ML20212D999

Proposed Tech Specs Identifying Proper Communication Channels for Reporting Requirements of Section 6.7 & Correcting Typo in Section 6.5(b)

ML20212D999
Person / Time
Site:	05000113
Issue date:	10/22/1997
From:	ARIZONA, UNIV. OF, TUCSON, AZ
To:
Shared Package
ML20212D991	List: ML20212D987 ML20212D999
References
NUDOCS 9711030015
Download: ML20212D999 (45)

Text

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.g a , page 30 6.5 Action to be Taken in the Event of a Reportable Occurrence '

in the event of a Reportable Occurrence, the following action shall be taken: '

a. The Reactor' Laboratory Director shall be notified and corrective action taken prior to resumption of the operation involved.

b. - A report shall be m6de which shallinclude an analysis of the cause of the

- occurrence,~;"kkray eficacy of corrective action and recommendations for measures to prevent or reduce the probability of reoccurrence. This report'shall be, submitted to the Reactor Committee for review.-

- c ' A report shall be submitted to the NRC in accordance w:th Section 6.7 of these specifications.

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6.7 Reporting Requirements.

in addition to the requirements of applicable regulations, and in no way substituting therefor,9 ports shall be made to the NRC as follows:

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a.L A report within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and telegraph or telefac (FAA) to the USNRC R:ibn "I Citi;G Non-Pcwer Reactors andDecommissioning Project Directorate of;

1. Any accidental off site release of radioactivity above limits pemiitted by 10 CFR m 20, whether or not the release resulted in property damage, personal injury, or exposure; .

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-2.. Any violation of a Safety Umit; and'

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[ y - of these specifications in writing.

. b. A written report within ten days to the U. S. Nuclear Regulatory Commission, Attn:

Document Control Desk, Washington D.C. 20555, with a copy to the Director, Ohisbn ci n;;; tor Cri;ti ;rd Pro}e;t3, Region YNon-Power Reactors aml

. DecommissioningProject Directorate, of :

1. Any significant variation of measured values from a corresponding predicted Lvalue _of previously measured value of safety connected operating

. characteristics occurring during operation of the reactor;

2. - Incidents or conditions relating to operation of the facility which prevented or L could have prevented the perfom1ance of engineered safety features as described in these specifications; c

. 3h Any reportable occurrences as defined in Section 1.0 of these specifications; and:

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4. Any violation of a Safety Limit.

. 5.' ~ Any accidental off site release of radioactivity above limits permitted by 10 CFR

,20, whether or not the release resulted in property damage, personalinjury, or exposure.

c. A written report within 30 days to the U.S. Nuclear Regulatory Commission, Attn:

Document Control Desk, Washington D.C. 20555, with a copy to the Director, O'visica cf Riactor Sof;ti and IrojGcts, Region IVNon-Power Reactors and q Decommissioning Project Directorate, of; wanes Uruversdy of Arizona Research Reactor -

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> 1. Any substantial variance from performance specifications contained in these

1. specifications or in the Safety Analysis Report;

~ 2. : Any significant change in the transient or accident analysis as described in the

Safety Analysis Report;-

3. f Any changes in facility organization; and 4.- -

Any bbserved inadequacies in the implementation of administrative or 2 orocedural controls! ,

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$d. A written report within 60 days after completion of startup testing of the reactor ' to

~ r the U.-S. Nuclear Regulatory Commission, Attn: Document Control Desk,

. #* ( Washington D.C. 20555, with a copy to the Oh;ter, Ordska of R::;tec Citj nd

[ 7;&l;;;;, n;ibe, ?l Non-Power Reactors andDecommissioning Project Directorate.

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.1.c An evaluation of facility performance to date in comparison with design g r,iradictions and specifications; and .

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2. . A reassessment of the safety analysis submitted with the license application in b = -

-_li3 ht of measured operating characteristics when such measurements indicate p y; sthat there may be substantial variance from prior analysis. .

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' e.f A wntten annual report within 60 days following the 30th of June each year to the

U.S. Nuclear Regulatory Commission, Attn: Document Control Desk, Washington D.

C.40555; with a copy to the Oh;ter, OPJsba of n;acter S;f;tj and Ic;jects,

- . n;ibe, ti Non-Power Reactors and Decommissioning Project Directorate. .

il. ? A brief narrative summary of (1) opeiating experience (including experiments performed), (2) changes in facility design, perfonnance characteristics, and operating procedures related to reactor safety and occuning during the lj  : reporting period, and (3) results of sur.eillance tests and inspections;

1. " 2.' : Tabulation of the energy output (in megawatt days) of the reactor, amount of

. pulse operation, hours reactor was critical, and the cumulative total energy.

output since initial criticality; .

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3. The number of emergency shutdowns and inadvertent scrams, including Treasons therefore;-

t4.J Discussion of the major maintenance operations performed during the period, including the effect, if any, on the safety of the operation of the reactor, and the -

reasons for any corrective maintenance required; 4

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L6.5 Action 16 be laken in the Event of a Reportable Occurrence -

'" In the event of a Reportable Oc'currence, the following action shall be taken.

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- a.= ' The Reactor Laboratory Director shall be notified and corrective action taken prior to resumption of the operation involved.

b. LA report shall be made'which shallinclude an analysis of the cause of the occurrence, efficacy of corrective action and recommendations for measures to

-: prevent or reduce the probability of reoccurrence. This report shall be submitted to

- 1: the Reactor Committee for review.--

. c. LA report shall be submitted to the NRC in accordance with Section 6.7 of these

' specifications. .

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lIn addition to the requirements of applicable regulations, and in no way substituting g -

therefor, reports shall be made to the NRC as follows:

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O A' report within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and telegraph or telefax (FAX) to the Non.

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, } Power Reactors and Decommissioning Project Directorate of:

t: il' f Any accidental off site release of radioactivity above limits' permitted by 10 CFR 7 % ,3 - m. J 20, whether or not the release resulted in property damage, personal injury, or y Ni - e , _ exposure;v ,

1. m M k2 Mans v'olation i of a Safety'Umit; aridi g -

p,m (3E Any reportable occurrences as defined in Section 1.0 (Reportable Occurrence)

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. , Lof these specifications in writing.L g M< xb. i Alwntten report within ten days to the U. S. Nuclear Regulatory Commission, Attn:

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L Document Control Desk, Washington D.C. 20555, with a copy to the Non. Power m _

L Reactors and Decommissiori.ng Project Directorate, of :

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TIR Any significant variation of measured values from a corresponding predicted

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- value of previously measured value of safety. connected operating characteristics occuning during operation of the reactor; t 2.finbidents or conditions relating to operation of the facility which prevented or could havs prevented the performance of engineered safety features as

?describedin these specifications; bJ f3. I Any reportable occurrences as' defined in Section 1.0 of these specifications;

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, 6e (24. Any violation of a Safety Limit.'

