ML20071G387
| ML20071G387 | |
| Person / Time | |
|---|---|
| Site: | Berkeley Research Reactor |
| Issue date: | 01/18/1979 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | Pigford T CALIFORNIA, UNIV. OF, BERKELEY, CA |
| Shared Package | |
| ML20071G392 | List: |
| References | |
| NUDOCS 7902060007 | |
| Download: ML20071G387 (61) | |
Text
x 6124 g eercy
,P UNITED STATES 3
[.
, y 's NUCLsAR REGULATORY CCMisslON
]
(.]
j WASHINGTCN, O. C. 2C555 January 18, 1979 v.[
Occket No.
50-224 Mr. Thcmas H. Pigford Reactor Administrator Department of Nuclear Engineering The Regents of the University of California Serkeley, California 94720
/
(
l Cear Mr. Pigford:
a We have completed a thorough review of your license renewal application and find that additional information is required.
These items were discussed in detail in your conversation with Mr. Ramos on Wednesday, January 3,1979, and are reiterated herein.
We informed you by letter on October 17, 1978, that the physical security plan (PSP) for your reactor facility had been reviewed and that although our evaluation showed it was acceptable, requested that it be reconciled and resubmitted in a single document in loose-leaf fo rma t.
In your discussions with Mr. Ramos, we were to send you a copy of proposed Regulatory Guide 5.XX, " Standard Format and Content for the License RSP..." Because of some changes being made to the Regulatory Guide, we are deferring sending it to you at this time.
In addition, pursuant to 10 CFR 50.54, we plan to include your PSP in the license conditions by reference.
Because of the sensitive nature ls) of the security plan, the actual PSP will not be attached to the license
'd and will be withneld from public disclosure in accordance wittt 10 CFR 2.790(d).
The folicwing is an example of such a license condi; tion:
n
,n "The licensee shall maintain in effect and fully implement;all.
1 provisions of the NRC Staff-approved physical security: plan, t including amendments and changes made pursuant to the (uthority of 10 CFR 50.54(p). The approved security plan consists 'of '
documents withheld from public disclosure pur uant to 10,CFR..
2.790, collectively titled, " University of California at Berkeley Security Plan," as follows:
Original, submitted with letter dated May 31, 1973 Revision 1, submitted with letter dated November 26, 1973 Revision 2, submitted with letter dated January 14, 1974 Revision 3, submitted with letter dated March 11, 1974" This, of course, is only an examcle and does not reflect your actual PSP.
7902060007 4)
(
.\\
~
/
q i
Thomas H. Pigford -
The second item discussed relates to the Department of Energy and State Department program to implement the Nonproliferation Act of March 10, 1978, by reducing the enrichment of fuels in nonpower reactors. Concomitant to this, the proposed Regulation 8 73.17 is designed to implement the US/IAEA Agreement when approved by the Senate. Both of these actions are keyed to the enrichment of fuel and other SNM. Therefore, your license, which authorizes certain maximum possession limits of SNM (U235, Pu, U233), should be changed to reflect not only the total amount of SNM, but the percent enrich-ment of each; the amount of SNM exempt and how exempt (i.e.,10 CFR 73.6(b)); and the amount of SNM non exempt. This will fonn the basis for establishing the level of protection of your PSP.
O For your information, in September 1975, a letter was sent to all licensees authorized to possess SNM in excess of 10 CFR 73.l(b) quanti-ties requesting that they review their requirements and provide justi-cation for the " lowest acceptable quantity" nece!.sary to sustain current operations and those projected for the ensuing twelve months.
There are still a number of licensees that are authorized to possess quantities in excess of 73.l(b) quantities.
In view of the foregoing, you are requested to review your SNM require-ments and provide:
1.
The maximum amounts of SNM and types.
2.
The enrichment of each item in 1.
3.
The amounts of each SNM exempt and how exempt.
4.
The amounts of each SNM non exempt, to be included in your license.
O to re' tere te o"e 'te= or o"r d'=c"=='o" re' t'"9 to exe at e"d "o" exempt SNM, this definition in Part 73 refers to whether or not the SNM is exempt or not exempt frem meeting certain regulations in Part 73.
These definitions do not apply to possession limits.
The third item discussed was your Technical Specifications (TS). Our review reveals that your TS's should be upgraded to current standards.
As agreed, we are providing a copy of theTexas A&M/ Washington State University TS for you to use as a guide in revising yodrs.
The final item discussed was the need to update your safety hazards report. This will not hold up the license renewal, but will be an item left open. It was agreed that this updating would be due following comple-tion of the license renewal.
3
,w,,-
w-3c 9-,
u.-w
~ -.
--*_---e
,- - - - y
9
+,
Thomas H. Pigford.
As we agreed, you will submit the Technical Specifications as the first priority item and the PSP second.
It is requested that tne Technical Specifications and your S!;M limits be submittec within 60 days, if possible.
Please do not hesitate to contact Steve Ramos (301 J92-7435) regarding these matters.
Sincerely, j
e 1^0
- }n adr O.A,/
Robert W. Reid, Chief
,]'
Operating Reactors Branch #4 Division of Operating Reactors
Enclosures:
1.
Sample Technical Specifications 2.
Guidance on Administrative Controls 3.
Draft ANS 15.18 Standard for Administrative Controls e
8 i
i
't
'l APPENDIX A CHANGE NO. 11 TO THE TECHNICAL SPECIFICATIONS FACILITT LICE SE NO. R-83 i
TOR THE NUCLEAR SCIENCE CETIER REACTOR i
0F TEXAS A&M UNIVERSI!!
DOCKET No. 50-IIS O
3ese Technical Specifications have been modified to incorporate guidance from NRC Regulatory Guide 1.16 and salient features of Washington State University Technical Specifications for their TRIGA Docket No. 50-27, License No. R-76, Also attached is Guidance for Section 6.0 Administrati'te Controls which differs from Section 6.0 herein.
Transmitted v/Amendnent No. 4, dated 6/26/73 e
1 TABLE OF CONTENTS P,, age, 1
I 1.0 Definitions 1
1.1 Reactor Shetdown 1
1.2 Reactor Secured 1
1.3 Reactor Operation 1
1.4 Cold Critical 1
1.5 Steady State Mode 1
1.6 Pulse Mode 2
1.7 Shutdown Margin 1.8 Abnormal Occurrence 2
I
( )
1.8a Reportable Occurrence 2
2 1.9 Experiment 1.10 Experimental Facilities 3
1.11 Shia-Safety Rod 3
1.12 Transient Rod 3
1.13 Regulating Rod 3
1.14 Fuel Element 3
1.15 Fuel Bundle 3
1.16 Core Lattics Position 3
1 1.17 Instrumented Elemenc 4
1.18 Standard Core 4
1.19 Mixed Core 4
4 1.20 Flip Core 4
1.21 Operational Core 4
1.22 Safety Limit 1.23 Limiting Safety System Setting 4
1.24 operable 5
1.25 Reactor Safety Systems 5
O 1.26 Experiment Safety Systems 5
1.27 Measured Value 5
1.28 Measuring Channel 5
1.29 Safety Channel 5
1.30 Channel check 5
1.31 Channel Test 5
1.32 Channel Calibration 6
2.0 Limitine Safety S? stem Settina 6
2.1 Safety Limit-Tuel Element Temperature 6
2.2 Limiting Safety System Setting 7
1 3.0 Limiting Conditions for Ceeration 8
3.1 Reactivity Limitations 8
8a 3.2 Fulse Mode Operation 3.3 Control and Safety System 9
3.4 Radiation Monitoring System 9a 1C 3.5 Engineered Safety Feature - Ventilation System 13 3.6 Limitations on Experiments l
l
- _. ~ _,
11
- 2. age 3.7 Argen-41 Discharge Limit 14 3.3 Engineered Safety Feature - 7entilation System 14 3.9 Lisitations on Experiments 15 3.10 Irradiations 17 4.0 Surveillance Requirements 17a 4.1 General' 17a 4.2 Safety Limit - Fuel Element Temperature 18 4.3 Limiting Conditions for Operation 18 4.4 Reactor Fuel Elements 22 i
5.0 Design Features 23 5.1 Reactor Fuel 23 5.2 Reactor Core 24 5.3 Control Rods 26 5.4 Radiation Monitoring System 27 28 5.5 Fuel Storage 5.6 Reactor Building and Ventilation System 28 5.7 Reactor Pool Water Systems 29 6.0 Administrative Controls 31 6.1 Organization 31 6.2 Review and Audie 32 6.3 Action to be Taken in the Event a Safety Limit 33 is Exceeded 6.4 Action to be Taken in the Event of an Abnormal 33 Occurrence 6.5 Cperating Procedures 34 6.6 Facility Operating Records 34 6.7 Reporting Requirements 35 Y
l l
l l
L
.s Included in this document are the Technical Specifications and the " Bases" for the Technical Specifications. These bases, which provide the technical support for the individual technical specifications, are included for infor-mation purposes only. They are not part of the Technical Specifications, and they do not constitute limitations or requirements to which the licensee must adhere.
Reference NRC Regulatory Guide 1.16 and ANSI N378-1974.
1.0 DEFINITIONS REACTOR OPERATING CONDITIONS 1.1 REAC"'OR SHU DOWN The reactor is shut down when the reactor is subcritical by at least one dollar of reactivity.
i O 1.2 REACTOR SECURED The reactor is secured when all the following conditions are satisfied:
a.
The reactor is shut down, b.
The console key switch is in the "off" position and the key is removed from the console and under the control of a licensed operator or stored in a locked storage area, and c.
No work is in progress involving in-core fuel handling or refueling operations, maintenance of the reactor or its control mechanisms, or insertion or withdrawal of in-core experiments.
I 1.3 REACTOR OPERATICN Reactor operation is any condition wherein the reactor is not secured.
1.4 COLD CRI"'! CAL The reactor is in the cold critical condition when it is critical with the fuel and bulk water temperatures both below 40 C.
0 1
l 1.5 STEADY STATE MCDE Steady state mode operation shall =ean operation of the reactor with the mode selector switch in the steady-state position.
l.6 PULSE MCDE Pulse mode operation shall sean any operation of the reactor with the mode selector switch in the pulse position, i,
6 4
4
,y_.
,,.,,_.,_,,,,,y,_
s I
1.7 SHUT:XNN MARGIN Shutdown margin shall sean the sisimum shutdown reactivity necessary to provide confidence that the reactor can be sade suberitical by means of the control and safety syscens, starting from any permissible operacing condicions and that the reactor will remain suberitical without further operator action.
1.3 A3 NOM'E CCCURRENCE An " Abnormal Occurrence" is defined for the purposes of the i
reporting requirements of Section 208 of the Energy Reorganization Act of 1974 (p.L.93-438) as an unscheduled incident or event which the Nuclear Regulatory Commission determines is signifi-cant from the standpoint of public heal:h or safety.
1.3a RE?CRTABLE CCCURRENCE A reportable occurrence is any of the following which occurs during reactor operation:
1
.I Operation with any safety system setting less conservative a.
than specified in Section 2.2, Limiting Safety System Settings:
b.
Operation in violation of a Limiting Condi:1on for Operacion; c.
Failure of a required reactor or experiment safety system component which could render the system incapable of per-forming its intended safety fune:1on; d.
Any usanticipated or uncontrolled change in reactivi:y greater than one dollar; O
e.
An observed inadequacy in the implementation of either administrative or procedural controls, such that the inadequacy could have caused the existence or development of a con-dition which eculd resul: in operacion of the reac:er outside the specified safety lisi:s; and f.
Release of fission products from a fuel element.
l REACTCR EXPERIMENTS l
l 1.9 EXPERIMENT Experiment shall mean (a) any apparatus, device, or sacerial l
which is not a normal part of :he core or experimen:21 facili:ies, but which is inserted in these facilities or is in line with a beam of radiation originating from the reactor core; or (b) any operation designed to measure reactor para =ecers or characteristics.
I i
l e
l i
1.10 E:C'ERIMENTAI. FACII.ITIES 1
1 j
Experimental facilities shall mean beam ports, including t
extensica tubes with shields, thermal columns with shields, i
vertical tubes, through tubes, in-core irradiation baskets, i
irradiation cell, pneumatic transfer systems and in-pool 4
irradiation facilities.
i REACTCR CCMPCNENTS 2
I 1.11 S' DIM-SAFETT ROD
~
4 A shim-safety tod is a control rod having an electric motor j
drive and scram capabilities. It may have a fueled follower j
section.
