ML20154G052

Proposed Tech Specs Supporting Safstor Option for Decontamination & Decommissioning of Univ of Il Nuclear Reactor Lab

ML20154G052
Person / Time
Site:	University of Illinois
Issue date:	10/05/1998
From:	ILLINOIS, UNIV. OF, URBANA, IL
To:
Shared Package
ML20154G045	List: ML20154G042 ML20154G052
References
NUDOCS 9810130041
Download: ML20154G052 (45)

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Technical Specifications Changes with Revision Bars in Support of SAFSTOR for the UIUC Nuclear Reactor Laboratory l

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!!A 188 u Zia8874 P PDR

-i-University ofIllinois Technical Specifications l TABLE OF CONTENTS Pace 1.0 DEFINITIONS 1 2.0 SAFETY LIMITS AND SAFETY SYSTEM SETTINGS 2.1 Safety Limit - Fuel Element Temperature 4 2.2 Limitine Safety System Setting 5 3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Reactivity 8 3.2 Hich Power Operation 10 3.3 Pulse Operation 11 3.4 ReactorInstrumentation 12 i

i 3.5 Reactor Safety System 14 3.6 Release'of Arnon-41 16 l 3.7 Ventilation System 17

' 3.8 Limitations on Exneriments 18 3.9 Suberitical Experiments and Fuel Storage  !

I Usine the Bulk Shieldine Facility 19 3.10 Primary Coolant Ouality 20b 4.0 SURVEILLANCE REOUIREMENTS 4.1 Fuel 21 4.2 Control Rods 22 4.3 Reactor Safety System 23 4.4 Emercency Spray Cooline System 24 4.5 Radiation Monitorine Esuipment 25 l

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4.6 Maintenance 26 4.7 Suberitical Exoeriments andFuelStorare Using the Bulk Shieldine Facility 26a 4.8 Primary Coolant Ouality 26b 5.0 D_ESIGN FEATURES 5.1 Reactor Fuel 27 1

5.2 Reactor Building 28 l

5.3 Fuel Storage 29 5.4 Emercency Removal of Decay Heat 30 6.0 ADMINISTRATIVE CONTROLS .

6.1 Organization 31 6.2 Review and Audit 33 6.3 Radiation Safety 35 6.4 Procedures 36 ,

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6.5 Experiments Review and Approval 37 1

6.6 Action to be taken in the Event a Safety limit is Exceeded 39 l

6.7 Action to be taken in the Event of an Abnormal Occurrence 40 6.8 Reporting Requirements 41 6.9 Plant Oneratine Records 43 l

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3.7 Ventilation System Applicability This specification upplies to the operation of the reactor facility ventilation system.

Obiective

. The objective is to assu e that the ventilation system is in operation to mitigate the coc2 g:ences of the possible release of radioactive materials resulting from reactor operation or duringfuel movements.

Specification The reactor shall not be operated andfuelshall not he moved unles.c ilie racility ventilation system is l in operation, except for periods of time not to exceed two days to permit repairs to the system.

During such periods of repair:

a. The reactor shall not be operated at power levels above 1 MW;

b. The reactor will not be operated in the puk;e mode; and

c. The reactor shall not be operated with experiments in place whose failure could result in the release of radioactive' gases or aerosols, and- l

d. Fuelshallnot be moved. l 7  ;

Basis It is shown in Chapter 13 of the SAR that operation of the ventilation system sufficiently reduces ofi-  ;

site doses to below 10 CFR Part 20 limits in the event of a TRIGA fuel element failure. The specifications go,verning operation of the reactor while the ventilation system is undergoing repair preclude the lik'elihood of fuel element failure during such times. It is shown in Chapter 5 of tb; SAR that, if the reactor were to be operating at a power level of 1 MW, fuel element failure will no. occur, even if all the reactor tank water were to be lost.

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l I 3.9 ' Syberitical Exneriments and Fue/ Storare Using the Bulk Shielding Facility l Applicability This specification applies to suberitical arrays and str- rge of fuel elements located external to the l reactor in the bulk shielding facility.

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Obiective The objective is to assure that accidental criticality of the storedfuel or suberitical experiment will not occur, proper radiation monitoring is present andpool level is maintainedfor radiation protection, l I l Snecifications

a. The effective multiplication constant (keg) of the suberitical facility shall not exceed 0.95 for assemblies of fuel elements using natural uranitun fuel and shall not exceed 0.99 for assemblies of.

TRIGA fuel elements.

b. For an assembly where it is expected that kerr could exceed 0.90, a step-wise procedure, in which kerr is determined using the inverse multiplication method, shall be followed for the initial loading of the assembly.

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c. During the nrst loading of any TRIGA fuel subcritical asetbly, a safety control rod worth at least 80 cents in the final assembly shall be provided in the ass m ab . The control rod shall be held in the withdrawn position by an electromagnet, and shall have se mm capability provided by manual switches and by a high radiation signal from a monitor located near the assembly. The naimum setpoint for the high radiation scram shall be 100 mr/hr.

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d. The initial use of the reactor as a source of neutrons for the suberitical assembly shall fotow a step-

. wise procedure for steady-state power increases and power transients.

e. A portable radiation monitor shall be used during the initial assembly and startup of the experiment l to determine dose rates in its vicinity,

e. Duringperiods when the Bulk Shielding Facility (BSF) or 1RIGA poolis usedforfuelstorage a continuous air monitor shall be in operation in the reactor hay and an area radiation monitor  ;

shall be in operation above the pool. 1he continuous air monitor andbr area radiation l monitor (s) may be out ofservicefor up to ten days provided that nofuel handling takesplace.

f Duringperiods when the Bulk Shielding Facility (BSF) or TRIGA poolis usedforfuelstorage the

. pool level will be maintained at a level at least six (6) feet above the top of thefuel elements.

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l Basis

. The performance of suberitical experiments external to the reactor was evakiated and authorized for the original TRIGA Mark 11 reactor at The University ofIllinois by amendment No. 6 to License No.

R-69. It was concluded at that time, and subsequently shown by actual operation, that the above specifications provided adequate assurance of safe operation. Since it has been shown that the l . presence of the suberitical assemblies external to the reactor had negligible effect on reactor operation, l it is ' concluded that such experiments can be performed with a similar degree of safety adjacent to the IIlinois Advanced TRIGA. Experience has shown through historical usage of thefuelstorage racks

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[ that a minimum levelofsixfeet ofu r above thefuelprovides adequate radiation shielding.

