ML20062F334

Amend 3 to Oper Lic R-87.Changing Tech Specs to Provide Concise & Definitive Statement of Limits of Reactor Oper Parameters & Surveillance Measures Req to Assure That Limits Are Not Exceeded

ML20062F334
Person / Time
Site:	Purdue University
Issue date:	11/28/1978
From:	Reid R Office of Nuclear Reactor Regulation
To:
Shared Package
ML20062F331	List: ML20062F328 ML20062F334 ML20062F352
References
R-087-A-003, R-87-A-3, NUDOCS 7812180079
Download: ML20062F334 (44)

Text

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PURDUE UNIVERSITY DOCKET NO. 50-182 AMENDED FACILITY OPERATING LICENSE Amendment No. 3 License No. R-87

1. The Nuclear Regulatory Comission (the Comission) has found that:

A. The application for amendment by Purdue U61versity (the licensee) dated May 30,1975, as supplemented March 15, 1977, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; 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 Conrnission's regulations; D. The licensee is technically and financially qualified to engage in the activities authorized by this operating license in accordance with regulations of the Comission; E. The licensee is a nonprofit educational institution and will use the facility for the conduct of educational activities, and has satisfied the applicable provisions of 10 CFR Part 140, " Financial Protection Requirements and Indemnity Agreements", of the Comission's regulations; F. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; 7 812180 07 $

G. 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; H. The receipt, possession, and use of byproduct and special nuclear materials as authorized by this license will be in

. accordance with the Commission's regulations in 10 CFR Parts 30 and 70, including 10 CFR Sections 30.33, 70.23, and 70.31; and I. Publication of notice of this amendment is not required since it does not involve a significant hazards consideration nor amendmcnt of a license of the type described in 10 CFR t' Section 2.106 (a)(2).

2. Facility Operating License No. R-87 issued to Purdue University is hereby amended to read as follows:

A. The license applies to the pool-type nuclear reactor (the facility) owned by Purdue University. The facility is located on the licensee's campus in West Lafayette, Indiana, and is described in the licensee's application dated January 25, 1961 and subsequent amendments and supplements thereto, including the application for amendment dated May 30, 1975 and supplement thereto dated March 15, 1977.

B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses Purdue University:

(1) Pursuant to Section 104c of the Act and 10 CFR Part 50,

" Licensing of Production and Utilization Facilities",

to possess, use, and operate the facility at the designated location in West Lafayette, Indiana, in accordance with the procedures and limitations set forth in this license; (2) Pursuant to the Act and 10 CFR Part 70, "Special Nuclear Material", to receive, possess, and use up to 3 kilograms of uranium-235 contained in uranium enriched in the isotope uranium-235 and up to 80.0 grams of plutonium contained in encapsulated plutonium-beryllium sources, both in connection with operation of the facility; and (3) Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This license shall be deemed to contain and is subject to the conditions specified in the following Comission regulations in 10 CFR Chapter I: Part 20. Section 30.34 of Part 30, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximun Power Level The licensee is authorized to operate the facility at steady state power levels not in excess of one kilowatt (thermal ) .

(2) Technical Specifications The Technical Specifications contained in Appendix A attached hereto are hereby incorporated in this license.

The licensee shall operate the facility in accordance with the Technical Specifications.

D. The licensee shall maintain in effect and fully implement all provisions of the NRC approved physical security plan, including amendments and changes made pursuant to the authority of 10 CFR 50.54(p). The approved security plan consists of the physical security plan (proprietary) submitted by the licensee's letter dated January 2,1974, as supplemented by the licensee's letter dated August 22,1974.

