall be measured semi-annually. The !
scram time (inflight time) shall additionally be measured after any work is performed which could affect the scram time or rod travel time.
4.2.3.6
. _ Checks, tests and calibrations of the reactor safety system shall be performed as
-specified in Tables 4 I and 4-2.
4.2.3.7
,~
A thermal power verification shall be performed monthly when the reactor is
. operating above 50 kW.
Table 4-1
- Suiveillance' Requirements of Reactor Safety System Scram Instruments Item. '
No. Item Surveillance Frequency *
' Linear System 1.
~
~[Nk0d Channel Check (neutron source check) Daily "M' , qo. p Channel Test (high level trip test Daily
^^
Channel check (comparison against a' heat balance) Semi-annual Channel Calibration Annually
- 2. Log;N; System Channel Test Daily Channel Check Monthly Channel Calibration Annually
'3..
Primary Coo)*nt Channel Test Daily Temperattze Channel Calibration Annually
. 4.. . Primary Coolant Flow Channel Check Daily Channel Test Daily g3 Channel Calibration Annually
- 5. Manual' Channel Test Daily
- 6. Electrical Power Channel Test Daily
- 7. . Seismic Channel Test Daily 28
__ ,. .. , - , , , - - - - ---------e-
- Prior to placing into service an instrument whi ch has been repaired, the instrument check, or test or calibration, as appropriate will be performed.
e 0
29 ,
(
. NEDD-12725A A
Table 4-2 Surveillance Requirements of Reactor Safety System y
Information Instruments Item
'No. Item Surveillance Frequency
- i
- 1. Reactor Cell Pressure Channel Test Quarterly
- 2. Fuel Loading Tank Water Level Channel Test Quarterly
- 3. Primary Coolant Temperature Channel Test Quarterly Channel Calibration Annually
- 4. Primary Coolant Conductivity Channel Check Quarterly Channel Calibration Annually 5 .' . Primary Coolant Core Channel Check Monthly
^
A Temperature Channel Calibration Annually
- 6. . Reactor Cell Radiation Monitor Channel Check Daily y Channel Test Monthly Channel Calibration Annually y 7. Stack Radioactivity Channel Check Daily (Gas sad particulate channels) Channel Test Monthly Channel Calibration Annually 8.- Linear Power Channel Test Monthly
- 9. Log Power Channel Test Quarterly
- Prior to placing into service an instrument which has been repaired, the instrument check, test or calibration, as appropriate, shall be performed, w
- e
a NEDO-12725A 4.2.4 Bases Specification 4.2.3.1 ensures that each safety and control rod are maintained
. operable.
-Specification 4.2.3.2 ensures that the safety rod interlock preventing the simul-teneous withdrawal of acre than one safety rod functions properly.
' Specifications 4.2.3.3 and 4.2.3.4 ensure that the control and safety rod with-drawal rates are within limits.
Specification 4.2.3.5 provides for the periodic measurement of safety rod in-
-sortionLtimes to ensure they are within limits.
Specification 4.2.3.6 ansures that the safety system is periodically tested and checked to maintain all instruments operable.
~
. ;. 4.3 REACTOR COOLANT SYSTEM Specifications regarding surveillance requirements of the reactor coolant system are included in the reactor safety system, Specification 4.2, Tables 4-1 and 4-2.
4.4 REACTOR CELL AND VENTILATION SYSTEM 4.4.1 Applicability This specification applies to the surveillance requirements for the reactor cell and ventilation system.
4.4.2: Objective
, . The objective of this specification is to ensure that the reactor ventilation system is in satisfactory condition to provide adequate. confinement and to control the release of radioactivity to the environment.
31
'4.4.3 Specification 4.4.3.1 The reactor cell negative pressure, with respect to the contrcl room, shall be verified prior to the first reactor startup of each day.
4.4.3.2 Surveillance requirements of the instrumentation and equipment required to comply with Specifications 3.4.3.2, 3.4.3.3 and 3.4.3.5 shall be as listed in Specifica-tion 4.2, Table 4-2, 4.4.4 Bases
', Operation in accordance with Specification 4.4.3.1 ensures that contaminated reactor cell; air is exhausted through the ventilation system._ This minimizes the possibility of airborne contamination release to surrounding areas.
