ML19320D499

Tech Specs 3.3,4.1,4.2,4.8,6.2 & 6.3 Re Safety Limits, Limiting Conditions for Operation & Surveillance Requirements

ML19320D499
Person / Time
Site:	University of Virginia
Issue date:	07/15/1980
From:	VIRGINIA, UNIV. OF, CHARLOTTESVILLE, VA
To:
Shared Package
ML19320D495	List: ML19320D494 ML19320D499
References
NUDOCS 8007210433
Download: ML19320D499 (9)

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APPENDIX A FACILITY LICENSB NO. R-66 TECIINICAL SPECIFICATIONS FOR Tile UNIVERSITY OF VIRGINIA DOCKET NO. 50-62 DATE: J ULY, 1980 t

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• e TABLE OF CONTENTS Page 1.0 1 DEFINITIONS' 1 2.0 SAFETY LDlITS AND LIMITING SAFETY SYSTEM SETTINGS 5

-2._1 -Safety-Limits 5 2.2 Limiting Safety System Settings 8 3.0 LDlITING' CONDITIONS FOR OPERATION 9 3.1 Reactivity. 9 3.2 Reactor Safety System 11 3.3 Reactor Instrumentation 13

3.4 Radioactive. Effluents 15 3.S Confinement 16

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3.6 ; Limitations on Experiments 17

-3.7 Operation with. Fueled Experiments 19 3.8 IIcight of Water Above the Coro in Natural 21 Convection Mode of Operation i 3. 9 Rod Drop Times 22 3.10 Emergency Removal of Decay lleat 23 4.0 -= SURVEILLANCE REQUIREMENTS 24

. 4.1 ~ Safety Rods 24 1 4 .-2 Reactor Safety System 25-

4.3 Emergency Core Spray' System - 26 4.4' -Radiation Monitoring Equipment 27. .

? 4.S I Maintenance 28

-4.6 0 Con finement' -

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. 4. 7; ' Airborne ' Effluents 30.

[ -4~.8 -Reactor Fuel-Dosc' Measurements .30a {l

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3.3 Reactor Instrumentation

Applicability ~-

This application applies to the instrumentation which must be operable

. for safe operation of the reactor.

Objective l

The objective is to -require that sufficient information is available l to the operator to assure safe operation of the reactor.

Specification l

The reactor shall not-be operated unless the measuring channels described in Section 3.2 " Reactor Safety Systems" and in the following l tabic are operable.

Measuring Minimum Operating Mode in Channel No. Operable hhich Required Linear Power 1 All Modes Log N and- Period 1 All Modes Coro Gamma Monitor

• 1 All Modes Reactor Room Constant *-

Air Monitor- 1 All Modes

• _ Bridge Radiation Monitor 1 All Modes -

l 1 Reactor Face Monitor

• 1- All Modes

• Pool Water Level Monitor 2 Forced Convection Mode Pool '.later Temperature 1 All Modes

~ Primary Coolant Flow 1 Forced Convection Mode Start-Up Count Rate ,

1 Reactor Start-Up l

Reactor Power Level -2 All Modes l l

• The reactor room constant air monitor, reactor face monitor, and

- core gamma monitor may be out of service for a period not to exceed I

7. days wi.hout requiring reactor shutdown. If the reactor face monitor.cannot_bc repaired within 7 days, it may be replaced by a I locally alarming monitor of similar range for up to 30 days without '

requiring a reactor shutdown.

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3.10 Emergency Removal.'of Decay Heat Applicability This l specification applies to the emergency removal of decay heat.

_ Objective

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The objective is to assure that the flow rate from this system is sufficient to prevent. overheating of the fuel elements. subsequent to a total loss of primary water'from the core.

Specification

- There shall be two separate emergency core spray systems, each capable of maintaining a flow rate of at least 10 gpm over the 64 fuel element positions for.theffirst 30 minutes, and at least 7-1/2 gpm over the 64 fuel element positions

.for the next 60 minutes following a total loss of coolant._ These system shall be operable when the reactor is operated in the forced convection flow mode.

Bases

. Either of the tow spray system aa specified will provide sufficient cooling to maintain the-fuel tmeperature below its melting point as demonstrated by the evaluation in Section 9.9 of the SAR.

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4.0 . SURVEILLANCE REQUIREMENTS 4.1 . Shim Rods .

Applicability

. This specification applies to he surveillance requirements for the shim rods.

. Objective-To assure that the shim . ods are capabic of performing their function and-that no significant physical dcBradation in the rods has occurred.

Specification

a. Shim rod drop times shall be measured at' intervals not to exceed five months. Shia . rod drop times shall also be measured if the control assembly is moved to a new position in the ' core or if maintenance is performed on the mechr.nism.

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b. The shim: rod reactivity worths shall be measured whenever the rods are installed in a new core configuration.

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c. The . shim rods shall be visually. inspected at intervals not to exceed thirteen months, and when rod drop times exceed the limiting conditions for operation,-Section 3.9 of these specifications. If

.the shim rod is found to be deteriorated or - to have a crack of more than .1/4 inches in 'len ,ch,- it shall be removed from service.

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The. reactivity worth of. the _ chim rods is measured to assure that the required shutdown' margin is available' and to provide means for determining

the reactivity worth of experiments inserted in'.the - core. The visual inspection of. 'the shim rods and' measurement of their drop timos are made to determine whether the shim' rods are capabic of-performing properly. _

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4.2 Reactor Safety System Applicability This . specification applies to the sn -veillanco requirements for the rear. tor safety system of the reactor.

