ML20062F067

Amend 13 to License R-75,increasing Authorized Steady State Reactor Power from 10 Kw to 500 Kw & Allowing Possession of SNM

ML20062F067
Person / Time
Site:	Ohio State University
Issue date:	11/14/1990
From:	Murley T Office of Nuclear Reactor Regulation
To:
Shared Package
ML20062F064	List: ML20062F061 ML20062F067 ML20062F077 ML20062F086 ML20062F099
References
R-075-A-013, R-75-A-13, NUDOCS 9011270080
Download: ML20062F067 (46)

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OHIO STATE UNIVERSITY DOCKET NO. 50-150-AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 13 License No. R,

1.

The Nuclear Regulatory Commission (the Commission) has found:

A.

The appiication for amendment to Facility Operating License No..

R-75, filed by the Ohio State University (the licensee), dated

+

October 7,1987, as supplemented on May 26,1989, February 28 and June 12, 1990, complies with'the standards and requirements of'the Atomic Energy Act of 1954, as amendea (the Act), and the Commission's regulations as set-forth in 10 CFR Chapter,I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; 1

C.

~There is reasonable assurance (i) that the activities authorized by' this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will_be.

conducted in compliance with the Commission's regulations set forth in 10 CFR_ Chapter I; D.

The issuance of this amendment will not be inimical to:the common defense and security or to the health 1and safety of the public; and' E.

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

r-2.

Accordingly, paragraph 2.B.' of License No. R-75 _is hereby amended to read as follows:

B.

Pursuant to the Act and 10 CFR Part 70, "Special Nuclear _Materiali" to receive, possess and use in connection with operation--of.the reactor 80 grams of plutonium contained in encapsulated plutonium--

beryllium sources, up to 10 grams of contained Uranium-235 enriched to 93 percent in the form of fission chamber linings,1 foil targets and other research applications, and up _to 5.2 kilograms _of contained.

Uranium-235 at enrichments equal.to or'less' than 20 percent, and to.

possess but not separate such special nuclear material as may~ be produced by operation of the' reactor.

9011270080 901114 PDR ADOCK 05000150 P

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Accordingly, paragraph 3.A. of License No. R-75 is.hereby amended to read as follows:

A.

Maximum Power Level The licensee is authorized to operate the reactor at steady state power levels up to a maximum of 500 kilowatts thermal.

4.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the enclosure to this license amendment, and paragraph 3.B. of License No. R-75 is hereby amended to read as follows:

B.

The Technical. Specifications contained in Appendix A,.as reviseu through Amendment 13, are hereby-incorporated in the license.

The. licensee shall operate the reactor in accordance with these technical specifications.

5.

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

FOR THE NUCLEAR REGULATORY COMMISSION

'; M --

Thomas E. Murley, Director Office of Nuclear Reactor Regulation

Enclosure:

Appendix A Technical Specifications Date of Issuance:

November 14, 1990 i

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ENCLOSURE TO LICENSE AMENDMENT NO.-13 TO FACILITY OPERATING LICENSE NO. R-75 DOCKET NO. 50-150 The Technical Specifications (TS) have been reformatted in their entirety.

Changes made to the TS ara identified by vertical lines indicating the areas of change.

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e TABLC OF CONTENTS 1.

INTRODUCTION.

1 1.1 Scope 1

1.2 Application I

1.2.1 Purpose I

1.1.2 Format 1

1.3 Definitions 2

2.

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

7 2.2 Limiting Safety System Settings 7

3.

LIMITING CONDITIONS FOR'0PERATION 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 Heactivity insertion Rate Il 3.2.3 Minimum Number of Scram Channels 11 3.3 Coolant System.

14 3.3.1 Pump Requirements

. E. -

14 3.3.2 Coolant i. eve.

14 3.3.3 Water Chemistry Requirements 14 l

3.3.4 1.eak or Loss of Ct.olant Detection.

15 3.3.5 Primary and Secordary Coolant Activity I.imits.

'15 3.4 Confinement Isolation 16 l

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

17 3.6.1 Radiation Monitoring 17 1

3.6,2 Radioactive Effluents.

18 3.7 Experiments 19 3.7.1 Reactivity Limits.

19 3.7.2 Design and Materials 20 l

4.

SURVEILLANCE REQUIREMENTS 21 4.1 Reactor Core Parameters 21 4.1.1 Excess Reactivity 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 34.6.3 Area Radiation Monitors (ARMS) 25:

4.6.4 Portable Survey Instrumentation 26

'l i

i

4 A

B 5.

DESIGN FEATURES 27 5.1 Site and Facility Description 27 6.1.1 Facility Location.

27' 5.1.2 Exclusion and Restricted Area.

27 5.2 Reactor Coolant Systea.

27?

5.2.1 Primary Co lant Loop 27 5.2.2 Secondary and Tertiary Coolant Loops 27 5.3 Reactoc Core and Fuel

'27 5.4 Fuel Sterage.

28 P.5 Fuel Haniling Tools 2

28 6.

ADMINISTRATIV3 CONTROLS 29 6.1 Organization.

29 6.1.1 Structure.

29 6.1.2 Responsibility

-29 6.1.3 Staffing 29-6.1.4 Selection and Training of Personnel 31.

6.2 Review and Audit.

31 6.2.1. Composition and Qualifications of the ROC.

31' 6.2.2 ROC Meetings

..,....................- 131 6.2.3 Sub-committees, 32 6.2.4 ROC Review and Approval Funct1on 32 I

6.2.5 ROC Audit Function 33 1

6.3 Procedures.

34-6.3.1 Reactor Operating Procedures 34 6.3.2 Administrative Procedures 35 6.4 Experiaent Review and Approval..

35-6.4.1 Definitions of Experiments 35' 6.4.2 Approved Experiments-35 j

6.4.3 -New Experiments.

36 l

6.5 Required Actions.

36 6.5.1 Action To Be Taken-In the Event A Safety Limit is Exceedad 36 i

6.5.2 Actic

'o Be Taken in The' Event Of A Reportable Occur.. ce 36 6.6 Reports 37

6. 6 '.1 Operating Reports.

38.

6.6.2 Special Reports.

38 6.7 Records 40 L

6.7.1. Records to be Retained for a Period of at Least Five.

Years 40 6.7.2 Records to be Retained for at'Least One Requalification Cycle.

'40 I

6.7.3 R,ecords to be Retained for the Life of the Facility.

40 i

1 13 e

4 1.

Ih'RODUCTION 1.1 Scope This decument constitutes the Technical SpecifJcations for Facility License No. R-75 and supersedes all prior Technical Specifications.

