ML20005E216

Amend 11 to License R-106,reflecting Increase in Authorized steady-state Reactor Power from 1,000 Kw Thermal to 1,100 Kw Thermal

ML20005E216
Person / Time
Site:	Oregon State University
Issue date:	12/21/1989
From:	Weiss S Office of Nuclear Reactor Regulation
To:
Shared Package
ML20005E214	List: ML20005E213 ML20005E216 ML20005E218
References
R-106-A-011, R-106-A-11, NUDOCS 9001040102
Download: ML20005E216 (9)

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UNITED STATES

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.g-NUCLEAR REGULATORY COMMISSION

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e WASHINGTON, D. C. 20666 s'=

'j OREGON STATE UNIVERSITY DOCKET NO. 50-243

- 1 AMENDMENT TO FACILITY OPERATING LICENSE i

Amendment No. 11 License No. R-106-l 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment to Facility Operating License.

No. R-106 filed by Oregon State University (the licensee), dated November 7, 1988, as supplemented on November 6, 1989, complies with i

the standards and requirements of the Atomic Energy Act of 1954, as 4

amended (theAct),andtheCommission'sregulationsassetforthin i

10 CFR Chapter I; D..

The ' facility will o>erate in conformity with the ap)11 cation, the provisions of t1e Act, and the regulations of t1e Comission; C.

There is reasonable assurance:

(i) that the activities authorized by this amendment can be conducted without endangering the health i

and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied; and F.

Prior notice of this amendment was not required by 10 CFR 92.105(a)(4) and publication of notice of this amendment is not required by 10CFR$2.106(a)(2).

9001040102 891221 DR p

ADOCK 05000243 PDC

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2.

Accordingly, paragraph 2.C.(1) of License No R-106 is hereby. amended to read as follows:

(1) Maximum Power Level The licensee may operate the facility at steady state power levels not in excess of 1100 kilowatts (thermal) and withreactivityinsertionsnottoexceed2.55$.inthepulsemode,

3. -

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the enclosure to this license amendment, and paragraph 2.C.(2) of License No. R-106 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.11, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

4.-

This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION i

Seymour H. Weiss, Director Non-Power Reactor, Decommi.ssioning and

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Environmental Project Directorate Division of Reactor Projects - III, IV, V and Special Projects Office of Nuclear Reactor Regulation i

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Enclosure:

Appendix A Technical Specifications Changes Date of Issuance:

December 21, 1989 l

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ENCLOSURE TO LICENSE AMENDMENT.WD. 11 FACILITY OPERATING LICENSE NO. R-106 DOCKET NO. 50-243 l

Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contains vertical lines-indicating the areas of change.

Remove Insert 6

6 7

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8 12 12 32 32 35 35 l

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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i

2.1

-SAFETY LIMIT-FUEL ELEMENT TEMPERATURE v

ADolicability. This specification applies to the temperature of the

. reactor. fuel.

Obliective.

The objective is to define the maximum fuel element tem-perature that can be permitted with confidence that no damage to the fuel element cladding will result.

Specifications a.

The temperature in a TRIGA-FLIP fuel element shall not exceed 2100*F (1150'C) under any condition of operation.

b.

The temperature in a standard TRIGA fuel element shall not exceed _1830'F (1000'C) under any condition of operation.-

l Bases.

The important parameter for a TRIGA reactor is the fuel ele-3 ment temperature. This parameter is well suited as a single specifi-cation especially since it can be measured.

A loss in the integrity of the fuel element cladding could arise from a build-up of excessive pressure between the fuel-moderator and the cladding if the fuel tem-perature exceeds the safety limit.

The pressure is caused by the presence of air, fission product gases, and hydrogen from the dissoci-ation of the hydrogen and zirconium in the fuel-moderator.

The magni-tude of this pressure is determined by the fuel-moderator temperature and the ratio of hydrogen to zirconium in the alloy.

The safety limit for the TRIGA-FLIP fuel element is based on data which indicate that the stress in the cladding due to the hydrogen pressure from the dissociation of zirconium hydride will remain below the ultimate stress provided the temperature of the fuel does not exceed 2100'F (1150'C) and the fuel cladding is water cooled.

