ML20064K028

Proposed Tech Specs for Univ of Ma Lowell Reactor W/Low Enrichment Fuel,Correcting Editorial Oversight & Providing Clarification as Result of Discussions W/Nrc

ML20064K028
Person / Time
Site:	University of Lowell
Issue date:	02/28/1994
From:	MASSACHUSETTS, UNIV. OF, LOWELL, MA (FORMERLY LOWELL
To:
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NUDOCS 9403220177
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j PROPOSED 1

TECHNICAI SPECEICATIONS FOR THE )

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UNIVERSITY OF MASSACHUSETFS LOWELL REACTOR {

l WITH LOW ENRICHMENT FUEL l

February, 1994 1

9403220177 940317 PDR ADOCK 05000223 P PDR

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i TABLEOFCONTENTS l 1.0 D E FINITI O N S . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . .. . . . . .... .... .. 1. . . . . . . .

1.1 AB NORMAL OCCURRENCES ............................. ................... ..1 1.2 CHANNEL................................................................................. .I 1.3 CH ANNEL CA LI B R ATION .. .. ............ .... ................ .. ........... .. ..... .

1.4 CHANNEL CH ECK ... . . .... . .. . . . . . . . ....... .. .. ... . .. ... .... ...... .... ... .. . . ... ... ... ... 2 1.5 CHA NNEL TEST . . .. .. . . . .. .. .. .. . . . .. . . .... .. ..... . . .. .. .... . .. . . . .. . .. .. . . .. ... .. .. . 2 1.6 CONTAINMENT BUILDING INTEGRITY ..... .. ......................... 2 1.7 CONIR OL RO D . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . .. . . . . . . . . . . . . . . . .. . . . . .. . . . . . . . . .. . . .. . . . .. . . . . . 2

, 1.8 EXCE S S REA GI VITY . .. . .. .. .. . . .. . . .. .. . .. ... .. .. .... .... .. ... ... .. . . .. .. .. .. .. .. . . .. . 3 1.9 EXP E R IM ENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.10 M EA S URED VA LUE .. .. . .. .. . . ... . . . ... . .. . . . . .. . .. .. . .. . . . .. . .. .. .. .. . .. .. . . .. .. .. . . .3 1.11 M OVA B LE EX PER IM ENT .... .. ... .... . .... . .... . ..... . ........ ............. .. 3 1.12 OPERABLE..................................................................................3 1.13 O P E R ATI N G . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .. . 3

1. I 4 P R OTEGIVE CHANNEL . .. .. . . . . .. .. .. .. . . . . . . .. .. ... . .. ... . . . .. .. .. ... . .. . . .. .. . . .. . . . 4 1.15 REACTOR OPER ATING MODE ......... .... ....................................... 4 .

1.16 R EAGO R OPE RATING .... .. .. . . .. .. .. .. .. . . .. .. .. .. . . . . . .. .. .. ... . . .. . . .. .. .. . . ... . . . 4 1.17 REAGOR S AFETY S YSTEM .. ........................... ............ .. ........... 4 1.18 REAGOR SECURED ..... .. . ...... ............................................4

1. I 9 REAGO R S HUTDOWN . . ... ... .. .. .. . . . ..... . ... .. .. .. ..... .. ... ... ... .. .. . ... ... .. 5 1.20 REFERENCE CORE CONDITION ..... . ..... .. ..... .............................. 5 1.21 R EG ULATING RO D .. .. .. .. . . .. . .... ... .. . .. .. . .. .. . .... .... .. .... . . . .. .. .... .... . .. .. .. . 5 1.22 S AFETY CHA NNEL ... ... . .. . ... . . . .. .... .. ... .. . ... .... .. ... . .. .. .. ... .. .. . . . . .. . . .. .. . . .. . . 5 1.23 S ECURED EXPERIM ENT ... ................... .......... ........... . ..... ..... ... ........ 6 1.24 S HALL, S HOULD, AND M A Y ..... ............. .......... .... .... .................. 6 1.25 S H UTDO WN MA R G I N .. . . . .. .. . .. .. .. . .. . .. . . . .. .. .. .... ...... . .. .. .. . .. . .... . .... . .. .. . 6 1.26 S URVEILLANCE INTERVALS ............... ...................................... 6 1.27 TR U E V ALU E . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . .. . . . .. . 7 2.0 SAFETY LIMITS AND LIMITING S AFETY SYSTEM ... ....................