Included are the " Specifications" and the

" Bases" for' the Technical Specifications.

These bases, which provide the technical support for the individual technical specifications, are included for information purposes only.

They are not part of the Technical Specifications, and they do not constitute limitations or requirements to ' which the-licensee must adhere.

This document was. written to be in conformance with ANSI /ANS-15.1-1982.

The content of the Technical Specifications includes:

Definitions, Safety Limits, Limiting Safety System Settings, Limiting Conditions 'for Operation, Surveillance Requirements, Design Features, and Administrative Controls.

1.2 Application 1.2.1 purpose These Technical Specifications have been written specifically for The Ohio State University Research Reactor (OSURR).

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

Nuclear Regulatory Commission on administrative controls, equipment. a va i l a b i'l i t y, and operational parameters, Specificat ions 'are limits. and equipmer,t i requirements for safe I

reactor operation and. for dealing with abnormal situations.

They are typically derived fromi the Safety.. Analysis, Report (SAR). These specifications represent a comprehensive' envelope for safe operation.

-Only. those operational parameters. and equipment requirements'directly.related-to preserving'that safe envelope are listed.

1.1.2. Format-The format of this document is in. general-accordance : with ANSI /ANS-15.1-1982.

L 4

E

=

1.3 Definitions Administrative Controls - those organizational ' and procedural requirements established.by=the Commission and/or the facility-management.

ALARA - as low as'is reasonably achievable.

't 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 - an adjustment of the channel such that its cutput corresponds.with acceptable accuracy to'!,nown values.

of the measured parameter.

Calibration shn!1 encompass the entire 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 verification, where possible, shall -include. comparison 1of the channel with other independent channels. or systems measuring the same variable.

Channel Test the introduction of a signal into the! channel for verification that it is operable.

Cold Clean Core - when the core is at ambient temperature ana 4

the reactivity worth af xenon is negligible.

Commission - tha U.S. Nuclear Regulatory Commission-(or NEC).

'i confinement - a closure on the overall facility'which ~ controls' i

the movement of air-into it.and out of ittthrough a controlled ~-

i path.

J a

f Containment - a testable enclosure which can support a defined'

~

j pressure differential and which.is normally closed.

l-l Control Rod - a device f abricated [from neutron absorbing-

)

material which is used to establish neutron flux l changes.

Control Rod Fuel Element

=a fuel element' capable of holding a control rod.

Controls - mechanisms used L to regulate the operation of-the reactor.

Core - the general arrangement of fuel elements, and control rods.

Critical - when the ef fective multiplication' fa6 tor (k I

I the reactor is equal to unity.

eff~

2 t

e

=

1 Direct Supervision.- in visual'and audible contact..

Excess Reactivity - 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 l

the licensee has the authority to determine all activities as y

per 10CFRiOO.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 interaction parameters of materials.

l Experimental Facility - any structure or device associated with the reactor that is intended to guide, orient,. position, manipulate, or otherwise facilitate completion;of experiments.

Explosive Material any materin) that-is-given an Identification of Reactivity (Stability) Index of 2. 3,'or_4 by the National Fire Protection Association / in its publication

(

704-M, Ident i ficat ion System for Fire Hazards of 'Materlaj,s, or is enumerated in the Handbook for Laboratory Safety Nblished by the Chemical Rubber Company (1967).

J l

Facility - the Reactor Building including ' of fices and laboratories.

Fueled Experiment,- any experiment that contains U-235.or U-233' or Pu-239, not including the normal. reactor fuel elements.

Licensee - The Ohio State University.

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

LCC' are administrativel; established constraints on eqtipment and opert.G ona]

characteristics.

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

Where:a limiting safety system -

setting is specified for a variable on'which a_ safety limit has been placed,' the setting shall' be so chosen ' that ~ automatic protective action will correct the abnormal situation before a safety limit is exct.eded.-

Neasured Value - the value of a parameter as it appears on the output of a channal.

3

i I

b Movable Experiment - one for which it le intended that all.or part of the expe: 4 ment may - be moved in ' relation to the core while the ree:,cor is operating.

Nuclear Regulatory Commission - (NRC).

Onset of Nucleate Bolling - (ONB).

Operable

,.a component or system which is capable of performing its intended functions in a normal manner, 1

Operating a component or system which is performing.its intended function.

Protective Action - the initjation of a signal or the operation of equipment within the reactor safety system in response to a variable or condition of the reactor fac1]!ty having reached a specified limit.

i i

Reactivity Limits - those limits imposed on reactor core excess i

reactivity based upon a reference core condition.

Reactivity Worth of an Experiment - the. maximum absolute value of the reactivity change that would occur as a result of intended or anticipated. changes -or credible malfunctions that alter an experiment's position or configuration.

Reactor - the combinatian of core, permanently installed experimental facilities, control rods, and connected control instrumentation.

Reactor Operating whenever the ' reactor 1is not secured or shutdown.

~ Reactor Operations Committee - (ROC),

t Reactor. Operator (RO) - an individual who is '~ licensed to-manipulate the controls of the. reactor in accordance with

~

10CFR55.

Reactor' Safety Systems those systems, _ includ'ng their s

associated ' input channels, - which; are = designed to: initlate automatic reactor-protection or..to provide informa:lon for l

Initiation of manual protective action.

Reactor % cured - whenever (1) all ' shim / safety r.ods al e fully inserted. (2) the console key.is in the OFF positior and' is removed from the lock, and (3) no in-core work isiin progress involving fuel or experiments or - maintenance of. the core structure, control rods, or control rod drive' mechanisms.

t Reactor Shutdown - when the reactor is suberit.ical by-at least l

1% delta k/k in the.-cold clean core condition.

1

-4

i 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.

a Reportable Occurrence any of the conditions described.i n a

Section G.5.2 of these specifications.

Restejcted Area - the Reactor Building to which access is controlled for purposes of protection of individuals from exposure to radiation and radioactive materials.-

Safety Analysis Report - (SAR).

Safety Channel a measuring or protective channel in the reactor safety system.

Safety Limits (SL) - limits on important. process variables -

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 jnsertion of the shim / safety rods. into the 7

reactor for the purpose of quickly shutting'down the reactor.

I Scram Time - the elapsed time between reaching a: limiting:

safety system setting and the. time when a control rod'is fully inserted.

Secured Experjment - any experiment. experimental facillty, or 5

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 I

which the experiment 1 might be ' subjected from the normal l

environment of the experiment or by forces which can result from credible malfunctions'.

l Senior Reactor Operator (SRO) - an individual who:is licensed to direct the activities of reactor operators.

