ML20198B213

Amend 17 to License R-52,updating Reporting Requirements for Change in NRC Organization & Correcting Typos

ML20198B213
Person / Time
Site:	05000113
Issue date:	12/18/1997
From:	Weiss S NRC (Affiliation Not Assigned)
To:
Shared Package
ML20198B204	List: ML20198B201 ML20198B213 ML20198B220
References
R-052-A-017, R-52-A-17, NUDOCS 9801060233
Download: ML20198B213 (11)

Text

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j \ UNITED STATES s* } NUCLEAR REGULATORY COMMISSION ,

g WASHINGTON. D.C. mM

% e , , , **j UNIVERSITY OF AHl2ONA DOCKET NO. 50113 AMENDMENT TO AMENDED FACILITY OPERATING LICENSE Amendment No.17 License No R 52

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for an amendment to Amended Facility Operating License No. R 52 filed by the University of Arizona (the licensee) on October 22,1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the regulations of the Commission as set forth in Chapter I of Title 10 of the Code of Federal Regulations (10 CFR): -

B. The f acility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public and (ii) that such activities will be conducted in compliance with the regulations of the Commission:

D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; E. The issuance of this amendment is in accordance with the regulations of the Commission as set forth in 10 CFR Part 51, and all applicable requirements have been satisfied; and F. Prior notice of this amendment was not required by 10 CFR 2.105 and publication of notice for this amendment is not required by 10 CFR 2.100.

k 9901060233 971218 PDR ADOCK 05000113 P PDR ,

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i University of Arizona Docket No, 50-113 -

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Office of the Mayor l P. O. Box 27210  !

Tucson, Arizona 85726 7210

.o Arizona Radiation Regulatory 3 Agency --  ;

4814 S. 40 Street  ;

' Phoenix, Arlsons .85040 I

Mr. Harry J. Doane, Reactor Supervisor -

Nuclear Reactor Laboratory -

University of Arizona Tucson, Arizona 85721. '

Dr. Michael Cusanovich Vice President for Research and Graduate Studies 1 University of Arizona .

. Administration Building, Room 601 P.O. Box 210066 -

Tucson, Arizona 85721 t

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2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the enclosure in this license amendment, and paragraph 2.C(2) of Amended Facuity Operating License No. R 52 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.1'/, are hereby incorporated into the license. The licensee shall '

operate the facility in accordance with the technical specifications.

3. This licenso amendment is effective on the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION e

Seymour H. Weiss, Director Non Power Reactors and Decommissioning Project Directorate Division of F] actor Program Management Office of Nuclear Reactor Regulation

Enclosure:

Appendix A Technical Specifications Changes Date of Issuance: Decernber 18, 1997 l

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i ENCLOSURE TO LICENSE AMENDMENT NOil7 AMENDED FACILITY OPERATING LICENSE NO. R 52 DOCKET NO. 50113

' Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain varticallines indicating the areas of change.

Remove . insert

- Table of Contents Table of Contents 30 30 32' 32 33 33 34 34 35 35 36 36 37 37 I

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TABLE OF CONTENTS page 1.0 DEFINITION S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2.0 SAFETY UMITS AND UMITING SAFETY SYSTEM SETTINGS . . . . . . . . . . . . . . . . 5 2.1 Safety Limit . Reactor Power Level . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2.2 Umiting Safety System Setting . Steady State Reactor Power Level . . . . . . . . . 6 2.3 Umiting Safety System Setting . Pulse Mode Reactor Power Level . . . . . . . . . . 7 3.0 UMITING CONDITIONS FOR OPERATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3.1 Reactivity Umits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8  !