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51 [Any accidental off site release of radioactivity above limits permitted by 10 CRt g*  ; 20, whether or not the release resulted in property damage, personal injury, or pgr . exposure.:

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c. i A written report within 30 days to the U.S. Nuclear Regulatory Commission, Attn:

Document Control Desk, Washington D.C. 20555, with a copy to the Non. Power

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Reactors and Decommissioning Project Directorate, of:

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. 1 specifications orin the Safety Analysis Report; wnia UrhersQofMaone Research Reactor a

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2. Any significant change in the transient or accident analysis as described in the

. . Safety Analysis Report; .

l L 3. : Any changes in facility organization; and '

4.1 Any observed inadequacies in the implementation of administrative or

procedural controls.

.d. . A written report within 60 days after completion of startup testing of the reactor to the U. S. Nuclear Regulatory Commission, Atin: Document Control Desk:

. Washington D.C. 20555, with a copy to the Non Power Reactors and Decommissioning Project Directorate.

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, li An evaluation of facility performance to date in comparison with design predictions and specifications; and

2. - A reassessment of the safety analysis submitted with the license application in

? light of measured operating characteristics when such measurements indicate Ethat there may be substantial variance from prior analysis, y Le.' A written annual report within 60 days following the 30th of June each year to the

.. U.S; Nuclear Regulatory Commission, Attn: Document Contro; Desk, Washington D.

C. 20555, with a copy to the Non Power Reactors and Decommissioning Project i Directorate.

1. , A brief narrative summary of (1) operating experience (including experiments performed), (2) changes in facility design, performance characteristics, and

operating procedures related to reactor safety and occurring during the 79 reporting period, and (3) results of surveillance tests and inspections; (2. Tabulation of the energy output (in megawatt days) of the reactor, amount of pulse operation, hours reactor was critical, and the cumulative total energy output since initial criticality;

3. The number of emergency shutdowns and inadvertent scrams, including F reasons therefore; (4. ' Discussion of the major maintenance operations performed during the period, including the effect, if any, on the safety of the operation of the reactor, and the reasons for any corrective maintenance required;

~ 5. I A brief description including a summary of the safety evaluations of changes in the facility orin procedures and of tests and experiments carried out pursuant to Section 50.59 of 10 CFR Part 50; i Urhersity of Aritone Research Reactor

-imamen .m

Y TECHNICAL SPECIFICATIONS

,: UNIVERSITY OF ARIZONA RESEARCH REACTOR 6

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- TABLE OF CONTENTS s &~ page

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,f . 1.0 DEFINITIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1 . . . . . . .

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• e 42.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS - . . . . . . . . . . . . . . 5.

e s, t2.1 Safety Limit < Reactor Power Level '. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 C2.2 Limiting Safety System Setting . Steady State Reactor Power Level . . . . . . . . . 6 JU 2.3 Limiting Safety System Setting Pulse Mode Reactor Power Level . . . . . . . . . 7 M

E3.0 LIMITING CONDITIONS FOR OPERATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8

13.1 Reactivity Limds . . . . . . . . . . . . . .. ....:.............................. 8 g

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.T32 High Power 0peration . . . . . . . . . < . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .10

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P 3.3 Pulse Operation . . . . . . . . . .x . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 13.4 Reactor Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12

" (3.5 Reactor Safety System L . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 -

g + 3.6 Ventilata System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

j M 3.7. Expe:iments J. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .16 .

. . . . . . . . . . , 15 D s n V,

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• 40 SLANEILLANCE REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 ......

w 14.1 Fuel l. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 p# e '

' 4 2 Control Rods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 14.3 Reactor Safety System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 14.4 Radiation Monitoring Equipment . . . -. . . . . . . . . . . . . . . . . . . .-. . . . . . . . . . . .

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- 4.5 Maintenance , . . '. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 3 J 4.6 Pool Water Conductuty ' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 w

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l 5.0 DESIGN FEATURES ' . . . . . . . . . . . . . . . . .- . . . . . . . . . . . . . . . . . . . 23 5.1 Reactor Fuel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 15 2 Reactor Buddmg . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 24 5.3 Fuel Storage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 5 4 16.0 ADMINISTRATIVE CONTROLS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26

,  : 6.1 Organization ! . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.6 x *, 16 2 Review - . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .27 . . ...

, L 6.3 Operationt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28

, a L a. - Operating Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28

', . _ . b. ALARA Program . e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 6.4 Action to be Taken in the Event a Safety Limit is Exceeded . . . . . . . . . . . . . . . 29

• !6.5' Action to be Taken in the Event of a Reportable Occurrence . . . . . . . . . . . . . . 30 L 6.6 Plant 0perating Records . . . -. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31

6.7 Reporting Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32

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6.8 Review of Experiments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 s

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Amendment 17 1

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1.0 DEFNITIONS

u 4% Channel 4 A channel is a combination of sensors, electronic circuits, and output devices

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connected by the appropriate communications network in order to measure and display the value of a parameter.

v Char nel Cahbration i A channel' calibration is an adjustment of a channel such that its output

corresponds with acceptable accuracy to known values of the parameter which the channel i

'measuresh Cahbration shall encompass the entire channel, including equipment, actuation, )

., alarm,'or trip and shall include a Channel Test.L

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Channel Check A channel check is a qualitative verification of acceptable performance by i W'y. observabon of channel behavior. The verification shallinclude comparison of the channel l

' 4 output with previous ' readings or performance or with other independent channels or

systems measurirgthe same variable, wt.enever possible.

k . Channel Test A channel test is the introduction of a signal into the channel for verification l[* '

7that it is operable..-

Cold Critical The reactor is in the cold cribcal condition when it is critical with the fuel and

bulk _ water temperatures the same (-20*C).

Experiment An experiment is any device or material, not normally part of the u .ctor, which

, is introduced into the reactor for the purpose of exposure to radiation, or any operation iwhkhLis designed to investigate non routine reactor characteristics.-

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Expanmental Facihties Experimental facilities are the thermal column, pneumatic transfer J systems, central thimble, rotary specimen rack,-beam tube, and the in core facilities.

,w 4* Limiting Conditions for Operation Limiting Conditions for Operation (LCO) are -

administratively estabhshed constraints on equipment and operational characteristics which

-p Eshall be adhered to during operation of the reactor.

Limitino Rafety System Settino A RSS) The LSSS_is the actuating level for automatic

, Av; ctive devices'related to those variables having significant safdy functions.

