ML20199K534

Proposed Tech Specs,Revising Surveillance Requirements,Site Description & Fuel Assemblies

ML20199K534
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Site:	Purdue University
Issue date:	06/30/1986
From:	PURDUE UNIV., WEST LAFAYETTE, IN
To:
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ML20199K530	List: ML20199K525 ML20199K534
References
NUDOCS 8607090185
Download: ML20199K534 (42)

Text

LS.

i.0 SURVEILLANCE REOUIREMENTS a.1 Reactivitv Limits Apolicability

• his specificacion applies to the surveillance requirements for reactivity 11mics.

Obiective The objective is to assure that the reactivi:7 limits of Specificacion 3.1 are not exceeded.

Soecificacion

a. The shim-saf ecy rod reactivity worths shall be measured and the shuc-down margin calculated at intervals noc to exceed 15 months , and when-ever a core configuracion is loaded for which shim-saf ety rod worths have noc been measured.

b. The shim-saf ety rods shall be visually inspected at intervals not to exceed 15 monchs. If the rod is found to be deceriorated, it shall be replaced'vich a rod of equivalent or greater vorth,

c. The reactivity worth of experiments placed in the ?UR-1 shall be measured during the first startup subsequenc to the experiment's inser-ion and shall be verified if core configuracion changes cause increases in experiment reactivity worth which may cause the experiment worth to l

l l exceed the values specified in Specificacion 3.1.

3ases Specification a.1.a vill assure : hat shim-saf ety rod reactivity wor hs are ucc degraded or changed by core manipulacions which cause these rods to operace in regions where : heir effectiveness is reduced.

• he boron scainless steel shim-safety rods have been in use at the ?.UR-1 since 1062, and over :his period of ime, no cracks or other evidence of f

l decerioracion have been observed. 3ased on this performance and the 8607090185 860630 PDR ADOCK 05000182 P PDR

• 3.0 DESIGN TEATUR2S 25.

5.1 31:e Descriocion 5.1.1 The reactor is located on :he ground floor of :he Duncan Annex of the Ilectrical Engineering Su11 ding, Purdue University, vesc Tafayec:a, :ndiana.

5.1.2 The scheat of Nuclear Ingineering concrois approximacely 5000 scuare f asc.

3.1.3 Access :o :his area is rescricted excepc when classes are held here.

3.l.i The reac:or room :emaias locked ac all cises excepc for :he enc:7 or ext: of auchert:ed personnel.

5.l.3 The ?UR-! is housed in a closed :aom designed .o restric: eakage.

3.1.5 The ninimum free volume si the reactor room shall be 13,000 cubic feec.

5.1.7 The vencilacton system is designed to exhause air or other gases f cm the r eac:or room :hrough an exhaust vene ac a sidimum of 50 feet acove the ground.

3.1.3 Openings into the reactor room consisc of :he following:

a. Three personnel doors

b. Two locked cransformer vaule doors

c. Air incake

d. Air exhause

e. Sewer venc 5.2 ruel Assemblies i

j 5.2.1 The fuel assemblies shall be MTR :7pe consisting of aluminum clad places enetched in the U-235 isotope.

5.2.2 A scandard fuel assembly shall consist of up to 10 fuel plates containing uranium enriched in the U-235 isotope.

5.2.3 A conc:ol fuel assembly shall consisc of up to 6 fuel plates containing uranium enriched in the U-235 isotope.

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APPENDIX A FACILITY LICENSE.NO. R-87 TECHNICAL SPECIFICATIONS FOR THE PUiJUE UNIVERSITY REACTOR

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DOCKET NO. 50-182 i

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4 April ,1977 l

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t t TABLE OF CONTENTS 1.0 D EF INI TIO NS , , , , , , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2.0 SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTING. . . . . . . . . . . . . . . . . . 7 2.1 Safety Limit................................................. 7 2.2 Limiting Safety System Setting............................... 3 3.0 LIMITING CONDITIONS FOR OPERATION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 3.1 R e ac t ivi ty L imi t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 3.2 Reactor Safety System.......................,,,.............. 11 3.3 Primary Coolant Conditions................................... 13 3.4 C o n t a in men t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 3.5 Limi tations on Experimen t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 4.0 SURVEILL ANCE REQ UIREMENTS , , , , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 4.1 Reactivity Limits............................................ 18 4.2 Reactor Safety System........................................ 20 4.3 Primary Coolant Conditions................................... 22 4.4 Containment.................................................. 23 4.5 Exp e r imen t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 3 5.0 D ES I GN FE ATURE S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 5 5.1 Site Descrtption............................................. 25 5.2 Fuel Assemblies.............................................. 25 5.3 Fuel Storage................................................ 26 6.0 ADMINI STRATIVE CONTR0 LS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 7 6.1 O r g an i z a t io n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 7 6.2 R e v i ew and Aud i t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1 6.3 Safety Limit Vtolation . .* ................................ 34 6.4 Op er a tin g Pro cedur e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 6.5 Operating Records........................................... 35 6.6 Reporting Requirements...................................... 36 7.0 CHANCES................................................... .... 40 i

t t L.

1 1.0 DEFINITIONS The ,following frequently used terms are to aid in the uniform inter-I pretation of these specifications:

1 1.l' , Reactor Shutdown - That suberitical condition of the reactor where the negative reactivity of the cold, clean core is equal to or greater than the shutdown i

i margin.

! 1.2 Reactor Secured - That overall condition where all of the following conditions I are satisfied:

a) Reactor shutdown Electrical power to the control rod circuits is switched off and the switch i

{ b)

key is in proper custody.

I c) No work in progress involving in-core components, experiments , or installed 4

control rod drives.