O 1 12
- x^=stzst noo I
j j
The transient rod is a control rod with scram capabilities that i
can be rapidly ejected from the reactor core to produce a pulse.
i It may have a voided follower.
1 1
1.13 REGUI.ATING RCD
~
The regulating rod is a low worth control rod that need not l
have scram capability and say have a fueled folicwer. Its 1
position may be varied manually or by the servo-cancro11er.
i I
i 1.14 FUEI. ELW_
A fuel element is a single TRICA fu1tl rod of either standard j
or FI.I? cype.
1
]
1.15 FUEL. BUNDLE 1
1 A fuel bundle is a cluster of three or four fuel elements secured 2
in a square array by a top handle and a bottom grid place i
adaptor.
t 1.16 CORE I.ACICE pCSITION 1
The core lattice position is that region in the core (approxi-mately 3" x 3") over a grid plug hole. It may be occupied by a fuel bundle, an experiment, or a reflector element.
1 4
3 1
1 i l
4 x...,
i j-4 e
i 1
l 1.17 INSTRUMETIED EL
.EN"_
An instrumented element is a special fuel element in which a sheathed chromel-alumel or equivalent thermocouple is J
embedded in the fuel near the horizontal center plane of the i
fuel element at a point approximately 0.3 inch from the center l
of the fuel body.
1.18 STANDARD CCRI I
A standard core is an arrangement of standard TRIGA fuel in the reactor grid place.
~
I-1.19 MIXED CCRE A mixed core is an arrangement of standard TRIGA fuel elements with at least 23 TRICA-FLI? fuel elements located in a central j
region of the core.
1.20 FLIP CORE I
A FLI? core is an arrangement of TRIGA-FLI? fuel in the reactor grid plate.
1.21 CPERATIONAL CORE An operational core may be a standard core, mixed core, or FLIP core for which the core parraeters of shutdown margin, fuel temperature, power calibration, and saximum allowable reactivity insertion have been deceru.ined to satisfy the requirements of the Technical Specifiestions.
REACTOR INSTRUMENTATICN 1.22 SAFETT LIMIT 1
Safety limits are limits on important process variables which are found to be necessary to reasonably protect the integrity of certain of the physical barriers which guard against the uncontrolled release of radioactivity.
~
1.23 LIMITING SAFETT SYSTEM SETTING Limiting safety systems setting is settina for automatic protective devices related to those variables having signi-ficant safety functions.
I*
- ,i.,
I r
i 1.24 OPERA 3LZ A system, device, or component shall be considered operable when it is capable of performing its intended functions in a normal unner.
1.25 REACTOR SAFETT SYSTEMS r
Ratscor safety systems are chose systems, including their associated input circuits, which are designed to initiate a reactor scram for the primary purpose of protecting the reactor i
or to provide information which requires manual protective action to be initiated.
7 1.26 EXPERIMENT SAFETT SYSTEMS t
Experiment safety systems are chose systems, including their associated input circuits, which are designed to initiate a scram for the primary purpose of protecting an experiment or
[
l to provide information which requires manual protective action I
to be initiated.
1.27 MEASURED VALUE r
The measured value is the magnitude of that variable as it appears on the output of a measuring channel.
t L
1.28 MEASURING CHANNEL
[
A measuring channel is the combination of sensor, interconnecting cables or lines, amplifiers, and output device which are connected for the purpose of measuring the value of a variable.
1.29 SAFETT CHMINEL O
A safety channel is a seasuring channel in the reactor safety system.
F i-1.30 CHANNEL CHECX A channel check is a qualitative verificacion of acceptable performance by observation of channel behavior.
1.31 CHANNEL TEST A channel test is the increduction of a signal into the channel to verify that it is operable.
i
--~w m er w-y v
y p-4 4
1.32 CHANNEL CALI3 RATION A channel calibration consists of comparing a measured value from the measuring channel with a correspending known value of the parameter so that the measuring channel output can be adjusted to respond with acceptable accuracy to known values of the measured variable.
2.0 SAFETT LIMITS AND LIMITING SAFITY SYSTEM SETTINGS 2.1 SAFETY LIMIT-FUEL EL."J.NT TEMPER.WJRE Applicability This specification applies to the temperature of the reactor fuel.
O es4eceive The objective is to define the maximum fuel elwent temperature that can be permitted with confidence that no damage to the fuel element cladding vill result.
Specifications a.
The temperature in a TRICA-FLIP fuel element shall not exceed 2100*F (1150*C) under any conditions of cperation.
5.
The temperature in a standard TRIGA fuel element shall not exceed 1330*F (1000*C) under any conditions of operation.
Bases The important parameter for a TRICA reactor is the fuel element temperature. This parameter is well suited as a single specifi-cation especially since it can be measured. A loss in the integrity of the fuel element cladding could arise frem a build up of excessive pressure between the fuel-moderator and the cladding if the fuel temperature exceeds the safety limit. The pressure is caused by the presence of air, fission pr: duct gases, and hydrogen frem the dissociation of the hydrogen and =ircenium in the fuel-moderator. The magnitude of this pressure is deter-mined by the fuel-moderator temperature and the ratio of hydrogen to zirconium in the alloy.
The safety limit for the TRICA-FLI? fuel element is based on data which indicate that the stress in the cladding due co the t
I.
e
,_.r--,
.i,
I I
l hydrogen pressure from the dissociation of circonium hydride will remain below the ultimate stress provided the temperaturts of the fuel dcas not exceed 2100*? (1150*C) and the fuel cladding is water cooled.
(SAR II, pg. 4)*
The safety limit for the standard TRICA fuel is based on data, including the large mass of experimental evidence obtained during high performance reactor tests on this fuel. These data indicate that the stress in the cladding due to hydrogen pres-sure from the dissociation of circonium hydride vill remain below the utlimate stress provided that the temperature of the fuel does not exceed 1830*? (IC00*C) and the fuel cladding is water cooled.
(SAR II, pg. 4) 2.2 LDfITING SAFETT SYSTEM SETTINGS Applicability This specification applies to the scram settings which prevent j
the safety limit from being reached.
Objective The objective is to prevent the safety limits from being reached.
Specification The limiting safety system settings shall be 400*C (750*?) as measured in an instrumented fuel element. For a mixed core, the instrumenced element shall be located in the region of the core containing FLI? cype elements.
Basis O
The limiting safety system setting is a temperature which, if exceeded, shall cause a reaccer scram to be initiated preventing the safety liste from being exceeded. A setting of 400*C pro-vides a safety nargin of 750*C for FLI? type fuel elements and a margin of 600*C for standard !RICA fuel elements. A part of the safety margin is used to account for the difference between References to the Safety Analysis Report and its amendments l
will be abbreviated as:
l SAR - Safety Analysis Report, August 1967 SAR I - Amendment I to SAR, April 1968 SAR II - Amendment II to SAR, December 1972.
i L
y=.
(
I 1
i i
the true and measured temperatures resulting from the actual T
location of the thermoccuple. If the thermocouple element is located in the hottest position in the core, the difference i
between the true and measured temperatures will be only a few degrees since the thermocouple junction is at the mid-plane of the element and close to the anticipated hot spot.
If the thermocouple element is located in a region of lower temperature, j
such as on the periphery of the core, the seasured temperature will differ by a greater amcune from that actually occurring at the core hot spot. Calculations indicate that, for this case, the true temperature at the hottest location in the core will differ from the sessured temperature by no more than a 4
l factor of two.
Thus, when the temperature in the thermocouple element reaches the trip setting of 400*C, the true temperature
]
at the hottest location would be no greater than 800*C providing a margin to the safety limit of at least 200*C for standard fuel i
elements and 350*C for yLI? cype elements. These margins are
(}
ample to account for the remaining uncertainty in the accuracy 4
of the fuel temperature measurement channel and any overshoot in reactor power resulting from a reactor transient during steady state mode operation. For a sized core (i.e., one containing both standard and FLI? cype elements), the requirement that the i
instrumented element be located in the FLI? region of the core provides an even greater margin of safety since the peak to average power ratio within that region will be smaller than over j
an entire core composed of elements of the same type.
i In the pulse mode of operation, the same limiting safety system setting vill apply. However, the temperature channel will have no effect on limiting the peak powers generated because of its relatively long time constant (seconds) as compared with the width of the pulse (milliseconds). In this acde, however, the temperature.crip will act to reduce the amount of energy generated in the entire pulse transient by cutting cf the " tail" of the energy transient in the event the pulse red remains stuck in O
the fully withdrawn position.
I 3.0 LIMITING CONDITIONS FOR OPERATION 3.1 S""IADY STA*E CpERATION Applicabiliev This specification applies to the energy generated in the reactor during steady state operation.
Objective The objective is to assure that the fuel temperature safety 1121: will not be exceeded during steady state operation..
g---,
w-
,m,v-,,,
,r,-,,---
-ww
-m-----w-e--,
m,
,-,e,g--er, e,r.,n,-e.4 m
--~-m ew
.-w
- e,--eowe
i Soecifications The reactor power level shall not exceed 1.3 megawatts under any condition of operation. The normal steady state operating power level of the reactor shall be 1.0 megawatts. However, for purposes of testing and calibration, the reactor may be operated at higher power levels not to exceed 1.3 megawatts during the testing period.
P 3ases Thermal and hydraulic calculations indicate that TRICA fuel may be safely operated up to power levels of at least 2.0 megawatts with natural convection cooling.
l 3.2 REACTIVITY LIMITATIONS O
Applicability These specifications apply to the reactivity condition of the reactor and the reactivity worths of control rods and experiments.
They apply for all modes of operation.
6 Objective The objective is to assure that the reactor can be shut dcwn at all times and to assure that the fuel temperature safety
~
limit will not be exceeded.
[
i I
i O
i
(
I I
P i
l
- Sa -
i i
e i
-~
.e e--
,~e.
y 7.~
y v
Specifications The reactor shall not be operated unless the shutdown margin provided by control rods shall be greater than 0.25 dollar with:
the highest worth non-secured experiment in its most s.
reactive state, b.
the highest control rod and the regulating rod (if not scrammable) fully withdrawn, and c.
the reactor in the cold critical condition without sanon.
Bases O
a.
The value of the shutdevn sargin assures that the reactor can be shut down from any operating condition even if the highest worth control rod should renain in the fully with-drawn position. If the regulating rod is not scrammable, its worth is not used in determining the shutdown reactivity.
3.3 FULSE MODE OPERATICN Applicability This specification applies to the energy generatad in the reactor as a result of a pulse insertion of reactivity.
Objective The objective is to assure that the fuel temperature safety limit vill not be exceeded.
Specification The reactivity to be inserted for pulse operation shall be determined and limited by a mechanical block on the pulse rod, such that the reactivity insertion will not exceed 2 dollars.
Bases Measurements perforned on the Puerto Rico Nuclear Center TRIGA-FLI? reactor indicated that a pulse insertion of reactivity of 2 dollars resulted in a maximum temperature rise of approxinately 400*C.
With an ambient water temperature of approx 1=ately ICO*C, the maximum fuel camperature would be approximately 500*C resulting l
in a safety margin of 500*C for standard fuel and 650*C for FL:?
cype fuel. These =argins allow a= ply for uncertainties due to the accuracy of seasurement or location of the instrumented fuel element or due to the extrapolation of data from the ?RNC reactor.
9 l
l
~
4 3.4 CORE CONFIGURATION LIMITATION J
j Applicability i
This specification applies to saxed cores of FLIP and standard types of fuel.
Objective The objective is to assure that the fuel temperature safety limit will not be exceeded due to power peaking ef fects in a sized core.
Specificatiens a.
The FLI? fueled region in a mixed core shall contain at
()
least 22 FLI? fuel rods in a contiguous block of fuel 2
in the central region of the reactor core. Water holes in the FLI? region shall be limited to single red holes.
b.
The FTR as defined by 2.36 and as calculated by the method i
used in Amendment I to the S. A.R. shall not exceed 1.5 for i
an operational core.
Bases j
a.