SE'ITINGS.......................................................................................................8 2.I S A FET Y LIM I TS . . . .. . . .. .. .. .. . . . ... .. .. .. . . . . .. ... .. .. .. .. . ... . . .. .... .. .. .. . . .. . ... .. ... .... . 8 2.1.1 Safety limits in the forced convection mode of o p e r a t i o n . . . . . .. . . . . . . . . . . . . . . . . . . . . . .. . . .. . . . . . . . . . . . .. . . . . .. . . . . . . . . . . . . .. . . . . . . . . . . 8 l

2.1.2 Safety Limits in the natural convection mode )

o f ope ra t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 2.2 LIMITING S AFETY SYSTEM SE'ITINGS......................... ..........10 '

2.2.1 Limiting Safety System Settings in the forced convection mode of operat10n........................................1 0 2.2.2 Limiting Safety System Settings in the natural convection flow mode of operation. ............................12 -

3.0 LIMITING CONDITIONS FOR OPERATION............. ... ..... ....................I 4

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1 3.1 RE A CTIV ITY. .. . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . 1i 3.2' REACTOR INSTRUMENTATION ... .................................. .... ......1 8  !

3.4 R ADIATION MONITORING EQUIPMENT........................... ...... 2 2 3.5 CONTAINMENT AND EMERGENCY EXHAUST SYSTEM ....... 2 4 3.6 LIMITATIONS OF EXPERIMENTS.................................... .......... 2 6 3.7 G A S EOU S EFFLUENTS . . . . .. .. . .. ... . .. . . . . ... .. . .. . . . . . ..... .. . .. .... . .. ... ... .. ... . .. . . 2 9 3.8 COOLANT SYSTEM.............. .....................................................29 4.0 SURVEILLANCE REQUIREMENTS .. ........... ..... . .................................... 3 1 4.1 CONTROL AND REGULATING RODS............. ................... .... ..... 3 1 4.2 REACTOR S AFETY SYSTEM ....... ...................... .... .. ............. ...... 3 2 4.3 R ADIATION MONITORING EQUIPMENT.................................. 3 4 4.4 CONTAIN MENT B UILDING....... ................ ..................... ....... ..... 3 5 4.5 PO O L WATE R . . .. . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . .. . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . 3 7 4.6 SCRAM BY PROCESS VARIABLE EFFECT. ........ ...................... 3 8 4.7 FUEL S U RVE IL LANCE .. . . . .. .. .. .. . . . . .. .. . . .. .. .. ... . .. . ... .. .. .. . . .. .. .. .. .... . . .. 3 9 5.0 DES I GN FEATURES . . . . .. .. . . .. . . . . . . .. . ... . . . .. .. ... .. .. . ... . .. . .. . . .. .. . . .. . .. .. . . . . . . ... .. .. . 4 0 5.1 REACTO R FUEL ... . .. . . .. .. . .. .. . ... .. ... . . .. . . .. ... . . .. . . . .. . ... .. ... ... ... . ... ... . 4 0 5.2 REA CTOR CO R E... .. . . ... ... . . ... . . .... .. .. ... . ... ... .... .. .. . .. .. ..... . . .... .... .. .. 4 0 5.3 REACTO R B UILDING ... .. .. .. . . .. . . . .. .. . .. . . .. .. . ... .. .. .. ...... .. ... . .. .. .... .. .. . . .. . 4 1 5.4 FUEL STORAGE . .. .. .... . . . . . . ... . . .. .. . . .. .. . . . ... .. .. . . .... .. . . . .. .. .. .. .. . . .. .. .. . .. .. .. . 4 1 6.0 A DMINISTR ATIVE CONTR OLS .... .................... ......................... ............ 4 2 6.1 ORGANIZATION AND MANAGEMENT. .................................... 4 2 6.2 R E V IEW AN D A UD IT ... .. .. . . .. . . .. .. .. .. . .. . .. .. . . . . .. .. .. .. .. .. . . .. .. . . .. .. .. . . . . . . . 4 3 6.3 OPERATING PROCEDURES.............................. ............ .................. 4 5 6.4 ACTION TO BE TAKEN IN THE EVENT OF AN .......................

ABNORMAL OCCURRENCE ...................... ............. .. ..................... 4 6 6.5 ACTION TO BE TAKEN IN THE EVENT A SAFETY ................