, J Manual Mode The' reactor is in the manual mode when the reactor mode selection switch is a lin the. manual or automatic position. : In this mode, reactor power is held constant or is

,m changed on periods of approximately one second orlonger.

Measured Value i The Measured Value is the value of a parameter as it appears on the output of a channel.

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page 2 Movable Exoeriment An experiment is movable when it is intended that all or part of the experiment may be moved in or near the core or into and out of the reactor while the reactor

-is operating.

Operable Operable means a component or system is capable of performing its intended function. -

Operating Operating means a component or system is performing its intended function.

Pulse Mode The reactoris in the pulse mode when the reactor mode selection switch is in the pulse position. In this mode, reactor power may be increased on periods less than one second by motion of the transient control rod.

Reactivity Worth of an Exoeriment The reactivity worth of an experiment is the maximum value of the reactivity change that would occur as a iesult of planned changes or credible

, malfunctions that alter experiment position or configuration.

Reactor Committee The group of persons at the University who are assigned responsibility for review and audit of facility operation and review of changes and experiments in accordance with 10 CFR 50.59.

Reactor Ooerating . The reactor is operating whenever it is not secured or shutdown.

Reactor Safety Svstems . Reactor Safety Systems are those systems, including associated

- input chonnels, which are designed to initiate automatic reactor protection or to provide information for initiation of manual protective action.

Reactor Secured . The reactor is secured when:

a. it contains insufficient fissile material or moderator present in the reactor, adjacent experiments or control rods, to attain criticality under optimum available conditions of moderation and reflection, or

b. fl. The minimum number of neutron absort>ing control rods are fully inserted or

. other safety devices are in shutdown position, as required by technical specifications, and

2. The console key switch is in the off position and the key is removed from the lock, and

3. No work is in progress involving core fuel, core structure, installed control rods,

- or control rod drives unless they are physically decoupled from the control rods, and naam Unkers6ty of Arinone Research Reactor a Amendment 17

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! 4. :: No experiments in or near the reactor are being moved or serviced that have, Eon movement, a reactivity worth of one dollar or more. .

Reactor Shutdown The reactor is in a shutdown condition when sufficient control rods are inserted to assure that itis subcritical by at least $1.00 of reactivity.

Reportable Occurrence . A Reportable (kmrrence is any of the following which occurs

during reactor operation

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a. l Operation with actual safety system settings for required systems less conservative

f. ;than the limiting safety system settings specified in Technical Specification 2.2.

bD Operation in violation of hmeting conditions for operation established in the Technical

- Specifications,-

c. J A' reactor safety system component malfunction which renders or could render the reactor safety system incyuda of performing its intended safety function unless L the malfunction or condition is discovered during maintenance tests or periods of reactor shutdown.-

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d> "Any unanticipated or uncontrolled change in reactivity greater than one dollar.

. e. Abnormal and significant degradation in reactor fuel, cladding, or coolant boundary .

/which could result in exceeding of prescribed radiation exposure or release limits.

4 Lf. i An observed inadequacy in the implementation of either administrative or procedurt' controls which could result in operation of the reactor outside the limiting conditions f for operation, g.iRelease of radioactuty from the site above limits specified in 10CFR20.

Control Rod . A control rod is a device fabricated from neutron absorbing material or fuel

& which is used to establish neutron flux changes and to compensate for routine reactivity

@ losses. A control rod may be coupled to its drive 'unit aliowing it to perform a safety function

• - when_the coupling is disengaged. .

Transient Rod . The transient rod is a control rod with scram capabilities that is capable of

.< f providing rapid reactuty insertion to produce a pulse.

m Safety Limit A Safety Umit is'a limit on an important process variwie which is found to be inecessary to reasonably protect the integrity of certain of the physical barriers which guard

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, ; against the uncontrolled release of radioactivity. The principal physical barrier is the fuel element cladding. -

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page 4 Secured Expenment A Secured Experiment is any experiment, experimental facility, or component of an experiment that is held in a stationary position' relative to the reactor by f mechanical means. The restraining forces must be substantially greater than those to which l

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the experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces which "are normal to the operating environment of the experiment, or by forces which can arise as

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l the result of credible malfunctions. _ ,

y Shall. Should. and Mav The word "shall"is used to denote a requirement, the word "should" denotes a recommendation, and the word "may" denotes permission, neither a requirement

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1 nor a recommendation.-

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Shutdown Margin Shutdown Margin is the reactivity existing when the most reactive control

rod is fully withdrawn from the core and the other control rods are fully inserted into the

. Cort. -

Time interval J Specification of the time between successive events as follows:

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, a. - Biennially f at two year intervals Gnterval not to exceed 30 months)

L b. Annually at one yearintervals Onterval not to exceed 15 months)-

c. l Semiannually at 6 month intervals Onterval not to exceed seven and one half

months)

c. ji: Quarterly at 3 month intervals 6nterval not to exceed four months) 1

e. Monthly at one monta intervals Onterval not to exceed six weeks) c f.'- --Weekly at'seven day intervals Gnterval not to exceed ten days) .

1 E g.; Daily 2 (must be done during the calendar day)

.1ktried.Exceriment An untried experiment is any experiment not previously performed in

': this reactor.

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Amendment 17

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t page 5 f2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1. Safety Limit Reactor Power Level

. Applicabildy

, This_ specification apples to the steady. state reactor power level.

Objective

. L The objective is to define a power level below which it can be predicted with confidence

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that no damage to the fuel elements will occur. - '

Specincation . _

7 The power level of the reactor in the steady. state mode shall not exceed 1000 kW under any conditions of operation.

Basis L  : It has been shown by expensive experience that operation of TRIGA reactors at a power

!-  : level of .1000 kW will not result in damage to the fuel. Reactors of this type have -

t ) operated successfully for decades at power levels in the range from 1000 to 1500 kW

- and above. The criterion for assuring the liitegrity of a TRIGA fuel element is that the I fuel temperature be maintained below 1000*C. It has been shown by analysis and by

i. & ' measurements on other TRIGA reactors that a power level of 1000 kW corresponds to a L . peak fuel temperature of approximately 400*C. Thus, a Safety Limit on power level of L- (1000 kW pmvides an ample margin of safety for operation.

4' Amendment 17

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12.2 Limiting Safety System SeTngi Steady State Reactor Power Level

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! This specification applies to the reactor power level safety system setting for steady

,yg?, state operation. '

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a eThe objective is to assure that the Safety Limit is not exceeded.

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4y SnarJr.ation _ ..

y c e , '.The110 setting for the power level scram in steady state operation shall be no grea 7N 3 kW. .

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Calculations and measurements show that at 110 kW, the peak fuel temperature in the j  : core will be less than approximately 150*C which is well below the safety criterion of fy A l1000:C and provides an ample safety margin to accommodate errors in measurement V' and anticipated operational transients.