3.2 High Power Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10

• 3.3 Pulse Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 3.4 Reactor Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 3.5 Reactor Safety System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 3.6 Ventilation System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 3.7 Experiments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 4.0 SURVEILLANCE REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 4 .1 Fu el . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 4.2 Control Rod s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 4.3 Reactor Safety System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 4.4 Radiation Monitoring Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 4.5 Mainte nanc e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 4.6 Pool Water Conductivity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 5.0 DESIGN FEATURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 5.1 Re actor Fuel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 5.2 Reactor Building . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 5.3 Fuel Storage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 6.0 ADMINISTRATIVE CONTROLS . . . . . . . . . . , . . . . . . , . . . . . . . . . . . . . . . . . . . 26 6.1 Organization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 6.2 Re view . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 7 6.3 Operation s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 a._ Operating Procedure s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28

b. ALARA Pro gram . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 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 6.6 Plant 0perating Records . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 6.7 Reporting Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 6.8 Review of Experiments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36

o 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 made whlch 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 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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page 32 6.7 Reporting Requirements in addition to the requirements of applicable regulations, and in no way substituting therefor, reports shall be made to the NRC as follows: '

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. 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;

2. Any violation of a Safety Limit; and

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

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

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, personal injury, 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 Non Power Reactors and Decommissioning Project Directorate, of:

1. Any substantial variance from performance specifications contained in these specifications or in the Safety Analysis Report; .

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2. Any signWicant change in the transient or accident analysis as desenbod in the Safety Analysis Report; l

3. Any changes in facility organization; and

4. - Any observed inadequacies in the implementation of administrative or proceduralcontrols. 1 t

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

. the d. S. Nuclear Regulatory Commission, Attn: Document Control Desk, Washington D.C. 20555, with a copy to the Non Power Reactors and l Decommissioning Project Directorate.  !

1. ' An evaluation of facility performance to date in comparison with design iWstions and specifications; and j i

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

e. 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. 20555, with a copy to the Non Power Reactors and Decommissioning Project  !

Directorate.  :

1. A brief narrative summary of (1) opt Ming experience (including experiments .

i performed), (2) changes in facility wsign, perfonnance characteristics, and t

operating procedures related to reactor safety and occuning during the 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 -l output since initial criticality; e

3. The number of emergency shutdowns and inadvertent scrams, including I

reasons therefore;

4. Discussion of the major maintenance operations perfonned 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. A 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 to Section 50.59 of 10 CFR Part 50; J I

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6. A summary of the nature and amount of radioactive affluents released or dise"ed to the environs beyond the efiective ' control of the licensee as m j at or prior to the point of such release or discharge; I innid Watte (summarized on a monthly basis)

(a) Radioactivity discharged during the reporting period.

(1) Total radioactivity released On curies).

I (2) Tbe MP6 limiting concentrations (10CFR20, fppendix 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 of release.

GasenuSEaste (summarized on a monthly basis)

(a) Radioactivity discharged during the ieporting period on curies) for:

(1) Gases.

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

(b) The MPG limiting concentrations (10CFR20, AppendixB) used and the b estimated activity On curies discharged du:ing 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 On cubic feet).

(b) The total activity involved On curies).

(c) The dates of transfer or shipment and disposition ummm me WihofMnone Reneeroh Menotor -,wr

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7. A swnmary of radiation exposures received by facility personnel and visitors,  !

I includmg dates and time of significant exposures, and a summary of the results of radiaton and contaminaton surveys performed within the facility; and l

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

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page 36 l l 6.8 Review of Experiments i

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

experimenter and the Reactor Committee. The evaluation 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

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; and

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

5. A determination that for the maximum planned or inadvertent pulse, no cred%Ie 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; 1

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 applicant 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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d. In evaluating experiments, the following assumptions should be used for the purpose of determining whether failure of the experiment would cause the appropriate limits of 10 CFR 20 to be exceeded:

1. W the possibility exists that airtome concentrations of radioactive gases or aerosols may be released within the facility,100 percent of the gases or aerosols will escape;

2. N the effluent exhausts through a fiiter installation designed for greater than 99 percent efficiency for 0.3 micron 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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