ML20043G881

Proposed Tech Specs & Bases Re Changes to Reactor Operating Power

ML20043G881
Person / Time
Site:	Ohio State University
Issue date:	06/12/1990
From:	OHIO STATE UNIV., COLUMBUS, OH
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ML20043G879	List: ML20043G878 ML20043G881
References
NUDOCS 9006210234
Download: ML20043G881 (44)

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TAltl.E OF CONTENTS 1.

INTRODUCTION.

1 I

1.1 Scope I

1.2 Application 1

1.2.1 Purpose 1

1.1.2 Format 1

1.3 Definitions 2

2.

SAFETY LIMIT AND I.IMITING SAFETY SYSTEM SETTINGS (LSSS) 7 2.1 Safety Limit.

7 7

2.2 Limiting Safety System Settings 3.

LIMITING CONDITIONS FOR OPERATION 9

3.1 Reactor Core Parameters 9

3.1.1 Reactivity 9

3.2 Reactor Control and Safety System 11 3.2.1 Control Rod Drop Times 11 3.2.2 Maximum Reactivity insertion Rate.

Il 3.2.3 Minimum Number of Scram Channels 11-3.3 Coolant System.

14 3.3.1 Pump Requirements 14 3.3.2 Coolant f.evel 14

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3.3.3 Water Chemistry Requirements 14 3.3,4 Lenk or Loss of Coolant Detection.

15 3.3.5 Primary and Secondary Coolant Activity !.imits 15 d

3.4 Confinement Isolation 16

.3.5 Ventilation Systems 16 3.6 Radiation Monitoring Systems and Radioactive Effluents.

17 3.6.1 Radiation Monitoring 17 3.6.2 Radioactive Effluents 18 3.7 Experiments 19 3.7'1 Reactivity Limits 10-3.7.2 Design and Materials 20 4.

SURVEILLANCE REQUIREMENTS 21 i

4.1 Reactor Core Parameters 21 4.1,1 Excess React ivit y and Shutdown Margin 21 4.1.2 Fuel Elements 21 4.2 Reactor Control and Safety Systems.

22 4.2.1 Control Rods 22 4.2.2 Reactor Safety System.

22 4.3 Coolant System.

23 4.3.1 Primary Coolant Water Purity 23 4.3.2 Coolant System Radioactivity 24 4.4 Confinement 24 4.5 Ventilation System 24 4.6 Radiation Monitoring Systems and Radioactive Effluents 25 4.6.1 Effluent Monitor 25 4.6.2 Rabbit Vent Monitor.

25 4.6.3 Area Radiation Monitors (ARMS) 25 4.6.4 Portable Survey Instrumentation 26 i

5.

11ES16N FLATURES 27 5.1 Site and Facility Description 27 5.1.1 Facility I,ocation 27 5.1. 2.

Exclusion and Restricted Area 27 5.2 Reactor Coolant System.

27 S.2.1 Primary coolant Loop 27 5.2.2 Secondary and Tert inry Coolant Loops 27 5.3 Reactor Core and Fuel 27 5.4 Fuel Storage.......

28 5.5 Fuel Handling Tools 28 6.

ADMINISTRATIVE CONTROLS 29 6.1 Organization..

29 6,1,1 St ruct ure 29 6.1.2 Responsibility 20 6.1.3. Staffing 29 6.1.4 Selection and Training of Personnel 3) 6.2 Review and Audit.

31 6.2.1 Composition and Qualifications of the ROC 31 6.2.2 ROC Meetings 31 6.2.3 Su,1; Committees 32 3

4 6.2.4 ROC Review and Approval Funct ion 32 6.2.5 ROC Audit Function 33 6,3 Procedures.

34 6.3,1 Reactor Operating Procedures 34 6.3.2 AdministralIvo Procedures 35 G 4 Experiment Review and Approval,

35 6.4.1 Definitions of Experiments 35

~

6.4.2 ~AJproved Experiments 35 0.4.3 New Experiments 36 6.5 Required Actions.

36 6.5.1 Action To Be Taken in the Event A Safety Limit is Exceeded 36 6.5.2 Action To Be Taken in The Event Of A Reportable Occurrence 36 6.6 Reports 37 6.6.1 Operating Reports 38 6.6.2 Special Reports 38 6.7 Records 40 6.7.1 Records to be Retained for a Period of at I,ea s t Five Vents 40 6.7.2 Records to be Retained for at Lenst One Requalificatton Cycle

'40 6,7.3 Records to be Retained for the Life of the Facility 40 13 m

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J 1.

INTRODUCTION 1,1-Scope This document constitutes the Technical Specificatjons for Facil i ty I,1 cense No. R-75 and supersedes all prior Technical Specifications.

Included are the " Specifications" and the l

" Rases" for the Technical Specifications.

These buses, which provide the technical support for the individual tcchnichl j

specifications, are included for information purposes on]y.

They are not part of the Technical Specifications, and they do not constitute 1 imitations or requirements to which the

. licensee must adhere.

1 i

This document was written to be in conformance with ANSl/ANS-15.1 --~19 8 2.

The content of the Technical Specli'jcations includes; Definitions, Safety himits, himit ing Safety System Settings, himiting Conditions for Operation, S u rve i l l an c e.

Requirements DesfEn Features, and Administrative Controls.

1 i

1.2 Application

- i i

1. 2.1.

purpose These Technical Specifications have been (Citten specifically for The Ohio State University Research Reactor (OSURR).

The Technical Specifications represent the agreement between the licensee and the U.S.

Nuclear Regulatory Commission on administrative controls,. equipment availability,- and operational parameters.

q Specifications are limits and equipment requirements for safe j

reactor operation and for dealing with abnormal situations.

They are typically derived from the Safety Analysis Report (SAR).

These specif(catlons represent a comprehensive envelope.

l for safe operation.

Only those operational parameters and

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equipment requirements directly related to preserving that safe

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envelope are listed, i

1.1.2 Format i

The formnt of this document is in general accordance with j

ANS T /ANS -15.1 1982 ;

i 1,

1.3 Definitions-Administrative Controls - those organizat ional-and procedural reqairements established by the Commission and/or the facility management.

AI,AllA - as low as is reasonably achievable.

Channel _ - the combination of sensor, line, amplifier, and output devices which are connected for the purpose of measuring the value of a parameter.

Channel Calibration

.un adjustment of the channel such that Its_ output corresponds with acceptable accuracy to known values of the measured parameter.

Calibration shall encompass the ent ire channel, -including equipment actuation, alarm, or trip settings, and shall be deemed to include a channel test.

Channel check - a qualitative verification of acceptable performance by observation of channel behavior.

This verlfication, where possible, shall -include comparison of the channel with other independent channels or systems measuring the same variable.

Channel Test - the introduction of a signal into the chunnel for veri #ication that it is operable.

Cold Clean Core - when'the core is at ambient temperature and

.the reactlvity worth of xenon is negligible.

Commission - the U.S. Nuclear Regulatory Commission (or NRC).

Confinement - a closure on the overall facility which controls the movement of air into it'and out of it through a controlled path.

Containment - a testable enclosure which can support a defined pressure differential and which is normally closed.

Control Rod - a device f abr icat ed from neutron absorbing material which is used to establish neutron flux changes.

Cont rol Rod Fuel Element - a' fuel element capable of holding a control rod, b

Controls - mechanisms used to regulate the operation of the=

reactor Hi Core the general arrangement of fuel elements und control rods.

Cri t leal - when the offeetIve mult1plicatton Iactor (keff) f L

the reactor is equal to unity.

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Ilirect Supervision - in visual and audible contact.

Excess Reactivit y - that amount of reactivity that would exist if all control rods were removed from the core.

Exclusion Area - that area around the reactor building in which the licensee has the authority to determine all activities as per 10CPR100.3, Experiment - any operation, or any apparatus, device, or material installed in or near the core or which could conceivably have a reactivity effect on the core and which itself is not a core component or experimental facility,.

Intended to investigate non-routine reactor parameters or radiation :nteraction parameters of materials.