Operation in accordance with' Specification 4.4.3.2 ensures that all required channels are operational and that proper notification and surveillance will occur.
4 4.5 EXPERIMENTS ~
Specific surveillance activities shall be established during the review and x
approval process as'specified in Section 6.2.3 " Review Function" and are not part
- of the Technical Specifications.
- 4.6~ FREQUENCY OF TESTING 4.6.1- Applicability
' This specification applies to all surveillance requirements of Section 4 of these
- Technical Specifications.
5 .
32
' ^
.m. . . . - _,,, __.,__,_m., _ ._ , - . . , _ . , .m,,__,_.--,-,___,.-.
- ,x l
NED0-12725A i
,s l
l
'4.6.2 Objective The objective of this specification is to establish maximum time intervals for surveillance periods. It is intended that this specification provides operational iflexibility and no't to reduce surveillance frequency.
4.6.3- , Specifications
.4'.'6.3.1 Time' intervals used elsewhere in these specifications shall be defined as follows*
- a. Biennially - Interval not to exceed 30 months.
- b. . Annually - Interval not to exceed 15 months.
- c. Semi-annual - Interval not to exceed 32 weeks
- d. Quarterly - Interval not to exceed 18 weeks
- e. Monthly - Interval not to exceed 6 weeks
- f. Weekly - Interval not to exceed 10 days
- g. Daily - Must be done prior to the first startup of_the calendar day following a shut down greater than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
33
NED0-12725A
'4'.6.3.2 Surveillance tests (except those required for safety while the reactor is shut down) may be deferred during a reactor shutdown. Deferred surveillance tests must be completed prior to reactor startup.
.4.6.3.3 Surveillance tests scheduled to occur during reactor continued operation which cannot be performed with the reactor operating may be deferred until the subsequent scheduled reactor shutdown.
4.6.4 Bases' Specification 4.6.3.1 establishes maximum time intervals for surveillance require-ments which define the terms and makes them objectively quantifiable.
Specification 4.6.3.2 permits deferring tests which are not required if the reactor will not be operating.
l Specification 4.6.3.3 permits deferring tests which might require a reactor shutdown for the sole purpose of performing the test.
., w 1
~ ~
T I
34
- ,ma e- ,- , ,w -- -w<-- - , . - - -- - - - - -
NEDO-12725A 5.0 DESIGN FL'TURES 5.1 SITE AND FACILITY DESCRIPTION 5.1.1 I The' Nuclear Test Reactor (NYR) facility shall be located on the site of the
-Vallecites Nuclear Center (VNC) which is owned and controlled by the General Electric Company.
5.1.2 The minimum distance from the reactor to the posted site boundary shall be approx-instely 1600 feet. The restricted area, as defined in 10CFR20 of the Commission's regulations, shall be the Vallecitos Nuclear Center.
5.1.3
- The fuel assemblies shall be positioned in a'rcel assembly irside the core tank.
The core reel assembly shall be rotated only when the reactor is shut down and by manual operation of a crank inside the NTR cell.
~ '
f5.1.4 The control system shall consist of four scrammnile, spring-actuated safety rods, three nonscrammable control rods, and a number of manual poison sheets. When the poison rods and sheets are inserted, they shall be located in the graphite reflec-tor at the outer periphery of the core tank. The safety and control rods shall be boron carbide clad in stainless steel. The manual poison sheets shall contain
. metallic cadmium.
5.1.5 The discharge of the gaseous effluent stack shall be approximately 45 feet above grade level of Building 105.
Y .
35 e
' l gr *- e ++- -vw- = .-w_- .c--. e--. - . ----,----r--- r,- .-- - ---- - , we- --+-e--%--_ - - -- -v--------w----- - - - -r-- ---
NEDO-12725A 5.2 REACTOR PRIMARY COOLANT. SYSTEM 5.2.1 Above 0.1 kW the reactor shall be cooled by light water forced coolant flow mosinally 18 spm. Below 0.1 kW forced coolant flow is not required.