E. This amended license is effective as of the date of issuance and shall expire at midnight, August 7, 1986.

FOR THE NUCLEAR REGULATORY COMMISSION 0

WOLcCed Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors i

Attachment:

Appendix A - Techaical Specifications Date of Issuance: November 28, 1978 I

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APPENDIX A FACILITY OPERATING LICENSE NO. R-87 TECHNICAL SPECIFICATIONS FOR THE PURDUE UNIVERSITY REACTOR DOCKET No. 50-182 Amendment No. 3 Dated: November 28, 1978

TABLE OF CONTENTS 1.0 D EF INI TIO N S, , , , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2.O SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTING. . . . . . . . . . . . . . . . . . 7 2.1 Safety Limit................................................. 7 2.2 Limi t in g S a f e ty Sy s t em S e t ting. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3.0 LIMITING CONDITIONS FOR OPERATION................................. 9 3.1 Reactivity Limits............................................ 9 3.2 Reactor Safety System........................................ 11 3.3 Primary Coolant conditions................................... 13 3.4 Containment.................................................. 1-4 3.5 Limit ations on Experimen ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 4.0 SURVEILLANCE REQUIREMENTS , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 4.1 Reactivity Limits............................................ 18 4.2 Reactor Safety System........................................ 20 4.3 Primary Coolant Conditions................................... 22

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4.4 Containment.................................................. 23 4.5 Ex p e r imen t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...................23 5.0 DESIGN FEATURES................................................... 25 5.1 Site Des:stption............................................. 25 5.2 Fu e l As s emb l i e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 5 5.3 Fuel Storage................................................ 26 6.0 ADMINI STR ATIVE CONTR0 LS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 7 6.1 Organization ............... ...............................27 6.2 R ev i ew and Aud i t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 6.3 Safety Limit Violation..... ........ . ....... ...........34 6.4 Operating Procedures........................................ 34 6.5 Op er ating R ec ord s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 5 6.6 Reporting Requirements...................................... 36 1..

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

1.1 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 margin.

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

a) Reactor shutdown b) Electrical power to the control rod circuits is switched off and the switch key is in proper custody.

c) No work in progress involving in-core components, experiments, or installed control rod drives.

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

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

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

1.6 Reactor Safety System - The reactor safety system is that combination of measuring channels and associated circuitry which forms the automatic pro-tective system of the reactor, or provides information which requires manual protactive action to be initiated.

1.7 Operable - A system or component is operable when it is capable of perfer=ing its intended function in a normal manner.

1.8 Ooeratine - A system or component is operating when it is performing its intended function in a normal canner.

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1.9 Channel Check - A channel check is a qualitative verification of acceptable performance 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 Repoftable 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-cive 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 0.3% a k/k; (2) exceeding excess reactivity limit; (3) shutdown margin less conservative than specified in technical specifications; 9

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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 0.5% a k/k or any unplanned criticality.

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

(f) Personnel 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 Hazards Summary Repret 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 cparation

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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 specifications that require remedial action or corrective measures to prevent the existence or i development of an unsafe condition.

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I 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 epecifically 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

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(measured horizwatally) 'from the graphite reflector.

l.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, benyant, 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 movable experiment is ons where it is intended that the entire experiment may be moved in or 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 is 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 Reactivity Worth - As used herein, the static reactivity worth of an experiment is the absolute value of the reactivity change which is measur-able by calibrated control or regulating rod comparison methods between two defined terminal positions or configurations of the experiment. For remov-t able experiments, the ter=inal positions are fully removed from the reactor and fully inserted or installed in the normal functioning or intended position.

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1.26 Potential Reactivity Worth - The potential reactivity worth of an experi-ment is the maximum absolute value of the reactivity change that would occur as a result of intended or anticipated changes or credible malfune-tions that alter experiment position or configuration.

The evaluation must consider possible trajectories of the experiment in motion relative to the reactor, its orientation along each trajectory, and circu= stances which can cause internal changes such as creating or 4

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fil;.ing 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 Material - 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 'i.1. 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 ::azards of Materials,"

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

1.28 neadily Available on Call - Readily available on call shall mean the licensed senior operator shall insure that he is within a reasonable driving time ( hour) from the reactor building.

1. 2) 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 SAFETI SYSTEM SETTING i 1

1 2.1 Safetv Limit )

l Apolicability l l

l This specification applies to the steady state power level.

Obiective 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 plate temperatures at this power level no damage to the fuel elements will occur. The aluminum alloy cladding will not melt below 1100 F and is expected to maintain its integrity and retain essentially 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 temperature would rise to less than 130 F. This coolant temperature could 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.