LIMIT IS EXCEEDED.. ... . .. . . . . . .. .. . . . . . .. .. . ... ... . . .. .. . . . . . .. .. .. .. . .. .. .. ... .. .. . . .. . . 4 7 6.6 REPORTING REQUIREMENTS ........................................................ 4 7 6.7 PLANT OPERATING RECORDS ...................................................... 5 2 6.8 APPROVAL OF EXPERIMENTS..... ................................. ............. 5 4

TECHNICAL SPECIFICATIONS - 1.0 DEFINITIONS l l

1.1 ABNORMAL OCCURRENCES -

An abnormal occurrence is any of the following: l

a. Any actual safety system setting less conservative than 1

specified in Paragraph 2.2 of these Technical Specifications;

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

c. Safety system component malfunction or other component I

or system malfunction which could, or threatens to, render 1 the system incapable of performing its intended function;

d. Release of fission products from a fuel element in a quantity that would indicate a fuel element cladding failure; q l

e. An uncontrolled or unanticipated change in reactivity greater than 0.5% delta k/k;

f. An observed inadequacy in the implementation of either administrative or . procedural controls, such that the .

inadequacy could have caused the existence or development l of an unsafe condition in connection with the operation of l the reactor;

g. Conditions arising from natural or offsite manmade events that affect or threaten to affect the safe operation of the facility.

1.2 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. Such a channel is also referred to as a l

measuring channel. It may or may not be a. safety channel.

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TECHNICAL SPECIFICATIONS 1.3 CHANNEL CALIBR ATION -

A channel calibration is an adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the parameter which the channel measures. Calibration shall l encompass the entire channel, including equipment actuation, alarm, or trip and shall be deemed to include a ' Channel Test.

1.4 CHANNEL CHECK -

A channel check is a qualitative verification of acceptable performance by observation of channel behavior. This verification, where possible, shall include comparison of the channel with other independent channels or systems measuring the same variable.

1.5 CHANNEL TEST -

A channel test is the introduction of a signal into the channel for verification that it is operable.

1.6 CONTAINMENT BUILDING INTEGRITY - 4 Integrity of the containment building is said to be maintained ,

1 when all isolation system equipment is operable or secured in an 1 i

isolating position. I J

1.7 CONTROL ROD - "

A control rod is a device fabricated from neutron absorbing material which is used to establish neutron flux changes and to .

compensate for routine reactivity losses. A control rod is coupled to its drive unit allowing it to perform a safety function when the coupling is disengaged.

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Q TECHNICAL SPECIFICATIONS i

1.8 EXCESS REACTIVITY - l l

. Excess reactivity is that amount of reactivity that would exist if all j 1

, control rods (control and regulating) were moved to the maximum '

l reactive condition from the point where the reactor is exactly critical (keff=1).

[ 1.9 EXPERIMENT -

An experiment is any operation, hardware, or target which is designed to investigate non-routine reactor characteristics or which is intended for irradiation within the pool, on or in a l beamport or irradiation facility and which is not rigidly secured to a core or shield structure so as to be a part of their design.

1.10 MEASURED VALUE -

1 j The measured value is the value of a parameter as it appears at the output of a channel.

1.11 MOVABLE EXPERIMENT -

A movable experiment is one in which the entire experiment may be moved into or out of the core or core region while the reactor is l

operating.

1.12 OPER ABLE -

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

1.13 OPERATING - ,

1 Operating means a component or system is performing its

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intended function.

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I TECHNICAL SPECIFICATIONS 1.14 PROTECTIVE CHANNEL -

A protective channel is a channel in the reactor safety system which is not merely a measuring channel.

1.15 REACTOR OPER ATING MODE -

Reactor operating mode refers to the method by which the core is 1

cooled, either natural convection mode of operation or forced l l

convection mode of operation. .

1.16 REACTOR OPERATING -

The reactor is operating whenever it'is not secured or shutdown.

1.17 REACTOR S AFETY SYSTEM -

The reactor safety system consists of those systems, including their associated input channels, which are designed to initiate automatic reactor protection or to provide information for~

initiation of manual protective action.

1.18 REACTOR SECURED -

The reactor is secured when: i I

(1) It contains insufficient fissile material or moderator present )

l in the reactor, adjacent experiments or control rods, to attain criticality under optimum available conditions of moderation and reflection, or (2) A combination of the following:

a. The minimum number of neutron absorbing control I.

l rods are fully inserted or other safety devices are in the shutdown position, as required by technical specifications, and TS-4

l TECHNICAL SPECIFICATIONS 1

1 E i L b. The console key switch is in the off position and the l l

I~ key is removed from the lock, and

c. No work is in progress involving core fuel, core l structure, installed control rods, or control rod drives unless l L

they are physically decoupled from the control rods, and L d. No experiments in or near the reactor are being moved t

I or serviced that have, on movement, a reactivity worth t

exceeding that maximum value allowed for a single i experiment or one dollar, whichever is smaller.

1.19 REACTOR SHUTDOWN -

1 The reactor is shut down if it is subcritical by at least 0.7% delta k/k in the Reference Core Condition plus the absolute reactivity worth of all experiments.