5.2.2
-The reactor coolant system shall be protected from overpressure by a vent line to the atmosphere of the cell.
-5.3 -REACTOR CORE AND FUEL
-,g The. core shall consist of 16 fuel element assemblies. Each fuel element assembly
,shall consist of 40 disks spaced on an aluminum support shaft. Other nominal specifications of the assemblies shall include the following:
- a. Fuel 23.5% by weight uranium - 76.5% by weight aluminum.
- b. Enrichment approximately 93% U-235 (unburned)
- c. Cladding Aluminum, 0.022 in. thickness.
- d. Fuel disk' active diameter 2.685 in.
- e. Fuel disk spacing'on shaft 0.35 to 0.45 in., center-to-center
-5.4 FISSIONABLE MATERIAL STORAGE o
= Fuel including fueled experiments and fuel devices not in the reactor shall be stored in a geometrical array where k,gf is'no greater than 0.9 for all conditions
.of moderation and reflection using light water.
36 L
. 4.
NEDO-12725A 6.0 ADHlNISTRATIVE CONTROLS 6.1 : ORGANIZATION AND STAFFING 6.1.1 S'tructure The NTR shall be owned and operated by the General Electric Company with manage-ment and cperations organization as shown in Figure 6-1 or equivalent.
6.1.2 Responsibilities 6.1.2.1 The section-level manager shall be responsible for the NTR facility license.
6.1.2.2 i
s
.- The-subsection-level' manager (Operations) is designated the facility manager and
,- .shall be responsible for the overall safe operation and maintenance of the facili-ty.
6.1.2.3 m.
The unit-level manager is responsible for the routine safe operation and mainte-nance of the facility in accordance with the license, regulations and established
' written procedures.
6.1.2.4 The Reactor Supervisor (if utilized) is the individual responsible for supervising the' daily oparations. In the absence of this position, the unit-level manager is responsible for supervising the daily operations.
k 37
- -., n ~ - .,,,-
l l
GENERAL ELECTRIC COMPANY UPPER MANAGEMENT LEVEL SECTION LEVEL (MANAGER) r- - - - - - - -
y . _ _ . _ . . . _
l ' Nb A E NO l SUBSECTION LEVEL g VALLECITOS l QUA!.lTY ASSURANCE l (FACILITY MAN AGER) TECHNOLOGICAL L--- --.J L _ SAFETY __ _ COUNCIL l
r -- -- - - - - -- -- 7 r ,
l INTERN AL REVIEW ENGINEERING AND ! ! AND
! AVOIT FUNCTION l MAINTENANCE l l L _ .J LJ UNIT LEVEL q
NUCLEAR SAFETY l UNIT LEVEL FUNCTION l (MANAGERI
_ _ -. . ._J p - _ . _ __ -
F ---
NUCLEAR SAFETY . ERING
-- 7
~
l REACTOR g FUNCTION }
l SUPFRVISOR g MAINTENANCE !
t.___.._._. u _. __ __1 NTR OPERATORS NOTE: DASHED LINES REPRESENT ALTERNATE /OFTIONAL ORGANIZATION FUNCTIONS Figure 6.1. Facility Organization 38 g - .
- 6.1.2.5 Responsibilities of one level may be assumed by alternates when designated in writing.
6.1.2.6-Functions performed by one level may be performed by a higher level, provided the minimum qualifications are met (e.g., Senior Reactor Operator's license).
6.1.3 Staffina 6.1.3.1 The minimum staffing when the reactor is not secured shall be composed of:
- a. A licensed operator in the control room.
- b. A second person present at the site familiar with NTR Emergency Proce-dures and capable of carrying out facility written procedures.
- c. A licensed Senior Reactor Operator shall be present at the NTR Facility or readily available on call.
6.1.3.2 A licensed Senior Reactor Operator shall be present at the NTR Facility during the following events:
- a. During the recovery from an unscheduled shutdown. *
- b. During reactor fuel loading or reactor fuel movement.