Experimental Facility - any structure or device associated with the reactor that is intended to guide, orient, position,

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manipulate, or otherwise facilitate completion of experiments, l

Explosive Material - any material that is given an Identification of Reactivity (Stability) index of 2, 3, or 4 by the National Fire Protection Assocjat.lon in its pub.li ca tion -

704-M, Ident ificat ion System for Fire Hazards of Materials, or i

is enumerated in the Handbook for Laboratory Safety published f

by the Chemica1. Rubber Company (1967).

j I

Facility - the Reactor Hullding including offices and 1

laboratories.

Fueled Experiment - any experiment that contains U-23f> or 11-233 or Pu-239, not including the normal reactor fuel elements.

ll

. Licensee - The Ohio State University.

Limiting Conditions for Operation (LCO) - the lowest functional capability or performance levels of equipment required for safe operation of the facility.

LCO are administratively 4

established constraints on equipment and operational characteristics.

Limitlug Safety System Settings (LSSS) - settings for automatic protective devices. related to those variables having significant safety functions.

Where a limit ing safety system setting is specified for a variable on which a safety limit has been placed, the setting shall be so chosen that automatic L

protective action will correct t he abnormal situation before ~ a l,

safet y 1imit is exceeded.

Measured Value -- the value of a parameter as it appears on the output of a channel.

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1 Movable Experiment - one for which it is intended that all or part of the experiment may be moved in relation to the core while the reactor is operating.

Nuclear Regulatory Commission - (NRC).

Onset of Nuc1eate 11011ing - (ONil).

Operable - a component or system which is capable of performing i

its intended functions in a normal manner.

Operating - a component or system which is performing its intended functlon.

i protective Action - the initiation of a signn) or the operation of equipment within the reactor safety system in response to a l

varlable or condition of the reactor f acility having reached a specified limit.

I 4

Reactivity Limits - those limits imposed on reactor core excess i

reactivity-based upon a reference core condition.

l Henct tvity Worth of an Experiment - the maximum. absolute value of the reactivity change that would, occur as a result of i

intended or anticipated changes or credible malfunctions that alter an experiment's position or configuration.

l Henctor - the combination of core, permanently _

control 1

installed experimental facilities, control rods, and connected instrumentation.

Henetor Operating - whenever the reactor is not secured or shutdown.

Heact or Operat ions Committee - (ROC).

Reactor Operator (RO) - an individual who la licensed to manipulate the controls of the reactor in accordance - with

10CFR55, Reactor Safety Systems - those systems, including their

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associated input channels, which are designed to initiate automatic reactor protection or to provide information for initintion of manual protective action.

React or Secured - whenever (1) all shim / safety rods are fully inserted, (2) the console key is in the OFF position and is removed from the lock, and (3) no in-rore work is in progress involving fuel or experiments or m a i n t.e n c..e c e of the core structure, cont rol rods, or control rod drive mechanisms.

Reactor Shudown - when the reactor is suberit leal hy at least

.1% delta k/k in & uotd clean core condition.

4

Regulating Rod - a low reactivity-worth control rod used primarily to maintain an intended power level.

Its position may be varied either by manual control or by the automatic servo-controller, Reportable Occurrence - any of the conditlons described in c

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Section 0.5.2 of these specifications.

Restricted Area - the Reactor Dul lding -tr.

which access is controlled for purposes of prot ec t i o n of-individuals from exposure to radiation and-radioactive materials.

Safety Analysis Report - (SAR).

Safety Channel -

a measuring or procective channel in the reactor safety system.

Safety Limits (Sh) limits on important process verlables which are found to be necessary to reasonably protect the integrity of certain physical barriers which guard against the uncontrolled release of radioactivity.

Scram - the rapid insertion of the shim /sufety rods into the reactor for the purpose of quickly shutting down the reactor.

the elapsed time between reaching a limiting Scram-Time safety system setting and the time when a control rod is fully inserted.

Secured F.xperiment - any experiment, experimental facility, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means.-

The restraining forces must be substantially-greater than those to which the experiment might. be subjected from the normal environment of the experiment or by forces which can result from credible malfunctions.

Senior Reactor Operator (SRO) an Individual who is licensed to direct the activities of reactor operators.

Such an individual may also operate the controls of the reactor pursuant to 10CFR55.

Shall, Should, and May - the word "shall" is used to denote a requirement; the word "should" to denote a recommendation; and the word "mny" to denote permission, which is neither a requirement nor a recommendation.

Shim / Safety Rods - hi gh -react iv i t y worth control rods used prima r i l y to provide coarse reactor control.

They are connected elect romagnet ically to their delve mechanisms and have scram capabilities.

5

E ter.1-Shutdown Margin - the shutdown reactivity necessary to provide confjdence that the reactor can be made suberitical by,means of the control and safety systems with the most reactive shim / safety-rod and the regulating rod in the most = reactive position (fuljy wit hdrawn) and that the reactor will remain suberitical without further operator action.

Standard Fuel Element an element to be used or stored in the core, fuel storage pit or other approved area, but not a control rod element.

Start up Source - a spontaneous source of neutrons which is used to provide a channel check of the startup (fission chamber) channel, and provide neutrons for. subcritical multiplication during reactor startup.

Surveillance Time Intervals - The average over any extended period for each surve.Illance time interval shall be closer to q

the. normal-surveillance time, e.g.

for the two year interval the averag-t shall be closer to two years rather_than 30 months.

two-year (in'terval not to exceed 30 months),

annually (interval not to exceed 15 months).

semiannually (interval not to exceed 7-1/2 months),

i quarterly (interval not to exceed 4 months).

monthly (interval not to exceed G weeks).

weekly (interval not to exceed'10. days).

daily (shall be done during the same working day).

Any extension of_ these intervals shall be ' occasional and for a valid reason. and shall not affect the average _ as 1

defined, a

True Value - the actual value of a parameter.

m Unscheduled Shutdowns - any unplanned _ shutdown of the reactor caused by actuation of the reactor safety' systems, operator error, equipment malfunction, or a manual shutdown in response l'

to condit-lons which could adversely affect safe operation.

11 ;

.They do not include those shutdowns resulting from expected W

_ testing operations, or planned shutdowns, whether initiated by 4

' controlled' insertion of control rods or planned manual scrams.

l

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1 6

.w 2.

SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTINGS (LSSS) 2.1 Safety Limit Applicability:

This specification applies to the melting

. temperature of the aluminum fue) cladding.

Objective:

The objective la to arsure that the integrity of the fuel cladding is maintained.

Specification:

The reactor fuel temperature shall be. less than 550 C.

Bases: The melting temperature of aluminum is 660 C (1220 F).

The blister threshold temperature for U yj dispersion fuel

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g has been measured as approximately 550 C.2 (ANL/RERTR/TM-10, October 1987, NRC NUREG 1313).

Because the objective of this npocification is to prevent release of fission products, any fuel whose maximum temperature reaches 550"C.

~

is to be treated as though the safety limit has = been reached until shown otherwise.

2.2 Limiting Safety System Settings Applicability:

This SpeciflCatlOn applies to the folloWing items associated with core thermodynamics:

(1) Reactor Thermal power Level and (2) Reactor' Coolant inlet Temperature.

Objective:

To assure that the fuel cladding integrity is maintained.

Specificatlon:

(1) Steady state power level shall not exceed 500 kw thermal.

(2) Henctor safety systems set tings shall. Ini t. late automatic protective action so that reactor thermal power level shall not exceed 600 kw (120% of full power) during a translent.

(3) Reactor safety systems settings shal1 initlate automatic protective action so that core inlet water temperature shall not exceed 35"C Buses:

The eriterion for these safety syst em settings is established as the fuel integrity.

If the temperature of the clad is maintained below that for blister threshold then cladding integrity is maintained.

This is the case for a powee level of 600 kw and a core inlet temperature of 35"O (normal 7

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(

inlet temperature j a = 20-25'C).

The taximum credible accident

. analysis is provided in Section 8.4.3J of-the Safety Analysis Heport.

The-maximum credible accident assumes steady state operation at 600 kw and a transjent to 750 kw.

The maximum temperature of the cladding reaches 91 C (SAR 8.4.3 J).

One may also reference SAR Sections 4.8.1, 4.8.2 for an 'astimate of cladding temperature during steady state operathn at 500 kw (56.5"C).