(SAR I)*

The safety limit for the standard TRIGA fuel is based on data

' including-the large mass of experimental evidence obtained during high performance reactor tests on this fuel. These data indicate that the stress in the cladding due to hydrogen pressure from the dissociation of zirconium hydride will remain below the ultimate stress provided that the temperature of the fuel does not exceed 1830*F (1000*C) and the fuel cladding is water cooled.

(SAR I)

• References to the Safety Analysis Report and its amendment will be abbreviated as:

SAR

- Safety Analysis Report, August 1968 SAR I - Amendment No. 4 to SAR, September 11, 1975 l Amendment No. 11 l

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• 2.2 LIMITING SAFETY SYSTEM S'TTINGS E

Apolicability. This specification epplies to the scram settings which prevent the safety limit from being reached.

Ob.iective.

The objective is to prevent the safety limits from being reached.

-q Specification. The limiting safety system setting shall be 510*C (950*F) as measured in an instrumented fuel element. The in:trumented fuel element shall be located in the B-ring.

Bases.

The~1imiting-safety system setting is a temperature, which, if exceeded shall cause'the reactor safety system to-initiate a reactor scram.

This setting applies to all modes of operation.

In steady-state operation up to 1.1 megawatts ample margins exist between I

this setting and the safety limits of Il50'C and 1000*C for FLIP and standard fuel, respectively.

The highest fuel temperatures are experienced during pulse transients, initiated from low power. The fuel temperature scram, to which the limiting safety system setting applies, can prevent reaching the safety limit of-the fuel by reducing the energy released in the " tail" l

of the pulse.

A setting of 510*C is conservatively estimated to provide the largest permissible pulses.

These estimates are obtained from calculations based on an adiabatic-reactor kinetics model. This model when applied to existing cores yields characteristics which are in good agreement with measured values.

Input parameters to this model for predicting mixed and full FLIP cores are obtained from flux profiles and prompt reactivity coeffi-4 cients calculated for these cores, and from information concerning cell parameters and prompt neutron life time, which have been established l

for similar cores elsewhere (Torrey-Pines, GA 9350, PRNC). The cal-culations leading to this value of the limiting safety system setting are outlined in detail in SAR I.

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3.

LIMITING CONDITIONS OF OPERATION

3.1 STEADY STATE OPERATION Anolicability.

This specification applies to the energy generated in the reactor during steady state operation.

Objective. The objective is to assure that the fuel temperature safety limit will not be exceeded during steady state operation.

Soecifications..The reactor power level shall not exceed 1.1 megawatts except for pulsing operations.

Basis.. Thermal and hydraulic calculations indicate that TRIGA fuel may be safely operated up to power levels of at least 2.0 megawatts with natural convection cooling.

3.2 REACTIVITY LIMITATIONS Acolicability.

These specifications apply to the reactivity condition of the reactor and the reactivity worths of control rods and experiments.

They apply for all modes of operation.

Objective.

The objective is to assure that the reactor can be shut down at all times and to assure that the fuel temperature safety limit will not be exceeded.

Soecifications. The reactor shall not be operated unless the following conditions exist:

The shutdown margin provided by control rods shall be greater than-$0.57 with:

1.

experimental facilities and experiments in place and the highest worth non-secured experiment in its most reactive

state, 2.

the most reactive control rod fully withdrawn, and 3.

the reactor in the cold condition without xenon.

Bases. The value of the shutdown margin assures that the reactor can be shut down from any operating condition even if the most reactive control rod should remain in the fully withdrawn position. Amendment No. 11

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TABLE I Minimum Reactor Safety Channels Effective Mode Safety Channel Function 5.5.

Pulse 5.W.

Fuel Element -Temperature SCRAM 9 510*C X

X X

Safety Power Level SCRAM 91100 kW(t) or less X

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Percent Power Level SCRAM e 1100 kW(t) or less X

X Console Scram Button SCRAM X

X X

Wide-Range Log Power SCRAM 9 period no less than X

Level 3 sec.

Preset Timer Transient rod SCRAM 9 15 sec X

or less after pulse High Voltage SCRAM 9 255 of nominal opera-X X-X ting voltage TABLE II Minimum Interlocks 2

Effective Mode Interlock Function 5.5.