- c. During any experiment or facility changes with a reactivity worth greater than one dollar.
39
i' NFDO-12725A 6.1.4 Selection and Training of Personnel i The selection, training and requalification of operations personnel shall meet or exceed the requirements of American National Standard for Selection and Training of Personnel for Research Reactors, ANSI /ANS 15.4-1977 Sections 4 through 6; Title 10 of the Code of Federal Regulations, Part SS, Appendix A; and the latest
, revision of the Facility Operator Requalification program.
9 G
40
m-
.g 77 NEDO-12725A
, 6.2 INDEPENDENT REVIEW 6.2.1 Independent review is performed by technically qualified individuals, responsible through optional intervening management, to the section level manager.
6.2.2 The independent review function shall be performed under a written charter or directive containing the following information as a minimum.:
- a. Subjects reviewed
- b. Responsibilities c.- Authorities
- d. Records
- e. Other matters as may be appropriate 6.2.3 Activities requiring independent review shall include the following:
- a. Proposed types of tests and experiments (or substantive changes thereto) including safety evaluations, that could affect core reactivity or result in an uncontrolled release of radioactivity, to be conducted without prior NRC approval, pursuant to 10CFR50.59, to verify the proposed activity does not constitute a change in the Technical Speci-fications or an unreviewed safety question.
- b. Proposed changes to the procedures or the facility, as described in the Safety Analysis Report, including safety evaluations, to be completed without prior NRC approval, pursuant to 10CFR50.59, to verify the activity does not consititute a change in the Technical Specifications or an unreviewed safety question.
- c. All new procedures and revisions thereto having safety significance required by the specifications in Section 6.3.
41 L
~
r NEDO-12725A d.~
Proposed changes to the Technical Specifications or the facility operat-ing license.
- e. . Violations of the Federal Regulations, Technical Specifications, and facility license requirements.
- f. Unusual or abnormal occurrences which are reportable to the NRC under provisions of the Federal Regulations or the Specifications in Section 6.6.
- g. Significant operating abnormalities or deviations from normal and expected performance of facility equipment that affect, or could affect, nuclear safety.
6.3 PROCEDURES 1*
6.3.1
' Written procedures shall be prepared for the following activities as required:
- a. Startup, operation, and shutdown of the reactor.
.b. Defueling, refueling, and fuel transfer operations when required.
c.
Preventive or corrective maintenance which could have an effect on the safety of the reactor.
- d. Off-normal conditions relative to reactor safety for which an alarm is received.
- e. Response to abnormal reactivity changes, f.
Surveillance testing, and calibrations required by the Technical Speci-fications.
3 Emergency conditions involving potential or actual release of radioac-tive meterials.
- h. Radiation protection consistent with 10CFR20 requirements.
- 1. Review and approval of changes to all required procedures.
j.
Security plan, the operator requalification program, and emergency procedures.
3 42
p- .=-
- k. : Operation and maintenance of experiments that could affect reactor safety or core reactivity.
>6.3.2
'The facility manager shall approve all procedures (including revisions) required by specification 6.3.1 before implementation.
6.3.3
' Minor changes to the original procedures which do not change their original intent
.may be.made by the Reactor Supervisor or unit level manager. These changes must
.'be subsequently approved by the facility manager.
16.3.4 Temporary. deviations from established procedures may be made by a Licensed Senior Reactor Operator in order to deal with special or unusual circumstances. These
- deviations shall be documented and reported to the facility manager.
-6.4 -REQUIRED ACTIONS Action to be taken in the event of an occurrence of the type identified in Section 6.5.2.
i
-6.4.1 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 Manager.
jf7.j' '6.4.2. Occurrence shall be reported to the Facility Manager and to the NRC as
. required.
- 614.3 Occurrence shall be reviewed by the Nuclear Safety group.
S 43
NEDo-12725A 6.5 REPORTS 6.5.1 operatina Reports Annual operating report (s) shall be submitted to the NRC. The report (s) shall include the following:
- s. A narrative summary of reactor operating experience including the hours
'the reactor was critical.