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4 8

o-3.

LIMITING CONDITIONS FOR OPERATION 3.1 Reactor Core Parameters P.1.1 Reactivity Applicability:

These specifications apply to the reactivity condition of the reactor and the reactivity worths of the shim /sufety rods and regulating rod under any operating conditions.

Objectlve: To ensure safe shutdown of the reactor and that the safety 1imits are not exceeded.

Specification:

The reactor shall be operated only if the following conditions exist:

(1)- The reactor core shall be loaded so that the excess reactivity, including the effects of installed experiments does ' not exceed 2.64.

delta k/k under any operating condition.

(2) The minimum shutdown margin under any operating conditton with the maximum worth shim / safety rod and the regulating rod full out shall be no less than 1.0%, delta k/k.

(3) The total reactivity worth of the regulating rod shall be less than 0.% delta k/k.

(4)- All core grid' positions internal to the active fuel boundary shall be occupied by a standard, control, regulating rod,. instrumented,.or blank fuel element; or by an experimental facility.

-(5) The moderator temperature coefficient shall be negative and-shall havy5 a m n mum a s ue reacuvl y value of at Jeast 2 x 10 / C across the active core at all normal operating temperatures.

(6)

The moderator void coefficient of reactivity shall be negative g/1% void across the active core.

nd shall'have a minimum value of at least 2.8 x 10 Bases:

(1 ) - The. maximum allowed excess reactivity of 2.6% delta k/k provides sufficient reactivity to accommodate fuel burnup, xenon bulldup. experiments. control requirements, and' fuel and moderator temperature feedback (Section 4.2 of the SAR). Also, calculations show that this excess reactivity assures that the maximum temperature of the surface of the cladding will t'a well below the blist er threshold of the U S1 fuel dur ing a design basis accident (SAR 8.4.3.2).

3 2 9

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=c; (2) The minimum shutdown margin ensures that the reactor can be shutdown from any operating condition. and remain-shutdown after cooling and xenon decay - even with the highest worth rod and the regislating rod fully withdrawn, (3) I.imiting tim reactivity worth of the regulating rod to a value less than the effective delayed neutron fraction assures that a failure of the automatle servo control system cannot result in a prompt critical condition.

(4) The requirement that all grid positions be filled during l

reactor operation assures that the volume flow. rate ' of primary coolant which bypasses the heat. producing elements

-will be within the range specified in Section 4~.8 of the SAR.

Furthermore, the possibility of accidentally dropping an object into a grid posi tion-and causing increuse of reactivity is precluded.

.i (5) A,negat Ive moderator temperature coef ficient. of reactivity assures that any moderator temperature rise will. cause a

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decrease in reactivity.

The U Si fuel, also ' has a significantnegativetemperaturecockf1clentofreactivity-2 due to the g ppler broaduning of neutron capture

. resonances in U, but no credit is taken for this effeet in our Safety analyses.

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A' ~ nega t ive void coefficient of reactivity helps provide reactor stability in the event of moderator displacement by experimental devices or other means.

10

c 3.2 Reactor Control and Safety System 3.2.1-Control Rod Drop Times Applicability:

This'specjfication upplies to-the time from the receipt of a safety signal to the time it takes for a

= shim / safety rod to drop from fully withdrawn to fully inserted.

s ObjectJve: To ensure that the reactor can be shutdown within a spectried period of time.

Specification:

The reacto' will not be operated unless the drop time of each of the three shim /sufety rods is less than GOO msec.

flases:

Control rod c' rop times as specified ensure that the safety limit will not be exceeded in a short period transient.

The analysis for=this is given in Section 4.3.3 of the SAR.

3.2.2, Maximum Reactivity insertion Rate Applicabl]ity:

This-upplies to the maximum positive reactivity insertion rate by the most reactive shim / safety rod and the regulating rod simultaneously.

Objective:

To ensure the reactor is operated safely and the safety limit is not exceeded due to a short period.

Specification:

The reactor will not be operated unless the maximum reactivity insertion rate is less than 0.02% delta k/k per second.

Itas i s : This maximum reactivity insertjon rate assures that the Safety Limit will not be exceeded during a startup accident due i

to a short period generated by a continuous linear reactivity

'insertlon.

3.2.3 Minimum Number of Scram Channels Applicability:

This specification applies to the reactor safety system channels.

Objective:

To stipulate the minimum number of reactor safety system channels that shall be operable to ensure the Safety Limits are not exceeded by ensuring the reactor can be shutdown at all times.

i Specification:

The reactor shall not be operated unless the safety system channels described in the following table are operable.

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. Reactor Safet'y System Minimum Function Component Required 1.

Core 11 0 Inlet Temp.

1 Slow scram if temp. 135 C 5

2 2.

Reactor Thermal 2

Fast scram-if thermal power power level 1 000 kw, as indicated on

=(Safety Channels) calibrated junization chamber channels, j

3.

hoactor Period 3

Fast scram if period 5 1 sec l

4.

Reactor Thermal 1

Slow scrum if coolant system

^

power level / coolant pumps not on by 1 320 kw system pumps thermal power

-5.-

Coolant Flow Hate 1

Slow scram if coolant system has no flow (primary) by-1 120 kw thermal' power 0.

Pool Water Level 1

Slow scram if pool level 5 20 feet (15 feet above core)

.7.

Switches 6

Slow scram jf any one switch n.

Magnet Power Key "On" is not properly set, at. the b.

Startup Cal-Use position indicated in

' In "Use" quotes.

(Also prohibits c.

Perlod Generator startup).

Switch."Off" d.

LOG-N Amp Calibrate Switch " Norm" e.

LOG-Perjud Amp Calibrate Switch " Norm"

'f.

Elfluent Monitor Compressor "On" 8.

Recorders 5-Slow scram if power 1s lost

-a.

LOG-N to any one of - the listed b.

Linear' Level recorders c.

Start-Up Channel d.

Period U

e.

Effluent Monitor

-N.

ManuulfScrams 5

Slow scram upon 'act ivat io..

,o a.

Control Room Console of any one manual scram J

b.

Pool Top Catwalk swltch c.

BSP Catwalk d.

Rabbtt/ IIP Area e.

- Thermal Column /BP Area 10 Compensated Ion Chambers 2

Slow scram if voltage drops below operationa1 specifIcat1ons 12 w

'f

Reuctor-Safety System Minimum Function

Component / Channel-Required-11 Safety Set points On 4

Slow scram if associated Recorders recorder values are exceeded a,

period 1 5 see b.

Linear Level 1 120% of licensed power.

c.

Linear Level Servo _ deviation 1 Set point (nominal 10%).

d.

Start-ilp Channel 5 2 cts /sec (may be bypassed if K 77 < 0.9) 12.

Safety System 2

Slow scram in case of a safety amp fault or-if system is discontinuous 13.

Backup Shutdown 3

Rod drop will occur for Mechanisms any control rod which has excess magnet current 2 60 ma Bases:

1.

Assures' safety limit is not exceeded; core inlet temperature is same as cooling system outlet

~2.

Assures safety limit is-not exceeded 3.

-Assures safety limit is not exceeded 4.

Assures coolant-system pumps are functional before raising power >

120 kw.

5.

Assures there is always primary coolant flow when greater than 120 kw.

G.

Assures _there is enough primary coolant for natural convection cooling

~7.

. Assures nuclear Instrumentation is in proper mode for operation

.8.

Annures jnformation ' is 'available for observation by the reactor operator during operation, and is recorded if required as a record of reactor operations-

-9.

' Assures'that the reactor can be shut down by the reactor operator in the control room or at other locations near experimental facilities if deemed necessary by,other reactor staff 10.

Assures shutdown if nuclear inst rumentation falls J

11.

Assures backup shutdown capability' from short period or high power level.

Assures ' shutdown if servo operation varies too greatly.

Assures shutdown if count rate is too low to provide meaningful startup Information, The startup interlock may be bypassed if K is <.9 12.

Assures all components of the safety system are "fr$ stalled and operatlonal 13.

Assures that any control rod exhibiting excess magnet current will be released und fall to the bottom due to gravity l

13

]

a

r 3.3 Coolant System 3.3.1 pump Requirement s '

Applicability:

Thi specification applies to the operation of a

pumps for both-t'ie primary and secondary coolant loops.