Pulse 5.W.

l Wide-Range Log Power Prevents control rod X

Level Channel withdrawal 9 less than 2 cps Transient Rod Cylinder-Prevents application of X

air unless fully inserted 1 kw Pulse Interlock Prevents pulsing X

above 1 kw Shim, Safety, and Prevents simultaneous X

X Regulating Rod Drive withdrawal of two rods Circuit Shim. Safety, and Prevents movement of X

Regulating Rod Drive any rod except transient Circuit rod

- Transient Rod Cylinder Prevents pulse insertion of X

X Position reactivity greater than $2.55

-12 Amendment No.11

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2. _ Review and approval of all proposed changes to the facility, F

procedures, and Technical Specifications; 3.

Determination of whether a proposed change, test, or experiment would constitute an unreviewed safety question or a change in the Technical Specifications; 4.

Review of the operation and operational records of the facility; 5.

Review of abnormal performance of plant equipment and operating anomalies; 6.

Review of all events which are required by regulations or Technical Specifications to be reported to the NRC in writing within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; and 7.

Approval of individuals for the supervision and operation of the reactor.

6.3 ACTION TO BE TAKEN ifi THE EVENT A SAFETY LIMIT IS EXCEEDED In the event a safety limit (fuel temperatures) is exceeded:

a.

The reactor 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 Chair-man, ROC, and reports shall be made to the NRC in accordance with Section 6.7 of these Specifications; c.

A report shall be prepared which shall include an analysis of the causes and extent of possible resultant damage, efficacy of

-corrective action, and recommendations for measures to prevent or reduce-the probability of recurrence.

This report shall be sub-mitted to the ROC for review and then submitted to the NRC when authorization is sought to resume operation of the reactor.

6.4 ACTION TO BE TAKEN FOR REPORTABLE OCCURRENCES For all events which are required by regulations or Technical Specifications to be reported to NRC in writing within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the following action shall be taken:

a.

The Reactor Administrator or his designated alternate shall be notified and corrective action taken prior to resumption of the operation involved.

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.

This report shall be submitted to the Reactor Operations Committee i

for review.

c.

Where appropriate, a report shall be submitted to the NRC in accord-ance with Section 6.7 of these specifications. Amendment No. 11

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1 2.. Th~ose events reported,as required by Sections 6.7.a.2 through

~6.7.a.8.

1 c.

A report within 30 days in writing to the NRC, Document Control Desk, I

Washington,.D.C., with a copy to the NRC, Region V.

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Any significant variation of measured values from a corresponding 1

predicted or previously measured value of safety-connected operating characteristics occurring during operation of the reactor; 1

2.

Any significant change in the transient or accident analyses as described in the Safety Analysis Report; 3.

Any changes in facility organization or personnel; and 4.

Any observed inadequacies in the implementation of administrative or procedural controls such that the inadequacy causes or could have caused the existence or development of an unsafe condition with regard to reactor operations.

d.

A report within 90 days after completion of starting testing of the reactor (in writing to the NRC, Document Control Desk, Washington, D.C.

and a copy to NRC, Region V) upon receipt of a new facility license, or an amendment to the license authorizing an increase in reactor power level, describing the measured values of the operating conditions or characteristics of the reactor under the new conditions includirg:

1.

An evaluation of facility performance to date in comparison with design predictions and specifications.

2.

A reassessment of the safety analysis submitted with the license application in light of measured operating charac-teristics when such measurements indicate that there may

.be substantial variance from prior analysis, e.

An annual report by November 1 of each year (in writing to the NRC, l

Document Control Desk, Washington, D.C. and a copy to the NRC, Region V).

1.

A brief summary of operating experience including experiments l

performed and changes in facility design, performance charac-teristics and operating procedures related to reactor safety occurring during the reporting period, and results of sur-veillance test and inspections.

2.

A tabulation showing the energy generated by the reactor (in megawatt-hours), hours reactor was critical, and the cum-ulative total energy output since initial criticality.

3.

The number of emergency shutdowns and inadvertent scrams, including 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 cor-rective maintenance required. Amendment No. 11 1

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