Objective.

The objective is to assure that the reactor safety system is operable as required by Specification 3.2.

Specification

a. A channel test of each of the reactor safety system measuring 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 each of the reactor safety system measuring channels shall be performed daily when the reactor is in operation.

c. A. channel calibration of the reactor safety measuring channels shal) he per-formed at intervals not to exceed eight months.

d. The power range channels 1 and 2 shall be checked against a_ primary system heat balance at least once each week the reactor is in operation above 100 kil'owatts in the forced convection mode,

e. The following items which are listed in sect'on 3.2 are not con-

.Sidered to be reactor safety measuring channels: Power to primary coolant' pump, manual button, header air pressure, and pool water level monitor. Operation of these systems will be checked prior to each days operation or prior to each operation extending more than Jone day.-

Bases- ,

The daily channel ' tests and channel checks will assure that the safety channels are operabic. The semi-annual calibration will permit any long-term drift of the channels- to be corrected. The weekly calibration of the power measuring channels will correct for drift and assure operation within the requirements of. the license.

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4.8 Reactor Fuel' Dose Measurements .

Applicability This specification applies to reactor fuel possessed under the Reactor' Facility Licenses.

Objective The objective of this specification is to ensure that the maximum quanity of special . nuclear material does not exceed the limits specified in the Facility licenses.

Specification A. The amount of special nuclear material (SNM) possessed at the Reactor Facility will be determined as necessary to ensure-that limits specified by the Facility licenses are not exceeded. As a minimum a evaluation,will be completed and documented ever six months.

B. Fuel elements will be irradiated as a part of the core or shipped away from the Reactor Facility as necessary to.

4 ensure that the quanity of nonexempt SNM (as defined in 10 CFR Part 73) does not exceed that allowed by the Facility licenses. If the amound of nonexempt SNM exceeds 5~.0 kg the Reactor Safety Committee will be informed.

C. Whenever fuel elements which have not been irradiated .as a part of the core for at least one month, adequate dose rate measurements.of representative fuel elements will be made as necessary., to determine which fuel elements have dose rates higher than specified by 10 CFR Part 73.67 (b).

Basis The specification will provide a high degree of assutance that the amount of SNM and nonexempt SNM does not exceed the' license limits. The amount of nonexempt SNM will normally lua . maintained at less that 5.0 kg. In the event that this

quanity is - exceeded - the Reactor Safety Committee will be informed and actions necessary to reduce the amount or

. other appropriate actions as defined in :the Physical ' curity

-Plan will'be-defined.. -

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fi 6.2 Feview and-Audit L-i'

a. There .shall be a Reactor Safety Committee which shall review and

' audit reactor operations to assure that the facility is operated in

a manner consistent with public safety and within the terms of the facility license. The Reactor Safety Committee shall report to the i

President of the University and advise the Chairma'n, Department of

- Nucicar Engineering and the Reactor Facility Director on those areas of responsibility specified below.

b. The Committee shall be composed of at 1 cast five members, one of whom shall be the Radiation Safety Officer of the University. No more than two members will be from the organization responsibic for

, Reactor Operations.

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The membership of the Committee shall be such l as to maintain a degree of technical proficiency in areas relating to' reactor operation and reactor safety.

c. A quorum of. the C o mmittee shall consist of not less than a majority of the full committee and shall include the Chairman or his designee.

d.

The Committee shall meet at least once every six months and on call-i by- the Chariman. Munutes of all meetings shall be disseminated to 1

I responsthic personnel as designated by the Committee Chairman.

c. -The Committee shall have a written statement defining such matters as the authority of the Committee, the subjects within its purview, l

and other such administrative provisions as are required for ~

cffective functioning of the Committee.

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' include minimum the responsibilities of the Reactor Safety Committee following:

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(1)[ Review and approval of un_tried experiments and tests which are 1

( 'significantly different from those previousiv used or tested in  !

! .the reactor as determined by the Facility Direc'or.

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i (2)' Review and approval of changes to the reactor core, reactor .

systems or- design features which may affect the safety of the reactor.

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6.3 Operating Procedures

a. Written procedures reviewed and approved by the Reactor Safety Committee, shall be in effect and followed for the items listed below. These procedures shall be adequate to assure the safe oper-ation of the reactor, but should not preclude the use of independent judgement and action should the situation require such.

(1) Startup, operation, and shutdown of the reactor, (2) Installation or removal of fuel elements, control rods, experiments, and experimental facilities, (3) Actions to be taken to correct specific and foreseen potential malfunctions of systems or components, including responses to alarms, suspected primary coolant system Icaks, abnormal reactiv-ity changes.

(4) Emergency conditions involving potential or actual release of radioactivity, including provisions for evacuation, re-entry, recovery, and medical support.

(5) Preventive and corrective maintenance operations which could have an effect on reactor sa fety.

(6) Periodic surveillance (including test and calibration) of reactor instrumentation and safety systems.

b. Radiation control procedures shall be maintained and made availabic to all operations personnel.

c. Substantive changes to the approved procedures shall be made only with the approval of the Reactor Safety Committee. Changes to the procedures which do not change their original intent may be made with the approval of the Facility Director. All such minor changes to procedures shall be documented and subsequently reviewed by the Reactor Safety Committee.

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