Obiective 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 oeen 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 ikW with a 1-second period and terminated by dropping 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 i. trument uncertainties associated with measuring the above parameter.

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3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Reactivity Limits Apolicability 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 limit 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.wirb the most reactive shim rod fully withdrawn, and the regulating r6d fully withdrawn shall be at least 0.01 Ak/k

b. The reactor shall be suberitical by more than 0.03 ok/k during loading changes.

c. No shim-safety rod shall be removed from the core if the shutdown mar-gin is less than 0.01 ok/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.

j e. The reactivity worth of each experiment shall be limited as l

follows:

Exoeriment Maximum Reactivitv Worth Movable .003 ak/k i Unsecured .003 ak/k i

Secured .004 ok/k t

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.

Bases 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 suberitical 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. TPis limit assures that the consequences of reactivity transients will not be increased relative to transients previously reviewed, and assures reactor periods of sufficient 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 lenger 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

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consequences than the maximum credible accident analyzed in the Hazards Sumary Report.

Specification 3.1.g along with 3.1.a assures that the reactor is cap-able 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.

Obiective 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.

Specification 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 600 milliseconds.

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TAB LE I . SAFETY CHANNELS REQUIRED FOR OPERATION Minimum Channel Number Setooint Function Recuired Log count rate 1(*) 2 cps 2 eps rod withdrawal and period interlock 12 sec. period Setback 7 sec. period Slow scram

) 12 sec. period Setback Log N and period 1 7 sec. period Slow scram 7 sec. period Fast scram 120% power Slow scram Linear 1 1107. range Setback 1207. range Slow scram Safety 1( } 110% power Setback 120",. power Fast scram Manual Scram (console) 1 Slow scram (hallway) 1 Slow scram

(*)Not required after Log N-Period channel comes on scale.

(b) Required to be operable but not on scale at startup.

TA3LE II. SAFETY-RElATED CHANNELS (AREA RADIATION MONITORS)

Minimum Channel Nu=ber Setooint Function Recuired (a)

Pool top monitor 1 50 mR/hr or Slow scram 2x full power background Water process 1 7 mR/hr Slow scram

! Console monitor 1 7% cR/hr Slow scram Continuous air 1 Air sampling sampler (a)For periods of time, not to exceed 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of operation, a radiation monitor may be replaced by a gamma sensitive instrument which has its own alarm or is kept under periodic visual observation.

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Bases The neutron flux level scrams provide redundant automatic protective action to prevent exceeding the safety limit on reactor power, and the period scram conservatively limits the rate 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.

I 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.

I 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 1

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 insert _ed position.

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

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. 3 14 Objective 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.St 1.0.

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

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

- Bases Experience at the PUR-1 and other facilities has shown that the main-tena-e.e 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 guarantee a continuous coolant path.

3.4 Containment Aeolicability This specification applies to the integrity of the reactor rocm.

Obiective The objective is to minimi:e 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 cnd 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.

Bases 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 r 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 vill further reduce the release to the environment.

3.5 Limitations on Experimenta Applicability 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 of radioactive materials in the event of an experiment failure, and to assure the safe cperation 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 corrosf.cn resistant for the duration of their residence in the pool.

b. All experiments will follow procedures approved by the Co==1ttee on Reacter 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.

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

f. The radioactive material content, 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-restricted area continuously for a period of two hours starting at the 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 Specifications 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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i new and untried experiments by the facility personnel and the Comnittee on Reactor Operations.

Specifications 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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18, 4.0 SURVEILLANCE REOUIREMENTS 4.1 Reactivity Limits Acolicability This specification applies to the surveillance requirements for reactivity limits.

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

Soecification

( a. The shim-safety rod reactivity worths shall be measured and the shut-dawn margin calculated at intervals not to exceed 14 months, and when-ever a core configuration is loaded for which shim-safety 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 experiments placed in the PUR-1 shall be measured during the first startup subsequent to the experiment's 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 PUR-1 since 1962, and over this period of time, no cracks or other evidence of j deterioration have been observed. Based on this performance and the l

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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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20, 4.2 Reactor Safety Svstem Aeolicability This specification applies to the surveillance of the reactor safety system.

