ML20151R203

Amend 23,renewing License R-2 for 20 Yrs

ML20151R203
Person / Time
Site:	Pennsylvania State University
Issue date:	01/27/1986
From:	Miraglia F Office of Nuclear Reactor Regulation
To:
Shared Package
ML20151R200	List: ML20151R203 ML20151R208 ML20151R217 NUREG-1158, Forwards Amend 23 to License R-2,notice of Renewal,Environ Assessment & SER (NUREG-1158).Amend Renews OL for 20 Yrs from Date of Issuance.W/O SER
References
R-2-A-023, R-2-A-23, NUDOCS 8602050462
Download: ML20151R203 (57)

Text

'./ 'o UNITED STATES

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- 8 o NUCLEAR REGULATORY COMMISSION g ;j WASHINGTON, D. C. 20555

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UNITED STATES NUCLEAR REGULATORY COMMISSION DOCKET NO.50-005 PENNSYLVANIA STATE UNIVERSITY RENEWAL OF THE FACILITY OPERATING LICENSE Amendment No. 23 License No. R-2

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

A. The application for amendment by Pennsylvania State University dated March 1, 1985, as supplemented, 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. Construction of the reactor facility was completed in substantial conformity with Construction Permit No. CPRR-71 dated October 2, 1962, the provisions of the Act, and the rules and regulations of the Comission; C. The facility will operate in conformity with the applicatica, the provisions of the Act, and the rules and regulations of the Comission; D. There is reasonable assurance (i) that the activities authorized by this license can be conducted without endangerirg the health

, and safety of the public, and (ii) that such activities will be .

conducted in compliance with the Comission's regulations; E. The licensee is technically and financially qualified to engage in the activities authorized by this operating license in accord-ance with the rules and regulations of the Comission; F. 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; G. The issuance of this license will not be inimical to the comon defense and security or to the health and safety of the public; H. The issuance of this license is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied; and 8602050462 060107 gDR ADocKOS00g5

I. The receipt, possession and use of the 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 Sections 30.33, 70.23 and 70.31.

2. Facility Operating License No. R-2 is hereby amended in its entirety to read as follows:

A. This license applies to the TRIGA Mark III nuclear reactor owned by Pennsylvania State University (the licensee), and located at University Park, Pennsylvania, and described in the licensee's application for renewal of the license dated March 1, 1985, as supplemented.

B. Subject to the conditions and requirements incorporated herein, the Coninission hereby licenses Pennsylvania State University:

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

" Domestic Licensing of Production and Utilization Facilities,"

to possess, use, and operate the facility at the designated location in accordance with the procedures and limitations set forth in this license.

(2) Pursuant to the Act and 10 CFR Part 70, " Domestic Licensing of Special Nuclear Material," to receive, possess and use up to nine (9) kilograms of contained uranium-235 in connection with operation of the reactor, and 0.900 kilograms of contained uranium-235 in MTR-type fuel elements, and (3) Pursuant to the Act and 10 CFR Part 30, " Rules of General Applicability to Domestic Licensing of Byproduct Material,"

to receive, possess, and use two 50 curie sealed antimony-beryllium neutron sources, either or both of which may be used o

for reactor start-up, or to use a 0.235 milligram californium- -

252 neutron source for operation of the reactor and to possess, but not to separate such byproduct material as may be produced by operation of the reactor.

C. This license shall be deemed to contain and is subject to the conditions specified in Parts 20, 30, 50, 51, 55, 70 and 73 of 10 CFR Chapter I, to ~all applicable provisions of the Act, and to the rules, regulations and orders of the Commission now or hereafter in effect and to the additional conditions specified below:

(1) Maximum Power Level The licensee may operate the reactor at power levels not in excess of 1 megawatt (thermal) and in the pulsing mode with reactivity insertions not to exceed 2.31% Ak/k.

(2) Technical Specifications The technical specifications contained in Appendix A, as revised through Amendment No. 23, are hereby incorporated into the license. The licensee shall operate the facility in accordance with the technical specifications.

(3) Physical Security Plan The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security plan, including all amendments and revisions made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p), which are part of the license. This plan, which contains information withheld from

_ - public disclosure under 10 CFR 2.790, is entitled "The Physical Security Plan-for the Pennsylvania State University Breazeale Reactor Facilities Possessing Special Nuclear Material of Low Strategic Significance" with revisions submitted through December 10, 1984.

D. This license is effective as of the date of issuance and shall expire twenty years from its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION h Fran J. Miraglia, Director Division of PWR Licensing-B

Enclosure:

Appendix A Technical

. Specifications DATE OF ISSUANCE: January 27, 1986

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! FACILITY OPERATING LICENSE R-2

TECHNICAL SPECIFICATIONS .

FOR THE PENNSYLVANIA STATE UNIVERSITY

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BREAZEALE REACTOR r

DOCKET NO.50-005 JANUARY 1986 f

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J I Amendment No. 23 i

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TECHNICAL SPECIFICATIONS FOR THE PENN STATE BREAZEALE REACTOR (PSBR)

FACILITY LICENSE NO. R-2 TABLE OF CONTENTS

1.0 INTRODUCTION

............................................................ 1 1.1 D e f i n i t i o ns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.1.1 1.1.2 ALARA ....................................................... 1 Aut on at i c O p er at i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.1 3 Channel.....................................................

1.1.4 C hann el C al i b r a t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1 1

1.1.5 Channel Check ............................................... 1 1.1.6 Channel Test ................................................ 2 1.1.7 C o l d C r i t i ca l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.1.8 Cl os e Pa c k e d A r r ay . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.1.9 Confinement .................................................

. 2 1.1.10 Cor e Lat ti ce Pos i ti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.1.11 Excess Reactivity ........................................... 2 1.1.12 Experiment .................................................. 2 1.1.13 Exp erim ent al Facili t y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.1.14 Instrumented Element ........................................ 3 1.1.15 Limiting Condi tions f or Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.16 Limi ti ng Saf et y Sys tem Set ti ng . . ; . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.17 Man ua l O p er a t i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.18 Me as ur e d Val ue . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.19 Movable Experiment .......................................... 3 1.1. 20 Norm al i z e d P ow er . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1 .1 . 21 O p er a bl e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.1.22 Operating ................................................... 3 1.1.23 Pulse Mode .................................................. 4 1.1. 2 4 R e a c t i vi t y L i mi ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.1.25 Reactivity Worth of an Experiment ........................... 4 1.1.26 Reactor Interlock ........................................... 4 1.1.27 Reactor Operating ....,....................................... 4 1.1. 28 R ea c t o r S e c ur e d . . . . ; . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.1.29 Reactor Shutdown ............................................ 5 1.1.30 Reactor Safety Systems ...................................... 5 1.1. 31 Ref er ence Cor e Conditi on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1.1 3 2 R es e ar ch R e a c t o r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1.1. 33 R epor t a bl e O cc urr en ce . . . . . . . . . . . . . . . . . . . .'. . . . . . . . . . . . . . . . . . . 5 1.1.34 Rod-Transient ............................................... 6 1.1.35 Safety Limit ................................................ 6 1 . 1 . 3 6 Sc r am T i m e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.1.37 Secured Experiment .......................................... 6

.1.1.38 Secured Experiment wi th Movable Parts . . . . . . . . . . . . . . . . . . . . . . . 6 1.1. 39 Shall , Sho ul d an d May . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.1.40 Shim, Regulating, Safety Rods ............................. 6 1.7.41 S hu t do w n M ar g i n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.1. 4 2 Squar e Wave O per at i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 1.1. 4 3 TRI G A F uel E l em en t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 Amendmeny No. 23

2.0 SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTING ......................... 7 2.1 Saf et y Limit-Fuel El ement Temperat ure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 2.2 Limiting Safety System Setting (LSSS) .............................. 8 30 LIMITING CONDITIONS FOR OPERATION ....................................... 9 3.1 Reactor Core Parameters ............................................

9 3.1.1 constant Power and Square Wave operation .................... 9 3.1.2 Reactivity Limitation ....................................... 10 3.1.3 Snutdown Margin ............................................. 10 3.1.4 Pulse Mode Operation ........................................ 11 3.1.5 core confi gura ti on Limi tat i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 3.1.6 TR IG A F ue l El em en t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 32 contro l and Saf et y Sys t em . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 3.2.1 Reactor control Rods ........................................ 14 3.2.2 Manual and Automati c control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 3 2.3 Reactor Control System ...................................... 15 3 2.4 Reactor Saf et y System and Interlocks . . . . . . . . . . . . . . . . . . . . . . . . 16 3.2.5 core Loading and Unloading Operation . . . . . . . . . . . . . . . . . . . . . . . . 18 3 2.6 S cr am T i m e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 33 c o o l an t S y s t e m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 3 3.1 c ool ant Leve l Lim i ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 3 3.2 Detection of Leak or Less of coolant ........................ 19 3 3.3 Fission Product Activity .................................... 20 3 3.4 Pool Water Supply for Leak Protection ....................... 20 3 3.5 cool ant conducti vit y Limi ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 3 3.6 coolant Tem per at ur e Limi t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 3.4 confinement ........................................................ 22 3.5 Engineered Safety Features - Facility Exhaust System and . . . . . . . . . . . 23 Energency Exhaust System 3.6 Radiation Monitoring System ........................................ 23 3.6.1 Radiation Monitoring Information . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 3.6.2 Eva c ua t i o n Al arm . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 3.6.3 Argon-41 Discharge Limit .................................... 25 3.6.4 ALARA ....................................................... 25 37 Limitations of Experiments.......................................... 25 4.0 SUR VE IL L AN C E R EQ U I R EM ENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 4.1 Reactor 4.1.1 Parameters ................................................. 29 Reactor Power Calibration ................................... 29 4.1. 2 Reactor Excess Reactivity ...........................'........ 29 4.1 3 TR I G A Fue l El em en t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 4.2 Reactor Control and Saf ety System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 4.2.1 Reactivity Worth ............................................ 30 4.2.2 Reactivity Insertion Rate ................................... 31 4.2.3 Reactor Safety and Control Systems .......................... 31 4.2.4 Reactor Interlocks .......................................... 32 4.2.5 overpower Scram ............................................. 33 4.2.6 Transient Rod Test .......................................... 33 4.3 coolant System ..................................................... 34 4.3 1 F i r e Hos e I ns p e c t i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 4.3.2 Pool Wat er Temperat ur e . . . . . . . . . . . . . ........................ 34 4.3 3 Pool Wat er Conducti vi t y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 434 Poo l Wat e r L e ve l A l arm . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 4.4 Confinement ........................................................ 36 Amendment No. 23

4.4.1 Reactor Bay Doors ........................................... 36 4.5 Facility Exhaust System and Emergency Exhaust System ............... 36 4.6 Radiation Monitoring System and Effluents .......................... 37 4.6.1 Radiation Monitoring System and Evacuation Alarm . . . . . . . . . . . . 37 4.6.2 Argon-41 .................................................... 37 4.6 3 ALARA .....................................................