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page 7 o"' 2.3 Limiting Safety System Setting;. Pulse Mode Reactor Power Level Applicabihtv . . .

(This specification app!ies to the reactor power level safety system setting for pulse

,; , mode operation.>

7 Objectiva tThe objective is to assure that the fuel temperature specified by the Safety Limit is not

, - exceeded in pulse mode operation.

I, ' Specification _._..

f ,The setting for the peak power level scram in pulse mode operation shall be no greater

! than 1100 MW.

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Basis.

y" Calculations and measurements show that at a peak power of 1100 MW in pulse mode

" operation, the peak fuel temperature in the core will be less than appmximately 400 *C, iThis_ provides an ample safety margin to accommodate errors in measurements and anticipated operational transients, m

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,.Si j_ < j 3.0 LMITING CONDITIONS FOR OPERATION L

' 3.1_ React l#j Limits yApplicabikty

? These specificatens apply to the reactivity condition of the reactor, and the reactivity worths of control rods and experiments, and apply for all modes of reactor operation.

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LThe objective is to assure that the reactor can be shut down at all times and to assure

that the safety limit will not be exceeded.

y Spacincations LThe reactor shall not be operated unless the following conditions exist:

n aNThe shutdown margin refened to the cold xenon free condition is greater than

' 80.50.with the highest worth rod fully withdrawn and with the highest worth

? , I non. secured experiment in its most positive reactivity state.

1. VIAny experiment with a reactuty worth greater than $1.00 is secured so as to prevent unplanned reactivity removal from or insertion into the reactor;

[2 J cNThe reactuty available to be inserted by the pu!r M is determined and is limited h Lby a _mecharucal block to'a maximum of $2.50.

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[d.J The reactuty worth'of an individual experiment 1.5 not more than $3.00; -

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!e.JThe total 6f the absolute values of_the reactivity worth of all experiments in the Lreactoris.less than $5.00; '

N- .

, e f. _ A ramp or_ oscillahng rod placed in the reactor cannot add more than $1.50 of s

1 reactivity; y

g!?The drop time of each standard control rod from the fully withdrawn position to 90

percent of full reactivity insertion is less than one second; and-na ht The neutron count rate on the startup channelis greater than one count per second.

[* 1 iThe maximum' reactivity insertion rate by control rods for non pulsed operation is 7

p less than _$0.20/second.1

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- Umantemome? M0fh M M so.7 t

Amendment 17

page 9 Basta .

The shut down margin required by specification 3.la is necessary so that the reactor can be shutdown from any operating condition and remains shutdown after cooldown

'and xenon decay even if one control rod should stick in the fully withdrawn position.

- Specification 3 lb is based on pulse measurements and analysis at the University of Arizona which indicate that as much as $3.00 resctivity could be inserted without increasing fuel temperature by more than 415'C. By restricting each non. secured experiment to a reactivity worth of one dollar, an ample margin is provided.

Specifications 3.1c through 3.lf are intended to provide additional margins between t". _.' those values of reactuty changes encountered during the course of operations involving experiments and those values of reactivity which, if exceeded, might cause a safety limit

^

= o be exceeded.

Specification 3.lg is intended to assure prompt shutdown of the reactor in the event a scram signalis rec 6ed.

Specification 3.1h is intended to assure that sufficient neutrons are available in the core to provide a signal at the output of the startup channel during approaches to criticality, i:

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\TMinittle 850 1047 Amendment 17

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page 10 3.2 High Power Operation.

Applicability _

? This specifcation applies to' operation of the reactor at high steady state power.

Objective The objective is to prevent inadvertent' pulse operation of the reactor while it is at a high

powerlevel, f ,

)

Soecificahon The reactor shall not be operated in the steady state mode at powerlevels above 10 kW L unless, in addition to the conditions of Section 3.1, the transient rod is fully withdrawn.

Basis .. - .

This specification is intended to prevent inadvertent pulse operation when the fuel temperature is above 50*C (corresponding to a power level of 10 kW) as measured in the B ring.L See Specification 3.3b.

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Amendment 17

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. page 11

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l3.3 Pulse 0peration Aoplicability

. ' The specifications apply to operation of the reactor in the pulse mode.

Obbctive

' ~

The objective is to prevent the fuel temperature safety !imit from being exceeded during pulse mode operation.

c

, w Specifications The reactor shall not be operated in the pulse mode unless, in addition to the requirements of Section 3.1, the following conditions exist:

a. ; The transient rod is set such that the reactivity worth upon withdrawal is not greater

. than $2.50; and O b.3 The temperature of the fuelimmediately priorto the pulse is essentially in equilibrium with the bulk water temperature. This is controlled by limiting the

' reactor power priorto pulsing..

Basis The specification 3.3a will maintain the maximum temperature of the fuel after a $2.50

. pulse belo'w 400*C above the bulk pool temperature, and thus well below the 1000*C g fuel safety criterion.

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i q;G muratsmossa M0fh M M too7 Amendment 17

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_ , page 12 13.4 Reactorinstrumentation m .

j This specification applies td the information which must be available to the reactor operator during reactor operation.-

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-: The objective is to require thM sufficient information is available to the operator to (assure safe operation of the reactor.

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The reactor shall not be operated unless the measuring channels described in the -

m , following table are operable and the information is available in the control room:

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Gi . MINIMUM. (OPERATING MODE:-

• > MEASURNGCHANIEL .

NUMBERL blN WHICHi ~

.ww WW "

• (OPERABLEi  :!REQUIREDT

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1 reactorpowerlevel(linear) l' steady state -

f' , wide range log powerlevel '

I steady state s

(startup count rate)

I.

(.. J reactor powerlevel0iigh range) l' pulse mode reactortank watertemperature 1- all modes

, area radiation monitors <2 ,

all modes

~

parbculate air radiation monitor-- 1 all modes -

r reactorwater activity monitor -1 - all modes

$5 s

i The neutron detectors assure that measurements of the reactor power level are L adequately covered at both low and high ranges.

1The radiabon monitors provide information to operating personnel of radiation above a

preset level so that there will be sufficient time to evacuate the facility or take action to 1 prevent the release of radioactivity to the surroundings.

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Amendment 17

4 page 13 F3,5 Reactor Safety System =

Anplicabilh This specification applies to the reactor safety system channels and interlocks. '

l iThe objective is to require the minimum number of reactor safety system channels and

~

interlocks that must be operable in order to assure that the safety limits are not T texceeded.

Specification "

Thec reactor shall not be operated unless the safety system channels and interlocks

! described in the following tables are operable.

urg / 1 IEASLRING ' . MINIMI.N NLNBER# FUNC g ; iOPERATING MODE IN?