I 1.3 True Value - The true value of a parameter is its exact value at any instant.

i i 1.4 Measured Value - The measured value of a parameter is the value as it appears t at the output of a measuring channel.

l 1.5 Measuring Channel - A measuring channel is the combination of sensor, lines, amplifiers, and output devices which are connected for the purpose of measuring the value of a process variable.

j 1.6. Reactor Safety System - The reactor safety system is that combination of j measuring channels and associated circuitry which forms the automatic pro-1

ective system of the reactor, or provides information which requires manual I

l protective action to be initiated.

1.7 Operable - A system or component is operable when it is capable of performing 1

l its intended function in a normal manner.

1.8 Operating - A system or component is operating when it is performing its intended i function in a normal manner.

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1.9 channel Check - A channel check is a qualitative verification of acceptable perfonnance by observation of channel behavior. This verification may in-clude comparison of the channel with other independent channels or methods of measuring the same variable.

1.10 Channel Test - A channel test is the introduction of a simulated signal into a channel to verify that it is operable.

1.11 Channel Calibration - A channel calibration is an adjustment of the channel such that its output responds, within acceptable range and accuracy, to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip.

1.12 Reportable Occurrence - A reportable occurrence is any of the following:

(a) Failure of the reactor protection system or other systems subject to limiting safety system settings to initiate the required protec-tive function by the time a monitored parameter reaches the setpoint specified as the limiting safety system setting in the technical specifications or failure to complete the required protective function.

(b) Operation of the reactor or affected systems when any parameter or operation subject to a limiting condition is less conservative than the limiting condition for operation established in the tech-nical specifications without taking permitted remedial action.

(c) Abnormal degradation discovered in a fission product barrier, i.e. ,

fuel cladding, reactor coolant boundary, or containment.

(d) Reactivity balance anomalies involving:

(1) disagreement between expected and actual critical positions of approximately 0.37. a k/k; (2) exceeding excess reactivity limit; (3) shutdown margin less conservative than specified in technical specifications;

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'(4) unexpected short-term reactivity changes that cause a period of 10 seconds or less; (5) if sub-critical, an unplanned reactivity insertion of more than approximately 0.3% a k/k or any unplanned criticality.

(e) Failure or malfunction of one or more componencs which prevents or could prevent, by itself, the fulfillment of the, functional requirements of system (s) used to cope with accidents analyzed in the HSR.

(f) Fersonnel error or procedural inadequacy which prevents, or could prevent, by itself, the fulfillment of the functional requirements of systems required to cope with accidents analyzed in the HSR (g) Unscheduled Conditions arising from natural or man-made events that, as a direct result of the event require reactor shutdown, operation of safety systems, or other protective measures required by techni-cal specifications. .

(h) Errors discovered in the transient or accident analyses or in the methods used for such analyses as described in the safety analysis report or in the bases for the technical specifications that have or could have permitted reactor operation in a manner less conservative than assumed in the analyses.

(1) Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analyses in the Hazards Summary Report or technical specifications bases; or discovery during plant life of conditions not specifically considered in the Hazards Summary Report or technical specification that require remedial action or correceive measures to prevent the existence or development of an unsafe condition.

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4 1.13 Experiment - An experiment shall mean:

a) any apparatus, device, or material installed in a core or experimental facility, or b) any operation to measure reactor parameters or characteristics, or c) any operation using the reactor as a source of radiation in conjunction with a) above.

1.14 Experimental Facility - Experimental facilities are:

a) those regions specifically designated as locations for experiments or b) systems designed to permit or enhance the passage of a beam of radia-tion to another location.

1 . 15 New Experiment - A new experiment is one whose nuclear characteristics have not been experimentally determined.

1.16 Tried Experiment - A tried experiment is:

a. An experiment previously performed in this facility, or

b. An experiment of approximately the same nuclear characteristics as an experiment previously tried.

1.17 Core Exoeriment - A core experiment is one placed in the core, in the graphite reflector, or within six inches (measured horizontally) of the reflector. This includes any experiment in the pool directly above or below the core.

1.18 Pool Experiment - A pool experiment is one positioned more than six inches (measured horizontally) from the graphite reflector.

1.19 Secured Experiment - Any experiment, experimental facility, or component of an experiment is deemed to be secured, or in a secured position, if it 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 by hydraulic, pneumatic, bonyane, or other forces which are normal to the operating environment of the experi-ment, or by forces which can arise as a result of credible malfunctions.

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1.20 Nonsecured Experiment - Any experiment, experimental facility, or com-ponent of an experiment is considered to be unsecured when it is not secured as defined in 1.19 above.

1.21 Removable Experiment - A removable experiment is any. experiment, experi-mental facility, or component of an experiment, other than a permanently attached appurtenance to the reactor system, which can reasonably be anticipated to be moved one or more times during the life of the reactor.

1.22 Movable Experiment - A movaole experiment is one where it is intended that the entire experiment may be moved in oh near the core or into and out of the core while the reactor is operating.

1.23 Experiment With Movable Parts (Secured or Nonsecured)_- An experiment with movable parts in an experiment that contains parts that are intended to be moved while the reactor is operating.

1.24 Fueled Experiment - A fueled experiment is any experiment which contains uranium 233, uranium 235, plutonium 239, or plutonium 241.

1.25 Static Reactivitv Worth - As used herein, the static reactivity worth of an experiment is the absolute value of the reactivity change which is measur-p able by calibrated control or regulating rod comparison methods between two defined terminal positions or configurations of the experiment. For remo p able experiments, the terminal positions are fully removed from the reactor and fully inserted or installed in the normal functioning or intended position.

1.26 potential Reactivity Worth - The potential reactivity worth of an experi-

' men t is the maximum absolute value of the reactivity change that would t

occur as a result of intended or anticipated changes or credible malfunc-tions that alter experiment position or configuration.