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-20b-3.11 Primary Coolant Ouality  !

l Aeolicability .

i This specification applies to the quality of the primary coolant water in contact with the cladding of l the fuel in the Advanced TRIGA eore/xiol and in the Bulk Shielding Facility. l Obiective i l

a) To limit the possibility for corrosion of the cladding on the fuel elements. l b) To limit the concentration ofdissolved materials which could be activated by neutron exposure.

l Specification l- i The Advanced TRIGA and/or the suberitical assembly in the Bulk Shielding Facility shall not be operated if the conductivity of the primary coolant water in the associated tank is higher than 4 l g:nho/cm.

Bish l l c) Corrosion may occur continuously in a water-metal system. In order to limit the rate of corrosion, and thereby extend the life and integrity of the fuel cladding, a water clean-up system is required.

Experience with water quality control at many reactor facilities has shown that maintenance within j the specified limit provides acceptable corrosion control. <

l b) Limiting the concentration of material dissolved in the water limits the radioactivity of neutron activatics products. This tends to decrease the inventory ofradionuclides in the entire coolant system, which w21 decrease personnel radiation exposure during both maintenance and operations.

This trend is consistent with the ALARA principle.

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4.0 SURVEILLANCE RE_QUIREMENTS 4.1 Fuel Applicability

'1S specification applies to the surveillance requirement for the fuel elements.

Obiective The objective is to assure the dimensions of the fuel elements remain within acceptable limits.

4 Specifications

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a. The standard fuel elements shall be measured for length and bend at intervals separated by not more than 1000 pulses ofmagnitude greater than $1.00 of reactivity or by an integrated reactivity of

$3,000. Fuel !cments in the B and C hexagonals sha!! be measured annua!!y not to exceed 14 monthshetween-measurements. Low hydride elements shall be measured annually not to exceed 14 months or at intervals separated by not more than 50 pulses, whichever is the lesser, if they are

, used for pulsed operation in the TRIGA core. New standard fuel elements shall be measured at intervals not to exceed 500 pulses until 1000 pulses have been exceeded.

- b. Standard thermocoupled fuel elements shall be checked at the same intervals as in above by the removal of the element from the core region and a visual check of the cladding.

c. A fuel element indicating an elongation greater than 1/4 af an inch over its original length or a lateral bending greater than 1/16 of an inch over it: original bending shall be considered to be damaged and shall not be used in the core for further operation.

d. Fuel elements in the B- and C-hexagonals shall be measured for possible distortion in the event that there is indication that fuel temperatures greater than the limiting safety system setting on i temperature may have been exceeded.

- Basis -

The most sever stresses induced in the fuel elements result from pulse operation of the reactor, during

. which differential expansion between the fuel and the cladding occurs and the pressure of the gases

. within the elements increases sharply. It is shown in Section Ill of the SAR that 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 i

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4.2 _Coptrol Rods  !

Apolicability j 1

i This specification applies to the surveillance require:nents for the control rods. i Obiective  !

1 The objective is to assure the integrity of the control rods.

' Specifications a-The+eactivity worth ofeach-contrel+od+ hall be determined +emiannua!!y, but at inte: val not to exceed eight =caths.

b. Contrel+od-drop-time + hall-bedetermined hemi-annually, but at4ntervats-not to exceed eight months.

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c. The control rods shall be visually inspected for deterioration biennially not exceed 30 months.

d-On each day that pulse mode + pere' ion-of-the-reactor 4; p!==d, a fbnetic=1 pc6crmance cheek-of thear= ient (pulac) red system cha!! be performed.

e-Semi =n=lly, at interval; not to eneeed eight months, thear= !:nt (pu! e) rod 4 rive-eylinders and the-assceinted air supply system sha'! be inspected, cleaned-and lubricated as neec :arj.

Bases The reactivity-worth +f4he-centrol-reds-luneasured4o-assure 4 hat the required shutdowrunargirris available =d te provide a me=: for determining the reactivity worths of+xperiments inserted in the core-The-visual inspection +f4he+cniro! red and-measurenwnts-ef4heirdrepaimesereemde4e determine whether4he+ontrol rods are capable of performingpoperly: . 7he risua/ inspection of thefuelfallower control rods specified has been shown to be adequate based on prior experience  ;

with a lack offuel cladding deterioration over time.

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I 4.3 Reactor Safety System Apolicability 1

. This specification applies to the surveillance requirements for the measuring channels of the reactor safety system.

Objective The objective is to assure that the safety system will remain operable and will prevent the fuel l

temperature safety limit from being exceeded.

l Specifications

a. A channel test of each of the reactor safety system channels 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 fuel element temperature measuring channels shall be performed daily whenever the reactor is in operation at power levels greater than 50 kw or when pulse operation is planned.

c. A channel check of the power level measuring channels shall be performed daily whenever the reactor is in operation.

. d-Adannel ee!ibration by the calorimetric method-shall be made of the power-leve! monitoring channels cmi annually, bat +t-intervals-net-tc exceed eight months-e-A-calibration-of-the temperature-measurmg<hannel shall be-performed semi annuauyrbut-at intervals-not-to-exceed e:ght-months. This calibrationsha4 consist ofintroducing c!cetric potentials in-placeef-the thermoccuple input te4hc channels-Basis The daily tests and channel check will' assure that the safety channels are operable. The semiannual calibrations and verifications will permit any long-term drift of the channels to be corrected.

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i- 4 Section 4.4 Emergency Spray Cooling System intentionally deleted.  !

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25-1 4.5 Radiation Monitoring Equipment l l

Apolicability l

.l This specification applies to the radiation monitoring equipment required by Section 3.4 and 3.9 of these specifications.

1 Obiective The objective is to assure that the radiation monitoring equipment is operating and to verify the l appropriate alarm settings. I l

Specifications l

The alarm set points for the radiation monitoring instrumentation shall be verified weekly-during periods when the reactor is in operation-monthly not to exceedsix weeks.

A Basis l Because of the redundancy of radiation monitoring instrumentation provided, weeklymonthly surveillance of the equipment will be adequate to assure that suflicient protection against radiation is available.

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-26a-l 4.7 Subcritical Experiments andFuelStorage Usine the Bulk Shieldine Facility Applicability This specification applies to the surveillance requirements associated with the suberitical assembly and storage offieel elements in the bulk shielding facility.

Obiective To ensure safe operation of the subcritical assembly . to ensure that the radiation monitoring equipment is j operating properly, and that the pool level is maintainedfor radiation protection.

Specification a) The reactivity worth oiti e control rod shall be determined annually (interval not to exceed fifteen months).