4.7 Experiments ...................................................... . 38

. 38 5.0 D ES I GN F E ATUR ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 5.1 Reactor Fuel ....................................................... 39 5.2 Reactor Core ..................................................... 39 53 Control Rods ..................................................... .

5.4 Fuel Storage ....................................................... 39

. 40 5.5 Reactor Bay and Exhaust Systems .................................... 40 5.6 Reactor Pool Water Systems ......................................... 40 6.0 ADMINISTRATIVE 6.1 CONTROLS ................................................. 41 Organization ....................................................... 41 6.1.1 Structure ................................................... 41 6.1.2 Responsibility .............................................. 41 6.1 3 Staffing .................................................... 43 6.1.4 Selection and Training of Personnel ......................... 43 6.2 Review and Audit ................................................... 43 6.2.1 ' Safeguards Committee Composition ............................ 43 6.2.2 Charter and Rules ...........................................

6.2.3 44 6.2.4 Review Function ............................................. 44 "

Audit ....................................................... 45 6.3 Operating Procedures ............................................... 45 6.4 Review and Approval of Experiments ................................. 46 6.5 Required Action .................................................... 47

. 6.5.1 Action To Be Taken in the Event a Safety Limit is Exceeded .. 47

- 6.5.2 Action To Be Taken in the Event of a Reportable Occurrence .. 47 6.6 Reports ...*......................................................... 48 6.6.1 Operating Reports ......................'..................... 48 6.6.2 Special Reports ..............:.............................. 49 6.7 6.7.1 Records ............................................................ 50 Records To Be Retained for at Least Five Years .............. 50 6.7.2 Records To Be Retained for at Least One Training Cycle ...... 50 6.7 3 Records To Be Retained for the Life of the Reactor Facility . 51 l

Amendment No. 23

TECHNICAL SPECIFICATIONS FOR THE PENN STATE BREAZEALE REACTOR (PSBR)

FACILITY LICENSE NO. R-2

1.0 INTRODUCTION

Included in this document are the Technical Specifications and the bases for the Technical Specifications. These bases, which provide the technical support for the individual technical specifications, are included for infonsation purposes only. They are not part of the Technical Specifications, and they do not constitute limitations or requirements to which the licensee must adhere.

1.1 DEFINITIONS 1.1.1 ALARA The ALARA (As Low As Reasonably Achievable) program is a program for maintaining occupational exposures to radiation and release of radioactive effluents to the environs as icw as reasenably achievable.

1.1.2 AUTCMATIC OPERATION Automatic operation shall mean operation of the reactor with the -

mode selector switch in the automatic position. In this mode, the reactor operates under the control of the servo system.

1.1 3 CHANNEL A channel is the combination of sensor, line, amplifier, and output devices which are connected for the purpose of measuring the value of a parameter.

1.1.4 CHANNEL CALIBRATION A channel calibration is an adjustment of the channel such that its output responds, with acceptable range and accuracy, to known -

values of the parameter which the channel measures. Calibration shall encanpass the entire channel, including equipment actuation, alarm, or trip and shall be deemed to include a Char.1el Test.

1.1.5 CHANNEL CHECK A channel check is a qualitative verification of acceptable performance by observatien of channel behavior. This verification, where possi;le, shall include comparison of the channel with other independent channels or systems measuring the same variable.

$mendmentNo.23

2 1.1.6 CHANNEL TEST A channel test is the introduction of a signal into the channel to verify that it is operable.

1.1.7 COLD CRITICAL The reactor is in the cold critical condition when it is critical with the fuel and bulk water temperatures both below 100*F (37.8'c).

1.1.8 ~CLOSE PACKED ARRAY An arrangement of fuel elements wherein no empty grid positions are comp.letely surrounded by fuel elements.

1.1.9 CONFINEMENT Confinement means an enclosure on the overall f acility which I controls the movement of air into it and out through a controlled path.

1.1.10 CORE LATTICE POSITION The core lattice position is that region in the core over a grid plate hole used to position a fuel element. It may be occupied by a fuel element, a control rod, an experiment, an experimental facility, or a reflector element.

1.1.11 EXCESS REACTIVITY Excess reactivity is that amount of reactivity that would exist if all control rods (safety, regulating, etc.) were moved to the maximum reactive condition from the point where the reactor is exactly critical (k rr=1) e in the reference core condition.

1.1.12 EXPERIMENT Experiment shall mean (a) any apparatus, device, or material which is not a normal part of the core or experimental f acilities, but which is inserted in these facilities or is in line with a beam of

• radiation originating from the reactor core; or (b) any operation designed to measure reactor parameters or characteristics.

1.1.13 EXPERIMENTAL FACILITY .

Experimental facility shall mean beam port, including extension tube with shields, thermal column with shields, vertical tube, central thimble, in-core irradiation holder, pneumatic transfer system, and in-pool irradiation facility.

Amendment No. 23

3 1.1.14 INSTRUMENTED ELEMENT I

An instrumented element is a TRIGA fuel element in which sheathed chromel-alumel or equivalent thermocouples are embedded in the l fuel, i

1.1.15 LIMITING CONDITIONS FOR OPERATION Limiting conditions for operation of the reactor are those

)

constraints included in the Technical Specifications that are required for safe operation of the facility. These limiting conditions are applicable only when the reactor is operating

unless otherwise specified.

1.1.16 LIMITING SAFETY SYSTEM SETTING

, A limiting safety system setting is a setting for an automatic 1

protective device related to a variable having a significant safety function.

i i 1.1.17 MANUAL OPERATION

' Manual operation shall mean operation of the reactor with the mode selector switch in the steady state position which means that the

, power level is established by the operator adjusting the control red positions.

, 1.1.18 MEASURED VALUE The measured value is the value of a parameter as it appears on the output of a channel.

l 1.1.19 MOVABLE EXPERIMENT t

A movable experiment is one where it is intended that the entire experiment may be moved in or near the core or into and out of the

reactor while the reactor is operating.

i 1.1.20 NORMALIZED PCWER e 3

The normalized power, NP, is the ratio of the power of a fuel element to the average power per fuel elenent.

i 1.1.21 OPERABLE t

Operable means a component or system is capable of performing its intended function.

1.1.22 OPERATING Operating means a component or system is performing its intended function.

! Amendment No. 23

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1.1.23 PULSE MODE 4 Pulse mode operation shall mean operation of the reactor with the i

mode selector switch in a pulse position which allows the operator to insert preselected reactivity by the ejection of the transient rod.

, 1.1.24 REACTIVITY LIMITS i

The reactivity limits are those limits imposed on reactor core

} reactivity. Quantities are referenced to a Reference Core i

Condition.

1.1.25 REACTIVITY WORTH OF AN EXPERIMENT l

The reactivity worth of an experiment is the maximum absolute value of the reactivity change that would occur as a result of intended or anticipated changes or credible malfunctions that alter experiment position or configuration. .

1.1.26 REACTOR INTERLOCK i

A reactor interlock is a device which prevents some action, associated with reactor operation, until certain reactor operation conditions are satisfied.

1 1.1.27 REACTOR OPERATING The reactor is operating whenever it is not secured or shutdown.

1.1.28 REACTOR SECURED 4

The reactor is secured when:

a. It contains insufficient fissile material or moderator present in the reactor, adjacent experiments, or control rods, to attain criticality under optimum available conditions 'of moderation and reflection, or

b. A combination of the following:

(1 ) The minimum number of neutron absorbing control rods are

fully inserted or other safety devices are in shutdown positions, as required by technical specifications, and

{. (2) The console key-switch is in the off position and the key is removed from the lock, and i

I (3) No work is in progress involving core fuel, core structure, installed control rods, or control rod drives unless they are physically decoupled from the control

, rods, and

! (4) No experiments in or near the reacter are being moved or serviced that have, on movement, a reactivity worth

' exceeding the maximum value allowed for a single experiment or one dollar whichever is smaller.

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5 1.1.29 REACTOR SHUTDOWN -

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The reactor is shutdown if it is subcritical by at least one dollar in the Reference Core Condition and the reactivity worth of all experiments is accounted for.

1.1 3C gEigTOR SAFETY SYSTEMS Reactor safety systems are tho:e systems, including their associated input channels, which are designed to initiate automatic reactor protection or to provide information for initiation of manual protective action.

1.1 31 REFERENCE CORE CONDITION i l

The condition of the core when it is at ambient temperature (cold) and the reactivity worth of xenon is negligible (< 30 dollars).

1.1 32 RESEARCH REACTOR A research reactor is defined as a device designed to support a self-sustaining neutron chain reaction for research, development, 1 educational, training, or experimental purposes, and which may have provisions for the production of radioisotopes.

1.1 33 REPORTABLE OCCURRENCE A reportable occurrence is any of the following which occurs {

.. during reactor operation:

a. Operation with the safety system setting less conservative

,. than specified in Section 2.2, Limiting Safety System Setting.

b. Operation in violation of a limiting condition for operation,

c. Failure of a required reactor safety system component which i ,

could render the system incapable of performing its intended

safety function.

d. Any unanticipated or uncontrolled change in reactivity greater j than one dollar. -

e. An observed inadequacy in the implementation of either administrative or procedural controls which could result in operation of the reactor outside the limiting conditions for operation.

f. Release of fission products from a fuel element.

g. Abnormal and significant degradation in reactor fuel, cladding,

, coolant boundary or containment boundary that could result in l exceeding 10 CFR 20 exposure criteria.

i Amendment No. 23

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1.1 34 ROD-TRANSIENT The transient rod is a control rod with scram capabilities that is capable of providing rapid reactivity insertion to produce a pulse or square wave.

1.1 35 SAFETY LIMIT Safety limits are limits on important process variables which are found to be necessary to reasonably protect the integrity of certain of the physical barriers which guard against the uncontrolled release of radioactivity. The principal physical barrier is the fuel element cladding.

1.1 36 SCRAM TIME Scram time is the elapsed time between reaching a limiting safety system set point and a specified control rod movement.

1.1 37 SECURED EXPERIMENT A secured experiment is any experiment, experiment facility, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means. The restraining forces must be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the experiuent, or by forces which can arise as a result of credible malf unctions .

1.1 38 SECURED EXPERIMENT WITH MOVABLE PARTS

• A secured experiment with movable parts is one that contains parts "

that are intended to be moved while the reactor is operating.

1.1 39 $RALL, SHOULD AND MAY The word "shall" is used to denote a requirement; the word "should" to denote a recommendation; and the word ."may" to denote

! permission, neither a requirement nor a recommendation.