- b. The unscheduled shutdowns including, where applicable, corrective action taken to preclude recurrence
- c. Tabulation of major changes in the reactor facility and procedures, and
, tabulation of new tests and experiments, that are significantly differ-ent from those performed previous sud are not described in the Safety
, Analysis Report, including provisions that no unreviewed safety questions were involved.
- d. A summary of the nature and amount of radioactive effluents released or
' discharged to environs beyond the effective control of the own-er-operator as determined at or before the point of such release or discharge.
- e. ' A summarized result of environmental surveys performed outside the facility.
- f. A summary of exposures received by facility personnel and visitors where such exposures are greater than 25% of that allowed or recommended.
. 44 .
IV '
1 l
6.5.2 Special Reports Special reports are used to report unplanned events as well as planned major facility and administrative changes. The following special reports shall be made as appropriate:
- a. There shall be a report not later than the following working day by telephone and confirmed in writing by telegraph or similar conveyance to the NRC to be followed by a written report that describes the circum-stances of the event within 14 days of any of the following:
(1) Release of radioactivity from the site above allowed limits. (See 6.4.2)
(2) Any of the following: (See 6.4.2)
- Operation with actual safety-system settings for required systems less conservative than the limiting safety-system settings specified in the technical specifications.
Operation in violation of limiting conditions for operation established in the technical specifications unless prompt remedial action is taken.
A reactor safety system component malfunction which renders or could render the reactor safety system incapable of performing its intended safety function unless the malfunction or condition is discovered during maintenance tests or periods of reactor shutdowns. (Note:
Where components or systems are provided in addition to those required by the technical specifications, the failure of the extra
~
components or systems is not considered reportable provided that the minimum number of components or systems specified or required perform their intended reactor safety function.)
An unanticipated or uncontrolled change in reactivity greater than one dollar.
45 1
e NEDO-12725A Abnormal and significant degradation in reactor fuel, cladding, or coolant boundary, which could result in exceeding prescribed radiation limits for personnel or the environment.
' An observed inadequacy in the implementation of administrative or procedural controls such that the inadequacy causes or could have caused the existence or development of an unsafe condition with regard to reactor operations.
! b. _ A written report within 30 days to the NRC for the following:
(1) Permanent changes in the facility organization involving Section or Subsection management.
(2)- Significant changes in the transient or accident analysis as described in the Safety Analysis Report.
c' j'.
I
- 46 k: '"
p 45 7 -
y.-
\ \;f cm , ,
., n. ') *
.h,9 _-
1, o NEDO-12725A i
- . s H .i. , .
. . . . . 6.6 RECORDS
- es . 4 s nr
-Records may be in the form of' logs, data sheets, or other suitable forms. The required informatibn may be contained in single, or multiple records, or a com-bination thereof.
,6 6.6.1 ;, ,
s Records to be Retained for n Period of at Least Five Years or for the Life of the Component whichevar is least-a.. Moraal reactor facility operation (supporting documents such as check-lists, los sheets,- etc., shall be maintained for a period of at least
,' , ene year).,
- b. Pr.incipal; maintenance operations.
~ c. Reportable. occurrences.
,, .,_~-- %
' ' d. Sorve111ance activitieu required by the Technical Specifications.
),.
- e. / Reactor f acility radiation and contamination surveys where required by
'. ,; siplicable regulations.
(! P
't
- f. Experime;nts. performed with the resctor.
( -
g ., Tuel inventories.f receipts, and shipments.
~
h.. _
Approved changes in operating procedures.
t . ,
,. 3
- 1. etca rds of seating reports of the review group.
J 9 g
?
.o 4
+-
t i
. 4 i -
47
- e-e
-t NZDO-12725A 6.6.2 4
Records of the requalification program shall be maintained for a period of two years from the date of the recorded event.
6.6.3 Records to be Retained for the Lifetime of the Reactor Facility (Note: Applicable annual reports, if they contain all of the required informa-tion, may F* used as records in this section.)
- a. Caseous and liquid radioactive effluents released to the environs.
- b. Radiation exposure for all personnel monitored.
, c. Drawings of the reactor facility.
a 48
._