Such an L

individual may also operate the controls. of the reactor-j h

pursuant to 10CFR55.

1:

l-Shall, Should,. and May the word "shall" is.used to denote. a

~

L requirement; the word "should" to denote a recommendation;;and the word " m a y to denote permisalon, which -is neither a requirement'nor a recommendation.

i

(

Shim / Safety Rods - high-react ivi t'y worth control rods used I-primarily to provide coarse reactor control.

They are-connected electromagnet 1cally to their drive': i

'ianisms and-have scram capabilities.

5 s

Shutdown Margin - the shutdown reactivity necessary to' provide-confidence that the reactor can be made subcritical by means of the control and safety. systems -with the most reactive shim / safety rod and the regulating. rod in the most reactive position (fully withdrawn) and that the: reactor w1)) remain-subcritical 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.

Startup Source - a spontaneous source of neutrons which is used to provide a channel. check of the i startup (fission chamber)-

channel,.and provide neutrons for subcritical multiplication 7

during reactor.startup.

Surveillance Time Intervals - The -average over any extended period for-each. survelliance time interval, shall be closer to the normal surveillance time, e.g.

for the two year jnterval the average shall be closer to two years rather than 30 months.

l two-year (interval not to exceed 30 months),

l annually (interval not to exceed 15 months),-

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

quarterly (interval not to exceed 4 months).

monthly (Interval not.to exceed 6 weeks).

weekly (interval not to. exceed 10 days),

daily (sha)) be donc during the same working day).

Any extension of these intervals: shall be occasional and for a valid reason. rad shall not. affect the average as defined.

True Value - the e.:tual.value of al parameter.

l l-Unscheduled Shutdowns - any unplanned ' shutdown of the reactor caused by actuation of the reactor ; safety systems, operator i

error, equipment malfunction, or a' manual shutdown in response.

to conditions which could adversely affect safe _ operation.

They do not' include those shutdowns Jresulting from; expected testing operations, or planned shutdowns. whether Initiated by.

l l

controlled insertion of control rods or planned manual! scrams.

1 I

f 6

i

2.

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

This specifjeation ' applies to the melting temperature of the aluminum fuel cladding.

Objective:

The objective is to assure that the integrity of the fuel cladding is maintained, s

Spegfrication:

The reactor fuel temperature shall be less than 550 C.

Bases:

The melting temperature of aluminum is 600 C (1220 F).

The blister threshold temperature for U S1 dispersion fuel g

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

Because the objective of this specification 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 untj] shown otherwise.

2.2 Limiting Safety System Settings Applicability:

This specification applies to the following items assocjated with core thermodynamics:

(1) Reactor Thermal' Power Level'and t

(2) Reactor Coolant Inlet Temperature.

Objective:

To assure that the fuel cladding. integrity is maintained, Specification:

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

j.

thermal, l

(2) Reactor safety systems settings shall'. initiate automatic

~

protective action so.that reactor thermal power. level:

l shall not exceed 600 kw. (120%; of - full power) during a

]

-transient.

I~

(3) Reactor safety systems settings shall' initiate automatic protective action so that core ' inlet watet temperature L

shall not exceed 35 C' Bases:

The criterion for ih: re cafety system. 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 power level of 600 kw and a core inlet temperature of ' 35 C ' (normal;

[

7

I inlet temperature ja = 20-25'C),

The maximum credible accident analysis is provided in Section 8.4.3.2 of the Safety Analysis Report.

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

The maximum.

0 temperature of the cladding reaches 910 (SAR 3. 4.3.3 ).

One say also reference SAR Sections 4.8.1. 4.8.2 for an estimate of:

cladding temperature during steady state-operation at 500 kw (56.5"C).

e i

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} sy 3.

LIMITING CONDITIONS FOR OPERATION 3.1 Reactor Core Parameters 3,1.1 Reactjvjty i

Applicability:

These specifications apply to the reactivity condition of the reactor and the reactivity worths of the

~

shim / safety rods and regulating rod under any operating.

conditions.

Objective:

To ensure safe shutdown of the reactor and that the t

safety limita are not exceedt.d.

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 2.6% delta k/k underi anyL operating.

l condition.

(2) The minimum shutdown margin under any operating condition 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 totnl reactivjty worth of th'e regulating rod shall be less than 0.% delta k/k.

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

l (5) The moderator temperature coefficient - shal1 -- be - negative i

and shall have a min um absolute macMy value of at least 2 x 10~s/ C across the active' core at all normal operating temperatures.

(6)

The moderator vold 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 j

Dases:

l-(1) The - maximum allowed. excess ' reactivity ' of 2f."6% delta >k/k

.l provides sufficient reactivity to, accommodate fucl burnup, xenon. buildup, experiments, control' requirements, and fuel

'and moderator temperature feedback,(Sectioi 4.2 of the SAR).

Also, calculatio's show that this excess reettivity

- anores that the maximum. emperature of the surface of the cladding will-be well belok the bl.ister threshold ; of the U S1 fuel during a design basis accident (SAR 8.4.3.2).

3 2 9

4

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j (2) The minimum shutdown margin ensures that. the reactor can be shutdown from any operating condition and remain-shutdown' af ter cooling and xenon-decay even with the highest worth rod and:the regulating rod fully withdrawn.

(3) Lim! ting the reactivity worth of the regulating rod to a value less than the effective delayed neutron fraction, assures that a f ailure. of the automatic servo control system cannot result in a prompt critical condition.

(4) The requirement that all grjd. positions be filled during reactor operation assures that -the volume flow rate of primary coolant which bypasses the heat producing. elements

't will be within the range specified in Section 4.8 of the SAR.

Furti ormore, the possibility of accidentally'.

dropping an object into a grjd ' position and. causing increase of reactivity 18 precluded.

(5) A negative moderator temperature coefficient.of reactivity assures that any moderator ~ temperature'rlse elll cause a decrease in reactivity.

The U Si fuel also has a.

significantnegativetemperaturecoekfiklentofreactivity'

~

due to the g ppler broadening of neutron capture resonances in U, but no credit is taken for this'effect in our Safety analyses.

l (6)

A negative void _ coefficient. of reactivity helps ' provide i

reactor stability in the event of.' moderator' displacement by experimental devicea or other means.

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w.

l 10 l

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j 3.2 Reactor Control and Safety System i

3.2.]

Control Rod Drop Times i

Applicability: This specification applies 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.

Objective:

To ensure that the reactor can be shutdown within a specified period of time.

