ML19327C053

Proposed Tech Specs,Including Section 2.1.b,Page 6,re Safety limit-fuel Element Temp & Limiting Safety Sys Settings

ML19327C053
Person / Time
Site:	Oregon State University
Issue date:	11/06/1989
From:	Oregon State University, CORVALLIS, OR
To:
Shared Package
ML19327C040	List: ML19327C039 ML19327C042 ML19327C053
References
NUDOCS 8911150165
Download: ML19327C053 (7)

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7. . Proposed New Pages- .

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R,! . . ' Amendment No,' ll','0STR Technical Specifications j

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'2. SAFETY LIMITS AND LIh! TING SAFETY SYSTEM SETTINGS f 2,1 SAFETY LIMIT-FUEL ELEMENT TEMPERATURE Applicability. This specification applies to the temperature of the reactor fuel. ,

Objective. The objective is to define the maximum fuel element tem- -

r'erature that can be pennitted with confidence that no damage to the i 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, r

-Bases. The important parameter for a TRIGA reactor is.the fuel ele-ment temperature. This parameter is well suited as a single specifi-cation especially since it can be measured. A loss in the integrity 1 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-()

m perature exceeds the safety limit. The pressure is caused by the presence of air, fission product gases, and hydrogen from the dissocia-tion of the hydrogen and zirconium in the fuel-moderator. The magni-tude of this pressure is detennined 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

.not exceed 2100 F (1150'C)provided the temperature and the fuel cladding is water of the fuel does 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

(')T SAR I - Amendment No. 4 to SAR, September 11, 1975 Amendment No. 11

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2.2 LIMITING SAFETY SYSTEM SETTINGS

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Ayplicability. This specification applies to the scram settings \

which prevent the safety limit from being reached. -

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

. Specification. The limiting safety system setting shall be 510*C '

(950*F) as measured in an instrumented fuel element. The instrumented fuel element shall be located in the B-ring.

' Bases. The limiting 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 this setting and the safety limits of 1150'C and 1000*C for FLIP and standard fuel, respectively. ,

The highest fuel temperatures are experienced during pulse transients, initiated frora 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" of the pulse. . A setting of 510*C is conservatively estimated to I provide the largest permissible pulses. These estimates are obtained i_ from calculations based on an adiabatic reactor kinetics model. This r

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( model when applied to existing cores yieltis characteristics which are in good agreement with measured values.

Input paramaters to this model for predicting mixed and full FLIP cores are obtained from flux profiles and prompt reactivity coeffi-cients calculated for these cores, and from information concerning cell parameters and prompt neutron. life time, which has 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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4, Amendment No. 11 1

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3. LIMITING CONDITIONS OF OPERATION 3.1 STEADY STATE OPERATION A>plicability. This specification applies to the energy generated in o tie reactor during steady state operation.

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

Specifications. The reactor power level shall not exceed 1.1 megawatts except for puTsing 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.

(oV)'

3.2 REACTIVITY LIMITATIONS Applicability. These specifications apply to the reactivity condition >

of the reactor and the reactivity worths of control rods and experiments.

They apply for all modes of operation.

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

l' Specifications. The reactor shall not be operated unless the following I 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 -

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3. the reactor in the cold condition without xenon.

?

Bases. The value of the shutdown margin assures that the reactor can i,

l 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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'Y 2. Review and approval of all proposed changes to the facility, J 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 abnornial performance of plant equipment and operating anomalies;

6. Review of all events which are required by regulations or v 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 IN 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; q

U 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- 1 mitted to the ROC for review and then submitted to the NRC when l authorization.is sought to resume operation of the reactor. j 6.4 ACTION TO BE TAKEN FOR REPORTABLE OCCURRENCES l i

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 n for measures to prevent or reduce the probability of reoccurrence, U This report shall be submitted to the Reactor Operations Committee for review.

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c. Where appropriate, a report shall be submitted to the NRC in accord-ance with Section 6.7 of these specifications.

-32 Amendment No. 11

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2. Those ever.ts reported as required by Sections 6.7.a.2 through i 6.7.a.8. l c/. A report within 30 da Washington,D.C.,witfsinwritingtotheNRC,DocumentControlDesk, L -

a copy to the NRC, Region V.  !

1

1. Any significant variation of measured values from a corresponding )

predicted or previously mea:ured value of safety-connected l operating characteristics occurring during operation of the j reactor; q

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

3. Any changes in facility organization or personnel; and l 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 (inwritingtotheNRC,DocumentControlDesk, Washington,D.C.

ano a copy to NRC, Region V) upon receipt of a new facility license, or e 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 including ,

'O An evaiuation of faciiity nerformance to date in comparison 1.

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, Document Control Desk Washington, D.C. and a copy to the flRC, ,

Region V).

1. A brief summary of operating experience including experiments 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-ve111ance test and inspections.

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

3. The number of emergency shutdowns and inadvertent scrams, including reasons therefore. l 1

O 4 Discussion of the major maintenance operations performed during the period, including the effect, if any, on the safety i of the operation of the reactor and the reasons for any cor-rective maintenance required.

Amendment No. 11

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