The limitation of the allowable core configurations as set forth in Section 5.0 of Amendment I to the W.S.U.
j IRIGA reactor S.A.R. limits power peaking effects. The limitation on power peaking effects insures that the fuel temperature limit will not be exceeded in a mixed core.
i
~
b.
A 5000C safety system setting and a 1.5 PTR limit the maxi-l sum possible steady state fuel temperature in the FLIP
-s,,)
region to below $00cc.
i i
i s
t 4
- 9a -
i e
9 9
9 t-
J
?
i 3.3 CONTROL AND SAFETY SYSTEM 3.3.1 Scram Time t
[
Applicability g
J This specification applies to the time required for the 4
scrammable control rods to be fully inserted from the instant that a safety channel variable reaches the Safety System l
Setting.
Objective t
l The objective is to achieve prempt shutdown of the reactor to prevent fuel damage.
f()
Specification 2
The scraa time measured from the instant a simulated signal reaches 1
the value of the LSSS to the instant that the slowest scrs=nable l
control rod reaches its fully inserted position shall not exceed 2 seconds.
l Basis
,j This specification assures that the reactor will be promptly l
shut down when a scram signal is initiated. Experience and
]
analysis have indicated that for the range of transients anti-1
{
cipated for a TRIGA reactor, the specified scram time is adequate to assure the safety of the reactor.
I 3.1. 2 Reactor Centrol System i
Applicabilit?
This specification applies to the information which sust be available to the reactor operator during reactor operation.
Objective The objective is to require that sufficient information is avail-able to the operator to assure safe operation of the reactor.
4 Specifiestion The reactor shall not be operated unless the measuring channels listed in the fo11 ewing table are operable.
10 -
e
,- -. -...., = - <,.,,,,
n.
a
,. ~,. -,..
.e,.
e
i Min. No.
Effec _tive, Mode Measurina Channel Oeerable S.S.
puise Fuel Element Temperature 1
X X
I Linear Power Level 1
X Los Power Level 1
I Integrated Pulse Power 1
I Bases Puel temperature displayed at the control console gives continuous information on this parameter which has a specified safety limit.
The power level monitors assure that the reactor power level is adequately monitored for both steady state and pulsing nedes of operation. The specifications on reactor power level indi-catica are included in this section since t.he power level is I
related to the fuel temperature.
3.5.3 Reactor Safety System Applicabiliev This specification applies to the reactor safety system channels.
Objective The objective is to specify the minimum number of reactor safety system channels that must be operable for safe operation.
Specification The reactor shall not be operated unless the safety channels described in Table 1 are operable.
3ases The fuel temperature and power level scrams provide protection I
to assure that the reactor can be shut down before the. safety limit on the fuel element temperature will be exceeded. The zanual scram allows the operator to shut down the system if an unsafe or abnormal condition occurs. In the event of failure of the power supply for the safety chambers, operation of the reactor without adequate instrumentatien is prevented. The preset timer insures that the reactor power level will re@2ag-#;,.
co a low level after pulsing.
j 11 -
e
..m..
_,,,_p,,
g
l l
TABLE I a
j Minimum Reactor Safety Channels Number Effective Mode t
l Safety channel Operable Function S.S.
Pulse
?
Fuel Element 1
JCRAM @ LSSS X
X Temperature Safety #1 2
SCIUM @ 125:
I I
(Power Level)
Console Scram 1
SCRAM I
X Buccon i
Safecies #1 &
1 SCRAM on loss of I
1
}
l
- 2 Detector supply voltage i
Power Supply i
Presec Timer 1
15 seconds or less i
afcar pulse l
Log Power 1
Prevent withdrawal X
l of shia-safeties at (4 x 10-3 vaces Log Power 1
Prevent ptrising I
dwe I W I
]
Transient Rod 1
Prevent applicacion X
i Posicion of air unless fully inserted j
I Shim-safecies &
1 Prevent withdrawal X
Regulating Rod Posicion l
Pool Level 1
Alars at 90% no mal X
X i
operating level i
1 i
4 A i i
e, r-.
~
b' The interlock to grevent startup of the reactor at power levels less chan 4 x 10- watts which corresponds to approximately 2 cps assures that sufficient neutrons are available for proper startup.
a The, interlock to prevent the initiation of a pulse above 1 kW is to assure that the magnitude of the pulse vill not cause the fuel element temperature safety limits to be exceeded. The interlock to prevent application of air to the transient red 1
unless the cylinder is fully inserted is to prevent pulsing the l
reactor in the steady state mode. The interlock to prevent withdrawal of the shim-safeties or regulating rod i= the i
pulse mode is to prevent the reactor frem being pulsed while 4
on a positive period. The pool level alarm is intended to alert the operator of any significant decrease in the pool level.
I s
3.6 RADIATION MONITORING SYSTEt
- f Applicability
)'
This specification applies to the radiation monitoring infor-1 macion which must be available to the reactor operator during j
reactor operation.
i Objective I
The objective is to assure that sufficient radiation monitoring i
information is available to the operator to assure safe operation j
of the reactor.
l l.
Specification l
The reactor shall not be operated unless the radiation monitoring channels listed in the following table are operable.
1 Radiation Monitoring Channels
- Function Number i
Area Radiation Monitor Monitor radiation levels 1
i within the reactor room Continuous Air Radiation l
l Monitor Exhaust Gas Radiation Monitor radiation levels 1
Monitor in the exhaust air stack Exhaust Particulate 1
Radiation Monitor 4
For periods of time for saintenance to the radiation monitoring channels, the intent of this specification will be' satisfied if they are replaced with portable grama sensitive instru=ents having their own alar =s or which shall be kept under visual observation..
p
=
a m
\\
3ases i
The radiation monitors provide information to operating personnel i
of any impending or existing danger f rom radiation so that there will be sufficient time to evacuate the facility and take the necessary steps to prevent the spread of radioactivity to the surroundings.
3.7 ARGON-41 DISCHARGE LIMIT Aoplicability This specification applies to the concentration of Argon that may be discharged from the TRIGA reactor facility.
Chiective
()
To insure that the health and safety of the public is not endan-gered by the discharge of Argen-41 from the TRIGA reactor
- facility, s
Soecification The concentration of Argon-41 in the effluent gas from the facil-icy as diluted by atmospheric air in the lee of the facility due to the turbulent wake effact shall not exceed 4.8 X 10-8 uci/a1 averaged over one year.
Bases The maximum allowable concentration of Argon-41 in air in unre-stricted areas as specified in Appendix 3, Table II of 10 CFR 20 is 4.8 x 10-8 C1/ml. Section 6.5 of Amendment 1 to the 5.A.R.
for the TRIGA reactor facility substantiates a 3.4 x 10-3 atmos-(/)
pheric dilution factor for a 4.4 g h wind speed.
A somewhat A-sore conservative value of 4 x 10- has been selected for the calculation of the Argon-41 dilution.
3.8 ENGBTZ?ED SAFETY FEX"URE - VENTTI.ATION SYSTDi Applicability This specification applies to the operation of the f acili:y ventila:ica system.
Objective The obje::ive is to assure that the ventilation systen is in opera: ion :o si: iga:e :he consequences of the possib.'e release of radioac:ive materials resul:ing from reactor operation.
14 -
e w
n
,a
1 Specification l
The reactor shall act be operated unless the facility van:ila-tion system is cperable except for periods of time necessary to persi: repair of the system. In the event of a substan:ial l
release of airborne radioactivity, :he ven:11arion syste= will be secured au:omatically by a signal f rom an exhaust air l
radiation monitor.
3ases During normal operation of the ventilation system, the concen-tration of Argon 41 in unrestricted areas is below MFC (SA.R.
pg. 101). In the event of a subs:antial release of airborne
{s-}/
radioac:1vity, the ventilation system vill be secured aute atically.
Therefore, operation of the reactor with the ventilation syste:
shui dcwn for shor: perices of time to make repairs insures the i
same degree of control of release of radioactive =aterials.
Moreover, radia: ion meni: ors wi:hin the building independen:
of those in the ventila: ion system will give warning of high levels of radiation that migh: occur during operati:n with the ventila: ion system secured.
h i
i t
I l
f l
l l
l
- 14a -
i v
'v s
~s, e
i f
3.9 trKITATIONS ON EXPERIMENTS I
Applicability
[
t This specification applies to experiments installed in the reactor and its experimental facilities.
I objective i-The objective is to prevent damage to the reactor or exces-t sive release of radicactive materials in the event of an
{'
experiment failure.
t t
Specifications
{
The reactor shall not be operated unless the following condi-()
tions governing experiments exist.
j Non-secured experinents shall have reactivity worths less i
a.
than 1 dollar.
i b.
The reactivity wceth of any single experiment shall be less than 2 dollars.
c.
Explosive materials, such as gunpowder, INT, nitroglycerin, or PETN, in quantities greater than 25 milligrams shall not be irradiated Le the reactor or experimental facili:1es.
Explosive materials in quantities less than 25 milligrams
+
may be irradiated provided the pressure produced upon detonation of the explosive has been calculated and/or experimentally demonstra:ed to be less than the design pressure of :he container.
i d.
Experiment zaterials, except fuel sacerials, which could I
O off-gas, sublime, volatilize, or produce aerosols under i
(1) normal operating :endiciens of the experiment or reactor, (2) credible accident condicions in :he reac:or, or (3) possible accident condi: ions in :he experimen: shall
[
be limited in ac:1vi:y such that if 100% of :he gasecus l
activity or radioactive aercsols produced escaped :o :he reactor room or the atomsphere, the airborosconcen::a: ion i
of radioactivity averaged ever a year would not exceed the limit of Appendix 3 of 10 CFR Part 20.
t e.
In calculations pursuant to d. above, the following assumptions i
shall be used.
t I
l i
[
l -
r i
i
1 i
4 4
(1) If the effluent frem an experimental facili:y exhausts through a holdup tank which closes auccmatically on high radiation level, at least 10% of :he gaseous activi:y or aerosols produced will escape.
i
,(2) If the affluent from an experimental facili:y exhausts through a filter installation designed for greater than 99 efficiency for 0.3 micron particles, at least l
10% of these vapors can escape.
(3) For materials whose boiling point is above 130*F and where vapors formed by boiling this material can escape j
only through an undisturbed column of water above the core, at least 10% of these vapors can escape.
f.
Each fueled experiment shall be controlled such : hat the 4
l
{'
total inventory of iodine isotopes 131 through 135 in :he experiment is no greater than 1.5 curies.
I g.
If a capsule fails and releases nacerial which could da= age the reactor fuel or structure by corrosion or other =eans, removal and physical inspection shall be performed to
+
decernine the consequences and need for correceive action.
The results of :he inspection and any corrective action taken shall be reviewed by the Director or his designated alternate and determined to be satisfactory before operation of the reactor is resumed.
Bases This specification is intended t a provide assurance tha:
l a.
the worth of a single unfastened experiment will be limited to a value such that the safety limit will not be exceeded if the posi:1ve worth of the experiment vere to be suddenly inserted (SAR II, pg. 24).
b.
The maximum worth of a single experiment is li=ited so :ha:
its removal from the cold cri:ical reac:or will not resui:
in the reactor achieving a power level high anough :o exceed the core camperature saf ety limit. Since expe-i=en:s of such worth must be fastened in place, its removal frem :he reactor operating at full power would result in a rela:ively slow power increase such : hat the reactor protec:ive systems would act to prevent high power levels frem being attained (SAR II, pg. 21).
. l l
t
\\
l
$9,
c.
This specification is intended to prevent damage to reactor components resulting from failure of an experiment involving explosive materials.
r d.
This specification is intended to reduce the likelihood
.that airborne activities in excess of the limits of Appendix 3 of 10 CFR Part 20 will be released to the aczosphere outside the facility boundary, e.
The 1.5-curie limitation on iodine 131 through 135 assures that in the event of f ailure of a fueled axperiment leading to total release of the iodine, the exposure dose at the exclusion area boundary will be less chan that allowed by L
10 CFR Part 20 for an unrestricted area.
f.
Operation of the reactor with the reactor fuel or structure
()
damaged is prohibited to avoid release of fission products.
3.10 IRRADIATIONS Apolicability This specification applies to irradiations performed in the irradiation facilities contained in the reactor pool as de-fined in Section 2.10.