Specification:

The reactor will not be operated unless the drop time of each of the three shim / safety rods is less than 600 maec.

Dases:

Control rod drop times as specified ensure that the safety 11mit 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 Applicability: This applies to the maximum positive reactivity insertion rate by the most reactive shim / safety rod and the regulating rod simultaneously.

l Objective:

To ensure the reactor is operated safely :and the i

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.

Basit:

This maximum reactivity insertjon rate assures-that the Safety Limit will not be exceeded during a startup accident due-to a short period generated by La continuous linear reactivity -

l insertion.

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 i

l system channels that shall be operable to ensure the Safety l

Limits are not exceeded by ensuring the reactor can be shutdown

)

at all times.

Speci fica tion:

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

11

Reactor Safety System Minimum Function Component Required 1.

Core !! 0 Inlet Temp.

1 Slow scram if temp. 1 35 C 2

2.

Reactor Thermal 2

Fast scram if thermal power power level 1 600 kw. as indicated on (Safety Channels) calibrated Jonization.

chamber channels.

3.

Reactor Period 1

Fast scram if period $ 1 sec 4.

Reactor Thermal 1

Slow scram if coolant system power level / coolant pumps.not.on by > 120 kw 1

system pumps thermal power 5.

Coolant Flow Rate 1

Slow scram if coolant system nas no-flow (primary) by a

1 120 kw theraal power 6.

Pool Water Level 1

Slow scram if pool level 3 20 feet (15 feet above core) 7.

Switches 6

Slow-scram if any one. switch u.

Magnet Power Key "On" la not properly setL at the

,I b.

Startup Cal-Use position indicated.in in "Use" quotes..(Also prohibits c.

Period Generator startup)

Switch "Off" d.

LOG-N Amp Calibrate Sw?tch " Norm" j

e.

LOG-Period Amp Calibrate l

Switch " Norm" f.

Effluent Monitor Compressor "On" 8.

Recorders 5

Slow scram if power is-lost l

a.

LOG-N to any one of the' listed l

b.

Linear Level recorders c.

Start-Up Channel d.

Period e.

Effluent Monitor 9.

Manual Scrams 5

Slow scram upon activation j

a.

Control Room Console of any one manual scram b.

Pool Top Catwalk switch c.

BSF Catwalk d.

Rabbit /BP Area e.

Thermal Column /BP Area 10.

Compensated lon Chambers 2

Slow scram if voltage drops below o p e r a t.1 o n a l specifications 12

]

Reactor Safety System Minimum Function Component / Channel Reaufred 12.

Safety Set Points On 4-Slow scram if associated Recorders recorder values are exceeded a.

Period I 5 sec.

b.

Linear Level 1 120% of licensed power c.

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

d.

Start-Up 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 a

Mechanisms any control rod which has t

excess magnet current 160 ma Bases:

1.

Assures safety limit is not exceeded; core inlet temperature is same as l

cooling system outlet 2.

Assures safety limit is not exceeded.

3.

Assures safety limit is not e.:ceeded 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.

6.

Assures there is enough primary coolant for. natural, convection cooling 7.

Assures nuclear instrumentation is in proper mode for' operation 8.

Assu.as information is available for observation. by the reactor q

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 l

deemed necessary by other reactor staff L

10 Assures shutdown if nuclear instrumentation-falls 11..

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

Assures shutdown i f ' set vo operation. varies tao L greatly.

Assures shutdown if count rate-is too low to provide meaningful startup information.

The startup interlock may be bypassed if K [ns'is <.9-components of the; safety. system. are

• talled and 12.

Assures all operational.

13.

Assures that any control rod exhibiting excess magnet

• current will be-released and-fall to the bottom due to gravity 13

O e

l 3.3 Coolant System 3.3.1 pump Reautrement9 Applicability:

Thih specification applies to the operation'of pumps for both the primary and secondary coolant loops.

Objectivo:

To ensure that both pumps are functioning whenever the reactor la operated above 120 kw.

I Specification:

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

Bases.

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

3.3.2 Coolant Level Applicability: This specification applies to the height of the-water in the Reactor Pool above the core.

Object tve:

To ensure there is adequate primary coolant in the Reactor Pool and sufficient c biological shielding ' above the

core, i

Specification:

The reactor shall not be operated unless there i

is 20 feet of water in the reactor pool and 15 feet of water l

above the core, Bases: With the pool full of water to a level of=20 feet there is adequate primary coolant for natural convection _ cooling.

l 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.

3.3.3 Water Chemistry Recuirements 1

Applicability: This specification appliet to the purity of the-l primary coolant-water.

l-l Objective; To minimize corrosion.of the cladding on the fuel i

elements, and to reduce the probability of neutron - activation of ions in the water.

14-

Specification:

(1) The conductivity of the pool water shall not' exceed the limit of 2.0 p sho/cm.

(2) The pil of the pool water shall not exceed 8'.0.

Bases:

Operation in a.'co rd a n ce with these specification' ensures aluminum corrosion s within acceptable limits, and that the concentration of dissolved impurities that could be-activated by neutron irradiation remains within acceptable

limits, 4

3.3.4 Lenk or Loss of Coolant Detection

-l 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.

Specification:

The pool water. level shall be,at least 15 feet above tiie top of the fueltin the: core.

Bases:

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

Desj n criteria of. the cooling -system to g

prevent large losses of pool water - due ' to siphoning are discussed in.Section 3.2.2.1 of the SAR.

3.3.5 primary and Secondary Coolant Activity Limits

(

Applicability:

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

Objective: To ensure there is a level low enough so'as.not-to exceed ~10CFR20 limits if' coolant is. released to the L sanitary sewer system.

Specification:

The primary and secondary coolant system shall be tmonitored for the buildup of. radioactivity. and. analyzed at L

least semiannually for increase in the_ concentration off radionuclides.

Dasis:

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

15 s

i 3.4 Confinement Isolation Applicability:

This specification applies to the capability of 1solating the reactor butiding from the unrestricted area outside.

Objective:

To. prevent the exposure of the public to airborne radioactivity exceeding the limits of 100FR20, and the ALARA I

principle.

Specification:

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

i (1) Ventilation fan operating-(2) Heactor Building bay door closed (3) Reactor Ballding 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 r

confined and 1imited to the extent analyzed-inLthe revised SAR of September 1987.

3.5 Ventilation Systems App]Icability:

This - specification applies to all heating, ventilating, and air conditioning systems that exhaust building j

air to the outside environment.

I

(

Objective: To provide for normal ventilation.and the' reduction of airborne radioactivity within the reactor building during -

normal reactor operation 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 2000 cfm shall.