+

i s M CHAfGEL '.e ? OPERABLE

'd WHICHREQUIREDt

, rextor powerlevel' 2L scram steady Wte m __

peak reactor power - - 1 ~ scram . pulse mode mancal scram : 1: scram all modes - i.

'I de 'M';' NMERM 4 s c N . . 3 OPERATING MODE:

W af@WTERLOCK ' NUMBER ,

FUNCTION. 4. \ flNWHICHC

' gn.W 96 ' ,

OPERABLE "

P. %n -

1REQtJREDT

.. startup Count rate; prevent Control rod interlock L withdrawalwhen neutron-I reactor startup count rate is less than .

,g. .1 per second transient rod preventwithdrawalof a interlock / transient rod when its steady state shock absorberanvilis not N fullyinserted -

c prevent simultaneous 4

simultaneous control rod withdrawal- 1. manualwithdrawalof two all modes w prohibit interlock : control rods -

e

, reactor powerlevell .

preventtransient rod interiock: -1: withdrawalwhen poweris pulse mode

> 10 kW -

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page 14 BASft1 The power level scrams are provided in all modes of operation as protection against abnormally high fuel temperatures and to assure that the reactor operation stays within  !

_ the licensed limits. The manual scram allows the operator to shut down the system if an unsafe or abnomial condition occurs.

The interlocks which prevent the withdrawal of the transient rod in the steady state mode l' and when the power level is greater than 10 kW prevent inadvertent pulses. The interlock to prevent startup of the reactor with less than one neutron per second indicated on the startup channel assures that sufficient neutrons are available to provide indicators on the measuring channels and to provide negative reactivity feedback through the fuel temperature coefficient of reactivity.

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page 15 m

= 3.6 Ventilation System mn ,

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, 3 - 4 This specdiccion apples to the operation of the reactor facility ventilation system.

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. , )The objectim is to assure that the venblation system is in operation to mitigate the consequences of the pcssible release of radioactive materials resulting from reactor opersbon.;

,4y; , . _ - - -

Specincation . . _. . .

'* 1The reactor shall not be operated unless the facility ventilation system is in operation except for periods of time not to exceed two days to permit repairs to the system.

. W Durire such periods of repair::

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.:a.1 The reactor shall not be operated at power levels above 10 kW and; i37

, ::bi The reactor shainot be operated with experiments in place whose failure could i result in the release ~of radioactive gases or aerosols.-

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iE Basis . . ..

n It is shown in The~ Safety Analysis Report that operation of the ventilation system reduces l doses due to argon 41lduring operation of the pneumatic transfer system, and also in p^2 Jihe event of a TRIGA fuel element failure. The specifications goveming operation of the i reactor whde the venblation system is ' undergoing repairs limit the generation of f T argon 41 and also' reduce the probability of fuel element failure during such times.

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Amendment 17 _ _ _ _

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page 16 i

? 3.7 Experiments M8bb .

! These specifications apply to experiments installed in the reactor and its experimental

facilities.- 6 w

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- (The objective is to prevent damage to the reactor or excessive release of radioactive s materials in the event of an experiment failure.

K ~

Spec fications .

The. reactor shall'not be operated un!ess the following conditions exist:

a. : fueled experiments shall be limited such that the total inventory of iodine isotopes

' 131 through 135 in the experiment is not greater than 1.5 curies and the strontium 90 inventory is not greater than 5 millicuries;

b. .txperiments containing materials corrosive to reactor components, compounds v . highly reactive with water, potentially explosive materials, or liquid fissionable

materials 'shall be doubly encapsulated.

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~ The limits of Specification 3.7a prevent the dose in unrestricted areas resulting from experiment failure from exceeding 10 CFR Part 20 limits. Specification 3.7b is intended Lto reduce the probability of damage to reactor components resulting from experiment .

t failure.

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, 7 mum m uosse - MOfh M M ios7 Amendment 17

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.o page 17 4.0 SURVEILLANCE REQUIREMENTS 4.1 Fuel Anolicabilrty

- This speciscation applies to the surveillance requirements for the fuel elements.

0 Obsciin l

< h The object've is to assure that the dirrensions of the fuel elements remain within

" acceptablelimits.

, specifications

- a. All fuel elements shall be removed from the core and visually inspectv for evidence oMeterioration of cladding, Oncluding at least corrosion, erosion, wear, cracking, ud weld integrity) at least once every five years.

b. The standard fuel elements shall be measured for length and bend at intervals separated by not more than 500 pulses of magnitude greater than $2.00 of reactivh.

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c. A' fuel element inJicating an elongation greater than 1/4 inch over its original length

( or a lateral bending greater than 1/1o inch over its original bending shall be considered to be d* ;.oged and shall not be used in the core for further operation.

d. Fuel elements in the B. and C. rings shall be measured for possible distortion in the event that there is indication that the Limiting Safety System Settings may have g

been exceeded, aazis L The most severe stresses induced in the fuel elements result from pulse operation with high reactmty input, during which differential expansion between the fuel and the cladding occurs and the pressure of the gases within the elements increases sharply. The above limits on the allowable distortion of a fuel element correspond to strains that are considerably lower than the strain expected to cause rupture of a fuel element.

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Amendment 17

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page 18 1

..4.2 Control Rods Annlicabatv This specincation applies to the surveillance reouirements for the control rods.

The objective is to assure the operability of the control rods. ,

lesC#iGalist l

a. The reactivity worth of each control rod shal! be determined annually.  ;

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m . b.s Control rod drop times shall be determined annu.Ily and after disassembly and i

' E reassembly of control rod drives or removal of control elements.

Lc. .- The control rods shaN be visually inspected for deterioration biennially.

id. On each day that pulse mode operation of the reactor is planned, a functional

?. performance pulse mode operation. check of the transient (pulse) rod system shall be per ,

e.; Semiannually, the transient (pulse) rod drive cylinder and the associated air supply e_ y  ; system sliall be inspected, ricaned aid lubricated as necessary.

%s ' fl The' maximum control rod reactivity insertion rates shall be determined annually. .

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• 9.; The reactivity worth of the control rods is measured to assure that the required ,

shutdown margin is available and to provide a means for determining the reactivity worths of experiments inserted in the core. - The visual inspection of the control rods and  :

e. measurement of their drop times are made to determine whether the control rods are - '

x capable of performing their functions property.

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a page 19 4.3 Reactor Safety System ADolicability The specification applies to the surveillance requirements for the measuring channels of the reactor safety system.

I ob>ctive The objective is to assure that the safety system will remain operable and will prevent the fuel tamperature safety limit from being exceeded.