Object ive:

To ensure that both pumps are functioning whenever the reactor is operated above 12p kw.

Specification:

The reactor will not be operated above 120 kw unless both the primary and secondary coolant pumps are actjvated and there is flow in the primary coolant loop.

l Bases:

Having both pumps operating and flow in the primary loop will ensure there is adequate cooling of the primary coolant so the Safety Limit is not exceeded.

,i 3.3.2 Coolant 1, eve l Applicability:

This specification applies to the height of the water in the Reactor Pool above the core.

I Object ive:

To ensure there is adequate primary coolant in the Reactor. pool and sufficient biological shielding above the core SpecifIcatlon:

The react or sha11 not ube. operated unless there la 20 feet of water jn the reactor. pool and 15 feet of water W

above the core.

Bases:

With the pool full of water t o a jovel of 20 feet there is adequate primary coolant for - natural convection cooling.

With 15 feet of water above the core there is sufficient shielding at-the_ licensed power level. -Section 7.1.1.4 of the SAR discusses this shielding.

f 3.3.3 Water Chemistry Requirements Applicability:

This specification applies to the purity of the primary coolant water.

Object ive; To minimize corrosjon of the cladding on the fuel elements, and to reduce the probability of neutron activation of ions in the water.

14

. e; Specification:

(1) The conductivity of the pool water shall not exceed the I'

limit of 2.0 mho/cm.

(2) The pil of the pool. water shall not exceed 8.0.

Bases:

-Operation in accordance with these specification

.[

ensures aluminum corrosion-is within acceptable limits, and that. the concentration of dissolved impurities that could be activated by neutron irradiation _ remains within acceptable limits.

3.3.4 I.eak or 1,oss of Coolant Detection Applicability: This specification _ applies to the capability of detecting and preventing the loss of primary coolant.

Objective:

To ensure there is adequate primary coolant In the Reactor pool and sufficient biological shielding above the core when the reactor is operating.

Specificutlon:

The pool water level shall be at least 15 feet above the top of the fuel in the core.

Dases:

The same system that functions _to scram the reactor on low pool nievel will also be used as the detection system for this specification.

Design criteria of the cooling system to prevent large losses al pool water due to siphoning are discussed -in Section 3.2.2.1 of the Salt.

3.3.5 primary and Secondary Coolant Activity I,imits Applicability:

This specification applies' to. the buildup of radioactive materials in the secondary. coolant system.

Objective:

To ensure there is a level low enough 80 as not to exceed 10CFR20 limits if coolant is released to the sanitary sewer system.

Spectficatlon:

The primary and secondary coolant system shall be monitored for t he buildup of radioactivity and analyzed at least seminnnually for increase in the concentration of radionuclides.

hasts:

The basis for this specification is to ensure releases are legal and consistent with the AI. ARA principal.

15

s 3.4 Confinement Isolation Applicability:

This specification applies to the capability of isolating the reactor building from the unrestricted area outside.

Objective:.To prevent the exposure of the public to airborne radioactivity exceeding the limits of 10CFR20, and the ALARA -

principle.

l Specification:

The reactor shall not be operated unless the following conditions are met:

(1) Ventflatton fan operating (2) Reactor Building bay door closed (3) Reactor Building front and rear personnel doors closed (4) Office windows closed bases:

By having the capability to isolate the Reactor Building, the release of airborne radioactive material may be confined and limited to the extent analyzed in the revised SAR of September 1987.

3.5 Ventilation Systems Applicability:

This npecification applies to all heating, ventilating, and air conditioning systems that exhaust building.

air to the outside environment.

Objective:

To provide for normal ventilation and the reduction

-of airborne radioactivity within the reactor building during normal reactor operatton and to provide a way to turn off all vent. systems quickly In order to isolate the building for emergencies.

Specification:

(1) An exhaust fan with a capacity of at least 1000 cfm shall be operable whenever t he reactor is operating.

(2) This fan, as wel1 as al1 other heating, ventilating, and air conditioning systems shall have the capability to be shut off from a alngle switch in the control room.

Bases:

In the unlikely event of a release of fission products or ather airborne re;' ' aa c t j vl t y, the vent flat ton system will reduce radioactivity inside the reactor building or be able to be isolated.

An analysis of fission product release is found in section 8.4.4 of 1he SAR.

16 I

l 1

[

4-3.6 Radiation Monitoring Systems and Radioactive Effluents 3,6.1 Radiation Monitqring Applicability:

This specification applies to the availability of radiation

r. coni tori ng equipment which shall be operable during reactor operatlon.

Objective:

'To assure that monitoring equipment-is available to evaluate radiation levels in restricted and unrestricted areas and to be consistent with ALARA.

Specification:

(1) When the reactor is operating, the building gaseous effluent monitor shall be operating and have a readout and alarm in the control room.

It may be used in either the

" normal" mode or " sniffer" mode.

(2) When the reactor is operating and the rabbit

-f experimentalfacility is used, the rabbit monitoring system sha)) be operating and have a readout and

-alarm in the control room.

(3) When the reactor is operating.

the following Area Radiation Monitors (ARMS) shall be operating and have both local and control room readouts a'id alarms.

a.

Pool Top b.

Primary Cooling System c.

Heam Port / Rabbit Area d.

Thermal Column Area 1 y (4) Portable survey instrumentation shall be available whenever the reactor is operating to measure beta-gamma exposure rates and neutron dose rates, y,

(5) Portable instruments, surveys, or analyses may be substituted for the jnstruments in the above sections (3.6.1.1, 3.6.1.2, - or 3.6.1.3) for periods up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.'

I

-Read-out and alarme from these temporary instruments shall be reported to the reactor operator on duty at least once per hour Bases:

I (1) The gaseous effluent monitor will detect Ar-41 levels in the reactor building.

During " normal" mode operation it p

will-sample and monitor air Just before it is released l;

from the reactor building.

(SAR 6.3.1)

During " sniffer" mode of operation it may be used for short periods to monitor in und around experimental facilities to determine local Ar-41 levels.

17

c f

L (2) The rabbit stack a mitor is used with the rabbit since the rabhit system uses a!* as its transport mechanism and Ar-41 production takes place.

This monitor will provide warning if Ar-41 levels being released in the bu!] ding are too high (SAR G.3.2 and G.3.4.3) o (3) The ARMS provide a cont inuing evaluation of the radiation levels within the Reactor Building (SAR 3,7).and provide a warning.lf levels are higher than anticipated.

(4) The availability of survey meters enables the Reactor Staff to independently confirm radiation levels throughout the building.

(5) in the ' event of instrument failure short term

=

substitutions will enable the safe continued operation of the Reactor, 3.6.2 Radioactive Effluents Applicability:

This specification applies to the monitoring of radioactive effluents from the facility.

Objectives:

(1) To - ensure that Ilyuld radloactIve releases are safe and legal.

_(2)

To ensure - t hat the release of Ar-41 beyond the site boundary does not result in - concentration above MPC for unrestricted area.

(3) To 'a s h. ire that the release of Ar-41 in the restricted area does not result in concentrations above MPC.

Specifications:

(1) The release rate for radioacttve ligulds beyond the site boundary shall not. - exceed the limits. as specified in 10CFR20 at the point of release.

(2) The concentration of Ar-41 at the 90 int of release into the unrestricted area shall not exceed the unrestricted area MPC when averaged over one year or 10 x MPC when averaged over one day.

(3) The concentration of Ar 41 in the restricted area shall' not exceed MPC when averaged over 7 consecutive days.

18

_o Bases:

(1) The basja for.this specification is found in Section 6.2 of the Safety Analysis Report.

(2) The basis for this specification is found in Section 6.3 of the. Safety Analysis Report.

(3) The - bl.si s for this specification is found in Section 0.3 ~ of the Safety. Analysis Report and 10CFR20.103(b)(2).

3.7 Experiments 3.7.1 Reactivity Limits Applicability:.This specification applies to experiments to be inpialled 'in or near.the reactor and associated experimental focilities.

Objectives:

To prevent damage to the reactor or excessive release of radioactive materials-in the event of an experiment futlure.