The evaluation must consi e possible trajectories of the experiment in motion relative to the reaccor, its orientation along each trajectory, and circumstances which can cause internal changes such as creating or

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filling of void spaces or motion of mechanical components. For removable experiments, the potential reactivity worth is equal to or greater than the static reactivity worth.

1.27 Explosive Macerial - Explosive material is any solid or liquid which is categorized as a Severe, Dangerous, or Very Dangerous Explosion Hazard in

" Dangerous Properties of Industrial Materials" by N. I. Sax, Third Ed.

(1968), or is given an Identification of Reactivity (Stability) index of 2, 3, or 4 by the National Fire Protection Association in its publica-tion 704-M,1966, " Identification System for Fire 'lazards of Materials,"

also enumerated in the " Handbook of Laboratory Safety" 2nd Ed. (1971) pub-lished by the Chemical Rubber Co.

1.28 acadily Available on Call - Readily available on call shall mean the licensed senior operator shall insure that he is within a reasonable driving time (11 hour1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />) from the reactor building.

1. 29 Reactor Facility - The reactor facility shall consist of that pcrtion of the ground floor of the Duncan Annex of the Electrical Engineering Building occupied by the Department of Nuclear Engineering. This consists of an area of approximately 5,000 square feet.

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2.0 SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTING 2.1 Safeev Limit Apolic ability This specification applies to the steady state power level.

Objective The objective is to define a power level below which it can be pre-dicted with confidence that no damage to the fuel elements will occur.

Specification The true value of the steady state power of the reactor shall not exceed 50kW.

Basis The Purdue University Reactor utilizes fuel of the same type as is used in several similar reactors such as the Lynchburg Pool Reactor, operated by the Babcock and Wilcox Company. These reactors use natural convection cooling and are routinely operated at power levels exceeding 50kW with no apparent damage to the fuel.

The steady state power of 50kW was chosen because calculations indicate that no boiling would occur at this level. With fuel place temperatures at this power level no damage to the fuel elements will occur. The aluminum f alloy cladding does not melt below 1100 F and is expected to maintain its integrity and retain essencially all of the fission fragments at temperatures below 1100 F. For a step input of reactivity equal to the available excess in the core, combined with a postulated failure of the scram mechanisms such that all control rods jam out of the core, it is estimated that the coolant i temperature would rise to less than 130 F. This coolant temperature would I restrict cladding temperatures well below 1100 F thus assuring retention of all fission fragments.

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2.2 Limiting Safety System Setting Apolicability This specification applies to the reactor power level safety system setting for steady state operation.

Objective The objective is to assure that the safety limit is not exceeded.

Soecification The measured value of the power level scram shall be no higher than 1.2 kW.

Basis The LSSS has been chosen to assure that the reactor protective system will be actuated in such a manner as to prevent the safety limit from being exceeded during the most severe expected abnormal condition.

The safety margin between the LSSS and the SL is sufficient to assure that the peak power achieved in a transient, starting at ikR with a 1-second period and terminated by drcpping a control rod, will not exceed 50kW.

The 1-second period corresponds to a reactivity of .006 ok/k, which is the maximum authorized to be loaded into the reactor.

The safety margin that is provided between the LSSS and the SL also allows for instrument uncertainties associated with measuring the above parameter.

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3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Reactivity Limits App licability These specifications apply to the reactivity conditions of the reactor, and the reactivity worths of control rods and experiments.

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

Soecification The reactor shall not be operated unless the following conditions-exist:

a. The shutdown margin, relative to the cold xenon-free condition with the most reactive shim rod fully withdrawn, and the regulating

• rod:

fully withdrawn shall be at least 0.01 ak/k

3. The reactor shall be suberitical by more than 0.03 ak/k during loading e.hanges.

c. No shim-safety rod shall be removed from the core if the shutdown mar-gin is less than 0.01 ak/k with the remaining shim-safety rod fully withdrawn.

d. The reactor shall be shutdown if the maximum positive reactivity of the core and any installed experiment exceeds 0.006 ak/k.

e. The reactivity worth of each experiment shall be limited as follows:

Experiment Maximum Reactivitv Worth Movable .003 ak/k Unsecured .003 ak/k Secured .004 ak/h

f. The total worth of all movable and unsecured experiments shall not exceed 0.003 ak/k.

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g. The total worth of all secured experiments shall not exceed 0.005 ak/k, Sases  !

The shutdown margin required by Specification 3.1.a assures that the reactor can be shut down from any operating condition and will remain shut down even if the control rod of the highest reactivity worth should be in the fully withdrawn position.

Specifications 3.1.b and 3.1.c provide assurance that the core will remain subcritical during loading changes and shim-safety rod maintenance or inspection.

Specification 3.1.d limits the allowable excess reactivity to the value assumed in the Hazards Summary Report. This limit assures that the consequences of reactivity transients will not be increased relative to transients previously reviewed, and assures reactor periods of sufficient I

. length so that the reactor may be shutdown without exceeding the safety limit.

Specification 3.1.e limits the reactivity worth of secured experiments to values of reactivity which, if introduced as a positive step change, are calculated not to cause fuel melting. This specification also limits the reactivity worth of unsecured and movable experiments to values of reactivity which, if introduced as a positive step change, would not cause-the violation of a safety limit. The manipulation of experiments worth up to 0.003 ak/k will result in reactor periods longer than 9 seconds. These periods can be readily compensated for by the action of the safety system without exceeding any safety limits.

A limitation of 0.003 ak/k for the total reactivity worth of all mov-able and unsecured experiments provides assurance that a common failure affecting all such experiments cannot result in an accident of greater

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110 consequences than the maximum credible accident analyzed in the Hazard Summary Report.

Specification 3.1.g along with 3.1.a assures that the reactor is capable of being shut down in the event of a positive reactivity insertion caused by the flooding of an experiment.