Specification

a. A channel test of each of the reactor safety system channels required in the a.. oporating mode to be followed shall be performed prior to each day's operation or prior to each operation extending more than one day,

b. A channel check of the power level measuring channels required in the operating mode to be followed shall be performed daily whenever the reactor is in operation. ,

c3 A channel calibration by the calorimetric method shall be performed for the reactor power level measuring channels annually.

g The daily tests and channel checks will assure that the safety channels are operable.

The annual calibration and verifications will permit any long term drift of the channels to be corrected. .

4 .

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..-._._ _ _ - - _ _ ^"?!dmont "

page 20 4.4 Radiation Monitoring Equipment M

This specification applies to the radiation monitoring equipment required by Section 3.4 of these specircations.

Obiedha - .

The objective is to assure that the radiation monitoring equipment is operating and to

. verify the appropriate alarm settings.

Specife.ation

a. The alarm set points for the radiation monitoring instrumentation shall be verified i prior to each day's run.

L L b The radiation monitoring equipment shall be calibrated annually.

L BaslE Verification of the alarm set points of radiation monitoring instrumentation will assure that sufficient information to provide protection against radiation exposure is available, i-

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.r unamit e see MOfh M M e.,ios7 Amendment 17 j

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page 21 4

4.5 Maintenance Applicability This specification applies to the surveillance requirements following maintenance of a control or safety system.

Objective .

1 The objective is to assure that a system is operable before being used '

aner maintenance has been performed.

Snacincation

a. Following maintenance or modification of a control or safety system or component, L it shall be verified that the system is operable prior to its retum to service. A L

system shall not be considered operable until after it is successfully tested.

L - b. Any additions, modifcations, or maintenance to the ventilation system, the core and its associated support structure, the pool or its penetrations, the poo! coolant system, the rod drive mechanism, or the reactor safety system shall be made and tested in accordance with the specifications to which the systems were originally

- designed and fabricated or to specifcations approved by the Reactor Committee.

c. A licensed reactor operator shall be present during maintenance of the reactor control and safety system.

Baals

' This specifcation relates to changes in reactor systems which could directly affect the

-safety of the reactor. Changes or replacements to these systems which meet the original l design speci'ications are considered to meet the presently accepted operating criteria.

s muuntieese M Of h M M = mer

_ Amendment 17

'n E

page 22 4.6 Pool Water Conductivity Anchcabihtv This specific action applies to surveillance of pool water conductivity.

05ctive The objective is to assure that pool water mineral content is maintained at an acceptable level.

Specification The conductivity of bulk coolant water shall be verified to be within specified limits at least monthly.

Basis Based en experience, in which pool water conductivity changes slowly with time, observation at these intervals provides acceptable surveillance of conductivity to assure that L accelerated fuel clad corrosion does not occur.

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\ TIE Witl@ pad fev 10f97 Amendment 17

c' page 23 5.0 DtiSIGN FEATlK,ES 5.1 Reactor Fuel Applicabilftv This specification applies to the fuel elements used in the reactor core.

Objective The objective is to assure that the fuel elements are of such a design and fabricated in suchf; manner as to permit their use with a high degree of reliability with respect to their mechanicalintegrity.

, , Specifications y

a. Standard Fuel Element: The standard fuel element shall be of the TRIGA type and shall contain uranium zirconium hydride, clad in 0.020 inch of 304 staintess steel. It chall contain a maximum of 9.0 weight percent uranium which has a maximum enrichment less than 20 percent. There shall be 1.55 to 1.80 hydrogen atoms to 1.0 zirconium atom.

l b. Loading: The elements shall be placed in a closely packed array except for experimental facilities or for positions occupied by control rods, elements fully loaded with graphite, a neutron startup source, or single positions within the array filled with water.

Basis This type of fuel element has a long history of successful use in TRIGA reactors.

Amendment 17

u e.

9 page 24 5.2 Reactor Building

' Anplicabikty This specification applies to the facility which houses the reactor.

Oyective The objective is to assure that provisions are made to restrict the radioactivity released j into the environment.

Specifications

a. . The reactor shall be housed in a clos 0d room of a facility designed to restrict leakage.

L b. The free volume of the reactor room shall be at least 6,000 cubic feet.

c. All air or other gases exhausted from the reacto room during reactor operation p shall be released at a minimum of 12 feet above ground level.

L

[ d. - The reactor facility shall be equipped with a ventilation system capable of exhausting 1  ; air.or other gases from the reactor room from a stack at a minimum of 50 feet b

above ground level under emergency conditions.

Basil . - .

In order that the movement of air can be controlled, the facility contains no windows that can be opened. Under emergencyconditions the room airis exhausted through a filter and

- discharged through a stack at a minimum of 50 feet above ground to provide dilution.

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page 25 l

, 5.3 Fuel Storage Applic@ .

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 supercritical and will not reach unsafe temperatures.

Specifications

a. All fuel elements shall be stored in a geomeMcal array where the value of kgs i less than 0.9 for all conditions of moderation and reflection using light water.

b. irradiated fuel elements and fueled devices shall be stored in an array which will permit sufficient natural convection cooling by water or air such that the fuel element or fueled device temper;.ture will not exceed 8s' 0+C.

~

Basis Specification 5.3a assures that unplanned criticality will not occur iri fuel storace racks.

0: Specification 5.3b is based on a fuel temperature limit of 950'C to assure fuel clad

o. integrity when the clad temperatures can equal the fuel temperature (Simnad, G. A. Report E 117 833, February 1980, p.41)

~

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_ Am?Pd m??E 17 - - - - -

L f .s page 26 6.0 ADMINISTRATIVE CONTROLS 6.1 Organization

a. Tre r;;. der fe;"Ri f " k ;n ;,4;; .; W, of te ek;ck;r ;ad Erargy Engh;;dag 0;;;;.t. eat of te C;";;; ef Engh;;di;g ;od ISree et te Ui;reRi cf Anzeno as 2,;;7, h ie dt;. 7 k';W.

& reactorfacility shall be operated by the Nuclear Reactor Laboratory (NRL) at the University ofArizona. & Nuclear Reactor Laboratory Director shall retwrt to the Vice Presidentfor Research and Graduate Studies at the University ofArizona as shown in the diapam below.

l b. The reactor facility shall be under the supervision of a licensed senior operator for

' ' the reactor. He shall be responsible for assuring that all operations are conducted

' in a safe manner and within the limits prescribed by applicable federal regulations, by the facility license, and by the provisions of the Reactor Committee, n

.c. There shall be a Health Physicist responsible for assuring the safety of reactor operations from the standpoint of radiation protection.

d.: An NRC. licensed operator must be present in the control room when the key switch h

' is on. An Operator and one other person authorized by the Reactor Supervisor must

. be present in the Reactor Laboratory whenever the reactor is not shut down.