1.1.40 SHIM REGULATING. SAFETY RODS A shim, regulating, or safety rod is a control rod having an electric motor drive and scram capabilities. It has a fueled follower section.

1.1.41 SHUTDCWN MARGIN Shutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence that the reactor can be made subcritical by means of the control and safety systems starting from any permissible operation condition although the most Amendment No. 23 i

7 reactive rod is in its most reactive position, and that the reactor will remain suberitical without f urther operator action.

1.1.42 SQUARE WAVE OPERATION Square wave (SW) operation shall mean operation of the reactor with the mode selector switch in the square wave position which allows the operator to insert preselected reactivity by the ejection of the transient red, and which results in a maximum power of 1 MW or less.

1.1.43 TRIGA FUEL ELEMENT A TRIGA fuel element is a single TRIGA fuel rod of standard type, either 8.5 wt% U-ZrH in stainless steel cladding or 12 wt% U-ZrH

,in stainless steel cladding enriched to less than 20% u' anium-235.

r 2.0 SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTING 2.1 SAFETY LIMIT-FUEL ELEMENT TEMPERATURE Applicability The safety limit specification applies to the maximum temperature in the reactor fuel.

Objective The objective is to define the maximum fuel element temperature that can be permitted with confidence that no damage to the fuel element

, and/or cladding will result.

Specifications The temperature in a water-cooled TRIGA fuel element shall not '

exceed 1150*C under any operating condition.

Basis

• The important parameter for a TRIGA reactor is the fuel element t emperature . This parameter is well suited as a single specification especially since it can be measured at a point within the fuel element. The measured fuel temperature is directly related to the maximum f uel temperature of the region. A loss in the integrity of the fuel element cladding could arise from a build-up

. of excessive pressure between the fuel-moderator and the cladding if the maximum fuel temperature exceeds 1150*C. The pressure is caused by the presence of air, fission product gases, and hydrogen from the dissociation of the hydrogen and zirconium in the f uel-moderator.

The magnitude of this pressure is determined by the fuel-moderator temperature, the ratio of hydrogen to zirconium in the alloy, and the rate change in the pressure.

Amen'dment No. 23

8

The safety limit for the standard TRIGA fuel is based on data, including the large mass of experimental evidence obtained during high performance reactor tests on this fuel. These data indicate that the stress in the cladding due to the increase in the hydrogen pressure from the dissociation of zirconium hydride will remain below the ultimate stress provided that the temperature of the f uel does not exceed 1150'C (2102'F) and the fuel cladding is water cooled. See Safety Analysis Report, Ref. 13 in section IX and Simnad , M.T. , F.C. Foushee , and G.B. West , " Fuel Elements f or Pulsed Reactors ," Nucl. Technology, Vol.28, p. 31-56 (January 1976) .

2.2 LIMITING SAFETY SYSTEM SETTING (LSSS)

Applicability The LSSS specification applies to the scram setting which prevents the safety limit from being reached.

Objective The objective is to prevent the safety limit (1150'C) from being reached.

1 Specifications The limiting safety system setting shall be a maximum of 700'C as measured with a 12 wt% U-ZrH instrumented fuel element. The instrumented fuel element shall be located in the B-ring and adjacent to an empty fuel position when an empty fuel position exists in the B-ring.

Basis The limiting safety system setting is a temperature which, if reached shall cause a reactor scram to be initiated preventing the safety limit from being exceeded. Experidents and analyses described in the Safety Analysis Report,Section IX - Safety Evaluation, show that the measured fuel temperature at steady state power has a simple linear relationship to the normalized power or power of the highest powered fuel element in the core. Maximum fuel temperature occurs when a new 12 wt1 U-ZrH fuel element is placed in the B-ring of the core. The measured fuel temperature during steady state operation is close to the maximum fuel temperature. Thus, 450*C of safety margin exists before the 1150'c safety limit is reached. This safety margin provides adequate compensation for using a depleted instrumented 12 wt% U-ZrH fuel element instead of an unirradiated one to measure the fuel temperature. See Safety Analysis Report,Section IX.

In the pulse mode of operation, the same limiting safety system setting shall apply. However, the temperature channel will have no effect on limiting the peak power generated, because of its relatively long time constant (seconds), compared with the width of the pulse (milliseconds). In this mode, however, the temperature trip will act i

Amendment No. 23

= - , - ,, -

to reduce the amount of energy generated in the entire pulse transient, by cutting the " tail" of the power transient if the pulse rod remains stuck in the fully withdrawn position with enough reactivity to exceed the temperature-limiting safety system setting.

3.0 LIMITING CONDITIONS FOR OPERATION The limiting conditions for operation as set forth in this section are applicable only when the reactor is operating. They need not be met when

.the reactor is shutdown unless specified otherwise.

31 REACTOR CORE PARAMETERS 3.1.1 CONSTANT POWER AND SQUARE WAVE OPERATION Applicability This specification applies to the maximum power generated during

. manual, square wave, and automatic operation.

Objective The objective is to assure that the safety limit (fuel temperature) will not be reached during manual, square wave, and automatic

. operation by providing a set point to automatically limit the maximum fuel temperature produced in the core and to limit the

2. energy produced in any seven (7) consecutive days to that used in

the LOCA analysis in the Safety Analysis Report.

Specification

a. The operating power level of the reactor shall not be intentionally raised above one megawatt except for pulse operation (see specification 3.1.4).

b. The reactor shall not be operated to produce more than 70 megawatt hours of energy in any seven (7) consecutive days.

Basis

a. Thermal and hydraulic calculations .and operational experience indicate that a compact TRIGA reactor core can be safely operated up to power levels of at least '1.15 megawatts with

-natural convective cooling. Power operation at 1.15 megawatts will not produce fuel temperatures which exceed 600*C using any allowed core configuration givin *g a large safety measure when the power of operation is limited to 1 MW. Thus, small variations can occur about 1 MW during ncrmal operation and still provide a large measure of safety in that the maximum fuel temperature remains well below the saf ety limit. This is true even if variations as high as 15% above 1 MW should occur. See Safety Analysis Report, section IX.

Amendment No. 23 .,

_. =_. - - - -

10

b. This specification limits the energy output of the PSBR to that used in the Safety Analysis Report (SAR) analysis of a Maximum Hypothetical Accident (MHA) and a total Loss of Coolant Accident (LOCA).

3.1.2 REACTIVITY LIMITATION Applicability This specification applies to the reactivity condition of the reactor and the reactivity worth of control rods, experiments, and

. experimental facilities. It applies to all modes of operation.

Objective The objective is to assure that the reactor is operated within the limits analyzed in the SAR and to assure that the safety limit will not be exceeded.

Specifications The maximum excess reactivity above cold, clean, critical plus samarium poison of the core configuration with experiments and experimental facilities in place shall be 4.9% ak/k (-37.00).

Basis Limiting the excess reactivity of the core to 4.9% ak/k prevents the fuel temperature in the core from exceeding 1150*C under any assumed accident condition as described in the Safety Analysis Report,Section IX - Satety Evaluation.

3.1.3 SHUTDOWN MARGIN Applicability This specification applies to the reactivity condition of the reactor and the reactivity worth of control rods, experiments, and experimental facilities. It applies to all modes of operation.

Objective The objective is to assure that the reactor can be shut down at all

. times and to assure that the safety limit will not be exceeded.

Specifications The reactor shall not be operated unless the shutdcwn margin provided by control rods is greater than 1.75 x 10-3 ak/k (-$0.25) with:

Amendment No. 23

22

a. All movable experiments, experiments with movable parts, and experimental facilities in their most reactive state.

b. The highest reactivity wo.ath control rod fully withdrawn.

Basis A shutdown margin of 1.75 x 10-3 ak/k assures that the reactor can be made subcritical from any operating condition even if the highest worth control rod should remain in the fully withdrawn position. In addition, the 1.75 x 10-3 ak/k can be easily measured, and therefore verifying that the operation meets this specification. The shutdown margin requirement supersedes Specifications 3.1.2.

3.1.4 PULSE MODE OPERATION Applicability This specification applies to the energy generated in the reactor as a result of a pulse insertion of reactivity.

Objective

~

IRue objective is to assure that the safety limit will not be exceeded during pulse mode operation.

Specification

. a. The stepped reactivity insertion for pulse operation shall not

.T exceed 2.315 Ak/k (-43 30) and the maximum worth of the poison

' section of the transient rod shall be limited to 2.59% Ak/k

(-43 70). When the core excess reactivity is less than 2.595 Ak/k (-43.70) the limit of the maximum worth of the poison section of the transient rod is not applicable.

b. Pulses shall not be initiated from power levels above 1 kw.

Basis

a. Experiments and analyses described in the SAR show that the peak pulse temperatures can be predicted for new 12 wt$ fuel placed in the B-ring. These experiments and analyses show that the maximum allowed pulse reactivity of 2 315 ak/k prevents the maximum fuel temperature from reaching the safety limit for any allowed core configuration, including those accidentally pulsed when operating at 1 MW.

The maximum worth of the pulse rod is limited to 2.595 ak/k to prevent exceeding the safety limit (1150*C). Accidental ejection of the transient rod during cold clean conditions will limit the maximum measured temperature to 720*C and the maximum fuel temperature to 1150*C. See Safety Analysis Report,Section IX. When the core excess reactivity is less than 2.59% Ak/k

(-43.70), it is not possible to pulse above 2.59% Ak/k (-$3.70)

Amendment No. 23

12 so that any possible pulse will not produce temperatures that exceed the safety limit, b.- If a pulse is initiated from power levels below 1 KW, the maximum allowed full worth of the pulse rod can be used without exceeding the safety limit.

3 1.5 CORE CONFIGURATION LIMITATION Applicability This specification applies to a core configuration with water-filled fuel positions producing large power peaking in some fuel elements.

Objective The objective is to asdure that the safety limit will not be exceeded due to power peaking effects in the various core geometries.

Specifications

a. The critical core shall be an assembly of either 8.5 wt%

stainless steel clad or a mixture cf 8.5 wt% and 12 wt5 stainless steel clad TRIGA fuel-moderator elements placed in water with a 1.7 inch center line grid spacing.

b. The fuel and fueled follower control rods shall be arranged in a close packed array except for (1) single positions and (2) positions the centers of which are greater than 4 inches from the center of the core where flux peaking dhd corresponding power densities produce fuel temperatures less than in the B-ring. -

c.' When the ek pt of the core is less than or equal, to 0.99 with all control rods at their upper limit, the ' fuel may or may not be arranged in a close packed array. The so,urce and detector shall be arranged such that the ek rf of the subcritical assembly shall always be monitored to assure compliance with kerr < 0.99 when all control rods are fully withdrawn.