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

Specification

a. A channel test of each of the reactor safety system channels 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 PROIDNGED 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:

'l. An electronic calibration will be performed at intervals not to exceed 14 months.

j 2. A power calibratien by foil activhtion will bs performed at inter-vals not to exceed 14 months.

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

21.

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

e. Shim-safety rod drop times will be measured annually, but at intervals not to exceed 14 months. These drop times shall also be measured prior to operation following maintenance 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 routinely records the readings of these monitors and will be aware of any reading which indicates lost 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 drift.

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

22.

4.3 Primarv Coolant System Aeolicability This specification applies to the surveillance of the primary coolant system.

Obiective -

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.

l' 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 corresive 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.

23.

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

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

Specification

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

( b. Operation of the inlet and outlet dampers shall be checked at inter-vals not to exceed 7 months.

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

d. Representative fuel plates shall be inspected at intervals not to exceed 14 months.

Bases Specifications a, b, end 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.

Objective ,

l The objective is te Kscure that the radioactive material content of

experiments does not exceed the limits of parts f and g of specification 3.5.

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24.

Specification

a. Calculations shall be made on samples of known composition to assure that the limits of specifications 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 sufficient to prevent exceeding prescribed limits.

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25.

5.0 DESIGN FEATURES 5.1 Site Description 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 Department of Nuclear Engineering controls approximately 5000 square feet.

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

5.1.4 The reactor room remains locked at all times except for the entry

," or exit of authorized personnel.

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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 15,000 cubic feet.

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

5.1.8 Openings into the reactor room consist of the following:

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 MTR type consisting of aluminum clad plates enriched to approximately 937. in the U-235 isotope.

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

5.2.3 .i centrcl fuel assembly shall consist of 6 fuel plates cencaining a maximum of 99 grams of U-235.

l 26.

5.2.4 Partially loaded fuel assemblies in which some of the fuel plates are replaced by aluminum plates 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 modaration and refleetion.

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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6.0 ADMINISTRATIVE CONTROLS 6.1 Organization 6.1.1 The reactor f acility shall be an integral part of the School of Nuclear Engineering of the Schools of Engineering at Purdue University as shown in Tigure 6.1.

6.1.2 The Reactor Supervisor shall be responsible for the safe operation of the PUR-1. He shall be responsible for assuring that all opera-tions 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 qualifica-tions are as follows:

a. Reactor Supervisor - At the time of appoint =ent to the active position, the reactor supervisor shall have a minimum of five years of nuclear experience. He shall have a baccalaureate degree or equivalent experience in an engineering 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 pcsition, a senior opera:or shall have a =inimum of a high school diploma or equivalent and should have four years of nuclear experience. A maximum of two years of experience

PRESIDENT PURDUE UNIVERSITY I ! _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _9 DEAN RADIOLOGICAL CONTROL l

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- - - - - 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 diplona or equivalent.

d. Operator Trainee - An operator trainee shall have all the qualifications to become a licensed operator except for possessing an op'erator'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 qualification for the Radiological Control Officer consists of certification by the American Boar d of Health Physics or its equivalent.

6.1.6 A licensed operator (LO) or licensed senior operator (LS0) pursuant to 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.

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

6.1.8 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 l

instances which result from the following:

a. A verified electrical power failure or interruption exclusive of internal power supply failures or interruption of the

[

reactor instrumentation, control, and safety systems;

I 30.

b. Accidencal manipulation of equipment in a manner which coes 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 i 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 menber 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 During fuel changes and movement of large-bulk experiments, an LSO will be present in the reactor room.

6.1.13 The Reactot- Supervisor or his designated alternate shall be respon-

- sible for the facility retraining and replacerent training program.

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31.

6.2 Review and Audit 6.2.1 A Committee on Reactor Operations (CORO) shall report to the Radiological .

Ccatrol 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.

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,

b. A senior radiological control officer. .

l c. The Purdue University Superintendent of Safety and Security,

d. The Reactor Supervisor.

c. Three senior scientific staff members.