The surveillance may be &ferred indefinitely when the suberitical assembly is not being utilized, but shall be the first operation performc3 when the subcritical assembly is to be operated.  ;

b) Control rod drop time shall be determined annually (interval not to exceed fifteen months). The drop time from the fully withdrawn to 03 percent of full reactivity insertion shall be less than one second. The .

surveillance may be deferral indefinitely when the suberitical assembly is not being utilized, but shall be ,

performed prior to operation of the assembly. j c) The radiation monitor utilized for a high radiation signal scram shall be calibrated and verified operable annually (interval not to exceed fifteen months). The surveillance may be deferred indefinitely when the suberitical assembly is not being utilized, but shall be performed prior to operation of the assembly.

l d) Approximately 210 % of the fuel elements in the suberitical assembly, or in wet storage racks, shall be visually inspected annually for any indication of deterioration or distortion (interval not to exceed fifteen months) such that all of the elements in the suberitical assembly are inspected over a fweten year period (interval not to exceed fweten and one half years). If any indication of deterioration or distortion is noted the element shall be removed from+erwee to other storage.

I e) The manual and high radiation scrams shall be verified operable daily prior to operation of the suberitical assembly. This specification is only applicable on days when the suberitical assembly is to be operated.

f) The lhdk Shielding Facilitypoollevel shall be checked on a weekly (twt to exceed ten days) basis.

. Basis The reactivity worth is measured to assure that control of the suberitical assembly can be maintained. The control rod drop time verifies the scram capability of the control rod. Calibration and verification of the operability of the radiation monitor verifies the scram capability of the monitor. The visual inspection of the fuel elements specified had been shown to be adequate based on prior experience with a lack of fuel deterioration over time.

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1 4.8 Primary Coolant Ouality  !

Applicability i

This specification applies to the surveillance of the quality of the primary coolant water in contact with

'the cladding of the fuel in the Advanced TRIGA corepooland in the Bulk Shielding Facility. l l

Obiective l

The objective is to ensure that the quality of the primary coolant water in contact with the fuel cladding does not deteriorate over extended periods of time even if the reactor is not operated.

Specification The conductivity of the primar" colant water in contact with the cladding of the fuelin the Advanced TRIGA cerepool and in the Bus Shielding Facility shall be measured at least once every two weeks l

.' (interval not to exceed 21 days) and shall not exceed 5 mho/cm for more than 5 consecutive days. If the conductivity of the water exceeds 4 mho/cm the sampling frequency shall be increased to daily until the conductivity drops below 4 pmho/cm. If the conductivity exceeds 5 mho/cm for more than five consecutive days the fuel shall be removed from the tank to storage until such time that the conductivity has been restored to below 4 mho/cm.

Basis Section 3.11 ensures that the water quality is acceptable during reactor operation. Section 4.8 ensures l that the fuel cladding is not exposed to a significantly more corrosive environment for an extended period of time in the event that the reactor is not actually operated.

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-3 0-j Section 5.4 Emergency Removal ofDecay Heat intentionally deleted. l 14-E_m_ergency Removal +f-Decav4 feel Applicability This specifica:ienSpphes4o4he-emergency +emovalef-decay-heat:

Obieetive i

The objective is to assure 4 hat 4here49-no-ftssionproduct4elease-dae4o4) eating +f4he fuel c!cments l subsequent 4o-a4etal4esseftrimary-water 4 rom-around4he-core-l l

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l T4 tere + hall-beanautomatically-operated-emergency coolinpystenwapable af-manual-operation which-will+ pray-water-ever-the-fuel-elementsat-a-rateefet least 2 gpm-for-an-inddmite-time: l l Basis i

T4:e system is ev;!uated4nGhapter4ef4he-SAR;-whereit-is-shown4 hat 4-gpmef-spray eccling can l keep 4he fuel temperatures wc!! helow the :;afety4imit: 1 l

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l 6.0 ADMINISTRATIVE CONTROLS l 6.1 Organization l 6.1.1 Structure and Re_sponsibility

a. The reactor facility shall be an integral part of the Department of Nuclear Engineering of the University ofIllinois. The reactor shall be related to the University structure as shown in Chart I.

b. The reactor facility shall be under the supervision of the Reactor Administrator who shall have been qualified as a licensed senior reactor 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 the facility license and the provisions of the Nuclear Reactor Committee.

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c. There shall be a Reactor Health Physicist responsible for assuring the day to day and routine radiological safety activities at the Nuclear Reactor Laboratory. The University ofIllinois Radiation Safety Ollicer shall be responsible for monitoring, planning, and promoting radiological safety at the Nuclear Reactor Laboratory. He has the responsibility and authority to stop, secure or otherwise control as necessary any operation or activity that poses an unacceptable i radiological hazard.  ;

CHARTI Head of Department of Division of Environmental Nuclear Engineering Health and Safety I !_________ ,

i Radiation Safety Oflicer Nuclear ___ l  ;

Reactor Admn. .ustrator l Reactor t________________, 1 Committee I j  !

I Operations-Supervisor Reactor Health Ph;sicist l

l Operation-StalT Reactor-Operation CHART 1: Administrative organization of the reactor facility. Dashed lines indicate repo; ting paths outside the operational chain of supervision, indicated by solid lines.

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-3 2-l 6.1.2 Staffima

a. The minimum staffing whenthe4eactor-inet-shutdown <rt the Nuclear Reactor Laboratory shall be: I l

l 1. A4icensed reactorsperatorskell-be-imtho<ontrotsoom: Reactor Adninistrator. This individual shall meet the requirements ofANSPANS-15.4 'American National Standardfor the Selection and Training ofPersonnelfor Resec::h Reactors "Jbr a Level Two individual. l 2.--A-second<lestg sted persomshallhpresent4nside4he4eactor-buildingebietoshutdown-the4eactor-in l

a+ emergency. Unexpectedebsencefor-es-kwas two hours-to-accommcdate a personalemergency 1 may-beeeceptabimovided4mmediate-action 4s4akemto-obtaima4eplacementReactor Health Physicist.

This individual shall meet the requirements ofANSVANS-15.4 "American National Standardfor the Selection and Traming ofPersonnelfbr Research Reactors "for a Level Three individual in addition to training in hxith physics. ,

3. A Senior-Reactor-Operator-shalhbe+eadilyevailableen<alles-defmed-imtheNuclear-Reactor i Laboratory-Rules-and-Regulations- i l

l b. A list of reactor facility personnel by name and telephone number shall be readily available isthe-control  :

room-for-use-by-thoeperatorto the ULUC Division ofPublic Safety dispatcher. One of these individuals \

shall be reachable and able to respond to thefacility within approximately one hour. The list shall l

! include:  ;

1. Operations-SupervisorCampus Radiation Safety Officer l

2. Reactor Administrator

3. Head, Department of Nuclear Engineering l

{ 4. Reactor Health Physicist i

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5. Licensed operators

( c. Events requiring the presence at the facility of a Senior Reactor Operator:

1. Initial startup and approach to power.

2. All fuel or control rod relocations.-within the+eactor< ore 4egion. l l

3. Relocation of any in-core experiment with a reactivity worth greater than one dollar.

l i 4. Recovery from unplanned or unscheduled shutdown or significant power reduction (In these instances, documented verbal concurrence from the Senior Reactor Operator is required).