Basis Calculations and experiments performed with the PSBR have shown that with only one empty fuel position in the central region of the core defined as lattice positions with centers less than 4 inches from the core center, the power peaking for any mixture of 12.0 wt% and 8.5 -wt% fuel remains less than 23 2 kw per fuel element, i.e., an NP < 2.2 (see Safety Analysis Report,Section IX - Safety Evaluation). This automatically limits the maximum fuel temperature, which always occurs in the B-ring, to well below 700*C for any mode of operation. The maximum fuel temperature possible in Amendment No. 23 -

13 a TRIGA core will occur in a. new 12 wt% fuel element in the B-ring adjacent to a water filled fuel position also in the B-ring.

When the keft of the core is less than or equal to 0.99 with all control rods at their upper limit, the core can not be taken critical. Hence, the requirement for close packed arrays is not necessary to prevent the core from attaining high fuel temperatures.

3.1.6 TRIGA FUEL ELEMENTS Applicability This specification applies to the mechanical condition of t?.e fuel.

Obj ective The objective is to assure that the reactor is not operated with damaged fuel that might allow release of fission products.

Specification The reactor shall not be operated with damaged fuel except to detect and identify the fuel elenent for removal. A TRIGA fuel element shall be removed from the core if:

a. In measuring the transverse bend, the bend exceeds 0.125 inch over the length of the cladding.

b. In measuring the elongation, its length exceeds its original length by 0.125 inch.

c. A clad defect exists as indicated by release of fission products.

Basis .

The limit of transverse bend has been shown to result in no difficulty in disassembling the core. Analysis of the removal of heat from touching fuel elements shows that there will be no hot spots resulting in damage to the fuel caused by this touching.

Experience with TRIGA reactors has shown that fuel element bowing that could result in touching has occurred without deleterious effects. This is because (1) during steady state operation, the maximum fuel temperatures are several hundred degrees Centigrade below 1150*C (the safety limit), and (2) during a pulse, the cladding temperatures remain well below their stress limit. The elongation limit has been specified to assure that the cladding material will not be subjected to strains that could cause a loss of integrity in the fuel containment and to assure adequate coolant flow .

i Amendment No. 23

14 3.2 CONTROL AND SAFETY SYSTEM 3.2.1 REACTOR CONTROL RODS Applicability This specification applies to the reactor control rods.

Objective The objective is to assure that sufficient control rods are operable to maintain the reactor subcritical.

Specification There'shall be a minimum of three operable control rods in the reactor core.

Basis The shutdown margin and excess reactivity specifications require that the reactor can be made subcritical with the most reactive control rod withdrawn. This specification helps assure it.

3.2.2 MANUAL AND AUTCMATIC CONTROL Applicability This specification applies to the maximum reactivity insertion rate associated with movement of a standard control rod.

Objective The objective is to assure that adequate control of the reactor can be maintained during manual and automatic operation.

Specification The maximum rate of reactivity insertion associated with movement of either the regulating, shim, or saf ety control rod shall be no greater than 0.12% ak/k (-17c) per second.

Basis This limits the insertion of reactivity to a rate much less than that during a pulse insertion of 2 31% ak/k reactivity. At a maximum insertion rate of 0.12% Ak/k/sec it takes almost 6 seconds to insert 0.007 ak/k (-$1.0) of r activity; the large negative temperature coefficient of the core (see page IX-29 of the SAR) limits the increase of the average core fuel temperature to 70*C due to the 0.007 ak/k insertion. Thus, the core temperature will conpensate for the rate of insertion of reactivity. In addition, the location of the maximum f uel temperature will be the same as that during constant power operation, i.e., near the position of the Amendment No. 23

15 thermocouple. Hence, when either the linear, percent power, or temperature scram occurs, the maximum f uel temperature will be f ar below the 1150'C safety limit.

f 3.2 3 REACTOR CONTROL SYSTEM Applicability This specification applies to the information which must be available to the reactor operator during reactor operation.

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

Specification The reactor shall not be operated unless the measuring channels listed in Table 1 are operable. (Note that MN. AU and SW are abbreviations for manual, automatic and square wave, respectively) .

Table 1

- Measuring Channels Min. No. Effective Mode Measuring Channel Operable MN.AU Pulse SW 2

Fuel Element Temperature 1 X X X Linear Power 1 X X Percent Power 1 X X

Pulse Peak Power 1 X Count Rate 1 X Log Power 1 X X Reactor Period 1 X Basis Fuel temperature displayed at the control console gives continuous information on this parameter which has a specified safety limit.

The power level monitors assure that the reactor power level is adequately monitored for the manual, automatic, square wave and pulsing modes of operation. The specifications on reactor power level and reactor period indications are included in this section to provide assurance that the reactor is operated at all times within the limits allowed by these Technical Specifications.

Amendment No'. 23

. . _ - _ - _9_

16 '

9 3 2.4 REACTOR SAFETY SYSTEM AND INTERLOCKS Applicability This specification applies to both the reactor safety system channels and the interlocks.

Objective The objective is to specify the minimum number of reactor safety system :hannels and interlocks that must be operable for safe operation.

Specification The reactor shall not be operated unless all of the channels and interlocks described in Table 2a and Table 2b are operable.

Table 2a Minimum PSBR Safety Channels Number Effective Mode Safety Channel Goerable Function MN,AU Pulse SW Fuel Temperature 1 'X SCRAM i 700*C X X

- High Power 2 X SCRAM i 110% of 1 MW X Detector Power 1 SCRAM on failure of X X Supply supply voltage Scram Bar on Console 1 Manual scram X- X X Preset Timer 1 Transient rod scram X 15 seconds or less after pulse Amendment No. 23 '

17 Table 2b Minimum PSBR Safety Interlocks Number Ef fective Mode Safety Interlocks Operable Function MN Pulse SW Source Level 1 Prevent rod withdrawal X with less than two neutron induced counts per second on the startup channel Log Power 1 Prevent pulsing from X levels above 1 kW Transient Rod 1 Prevent applications X of air unless cylinder is f ully inserted Shim, Safety, and 1 Movement of any rod X Regulating Rod except transient rod Simultaneous Rod 1 Prevents simultaneous X X Withdrawal manual withdrawal of two rods Basis A temperature scram and two power level scrams provide automatic protection to assure that the reactor is shut down before the safety limit on the fuel element temperature will be exceeded. The manual scram allows the operator to shut down the system in any mode.cf operation if an unsafe or abnormal condition occurs. In the event of failure of the power supply for the safety chambers, operation of the reactor without adequate instrumentation is prevented. The preset timer insures that the transient rod will be inserted and the reactor will remain at low power after pulsing.

In the pulse mode, movement of any rod except the. transient rod is prevented by an interlock. This interlock action prevents the addition of reactivity over that in the transient rod. The interlock to prevent startup of the reactor with less than 2 cps assures that sufficient neutrons are available for proper startup in all relevant modes of operation. The interlock to prevent the initiation of a pulse above 1 kW is to assure that the magnitude of the pulse will not cause the safety limit' to be exceeded. The interlock to prevent application.of air to the transient rod unless the cylinder is fully inserted is to prevent pulsing the reactor in the manual mode. Simultaneous manual withdrawal of two rods is prevented to assure the reactivity rate of insertion is not exceeded.

Amendment No. 23

18 3.2.5 CORE LOADING AND UNLOADING OPERATION Applicability This specification applies to the low count rate interlock.

Objective The objective of this specification is to eliminate interference with fuel loading procedures.

Specification During core loading and unloading operations when the reactor is subcritical, the low count rate interlock may be momentarily defeated using a spring loaded switch in accordance with the fuel loading procedure.

Basis During core loading and unloading, the reactor is suberitical.

Thus, momentarily defeating the count rate is a safe operation.

Should the core become inadvertantly supercritical, the accidental insertion of reactivity will not allow fuel temperature to exceed the 1150*C safety limit because no single TRIGA fuel element is worth more than 1% Ak/k in the most reactive core position.

.3.2.6 SCRAM TIME

- . Applicability This specification applies to the time required to fully insert any control rod to a full down position from a full up position.

Objective The objective is to achieve rapid shutdown of the reactor to prevent fuel damage.

Specification The time from scram initiation to the full insertion of any control rod from a full up position shall be less than 1 second.

Bas,1,s, This spacification assures that the reactor will be promptly shut down when a scram signal is initiated. Experience and analysis, 3

d Section IX, SAR, have indicated that for the range of transients anticipated for a TRIGA reactor, the specified scram time is adequate to assure the safety of the reactor. If the scram signal is initiated at 1.10 MW, while the control rod is being withdrawn, Amendment No. 23 o

-n-- . , ~ - , -, . . , . - _ _ . , , , , . , ._- . . _ , - , - , - - _ - , - - . . - , . - - - - - - - --

. 19 and the negative reactivity is not inserted until the end of the one second rod drop time, the maximum fuel temperature does not reach the safety limit.

]

33 COOLANT SYSTEM 3 3.1 coolant Level Limits Applicability This specification applies to operation of the reactor with respect to a required depth of water above the top of the bottom grid plate.

Obj ective The objective is to assure that water is present to provide adequate personnel shielding and core cooling when the reactor is operated, and during a L0cA.

Specification The reactor shall not be operated with less than 18 ft. of water above the top of the bottom grid plate.

} Basis When the water is more than approximately 18 ft. above the top of the bottom grid plate, the water provides sufficient shielding to protect personnel during operation at 1 MW, and core cooling is achieved with natural circulation of the water through the core.

Should the water level drop below approximately 17 f t. above the top of the bottom grid plate while operating at 1 MW, the radiation i

monitors on the bridge will automatically scram the reactor. Once this occurs it will take longer than 1000 sec before the pool water

)'

level drops 17 ft. in the peal. Tests and calculations show that '

1000 sec. is sufficient time during a LOCA (see Section IX,of the SAR) to prevent the fi:al temperature from reaching 900*C, The 900*C 1

is a conservative limit for the fuel temperature for an air cooled fuel element to prevent cladding rupture.

332 DETECTION OF LEAK OR LOSS OF COOLANT Applicability This specification applies to detecting a pool water loss.

Obj ective The objective is to detect the loss of a significant amount of pool water.

t Amendnent No. 23 -

20 Specification A pool level alarm shall be activated and corrective action taken when the pool' level drops 26 cm from a level where the pool is full.

Basis This alarm level occurs when the water level is approximately 18 f t.

above the top of the bottom grid plate . During reactor operation, the reactor staff will take action to keep the core covered with water according to existing procedures. The alarm is also transmitted to the Police Services annunciator panel which is monitored 24 hrs. a day. The alarm provides a signal that occurs at all times (see section III. SAR). Police Services and the reactor staff will move quickly during non-operation of the reactor to maintain water in the pool. Thus, the alarm provides time to initiate corrective action before the radiation from the core poses a serious hazard.