Specificatlon:

(1) The absolute value of the-reectivity worth of any single secured experiment shall not exceed 0.7% delta k/k.

(2) The absolute value of the reactivity worth of any single movable experiment shall not exceed 0.4% delta k/k.

(3) The absolute value of the reactivity worth of all movable s

experiments shall not exceed 0.6% delta k/k.

(4) The absolute value of the reactivity worth of experiments having moving parts shall be designed to have an insertion rate Jess than 0.05% delta k/k per second.

j (5) The nbsolute value of the reactivity worth of.any movable experiment that may be oscillated shall have a reactivity change of less than 0.05% delta k/k.

(6) The total react ivity worth of all experiments shall not be greater than 0.7% delta k/k.

19

cc t

6 Bases:

(1) The bases for specifications 1, 2,

3, and 6 are found in Section 8.4.3.2 of the SAR which - evaluates a step insertion of reactivity from an experiment.

(2) The bases for specifications 4 and 5 allows for certain reactor kinetics experiments to be performed but still limits the rat e of change of reactivity insertions to levels that have been-analyzed.

Section 8.4.3.2 of the,

SAR evaluat es a step insertion of reactivity - from, an experiment, m

3.7.2 Deslen and Materials Specification:

(1) No experiment shall be installed that could shadow the nuclear instrumentation, interfere with'the insertion of a control rod, or credibly result in fuel element damage.

(2) All materials to be irradiated in the reactor shall be either corrosion resistant or doubly encapsulated within corrosion resistant containers.

(3) Explosive materials shall not be allowed in experiments,-

except for neutron radiographic exposures of items performed outside uf the core and experimental facilities.

-The amount of explosive material contained in capsules used for radiographic exposures shall'not exceed 5 grains-of gunpowder.

Bases:

(1) Specification 1 assures no phyr.ical interference with Om operation of the reactor detectors, control rods, os physical damage to fuel element will take place.

(2) Ljmiting corrosive materials in Specification 2,

and explosives in Specification 3 reduces the likelihood of damage to reactor components and/or releases of radioactivity resulting from experiment failure.

(3) 1,i m i t i ng explosive materials to neutron radiographic exposures done outside of the core and experimental facilities reduces the likelihood of damage resulting for this experimental failure.

20 Q

' j.,.

}

t 4.

SURVEll, HANCE REQUIREMENTS 4.1 Reactor Core Parameters 4.1.1 Excess React ivit y and Shut down Marcin Applicability:

This specificalfon applies to survel))ance requirements for determining the excess reactivity of the reactor core and its shutdown margin.

?

i Objective:

To assure,that the excess reactivity and shutdown margin limits of the reactor are not exceeded.

Specifications i

(1) Whenever a net change in core configuration, for which the predicted change in reactivity is >. 2 %,

delta k/k, involving grid position is made, both excess reactivity and shutdown margin shall be determined.

(2) Both shutdown margin and excess reactivity sha)) be determined annually.

Dasos:

A determination of excess reactivity is needed to l

preclude operatjng without adequate shutdown mntgin.

Moving a component out. of the core and returning it to its same location is not a change in the core. configuration and does-not require a determination of excess react ivity.

4.1.2 Fuel Elements Applicability:

This specificat ion app'les to surveillance l

requirements for determining the physical condition or the rear".or fu-l.

Objective:

To e n e< u r e that visible deterioration, j'

are detect ed. In a t imely manner.

corrosion, or other physical changes.to the fuel elements Specification:

All fuel elements, both in-core and out, shall be visually inspected at least once every five years, by inspecting at least one f.ifth of the elements annually.

Hasis; if the water purity is continuously maintained within specified limits, it is projected that chemical corrosion of the fuel clad

'll proceed slowly.

However, raults in the basic mater!

or fabrication could lead to loss of cladding integrity.

21

a e

4.2 Reactor Control and Safety Systems 4.2.1 Control Rods Applicability:

This specification applies to the surveillance requirements, for the shim safety rods and the regulating rod.

Objecttve:

To assure that all rods are operable.

Specifications:

(1) The reactivity worth of the shim safety rods and regulating rod shall be determined annually and prior to the routine operation of any new core configuration.

(2) Shim kaf ety rod drop and drive times and regulating rod drive time shall be determined annually or after maintenance or modification in completed on a mechanism.

(3) The shim safety rods and regulating rod shall be visually inspected annually, for indication of corrosion, and indication of excessive friction with guides.

Bases:

The reactivity worth of the rods is measured to assure the required shutdown margin and reactivity insertion rates are maintained.

it also provides a means for determining the reactivity of experiments.

Measuring annually will provide corrections for burnup and after core cha..Aes assures that altered rod worths wl)) be known prior to continued operat ions.

The visuni inspectton of the rods and measurements of drive and drop times are mnde to assure the rods are capable of performing properly.

Verificatirn of operability after maintenance or modificutinn of the control system wi11 ensure proper reinstallation.

4.2.2 Reartor Safety SYR1em Applienbility:

This specification applies to the surveillance requirement s for the Henctor Safety System.

Objective:

To assure the reactor safety system channels will remain opernble and prevent safety limits from being exceeded.

Specification:

(1) A channel check of each mensuring channel shall be performed daily when the reactor is operating.

(2) A channel test of each measuring channel shall be performed prior to each day's operation, or prior to each

• operatinn extending more than one day, 22

(3) A channel salibration of the reactor power level measurjng channeln saall be made annually.

(I.inear 1.evel and 1.00-N.)

(4) A channel calibration of the bevel and period Safety Channels shnll be made annually.

Channel tests are done on these before each day's operatJon.

( fi) A channel ca l i brat lon of the following sha}] be made annually a.

Core inlet temperature measuring system b.

Pool water level measuring system c.

Coolant system pumps measuring system d.

primary coolant flow measuring system (6) The control room manual scram shall be verified to be operable prior to each day's operation.

All other manual scram switches shall be tested annually.

(7) Other scram channels shall be tested / calibrated annually.

(8) Any instrument channel replacement shall be calibrated after installation and before utilization.

(9) Any instrument repair or replacement shall have a channel test prior to reactor operation.

';m dally channel tests and checks will assure that the liases:

i scram channels are operable.

Appropriate annual tests or calibrations will assure that long term functions not tested beforo daily operation are operable.

4.3 Coolant System 4.3.1 primary Coolant Water purity Applicebility:

This specification applies to the conductivity of t he prir.iary coolant wat er.

Objective:

To assure high quality pool water, s

Specification:

The conductivity and pH of the pool water shell be measured weekly.

lin n e s :

This assures that changes that ofght increase the corrosion rate are detected in a timely manner and that the concentrations of impuritles that might be made radioactive do not increase significantly.

23 i.

Y

g-3 i

4. 3. 2 ' Coolant System Rudioactivity Applicability:

This specification applies to the radioactive materjnl Jn the primary coolant or secondary coolant.

Objective:

To identify radionuclides as potential sources of j

release to the sanitary sewer system.

= Specification:

Primary and secondary coolant shall be analyzed for radioacttvity quarterly or before release.

Itases :

Radionuclide analysis of the pool water or secondary coolant allown for determir.ation of any significant buildup of r

fission. or activatJon products and helps assure that radioactivity la not permitted to escape to the tertiary system in an uncontrolled manner.

4.4 Confinement j

Applicability: 'This specificatlun applies to the surveillance l

requirements for building confinement.

ObjertIve:

To assure that the building closure capability

exists, f

Specification:

A monthly test shall be made to assure that the building exhaust fan, bay door, front.und rear personnel doors, arnt office doors and windows are operable.

c flases :. Monthly surveillance of this equipment will verify that the confinement of the reactor bay, can be maintained if needed, i

4.5 Ventilation System Applicability:

This specification applies to the surveillance requirements for the building ventilation system.

Objective To assure that the ventilation ystem functions s

satisfactorily.

Spectficatlon; (1) Vent i tut ion fans and closures shall be checked for proper operation on a quarterly basis.