! 6.1.3 Selection and Trainina of Personnel l

The Reactor Administrator is responsible for the training and requalification of the facility reactor operators and j senior reactor operators. He selection, training, and requalification of operations personnel shall be consistent with all current regulations and guidelines. j

-3 3-l 6.2 Review and Audit 6.2.1 Charter and Rules l

a. He Reactor Committee shall be composed of at least five voting members, one of whom shall be a Health Physicist designated by the campus Radiation Safety Officer for the University, and one whom shall be the Reactor Administrator, and one whom shall be the Reactor Health Physicist. %reeThe remaining members shall be appointed by the Head of the Department of Nuclear Engineering, chosemfrom4he facuky of Nuclect Engineering-so as to maintain a balanced knowledge of reactor safety and regulation.

He Reactor-Healtidhys+ctst-shaltbe non votintpnember:

b. The Reactor Committee shall have a written statement defming such matters as the authority of the committee, the subjects within its purview, and other such administrative provisions as are required for the effective functioning of the Reacter Committee Minutes of all meetings of the Reactor Committee shall be kept. .

c. A quorum of the Reactor Committx shall be a majority ofnot less than one half of the members and th'e reactor operating staff shall not constitute a voting majority.

d. He Reactor Committee shall meet at least quaiterlysemiannually not to exceed nine months -witlethe interval betvecen n;cc:ings notwexceed4-months-6.2.2 Review Function ne review function of the Committee shall include, but is not limited to the following:

a. Determination that proposed changes in equipment, systems, tests, experiments, or procedures do not involve an unreviewed safety question.

b. All new procedures and major revisions thereto having safety significance, proposed changes in reactor l facility equipment, or systems having safety significance.

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c. All new experiments or classes of experiments for determination that an unreviewed safety question does not exist.

d. Proposed changes in the technical specifications or license.

e. Violations of technical specifications or license.

f. Operating abnormalities having safety significance.

g. Reportable occurrences as listed in 6.8.

h. Audit reports.

l 'A written report or minutes of the findings and recommendations of the Committee shall be submitted to the Head, Department of Nuclear Engineering, and the Reactor Committee members in a timely manner after each meeting.

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-3 6-6.4 Procedures Written procedures shall be prepared, reviewed and approved prior to initiating any of the activities listed in this section. The procedures shall be reviewed by the Reactor Committee and approved by the Reactor Administrator and such review and approval shall be documented in a timely manner. The procedures shall be adequate to assure the safety of the reactor, but should not preclude the use ofindependent judgment and action should the situation require such.

a. Startup, operation, and shutdown of the reactor.

b Installation or removal of fuel elements, control rods, experiments, and experimental facilities.

c. Actions to be taken to correct specific and foreseen potential malfunctions of systems or components, including responses to alarms, suspected primary coolant leaks, and abnormal rea;tivity changes.

d. Emergency cxditions involving potential or actual release of radioactivity, including provisions for evacuation, reentry-entry, recovery, and medical support.

e. Maintenance procedures which could have an effect on reactor safety.

f. Periodic surveillance of reactor instrumentation and safety systems, area monitors and continuous air monitors. i l

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g. Personnel radiation protection, consistent with applicable regulations or guidelines. The procedures shall include management commitment and programs to maintain exposures and releases As Low As is Reasonably Achievable (ALARA).

h. Implementation of physical security plan Substantive changes to the above procedures shall be made only after documented review by the Reactor Committee and approval by the Reactor Administrator. Minor modifications to the original procedures which do not change their original intent may be made by the Reactor Administrator 4perations Supervisor-but-t he-nmdifteationsinust4e-a pproved4y4he-Reactor-Ad minist rator-wit hin44 days.

Temporary deviations from the procedures may be made by the responsible Senior Reactor Operator or higher individual present, in order to deal with special or unusual circumstances or conditions. Such deviations shall be documented and reported to the Reactor Administrator.

-3 7-l 6.5 Experiments Review and Approv_al l

l a. All new experiments or class of experiments utilizing the reactor shall be evaluated by the experimenter and a staff member approved by the Reactor Committee. The evaluation shall be reviewed by a licensed senior reactor operator of the facility and the Reactor Health Physicist to I assure compliance with the provisions of the utilization license, the Technical Specifications and I l 10CFR20. If, in theirjudgment, the experiment meets with the above provisions and does not l_ constitute a threat to the integrity of the reactor, they shall submit it to the Operations SupervisorReactor Administrator for review. If the Operations-SupervisorReactor Administrator i agrees with the evaluation by the senior reactor operator and the Reactor Health Physicist he shall submit the experiment to the Reactor Committee for review as indicated in Section 6.2. The experiment shall be approved in writing by the Reactor Administrator prior to initiation. When pertinent, the evaluation shall include the following: l

1. The reactivity worth of the experiment.

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

3. Any physical or chemical interaction that could occur with the reactor components.

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4. Any radiation hazard that may result from the activation of materials or from external beams. l

b. The Reactor Committee review of an experiment shall be performed prior to the first experiment and shall be documented in writing and shall consider at a minimum the following:

1 The purpose of the experiment.

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2. A procedure for the performance of the experiment.

3. The evaluation approved by a licensed senior reactor operator.

4. Determination that the experiment does not involve an unreviewed safety question.

c. Substantive changes to previously approved experiments shall be made only after review by the Reactor Committee and approved in writing by the Reactor Administrator. Minor changes that do not significantly alter the safety analysis of the experiment may be approved by the Operntions-Supervisor or-Reactor Administrator,

d. For the irradiation of materials, the applicant shall submit a request to the Reactor Health l Physicist-enAOperations4upervisor. This request shall contain at a minimum information on the target material including the amount, chemical form, and expected radiological hazard for the desired irradiation period. For routine irradiations (which do not contain nuclear fuel or known explosive materials and which do not constitute a significant threat to the integrity of the reactor or to the safety ofindividuals), the approval for the Reactor Committee may be made by the Reactor Health Physicist I and-the Operations-Supervisor.