V.P. Research &

Graduate Studies

.q Director, Radiation Control Office

- Director NRL g . . . . . Reactor . . .. Heaith gg Commdee Physicist Q:

Reactor Operations mannuome MNM M M e., w Amendment 17 .

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page 27 4

j6.2 Review

a. There shd be a Reactor Committee which shd review reactor operations to assure

, . that the facipty is operated in a manner consistent with public safety and within the terms of the facility license.-

b. Tne responsibility of the Committee includes, but is not limited to, the following: ,

1. Review and approval of experiments utilizing the reactor facilities;

~ 2. Review and approval of aN proposed changes to the facility, procedures, and L TechnicalSpecifications;- ,

3. Determination of whether a proposed change, test, or experiment would consdtute an unreviewed safety question or a change in the Technical  ;

Specifications as required by 10 CFR 50.59, and review and approval of -

. required safeiy analyses;

/ -

4. Review of the operation and operational records of the facility;

5. Review of abnormal performance of plant equipment and operating anomalies;

.and L 6 ' Review of unusual or abnormal occurrences and incidents which are reportable e under 10 CFR 20 and 10 CFR 50.-

7. Review and audit of the retraining and requalification program for the operating

,e staff.

8. Biennial audst of the Emergency Plan.

c. The Committee shaN be composed of at least five members, and shall include a -

health physicist and members competent in the field of reactor operations, radiation science, or reactor engineering. The membership of the Committee shall be such

. 7as to maintain a high level degree of technical proficiency.

d. - The Committee shal establish a written charter defining such matters as the m authortly of the Committee, review and audit functions, and other such 7 ~

-*4 , administrative provisions as are required for effective functioning of the Committee.

Mmutes of all meetmas of the Committee shall be kept and submitted to committee members and to the HeedeHhe%;;;.;c.;;;t of t'ar er.d Crergi Er.glr.n.ir.g Yke Preddentfor Renarch andGraduate Studies in a timely manner.

[;

L e. ? A quomm of the Committee 'shall consist of not less than three members of the

Committee and shau include the chairman or his designee.

1 j tuumsts)qoses h$0fh M M m=7 j i1 Amendment 17

3' page 28

f. The Committee shall meet at least quartedy.

63 Operations 6.3.1 Ope. %g Precedures

a. OperaungProcedures Written procedures, reviewed and approved by the Reactor Committee, shall be in effect and followed for the following items. The procedures shall be adequate to assure the safety of the reactor, but should not preclude the use of independent judgment and

. action should the situation require such.

e: A Startup, operation, and shutdown of the reactor, c . b: 2. Installation or removal of fuel elements, control rods, experiments, and experimentalfacilities.

e: 3. Actions to be taken to correct specific and foreseen potential malfunctions of systems or components, including responses to alarms, suspected primary coolant system leaks, and abnormal reactivity changes, d:). Emergency conditions involving potential or actual release of radioactivity, iiicluding provisions for evacuation, re entry, recovery, and medical

. support.

e: 5. Maintenance procedures which could have an effect on reactor safety, f: 6. Periodic surveillance of reactor instrumentation and safety system, area monitors and continuous air monitors.

. Substantive changes to the above procedures shall be made only with the approval of the Reactor Committee. Temporary changes to the procedures that do not change their originalintent may be made with the approval of the Reactor Laboratory Director. All

_ such temporary changes to procedures shall be documented and subsequently reviewed by the Reactor Committee.

S.3.2 f1A'". Pres.c.m

. b. ALARA Program A program shall be established to assure that radiation exposures and releases are kept as low as reasonably achievable.

k p endment 17

page 29

- 6.A Acbon to be Taken in the Event a Safety Limit is Exceeded in the event a safety limit is exceeded, orthought to have been exceeded:

a. The reacter shall be shut down and reactor operation shall not be resumed until authorized by the NRC.

b. An immediate report of the occurrence shall be made to the Chairman of the

. Reactor Committee and reports shall be made to the NRC in accordance with Section 6.7 of these specifications. l

c. A report shall be made which shallinclu6 an analysis of the causes and extent of possible resultant damage, efficiency of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence. This report shall be submitted to the Reactor Committee for review, and a similar report submitted to the NRC when authorization to resume operation of the reactor is requested.

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1 muunts.nesse . MMM M miw i

Amendment 17

,p.

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page 30 6.5 Action to be Taken in the Event of a Reportable Occurrence in the event of a Reportable Occurrence, the following action shall be taken:

a. The Reactor Laboratory Director shall be notified and corrective action taken prior to re.umption of the operation involved.

b. - A report shall be made which shallinclude an analysis of the cause of the v occunence, efficacy of corrective action and recommendations for measures to prevent or reduce the probability of reoccurrence. This report shall be submitted to the Reactor Committee for review,

c. A report shall be submitted to the NRC in accordance with Section 6.7 of these specifications.

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. Usuositte ne . M0fh M M ==

Amendment 17

p page 31 6.6 Plant Operating Records in addition to the requirements of applicable regulations, and in no way substituting therefor;tecords and logs of the fol;owing items shall be prepared and retained for a period of at least 5 years (except as otherwise specified in the Commission's regulstions);

a. Normal plant operation (but not including supporting documents such as checklists, and recorder charts, which shall be maintained for a period of at least one year);

b. Principal maintenance Ntivities;

c. Reportable Occurrences; I

d. Equipment and component surveillance activities required by the Technical Specification; L

e. Experiments performed with the reactor; Logs and records of the following items shall be prepared and retained for the life of the facility._

f. GaseoJs and liquid radioactive effluents rele8 Sed to the environs;

g. Off site environmental monitoring surveys; o;

h. Fuelinventories and transfers;

i. Facility radiation and contamination surveys; e ,

J. Radiation exposures for all personnel; and

k. Updated, corrected, and as built drawings of the facility.

Amendment 17 .

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. 4 . .r w page 32

- 6.7 Reporting Requirements in adddion to the requirements of applicable regulations, and in no way sut'stituting I therefor, reports shall be made to the NRC as follows:

l L

• _ a. - A report within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and telegraph or telefax (FAX) to the Non-Power Reactors and Decommissioning Project Directorate of:

1.1 Any accidental off site release of radioactivity above limits permitted by 10 CFR

~ 20, whether or not the release resulted in property damage, personal injury, or :

exposure; i

! 2.1 Any violation of a Safety Umit;and w

3. Any reportable occurrences as defined in Section 1.0 (Reportable Occurrence)

.. of these specifications in wnting.

b. J A written report within ten days to the U. S. Nuclear Regulatory Commission, Attn

.

Document Control Desk,' Washington D.C. 20555, with a copy to the Non Power Reactors and Decommissioning Project Directorate, of : .