Irradiations are a subclass of exper-iments that fall within the specifications hereinafter stated in this section..The surveillance requirements for irradiations are given in Section 5.3.5.b.
Ob4ective The objective is to prevent damage to the reactor, exces,sive relea'se of radioactive nacerials, or excessive personnel radiation exposure during the performance of an irradiation.
Scecifications A device or nacerial shall not be irradiated in an irradiation facility under the classificatien of an irradiation unless the following conditions exist:
a.
The irradiation meets all the specifications of Section i.9.L for an experi=ent, b.
The expected radiation field produced by the device or sample upon removal from the reactor is not more than 10 res/hr at One foot, otherwise it shall be classed as an experiment..
1
+.
i 1
c.
The device or material is encapsulated in a suitable container, d.
The reactivity worth of the device or material is $0.25 or less, otherwise it shall be classed as an experiment, and e.
The device or material does not remain in the reactor for a period of over 15 days, otherwise it shall be classed as an experiment.
4 Bases i
This specification is intended to provide assurance that the i
special class of experiments called irradiations will be per-formed in a manner that will not permit any safety limit to l
be exceeded.
/
4.0 SURVEII. LANCE REQUIREMENTS j
4.1 CENERAL i
l Applicabiliev i
This specification applies to the surveillance requirements of any system related to reactor safety.
1 Objective The objective is to verify the proper operation of any system related to reactor safety.
Specifications Any additions, modifications, or maintenance to the ventilation system,'the core and its associated support structure, the O
pool or its penetrations, the pool coolant system, the rod j
drive mechanism, or the reactor safety system shall be made I
and tested in accordance with the specifications to which the systems were originally designed and fabricated or to specifi-cations approved by the Reactor Safety 3oard. A system shall not be considered operable until af ter it is successfully cested.
?
3ases This specification relates to changes in reactor systems which could directly affect the safety of the reactor. As leng as changes or replacements to these systems continue to meet the original design specifications, then it can be assumed that they meet the presently accepted operating criteria.
4
?
17a -
i
e i
f 4.2 SAFETY LIMIT - FUEL ELDfENT TEMPERATURE Applicability This specification applies to the surveillance requirements of 1
the* fuel element camperature seasuring channel.
]
Objective 1
The objective is to assure that the fuel element temperatures
{
are properly monitored.
i Specifications a.
Whenever a reactor scram caused by high fuel element temperatura occurs, an evaluation shall be conducted to determine whether the fuel element temperature safety limit was exceeded.
s i
b.
A calibration of the temperature measuring channels shall be performed semi-annually but at intervals not to exceed 8 months.
c.
A Channel Check of the fuel element temperature measuring channel shall be made daily whenever the reactor is operated by recording a measured value of a meaningful temperature indication.
Bases j
operational experience with the TRIGA system gives assurance that the thermocouple measurements of fuel element temperatures 4
have been suf ficiently reliable to assure accurate indication of this parameter.
()
4.3 LIMITING CONDITIONS FOR OPEKATION 1
4.3.1 Reactivity Requirements Applicability These specifications apply to the surveillance requirements for j
reactivity control of experiments and systems.
l Objective s
i The objective is to measure and verify the worth, perfor:ance, and operability of those systems affecting the reactivity of i
the reactor.
4 '
J I
f v
v a
s,,,
.-y
.n-
--n
l S pecifica tions The reae:ivity wor:h of each control red and :he shutdown a.
at intervals not sargin shall be determined annually but to exceed 14 mon:hs.
b.
The reactivity worth of an experimen: shall be estima:ed or measured, as appropriate, before reactor operation vi:h said experiment.
The control rods shall be visually inspected for deteriora-c.
tion at intervals not to exceed 2 years.
d.
The transier.: rod drive cylinder and associated air supply system shall be inspected. : leaned, and lubricated as necessary semiannually a: intervals not :: exceed 5 men:hs.
The reactor shall be pulsed semiannually to co= pare fuel
[~])
e.
temperature measurements and peak power levels vi:h those of previous pulses of the same reactivity value or the reactor shall not be pulsed until such comnarative pulse measuremen:s are performed.
3ases The reac:ivi:y worth cf the :entrol rods is censured :o assure that the required shn:devn margin is available and to previde an accurate means f or determining the rea::ivi:7 worths of experi-sents inser:ed in the core. Pas: experience vi:n IRIGA reae:crs gives assurance :ha: measuremen: ef the rea ::vi:7 worth en an annual basis is adequate :c insure ne signifi:an: changes in :he shu:devn margin. The visual inspec:i:n Of :he cent:01 rods is made to evaluate corresten and wear :naracteris:1:s :aused by operation in :he reac::. The rea::c is pulsed a: sui:able
)
intervals and a comparisen made vi:5 previcus similar pulses to deter:ine if changes in fuel or ::re charac:eris:1:s are taking place.
4.3.2 Contrei and Safe:v Svs:ac Aeplicabili:v These specifica: ices a;;1y :: :he surveillance requiremen:s f or measurements, tes:s, and :alibra:iens of the ::::rel and safe:y sys: ems.
Cbfee:1ve 5
The cbjective is :o verify :he perfrr:an:e and operat:11:y f those systees and compenen:s vni:h are dire::*y relatec :r reactor safety.
I
i Specifications a.
The scram time shall be measured annually but at intervals t
not to exceed 14 months.
f b.
A Channel Test of each of the reactor safety system channels j
'for the intended mode of operation shall be performed prior to each day's operation or prior to each operation extending acre chan one day, except for the pool level channel which j
shall be tested weekly..
l t
A Channel Calibration shall be made of the power level c.
aonitoring channels by the calorimetric method annually but at intervals not to exceed 14 months.
[
3ases
()
Measurement of the scram time on an annual basis is a check not only of the scram system electronics, but also is an indication
[
of the cap ability of the control rods to perform properly. The channel te sts will assure that the safety system channels are operable o n a daily basis or prior to an extended run. The pewer level channel calibration vill assure that the reactor vill be operated at the proper power levels. Transient control rod checks and semiannual maintenance insure proper operation of this control rod.
4.3.3 Radiation Monitoring Svstes Applicabiliev This specification applies to the surveillance require =ents for the area radiation monitoring equipment and the continuous air monitoring system.
()
Objective r
The objective is to assure that the radiation sonitoring equipment is operating and to verify the appropriate alarm settings.
i Soecification The area radiation sonitoring system and the continuous air l
monitoring systen shall be calibrated annually but at intervals j
1 not to exceed 14 months and shall be verified to be operable at weekly intervals.
f 3 asis Experience has shown that weekly verification of ares radiation and air monitoring system set points in conjunctica with annual calibration is adequate to correct for any variation in the system due to a change of operating characteristics over a long.
I time span.
r
- P y
s 4.3.4 Ventilation System Applicability This specification applies to the building confidement ventila-tion system.
Objective The objective is to assure the proper operation of the ventilation systen in controlling releases of radioactive sacerial to the uncontrolled environment.
Specifiestion It shall be verified weekly that the ventilation system is operable.
Bases Experience accumulated over several years of operation has demonstrated that the tests of the ventilation system on a weekly basis are sufficient to assure the proper operation of the systes and control of the release of radioactive material.
- 4. '3. 5 Experiment and Irradiation Limits Applicabiliev This specification applies to the surveillance requirements for experiments installed in the reactor and its experimental facilities and for irradiations performed in the irradiation facilities.
Objective Os The objective is to prevent the conduct of experiments or irradiations which may damage the reactor or release excessive amounts of radioactive materials as a result of failure.
Specifications a.
A new experiment shall not be installed in the reactor or its experimental facilities until a hazards analysis has been performed and reviewed for compliance with the Limita-tions on Experiments, Section 3.6, by the Reactor Safeguards 3 card. Minor modifications to a reviewed and approved experi-ment may be made at the discretion of the senior reactor operator responsible for the operation provided that the ha:ards associated with the nodifications have been reviewed and a determination made and documented that the modifications do not create a significantly different, a new, or a greater than the original approved experiment.
i.
m
t l
l t
i r
b.
An irradiation of a new type of device or material shall not be performed until an analysis of the irradiation has been performed and reviewed for compliance with the Limitations on Irradiations, Section 3.6. by a licensed senior operator qualified in health physics, or a licensed senior operator and a person qualified in health physics.
Bases f
F It has been demonstrated over a number of years of experience l
that experiments and irradiations reviewed by the Reactor Staff and the Reactor Safeguards 3oard as appropriate can be conducted without endangering the safety of the reactor or exceeding the limits in the Technical Specifications.
O t
r f
I r
I i
t I
e
- 21a -
l
e 4.4 REACTOR FUEL ELIMENTS Applicability This specification applies to the surveillance require =ents for the fuel elements.
Objective The objective is to verify the continuing integrity of the fuel element cladding.
Specifications All fuel elements shall be inspected visually for damage or deterioration and measured for length and bend at intervals not to exceed the sum of 3,500 dollars in pulse reactivity.
n.
(,)
The reactor shall not be operated with damaged fuel. A fuel element shall be considered damaged and must be removed frem the core if:
a.
In asasuring the transverse band, the bend exceeds 0.125 inch over the length of the cladding, b.
In measuring the elongation, its length exceeds its original length by 0.125 inch, or c.
A clad defect exists as indicated by release of fission products.
Bases The frequency of inspection and seasurement. schedule is based on the parameters sost likely to affect the fuel cladding of O
a pulsing reactor operated at moderate pulsing levels and utili:ing fuel elements whose characteristics are well known.
The limit of transverse bend has been shewn to result is no difficulty in disassembling the core. Analysis of the renoval of heat from touching fuel elements shews that there will be no hot spots resulting in dasage to the fuel caused by this touching. Experience with TRIGA reactors has shcwn that fuel element bowing that could result in touching has occurred withouc l
deleterious effects. The elongation limit has been specified to assure that the cladding nacerial will not be subjected to stresses that eculd cause a loss of integrity in the fuel con-tainment and to as,sure adequate coolant flow.
t I
l 1.
,.v,
,e
--,---,-e
,,,,--w-
i j
=
i
]
5.0 DESIGN FEATURIS i
5.1 REACTOR FUEL I
Applicability This specification applies to the fuel elements used in the i
reactor core, i
objective The objective is to assure that the fuel elements are of such a design and fabricated in such a manner as to permit their 4
use with a high degree of reliability with respect to their i
[
physical and nuclear characteristics.
Specifications a.
TRIGA-FLIP Fuel The individual unieradiated FLIP fuel elements shall have i,
the following characteristics:
i (1) Uranium content: maximum of 9 Wt-enriched to j
nominal 70T Uranium 235.
(2) Hydrogen-to-zirconium atom ratio (in the ZrHx):
nominal 1.6 H atoms to 1.0 Zr atoms.
1 (3) Natural erbium content (homogeneously distributed):
n=4 n = 1 1.5 Wt-%.
l i
(4) Cladding: 304 stainless steel, nominal 0.020 inch O
=*1=*-
1 (5) Identification: Top pieces of FLIP ele =ents will have j
characteristic markings to allow visual identification of FLIP elements employed in mixed cores.
b.
Standard TRIGA fuel The individual unirradiated standard TRIGA fuel elements i
shall have the following characteristics:
i (1) Uranium content: maximum of 9.0 Wt " enriched to a nominal 20 Uranium 235.
(2) Hydrogen-to-zirconium atom ratio (in the ::rH ):
3 nominal 1.7 H atems to 1.0 Zr atems.
23 -
o
,e-y cw-e
-m, e-,
+--,,
w
---r-a
i (3) Cladding: 304 stainless steel, nominal 0.020 inch j
i thick.
s t
Bases a.
A' maximum uranium content of.9 WC-% in a TRIGA-RI? element is about 6% greater than the design value of 5.3 We-%.
Such l
an increase in loading would result in an increase in power density of about 2%. Similarly, a minimum erbium content of 1.1% in an element is about 30% less than the design value.
This variation would result in an increase in power density of only about 6%. An increase in local power density of l
6% reduces the safety margin by at most can percent. The l
mavf== hydrogen-to-zirconium ratio of 1.65 could result in a maximum stress under accident conditions in the fuel element clad about a factor of two greater than the value resulting i
<==
8 4===
-= -=1=c 1=====1
=< t so-a
=. =*1-O.