6.2.3 The CORO or a subcommittee thereof shall meet at least once per calendar quarter. The CORO shall meet at least semiannually.

5.2.4 A quorum shall consist of not less than a majority of the full Committee and shall include the chairman or his designated alternace.

6.2.5 The CORO shall review:

a. Safety evaluations for 1) changes to procedures, equipment er syatens and 2) tests or experiments, conducted without NR2 aoproval under the provision of Section 50.59, 10 CTR, 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.

aor those that involve an unreviewed safety question as define.!

in Section 50.59, 10 CFR.

- c. Proposed tests or experiments which 'are significantly dif ferent

< 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, orde:s, Technical Specifications, license requirements, or of internal I

' 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 .T!C in writing.

h. Audit reports.

6.2.6 AUDITS Audits of facility activities shall be performed under the cognizance of the CORO but in no case by the personnel responsible for the item audited. Individual audits may be performed by one individual who need not be an identified CORO member. These audits shall examine the operating records and encompass:

a. The conformance of facility operation to tha ~echnical Specifica-tiens and applicable license conditions, at least enee per 12 I

months.

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33.

b. The performance training and qualifications of the entire facility staff, at least once per 12 months.

c. The results of all actions taken to correct deficiencies oc-curring in facility equipment, structures, systems or method of operation that affect nuclear safety, at least once per 6 to 12 months.

d. The Facility Emergency Plan and implementing procedures, at least once per 24 months.

( e. The Facility Security Plan and implementing procedures, at least once per 24 months.

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

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 sections 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 Section 6.2.6 above, shall be forwarded to the CORO Chairman and to the management respon-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 regional USNRC Office of Inspection and Enforcement, the Reactor Supervisor and to 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 C00.0 and the Reactor Supervisor within 14 days of the violation.

6.4 coerating 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.

35.

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.

Operations which require the sequential performance of a series i

of sub-tasks 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 the CORO.

6.5 Operating Records 6.5.1 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.

d. Records of experiments performed.

e. Records of radioactive shipments.

f. Changes to operating 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-built facility drawings.

f. Minutes of the CORO meetings.

g. Records of transient or operational cycles for those g

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 or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.

6.6 Reporting 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. equipment 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 required authorization from the Commission pursuant to 10 CFR 50.59(a) - -

b. Power Generation--A tabulation of the thermal output of the facility 1

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 summary of the safety evaluation for those changes, tests, and experiments which were carried out without prior Coani.ision approval, pursuant to the requirements of 10 CFR Part 50.59(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 co. trol of the licensee as measured at or prior to the point of such release or discharge ~.

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38.

6.6.2 Non-Routine Reports ,

a. Reportable Occurrence Reports In the event of a reportable occurrence (defined in 1.0) notification shall be made 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 appropriate regional USNRC Office of Inspection and Enforcement, followed by a written report within 10 days to the appropriate regional USNRC Office of Inspection and Enforcement with copies to the Director, Office of Inspection and Enforcement, USNRC, Washington, D. C. 20555

( and to the Director, Office of Management Planning and Analysis, USNRC, Washington, D. C. 20555. The written report on these reportable occurrences, and to the extent possible the preliminary telephone and telegraph notification shall: (a) describe, analyze, and evaluate safety implications, (b) outline the measures taken to assure that the cause of the condition is determined, (c) indicate the corrective action (including any changes made to the procedures and to the quality assurance program) :.aken to prevent repetition t of the occurre::,c= and of similar occurrences involving similar components or systems, and (d) evaluate the safety implications of the incident in light of the cumulative experience obtained from the record of previot failures and malfunctions of similar systems-and components.

• Telegraph notification may be sent on the next working day in the event of a reportable occurrence during a weekend or holiday period.

39.

b. Unusual Events A written report shall be forwarded within 30 days to the appropriate regional USNRC Office of Inspection and Enforcement in the event of:

1. Discovery of any substantial errors in the transient or accident analyses or in the methods used for such analyses, as described in the Hazards Summary Report or 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 failure 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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