3.2 Reactor Safety System Applicability This specification applies to the reactor safety system and other safety-related instrumentation.

Objective The objective is to specify the lowest acceptable level of performance or the minimum number of acceptable components for the reactor safety system and other safety related instrumentation.

Soecification The reactor shall not be made critical unless the following condi-tions are met:

a. The reactor safety channels and safety-related instrumentation are operable in accordance with Tables I and II including the minimum number of channels and the indicated maximum or minimum set points.

b. Both shim-safety rods and the regulating rod shall be operable.

c. The time from the initiation of a scram condition in the scram circuit until the shim-safety rod reaches the rod lower limit switch shall not exceed one second.

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T.AS G 1. SAFT"Y CHANNELS REQUIRED FCR OPERX"!CN Minimm Channe! Numoer Secooine Funecion Raouired L(*} 2 cps 2 cas rod vi:hdrawal Los counc race incerlocic and period 12 sec. period Seeback 7 .iec. period Slow scram

'og N and period ' 1' sec. period Seeback 7 sec. period Slow scram 7 sec. period Tase scram 1 07. power Sicv scram Linear

' 17 07. range Seeback 1007. range Slow scram 1( 1107. power Seeback Saiacf 1207. power Fasc scram Manual Scram (console) 1 Slow scram (hallway) i S1wscram

(#}Noc reo,uired af ter Tcs N-?eriod channel comes on scale.

(b) Rec.uired :o be operable but noc on scale ac scar:up.

TA3 G !! . SAEr"Y-REIt'ED CHANNELS (AREA RADIATION MONITORS)

Minimm Channe1 Numirer Seenciar Funecion Recuired (c) (

Pool :op meni:ce 1 50 m /hr or Slow scram 2:e full power background

'Jacer peacess  ! 7% mR/hr Slow scram Console =cni:or 1 7% mR/hr Slow scram Air samoling Concinucus air i samoler

(*) For periods oc :1:ne, noc :o exceed 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> o: operacion, a radiacion moni or may bc replaced by a gamma sensicive inscrument wnien has ics own

.laem or is kept under periodic visual observation.

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Bases The neutron flux level scrams provide redundant automatic protective actiott to prevent exceeding the safety limit on reactor power, and the period scram conservatively limits the race of rise of the reactor power to periods which are manually controllable without reaching excessive power levels or fuel temperatures.

The rod withdrawal interlock on the Log Count Rate Channel assures that the operator has a measuring channel operating and indicating neutron flux levels during the approach to criticality.

The manual scram button and the " reactor on" keyswitch provide two methods for the reactor operator to manually shut down the reactor if an unsafe or abnormal condition should occur and the automatic reactor protec-tion does not function.

The use of the area radiation monitors (Table II) will assure that areas of the Purdue University Reactor (PUR-1) facility in which a potential high radiation area exists are monitored. These fixed monitors initiate a scram whenever the preset alarm point is exceeded to avoid high radiation conditions.

Specifications 3.2.b and 3.2.c assure that the safety system response will be consistent with the assumptions used in evaluating the reactor's capability to withstand the maximum credible accident.

In specification 3.2.c the rod lower limit switch is positioned to measure, as close as possible, the fully inserted position.

3.3 Primarv Coolant conditions Apolicability This specification applies to the limiting conditions for the primary coolant.

14 Obiective The objective is to assure a compatible environment, adequate shielding, and a continuous coolant path for the reactor core.

Soecification

a. The primary coolant pH shall be maintained at 5.5t 1.0.

b. The primary coolant resistivity shall be maintained at a value greater than 330,000 ohn-em.

c. The primary coolaat shall be maintained at least 13 feet above the core.

Bases Experience at the PUR-l and other facilities has shown that the main-tenance of primary coolant system water quality in the ranges specified in specification 3.3.a and 3.3.b will minimize the amount and severity of corrosion of the aluminum components of the primary coolant system and the fuel element cladding.

The height of water in specification 3.3.c is enough to furnish adequate shielding as well as to guacantee a continuous coolant path.

3.4 Containment Apo licability This specification applies to the integrity of the reactor room.

Objective The objective is to minimize the release of particulate radioactive material from the reactor room.

Soecification

a. During reactor operation the following conditions will be met:

1. The reactor room will be maintained at a negative pressure of 0.05 inch of water or less.

2. All exterior doors in the reactor room shall remain closed except as required for personnel access.

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b. All inlet and exhaust air ducts and the sewer vent shall contain an AEC

1. 1 absolute filter or its equivalent.

c. Dampers in the ventilation system inlet and outlet ducts are capable of being closed.

d. The air conditioner can be shut off.

~3ases The PUR-1 does not rely on a containment building to reduce the levels of airborne radioactive material released to the environment in the event of the design basis accident. However, in the event of such an accident, a significant fraction of the airborne material will be confined within the reactor room, and the specifications stated above will further reduce the release to the environment.

3.5 Limitations on Exoeriments Apolicability This specification applies to experiments installed in the reactor and its experimental facilities.

Obiective The objective is to prevent damage to the reactor or excessive release 1

of radioactive materials in the event of an experiment failure, and to assure the safe operation of the reactor.

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

a. All experiments shall be constructed of material which will be corrosion resistant for the duration of their residence in the pool.

b. All experiments will follow procedures approved by the Committee on Reactor Operations.