333 FISSION PRODUCT ACTIVITY Applicability This specification applies to the detection of fission product activity. ,

Objective The objective is to assure that fission products from a leaking fuel 1 element are detected to provide opportunity to take protective action.

Specification An air particulate monitor shall be operating in the reactor bay whenever the reactor is operating. An alarm on this unit shall activate a building evacuation alarm.

Basis This unit will be sensitive to airborne radioactive particulate matter containing fission products and fission gases and will alert personnel in time to take protective action.

3 3.4 PCOL WATER SUPPLY FOR LEAK PROTECTION Applicability j

This specification applies to pool water supplies for the reactor pool for leak protection.

Amendment No. 23

21

' Objective The objective is to assure that a supply of water is available to replenish reactor pool water in the unlikely event of pool water leakage.

Specification A source of water of at least 100 GPM shall be available either from the University water supply or by diverting the heat exchanger secondary flow to the pool.

Basis Provisions for both of these supplies are in place and will supply more than the specified flow rate. This flow. rate will be more than sufficient to handle leak rates that have occurred in the past or any anticipated leak that might occur in the future.

3.3.5 COOLANT CONDUCTIVITY LIMITS Applicability

_. This specification applies to the conductivity of the water in the pool.

Obj ective

The objective is:

a. To prevent activated contaminants from becoming a radiological hazard.

b. To help preclude corrosion of fuel cladding and other primary system components.

Specification The reactor shall not be operated if the conductivity of the bulk pool water is greater than 5 micromhos/cm.

Basis Experience indicates that 5 micromhos/cm is an acceptable level of water contaminants in an aluminum / stainless steel system such as 1

that at the PSBR. Based on experience, activation at this level does not pose a significant radiological hazard, and significant corrosion of the stainless steel fuel cladding will not occur when the conductivity is in this range.

t Amendment No. 23

22 "

3 3.6 COOLANT TEMPERATURE LIMITS Applicability This specification applies to the pool water temperature.

Objective The objectiv,e is to maintain the pool water temperature at a level that will not cause damage to the demineralizer resins.

Specifications An alarm shall annunciate and corrective action shall be taken if during operation the bulk pool water temperature reaches 100*F (37.8*c).

Basis This specification is primarily to preserve demineralizer resins.

Information available indicates that temperature damage will be minimal up to this temperature.

34 CONFINEMENT Applicability This specification applies to reactor bay doors.

Objective The objective is to assure that no large air passages exist to the reactor bay during reactor operation.

Specification The reactor bay truck door shall be closed when the reactor is operating. Personnel doors to the reactor bay shall not be blocked open and left unattended when the reactor is operating.

Basis This specification helps to assure that the air pressure in the reactor bay is lower than the remainder of the building and the outside air pressure. Controlled air pressure is maintained by the air exhaust system and assures controlled release of any airborne radioacti vity .

Amendment No. 23

23 4

3.5 ENGINEERED SAFETY FEArURES - FACILITY EXHAUST SYSTEM AND EMERGENCY EXHAUST SYSTEM Applicability This specification applies to the operation of the facility exhaust system and the emergency exhaust system.

Objective The objective is to mitigate the consequences of the release of airborne radioactive materials resulting from reactor operation.

Specification i

The facility exhaus't system shall be operating, and the emergency exhaust system shall be maintained in an operable condition when the reactor is operating except for periods of. time less than 48 hrs, when it is necessary to permit maintenance and repairs.

Basis During normal operation, the concentration of airborne radioactivity '

in unrestricted areas is below MPC as described in the Safety Analysis Report,Section IX - Safety Evaluation. In the event of a substantial release of airborne radioactivity, an air radiation

'monitcr and/or an area radiation monitor will sound a building evacuation alarm which will automatically cause the facility exhaust

t system to close and the exhausted air to be passed through the emergency exhaust system filters before release. This reduces the radiation within the building. The filters will reduce to <10% all of the particulate fission products that escape to the atmosphere.

Radiation monitors, Section 3.6.1 of these Technical Specifications j

and Section VII.F of SAR, within the building, independent of the exhaust systems, will give warning of high levels of radiation that might occur during operation with the exhaust systems out of -

service.

3.6 RADIATION MONITORING SYSTEM 3.6.1 RADIATION MONITORING INFORMATICN Applicability This specification applies to the radiation monitoring information which must be available to the reactor operator during reactor operation.

Objective The objective is to assure that sufficient radiation monitoring information is available to the operator to assure safe operation of the reactor.

A Amendment No. 23

24 Specification The reactor shall'not be ope"ated unless the radiation monitoring channels listed in Table 3 are operating.

Table 3 Radiation Monitoring Channels Radiation Monitoring Channels Function Nu=ber Area Radiation Monitor Monitor radiation levels 1 in the reactor bay Continuous Air Monitor radioact'ive 1 (Radiation) Monitor particulates in the reactor bay Beashole Laboratory Monitor radiation in the 1 Monitor Beamhole Laboratory only when the laboratory is in use.

Basis The radiation monitors provide information to operating personnel of any impending or existing danger from radiation so that there will

} be sufficient time to evacuate the facility and to take the necessary steps to control the spread of radioactivity to the surroundings .

The area radiation monitor in the Beam Hole Laboratory provides information to the user-and to the reactor operator when this laboratory is in use.

3.6.2 EVACUATION ALARM Applicability This sp'ecification applies to the evacuation alarm.

Obj ective The objective is to assure that all personnel are alerted to evacuate the PSBR building when a potential _ radiation hazard exists within th15 building.

Scecification z

The reactor shall not be operated unless the evacuation a?. arm is operable and audible to personnel within the P.SdR building when activated by the radiation monitoring channel 3 in Table 3 or a manual switch.

Amendment No. 23

25 Basis The evacuation alarm produ 'es a loud pulsating sound .throughout the PSBR building when there is any impending or existing danger from radiation. The sound notifies all personnel within the PSBR building to evacuate '.he building as prescribed by the PSBR emergency procedure.

3.6.3 ARGON-41 DISCHARGE LIMIT Applicability This specification applies to the concentration of argon-41 that may be discharged from the PSBR.

Objective The objective is to insure that the health and safety of the public is not endangered by the discharge of argon-41 from the PSBR.

Specification -

All Ar-41 concentrations produced by the operation of the reactor shall be bel.ow the limits imposed by 10 CFR 20 when~ averaged over a year.

Basis The maximum allowable concentration of argon-41 in air in

?? unrestricted areas as specified in Appendix B, lable II of 10 CFR 20 is 4.0 x 10'OuCi/ml . Measurements of argon-41 have been made in the reactor bay when the reactor operates at 1 MW. These measurements e,

show that the concentrations averaged over a year produce much less than 4.0 x 10'OuC1/ml in.an unrestricted area (see Environmental Impact Appraisal, page 3).

3.6.4 ALARA Applicability This specification applies to all reactor operations that could result in occupational exposures to radiation or the release of radioactive effluents to the environs.

Obj ective The objective is to maintain all exposures to radiation and release of radioactive effluents to the environs as low as reasonably achierable.

Specification An ALARA program shall be in effect.

1 Amendment No. 23

26' Basis Having an ALARA program will assure that occupational exposures to radiation and the release of radioactive effluents to the environs will be as low as reasonably achievable. Having such a formal program will keep the staff cognizant of the importance to minimize radiation exposures and effluent releases.

3.7 LD4ITATIONS OF EXPERIMENTS Apolicability -

This specification applies to experiments installed in the reactor and its experimental facilities.

Objective The objective is to prevent damage to the reactor and to prevent excessive release of radioactive materials in the event of an experiment f ailure.

Scecifications The reactor shall not be operated unless the following conditions governing experiments exist:

a. The reactivity of a movable experiment and/or movable portions of a secured experiment plus the maximum allowed pulse reactivity shall be less than 2.59% ak/k (-$3.70). However, the reactivity of a movable experiment and/or movable portions of a secured experiment shall have a reactivity worth less than 1.4%

ak/k (-$2.00). When the movable experiment is worth more than 0.28% ak/k (-$0.41), the maximum allowed pulse shall be reduced below the allowed pulse reactivity insertion of 2 31% ak/k

(-$3 30) to assure that the sum is less 2.59% ak/k.

b. A single secured experiment shall be limited to a maximum of 2.31% ak/k (-$3 30). The sum of the ' reactivity worth of all experiments shall be less than 2.59% ak/k (-$3 70).

c.

When the keft of the core is less than 1 with all control rods at their upper limit and no experiments in or near the core, secured negative reactivity experiments may be added without limit.

d. Explcsive materials in quantities greater than 250 milligrams shall not be allowed within the PSBR facility. Irradiation of explosive materials shall be restricted as folicws:

i Explosive materials in quantities greater than 25 milligrams shall not be irradiated. Explosive materials in quantities of 25 milligrams or less may be irradiated provided the pressure l

produced upon detonation of the explosive has been calculated I

Amendment No. 23

27 and/or experimentally demonstrated to be less than the design pressure of the container ,

e. Experiment materials, except fuel materials, which could off-gas, sublime, volatilize, or produce aerosols under (1) normal operating conditions of the experiment and reactor, (2) credible accident conditions in the reactor, or (3) possible accident conditions in the experiment, shall be limited in activity such. that the airborne concentration of radioactivity averaged over a year shall not exceed the limit of Appendix B Table II of 10 CFR Part 20.

When calculating activity limits, the following assumptions shall be used:

(1) If an experiment fails and releases radioactive gases or aerosols to the reactor bay or atmosphere, 100% of the gases or aerosols escape.

(2) If the effluent from an experimental f acility exhausts through a holdup tank which closes automatically on high radiation level, at least 10% of the gaseous activity or aerosols produced will escape.

~

(3) If the effluent from an experimental facility exhausts through a filter installation designed for greater than 99%

efficiency for 0 3 micron particles, at least 10% of these vapors can escape.

J. (4) For materials whose boiling point is above 130*r and where

!. vapors formed by boiling this material can escape only

' through an undisturbed column of water above the core, at least 10% of these vapors can escape.

f.

Each fueled experiment shall be controlled such that the total inventory of iodine isotopes 131 through 135 in the experiment is no greater than 1.5 curies. In addition, any f ueled experiment which would generate an inventory of more than 5 millicuries (mC1) of I-131 through I-135 shall be reviewed to assure that in the case of an accident, the total relsase of iodine will not exceed that postulated for the maximum hypothetical accident (see Section IX.E. SAR).

g.