6.7 Action to be taken in the Event of an Abnormal Occurrence

In the event of an abnormal occurrence, as defined in Section 1.M of the specifications, the following action shall be taken:

a. The reactor shall be shutdown and the Operations &pervisoear4 Reactor Administrator shall be l notified and corrective action taken prior to resumption of operations. The Reactor Administrator shall authorize resumption of operations.

b. A report shall be made which shall include an analysis of the cause of the occurrence, eflicacy of corrective action and recommendations for measures to prevent or reduce the probability of reoccurrence.  !

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c. The occurrence shall be reviewed by the Reactor Committee at the next scheduled meeting. )

d. A report shall be submitted to the USNRC in accordance with Section 6.8 of these specifications.

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6.8 Beporting Requirements In addition to the requirements of applicable regulations, and in no way substituting therefore, reports shall be mad'e to the USNRC as follows:

a. There shall be a report not later than the following working day by telephone and confirmed in writing by facsimile or similar conveyance to the Regional Administrator, USNRC, Region III and the USNRC headquarters operations center to be followed by a written report that describes the l circumstances of the event that describes the event within 14 days to the Document Control Desk, USNRC Headquarters, and a copy to the Regional Administrator, USNRC, Region III of any of the following:

1. Release of radioactivity from the site above allowed limits.

2. Violation of safety limits.

3. Any significant variation from measured values from a corresponding predicted or previously measured value of safety-connected operating characteristics cccurring during operation of the reactor.

4. 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.

5. Any abnormal occurrences as at.aed in Section 1.14 of these specifications.

b. A report within 30 days (in writing to the Document Control Desk, USNRC Headquarters) of:

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

2. Significant changes in the transient or accident analysis as described in the Safety Analysis Report.

3. Permanent changes in the facility organization involving the4perations-SupervisorReactor Health Physicist, Reactor Administrator or Department Head.

c. A report .within 60 days afler criticality of the reactor (in writing to the Document Control Desk,

! USNRC Headquarters) upon receipt of a new facility license or an amendment to the license authorizing and increase in reactor power level or the installation of a new core, describing the measured values of the operating conditions or characteristics of the reactor under the new conditions, including:

l. Total control rod reactivity worth.

2. Reactivity worth of the single control rod of highest reactivity worth.

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Technical Specifications Changes "To Be Inserted" in Support of SAFSTOR l for the l UIUC Nuclear Reactor Laboratory 9/98 '

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University ofIllinois Technical Specifications l TABLE OF CONTENTS Page l 1.0 DEFINITIONS 1 2.0 SAFETY LIMITS AND SAFETY SYSTEM SETTINGS 2.1 Safety Limit - Fuel Element Temperature 4 2.2 Limiting Safety System Setting 5 3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Reactivity 8 3.2 High Power Operation 10 L

3.3 Pulse Operation 11 3.4 ReactorInstrumentation 12 3.5 Reactor Safety System 14 3.6 Release of Argon-41 16 3.7 Ventilation System 17 3.8 Limitations on Experiments 18 3.9 Subcritical Experiments and Fuel Storage

Using the Bulk Shielding Facility 19 3.10 Primary Coolant Ouality 20b l 4.0 SURVEILLANCE REOUIREMENTS l 4.1 Fuel 21 4.2 Control Rods 22 4.3 Reactor Safety System 23 l

l 4.4 Emergency Spray Cooling System 24 4.5 Radiation Monitoring Equipment 25 4

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. 4.6 Maintenance 26 l

4.7 Suberitical Experiments and Fuel Storagg Using the Bulk Shielding Facility 26a 4.8 Primary Coolant Ouality 26b 5.0 DESIGNFEATURES

5.1 Reactor Fuel 27 l

l' 5.2. Reactor Building 28 5.3 Fuel Storage 29

!4 5.4 Emergency Removal of Decav Heat 30 l

l 6.0 ADMINISTRATIVE CONTROLS 6.1 Organization 31 6.2 Review and Audit 33 6.3 Radiation Safety 35 I

6.4 Procedures 36

'6.5 Experiments Review and Approval 37 6.6 Action to be taken in the Event a Safety limit is Exceeded 39 6.7 Action to be taken in the Event of an Abnormal Occurrence 40 i-L 6.8 Reporting Reauirements 41 6.9 Plant Operating Records 43 7

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l 3.7 Ventilation System l

Applicability This specification applies to the operation of the reactor facility ventilation system.

Obiective The objective is to assure that the ventilation system is in operation to mitigate the consequences of l c

the possible release of radioactive materials resulting from reactor operation or during fuel l movements.

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. Specification  !

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The reactor shall not be operated and fuel shall not be moved unless the facility ventilation system is in operation, except for periods of time not to exceed two days to permit repairs to the system. During such periods of repair:

a. The reactor shall not be operated at power levels above 1 MW;

b. The reactor will not be operated in the pulse mode; and

c. The reactor shall not be operated with experiments in place whose failure could result in the release of radioactive gases or aerosols, and l

l d. ' Fuel shall not be moved.

l Basis i It is shown in Chapter 13 of the SAR that operation'of the ventilation system sufficiently reduces off-site doses to below 10 CFR Part 20 limits in the event of a TRIGA fuel element failure The l specifications governing operation of the reactor while the ventilation system is undergoing repair preclude the likelihood of fuel element failure during such times. It is shown in Chapter 5 of the SAR that, if the reactor were to be operating at a power level of 1 MW, fuel element failure will not occur, even if all the reactor tank water were to be lost.

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3.9 Suberitigal Experiments.and Fuel Storage Using the Bulk Shielding Facility Applicability This specification applies to suberitical anays and storage of fuel elements located external to the reactor in the bulk shielding facility.

Obiective ,

The objective is to assure that accidental criticality of the stored fuel or suberitical experiment will not

. occur, proper radiation monitoring is present and pool le'.el is maintained for radiation protection, Specifications

a. The effective multiplication constant (k.g) of the subcritical facility shall not exceed 0.95 for assemblies of fuel elements using natural uranium fuel and shall not exceed 0.99 for assemblies of TRIGA fuel elements.

b. For an assembly where it is expected that k.a could exceed 0.90, a step-wise procedure,'in which l k.a is determined using the inverse multiplication method, shall be followed for the initial loading of the assembly.