7 increase in the clad stress during an accident would not exceed the rupture strength of the clad.
l When standard and RI? fuel elements are used in sixed cores, visual identification of types of elements is necessary to verify correct fuel loadings. The accidental rotation of fuel i
bundles cont =fning standard and RI? elements can be detected by visual inspection. Should this occur, however, studies of a single RI? element accidently rotated into a standard j
fuel region indicate an insubstantialincrease in power gener-acion in the RIP element.
t l
b.
A maximum uranium content of 9 WC-% in a standard TRICA element is about 6% greater than the design value of 8.5 t
Wt-%.
Such an increase in loading would result in an increase in power density of less than 6%. An increase in local power density of 6% reduces the safety margin by at most 10%. The mmr hum hydrogen-to-circonium ratio of 1.8 will produce a maximum pressure within the clad during an accident well below the rupture strength of the clad.
l 5.2 REACTOR CORI j
Applicability i
This specification applies to the configuration of fuel and l
in-core experiments.
i 6,
i l
i
h Objective The objective is to assure that previsions are made to restrict the arrangement of fuel elements and experiments so as to provide assurance that excessive power densi:ias will not be produced.
Specifications a.
The core shall be an arrangement of IRICA uranium-i zirconium hydride fuel-moderator bundles positioned in the reactor grid place.
b.
The TRIGA core assembly may be scandard, FLI?, or a combination thereof (mixed core) previded that any FLIP fuel be comprised of at least eventy-three (23) fuel elements, located in a conciguous, can:ral region.
O r
c.
The reactor shall not be operated with a core lattice post-tion vacant except for positions on the periphery of :he core assembly.
d.
The reflector, excluding experiments and experi= ental l
facili:ies, shall be vacer or a combination of graphite and water.
Baseg a.
Scandard TRIGA cores have been in use for years and their characteristics are well documented. The Gulf Mark III all-FLIP core is operacional and characteristics are available'. Gulf has also performed a series of experi-I ments using standard and FLI? fuel in mixed cores.
In addition, studies performed at Texas A&M for a variety
[
of nixed core arrangements indicace that such cores with
()
zixed loacings would safely satisfy all operational recuire-l
=ents (SAR II).
b.
In mixed cores, it is necessary to arrange FLIP elements in a conciguous, central region of the core to contrar flux
[
peaking and pcwer generation peak values in individual elements.
i c.
Vacant core lactice positiens will contain experimen:s or an experimental facility to prevent accidental fuel additions i
to the reactor core. They will be persicted only on the i
periphery of the core :o prevent pcwer perturba:1ons in regions of high pcwer densi:y.
1
- 25 r-e r
' ~.
4 i
d.
The core will be assembled in the reactor grid place which is located in a pool of light water. Water in combination l
with graphite reflectors can be used for neutron economy and the enhancement of experimental facility radiation
{
requirements.
5.3 CCNEROL RCDS Applicability j
This specification applies to the control rods used in the j
reactor core.
1 Objective i
The objective is to assure that the control rods are of such O
a design as to permit their use with a high degree of reliability v
with respect to their physical and nuclear characteristics.
1 i
l Specification a.
The shim-safety control rods shall have scram capability and contain borated graphite, B g powder or boron and its 4
compounds in solid form as a poison in aluminum or stainless l
steel cladding. These rods may incorporate fueled fo11cwers j
which have the same characteristics as the fuel region in which they are used.
j b.
The regulating control rod need not have scram capability
)
and shall be a stainless rod or contain the materials as I
specified for shim-safety control rods. This red may incer-potate a fueled follower.
I c.
The transient control rod shall have scram capability and con-i ()
tain borated graphite or boren and its compounds in a solid l
form as a poison in an aluminum or stainless steel clad. The 4
j transient rod shall have an adjustable upper limit to allow a variation of reactivity insertions. This red nay incor-j porate an alwalnum or air follower.
]
3ases J
l The poison requirement 3 for the centrol rods are satisfied by l
using neutron absorbing borated graphite, 3 8 powder or boren 4
and its compounds. Since the regulating rod norsally is a lov worth rod, its function could be satisfied by using a solid stainless steel rod.- These materials must be contained in a t
r e
4
'~
suitable clad material, such as aluminum or stainless steel, to insure mechanical stability during movement and to isolate the poison from the pool water environ =ent.
Control rods that are fuel followed provide additional reactivity to the core and l
increase the worth of the control rod. The use of fueled followers in the FLl? region has the additional advantage of reducing flux peaking in the water filled regions vacated by the withdrawal of the control rods. Scram capabilities are provided for rspid insertion of the control rods which is the primary safety feature of the reactor. The transient control rod is designed for a reactor pulse. The nuclear behavior of the air or aluminum I
follower which may be incorporated into the transient rod is sfailar to a void. A voided follower may be required in certain core loadings to reduce flux peaking values.
5.4 RADIATION MONITORING STS"'EM
)
Applicability 1
This specification describes the functions and essential co=ponents of the area radiation monitoring equipment and the system for continously monitoring airborne radioactivity.
a Objective The objective is to describe the radiation monitoring equipment that is available to the operator to assure safe operation of the reactor.
Specification The radiation monitoring equipment listed in the following table will be available for reactor operation.
()
Radiation Monitoring Channel and Function Area Radiation Monitor (game:a sensitive instruments)
Function - Monitor radiation fields in key locations, alarm and readout at control console and readout in reception roca.
Continuous Air Radiation Monitor (beta, ganna sensitiva detector with air collection capability)
Tunction - Monitor concentration of radioactive particulate activity in building, alarm and readout at control console and readout'in reception reem.
6 l
l
( s g
l l
l
' ~.
Gas and Particulate Stack Radiation Monitor (gamma sensitive detector with air collection capability)
Function - Monitor concentration of radioactive particulate activity and radioactive gases in building exhaust, alarm and readout at control console and readout in reception room.
Basis The radiation monitoring system is intended to provide infor-nation to operating personnel of any impending or existing danger from radiation so that there will be sufficient time to evacuate the facility and take the necessary steps to prevent the spread of radioactivity to the surroundings.
5.5 FUEL STORAGE Applicability This specification applies to the storage of reactor fuel at times when it is not in the reactor core.
Objective The objective is to assure that fuel which is being stored will not become critical and will not reach an unsafe temperature.
Specifications a.
All fuel elements shall be stored in a geometrical array where the k-effective is less than 0.8 for all conditions of moderation.
()
b.
Irradiated fuel elements and fueled devices shall be stored in an array which will permit sufficient natural convec-tion cooling by water or air such that the fuel element or fueled device temperature will not exceed design values.
Basis The linics imposed by Specifications 5.5.a and 5.5.5 are conservative and assure safe storage.
l 5.6 REACTOR 3UILDDiC AND VDiTILATICN SYSTEM Applicability This specification applies to the building which houses the reactor..
l
' ~.
Objective j
i The objective is to assure that provisions are nade to restrict i
che amount of release of radioactivity into the environnent.
j Specifications a.
The reactor shall be housed in a facility designed to restrict leakage. The minimum free volume in the facility shall be 180,000 cubic feet.
b.
The re, actor building shall be equipped with a ventilation system designed to filter and exhaust air or other gases from the reactor building and release them from a stack at a =4 a 4 ="= of 85 feet frem ground level.
c.
Emergency shut down controls for the ventilation system O
shall be located in the reception room and the system shall be designed to shut down in the event of a substantial release of fission products.
Bases The facility is designed such that the ventilation system will normally maintain a negative pressure with respect to the atomsphere so that there will be no uncontrolled leakage to the environment. The free air volume within the reactor building is confined when there is an emergency shutdown of the ventila-tion system. Controls for startup, emergency filtering, and normal operation of the ventilation system are located in the receptica room. Proper handling of airborne radioactive nacerials (in emergency situations) can be conducted from the reception r:om with a =4a4=n= of exposure to operating personnel (SAR, pg.56).
5.7 REACTOR PCOL WATER SYSTD*.S Applicabilit?
This specification applies to the pool containing the reactor and to the cooling of.the core by the pool water.
Objective The objective is to assure that coolant water shall be available to provide adequate cooling of the reactor core and adequate radiation shielding.
6 e.
m l
1
Specifications a.
The reactor core shall be cooled by natural convective i
water flow.
- b.
- The pool water inlet and outlet pipe to the deminerali:er L
shall not extend more than 15 feet below the top of the reactor pool when fuel is in the core.
f Diffuser and skinner pumps shall be located no more than c.
15 feet below the top of the reactor pool.
d.
Pool water inlet and oucist pipe to the heat exchanger shall have emergency covers within the reactor pool for manual shut off in case of pool water loss due to external pipe system failure.
()
A pool level alarm shall indicate loss of coolant if the a.
po.s1 level drops approximately 2 feet below normal level.
Bases This specification is based on thermal and hydraulic cal-a.
culations which show that the TRIGA-FLIP core can operate in a safe manner at power levels up to 2,700 kW with natural convection flow of the coolant veter. A compari-son of operation of the TRIGA-FLIP and standard TRIGA Mark III has shown to be safe for the above power level.
Thermal and hydraulic characteristics of nixed cores are essentially the same as that for TRIGA-FI.I? and standard cores.
()
b, In the event of accidental siphoning of pool water through inlet and outlet pipes of the demineralizer system, the pool water level will drop no more than 15 feet frem the top of the pool, c.
In the event of pipe failure and siphoning of pool vater through the skinner and diffuser water systems, the pool water level will drop no more than 15 feet from the top of the pool.
d.
Inlet and outlet coolant lines to the pool heat exchanger terminats at the bottom of the pool. In the event of pipe failure, these lines aust be manually sealed from within i
l l
~.
the reactor pool. Covers for these lines will be stored in the reactor pool. Time required to uncover the reactor core due to f ailure of a single pool coolant pipe system 2
is 17 minutes, Loss of coolant alarm after 2 feet of loss required cor-f e.
i rective action. This alarm is observed in the reactor control room and outside the reactor building.
1 i
6.0 ADMINISTRATIVE CCEROI.S
\\
6.1 ORGANI7.ATION 4
The facility shall be under the direct control of the 5
a.
4 Director (NSC) or a licensed senior operator designated by him to be in direct control. The Director shall be respon-l sible to the Dean of the College of Engineering and Director i
of the Texas Engineering Experiment Station for safe operation and maintenance of the reactor and its associated equipment. The Director (NSC) or his appointee shall review and approve all experiments and experimental procedures prior to their use in the reactor. He shall enforce rules for the protection l
of personnel against radiation.
1 i
b.
The safety of operation of the Nuclear Science Center Reacect 2
shall be related to the University Administration as shown 1
in the following chart.
Office, President j
i Texas A&M University l
l Dean of 1
Radiological Engineering Reactor Safety and Safety 4
l Office Director TIES
_3 card i
I 6
I I
D1CC '
i L_
Nuclear Science Center
-J
+
~~
~~ --
i i
Reactor j
Operations i
r 4
d I
- 1 r
l l
1 6.2 REVIEW AND ALTIT a.
A Reactor Safety Board (RSB) of at least three (3) members knowledgeable in fields which relate to Nuclear Safety shall review, evaluate, and approve safety standards associated
,with the operation and use of the facility. The University Radiological Safety Of ficer shall be an ex-of ficio member of the Reactor Safety Board. The jurisdiction of the RSS shall include all nuclear operations in the facility and general safety standards, b.
The operations of the Reactor Safety Board shall be in accordance with a written charter, including provisions for:
(1) Meeting frequency,
()
(2) Voting rules.
(3) Qucrums, (4) Method of submission and content of presentation to the Committee, (5) Use of subcommittees, and (6) Review, approval, and dissemination of minutes, c.
The RSB or a Subcommittee thereof shall audit reactor operatiocs at least quarterly, but at intervals not to exceed four months.
d.
The responsibilities of the Board or designated Subcommittee thereof include, but are not limited to, the following:
()
(1) Review and approval of experiments utill:ing the reactor facilities, (2) Review and approval of all proposed changes to the facility, procedures, and Technical Specifications, (3) Review of the operation and operational records of the
- facility, (4) Review of unusual or abnormal occurrences and incidents which are reportable under 10 CFR Part 20 and 10 CFR Part 50, 6.
= = = -
~.