If a capsule fails and releases material which could damage the reactor fuel or structure by corrosion or other means, physical inspection shall be performed to determine the consequences and need for corrective action. The results of the inspection and any corrective action taken shall be reviewed by the Director or a designated alternate and determined to be satisfactory before operation of the reactor is resumed, i

l Amendment No. 23 l

28 Basis l

a. This specification limits the sum of the reactivities of a maximum allowed pulse and a movable experiment to the specified maximum reactivity of the transient rod. This limits the effects of a pulse simultaneous with the failure of the movable experiment to the effects analyzed for a 2.59% ak/k (-$3 70) pulse. In addition, the maximum power attainable with the ramp insertion of 1.4% ak/k (-$2.00) is less than 500 kw starting from critical.

b. The maximum worth of all experiments is limited to 2.59% ak/k

(-$3.70) so that their inadvertant sudden removal from the cold ,

critical reactor will not result in the reactor achieving a power level high enough to exceed the core temperature safety l limit. The worth of a single secured experiment is limited to j the allowed pulse reactivity insertion as an increased measure of. safety. Should the 2 31% ak/k reactivity be inserted by a ramp increase, the maximum power attainable is less than 1 MW.

c. Since the initial core is subcritical, adding and then 1 inadvertantly removing all negative reactivity experiments leaves the core in it.5 initial.subcritical condition.

d. The failure of an experiment involving the irradiation of up to 25 millig ams of properly contained explosive material in a reactor irradiation facility will not result.in damage to the reactor or the reactor pool containment structure.

! This specification is also intended to prevent damage to vital equipment by restricting the quantity of explosive materials to less than 250 milligrams within the reactor building.

e. This specification is intenfed to reduce the likelihood that airborne activities in exces s of the limits of Appendix B Table II of 10 CFR Part 20 will be eleased to the atmosphere outside the facility boundary.

f. The 5 mci limitation on I-131 through I-135 assures that in the event of failure of a fueled experiment, the exposure dose at the exclusion area boundary will be less than that postulated

for the maximum hypothetical accident (Section IX.E, SAR) even if the iodine was released in the air.

g. Operation of the reactor with the reactor fuel or structure damaged is prohibited to avoid release of fission products.

r Amendment No. 23

--,_,,---,--------,x - . - - . . - - - , -- - - , - , -

g . . - - , - ,

29 4

4.0 SURVEILLANCE REQUIREMENTS 4

4.1 REACTOR PARAMETERS 4.1.1 REACTOR POWER CALIBRATION Applicability This specification applies to the surveillance of the reactor power calibration.

Objective The objective is to verify the performance and operability of the power measuring channel.

Specifications A thermal power channel calibration shall be made on the linear power

. level monitoring channel annually, not to exceed 15 months.

Basis i

The thermal power level channel calibration will assure that the reacter is to be operated at the authorized power levels.

4.1.2 REACTOR EXCESS REACTIVITY Applicability f

This specification applies to surveillance of core excess reactivity.

Objective The objective is to assure that the reactor excess reactivity does not exceed the Technical Specifications and the limit analyzed in Section IX.F. SAR.

i Specification The excess reactivity of the core shall be measured annually, not to exceed 15 months, and following core or control rod changes equal to or greater than 0.75 Ak/k (-$1.00).

Basis Excess reactivity measurements on this schedule assure that no unexpected l changes have occurred in the. core and the core configuration does not exceed excess reactivity limits established in the Technical J

Specifications.

I I

Amendment No. 23 l

30 4.1. 3 TRIGA FUEL ELEMENTS Applicability This specification applies to the surveillance requirements for the TRIGA fuel elements.

Objective The objective is to verify the continuing integrity of the fuel element cladding.

Specifications All fuel elements and control rods with fuel followers shall be inspected visually for damage or deterioration and measured for length and bend before being placed in the core for the first time and at intervals not to exceed the sum of 3,500 dollars in pulse reactivity or two years , not to exceed 30 months, whichever comes first.

Bas,i s,

~

The frequency of inspection and measurement schedule is based on the parameters most likely to affect the fuel cladding of a pulsing reactor operated at moderate pulsing levels and utilizing fuel elements whose characteristics are well known.

4.2 REACTOR CONTROL AND SAFETY SYSTEM 4.2.1 REACTIVITY WORTH Applicability This specification applies to the reactivity worth of the control rods.

Objective The objective is to assure that the control rods are capable of -

maintaining the reactor suberitical.

I Specification The reactivity worth of each control rod and the shutdown margin for the core loading in use shall be determined annually, not to exceed 15 months, or following core or control rod changes equal to or greater than 0.7% ak/k (-$1.00) .

Basis The reactivity worth of the control rod is measured to assure that the required shutdown margin is available and to provide an accurate means

! for determining the core excess reactivity, maximum reactivity, Amendment No. 23

31 insertion rates, and the reactivity worth of experiments inserted in the core.

4.2.2 REACTIVITY INSERTION RATE Applicability This specification applies to control rod movement speed.

Objective -

The objective is to assure that the reactivity addition rate specification is not violated and that the control rod drives are f unctioning .

Specification The rod drive speed both up and down and the time from scram initiation to the . full insertion of any control rod from the full up position shall be measured annually, not to exceed 15 months, or when any significant work is done on the rod drive or the rod.

Basis This specification assures that the reactor will be promptly shut down when a scram signal is initiated. Experience and analysis have indicated that for the range of transients anticipated for a TRIGA

" reactor, the specified scram time is adequate to assure the safety of the reactor. It also assures that the maximum reactivity addition rate specification will not be exceeded.

4.2 3 REACTOR SAFETY AND CONTROL SYSTEMS Applicability These specifications apply to the surveillance requirements for measurements, channel tests, and channel checks of the reactor safety systems.

~

Objective The objective is to verify the performance and operability of the systems and components that are directly related to reactor safety.

Specifications -

a. A channel test of the scram f unction of the high power, fuel temperature, manual, and present timer saf ety channels shall be s

made on each day that the reactor is to be operated, or prior to each operation that extends more than one day.

b. A channel test of the detector power supply scram function shall be performed annually, not to exceed 15 months.

Amendment No. 23

32

c. Channel checks for operability shall be performed daily on fuel element temperature, linear power, count rate, log power and reactor period when the reactor is to be operated, or prior to each operation that extends more than one day.

d. The percent power channel shall be compared with other independent channels for proper channel .dication, when appropriate, each time the reactor is operated.

e. The pulse peak power channel shall be compared to the fuel temperature each time the reactor is pulsed, to assure proper peak power channel operation.

Basis TRIGA system components have proven operational reliability. Daily channel tests insure accurate scram functions and insure the detection of possible channel drif t or other possible deterioration of operating characteristics. The channel checks will make information available to the cperator to assure safe operation on a daily basis or prior to an extended run. An annual channel test of the detector power supply scen: will assure that this system works, based on past experience as recorded in the operation log book. Comparison of the percent power channel with other independent power channels will assure the detection of channel drift or other possible deterioration of its operational "

characteristics. Comparison of the peak pulse power to the fuel temperature for each pulse will assure the detection of possible channel drif t or deterioration of its operational characteristics.

4.2.4 REACTOR IKTERLOCKS Applicability This specification applies to the surveillance requirements for the reactor control system interlocks.

Objective The objective is to insure performance and operability of the reactor (ontrol system interlocks. .

Specifications

a. A channel check of the source interlock shall be performed each day that the reactor is operated or prior to each operation that extends more than one day.

b. A channel test shall be performed semi-annually, not to exceed 7V, months, on the log power interlock which prevents pulsing from power levels higher than one kilowatt.

Amendment No. 23 7

33 c.

A channel check shall be performed semi-annually, not to exceed Thi months, on the transient rod interlock which prevents application of air to ,the transient rod unless the cylinder is f ully inserted.

d.

A channel check shall be performed semi-annually, not to exceed 7hi months, on the rod drive interlock which prevents movement of any rod except the transient rod in pulse mode.

e. A channel check shall be performed semi-annually, not to exceed 7%,

months, on the rod drive interlock which prevents simultaneous manual withdrawal of more than one rod.

Basis

,The chant.el test and checks will verify operation of the reactor interlock system'. Experience at the PSBR indicates that the prescribed frequency is adequate to insure operability.

4.2.5 overpower Scram Applicability This specification applies to the high power and fuel temperature scram channels.

Objective The objective is to verify that high power and fuel temperature scram channels perform the scram functions.

Specification The high power and fuel temperature scrams shall be tested annually, not to exceed 15 months.

Basis '

Experienc.e with the Penn State TRIGA for more than a decade, as

' recorded in the operation log books, indicates that this interval is adequate to assure operability.

4.2.6 Transient Rod Test Applicability This specification applies to surveillance of the transient rod mechanism.

Objective The objective is to assure that the transient rod drive mechanism is maintained in an operable condition.

Amendment No. 23

34 1

Specification

a. On each day that pulse mode operation of the reactor is planned, a functional performance check of the transient (pulse) rod system shall'be performed. The transient (pulse) rod drive cylinder and the associated air supply system shall be inspected, cleaned, and lubricated as necessary annually, not to exceed 15 months,

b. The reactor shall be pulsed annually, not to exceed 15 months, to compare

" fuel temperature measurements and peak power levels with those of previous pulses of che same reactivity value or the reactor shall not be pulsed until such comparative pulse measurements are performed.

Basis Functional checks along with periodic maintenance assure repeatable performance. The reactor, is pulsed at suitable intervals and a comparison-made with previous similar pulses to determine if changes in fuel or core characteristics are taking place.

4.3 COOLANT SYSTEM 4.3 1 Fire Hose Inspection Applicability This specification applies to the dedicated fire hoses used to supply water to the pool in an emergency.

Objective The objectiv,e is to assure that these hoses are operable.

Specification .

The two (2) dedicated fire hoses that provide supply water to the pool in an

, emergency shall be visually inspected for damage and wear annually, not to exceed 15 months.

Basis This frequency is adequate to assure that significant degradation has not l

occurred since the previous inspection.

4.3.2 Pool Water Temperature Applicability This specification applies to pool water temperature.

Objective The objective is to limit pool water temperature.

Amendment No. 23

35 Specification The pool temperature alarm shall be calibrated annually, not to exceed 15 months.

Basis Experience has shown this instrument to be drift-free and that this interval is adequate to assure operability.

4.3 3 Pool Water Conductivity Applicability This specification applies to surveillance of pool water conductivity.

Objective The objective is to assure that pool water mineral content is maintained at an acceptable level.

Specification Pool water conductivity shall be measured and recorded daily when the reactor is to be operated, or at monthly intervals when the reactor is shut down for this time period.

Basis Based on experience, observation at these intervals provides acceptable surveillance of limits that assure that fuel clad corrosion and neutron activation of dissolved materials will not occur.

4 3.4 Pool Water Level Alarm Applicability This specification applies to the surveillance requirements for the pool level alarm .

Obj ective j The objective is to verify the operability of the pool-water level alarm.

Specification The pool-water level alarm shall be channel checked conthly, not to exceed 6 weeks, to assure its operability.

Basis Experience, as exhibited by past periodic checks, has shown that monthly checks of the pool-water level alarm assures operability of the system during the month.