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c. During the first loading of any TRIGA fuel suberitical assembly, a safety control rod worth at least 80 cents in the final assembly shall be provided in the assembly. The control rod shall be held in the withdrawn position by an electromagnet, and shall have scram capability provided by manual switches and by a high radiation signal from a monitor located near the assembly. The maximum setpoint for the high radiation scram shall be 100 mr/hr.

d. The initial use of the reactor as a source of neutrons for the suberitical assembly shall follow a step-wise procedure for steady-state power increases and power transients.

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e. A portable radiation monitor shall be used during the initial assembly and startup of the experiment to determine dose rates in its vicinity.

' e. During periods when the Bulk Shielding Facility (BSF) or TRIGA poolis used for fuel storage a continuous air monitor shall be in operation in the reactor bay and an area radiation monitor shall be in opera'. ion above the pool. The continuous air monitor and/or area radiation monitor (s) may

.be out of service for up to ten days provided that no fuel handling takes place.

f. During periods when the Bulk Shielding Facility (BSF) or TRIGA poolis used for fuel storage the pool level will be maintained at a level at least six (6) feet above the top of the fuel elements.

1 l i IliLSIS The performance of suberitical experiments external to the reactor was evaluated and authorized for the original TRIGA Mark II reactor at The University ofIllinois by amendment No. 6 to License No. 1 R-69. It was concluded at that time, and subsequently shown by actual operation, that the above  !

l specifications provided adequate assurance of safe operation. Since it has been shown that the I

presence of the suberitical assemblies external to the reactor had negligible effect on reactor operation, it is concluded that such experiments can be performed with a similar degree of safety adjacent to the Illinois Advanced TRIGA. Experience has shown through historical usage of the fuel storage racks that a minimum level of six feet of water above the fuel provides adequate radiation shielding.

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-20b- l 3.11 Primary Coolant Ouality Applicability This specification applies to the quality of the primary coolant water in contact with the cladding of the fuel in the Advanced TRIGA pool and in the Bulk Shielding Facility.

Obiective  ;

l a) To limit the possibility for corrosion of the cladding on the fuel elements.

b) To limit the concentration of dissolved materials which could be activated by neutron exposure.

Specification l The Advanced TRIGA and/or the subcritical assembly in the Bulk Shielding Facility shall not be operated if the conductivity of the primary coolant water in the associated tank is higher than 4 .

I pmho/cm.

' Basis a) Corrosion may occur continuously in a water-metal system. In order to limit the rate of corrosion, and thereby extend the life and integrity of the fuel cladding, a water clean-up system is required.  !'

Experience with water quality control at many reactor facilities has shown that maintenance within the specified limit provides acceptable corrosion control.

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b) Limiting the concentration of material dissolved in the water limits the radioactivity of neutron activation products. This tends to decrease the inventory of radionuclides in the entire coolant system, which will decrease personnel radiation exposure during both maintenance and operations.

This trend is consistent with the ALARA principle.

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21-4.0_jSURVEILLANCE REOUIREMENTS  ;

4.1 Ev.d Applicability f l

This specification applies to the surveillance requirement for the fuel elements.

Objective i

The objective is to assure the dimensions of the fuel elements remain within acceptable limits.

Specifications

a. The standard fuel elements shall be measured for length and bend at intervals separated by not more l than 1000 pulses of magnitude greater than $1.00 of reactivity or by an integrated reactivity of l

$3,000.. Low hydride elements shall be measured annually not to exceed 14 months or at intervals  :

separated by not more than 50 pulses, whichever is the lesser, if they are used for pulsed operation

'in the TRIGA core. New standard fuel elements shall be measured at intervals not to exceed 500 pulses until 1000 pulses have been exceeded.

b. Standard thermocoupled fuel elements shall be checked at the same intervals as in above by the l removal of the element from the core region and a visual check of the cladding. )

, c. A fuel element indicating an elongation greater than 1/4 of an inch over its original length or a -

lateral bending greater than 1/16 of an inch over its original bending shall be considered to be damaged and shall not be used in the core for further operation.

d. Fuel elements in the B- and C-hexagonals shall be measured for possible distortion in the event that there is indication that fuel temperatures greater than the limiting safety system setting on temperature may have been exceeded.

Basis The most sever stresses induced in the fuel elements result from pulse operation of the reactor, during l which differential expansion between the fuel and the cladding occurs and the pressure of the gases l within the elements increases sharply. It is shown in Section III of the SAR that 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 t

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l 4.2 Control Rods l 6pplicability This specification applies to the surveillance requirements for the control rods.

. Obiective 1 1

The objective is to assure the integrity of the control rods. j Soecifications

c. The control rods shall be visually inspected for deterioration biennially not exceed 30 months.

Bases

. The visual inspection of the fuel follower control rods specified has been shown to be adequate .

based on prior experience with a lack of fuel cladding deterioration over time.

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'4.3 B_e_

e actor Safety Syste_m Applicability This specification applies to the surveillance requirements for the measuring aannels of the reactor safety system.

Obiective The objective is to assure tht the safety system will remain operable and will prevent the fuel temperature safety limit from being exceeded.

Specifications-

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

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b. A channel check of the fuel element temperature measuring channels shall be performed daily whenever the reactor is in operation at power levels greater than 50 kw or when pulse operation is planned.

c. A channel check of the power level measuring channels shall be performed daily whenever the reactoris in operation.

basis The dri!y tests and channel check will assure that the safety channels are operable. The semiannual calibrations and verifications will permit any long-term drift of the channels to be corrected.

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I Section 4.4 Emergency Spray Cooling tyn.:m intentionally deleted.

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.~ . 1 4.5 Radiation Monitoring Equipment Appjitabil:ty This specification a,splies to the radiation monitoring equipment required by Section 3.4 and 3.9 of these specifications..

Obiective The objective is to assure that the radiation monitoring equipment is operating and to verify the approprhte alarm settings.

Specificati(ns  !

The alarm set points for the radiation monitoring instmmentation shall be verified monthly not to exceed six weeks.

Basis i 1

Because of the redundancy of radiation monitoring instrumentation provided, monthly surveillance of the equipment will be adequate to assure that sufficient protection against radiation is available.

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-26a-4.7 Syberitical Experiments and Fuel Storaae Usina the Bulk Shielding Facility Applicability His specification applies to the surveillance requirements associated with the suberitical assembly and storage of fuel elements in the bulk shielding facility.

Obiective To ensure safe operation of the suberitical assembly, to ensure that the radiation monitoring equipment is l orerating properly, and that the pool level is maintained for radiation protection. I Specification a) He reactivity worth of the control rod shall be determined annually (interval not to exceed fifteen months).