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c. Known explosive materials shall not be placed in the reactor pool.

d. Cooling shall be provided to prevent the surface temperature of an experiment from exceeding 100 C.

e. No experiment shall be placed in the reactor or pool that interferes with the safe operation of the reactor.

f. The radioactive material concent, including fission products, of any singly encapsulated experiment should be limited so that the complete release of all gaseous, particulate, or volatile components from the encapsulation will not result in doses in excess of 10% of the equivalent annual doses stated in 10 CFR 20. This dose limit applies to persons occupying (1) unrestricted areas continuously for two hours starting at time of release or (2) restricted areas during the length of time required to evacuate the restricted area.

g. The radioactive material content, including fission products, of any doubly encapsulated experiment or vented experiment should be limited so that the complete release of all gaseous, particulate, or volatile components from the encapsulation or confining boundary of the experi-ment could not result in (1) a dose to any person occupying an un-the restricted area continuously for a period of two hours starting at time of release in excess of 0.5 Rem to the whole body or 1.5 Rem to the thyroid or (2) a dose to any person occupying a restricted area during the length of time required to evacuate the restricted area in excess of 5 Rem to the whole body or 30 Rem to the thyroid.

Bases Specification 3.5.a through 3.5.e are intended to reduce the likelihood of damage to reactor components and/or radioactivity releases resulting from experiment failure and serve as a guide for the review and approval of

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new and untried experiments by the facility personnel and the Committee on Reactor Operations. .

Specification 3.5.f and 3.5.g conform to the criteria set forth in Regulatory Guide 2.2 issued in November,1973.

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4.0 SURVEILLANCE REOUIREMENTS 4.1 Reactivity Limits Applicability This specification applies to the surveillance requirements for reactivity limit s .

Objective The objective is to assure that the reactivity limits of Specification 3.1 are not exceeded.

Soecificacion I

a. The shim-safety rod reactivity worths shall be measured and the shut-down margin calculated at intervals not to exceed 14 months, and when-ever a core configuration is loaded for which shim-saf ety rod worths have not been measured.

b. The shim-safety rods shall be visually inspected at intervals not to exceed 14 months. If the' rod is found to be deteriorated, it shall be replaced with a rod of equivalent or greater worth.

c. The reactivity worth of experimeses placed in the PUR-1 shall be measured during the first startup subsequent to the experimends inser-tion and shall be verified if core configuration changes cause increases in experiment reactivity worth which may cause the experiment worth to exceed the values specified in Specification 3.1.

! Bases Specification 4.1.a will assure that shim-safety rod reactivity worths are not degraded or changed by core manipulations which cause these rods to operate in regions where their effectiveness is reduced.

The boron stainless steel shim-safety rods have been in use at the P.UR-1 since 1962, and over this period of time, no cracks or other evidence of deterioration have been observed. Based on this performance and the

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experience of other facilities using similar shim-safety rods, the specified inspection times are considered adequate to assure that the control rods will not fail.

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4.2 Reactor Safety System Apolicability This specification applies to the surveillance of the reactor safety system.

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

Soecification

'a. A channel test of each of the reactor safety system che.nnels listed in Table III shall be performed prior to each reactor startup following a shutdown in excess of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or if they have been repaired or de-energized.

TABLE III.

SAFETY SYSTEM CHANNELS CHECKED AFTER PROIONGED SHUTDOWN Log Count Rate (startup channel)

Log N-Period Linear Level Safety Channel

b. A channel check of each of the reactor safety system measuring channels in use or on scale shall be performed approximately every four hours when

' the reactor is in operation.

c. A channel calibration of the reactor safety channels shall be performed as follows:

1. An electronic calibration will be performed annually at intervals not to exceed 15 months.

2. A power calibration by foil activation will be performed annually at intervals not to exceed 15 months.

d. The operation of the radiation monitoring equipment shall be verified

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daily during periods when the reactor is in operation. Calibration of these monitors shall be performed semi-annually at intervals not to exceed 7 months,

e. Shim-safety rod drop times will be measured annually, but at intervals 4

not to exceed 15 months. These drop times shall also be measured prior ta operation following maintance which could affect the drop time or cause-movement of the shim-safety rod control assembly.

Bases A test of the safety system channels prior to each startup will assure their operability, and annual calibration will detect any long-term drif t that is not detected by normal intercomparison of channels. The channel check of the neutron flux level channel will assure that changes in core-to-detector geometry or operating conditions will not cause undetected changes in the response of the measuring channels.

Area monitors will sound an alarm when they sense they are not operating correctly. In addition, the operator rodtinely records the readings of these monitors and will be aware of any reading which indicates loss of function.

The area monitoring system employed at the PUR-1 has exhibited very good stability over its lifetime, and semi-annual calibration is considered adequate to correct long-term drif t.

The measured drop times of the shim-safety rods have been consistent for 14 years since the PUR-1 was built. An annual check of this para-meter is considered adequate to detect operation with materially changed drop times. Sinding or rubbing caused by rod misalignment could result from maintenance; therefore, drop times will be checked af ter such maintenance.

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6.3 primarv Coolant System Apolicability This specification applies to the surveillance of the primary coolant system.

Objective The objective is to assure high quality pool water, adequate shielding, and to detect the release of fission products from fuel elements.

Soecification

a. The pH of the primary coolant shall be recorded weekly.

b. The conductivity of the primary coolant shall be recorded weekly.

c. The reactor pool water will be at or above the height of the skimmer trough whenever the reactor is operated.

d. Monthly samples of the primary coolant shall be taken to be analyzed for gross alpha and beta activity.

Bases Weekly surveillance of pool water quality provides assurance that pH and conductivity changes will be detected before significant corrosive damage could occur.

When the reactor pool water is at the skimmer trough level, adequate shielding of more than 13 feet of water is assured.

Analysis of the reactor water for gross alpha and beta activity assures against undetected leaking fuel assemblies.

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4.4 Containment Apolicability This specification applies to the surveillance requirements for maintaining the integrity of the reactor room and fuel clad.

Objective The objective is to assure that the integrity of the fuel containment is maintained.