Amendment No. 23

+ - _

, -, h,

36 4.4 CONFINEMENT 4.4.1 REACTOR BAY DOORS Applicability This specification applies to reactor bay doors.

Objective The objective is to assure that reactor bay doors are kept closed as per specification 3.4.

Specification Doors to the reactor bay shall be locked or located within view of the receptionist and/or operator when the reactor is operating.

Basis Experience has indicated that with the current degree of surveillance, the doors in question are maintained such as to meet specification 3.4 4.5 FACILITY EXHAUST SYSTEM AND EMERGENCY EXHAUST SYSTEM Applicability This specification applies to the facility exhaust system and emergency exhaust system.

Obj ective The objective is to assure the proper operation of the facility exhaust system and emergency exhaust system in controlling releases of

- radioactive material to the uncontrolled environment.

~

Specification

a. It shall be verified monthly, not-to exceed 6 weeks, whenever operation is scheduled, that the emergency exhaust system is operable with correct pressure drops across the filters (as specified in procedures).

b. It shall be verified monthly, not to exceed 6 weeks, whenever operation is scheduled, that the facility exhaust system closes when the emergency exhaust system operates.

Basis Experience, based on periodic checks performed over a few years of operation, has demonstrated that a test of the exhaust systems on a monthly basis, not to exceed 6 weeks, is sufficient to assure the Amendment No. 23

37 proper operation of the systems. This provides reasonable assurance on the control of the release of radioactive material.

4.6 RADIATION MONITORING SYSTEM AND EFFLUENTS 4.6.1 RADIATION MONITORING SYSTEM AND EVACUATION ALARM Applicability This specification applies to surveillance requirements for the area radiation monitor, the beashole laboratcry radiation monitor, the air radiation monitor, and the evacuation alara.

Obj ective The objective is to assure that the radiation monitors and evacuation alarm are operable and to verify the appropriate alara settings.

Specification The area radiation monitor, the beashole laboratory radiation monitor, the continuous air (radiation) sonitor, and the evacuation alara system shall be channel tested monthly not to exceed 6 weeks. They shall be verified to be operable by a channel check daily when the reactor is to be operated, and shall be calibrated annually, not to exceed 15 months.

2 Basis i

Experience has shown this frequency of verification of the radiation monitor set points and operability and the evacuation alara operability is adequate to correct for any variation in the system due to a change of operating characteristics. Annual channel calibration insures that units are within the specifications defined by procedures.

4.6.2 Argon-41 Applicability This specification applies to surveillance of the argon-41 produced during reactor operation.

Objective To assure that the, production of argon-41 does not exceed the limits specified by 10 CFR 20.

Specification

.The production of argon-41 shall De seasured and/or calculated for each new experiment that is eitimated to generate sufficient argon .41 to produce more than 2x10-1 v microcuries/mi at the exclusion boundary.

Amendment No. 23

-..v...e., - - ,. - - - - -, , - - , . -- ,-n----,--.r-- . - , . . . < - ~ . . -

38 Basis f

Measurements have been made (Environmental Impact Appraisal, page 3) that show for the past several years the annual release of argon-41 from the PSBR operation is approximately 600 mC1. The calculated dose at the exclusion area boundary is estimated to be less than 1 mrem /yr due to the PSBR operation.

1 4.6 3 ALARA Applicability This specification applies to the surveillance of all reactor operations that could result in occupational exposures to radiation or the release of radioactive effluents to the environs.

Objective -

The objective is to provide surveillance of all operations that could lead to occupational exposures to radiation or the release of radioactive effluents to the environs.

Specification

, As part of the review of all operations, consideration shall be given to alternative operational modes that might reduce staff exposures, release of radioactive materials to the environment, or both.

Basis

, Experience has shown that experiments and operational requirements can,

in many cases, be satisfied with a variety of combinations of facility ,

options, core positions, power levels, time delays, and effluent or staff radiation exposures. Similarly, overall reactor scheduling achieves significant reductions in ' staff exposures. Consequently, ALARA must be a part of both overall reactor scheduling and the detailed experiment planning.

_ 4.7 EXPERIMENTS l

Applicability l This specification applies to surveillance requirements for experiments.

Obj ective 4

i The objective is to assure that the conditions and restrictions of l 3 Specification 3.7 are met.

] Amendment No. 23

. 39 Specification Those conditions and restrictions listed in specification 3.7 shall l i

be considered by the PSBR authorized reviewer before signing the irradiation authorization for each experiment.

Basis This specification has proven to prevent performance of unacceptable experiments in the past. Authorized reviewers are appointed by the facility director.

1 5.0 DESIGN FEATURES 5.1 REACTOR FUEL' Specifications The individual unirradiated TRIGA fuel elements shall have the following characteristics:

a. The uranium content shall be a maximum cl either 9.0 wt5 or 12.7 wt$ enriched to less than 20% uranium-235.

4

b. The hydrogen-to-zirconium atom ratio (in the ZrHx) shall be a nominal 1.65 H atoms to 1.0 Zr atom.

t c. The cladding shall be 304 stainless steel with a nominal 0.020 inch thickness.

5.2 REACTOR CORE Specifications '

a. The core shall be an arrangement of TRIGA uranium-zirconium hydride f uel-moderator elements positioned in the reactor griC i plates.

b. The reflector, excluding experiments and experimental facilities, shall be water, or D 0, 2 or graphite, or any combination of the three moderator materials.

l 53 CONTROL RODS i

Specification

a. The shim, sa'fety, and regulating control rods shall. have scram capability and contain borated graphite B C4 powder, or boron and its compounds in solid form as a poison in stainless steel i .-

or aluminum cladding. These rods may incorporate fueled followers i

which Amendment No. 23

, , . - - . , - , . -,%__,_--.-_-.,,,~-,y,,.y- , ,,y-v ,--y-v_-, ,.-- - , - - , - - , , - _ _ , - - _ . , ,,_~-.,..,,,...~,-,_-o

40 have the same characteristics as the fuel region in which they are used,

b. The transient control rod shall have scram capability and contain borated graphite, B C4 powder, or boron and its compounds in a solid form as a poison in an aluminum or stainless steel clad. When used as a transient rod, it shall have an adjustable upper limit to allow a variation of reactivity insertions. This rod may incorporate a voided or a solid aluminum follower.

5.4 FUEL STORAGE Specifications

a. All fuel elements shall be stored in a geometrical array where the ke pt is less than 0.8 for all conditions of moderation.

b. Irradiated fuel elements shall be stored in an array which shall permit sufficient natural convection cooling by water such that the fuel element temperature shall not reach the safety limit as -

defined in Section 2.1 of the Technical Specifications.

5.5 REACTOR BAY AND EXHAUST SYSTEMS

, Specifications

a. The reactor shall be housed in a room (reactor bay) designed to restrict leakage. The minimum free volume in the reactor bay shall be 2500 m3

b. The reactor bay shall be equipped with two exhaust systems.

Under normal operating conditions, the f acility exhaust system exhausts unfiltered reactor bay air to the environment releasing,

' it at a point at least 24 feet above ground level. Upon initiation of a building evacuation alarm, the previously mentioned system is automatically secured and an emergency exhaust system automatically starts. The emergency exhaust system is designed to discharge reactor bay air at a point at least 24 feet above ground level after passing it through a 3-stage filter system.

5.6 REACTOR PCOL WATER SYSTEMS Specifications The reactor core shall be cooled by natural convective water flow.

1

. Amendment No. 23

41 I

6.0 ADMINISTRATIVE CONTROLS 6.1 ORGANIZATION 6.1.1 STRUCTURE The University Vice President for Research and Dean of the Graduate School (level 1) has the responsibility for the reactor i

facility license. The management of the facility is the responsibility of the Director and the Deputy Director (level 2), who report to the Vice President for Research and Dean of the Graduate School through the Head of the Nuclear Engineering Department and the Dean of the College of Engineering.

Administrative and fiscal responsibility is within the offices of the Department Head and the Dean.

The minimum qualifications for the position of Director of the PSBR are an advanced degree in science or engineering, and 2 years experience in reactor operation. Five years of experience '

directing reactor operations may be substituted for an advanced degree .

The Director can at any time temporarily delegate his authority to the Deputy Director who can in-turn f urther delegate his authority to a qualified Senior Reactor Operator (level 3).

The reactor operators (level 4) report to the Senior Reactor Operator (level 3) for operational matters.

The University Health Physicist reports directly to the Office of the Vice President for Research and Dean of the Graduate I

School. The qualifications for the University Health Physicist position are the equivalent of a graduate degree in radiation protection, 3 to 5 years experience with a broad byproduct j material license, and certification by The American Board of l

Health Physics or eligibility for certification.

I

, 6.1.2 RESPONSIBILITY l

Responsibility for the safe operation of the reactor facility shall be within the chain of command shown in the organization chart. Individuals at the various management levels, in addition to having responsibility for the policies and operation of the reactor f acility, shall be responsible for safeguarding i

the public and facility personnel from undue radiation exposures and for adhering to all requirements of the operating license and technical specifications.

i In all instances, responsibilities of one level may be assumed l

i by designated alternates or by higher levels, conditional upon appecpriate qualifications.

1 i .

i Amendment No.-23 O ,,

,, _ _ -, -e, , , . - - - - ~ - - - < - - - 'e -

• ' ' ' ' ' ' ~ ^ " * " ~ " " ' ' ~

42 .

Vice President for Research and Dean of the Graduate School (Level 1)

Dean, College of Engineering Nuclear Engineering Department Head

• University Health Penn State Reactor Physicist Safeguards Committee

! l 1 l Director

_ _ _ . _ _ . _ _ Deputy Director Penn State Breazeale Reactor (Level 2) l Senior Reactor Operator (Level 3)

Operating Staff Reactor Operators (Level 4)

ORGANIZATION CHART Amendgent No. 23 m ,

43 6.1 3 STAFFING

a. The minimum staffing when the reactor is not secured shall be:

(1) A licensed operator present in the control room, in 1

accordance with applicable regulations .

(2) A second person present at the facility able to carry out prescribed written instructions.

(3) If a senior reactor operator is not present at the f acility, one shall be available by telephone and able to be at the facility within 30 minutes.

b.

A list of reactor facility personnel by name and telephone number shall be readily available in the control room for use by the operator. The list shall include:

(1) Management personnel.

(2) Radiation safety personnel.

2 (3) Other operations personnel.

1

_ c. Events requiring the direction of a Senior Reactor 1

Operator shall ir.clude:

._ (1) All fuel or control-rod relocations .within the reactor core region.

(2) Relocation of any in-core experiment with a reactivity worth greater than one dollar.

(3) Recovery from m1 planned or unscheduled shutdown (in this instance, documented verbal concurrence from a Senior Reactor Operator is required).