L

(?) The shutoff switch for all fans and air conditioning

(

systems shall be tested on n quarterly basis.

t lia s e s :

This surveillance will assure that durjng normal operations the althorne radioactivity will be minimized inside the building and that the building con be Isolated quickly if necessary to prevent uncontrolled esenpe of air-borne radioactivity to the unrestricted environment.

24 sp a

O:

t 4.6 Radiation Monitoring Systems and Radioactive Effluents 4.6.1 1:ffluent Monitor Applicability:

This specification applies to the surveillance requirement of the effluent monitor.

Objective:

To assure the effluent monitor is operational and providing accurate effluent readings.

Specification:

The effluent monitor shall have a channel calibration annually and a channel test before each days operatlon.

Bases:

The calibration will assure effluent release estimates are accurate and the test will askure the monitor is operable whenever the reactor is operating.

4.6.2 Rabbit Vent Monitor Applicability:

This specificat ion applies to the surveillance requirements of the rubbit vent monitor.

Objective:

To assure the monitor is operutional and providing meaningful information about effluent releases from the rabbit into the reactor building.

Specification:

The monitor shall have a channel calibration e:nnua l l y and a channel test before each day's reactor operatlon.

11ases :

The calibration will assure effluent releases inside the building are accurately estimated and the test will assure the monitor in operable before the rabbit is used, 4.G,3 Aren Radiation Monitors (ARMS)

Applicabjljty:

This specification applies to the area radiation monitoring equipment.

Objective:

To assure that radiation monitoring equipment is opetable whenever the reactor is operat ing.

Specificatlon:

A channel test of the ARMS shall be completed before each day's operation and a channel calibration shall be completed annually.

Huses:

Calibratton annually will insure the required reliability and a check on days when the reactor is operated will detect obvious malfunctions in the system.

25 i

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4

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4.6.4 portable Survey inst rumentat ion 1,

Applienhility:

.This specifjcation applies to the portable survey inst rumentation available to measure beta-gamma exposure rates and neutron dose rates, i

r,,

Objective: 'To assure that radiation survey instrumentation-Ja i

operable whenever the reactor is operating-Specification:

liet a-gamma and neutron survey meters shall _ be i

tested for operability each day: the reactor is to be operated j

and shall be calibrated annua))y.

Bases:

Tests on days when the reactor is operated will detect obvious detector deficiencies and an annual calibration will I

assure re11abl21ty.

,a

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5.

DI. SIGN Ff.ATURT.S 5.1 Site and Faci]Ity Description 5.1.1 Factilty I.oention The reactor and associated equipment is housed in a building at 1298 Kinnear Road, Columbus, Ohio. Theminimumfreeajrvoluge of the building housing the reactor will be,1 70,000 ft There is an exhaust fan with dampers prov.iding for control of release of uitborne radioacttvity.

It is in the area of The Ohio State University Research Center.

5.1.2 1:xclusion and Hest rict ed Area The fence surrounding the Research Center shall describe the exclusion area.

The restricted area as defined in 10CPH2O shall consist of the Reactor Building.

5.2 Reactor Coolant System 5.3.1 Primary coolant I,oop Natural convertIve cooling is the primary means of heat removal from the core.

Water enters the core at the bottom and flows upward t hrough t he flow channels in t he f uel element s.

5.2.2 secondary and Tertinry Coolant hoops The secondary coolant loop removes heat from the primary coulant.

The secondary coolant (ethylene glycol and water) passes through two separate heat exchangers to remove heat Jf necessary.

If the outside air temperature is < 78"F then an outside fan-forced drycooler 18 sufficient to remove all heat generat ed at 500 kw.

City water flow through the secondary side of an additional heat exchanger makes up the tertiary

loop, 11 provides additional coo]Jng for the secondary coolant.

5.3 Reactor Core and Fuel 11p to 30 posit tons on the core grid plate are available for use as fuel element positjons.

Contro] rod fuel elements occupy four of these positions and one is reserved for the Central Irradiation Facility flux trap.

Several arrangements for the

eold, clean, critleal core have been investigated, Approximately sixteen standard fuel elements in addition to the cont r ol rod fuel elements will be required.

Partial elements, core plugs, and graphit e elements may be utilized in various combinat ions t o achieve t he proper K excess.

27

e The reactor fuel in The DOE Standard uranium-silicide (U Si )

3 with a U-235 enrichment of less than 20%.

It is flat bla e fuel with a " ment

• thickness of 0.020" and aluminum cladding of 0.015*.

Standard fuel elements have a total of 16 fueled plates ured 2 outer pure aluminum plates.

The control rod fuel elements have eight of the inner fuel plates removed to allow the cont rol rods to enter, pure aluminum guide plates are on the inside of this gap.

The outer two plates for each control rod assembly are fueled.

partini elements are also available with 25, 40, 50, and N percent of the nominal loading of a standard element These partial fuel elements are prefabricated by the vendor with fixed numbers of plates.

(1) Refcrences:

?RC NURl;G 1313 (NL/RERTR/TM-10 ANI./RERTR/TM-11 5.4 Fuel Storage The fuel storage pit, located below the floor of the reactor pool and at the end opposite from the core, shall be flooded with water whenever fuel is present and shall be capable of storing a complete core loading.

When fully loaded with fuel

}

and filled with water, K shall not exceed 0.90, and natural convectIvecoolingshall"efsurethatnofueltemperaturesreach a point at which UND is possible:.

The two fuel storage racks located in the Hulk Shielding a

Facility storage pool shall each:

(a) Contain no more than 16 fuel elements spaced on a pitch of

[

at least 6 inches in a two by eight matrix.

(b) De placed no closer than 24 inches in any direction from each other or any other fuel storage facility.

(c)

Have u K less than C.90 when fully loaded with fuel and floodedONhwater.

5.5 Puel Handling Tools All tools desJgned for or capable of removing fuel from core posj t innu or storage rock positions shall be secured when not in use by a system controlled by the supervisor of reactor operations, or the senior reactor operator on duty.

28 I

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6.

ADMINISTRATIVE CONTHOLS i

G.i Organisation 6.1.1 Structure i

t The ' Ohio State University Research Reactor 18 a part of, the College of Engineering administered by the Ensti nee ri n g l

Experiment Station.

The organizationni structure is shown in Figure G.I.

r b-0.1.2 Responsibility h

L The Ilirector of the. Engineering Experiment Station (Level 1) is

')

L the contact person for communications between the U.S. Nuclear

)

Regulatory Commist.fon and The Ohio State University.

The Director of the Nuclear Reactor Laboratory (Level 2) will have overall responsibiljty for the management of the facility.

I The Associnte Director (or Manager of Reactor Operations)

(Level 3) shall be responsible for the day-to-day operation and for ensuring that all operations are conducted in a safe manner a

and wit hin -the limits prescribed by the facility license and Technical Specjfications.

During periods when the Associate l

Director 18 absent. his responsibilities are delegated to a Senior Reactor Operator (Level 4).

l 6.1.3 S t a f f i nt!

,o I

4-During Reactor Operations:

Nd (1) Two or more personnel, at,least one of whom is a licensed K.

reactor operator, shall be in the building during all l

L ';,,

react or operat ions.

The second shall be-capable of a

ll

'2' the reactor.

4 following simple written instructions for shutting down i

w (2) At least two licensed ~operntors should be in the building g

during any extended operations (longer than 60. minutes).

.-( 3 ) Two persons, one of whom shal1' be a 1iconsed - senior renetor-operator shalithe in the building for the first m,,

start-up of the day, y

.(4) Two persons, one_of whom shall be a licensed senior

'i reactor operator, shall be in the building during start-up-l

(,f ' -

after an. unplanned shutdown.

y (S)' !)nring all operations, n licensed operator shall be in the g

control room either as console operator or directing _ the notivilles of a student operator or t rainee.

29 h

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, i s (if -!W i n l --

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, President Vice President Provont.

for

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Health Services 1

'1 t

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Dean,

- College of Engineering w

j (

Reactor Operations Committee G,c..

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' ' '~

- Director. Engineering I

f

,)

Experiment Station j

(Level 1)

I I

I Director, Nuclear Reactor

Director, Laboratory Office Radiation Safety (Level 2) 4

' Associate Director -

Nuclear Reactor Laboratory.