(5) Determination of whether a proposed change, test, or experiment would constitute an unreviewed safety question or a change in the Technical Specifications, and
- (6) Review of abnormal performance of facility equipment and operating anomalies.
6.3 ACTION TO BE T.UCEN IN THE E7ENT A SAFETY I.IMIT IS EXCEEDED In the event a safety limit is exceeded:
The reactor shall be shut down and reactor operation shall a.
not be resumed until authorized by the AEC, b.
An 1:enediate report of the occurrence shall be made to the Chairman, Reactor Safety Board, and reports shall be made to the AEC in accordance with Section 6.7 of these specifications, and A report shall be prepared which shall include an analysis c.
of the causes and extent of possible resultant damage, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recur-rence. This report shall be submitted to the Reactor Safety Board for review and then submitted to the AEC when authorization is sought to rerums operation of the reactor.
6.4 ACTICN TO BE TAKEN IN THE EVENT OF A REPORTAETl OCC"omic?
In the event of a reportable occurrence, the following action shall be taken:
The Director (NSC) or his designated alternate shall be a.
notified and corrective action takan with respect to the operations involved, b.
The Director (NSC) or his designated alternate shall notify the Chairman of the Reactor Safety 3eard, A report shall be made to the Reactor Safety Board which c.
shall include an analysis of the cause of the occurrence, efficacy of corrective action, and recommendations for measures to prevent or re.uce the probability of recur-rence, and 6
_.33 -
7-d.
. A report shall be made to the NRC in accordance with Section 6.7 of these specifications.
6.5 OPERATING pROCEtt:RES Written operating procedures shall be adequate to assure the safety of operation of the reactor, but shall not preclude the use of independent judgement and action should the situa-tion require such. Operating procedures shall be in ef fect for the following items:
Testing and calibration of reactor operating instru-a.
mentation and controls, control red drives, area radia-tion monitors, and. air particulate monitors; b.
Reactor startup, operation, and shutdown; O.
Emergency and abnormal conditions, including provisions c.
for evacuation, reentry, recovery, and medical support; d.
Fuel element and experiment loading or unloading; e.
Control rod removal or replacement; f.
Routine maintenance of the control red drives and reactor safety and interlock systems or other routine maintenance that could have an effect on reactor' safety
- Actions to be taken to correct specific and foreseen g.
potential malfuncticas of systems or components, including responses to alarms and abnormal reactivity changes, and h.
Civil disturbances on or near the facility site.
Substantive changes to the above procedures shall be =ade only with the approval of the Reactor Safety Board. Temporary changes to the procedures that do not change their original intent may be made by the Director (MSC) or his designated alternate.
All such temporary changes shall be documented and subsequently reviewed by the Reactor Safety Board.
6.6 FACILITY OPERATING RECORDS In addition to the requirements of applicable regulations, and in no way substituting therefor, records and logs shall be prepared of at least the folicwing items and retained for a period of at least five years for items a through f and indefinitely for items g through k.
- 3/4 ep
___ _ _____ ____ _ _ _ _ _j
~~,
a.
Normal reactor operation, b.
Principal maintenance activities, c.
Reportable occurrences, d.
' Equipment and component surveillance activities required by the Technical Specifications, e.
Experiments performed with the reactor, f.
Gaseous and liquid radioactive effluents released to the
- environs, g.
Offsite environmental monitoring surveys, h.
Fuel inventories and transfers,
{}
1.
Facility radiation and contamination surveys,
- j. Radiation exposures for all personnel, and k.
Updated, corrected, and as-built drawings of the facility.
6.7 REPORTING REQUIREMENTS In addition to the requirements of applicable regulations, and in no way substituting therefor, reports shall be made to the NRC Region IV, Office of Inspection and Enforcement as follows:
A report - ithin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and' telegraph.
~
a.
()
(1) Any accidental release of radioactivity above per-missible limits in unrestricted areas whether or not the release resulted in property damage, personal injury, or exposure; (2) Any violation of the safety limit; and (3) Any reportable occurrences as defined in Sectica 1.11 of these specifications.
b.
A report within 10 days in writing of:
l I
f l
l I
(1) Any accidental release of radioactivity above per-missible limits in unrestricted areas whether or not the release resulted in property danage, personal injury, or exposure. The written report (and, to the extent possible, the preliminary telephone or tele-graph report) shall describe, analy:e, and evaluate safety implications, and outline the corrective =easures taken or planned to prevent reoccurrence of the event; J
(2) Any violation of a safety linic; and (3) Any deportable occurrence as defined in Section 1.11 of these specifications.
A report within 30 days in writing of:
c.
)
l I
l (1) Any significant variation of nessured values fron a corresponding predicted er previously =easured value 1
of safety-connected operating characteristics occurring during operation of the reactor; j
I (2) Any significant change in the transient or accident analysis as described in the Safety Analysis Report; i
(3) Any changes in f acility organization; and l
(4) Any observed inadequacies in the inplementation of administrative or procedural controls.
l 6.7.1 A report within 90 days after completion startup testing of
()
the reactor upon receipt of a new facility license or an amendment to the license authorizing an increase in reactor power level describing the measured values of the operating con-dicions or characteristics of the reactor under the new condi-tions including:
a.
An evaluation of facility performance to date in ecmparisen with design p'redicticas and specifications, and b.
A reassessment of the safety analysis submitted with the license application in light of seasured operating charac-teristics when such measurements indicate that there may be substantial variance frem prior analysis.
6.7.2 An annual report covering the operation of the unit during the previous calendar year submitted prior to March 31 of each year providing the following information:
- a.. A brief narrative summary of (1) operating experience (including experiments performed), (2) changes in facility design, perf ormance characteristics, and operating pro-cedures related to reactor safety and occurring during the reporting period, and (3) results of surveillance tests and inspections; b.
Tabulation of the energy output (in negawatt days) of the reactor, hours reactor was critical, and the cumulative total energy output since initial criticality; c.
The number of emergency shutdowns and inadvertent scrams, including reasons therefor; d.
Discussion of the major maintenance operations performed
, - on the during the period, including the effect, if any safety of the operation of the reactor and tae reasons for any corrective maintenance required; A brief description, including a summary of the safety e.
evaluations of changes in the facility or in procedures and of tests and experiments carried out pursuant to Section 30.59 of 10 CFR Part 50; f.
A summary of the nature and amount of radioactive affluents released or discharged to the environs beyond the ef fective
, control of the licensee as seasured at or prior to the point of such release or discharge.
Liquid Waste (summariced on a monthly basis)
(1) Radioactivity discharged during the reporting period.
(a) Total radioactivity released (in curies).
(b) The MFC used and the isotopic composition if greater than 1 x 10-7 alcrocuries/cc for fission and activation products.
(c) Total radioactivity (La curies), released by nuclide, during the reporting period based on representative isotopic analysis.
?
f (d) Average concentration at point of release (in microcuries/ c) during the reporting period.
l 37 -
pp
-e w
we
--e v
-g---,
w, e
i-w---em
+--1--m-
-t-
-* e is
-*-7 sC r-7 a~'u*'-y T*i-wy y
emp^
-y#+ew
-p-y'
(2) Total volume (in gallons) of effluent water (including dilutent) released during each period of release.
Caseous Waste (summari:ed on a monthly basis)
J (1) Radioactive discharged during the reporting period (in curies)
(a) Total estimated quantity of radioactivity released (in curies) determined by an appropriate sampling and counting method.
)
(b) Totil estimated quantity of Argon-41 released i
(in curies) during the reporting period based on data from an appropriate monitoring system.
f~)
(c) Estimated average atmospheric diluted concentration of Argon-41 released during the reporting period in terns of microcuries/cc and fraction of the appli-cabla MFC value.
(d) Total estimated quantity of radioactivity in par-ticulate form with half lives greater than eight days (in curies) released during the reporting period as determined by an appropriate particulate monitoring system.
(e) Average concentration of radioactive particulates with half lives greater than eight days released j
in microcuries/cc during the reporting period.
I (f) An estimate of the average concentration of other i
significant radionuclides present in the gaseous I
waste discharge in terms of microcuries/cc and
()
fraction of the applicable MPC value for the re-porting' period if the estimated release is greater than 20T of the applicable MPC.
Solid Waste (su=mari:ed on an annual basis)
(1) Total amount of solid waste packaged (in cubic feet)
(2) Total activity in solid waste (in curies)
(3) The dates of shipment and disposition (if shipped off site).
t
- 37a -
g.
An annual summary of the radiation exposure received by facility personnel and visitors in terms of the average radiation exposure per individual and greatest exposure per individual in the two groups.
Each significant ex-posure in excess of the limits of 10 CFR 20 should be reported including the time and date of the exposure as well as the name of the individual and the circumstances leading up to the exposure.
h.
An annual summary of the radiation levels and levels of i
contamination observed during routine surveys performed at the facility in terms of the average and highest levels.
l 1.
An annual summary of any environmental surveys performed outside the facility.
[ ()
(2) Total volume (in gallons) of effluent water (including i
dilutent) during periods of release, t
1 l
Gaseous Waste (summari:ed on a monthly basis) 1 Radioactivity discharged during the reporting period (in l
euries) for:
(1) Argon-41 l
l (2) Particula:es with half lives greater than eight i
days.
1 Solid Waste toummarized on an annual basis) j (1) The total amount of solid waste packaged (in cubic feet).
1 1 ()
(2) The total activity involved (in curies).
I i
(3) The dates of shipnen; and disposition (if shipped off s ite).
1 1
- j. A su= mary of radiation exposures received by facili:y per-l sonnel and visitors, including dates and tine of significan:
)
exposures and a su=sary of the results of radiation and i
contamination surveys perfor:ed within the facility; and k.
A description of any environnental surveys perforned outside the facility.
i
- 3a -
4 r
-e
1 w
i 1
1 1
1
\\
i i
1 i
i i
)
i I
GUIDIVCE l'OR s,
i l
1:0:!-l'0"ER 11 ACTOR 5 O
SEC110;; C.0 0
1 1
AD?.I!!ISTi>J.'llVE CO:!TROLS i
l e
l 4
O I
O ey I
k i
i I
i i
4 l
1
...A; w
6.0 ADMI!!ISTRATIVE C0?;TF.0LS 6.1 RESP 0::SIBILITY I
6.1.1 The (Facility Director ) shall be responsible for overall facility opera tion. He shall delegate in 1:riting the succession to this rest.on-sibility during his absence (or the' succession shall be cicarly presented in Figure 6.2-1).
6.2 ORGA :1ZATIO::
I'l-Facility Staff-b f
The organization for facility management and operation shall l
6.2.1 be as shown in Figure G.2 - 1.
a.
Each on duty shift shall be ccmposed of at least:
List cach si:r.Wer of the operating shift by position title or function, and idcntify those required to have an imC.licensa For example -
I I
2 Itinimun staff when the reactor is not secured shall includa-i
~
1.
Senior Operator (SRO) on call but not necessarly on site; i
2.
Radiation Control Technician on call; i
3.
Reactor Operotor (RC) at the controls;
(]}
4.
Anothcr personable to carry out emergency procedures in i
control rec:n.
At least ona licensed cperajor shall be at tije controls when b.
the recctor is not secured O
h O
e e
-7 A
s
~
An individual that is designated by the licensee as qualified c.
to implement routine radiation protection procedures shall k-present at the f acility v:iienaver the reactor is not securcd or tihenever any experiment or experimental facility is being serviced.
d.
All core alternations that cculd affect reactivity of the reactor shall be supervised by a licensed Senior Operator.
1.
Drachets or bor.cs are used either to set off guidance fro:n the technical specification or to identify areas of licensee option.
]dentifying positien titles contained t:ithin the bracket, here and else.:hece in the report, are representctive only.
Each posi-tion title is used consistently throughouc this guicance docum.ar.:.
P) 2.
A reactor is secured 1: hen (1-) all poison control rod.: are fully inserted, (2) t.he console key sv: itch is in the off patition and
'^'
the key is rer.:oveJ f rc:a the lock, and (3) na 1: ark is in prcgress involving fuel or incor: cxperiments, or maintenance of the cc:c structure, control rods or control red drivas.
e s
I I
i I
t 3
l
~=.
-wf*.
L
~
4 President of College Or i
Provost
+
i
)
i i
i
(
1 Dean of Engineering l
i O.