6.1.4 SELECTION AND TRAINING OF PERSONNEL The selection, training, and requalification of operations -

personnel shall meet or exceed the requirements of all applicable regulations and the American National Standard for

. Selection and Training of Personnel for Research Reactors, i ANSI / ANS-15. 4-1977, Sections 4-6.

6.2 REVIEW AND AUDIT .

6.2.1 _ SAFEGUARDS COMMITTEE CCMPOSITION A Penn State Reactor Safeguards Committee (PSRSC) shall exist to j

^ provide an independent review and audit of the safety aspects of reactor f acility operations . The committee shall have a minimum of 5 members and shall collectively represent a broad spectrum

' Amendment No. 23

44 '

of expertise in reactor technology and other science and engineering fields. The committee shall have at least one member with health physics expertise. The committee shall be appointed by and report to the Dean of the College of Engineering. The PSBR Director shall be an ex-officio member of the PSRSC.

6.2.2 CHARTER AND RULES The operations of the PSRSC shall be in accordance with a written charter, including provisions for:

a. Meeting frequency - not less than once per calendar year not to exceed 15 months.

b .* Quorums - at least one-half of the voting membership shall b'e present (the Director who is ex-officio shall not vote) and no more than one-half of the voting members present shall be members of the reactor. staff.

c. Use of Subgroups - the committee chairman can appoint ad-Hoc committees as deemed necessary,

d. Minutes of the meetings - shall be recorded, disseminated, reviewed, and approved in a timely manner.

6.2 3 REVIEW FUNCTION The following items shall be reviewed:

a .- Evaluation of proposed changes in equipment, systems, tests, or experiments if they involve an unr3 viewed safety question.

b. All new procedures and major revisions thereto having a significant effect upon safety.

c. All new experiments or classes of experiments that coul'd have a significant effect upon reactivity or upon the release of radioactivity,

d. Proposed changes in technical specifications, license, or charter.

e. Violations of technical specifications, license, or charter.

Violations of internal procedures or instructions having safety significance,

f. Operating abnormalities having safety significance,

g. Reportable occurrences listed in 6.6.2.

h. Audit reports.

Amendment No. 23

45 6.2.4 AUDIT The audit function shall be performed annually, not to exceed 15 months, preferably by a non-member of the reactor staff. The audit function shall be performed by a person not directly involved with the function being audited. The audit f unction shall include selective (but comprehensive) examinations of operating records, logs, and other documents. Discussions with operating personnel and observation of operations should also be used as appropriate. Deficiencies uncovered that affect reactor safety shall promptly be reported to the Nuclear Engineering Department Head and the Dean of the College of Engineering. The following items shall be audited:

a. Facility operations for conformance to Techn! 21 Specifications, license, and procedures (at least once per calendar year with interval not to exceed 15 months).

b. The requalification program for the operating staff (at least once every other calendar year with the interval not to exceed 30 months).

c. The results of action taken to correct deficiencies that may occur in the reactor facility equipment, systems, structures, or methods of operations that affect reactor

, safety (at least once per calendar, year with the interval i

not to exceed 15 months).

i

d. The reactor facility emergency plan and implementing procedures (at least once every other calendar year with the interval not to exceed 30 months).

6.3 OPERATING PROCEDURES .

Written procedures shall be reviewed and approved prior to th'e initiation of activities covered by themi in accordance with section 6.2.3. The procedures in this section preceded by an asterisk shall also 5e approved and initialed by a representative of the University Health Physics Office. Written procedures shall be adequate to assu're the safe operation of the reactor, but shall

! not preclude the use of independent judgment and action should the l situation require such. Operating procedures shall be in effect i

I and shall be followed for at least the following items:

a. Startup, operation, and shutdown of the reactor.

i

b. Core loading, unloading, and fuel movement within the reactor.

! c. Routine maintenance of major components of systems that j oculd have an effect on reactor safety.

I l

Amendment No. 23

46 - -

d. Surveillance tests and calibrations required by the -

technical specifications (including daily checkout procedure).

e. Radiation, evacuation, and alsrm checks.

• f. Release of Irradiated Samples.

• g. Evacuation.

'h. Fire or Explosion.

• i . Oaseous Release.

'j . Medical Emergencies.

• k. Civil- Disorder.

• 1. Bomb Threat.

• m. Threat of Thef t of Special Nuclear Material.

'n. Thef t of Special Nuclear Material.

'o. Industrial Sabotage.

'p . Experiment Evaluation and Authorization.

'q. Reactor Operation Using a Beam Port.

• r. D 20 Handling.

's. Healtt. Physics Orientation Requirements.

't. Hot Cell Entry Procedure. *

'u. Implementation of emergency and security plans.

v. Radiation instrument calibration

w. Loss of pool water.

6.4 REVIEW AND APPROVAL OF EXPERIMENTS

a. All new experiments which may present unreviewed safety questions shall be reviewed by the PSRSC and approved in writing by level 2 management or designated alternate prior to initiation.

b. Substantive changes tc experiments previously approved by the PSRSC shall be mace only af ter review and approval in writing by level 2 management or designated alternate.

Amendment No. 23

47 6.5 REQUIRED ACTION 6.5.1 ACTION TO BE TAKEN IN THE EVENT THE SAFETY LIMIT IS EXCEEDED In the event the safety limit (1150ac) is exceeded:

a. The reactor shall be shut down and reactor operation shall not be resumed until authorized by the U.S. Nuclear Regulatory Commission.

b. The safety limit violation shall be promptly reported to level 2 or designated alternates,

c. An immediate report of the occurrence shall be made to the Chairman, PSRSC and reports shall be made to the USNRC in accordance with Section 6.6 of these specifications,

d. A report shall be prepared which shall include an analysis of the causes and extent of possible resultant da= age, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence.

This report shall be submitted to the PSRSC for review.

6.5.2 ' ACTION TO BE TAKEN IN THE EVENT OF A REPORTABLE OCCURRENCE In the event of a reportable occurrence, (1.1.33) the following action shall be taken:

a. The reactor shall be returned to normal or shutdown. If it is necessary to shutdown the reactor to correct the occurrence, operations shall not be resumed unless

. authorized by level 2 or designated alternates,

b. The Director or a designated alternate shall be notified and corrective action taken with respect to the operations involved.

c. The Director or a designated alternate shall notify the Nuclear Engineering Department Head who, in turn, will notify the office of the Dean of the College of Engineering

• and the office of the Vice President for Research and  :

Dean of the Graduate School.

d. The Director or a designated alternate shall notif y the Chairman of the PSRSC.

e. A report shall be made to the PSRSC which shall include an analysis of the cause of the occurrence, efficacy of corrective action, and recommendations for measures to prevent cr reduce the probability of recurrence. This report shall be reviewed by the PSRSC at their next meeting.

Amendment No. 23

. 48 .

f. A report shall be made to the USNRC Director, Office of Nuclear Reactor Regulation, with a copy to the Regional Adminis trator .

6.6 REPORTS 6.6.1 OPERATING REPORTS An annual report shall be submitted within 6 months of the end of The Pennsylvania State University fiscal year te the USNRC Director, Office of Nuclear Reactor Regulation, with a copy to the Regional Administrator, including at least the fcllowing

items: e l a. A narrative summary of reactor operating experience l including the energy produced by the reactor, and the nunber of pulses > -$2.00 but less than or equal to $2.50 and the j number greater than $2.50.

b. The unscheduled shutdowns and reasons for them including, ,

where applicable, corrective action taken to preclude recurrence.

c. Tabulation of major preventive and corrective maintenance operations having safety significance.

d. Tabulation of major changes .in the reactor f acility and procedures, and tabulation of new tests and experiments, that are significantly different from those performed previously and are not described in the Safety Analysis 1

Report, including a summary of the analyses leading to the ,

conclusions that no unreviewed safety questions were .

j involved and that 10 CFR 50.59 was app'licable.

e. A summary of the nature and amount of radioactive effluents released or discharged to environs beyond the effective control of the owner-operator as determined at or before the point of such release or discharge. The summary shall include to the extent practicable an estimate of individual radionuclides present in the effluent. If the estimated average release after dilution or diffusion is less than 25 percent of the concentration allowed or recommended, only a statement to this effect need be presented.

a

f. A summarized result of environmental surveys performed outside the facility.

j g. A summary of exposures received by f acility personnel and

, visitors, in the form indicated in 10 CFR 20.407(b), where i such exposures are greater than 25 percent of that allowed 4

or recommended in 10 CFR 20.

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Agendment No. 23 J

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49 6.6.2 SPECIAL REPORTS i

Special reports are used to report unplanned events as well as planned major facility and administrative changes. These special reports shall contain and shall be communicated as follows:

a. There shall be a report no later than the following wor' king

day by telephone and confirmed in writing by telegraph or i

similar conveyance to the USNRC Regional Administrator to be followed by a written report to the offices given in 6.6.1 that describes the circumstances of the event within 14 days of any of the following:

(1) Violation of safety limits (See 6.5.1)

(2) Release of radioactivity from the site above allowed limits (See 6.5.2)

(3) A reportable Occurrence ( Section 1.1.33)

' .b . A written report shall be made within 30 days to the USNRC, and to the offices given in 6.6.1 for the following:

(1) Permanent changes in the facility organization involving level 1-2 personnel.

(2) Significant changes in the transient or accident analysis as described in the Safety Analysis Report.

6.7 RECORDS l To fulfill the requirements of applicable regulations, records and logs shall be prepared of at least the following items and

, retained:

6.7.1 RECORDS TO BE RETAINED FOR AT LEAST FIVE YE.*RS

a. Log of reactor' operation, and sumar.cy of ensrgy produced or hours the reactoc was critical.

b. Checks and calibrations procedure file.

c. Preventive and corrective electronic maintenance log.

d. Major changes in the reactor facility and procedures.

e. Experiment authorization file including conclusions that no unreviewed safety questions were involved for 4

new tests or experiments.

f. Event evaluation forms (including unscheduled shutdowns), and reportable occurrence reports.

Amendment No. 23

50 ,

g.

Preventive and corrective maintenance records of associated reactor equipment.

h. Facility radiation and contamination surveys.

1. Fuel inventories and transfers.

J. Surveillance activities as required by the Technical Specifications .

k. Records of PSRSC reviews and audits.

6.7.2 RECORDS TO BE RETAINED FOR AT LEAST ONE TRAINING CYCLE

a. Requalification records for licensed reactor operators and senior reactor operators.

6.7 3 RECORDS TO BE RETAINED FOR THE LIFE OF THE REACTOR FACILITY

a. Radiation exposure for all facility personnel and visitors. '

b. Environmental surveys performed outside the facility.

c. Radioactive affluents released to the environs,

d. Drawings of the reae.ar facility including changes.

4 J

e i

Amendment No. 23. .

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