He surveillance may be deferred indefinitely when the suberitical assembly is not being utilized, but shall be the first operation performed when the suberitical assembly is to be operated.

b) Control rod drop time shall be determined annually (interval not to exceed fifteen months). He drop time l from the fully withdrawn to 90 percent of full reactivity insertion shall be less than one second. He surveillance may be deferred indefinitely when the suberitical assembly is not being utilized, but shall be performed prior to operation of the assembly.

c) He radiation monitor utilized for a high radiation signal scram shall be calibrated and verified operable annually (interval not to exceed fifteen months). He surveillance may be deferred indefinitely when the suberitical assembly is not being utilized, but shall be performed prior to operation of the assembly, d) Approximately 10 % of the fuel elements in the suberitical assembly, or in wet storage racks, shall be j I

visually inspected annually for any indication of deterioration or distortion (interval not to exceed fifteen months) such that all of the elements in the subcritical assembly are inspected over a ten year period (interval .

not to exceed ten and one half years). If any indication of deterioration or distortion is noted the element )

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shall be removed to other storage.

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! e) ne manual and high radiation scrams shall be verified operable daily prior to operation of the suberitical l: assembly. %is specification is only applicable on days when the suberitical assembly is to be operated.

f) He Bulk Shielding Facility pool level shall be checked on a weekly (not to exceed ten days) basis.

Basis ne reactivity mrth is measured to assure that control of the subentical assembly can be maintained. He control rod drop time verifies the scram capability of the control rod. Calibration and venfie tion of the operability of the radiation monitor verifies the scram capability of the monitor. He visual msgion of the fuel elemetts specified had been shown to be adequate based on prior experience with a lack of fuel deterioration over time.

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-26b- 1 4.8 Primary Coolant Ouality Applicability This specification applies to the surveillance of the quality of the primary coolant water in contact with the cladding of the fuel in the Advanced TRIGA pool and in the Bulk Shielding Facility.

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. The objective is to ensure that the quality of the primary coolant water in contact with the fuel

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cladding does not deteriorate over extended periods of time even if the reactor is not operated.

Specification The conductivity of the primary coolant water in contact with the cladding of the fuel in the Advanced TRIGA pool and in the Bulk Shielding Facility shall be measured at least once every two weeks  !

l (interval not to exceed 21 days) and shall not exceed 5 mho/cm for more tha,5 consecutive days. If the conductivity of the wate:. exceeds 4 mho/cm the sampling frequency shall be increased to daily l

l, until the conductivity drops below 4 mho/cm. If the conductivity exceeds 5 mho/cm for more than

)- five consecutive days the fuel shau be removed from the tank to storage until such time that the l conductivity has been restored to below 4 mho/cm. l i

Basis i l

l Section 3.11 ensures that the water quality is acceptable during reactor operation. Section 4,8 ensures

! that the fuel cladding is not exposed to a significantly more corrosive environment for an extended  ;

period of time in the event that the reactor is not actually operated.

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( 6.0 ADMINISTRATIVE CONTROLS 6.1 Organization 6.1.1 Structure and_fesoonsibility

a. The reactos facility shall be an integral part of the Department of Nuclear EnFi neering of the University c fIllinois. The reactor shall be related to the University structu:c as shown in Chart I.

b. The reactrar facility shall be under the supervision of the Reactor Administrater who shall have been qualified as a licensed senior reactor operator for the reactor. He shal! be responsible for assuring that all operations are conducted in a safe manner and within the limits prescribed by the facility license and the provisions of the Nuclear Reactor Committee.

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c. There shall be a Reactor Health Physicist responsible for assuring the day to day and routine ,

radiological safety activities at the Nuclear Reactor Laboratory. The University ofIllinois Radiation Safety Officer shall be responsible for monitoring, planning, and promoting radiological l safety at the Nuclear Reactor Laboratory. He has the responsibility and authority to stop, secure i or otherwise control as necessary any operation or activity that poses an unacceptable j radiological hazard.

CHARTI Heaa . Department of Division of Environmental Nuclear Engineering Health and Safety

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l Radiation Safety Officer i Nuclear . _ _

l 1 Reactor Administrator l Reactor c________________, I ,

Committee I l

, l Reactor Health Physicist i

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! CIfART I: Administrative organization of the reactor facility. Dashed lines indicate reporting paths outside the operational chain of supervision, indicated by solid lines.

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-3 2-6.1.2 Staffi.ngn I i

a, The minimum staffing at the Nuclear Reactor Laboratory shall be:

1. Reactor Administrator. His individual shall meet the requirements of ANSI /ANS-15.4 "American National Standard for the Selection and Training of Personnel for Research Reactors" for a Level Two individual.

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2. Reactor Health Physicist. His individual shall meet the requirements of ANSI /ANS-15.4 "Amencan l National Standard for the Selection and Training of Personnel for Research Reactors" for a Level Three individual in addition to trahing in health physics.

b. A list of reactor personnel by name and telephone number shall be readily available to the UIUC Division of Public Safety dispatcher. One of these individuals shall be reachable and able to respond to the facility within approximately one hour. He list shall include:

1. Campus Radiation Safety Officer

2. Reactor Administrator

3. Head, Department of Nuclear Engineering

4. Reactor Health Physicist

5. Licensed operators l

c. Events requiring the presence at the facility of a Senior Reactor Operator; i 1. Initial startup and approach to power.

j 2. All fuel or control rod relocations..

3. Relocation of any in-core experincnt with a reactivity worth greater than one dollar.

4. Recovery from unplanned or unscheduled sh:itdown or significant power reduction (In these instances, documented verbal concurrence fmm the Senior Reactor Operator is required).

6.1.3 Selection and Trainina of Personnel he Reactor Administrator is responsible for the training and requalification of the facility reactor operators and senior reactor operators. he selection, training, and requalification of operations personnel shall be consistent with all current regulations and guidelines.

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6.2 Review and Audit 6.2.1 Charter and Rules

a. The Reactor Committee shall be composed of at least five voting members, one of whom shall be a Health Physicist designated by the campus Radiation Safety Officer for the University, one whom shall be the Reactor Administrator, and one whom shall be the Reactor Heahh Physicist. The remaining members shall be appointed by the Head of the Department of Nuclear Engineering, so as to maintain a balanced knowledge of reactor safety and regulation.