I Facility Director or Department licad
!!uc1 car C:/h'. c: 2 Green Review Audit Function Fur.cti::
.e.
l Radiation Facility Supertisor l
Safety or l
Reactor Supervisor 1
,i i
Operating Staff Figuro 6.2-1 FACILITY 0?,GA: 1Zl.T::.:.
i 1Responsible for fccilit.y operation I
i 4
1 e
6 1
p
-m.
A J..
~
.t I.
6.3 FACil.lTY STfFF OUALIFICATIO::5, Minitum c;ualifications for ec:r.bers of the facility staff may be specified by use of an overall qualificati;n statoir.ent; referencing an applicchio standard (AtiS 15. co:mittee is developing a standard titled,15.4, "Stcndrd for Selection cnd Training of Personnel for Research Reactors") or alternately, by specifying individual position qualifications.
Referencinc an applicable standard would be preferably when a standard exist:;.
Such a specification..ould be:
3 1
"Each member of the facility staff shall meet or excced the minicr; qualifications of (applicable standard) for comparable positions".
However, the follo.:ina raethed should be used whenever a cicar correlation cannot be drawn betueEn a standard and the reactor orgcriiaational struc'u.u.
Q, 6.3.l(a)
Facility Director / Reactor Supervisor I
At the til.2 of initial core loading or appointment to-the i
active position, the reactor Adainistrator/ Reactor Supervi:;or shall have a miniraua of five years of nuclear experiencc.
1:e shall have a baccalaurca;.e or higher degree in an engineering or other scientific field.
The degree niil fulfill four yctes of experience on a one-for-one ti:r.c basis.
Equivalent educc.icn or experience uay be substituted for a degree.
(b) - Senior / Supervisory Reactor Operator 4
At the time of initial core loadiag or appointen.nt to thi active position, a supcrvisor shall have a minitu: of a high school dipicu or equivalent and should have four years of nuclear cy.crience'. A mar.iuvr. of two yc'rs of experience m y O'
be fulfilled by related ecLdemic or technical training on a one-for-cnc time basic.
(c) Reactor Oparator At the tico of initial core loading or appointment to the active position, operators shall have a high sch::ol diploma cc equivalent.
6.4 TPAli:II:G 6.4.1 The (position title) shall be responsible for the facility ret:ai:.ir.:.
and replaccment training progreu.
k i
e 6
g
4 L.
l l
6.5 REVIE!! /J:D AUDIT 6.5.1 (NUCLEAP. P,EVIE!I A::D AUDIT Gp0!!P,l!!:AG)
The ecthod by :thich inde,,endent review and audit of facility cpercticus is accomplished say take one of sevccal forms. The licensee may either assign t.his function to an crgani: t.ional unit separate and indcpondens from the group having resp:ensibility for facility operation or may utilize a standing co=:ittcc corapcscd of individuals from )ithin and outside the group having ad.:iinistrative responsibility.
Irrespective of the method used, the licensce shall specify the details of each functional element provided for the independent review and cudit i
process for his particular facility as illustrated in the follcuing exaaple specifici.tions.
_,,_,j FilNCT10:1
("9 6.5.1.1 The (Mi:AS) shall function to provide independent revicu and l
audit of facility activities.
Arcas designated bolcw shall be considered:
l a.
nuclear operations 1
b.
nuclear engineering c.
cheaistry and radiochemistry d.
netallurgy c.
instrumentation' and control f.
radiological safety
,]
g.
mechanical and electrical engineering h,
tests and expericents i.
(other appropriate fields associated with the unique char-acteristics of the nuc1 car reactor) l C0!', POSITION AUD QUld.IFICATICC!S G.5.1.2 The (!:!'AG) shall be cc: posed of:
List all nc:bers by ;)osition title and indicate the Chairman.
Do not use given nur.es.
e Chairaan:
(PositionTitle)
Mem!lcr:
(Position, Title) i i
f 8
Maher:
(Position Title) e r-
,4.
>.. c
.,7._.__..
= '
s I
L l
The uinicum qualifications for persons on the (NRAG) shall be 5 years of professional work experience in the discipline or spccific field bc represents. A baccalaureate degree may fulfill 4 years of experience.
If a sepcrate orhani:ationci unit is used, the supervisor of the unit shall have a ninimum of 10 years of experience.
A baccitlaureato dcgrec may fulfill 4 years of experience.
i ALTER!:ATES
- 6. S.1. 3 Alternate tcebers may be appointed by the (!!P1,G Chairnen) to serve on a tcuporcry basisy cach appointr.'ont shall be in uriting.
ilo more than two altet nates shall participate on a voting basis in (NRAG) activitics at any one tirr.a.
11EETil:3 FP.EDUFNCY 6.S.1.4 The (NRA3) or a subcoanittee thereof shall ir.ect at least once per calendar quarter. The (hRAG) shall meet at least semi-annually.
.,QUORE1 6.S.1.5 A quoren of (NRAG) for revicu shall consist of the Chain:nn er his designdttd alternate end (tua) other w:mbers, or citernate uw.ters; houaver, a raajority of those present shall be regular cc:r.bers. ' The -
quoreu shtll have represcatution experienced in reactor operations cod radiction protection; houavcr the cporating staff shall not be a vM.ing raajority.
REVIEU 6.5.1.6 The (NRAG) shall revicu:
O a.
safety evaiention2 for 1) cnanoes to proccdurcs, c,uir=ent or systeus and 2) tests or expcrin:nts, conducted uithcut NT.C approval untr the provision of Section 50. 59,10 CFR,- to verify that such actions did not ccnstituto an unrovic.:cd safety qucstion.
b.
Proposed ~ch6nres to prccedures, equipracnt or systems that L
change tha original intent or use, and-arc 'non-conservative,,
l or those that involve an unrevicwed safety question as defintd in Section 50.59,10 Cf R.
L P
l' I
r-
i
)~
a-w
'i i
Proposed icsts or experir.:ents which are significcntly different.
c.
from previous approved tests or experiments, or those that involve an unrevic.:cd safety question as defined in Section j
50.59,10 CFR.
j d..
Proposed changes in Technical Specifications or licenses.
Violations of appliceble statutes, codes, regulations, orders, e.
Technical Specifications, license. requirements, or of internal
.l i
precedures or instructions having nuclear safety significcnce.
[
t f.
Significent operating ahnor:nalitics or deviations from norn.al l
and expected,perfor;;ance of facility equipent that affect l
nuclear safety.
r Events uhich have been reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the IRC.in
.)
9 writing.
e t
l h.
Audit reports.
AUDITS 6.5.1.7 Audits of facility activitics shril be perfor::nd under the cognizance of the (fiit?.G) but in no case by the personnel res;)cnsibic for i
t the iten audited.
Individual audits may be performed by one individr.1
~
who need not be en identificd (::RI.G) nuc.bor.
These audits snall excuine the operating records and encer. pass:
a.
The confen.2nce of facility operation to the Technical Specifica-tions and applicabic license conditions, at least.once per 12 i
raonths.
i O b.
The perforaance training and qualificaticns of the entire facility staff, at 1ccst once per 12 months.
.c.
The results of all actions taken to corecct deficiencies oc-curring in fccility equipmant, structures, syste:ns or mathcd of
[
operation that affect nacicar safety, at least once per G.tc 12.
j Donths.
d.
The facility Emergency Plan 'and imple:nanting. procedures, at-L least once per 24 months:
c.
The facility Scr.urity Plan and iuplc;:enting procedures, at-least once per 24 months.
-i f.
Any other area of facility operation considered appropriote by.
the (!!R,*S)-or the (f ccility Dircetar).
F I
l l
i
~-
V r
-a
.,n
-,y
-p.W- -....g.
.ww y.,
e-i e,,
y w
m.
3
+ gy y e-.-r w 1r9----gr 9
.4
.i L.
s AUTliORITY--
6.5.1.8 The (I;iMG) thall report to (! anage:nont at level above Facility,
Director) and advise the (Facility Director) on those areas of resp n-sibility specified in Sections 6.5.1.6 and 6.5.1.7.
RECORDS 6.5.1.9 Records of (l: RAG) activities shall be prepared and distributed as indicated below:
a.
Minutes of each (imAG) reeeting shall be prepared, and font;:r:'ed to the (Facility Director) within 30 days following each
- Idecling, b.
Reports of revicus encocpassed by Section 6.5.1.6 e, f, and g abaye, shall be preparcd and forwarded to the (Facility qV Director) i ithin 30 days following cen.pletion of the revicu.
c.
Audit re;; orts encompassed by Section 6.5.1.7 above, shall be forwarded to the (l:n.'.G Chair;.:an) and to the manage:cnt res; n-sible for tha areas audited within 30 days after com;)letion of
,the audit.
6.6 S5FETY Lli'.1T V10LATI.O.?:--
6.6.1 The following actions shall be tahen in the event a Safety Lir.:it is violated:
a.
The reactor will be shut'do.n imediately and recci.or operaticn will not be rc u :ed without authorizatica by the Cc:a:aission.
33
!'"/
b.
The Safety Linit violction shali be reported to the Director of the appropriate I:RC Regional Of fice of Inspection and Enforcemnt (or his desigt. ate), the (Facility Director) and to the (:: PAG) r.ot later than the next work day.
c.
A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the (l: RAG).
This repor's.shall describe (1) applict.ble circunstances preceding the violation, (2) cffects of.th: violation upon facility cergonents, syctn Or structures, and (3) corrcctive actica taken to preycat racurrcn:::.
d.
The Safety Limit Violation Report shall be sul::itted to the Cocaission, the (l;Pl,G) and the (Facility Director) within 11 days of the violation.
t I
7.- -..
A!
J.. L.
~,
i;'
'6.7 PROCEDURES
}
There shall be written operating procedures that cover the follcuing j
activitics. They shall be approved by the (facility. Director).
a.
Conduct of irradiations and experinents that could affect the operation or safety of the reactor.
1 b.
Startup, Operation, and Shutdo.;n of the Reactor.
c.
Fuel movement and changes to the core and experiments l
that can ef fect the reactivity.
d.
Preventive or correction maintenance which could have an effect on the safety of the reactor.
Surveillance, testing and calibration of instruments, com?cncr.ts i
c.
and systcr:s involving nuclear safety.
j O f.
Review and approval of changes to procedures.
g.
Personnel radiation protcction consistent with 10 CFR Part 20.
h.
Implementation of the Security Plan and Euergency Plan.
i.
Adrainistrative control of operation and maintenance.
e Though substantive ch:nges to the above proccdures shall 'be mMe only i
approval by the (Facility Director) te;nporary changes to the procci: ras that do not char.gc their original intent.ny 'be r.ade by 'the (Reactor Supervisor).
All such temporary changes shr.ll be documented, end'sub-4 j
sequently approved by the (Facility Director) within 14 days, i
i 6.8 EXPERIFI'lTS i O l
a.
Prior to initiating any new reactor experiment, e.g., cicss l
of experim:nts that could affect reactivi;.y of the reactor or result in' release of radioactive neterials, an experiment plan shall be prep:tred, rcvicued by (MRl5 or other identitied i
coarnittec), cnd approvcd by the (Reactor Supervisor).
b.
Ecch cr.periment plan shall (1) identify the type of experiment i
(previously a;iproved or rcceritly reviewcd pee 6.Ca), (2) i identify the cxperinanters and.(3)- have been approved by tbc licensed senior operator in charge of' reactor operation.
j l
4 j
g 4
3 7
4 0
c 4
s 4 -.
6.9 REPORT 1l:G RE0'J1RD'.D!TS In addition to the applicable reporting requirenents of Tit 1c 10, Coda of Federal Regulations, the following reports shall be submitted to tha Director of the appropriate l'RC Regional Office unicss othenvise noted.
G.9.1.
Routine Reports a.
Startup Renort.
A summary report of plant startup and powcr escalaMa testing shall be submitted following (1) receipt of an operating license, (2) amend:uent to the liccase involvina a planned increase in power level, (3) installation of fuel ti:at has a different design, and (4) modifications that may have significantly altered the nuclear, therm 1, or hydraulic-performanca of the plant. The report shall address each of the tests identified in the FS/2 and shi.11 in general includs a description of tha ceasured values of the operating'cor.ditier.s or characteristics obtained dering the tests program and a p'v-comparison of those values with design predictions and specific.a-c tions. Any corrective actions that tiere required to obtsin satisfactory operation shall also ba described.