1,' Any significant variation of measured values from a corresponding predicted

^

' ~

value of previously measured value of safety. connected operating characteristics occuning during operation of the reactor;

, . 2. ; incidents or conditions relating to operation of the facility which prevented or could have prevented the performance of engineered safety features as

described in these specifications; o

3.L Any reportable occurrences as defined in Section 1.0 of these specifications; land-  ;

4... Any violation of a Safety Limit.

IL' q 5.J Any accidental offisite release of radioactivity above limits permitted by 10 CFR 20, whether or not the release resulted in property damage, personal injury, or y exposure. . .

cdA written report within 30 days to the U.S. Nuclear Regulatory Commission, Attn:

.; > SM Document Control Desk, Washington D.C. 20555, with a copy to the Non Power

. , Reactors and Decommissioning Project Directorate, of:

L Ic Any substantial variance from performance specifications contained in these specPations orin the Safety Analysis Report;

_, . }manness*

Ofh M M '- =

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Amendment 17

)!' i page 33 2.' Any significant change in the transient or accident analysis as described in the 1 Safety Analysis Report;  !

3. Any changes in facility organization; and

4. Any observed inadequacies in the implementation of administrative or  ;

procedural controls. - .

d.' A written report within 60 days ofter completion of startup testing of the reactor to the U. S. Nuclear Regulatory Commission, Attn: Document Control Desk, l Washington D.C. 20555, with a copy to the Non Power Reactors and Decommissioning Project Directorate.

1. An evaluation of facility performance to date in comparison with design

- predictions and specifications; and

", 2. A' reassessment of the' safety analysis submitted with the license application in 3, light of measured operating characteristics when such measurements indicate that there may be substantial variance from prior analysis.

e.1 A written annual report within 60 dhys following the 30th of June each year to the

~ l U.S. Nuclear Regulatory Commissiot4 Attn: Document Control Desk, Washington D.

C. 20555, with a copy to the Non Power Reactors and Decommissioning Project Directorate.

L1; J A brief narrative summary of (1) operating experience (including experiments

, performed), (2) changes in facility design, performance characteristics, and operating procedures related to reactor safety and occuning during the reporting period, and (3) results of surveillance tests and inspections;

1. ' 2. Tabulation of the energy output (in megawatt days) of the reactor, amount of

- pulse operation, hours reactor was critical, and the cumulative total energy

+

output since initial criticality; i 3. The number of emergency shutdowns and inadvertent scrams, including reasons therefore;

" 4. Discussion of the major maintenance operations performed during the period,

~ includmg the effect, if any, on the safety of the operation of the reactor, and the 2

reasons for any corrective maintenance required;

, 5.fA brief description including a summary of the safety evaluations of changes in the facility or in procedures and of tests and experiments canied out pursuant m yto Section 50.59 of 10 CFR Part 50; umanaa : UnkereQofMnne Meneeroh Meector - ar E

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6. A summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the licensee as measured at or prior to the point of such release or dit...arge; Liould Waste (summarized on a monthly basis)

(a) Radioactivity discharged during the reporting period.

(1) Total radioactivity released On curies).

e (2) The MPG limiting concentrations (10CFR20, fpperulix B) used and the isotopic composition if greater than 1 x 10' microcuries/cc for fission and activation products.

(3) Total radioactivity On curies), released by nuclide, during the reporting period, based on representative isotopic analysis.

(4) Average concentration at point of release On microcuries /cc) during the reporting period.

(b) . Total volume On gallons) of effluent water Oncluding diluent) during periods y of release.

Gaseous Waste (sumraarized on a monthly basis)

[ (a) Radioactivity discharged during the reporting period On curies) for:

(1) Gases.

(2) Particulates with half lives greater than eight days.

(b) The MPG limiting concentrations (10CFR20, Apperulix B) used and the nstimated activity On curies discharged during the reporting period, by nuclide, for all gases and particulates based on representative isotopic analysis.

Solid Waste

. (a) The total amount of solid waste packaged (in cubic feet).

.(b) The total activity involved On curies).

(c) The dates of transfer or shipment and disposition 8:

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page 35 D 7. ' A summary of radiation exposures received by facility personnel and visitors, includng dates and time of significant exposures, and a summary of the results of radiation and contamination surveys perforTned within the facility; and

8. A description of any environmental surveys performed outside the facility.

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page 36 6.8 Review of Experiments

- a. All proposed new experiments utilizing the reactor shall be evaluated by the

. experimenter and the Reactor Committee. The evaluadon shall be reviewed by a licensed Senior Operator of the facility (and the Health Physicist when appropriate) to assure compliance with the provisions of the utilization license, the Technical

- Specifications,10 CFR 20, and the requirements of 10 CFR 50.59. If, in his judgment, the experiment meets with the above provisions and does not constitute a threat to the integrity of the reactor, he shall submit it to the Reactor Supervisor for scheduling or to the Reactor Committee.for final approval as indicated in Section 6.2 above. When pertinent, the evaluation shallinclude:

1. The reactivity worth of the experiment;

2. . The integrity of the experiment, including the effects of changes in temperature, pressure, or chemical composition;

3. Any physical or chemicalinteraction that could occur with the reactor n components; and

4. Any radiation hazard that may resuit from the activation of materials or from extemal beams.

5. A determination that for the maximum planned or inadvertent pulse, no credible mechanism exists which could cause the experiment to fail.

b. Prior to performing an experiment not previously performed in the reactor, it shall be reviewed and approved in writing by the Reactor Committee. This review shall consider the following information:

1. The purpose of the experiment;

2. A procedure for the performance of the experiment; and

3. The evaluation approved by a licensed Senior Operator.

c. For the irradiation of materials, the appl 5nt shall submit an " Irradiation Request" to

- the Reactor Supervisor. This request shall contain information on the target material including the amount, chemical form, and packaging. For routine irradiations (which do not contain known explosive materials and which do not constitute a significant threat to the integrity of the reactor or to the safety of individuals) the approval for the Reactor Committee may be made by the Reactor Supervisor.-

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page 37 J. In evaluating experiments, the following assumptions should be used for the purpose of determining whether failure of the experiment v :uld cause the appropriate limits of 10 CFR 20 to be exceeded:

1. If the possibility exists that airbome concentrations of radioactive gases or aerosols may oe released within the facility,100 percent of the gases or aerosols will escape;

2. If the effluent exhausts through a filterinstallation designed for greater than 99 percent eft.ciency for 0.3 mlcron particles,10% of the particulates will escape; and

3. For a material whose boiling point is above 130*F and where vapors formed by boiling this material could escape only through a column of water above the core,10% of these vapors will escape.

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Amendment 17