Soecification

a. The negative pressure of the reactor room will be recorded weekly.

b. Operation of the inlet and outlet dampers shall be checked semi-annually at intervals not to exceed 7h months. l

c. Operation of the air conditioner shall be checked semianr tally at intervals not to exceed 7h months. l

d. Representative fuel plates shall be inspected at intervals not to exceed 15 months. l Bases Specification a, b , and c check the integrity of the reactor room, and d the integrity of the fuel clad. Based upon past experience these intervals have been shown to be adequate for insuring the operation of the systems affecting the integrity of the reactor room and fuel clad.

4.5 Experiments Applicability This specification applies to the surveillance of limitations on experiments.

Ob j ec t ive The objective is to assure that the radioactive material content of experiments does not exceed the limits of parts f and g of specification 3.5

24 Specification

a. Calculations shall be made on samples of known composition to assure that the limits of specification 3.5.f and 3.5.g are not exceeded.

b. The mass of samples of unknown composition shall not exceed 10 grams.

Bases Past experience has shown that calculating the expected activity of a sample of known composition, or limiting the mass of a sample of unknown composition has been suf ficient to prevent exceeding prescribed limits.

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5.0 DESIGN FEATU?.ES 5.1 Site Descriotion 5.1.1 The reactor is located on the ground floor of the Duncan Annex of the Electrical Engineering Building, _ Purdue University, West Lafayette, Indiana.

5.1.2 The School of Nuclear Engineering controls approximately 5000 square feet.

5.1.3 Access to this area is restricted except when classes are held here.

5.1.1 The reactor room remains locked at all times except for the entry or exit of authorized personnel.

! 5.1.5 The PUR-1 is housed in a closed room designed to restrict leakage.

5.1.6 The minimum free volume of the reactor room shall be 13 ,000 cubic feet.

5.1.7 The ventilation system is designed to exhaust air or other gases from the r eactor room through an exhaust vent at a minimum of 50 feet above the ground.

5.1.8 Openings into the reactor room consist of the following:

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a. Three personnel doors

b. Two locked transformer vault doors

c. Air intake

d. Air exhaust

e. Sewer vent 5.2 Fuel Assemblies 5.2.1 The fuel assemblies shall be MIR type consisting of aluminum clad places enriched to approximately 937. in the U-235 isotope.

5.2.2 A standard fuel assembly shall consist of 10 fuel plates containing a maximum of 165 grams of U-235.

5.2.3 A control fuel assembly shall consist of 6 fuel plates containing a maximum of 99 grams of U-235.

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5.2.4 Partially loaded fuel assemblies in which some of the fuel plates are replaced by aluminum places containing no uranium may be used.

5.3 Fuel Storage 5.3.1 All reactor fuel assemblies shall be stored in a geometric array where k,gg is less than 0.8 for all conditions of moderation and reflection.

5.3.2 Irradiated fuel assemblies and fueled devices shall be stored in an array which will permit sufficient natural convection cooling by water or air such that the temperature of the fuel assemblies or fueled devices will not exceed 100 C.

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6. ADMINIS'IltATIVE C0ffrROLS 6.1 Ornanisation 6.1.1 The reactor facility shall be an integral part of the School of Nuclear Engineering of the Schools of Engineering at Purdue University as shown in Figure 6.1.

6.1.2 The Reactor Supervisor shall be responsible for the safe opera-tion of the PUR-1. He shall be responsible for assuring that all operations are conducted in a safe manner and within the limits prescribed by the facility license, including the technical specifications and other applicable regulations.

6 .1.3 In all matters pertaining to the operation of the reactor and the administrative aspects of these technical specifications, the Reactor Supervisor shall report to and be directly responsible to the Head of the School of Nuclear Engineering. In all matters pertaining to radiation safety he shall be responsible to the Radiological Control Committee.

6.1.4 Minimum Qualifications of Reactor Personnel - The minimum qualif-ications are as follows:

a. Reactor Supervisor - At the time of appointment to the active position, the reactor supervisor shall have a minimum of five years of nuclear experience. He shall have a bac-calaureate degree or equivalent experience in an engineering l or other scientific field. The degree may fulfill four years of experience on a one-for-one basis. The reactor supervisor shall possess a valid Senior Operator License,

b. Licensed Senior Operator - At the time of appointment to the active po sition, a senior operator shall have a minimum of a high school diploma or equivalent and should have four years of nuclear experience. A maximum of two years of experience l

PRESIDENT .

PuRDUE UNIVERSITY I!_________________

DEAN RADIOLOGICAL CONTROL SCHOOLS OF ENGINEERING j COMMITTEE I

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llEAD i RADIOLOGICAL I

8 CONTROL 0FFICER SC1100L OF NUCLEAR ENGlHEERING i  !

8 COMMITTEE ON REACTOR OPERATIONS l

! i f l REACTOR _ _ _ ._ _ _ _ _ _ _ _i

> 4

. SUPERVISOR up l REACTOR OPERATIONS i

i Figure 6.1 Primarily Adniinistration

- - - - - Primarily Safety

29.

may be fulfilled by related academic or technical training on a one-for-one time basis,

c. Licensed Operator - At the time of appointment to the active position, an operator shall have a high school diploma or equivalent. l

d. Operator Trainee - An operator trainee shall have all the qualifications to become a licensed operator except for possessing an operator's license.

6.1.5 A Radiological Control Officer who is organizationally independent of the PUR-1 operations group shall advise the Reactor Supervisor in matters concerning radiological safety. Minimum qualifications for the Radiological Control Of ficer (RCO) is a bachelor's degree or the equivalent in a science or engineering subject, including some formal training in radiation protection. The RCO should have at least five years of professional experience in applied radiation pro-tection. A master's degree may be considered equivalent to one year of professional experience, and a doctor's degree equivalent to two years of professional experience where course work re-laced to radiation protection is involved. At least three years of this professional experience should be in applied radiation protection work in a nuclear f acility dealing with radiological problems.

to 6.1.6 A licensed operator (LO) or licensed senior operator (LS0) pursuant 10 CFR 55 shall be present at the controls unless the reactor is shut down as defined in these specifications. During training opera-tions an unlicensed operator may operate the controls but only un-der the direct supervision of an LO or an LSO.