Any additionci specific details required in license conditions based on cther co:aitments shall be incitded i_n this report.
Startup reports shall be subnitted within (1) 90 days folic.dng co:rpletion of the startup test progrua, (2) 90 days follcuinc resu: ption or cc:r.renccrant of power cparation, (3) 9 trenths 3
following initial criticality, which21er is earliest.
If tbc Startup Report does not cover all thice events (i.e. initial criticality, completion of stat,' tup test program, and resur.p-l tion or comm0nccccnt of pcuer operation), suppic::antary rc:.6 :s shall be sub.aitted at leas.t every three months until all thcce events have been completed.
~
b.'
An'aval 00erating_Renart.
Routine cperating reports coverir.g the operation or tPa unit durir.g the previous calendar yese should be submitted prior to (March 31) of each year.
The annual operating reperts r.tade by licenscos shall provide a i
cceprehensive su:c.12ry of the operating experience having thfe.y significance that was gained during-the year, even th:Urh sc:2 repetition of previcusly reported information may be involve?.
References in the ennual epcrating report to p'reviously sub.:itted reports shall be clear.
t i
o end
M s - -
w?
g 4
Each annual operating report shall include:
1 (f) ' A bric f narrative sur.raary c. f:
(a) Changes in facility' design, perfonrance characteristics, l
and ~ operating procedules related to reactor safety, that occurred during the reporting period.
t 5
.(h) l'esults of aajor surveillance tests and inspections.
(2) A tabulation showing the energy generated by the reactor a!
' (in (;cncral r e rMentbly tabulation in 1.ugat:ai.t-hours end/or -
i
. hours the reactor is operatinj uill be satisfactory).
j.
(3) List of tha'unschcduled shutdovins, including the reasons thercroce and corrective action tchen,-if any.
1 (4)
Discussion of the najt.r safety related corrective tairi-i tenance perfort.ed during the period, ir;cluding the effects.
if any, on the safe op;iration of-the reactor, anti ti.a reasons for ths corrective ir.aintenance required.
i (5) A brief description of:
j
- (a)
Each change to the facility to the extent that it l
changcs a description of the facility in the Safety l
/.nalysis Repart.
i,
-(b)' Changes to tbc procedures es described in the Safety l
Analysis Rcport.
(c) Any new er untrica e>.pvients or t5sts perfercod i-i during-the reporting period that are not dc<;cribed 1
in the Safety Aaalysis P.eport.
(6) A suuary of the safety evaluation nade for. each ch?nge, test, or experir.:ent not su!raitted f or Co:.a:lssion approval pursuant to 10 CF!! M.Si! which clearly shoc s the~ reason lecding to the concli:sion that no unrevie.cd safety
~
. question ~ existed-arid that no technical specificatiop--
change 1::s rcquired.
(7)" 'A sum: dry of the natur6 cirf c::.00nt of redioa'ctive effl':: ds
~
relecsed or disci:arg:d 'to the environs '!:oyend the efketir
~
' control of the licence as-dete:v.:ined at:criprior to the point of such release er discharge.
The s ecificaticns leir:t.ifind. for 'this.r.ect f or, i.h:u1N used to the ektc.t ;2rac'.ict ! cru.cid: fring t'e i,c,iure of '
the o f fl::,,ats,
',r.. evn il. ' ele. equip:t nt, ad the s'; Sci fi.- '
i fac ai'y.
F
+
b 4
.... = _,,
1 4
'. w s
(a) Liquid Maste_ (suira:arized en a 3 month basis).
(1) - Total estimated quantity of radioactivity :
~
released (in curies)and Total vol'une (in liters) j l
l of cffluent water (including dilvent) released.
(2) An estimation of the specific activity for each detectable radionuclide present if the specific natcrial after dilution activityofthereleasci)microcuries/cc.
is greater than 1-~x 10~
1 (3) Sumary of the total release in curies of coch nuclide determined in (2). above for the re; orting period based on representative isotopic analysis.
(4)
Estimated average concentretion of tha. released radioactive naterial at the point of releasc-for the reporting period in terms of raicrocuries/cc (N
and fraction of the applicabic ilpC.
r)
(b) Air _ borne Uaste (su=arized on a 3 month basis)
(1) Total estin:ated quantity of radioactivity released (in-curies) determined by an approved sampling and counting method.
(2) Tocal esticated quantity of Argon 41 released (in curics) during the reporting period hsed on date fred en appropriate monitoring systan.
(3) Estincted averaga atr.iospheric diluted concentraticn of Argon-41 released during the reporting period in terms of microcuries/cc and fracti0n of the applicable MPC value.
O (4) Total estimeted quantity of radioectivity in particulate form with half lives greater th3n; eight days (in curics) released during.the reporting period as deterained by an appropriate particulate monitoring system.
(5) An estimate of the average concentration of other significant radionuclides.prqsent in the gascous wasto discharge in torns of uicrocu. ics/cc and f raction of the applicable liPC:value.for the reportin.j paried if the estinated relcase is grester than 20:'.of -the applicable MPC.
F t
4 l
F
~
a:
'- w
^
4 e.
(c) Solid l'.61.e (sue.:arixed on an ant.ual basis)'
(1) Total amount of solid waste packaged (in cubic meters)
.(2) Total activity in solid naste (in curics)
(3) The dates of shipments and disposition (if shippedoffsite).
(C) ' A description of the results of any environtsental radiologic.1 surveys ri.rior. sed outside the facility.
~'
6.9.2 Pg ortable Oqcu tences Reporteblo tr:currences, including causes, prob.'ble consequences, correctiva act.icns and : aasures to prevent recurrance, shall be reportcd to the IUTC.
Supplcrantel rcports may be required to fully describe final resolution of the occurrcace.
In case of corrected or supplew ntal
_n V
yeports, an. -en:Si licensee event ecport shall be cc:apleted end referen::c-shall be asde to the original report date, e
a.
prc.: pip fificatinn !!ith 1.'rit'en Foll:wuo.
The typ?s of-cvents listed i;oion snall be ref.oric'J as expeditiously as possibic by teicphone and conTirand i>y teier.raph, milr:rce, or f acsiv.ile transmission to ti.e Director cf the ap;>ropriato i;RC l-Regional Of fice, or his desisa.te no later thcn the first work day folloniat. the event, with a *.:ritten follci.cp rr.p',rt-within tue..'ceks.
The writteo follo.!99 report sh!.ll includa, i
as a r.tinimum, a completed copy of a licensoc cvent icport
- form, inforr.mtion provided on the. licer.':ce event report fora -
l shall b2 supplcl.:nted, as ne:ded, by c.de:itional narre!.ive natcrial to 1.'rovidc cc::plete explanit. ion of the circumstenccs
. surrounding th: cvant.
~
O ci) railure or ti>e reactor i,rotecticn sy:4 or oth::r
. systcr.5 subject to limiting safety sys:/ n settints to _
l initiate the ret,uiresi protective functie cy ti.e tinc a-Monitored p;rm:ter rcnch95 the setpeint specified as the limiting safety syste:a scLting in i.he 19:.hnicci
+
specificatiens or failuia t.o comp!eM tha requircJ protective funci' ion.
I l
(2) ' Operation of the reactor or af fccicci :.ys! cms wito ar.y paremeter or ep&ation.:ebje:.t to e li:.itint; con:lition.
is less con:: rvative thra::the limiting conditit-n for i
oper0tica citablished ir, tha tc:.hr.ic I sp crifications wit.hnut ta!.ing percrittcd rewiial cctiivi, t
L t
D t
p-
-.e(--
,w,r
_~
,,,._,,,..,-e s.
-e.--.
y
- - +,
z,,6.-,
-..-....,-v
-v-,y y-
3
+
(3') Abnormal degradation discovered in a fission product Larrier, i.e., fuel cladding, reactor codlant boundary, or containcent.
(4) Reactivity balance anomalics involving:
(a) disagrecrcont between expected and actual critical positions of ap;)roxi:nately 0.3% A h/k; i
l (b) exceeding excess reactivity limit; l
I
.(c) shutdoun margin less conservative than specified in j
technical specifications; I(d) unexpected short-term reactivity changes that cause a period of 10 seconJs or less; (c) if sub-critical, an unplanned reactivity.insortion of more than approximately 0.55 A k/k or any unplanned criticality.
(5) Failure or malfunction of one-or more components which prevents or could prevent, by itself, the fulfillc.cnt of ti.e functional requiratents of system (s) used to cope with cccidents analyzed in the SAR.
(6) ~ Pershnnel error or precedural inadequacy which prevents, or could prevent, by itself, the fulfilluant cf the functional requirer.nnts of systcms required to cope with accidents analyzed in the SAR.
.(7) Unscheduled Conditions arising fro.n nstural or man-c dc events th:t, as a direct result of the event require reactor shutdot.n, operation of safety systems, or other protective ucatures required by technical specifications, j O (8) errors discovered in the transient or accident ana,yses or in-the methods used for such. analyses as describcd I
j in the safety analysis report or in the bases for. the
~
technicc1 specificrticns that have or could have f
. permitted reacter operation in a manner less conscrvative i
than assu:::2d in the analyses.
(9) Performance of structurcs, systems, o'r compori;.nts thst -
rcquires ren:edial action or corrective measures to prevent operation in a mar.nce less conscrvat.ive than assuced in 1.
?
l
?
l I
.h l
3;}
~
1
.^
v r
._~q the accident analyses in the safety analysis report or
. technical sp6cifications basos; or discovery during plant life of conditions not sp;cifically conridered in the
.l safety analysis ' report or technical :.pccification: that require rcacdfal action or corrective 1.:casures to prevent the er.istence or. developuent of an unsaf e condition. <
SPECI W 'ftEPORTS 1
Special reports may be required covering inspections, tests and minte-nance activitics.
These special reports are detcrmined on an individual.
basis for ecch facility and their preparation and sut.r.ittal are designated
' ~
iI in the Tochaical Specifications, i
~
6.9.3 Special reports shall be sub:iitted to the Director of the apprcpriate itRC Regional Office hithin the timo period specified for cach report.
6.10 RECORD 18.' TENT 10?;
O 6,10.1 hecords to be Retsined for a Period'o at least five years:
l'
,a i
(a) Operating 1 cgs or data which shell-identify:
1 i
(1) Completicn of pre-startup chsi:out, startup, power chens-s, and shutdmin of the' reactor.
(2)
Instellation or rc=0 val,0f' fuel elsents, control reds ce experi;r, cats that caufd Lifect core reactiyity.
~
(3)
Installation or r.emoval of'iumpers, shecial tags or notice:,
or othel? temporary changos 'to reactorj safety circuitry.
'~
(4) Rod worth neasurcments and'oth$r reactivity rocasur'O.:nts.
Q (b) Principal miintenance operations.
4 x
(c)
Reportable occcrrenc.es.
m
\\
(d) Surveillance 'activitics reqGired by' technical specifications.
(e)
Facility radiation and contcaination su,rveys.
1 j
s 4
6 b
g F
l t
a l
I pr
' +
n 4
A
'hrt 7;
e,
.:.g..
/-
~
2;
~
e
-4 (f) Experiments performed with the reactor.
This requirement may be satisfied by the nornal operations log booh '
- plus, (1) records of radioactive material transferred from the facility as required by license.
(2)
Records required by the (!! RAG) for the performance of new or special experiments.
(g)
Changes to operating procedurcs.
6.10.2 Records to be Retained for the life of the Facility.
(a) Gascous and liquid radicactive effluents released to the environs.
(v3 (b) Appropriate off-site environ!:. ental monitcring surveys.
(c)
Fuel inventories and fuel transfers.
(d) Radiation exposures for all personnel.
(e) Updated as-built drawings of the facility.
(f) Recor6 cf transient er cperational cyclc.s for thosc cc::;cncnts designed for c limited nu:r. hec of transients or cycles.
(g)
Records of training cnd qualification for n:cmbers of the facility staff.
(h) Records of reviews performed for changes made to procedures or g) equipa3nt or revicws of tests and experiments pursuant to 10 y
CFR 50.59.
3 (i)
Records of meetings of the (I RAG).
~
Jh J
V