-be operable whenever-the reactor is operating.

(2) This fan, as well as all Other heating, ' ventilating, and air conditioning. systems shall have the capability. to be shut off from a single switch in the control Yoom.

Bases:

In the unlikely event of a release-of fission products or other airborne radioactivity, the ventilation 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 the SAR.

16 i

3.6 Radiation Monitoring Systems and Radioactive Effluents 3.6.1 Radiation Monitoring i

Applicability:

This specification applies to the availability-l of radiation monitoring equipment which shall be operable-during reactor operation.

Objective:

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

Specification:

1 (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 experimentalfacility is used, the rabbit monitoring system shull be operating and have a. readout and alarm in the control room.

(3) When the reactor is operating, the following Area Radjation Monitors (ARMS) shall be operating and have both local and control room' readouts and alarms, f

a.

Pool Top b.

Primary Cooling System c.

Beam Port / Rabbit Area d.

Thermal Column Area I

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

l (5) Portable instruments, surveys, or analyses may be-substituted for the instruments 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 />.

Read-out and alarms from-these temporary instruments shall be reported to the reactor operator on' duty at least once.

per hour, s

Bases:

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

During " normal" mode > operation. It will sample and monitor air; just before it L is released from the reactor building.

(SAR 6.3.1)

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

17

.l I

i (2) The rabbit stack monitor is.used with the-rabbit since the

{

rabbit system uses air as its transport mechanism and Ar -

1 41 production takes place.

This monitor will provide warning if Ar-41 levels being released in the building are-I too high (SAR G.3.2 and 6.3.4.3)

(3) The ARMS' provide a continuing evaluation of the radthtlon-i levels within the Reactor Building (SAR 3.7) and provide a.

warning if levels are higher than anticipated.

(4) The availability of survey meters - enables the Reactor.

Staff to independently confirm radiation levels,throughout-the building.

i (5) in the event of instrument f all'ure, s hor t term substitutions will enable the safe continued operation ~ of the Reactor.

3.6.2

_Radloactive Effluents Applicability -This specification applies to the monitoring of radioactive effluents from the facility.

Objectives:

1 (1) To ensure that-liquid radioactive releases' are safe and legal.

(2)- To ensure ihat the release of Ar-41.beyond. the site.

I boundary does not result in concentration above.MPC ' for unrestricted area.

(

(3) To assure that the release of Ar-41 in the

-(

restricted area does not result in' concentrations above MPC.

Specifications:

(1) The release rate for radioactive'11gulds beyond thh site' boundary shall: not exceed the limits as specified 'in 10CFR20 at the point of release.

(2) The concentration of Ar-41: at the point 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 I

consecutive days.

18

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3 Bases:

1 (1) The basjs for this specification is found in Section 6.2 l

of the Safety Analysis Report..

(2) The basis for this specification ja found in Section 6.3

{

of the Safety Analysis Report.

=;

i (3) The basjs for this specification - is found in l

Section 6.3 of the' Safety Analysis Report and j

10CFR20.103(b)(2).

-i 3.7 Experimente 3.7.1 Reactivity Limits Applicabi11ty: This spectfication applies to experiments to be-I installed in or near the reactor and associated experimental facilities.

Objectives:

To prevent damage to the reactor or excessive release of radioactive materiala in the event of an experiment

failure, i

Specification:

(1) The absolute' value of the reactivity worth of any single secured experiment shall siot exceed 0.7% delta k/k -

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

(3) The absolute value or=the reactivity worth of all movable-exportments 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 less than 0.05% delta k/k per.second.

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

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

m i

i 19

._m

Bases:

(1) The bases for specifications 1, 2, 3, and 6 are found in Section 8.4.3.2 of the SAR mhich evaluates a Otep insertion of reactivity from an experiment.

(2) The bases for specifications /. and 5 allows for certain reactor kinetics experiments to be performed but still 1.imits the rate of change of reactivity insertions to levels that have been analyzed.

Section 8.4.3.2 of the SAR evaluates a step insertion of reactivity from an experiment.

3.7.2 Design and Materials SpecifIcatlon:

(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.

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

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

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

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

of gunpowder.

Iluses:

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

(2) Limiting corrosive materials in Specification 2,

and explosives in Specification 3 reduces the likelihood of damage to reactor components and/or releases of i

radioactivity resulting from experiment failure.

(3)

Limiting exploelve materials to _ neutron,. radiographic exposures done. outside cf the core and experimental facilities rMuces the likelihood of damage resulting for this experimental failure.

20-

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SURVEILLANCE REQUIREMENTS 4.1 Reactor Core Parameters 4.1.1 Excess Reactivity and Shutdown Marcin Applicability:

This specification applies to surveillance requirements for determining the excess reactivity of the reactor core and its shutdown margin.

Objective:

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

Specifications:

(1) Whenever a net change in core configuration, for which the predicted change in reactivity 16

>.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 shall be determined annually, lia ses :

A determination of excess reactivity is needed to preclude operating without adequate shutdown margin.

Moving a component out of the coras and returning it to its same location la not a change in the core configuration and does not require a determination of excess reactivity, i

4.1.2 Fuel Element _n I

Applicability:

This specification applies to surveillance requirements for deiermining the physical condition of the reactor fuel.

Objective:

To chaure that visible deterioration.

corroulon, or other physical changes to the fuel elements are detected in a timely manner, Specification:

All fuel elements, both in-core and out,

'i shall be visually inspected at leact once every five years, by inspecting at least one fifth of the elements annually.

Basis:

If the water purity is continuously maintained within specified limits, it is projected that chemical corrosion of the fuel clad will proceed-slowly. " However, faults in the basic materials or fabrication could lead to loss of cladding integrity.

21-

4.2 Reactor Control and Safety Systess i

4.2.1 Control Roda Applicability:

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

Objective:

To assure that all rods are operable.

Specifications:

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

(2) Shim safety-rod drop and drive times and regulating rod drive tjme shall be determined annually or after maintenance or modification is completed on a mechanism.

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

Bases:

The reactivity worth of the rods in measured to assure i

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 changes assures that altered rod worths will be known prior to continued operations.

The visual inspection of the rods and measurements of drive and drop times are made to assure the rods-are capable of performing properly.

Verification of operability af ter I

maintenance or modification of the control system will ensure proper reinstallation.

4.2.2 Reactor Safety System Applicability:

This specifjcation applies to the surveillance requirements for the Henctor Safety System.

Objective:

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

l Specification:

(1) A channel check of each measuring channel shall be perfo med 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 operation extending more than one day.