30.

6.1.7 An LSO shall be present or readily available on call at any time that the reacter is operating.

6.1.3 The identity of, and method for rapidly contacting, the licensed senior operator on duty shall be known to the reactor operator at any time that the reactor is operating.

6.1.9 The presence of an LSO shall be required at the reactor f acility -

during recovery from unplanned or unscheduled shutdowns except in instances which result from the following:

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a. A verified electrical power failure or interruption exclusive 6

of internal power supply failures or interruption of the reactor instrumentation, control, and saf ety systems ;

b. Accidental manipulation of equipment in a manner which does not affect the safety of the reactor;

c. A verified practice or evacuation of the building initiated by persons exclusive of the reactor.

The LSO shall be notified of the shutdown and shall determine its cause. If due to one of the enumerated reasons above, he shall decide if his presence. is necessary for a subsequent start up.

6.1.10 The presence of an LSO at the reactor facility is unnecessary for the initial daily start up , provided the core remains unchanged from the previous run.

6.1.11 The minimum crew for operating the reactor shall consist of 2 (two) persons , one of which must be a licensed member of the PUR-1 opera-tions group. The unlicensed crew member must be instructed as to how to shut down the reactor in the event of an emergency.

6. 1. 12 Durtng fuel changes and movement at large bulk experiments , an LSO will be present in the reactor room.

6.L.13 The Reactor Supervisor or.his designated alternate shall be respon-sible for the facility retraining and replacement training program.

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6.1 Review and Audit 6.2.1 A Committee on Reactor Operations (CORO) shall report to the Radiological Control Committee on matters of Radiation Safety and the Head, School of Nuclear Engineering on matters of administration and safety.

CORO will advise the Reactor Supervisor on those areas of responsibility specified in Sections 6.2.5 and 6.2.6.

The minimum qualifications for persons on the CORO shall be five years of professional work experience in the discipline or specific field he represents. A baccalaureate degree may fulfill four years of experience. t 6.2.2 The CORO shall have at least 7 (seven) members of whom no more than a minority shall be directly concerned with the administration or direct use of the reactor. These members shall include the following:

a. The Chairman, a responsible, senior technical person, knowledge-able in the field of reactor technology, who does not have line responsibility for day-to-day operation of the reactor.

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b. A senior radiological control officer.

c. The Purdue University Superintendent of Safety and Security d The Reactor Supervisor.

e. Three senior scientific staff members.

6.2.3 The CORO or a subcommittee thereof shall meet quarterly at intervals l not to exceed 4 months. The CORO shall meet semiannually at intar-vals not to exceed 7h months.

l j 6.2.4 A quorum shall consist of not less than a majority of the full Committee and shall include the chairman or his designated alternate.

6.2.5 The CORO shall review:

- a. Safety evaluations for 1) changes to procedures, equipment or sysr. ems and 2) tests or experiments, conducted without NRC l

approval under the provision of Section 50.59, 10 CFR, to

32.

verify that such actions did not constitute an unreviewed safety question.

b. Proposed changes to procedures, equipment or systems that change the original intent or use, and are non-conservative, or those that involve an unreviewed safety question as defined in Section 50.59, 10 CFR.

c. Proposed tests or experiments which are significantly different from previous approved tests or experiments, or those that involve an unreviewed safety question as defined in Section 50.59, 10 CFR.

d. Proposed changes in Technical Specifications or licenses.

e. Violations of applicable statutes, codes, regulations, orders, Technical Specification, license requirements, or of internal procedures or instructions having nuclear safety significance.

f. Significant operating abnormalities or deviations from normal and expected performance of facility equipment that affect nuclear safety.

g. Events which have been reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the NRC in writing.

h. Audit reports.

6.2.6 AUDITS Audits at facility activities shall be performed under the cognizance of the CORO but in no case by the personnel responsible for the item audiced. Individual audits may be performed by one These individual who need not be an identified CORO member.

audits shall examine the operating records and encompass:

a.

The conformance of facility operation to the Technical Specifica-tions and applicable license conditic

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b. The performance training and qualifications of the licensed facility staff, to be done annually at intervals not to exceed 15 months.

c. The results of all actions taken to correct deficiencies oc-curring in f acility equipment, structures, systems or method of operation that affect nuclear safety, to be done annually at intervals not to exceed 15 months,

d. The Facility Emergency Plan and implementing procedures, to be done biennially at Latervals not to exceed 2h years.

e. The Facility Security Plan and implementing procedures, to be done biennially at intervals not to exceed 2h years,

f. Any other area of facility operation considered appropriate by the CORO or the Reactor Supervisor.

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6.2.7 RECORDS Records of CORO activities shall be prepared and distributed as indicated below.

a. Minutes of each CORO meeting shall be prepared and forwarded to the Reactor Supervisor within 30 days following each meeting.

b. Reports of reviews encompassed by section 6.2.5 e, f, and g above, shall be prepared.and forwarded to the Reactor Supervisor within 30 days following completion of the review.

c. Audit reports encompassed by Sec tion 6.2.6 above, shall be forwarded to the CORO Chairman and to the management respon-I sible for the areas audited within 30 days af ter completion of the audit.

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34 6.3 Safety Limit Violation The following actions shall be taken in the event the Safety Limit is violated:

a. The reactor will be shut down immediately and reactor operation will not be resumed without authorization by the Commission.

b. The Safety Limit Violation shall be reported to the Director of the appropriate NRC Regional Office of Inspection and Enforce-ment (or his designate), the Reactor Supervisor and ed the CORO not later than the next work day.

c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the CORO. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.

d. The Safety Limit Violation Report shall be submitted to the Commission, the CORO and the Reactor Supervisor within 14 days of the violation.