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b. The Reactor Committee shall have a written statement defining such matters as the authority of the committee, the subjects within its purview, and other such administrative provisions as are required for the ,

effective functioning of the Reactor Committee Minutes of all meetings of the Reactor Committee shall be l kept.

c. A quorum of the Reactor Committee shall be a majority of not less than one half of the members and the reactor operating staff shall not constitute a voting majority.

d. The Reactor Committee shall meet at least semiannually not to exceed nine months l

6.2.2 Review Function The review functLm c 'the Committee shall include, but is not limited to the following:

a. Determination that proposed changes ia equipment, systems, tests, experiments, or procedures do :M involve an unreviewed safety question.

b. All new procedures and major revisions thereto having safety significance, proposed changes in reactor facility equipment, or systems having safety significance.

c. All new experiments or classes of experiments for determination that an unreviewed safety question does not exist.

d. Proposed changes in the technical specifications or license.

e. Violations of technical specifications or license.

f. Operating abnormalities having safety significance.

g. Reportable occurrences as listed in 6.8.

h. Audit reports.

A written report or minutes of the findings and recommendations of the Conunittee shall be submitted to the Head, Department oGuclear Engineering, and the Reactor Committee members in a timely manner after each meeting.

-3 6-6.4 Procedures j

Written procedures shall be prepared, reviewed and approved prior to initiating any of the activities listed

' in this section.' The procedures shall be reviewed by the Reactor Committee and approved by the Reactor -l Administrator and such review and approval shall be documented in a timely manner. The procedures shall

. .be adequate to assure the safety of the reactor, but should not preclude the use ofindependent judgment and action should the situation require such.

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a. Stanup, operation, and sbutdown of the reactor.

b. Installation or removal of fuel elements, control rods, experiments, and experimental facilities.

c. ' Actions to be taken to correct specific and foreseen potential malfunctions of systems or components, including responses to alarms, suspected primary coolant leaks, and abnormal reactivity changes.

d. Emergency conditions involving potential or actual release of radioactivity, including provisions for evacuation, reentry-entry, recovery, and medical support.

e. Maintenance procedures which could have an effect on reactor safety.

f. PeOdic surveillance of reactor instrumentation and safety systems, area monitors and conunuous air monitors.

I' g. Personnel radiation protection, consistent with applicable regulations or guidelines. The procedures shall include management commitment and programs to maintain exposures and releases As Low As is Reasonably Achievable (ALARA).

! h. Implementation of physical security plan l

Substantive changes to the above procedures shall be made only after documented review by the Reador

! Committee and approval by the Reactor Administrator. Minor modifications to the original procedures I which do not change their original intent may be made by the Reactor Administrator. Temporary deviations from the procedures may be made by the responsible Senior Reactor Operator or higher individual present, in order to deal with special or unusual circumstances or conditions. Such deviations shall be documented and reported to the Reactor Administrator.

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-3 7-6.5 Ex_periments Review and Approval

a. All new experiments or class of experiments utilizing the reactor shall be evaluated by the experimenter and a staff member approved by the Reactor Committee. The evaluation shall be reviewed by a licensed senior reactor operator of the facility and the Reactor Health Physicist to assure compliance with the provisions of the utilization license, the Technical Specifications and 10CFR20. If, in theirjudgmerrt, the experiment meets with the above provisions and does not constitute a threat to the integrity of the reactor, they shall submit it to the Reactor Administrator for review. If the Reactor Administrator agrees with the evaluation by the senior reactor operator and the Reactor Health Physicist he shall submit the experiment to the Reactor Committee for review as indicated in Section 6.2. The experiment shall be approved in writing by the Reactor Administrator prior to initi ttion. When pertinent, the evaluation shall include the following:

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 chemical interaction that could occur with the reactor components.

4. Any radiation hazard that may result from the activation of materials or from external beams.

b. The Reactor Committee review of an experiment shall be performed prior to the first experiment and shall be documented in writing and shall consider at a minimum the following:

1. The purpose of the experiment.

2. A procedure for the performance of the experiment.

3. The evaluation approved by a licensed senior reactor operator.

4. Determination that the experiment does not involve an unreviewed safety question.

c. Substantive changes to previously approved experiments shall be made only after review by the l

! Reactor Committee and approved in writing by the Reactor Administrator. Minor changes that do not significantly alter the safety analysis of the experiment may be approved by the Reactor A Ninistrator.

d. For the irradiation of materials, the applicant shall submit a request to the Reactor Health I Physicist. This request shall contain at a minimum infonaation on the target material including the amount, chemical form, and expected radiological hazard for the desired irradiation period. For routine irradiations (which do not contain nuclear fuel or known explosive materials and which do not constitute a significant threat to the integrity of the reactor or to the safety ofindividuals), the approval for the Reactor Committee may be made by the Reactor Health Physicist.

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l 6.7 Action to be taken in the Event of an Abnormal Occurrence l

In the event of an abnormal occurrence, as defined in Section 1.14 of the specifications, the following action shall be taken: i

a. The reactor shall be shutdown and the Reactor Administrator shall be notified and corrective action taken prior to resumption of operations. The Reactor Administrator shall authorize resumption of i operations.

b. A report shall be made which shall include an analysis of the cause of the occurrence, efficacy of corrective action and recommendations for measures to prevent or reduce the probability of reoccurrence.

c, The occurrence shall be reviewed by the Reactor Committee at the next scheduled meeting.

d. A report shall be submitted to the USNRC in accordance with Section 6.8 of these specifications.

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6.8 RepArtingJteguirements In addition to the requirements of applicable regulations, and in no way substituting therefore, reports shall be mad'e to the USNRC as follows-

a. There shall be a report not later than the following working day by telephone and confirmed in l writing by facsimile or similar conveyance to the Regional Administrator, USNRC, Region III and the USNRC headquarters operations center to be followed by a written report that describes the circumstances of the event within 14 days to the Document Control Desk, USNRC Headquarters, and a copy to the Regional Administrator, USNRC, Region III of any of the following:

1. Release of radioactivity from the site above allowed limits.

2. Violation of safety limits.

3. Any significant variation from measured values from a corresponding predicted or previously measured value of safety-connected operating characteristics occurring during operation of the reactor.

4. 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.

5. Any abnormal occurrences as defined in Section 1.14 of these specifications.

b. A report within 30 days (in writing to the Document Control Desk, USNRC Headquarters) of:

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

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2. Significant changes in the transient or accident analysis as described in the Safety j l Analysis Report. 1 l

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3. Permanent changes in the facility organization involving theReactor Health Physicist, Reactor i Administrator or Department Head.

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c. ' A report within 60 days after criticality of the reactor (in writing to the Document Control Desk, USNRC Headquarters) upon receipt of a new facility license or an amendment to the license authorizing and increase in reactor power level or the installation of a new core, describing the

! measured values of the operating conditions or characteristics of the reactor under the new conditions, including:

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1. Total control rod reactivity worth. )

k 2. Reactivity worth of the single control rod of highest reactivity worth.

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