22

i (3) A channel calibration of the reactor power level measuring channels shall be made annually.

(Linear Leve] and LOO-N.)

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

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

(5) A channel calibration of the following shall 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 annua))y.

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

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

l (9) Any instrument repair or replacement shall have a channe]

l test prior to reactor operation.

Bases:

The daily channel tests and checks will assure that the 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 l

Applicability:

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

Objective:

To assure high quality pool water.

Specification:

The conductivity and pH of the poo) water sha))

be measured weekly.

Bases:

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

23

4.3.2 Coolant Svatem Radioactivity Applicability:

This specification applies to the radioactive material in the orlmary coolant or secondary coolant.

Objective:

To identify radionuclides as potential sources of release to the sanitary sewer system.

Specification:

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

Bases:

Radionuclide analysis of the pool water or secondary coolant allows for deterr nation of any significant buildup of fisslon or activatjou products and helps assure that radioactivity is not permitted to escape to the tertjary system in an uncontrolled manner.

4.4 Confinement Applicability:

This specification applies to the surveillance requirements for building confinement.

Objective To assure that the building closure capability exists.

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

11ases : Monthly surveillance of this equipment will verify that the confinement of the reactor bay can be maintained if needed.

?

4.5 Ventilation System Applicability:

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

Objective:

To assure that the ventilation system functions satisfactorily.

Specification (1) Ventilation fans and closures shall be checked for proper operation on a quarterly basis.

(2) The shutoff switch for all fans and air

• conditioning -

systems shall be tested on a quarterly basis.

Bases:

This surveillance will assure that during normal operations the airborne radioactivity will be minimized inside the building and that the building can be isolated quickly if necessary to prevent uncontrolled escape of ajr-borne radioactivity to the unrestricted environment, 24

l o

4.6 Radiation Monitoring Systems and Radioactive Effluents 4.6.1

,hffluent Monitor Applicability:

This specification applies to the surve11)ance 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 operation.

Bases:

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

4.6.2 Rabbit Vent Monitor App]!cability:

This specification applies to the survel))ance requirements of the rabbit vent monitor.

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

• Specification:

The monitor shall have a channel calibration annun))y and a channel test before each day's reactor operation.

Danes:

The calibration will assure effluent releases inside the building are accurately estimated and the test will assure the monitor is operable before the rabbit is used.

i 4.6.3 Area Radiation Monitors (ARMS)

Applicability:

This specification applies to the area radiation monitoring equipment.

Objective:.

To assure that radiation monitoring equipment is operable whenever the reactor is operating.

~

Specification:

A channel test of the ARMS shal1 be completed before each day's operation and a channel calibration shall be completed annually.

Bases:

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

25

d a

i l

4.6.4 Portable Survey instrumentation Applicability:

This specifjcation applies to the portable survey inntrumentation available to measure beta-gamma exposure rates and neutron dose rates.

Objective:

To assure that radiation survey instrumentation is operable whenever the reactor is operating.

Specification -

Beta-gamma and neutron survey meters shall be tested for operability each day the reactor is to be operated and shall be calibrated annually.

i Bases:

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

b b

9 W

5 26 1

5.

DI. SIGN FEE'URES 5.) Site and Fac.'lity Description 5.1.1 Facility I,ocation The reactor and associated equipment is housed in a building at 1298 Kinnear Road, Columbus, Ohio.

The minimum free air voluge of the building housing the reactor will be > 70,000 ft.

There is an exhaust fan with dampers providinct for control of release of airborne radioactivity, it is in the area of The Ohio State University Research Center.

5.1.2 Exclusion and Restricted Area The fence surrounding the Research Jenter shal] describe the exclusion area.

The restricted t.rea as defined in 10CFR20 shall consist of the Reactor Build'.ng.

5.2 Reactor Coolant System 5.2.1 Primary Coolant Loop Natural convectjve coo]Ing la the primary means of heat removal from the core.

Water enters the core at the bottom and flows upward through the flow channels in the fuel elements.

5.2.2 Secondary and Tertiary Coolant Loops The secondary coolant loop removes heat from the primary coolant.

The secondary coolant (ethylene glyco] and water) i passes through two separate heat exchangers to remoge heat if-necestary.

If the outside air temperature is S 78 F then an outside fan-forced drycooler is sufficient to remove all heat generated at 500 kw.

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

It provides additional cooling for the secondary coolant.

5.3 Reactor Core and Fuel Up to 30 positions on the core Erid plate are available for use as fuel element positions.

Control rod fuel elements occupy four of these positions and one is reserved fot. the Central Irradiation Facility flux trap.

Several arrangements for the

cold, clean, critica! core have been investigated.

Approximately six.een standard fuel elements in addition to the control rod fuel elements will be required.

Partial elements, core plugs, and graphite elements say be utilized 'in 'various combinations to echieve the proper K excess, 27

]

The reactor fuel is The DOE Standard uranium-silicide (U Si.,)

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

It is flat blate fuel with a " meat

• thickness of 0.020" and aluminum cladding of

)

0.015".

Standard fuel elements have a total of 16 fueled i

plates and 2 outer purs aluminum plates.

The control rod fuel elements have eight of the inner fuel pDites removed to allow the control rods to enttr.

Pure aluminum guide plates are on the inside of this gap.

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

vartial elements are =!ao available 1

with 25, 40, 50, and 60 percent of the nomira; loading of a standard element.

These partial fuel elements are prefabricated by the vendor with fixed numbers of plates.

(1)

References:

NRC NOALG 1813 ANL/RERTR/i?4-10 ANL/NERTR/TM-11 5.4 Fuel Storage The fuel storage pit located below the floor of the reactor pool and at the er.d opposite trom the core, shall be flooded with water wbcaever 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'erfsurethatnofueltemperaturesreach I

a point at which OND la possible, The two fuel storage racks located in the Bulk Shielding Facility storage pool shall each:

(a) Contain no more than 16 fuel elemente spaced on a pitch of at least 6 inches in a two by e18ht matrix.

(b) De placed no closer than 24 inches in'any direction from each other or ar.y other fuel storage fac1]ity.

(c)

Have a K less than 0.90 when fully loaded with fuel and floodedUNbwater.

5.5 Fuel Handling Tools All tools designed for or capable of removing fuel from core positions or storage rack positions shall-be secured when not in use by a system controlled by the supervisor of reactor operations, or the senior reactor operatot on duty.

28

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2 i

6.

ADMINISTRATIVE CONTR0!.S 6.1 Organization 6.1.1 St ructure The Ohio State University Research Reactor is a part of the College of Engineering administered by the Engineering Experiment Station.