6 .4 Operating Procedures Written procedures , including applicable check lists reviewed and approved by the CORO, shall be in effect and followed for the following operations:

6.4.1 Startup, operation, and shutdown of the reactor.

6.4.2 Installation and removal of fuel elements and control rods.

6.4.3 Actions to be taken to correct specific and foreseen potential mal-functions of systems or components, including responses to alarms and abnormal reactivity changes.

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l 6.4.4 Emergency conditions involving potential or actual release of radio-  ;

activity, including provisions for evacuation, re-entry, recovery, and medical support.

6.4.5 Maintenance procedures which could have an effect on reactor safety.

6.4.6 Experiment installation, operation, and removal.

6.4.7 Implementation of the Security Plan and Emergency Plan.

Non-routine operations which require the sequential performance of a series of sub-casks shall be carried out with the written procedure at the console. To assure adherence to the documentation of the procedure, each step will be entered in the log book as it is completed.

Substantive changes to the above procedures shall be made only with the approval of the CORO. The Reactor Supervisor may make changes to procedures which do not change the intent of the original procedure.

All such changes to the procedures shall be documented and subsequently reviewed by CORD.

6.5 Operating Records 6.3.1 The following records and logs shall be prepared and retained for at least five years: .

a. Normal facility operation and maintenance.

- b. Reportable occurrences.

c. Tests, checks, and measurements documenting compliance with surveillance requirements.

i d. Records of experiments performed.

e. Records of radioactive shipments.

f. Changes of opera..ng procedures.

g. Facility radiation and contamination surveys.

36.

6.5.2 The following records and logs shall be prepared and retained for the life of the facility:

a. Gaseous and liquid waste released to the environs.

b. Offsite environmental monitoring surveys.

c. Radiation exposures for all PUR-1 personnel.

d. Fuel inventories and transfers.

e. Updated, corrected, and as-b'uilt facility drawings.

f. Minutes of the CORO meetings.

g. Records of transient or operational cycles for those components designed for a limited number of transients or cycles.

h. Records of training and qualification for members of the facility staff.

i. Records of reviews performed for changes made to procedures to or equipment or reviews of tests and experiments pursuant 10 CFR 50.59, 6.6 Reoorting Requirements The following information shall be submitted to the USNRC in addi-tion to the reports required by Title 10, Code of Federal Regulations.

6.6.1 Annual Operating Reports--a report covering the previous year shall be submitted to the Director of the appropriate NRC Regional office by March 31 of each year. It shall include the following:

a. Changes in plant design and operation

1. changes in facility design

2. performance characteristics (e.g. equipe ... and fuel performance).

37.

3. changes in operating procedures which relate to the safety of facility operations 4 results of surveillance tests and inspections required by these technical specifications

5. a brief summary of those changes, tests, and experiments which reqaired authorization from the Conmission pursuant to 10 CFR 50.59(a)

b. Power Generation--A tabulation of the thermal output of the facility during the reporting period.

c. Shutdowns--A listing of unscheduled shutdowns which have occurred during the reporting period, tabulated according to cause, and a brief discussion of the preventive actions taken to prevent recurrence.

d. Maintenance--A discussion of corrective maintenance (excluding preventative maintenance) performed during the reporting period on safety-related systems and components.

e. Changes , Tests , and Experiments-- A brief . description and a sununary of the safety evaluation for those changes, tests, and experiments which were carried out without prior Coaniasion approval, pursuant to the requirements of 10 CFR Part 50.5 9 (b ) .

f. Radioactive Effluent Releases--A summary of the nature, amount, and maximum concentrations of radioactive effluents released or discharged to the environs beyond the effective control of the licensee as measured at or prior to the point of such release or discharge.

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• 3 Occupational Personnel Radiation Exposure--A sammary of radiation exposures greater than 500 sRem (50 = Rem f or pe r-sons under 13 years of age) received during the reporting period by facility personnel (faculty, students, or experi-menters).

6 .6 .2 Non-Routine Reports

a. Reportable Occurrence Reports In the event of a reportable occurrence (defined in 1.0) notification shall-be =ade within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and telegraph to the Director of the Regional 2essistory Operations Office (copy to the Director of Licensing), foi-lowed by a written report within 10 days to the Director of the Regional Regulatory Operations Of fice. The written report on these reportable occurrences, and to the extent f possible the preibninary telephone and telegraph notifica-tion shall: (a) describe, analyze, and ev21uste safety implica t ions , (b) outline the =easures tahan to assure that the cause of the condition is deter =ined. (c) indicate the corrective action (including any changes made to the pro-cedures and to the quality assurance program) esken to i

prevent repetition of the occurrence and of similar occurrences involving similar components or syste=s, and (d) l evaluate the sa f ety implications o f the incident in light of I

l the cesalat iv e experience obtained f rom the record of previ-ons failures and malfunctions of similar syste=s and com-ponents.

• Telegraph notifiestion =sy be sent on the ne:: working day in the event of a reportsble occurence durin; a weehend or holidsy period.

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b. Unusual Events A written report shall be forwarded within 30 days to the Director of the Regional Regulatory Operations Office in the event of:

1. Discovery of any substantial errors in the transient or a

accident analyses or in the methods used for such analyses, as described in the Hazards Summary Report on the bases for the Technical Specifications.

2. Discovery of any substantial variance from performance specifications contained in the Technical Specifications or in the Hazards Summary Report.

3. Discovery of any condition involving a possible single f ailure which, for a system designed against assumed single failures, could result in a loss of the capability of the system to perform its safety function.

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