(Level 3) mi Senior Reactor Operator (Level 4) o

\\

h i f(

Renetor' Operations Staff i

'q4 1

.w :

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t

.gWM Solid Lines-Paths of Direct Responsibility

' h:li;;;

,Das.hed Lines----------- Paths of Information.

77

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Figure 6.1

' Administrative Organization' 3

v.

30

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9 y!LM!? Si "

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A (6) A minimum of three people shall be present during fuel handling.

One shall be a licensed senior reactor operator, and one shall be at least a licensed reactor operator.

6.1.4 Select Jon and Traininit of Personnel The selection, training, and requalification of operations personnel shall meet or exceed the requirements of American National Standard for Selection and Training of Personnel for Hesearch Heact ors, ANSI /ANS-15.4-1977, Sections 4-6, 6.2 Review and Audit There shall be a Heactor Operations Committee (H00) which shall review and audit reactor operations to assure the facility is operat ing in a manner consistent with public safety and within the terms of the facility license.

The Committee advises the Director of t he NHl., and 18 responsible to the Provost of The Ohio State-(Inive rs i t y.

6.2.1 Comnosition and Qualificattons of the HOC Committee members shall be appointed annually by the Provost of The Ohio State 11niversity.

The Committee shall be composed of at least nine members including ex-officio members.

The Director and Associate Director of the Nuclear Reactor Laboratory, and the Director of the Office of Radiation Safety shall be ex-officio vot ing members of the Commit t ee.

The remaining Committee members shall be faculty, staff, and student representatives of The Ohio State (Iniversi t y, having professional backgrounds in engineering, physical, biological, or medical sciences, as well as knowledge of and interest in applications of nuclear technology and tonizing radiation.

0.2.2 HOC Meet inits The Committee shall Leet at least once each quarter, it should meet during the first two weeks of each calendar quarter, A quorum shall consist of at least 50 percent of the members who are not directly involved in or responsible for facility operations.

Ex-offielo members shulI be counted in the quorum as foilows:

(1) The Provost is an ex-officio member.

Since the Provost is not appointed as a member of the ROC, the Provost is not required to act as a member, is not counted as a member when counting a quorum, but does have the right to vote (2) hx-offlejo members who are under the authority of the Provost serve in the same capacity as those who are appointed by the

Provost, i.e they have the right to vote but they are not counted as members when counting a quorum if they are directly involved in or responsible for facility operattons.

31

1 (a) Ex-officio members, if any, who are not under the authority of the Provost, have the right to vote, but have no obligation to participate.

Accordingly, they are not counted as members when counting a quorum.

(4) All ex-officio members hold membership by virtue of their office.

They cease to be members when they cease to hold office.

6.2.3 Sub-Committeen The chairperson may appoint a Subcommittee from within the committee membership to act on behalf of the full committee on those ma tt ers which cannot await the regular quarterly meeting.

The fu)) Committee shall review the actions taken by the Subcommittee at the next regular meeting, 6.2.4 ROC Review and Approval Function The responsibilities of the ROC include, but are not limited to the following:

(1) Review and approval of experiments util.lzing the reactor facilitles (2) Review and approval of procedures

'(3) Review and approval of all proposed changes to the license and technical specifications (4) Determination of whether a proposed change, new test, or experiment would constitute an unreviewed safety question or require a change in the technical specifications per 10CFR50.59 (5) Review of audit reports (6) Review of abnormal performance of plant equipment and operating abnormalities having safety significance (7) Review of unusual occurrences and incidents which are reportable under 10CFR19, 20, 21, and 50, or Section 6.6,4 of this document, and (8) Review of violat ions of technical specifications, license, or procedures having safety significance.

Relat Ivo to it em (1), responsibility for review of experiments on a day-to-day basis shall lie with the Director of the Nuclear Reactor 1.aboratory or his designee.

This day-to-day review shall determine whether a specific experiment has previously been approved in the generic sense by the ROC.

A quarterly report of performed experiments shall be provided.for ROC review.

32

o o l

Relative to item (2), the NRh Director or his designee shall be responsible for approval of procedures or changes to procedures on a day-t o-day basis.

He shall provide a summary of all procedure i

changes to the ROC for their review and approval.

I A complete set of minutes of all Committee and Subcommittee 1

meetings, including copies of all document 6ry material reviewed, 1

and all approvals, disapprovals, and recommendationw shall be kept.

Minutes or reports of all Committee meetings or Subcommittee meet inn should be disseminated to the Committee members prjor to the next regularly scheduled meeting, and should be read for approval as the first item on each Agenda.

A copy of the minutes, or any reports reviewed, should also be forwarded to the Director of the Engineering Experiment Station in a timely manner.

G.2.5 ROC Audit Functton A t hree member Subcommittee shall meet annually to perform an audit of NRh operations and records or review the results of an-independent audit completed by another qualified agency.

At least two individuals on the Audit Subcommittee shall be ROC members.

The third may be a staff member from the Reactor Laboratory or another individual appointed by the ROC chairperson.

No member shall audit a function that he is responsible for performing.

Each person should serve for three consecutive audits, at which time he or she should be replaced by a new member.

In this way, each Subcommittee should consist of two holdovers and one new member.

The member serving for his or her second audit should be the Audit Subcommittee Chairperson.

The following items shall be audited:

(1) Reactor operations for adherence to facility procedures, Technical Specifications, and license requirements (2) The requalification program for the operating staff, (3) The facility Emergency plan and implementing procedures, i

l (4) The facility Security plan and implementing procedures, and (5) The results of actions taken to correct any deficiencies that affect reactor safety, and ll (6)

Conforrance with the ALAHA policy and the effectiveness of radio]rgic control.

Deficiencies found by the AudJt Subcommittee that affect React or Safety. shall be reported immediately to the Director of the Engineering Experiment Station.

A written report of audit findings should he submitted to the Director of the Engineering Experiment Station and the full Reactor Operations Committee within three months of the audit's completion.

33

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6.3 Procedures G.3.1 Renet or Operat inn procedures Writ ten procedures, reviewed and approved by the Director and the HOC, shnll be in effect and followed.

The procedures shall be adequate to assure the safety of the reactor, but should not preclude the use of Independent judgement and action should the situation require such. All new procedures and changes to existing procedures shall be documented by the NRh staff and subsequently reviewed by the ROC.

At least the following items shall be r

covered:

(1) Startup, operation, and shutdown of the reactor, (2)

Installation, removal, or movement 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 cooling system leaks, and abnormal reactivity changes, y

(4)

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

(: ) Preventive and corrective maintenance procedures for systems which could have un effect on reactor safety, (6) periodic surveillance of reactor instrumentation and safety systems, area monitors, and radiation safety equipment, (7)

Implementation of Security, Emergency and Operator training and requalification plans, and (8)

Personnel radiation protection.

34

o i

0.3.2 Adminiktrative procedures Procedures shall also be written and maintained to assure compliance with Federal regulations, the facility license, and commitments made to the ROC or other advisory or governing bodies.

As a minimum, these procedures shall include:

(1) Audits.

(2) Special Nuclear Material accounting, (3) Operator requalification, (4) Record keeping, and (5) procedure writing and approvn].

6.4 Experiment Review and Approval G.4.1 Definitions of Experiments Approved experiments are those which have previously been reviewed and approved by the ROC.

They shall be documented and may be included na part of the procedures Manual.

New experiments are those which hnve not previously been reviewed,- approved, and performed.

Routine tests and maintenance activities are not experiment 8.

6.4.2 Approved Experiments All proposed experiments ut ilizing the reactor shall be evaluated by the experimenter and a licensed Senior Reactor Operator to assure compliance with the provisions of the utilizat ion license,.

the Technical Specifications, and 10CFR Parts 20 and 50.

If, in the judgement of the Senior Reactor Operator, the experiment meets with the above provisions, is an approved experiment, and does not constitute a threat to the integrity of the reactor, it may be approved for performance.

When pertinent, the evaluation shall include considerations of:

!I (1) The reactivity worth of the experiment (2) The integrity of the experiment, including the effeet8 of changes in temperature, pressure, or chemical composition (3) Any physical or chemical interaction that could occur wit h t he reactor components, and (4) Any radiation hazard t hat may result from t he ac't ivat ion of materials or from external beams.

l 35 l

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6.4.3 New Fxperiments Prior to performing an experiment not previously approved for the reactor, the experiment shall be reviewed and approved by the Reactor Operat ions Commit t ee.