The organizational structure is shown in Figure 6.1.

6.1.0 gesponalbilit y The Director of the Engineering Experiment Station (Level 1) f

• the contact person for communications between the U.S. Nuclear Regulatory Commission and The Ohio State University.

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

The Associate Director (or Manager of Reactor Operations)

(!.evel 3) shull be responsible for the day-to-day operation and for ensuring that all operations are conducted in a safe manner and within the limits prescribed by the facility license and Technical Specifications.

During ' periods when the Associate Director is absent, his responsibilities are delegated to a Senior Reactor Operator (Level 4).

6.1.3 St a f finn During Reactor Operations:

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

reactor operator, shall be in the building during all l

reactor operations.

The second shall be capable of' I

following simple written instructions for shutting down the reactor.

(2) At least two licensed operators should be in the building during any extended operations (longer than 60 minutes).

(3) Two persons, one. of whom shall-be a licensed senior reactor operator, shall be in the building for the first start-up of the day.

a (4) Two persons,. one of whom shall be a lisensed senior reactor operator, shall be in the building during start-up af ter an unplanned shutdown.

(f>) During all operations, a licensed operator shall be in the control room either as console operator or directing - the activities of a student operator or trainee.

29

e l

)

President 1

3 Provost Vice-President 1

for Health Services

Dean, College of Engineering i

i Reactor Operations

+

Committee I

I Director, Engineering I

Experiment Station I

(Level 1)

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l l

Director, Nuclear Reactor I

Director,

~l Laboratory Office Radiation Safety I

(Level 2)

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Associate Director, Nuclear Reactor Laboratory (Level 3) i

[

Senior Reactor Operator (Level 4)

[

Reactor Operations Staff i

l.

i Solid Lines Paths'of Direot' Responsibility?

Dashed Lines ---------- Paths of.Information.

Figure 6,1

' Administrative Organization'-

1 I;

30 I

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l (6) A minimum of three people shall be present during fuel j

handling.

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

6.1.4 Selection and Training 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 Research Henctors, ANSI /ANS-15.4-1977, Sections 4-6, 6.2 Review and Audit There shall be a Henctor Operations Committee (HOC) which shall review and audit reactor operations to assure the f acili ty is operating in a manner consistent with public safety and within the terms of the facility license.

The Committee advises the Directrr of the NHL, and is responsible to the Provost of The Ohjo State

[

University.

6.2.1 Composit ion nnd Oualificationn of the ROC Committee members shall be appointe annually by the Provost of The Ohio State University.

The Committee sha)) be composed of at it ast nine members including ex-officio members.

The Director tnd Associate Director of the Nuclear Reactor Laboratory, and toe Director of the Office of Radiation Safety sht.11 be ex-officio voting members of the Committee.

The remaining Committee members shall be faculty, staff, and student representatives of The Ohio State University, having professional backgrounds in engineering, physical, biological, or med. cal sciences, as well as knowledge of and interest in opplIcat to is of nuclear technology and ionizing rndlation, t,.2.2 HOC Meetinca The Committee shall moot at Icast once each-quarter.

It should meet during the first t40 weeks of each calendar quartar. A quorum shall consist of at least 50 percent of the earloers who are not directly involved in or responsible for facliity operations.

Ex-officio members shall be counted in the quorum as follows:

(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 cornting a quorum, but does have the right to vote.

(2) Ex-officio members who are under the authority of the Provost serve in the same capacity as those who are appointed by the-

Provost, f.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 operations.

31

(3) 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.

G.2.5 sub-Committees The chairperson may appoint a Subcommittee from within the Committee membership to act on behalf of the full committee on those matters which cannot await the regular quarterly meeting.

The. full Committee shall review the actions ' taken by the Subcommittee at the next regular meeting.

6.2.4 ROC Review end Approval Function the responsibilltles of the ROC include, but are not' limited to the folJewing:

(1) Review and approval of experiments utllizing the reactor facilities (2) Review and approval of procedures (3) Review and approval of all prop.ised 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 100FR50.59 (5) Review of audit reports (6) Review of abnormal performance of. plant. equipment and operating abnormalities having safety significance (7) Review of unusual occurrences a'id incidents which are reportable under 10CFR19,- 20, 21, ar.d 50, or Section 6.6.4 of this document, and (8) Review of violations of technical specifications, license, or

(

procedures having safety significante.

Relative to item (1), responsibility for review of experiments on a-day-to-day basis shall lie with the Director of-the Nuclear Reactor Laboratory 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 perforned experiments shall be provided for ROC review, 32

l i

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

He shall provide a summary of all procedure changes to the ROC for their review and approval.

A complete set of minutes of all Committee and Subcommittee meetings, including copies of all documentary material reviewed, and all approvals, disapprovals, and recommendations shall be kept.

Minutes or reports of all Committee meetings or Subcommittee meeting should be disseminated to the Committee members prior 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 L

of the Engineering Experiment Station in a timely manner.

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

At lehnt two individuals on the Audit Subcommittee shall be RCC 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 pe' forming.

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

In this way, each Subcommitt ee should consist of two haldovers and one new member.

The ; ember serving for his or her second audit should be the Audit S.ebcommittee Chairperson. The following ittma 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, (4) The fuellity Security plan and implementing procedures, and (5) The results of actions taken to correct any deficiencies that affect reactor safety, and (6)

Conformance with the ALARA policy and the ef fectiveness of radiologic control, l

l.

l Deficiencies found by the Audit Subcommittee that affect Reactor 1

Salety, shall be reported 'immediately to the Director of the

)

Engineering Experiment Station.

A written report of audit findings should be submitted to the Director of the Engineering Experiment l

Station and the full Reactor Operations Committee within three months of the audit's completion.

33-

h 6.3 Procedures 6.3.1 Reactor _OperatinF Proceduren Written procedures, reviewed and approved by the Director and the ROC, shall 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 NRb staff and subsequently reviewed by the ROC.

At least the following items shall be covered:

(1) Startup, operition, and shutdown of the reactor, L

(2)

Installation, removal, or movement of fuel elements, control rods, experiments, and exper! mental 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, (4) EmerEency conditions involving potential or actual release of radioactivity including provisions for evacuation, re-entry, recovery, and medical support, (5) preventive and corrective maintenance procedures for-systems which could have an 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 end requalification plans, and (8)

Personnel radiation protection.

1 1

34

\\

6.3.2 Administrative procedures procedures shall also be written and maintained to assure t

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 includer (1) Audits, (2) Special Nuclear Material accounting, (3) Operator requalification, (4) Record keeping, and (5) procedure writing and approval.

6.4 Experiment Review and Approval 6.4.1 Definitions of Experiments-Approved experiments are those which have previously.been reviewed and approved by the ROC.

They sha)) be documented and may be included as part of the procedures Manual.

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

Routine tests and maintenance activities are not experiments.

6.4.2 Approved Experiments A)) proposed experiments utilizing the reactor shall be evaluated by the experimenter and a licensed Senior Reactor Operator to assure compliance with the provisions of the utilization license, the Technical Specifications, and 100FR parts 20 and 50.

If, in i

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 perlinent, the evaluation shall include considerations of:

(1) The reactivity worth of the experiment r

(2) The integrity of the experiment, including the effects of changes in temperature, pressure, or chemical composition l

(3) Any physical or chemical interaction that could occur with the reactor components, and (4) Any radiation hazard that may result from the activation'of materials or from external beams.

35-

6.4.3 New Experiments prjor to performing an experiment not previously approved for the reactor, the experiment shall be reviewed and approved by-the Reactor Operations Committee.

Committee review shall consider the following information:

(1) The purpose of the experiment, t

(2) The procedure for the performance of the experiment, and (3) The safety evaluation previously reviewed by a licensed Senior Reactor Operator.

6.5 Required Actions 6.5.1 Action To De Taken in the Event A Safety Limit is Exceeded (1) The reactor shall be shut down, and reactor operations shall l

not be resumed until authorized by the NRC.

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

(3) The safety limit violatJon 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 violation report shall be prepared. The report shall describe the following:

a.

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

factors, i

b.

Effect of the violation upon reactor facility components, l

systems, or structures and on the health and safety of personnel and the public, and c.

Corrective action to be taken to prevent recurrence.

(5) The report shall be reviewed by the Reactor Operations Committee and shall be submitted to the NRC within 14 working days when authorJzatJon ia sought to resume. operation of the reactor.

6.5.2 Act ion To Be Taken In The Event Of A Reportable Occurrence A report.able occurrence is any of the following conditions:

(1) Opsrating with any safety system setting less conservative ttan stated in these specifications, 36

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

(3) Safety 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 uncontro))ed or unanticipated increase in reactivity in excess of.4% delta k/k, (S) An observed inadequacy in the implementation of either administrative or procedural ;ontrols, such that the inadequacy could have caused the existence or development of an unsafe condition in connection with the oneration 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.'imits of personnel and/or the environment.

(7)

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

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

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

(2) The Director of the Reactor Laboratory. shall be notified as soon as possible and corrective action shall,be taken before resubing the operation involved.

(3) A written report of the occurrence shall be made which shall include an analysis of the cause of - the occurrence, the corrective action taken, and the recommendations for measures to preclude or reduce the probability of recu rence.

This report shall be submitted to. the Director and the Reactor l

Operations Committee for review and approval.

(4) A report shn)) be submitted to the Nuclear Regulatory Commission in accordance with Section 6.6.2 of these specifications.

6.6 Reports Reports shall be made to the Nuclear Regulatory Commission as follows:

37

L 6.6.1 Operatinc Reports An annual report shall be made by September 30 of each year to the Director, Office of Nuclear Reactor Regulation. NRC, Washington, DC 20555, with a copy to the NRC,f ' ton III, in accordance with 10CFR 50.4, providing the following it a,mation:

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

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

[

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

(3) The results of safety-related main'enance and inspections.

The reasons for corrective maintenance of safety related items sha)) be included.

(4) A table of unscheduled shutdowns and inadvertent scrams, including their reasons and the corrective actions taken.

(5) A summary of the Safety Analyses performed'in connection with changes to the facility or procedures, which affect reactor safety, and performance of tests or experiments carried out under the conditions of Section 50.59 of 10CRF50.

(0) A sunimary of the nature and amount of radioactive gaseous, liquid, and solid effluents released or discharged - to the environs beyond the effective control of the licensee as measured or calculated at or prior to the point of such release or discharge.

l l

(7) A summary of radiation exposures received by facility l

personnel and visitors, including the dates and times of significant exposures.

6.0.2 Special Reports (1) A telephone or telegraph report of the following shall be submitted as soon as possible, but no later than the next working day, to the NRC Region III Offices.

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

l (b) Any exceeding of the safety limit as defined in Section 2.1 of these specifications.

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

38

o v

t (2) A written report shall be submitted within 24 days to the Director, Office of Nuclear Reactor Regulation, US NRC, Washington, DC 20555 with a copy to the NRC Region III, in accordance with 10CFR 50.4, of the following:

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

(b) Any exceeding of the safety limit 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, Office in accordance with 10CFR 50.4, of the following:

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

(4) A report shall be subitted 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 r

Director, Office of Nuclear Reactor Regulation, U.S.

NRC, Washington, DC 20555, with a copy to the NRC, Region III 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.

l 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 i

the following:

(a) Total control rod reactivity worth, (b) Reactivity north of the single control rod of highest reactjvity worth, and I

(c) Minimum shutdown margin both at ' ambient and _ operating temperatures.

39

V (d) Excess reactivity (e) Calibration of operating power levels (f) Radiation leakage outside the biological shielding (g)

Release of radiocctive effluents to the unrep+ricted environment.

0.7 Records Records or logs of the items listed below shall be kept in a manner convenient for review, and ahall be retained for as long as indicated.

6.7.1 Hecords to t,e Retained for a period of at Leant Five Years (1) normal plant operation, (2) principal maintenance activities, (3) e:

• 'ments performed with the reactor, (4) re ble occurrences, (5) equ.pment and component surveillance activity, (G) facility radiation and contamination surveys, (7) transfer of radioactive material, (8) changes to operating procedures, and-(9) minutes of Reactor Operations Committee meetings.

6.7.2 Records to be Retnined for at Least One Requalification Cvele Regarding retraining and requalification of licensed operations i

personn.al, the records of the most recent complete requalification cycle shall be maintained-at all times the individual is employed, i

l 6.1.3 Records to be Retained for the Life of the Facility

[

L (3) gaseous and 11guld radioactive effluents released to the environment, (2) fuel inventories and transfers, (3) radiation exposures for all personnel, (4) changes to reactor systems, components', or equiprent that may affect reactor safety, (5). updated, corrected, and as-built drawings of the facility.

40-

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_l (6). records of-significant spills of' radioactivity,,and status,.

(7) annual operating reports provided to the NRC:.

(8) copies.cf NRC inspection reports.'and related correspondence.

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