Committee review shall consider the following information:

(1) The purpose of the experimen',

(2) The procedure for the performance of the experiment, and L

(3) The safety evaluntton previously reviewed by a licensed Senior Reactor Operator.

6.5 Required Actions 6.5.1 Act ion To De Taken in t he Event A Safet y 1,imit in Exceeded (1) The reactor shall be shut down, and reactor operatjons shall ~

not be resumed until authorized by the NRC.

(2) The safety limit violation shall be promptly reported to the Director of the Reactor baboratory.

(3) The safety limit violation shall be reported by telephone to the NRC within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />..

(4) A safety. limit violatJon report shall be prepared.

The report shall describe the following:

Applicable circumstances lending to the violation including, when known, the cause and contributing

factors, b.

Effect of the violation upon reactor facility components, systems, or structures and on the health and safety of personnel and the public, and c.

Corrective action to be taken ta prevent recurrence.

(5) The report shall be reviewed by the Reactor Operations Committee and shall be submitted to the N'tC within 14 worktre days when aut horizat lon is sought to resume operation of the reactor.

6.5.2 Act ion To He Taken in The I' vent Of A Reportable Occurrence A reportable occurrence is any of the following conditions:

(1)

Operating with any safety system setting less conservative than stated in these speelfications, 36

}o(dd g i

(2) Operating in violation of a Limiting Condition for Operation established in Section 3 of these specifjcations.

(3) Snfety system component malfunctions or other component or system malfunctions during reactor operation that could. or threaten to, render the safety system incapable of performing its intended function.

(4) An uncontrolled or unanticipated increase in reactivity in excess of.4% delta k/k, (5) An observed inadequacy in the implementation of either administrative or procedural controls, such that the inadequacy could have caused the existence or. development of an unsafe condition in connection with the operation of the reactor, and (G) Abnormal and. significant degradation in reactor fuel and/or cladding, coolant boundary, or confinement boundary.(excluding minor leaks) where applicable that could result in exceeding prescribed radiation exposure limits of personnel and/or the environment.

(7)

Any uncontrolled or unnuthorized release of radioactivity to the unrestricted environment.

. In t he event of a reportable occurrence, the following action shall be taken:

(1) The reactor conditions shull be returned to normal,.or the reactor shnll be shutdown, to correct the occurrence.

(2). The Dl cetor of the Reactor 1.aboratory shall be notified as soon as possible and-corrective action shall be taken before reauming the opernt.lon involved.

(3)- A we.itten report of the occurrence shall be made which shall include an analysis of the cause of the occurrence, the-correctivo actton taken, and the recommenduttons for measures to.' preclude or reduce the probability of recurrence.

.This report shall. be submitted to the Director and ' the Reactor Operations Committee for review and approval.

(4)

A. report. shall be submi tt ed to the Nuclear kegu l a t ory Commission in accordance with Section G.6.2 of. these spec 1fIenIlons, 6.6 Reports g

Reports shull be made to the Nuclear Regulatory Commission as' follows:

37 i

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• e 6.6.1 Operatinc Repor3 An annual report shall be made by September 30 of each year to the Director, Of fice of Nuclear Reactor Regulation, NR':, Washington, DC 20555, with a copy to the NRC, Region 111, in accordance with 10CFR 50.4, providing the following information:

(1) A narrative summary of operating experience (including experiments performed) and of changes in facility design, performance charact eristics, and operating procedures related to reactor safety occurring during the reporting period.

(2) A tabulation showing the energy generated by the reactor (in i

kilowatt hours) and the number of hours the reactor was in use.

l (3) The results of. safety-related maintenance and inspections.

The reasons for corrective maintenance of safety related jtems, shall be included.

(4 ) A table of unscheduled shutdowns and inadve-tent scrams, including their reasons and the corrective actione taken.

I (5) A summary of the Safety Analyses performed in connet ton with changes to the facility or procedures, which af fect reactor safety, and performance of tests or experiments carraad out under the conditions of Section 50.59 of 10CRF50.

(0) A summary of the nature and amount of radioactive ganeons, liquid, and solid effluents released or discharged to the environs beyond the e f fect ive control of the licensee as measured or calculated ' at or prior to the point of such release or discharge.

(7)- A summary of radiation exposures received by facility personnel and visitors, including the dates and times of significant exposures.

0.0.2 Special Reports j

(1) A telephone or telegraph report of the following shall be submitted as soon as posolble, but no 10. ' u than the next I

working day, to the NRC Region ll! ut f. ice:

(a) Any accidental offsite release of radioact ivity ' above authorized limits, whether or not the release resulted in property damage, personal injury, or known exposure.

i (b) Any exceeding of the safety.Ilmit as defined in 80ction 2.1 of t hese speci ficat ions.

(c) Any reportable occurrences as defined in Sect ion 0.5.2 of tbese specifications.

38 l

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1'c ) A written report shall be submitted within 14 days to the Director, Office of ' Nuclear Reactor Regulation, US NRC, Washington, DC 20555 with a copy to the NRC Region 111, in accordance with 100FR 50.4, of the following:

(a). Any accidental offsite release of radioactivity above permissible. ljmi ts, whether or not the release resulted in property damage, personal injury, or known exposure.

(b) Any exceeding of the safety 11mit as defined in Section 2.1.

(c) Any reportable occurrence as defined in Section 6.5.2 of these specifications.

(3) A written report shall be submitted within 30 days to the Director, Office of Nuclear Reactor Regulation, US NRC, Washington, DC 20555, with a copy to the NRC, Region 111,_

i Office in accordance with 10CFR 50.4, of the following:

(a) Any substantial variance from performance specifications contained in these specifications or in the SAR, (b) Any significant change in the transient or accident analyser as described in the SAR, and (c) Changes in personnel serving as D i rect or. Engineering Experiment Station, Reactor Director, or Reactor Associate Director.

(4) A report shall be submitted within nine months af ter initial criticality of the reactor or within 90 days of completion of the startup test program, whichever is earlier, to the Dilector. Office of Nuclear Reactor Regulation, U.S.

NRC, Washington, DC 20555, with n copy to tne NRC,- Region 111 upon receipt of a new facility license, an amendment to license authorizing an increase in power level or the installation of a new core of a different fuel element type or design than previously used, i

The report shall include the measured values of the operating conditions or characteristics of the reactor under the new conditions, and comparisons with predicted values, including the following:

i (a) Tot al cont rol rod react tvity wort h, (b) Reactivity wort h of the single cont rol rod of h i gi.a s t reactivity worth, and (c) Minimum shutdown margin both at umhjent and operating temperatures.

39

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t (d) Excess reactivity

!l.

(e) Calibration of operating power levels

-(f) Radiatton leakage outside the biological shielding (g)

Release of radioactjve effluents to the unrestricted environment.

6.7 Records Records or logs of the itsas listed below shall be kept jn a manner convenient for - review, 4nd shall be retained for as long as indicated, 6.7.1 Records to be Retained (or a period of at 1.en s t Five Yents

g.

_f (1) normal plant operation, (2) principal _ maintenance activities, (3) experiments performed with the reactor.

(4)- reportable occurrences, (5) equipment and component surveillance activity, (6) facility radiation and contamination surveys, (7) transfer of radionetIve material, (8) changes to operatinE procedures, and

-(9) minutes of Renctor Operations Committee meetings.

6,7.2 Records to be Retnined for at 1. cast One Requalifiention Cycle Regarding retraining and roqualification of licensed operations personnel, the. records of the most recent complete requalification

, cycle shall be maintained at u)! tines the individual is' employed.

0.7,3 Records to be Retnined for the Life of the Facility (1) :enseous and liquid-radianctive effluents releanud to the environment, (2)_ fuel inventorJen and transfers, (3) radintlon exposures for all personnel, (4) changes to reactor systems, component s, or: equipment that any l

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affect' reactor safety, (S) updat ed, corrected, and_ as-hullt drawings of t he facility.

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