1.20 REFERENCE CORE CONDITION -

t The condition of the core when it is at ambient temperature (cold)

L and the reactivity worth of xenon is negligible <.2% delta k/k.

1.21 REGULATING ROD -

The regulating rod is a low worth control rod, used primarily to maintain an intended power level, that does not have scram capability. Its position may be varied manually or by the servo-controller.  ;

1.22 SAFETY CHANNEL -

A safety channel is a measuring or protective channel in the -

reactor safety system.

L TS-5

_ - _ _ _ = _ _ - _ _ _ - - _ _ - _ _ _ _ _ _ _ - __ -

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TECHNICAL SPECIFICATIONS a

1.23 SECURED EXPERIMENT -

A secured experiment is an experiment, experimental facility,. or ='

component of an experiment that is held in a stationary position.

relative to the reactor by mechanical means. The retaining .

~ devices must be able to withstand the . hydraulic, pneumatic, buoyant, or other forces which are . normal to the operating environment of the experiment, or forces which can arise as a result of credible malfunctions.

1.24 SHALL. SHOULD. AND MAY -

The word shall, is used to denote a requirement; the word should to denote a recornmendation; and the word may to denote permission, neither a requirement nor a recommendation, l 1.25 SHUTDOWN MARGIN - l J

Shutdown margin shall mean the minimum shutdown ' reactivity necessary to provide confidence that the reactor can be made subcritical by means of control and . safety systems . starting. from any permissible operating condition although the most reactive rod is in its most reactive position, and that the reactor will remain subcritical without further operational' action. ,

1.26 SURVEILLANCE INTERVALS - I The average over any extended period for each surveillance' time-interval shall be closer to the normal surveillance time ' th'an the -

extended time. Any extension of these intervals shall be:

occasional and for a . valid reason, and shall not affect the average as defined. Allowable surveillance intervals ~ shall not exceed the i following: 1 TS-6

TECHNICAL SPECIFICATIONS

a. Five year (interval not to exceed six years)

b. Two year (interval not to exceed two and one half years)

c. Annual (interval not to exceed 15 months)

d. Semi-annual (interval not to exceed seven and one half months)

e. Quarterly (interval not to exceed four months)

f. Monthly (interval not to exceed six weeks)

g. Weekly (interval not to exceed ten days)

h. Daily (must be done during the calendar day).

l.27 TRUE VALUE -

The true value is the actual value of a parameter.

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TECHNICAL SPECIFICATIONS 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 S AFETY LIMITS -

2.1.1 Safety limits in the forced convection mode of operation.

Applicability This specification applies to the interrelated variables -

associated with core thermal and hydraulic performance with forced convection flow. These variables are:

j. P = Reactor thermal power W= Reactor coolant flow rate Ti = Reactor coolant inlet temperature L = Height of water above the center line of the core Qbjective

-To assure that the integrity of the fuel cladding is maintained.

Specification l Under the conditions of forced convection flow: J l

1. The combination of true values of reactor thermal l power (P) and reactor coolant flow rate (W) shall not exceed the limits shown in Figure 2.1 T.S under any operating conditions. The limits are considered exceeded if the point defined by the true values of P q l

and W is at any time above the curve shown in Figure '

2.1 T.S.

2. The true value of the pool water level (L) shall not be ,

1 less than 24 feet above the center line of the core.

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4-3 2

.5 t

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' 3-2-

l 1-0 i , , , ,

0 500 1000 1500 2000 2500 3000 l

l Flow Rate (gal. / min.)

Figure 2.1 T.S. Power-Flow Safety Limit Curve TS-8A

- - _ -___ - . _-_- __ - _ _________-_________________________ _ -_ ______________ = _______-_-_____________-_________________

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TECHNICAL SPECIFICATIONS

3. The true value of the reactor coolant inlet temperature (pool temperature, T p). shall not be greater than 1100F.

Bases In the region of full power operation, the criterion used to establish the safety limit was the onset of nucleate boiling (ONB) at the hot spot in the hot channel.' The analysis is-f given in Section 3.1.2.2 of the FSAR Supplement for L

! Conversion to Low Enrichment Uranium (LEU) Fuel.

2.1.2 Safety Limits in the natural convection mode of operation.

!- Anolicability This specification applies to the. interrelated variables associated with core thermal and hydraulic performance with natural convection flow. These variables are:

-l P = ' Reactor thermal power ,

, l Tp = Reactor pool temperature L =- Height of water above the center line of the core Obiective To assure that the integrity of the fuel cladding is R

maintained.

i Soecification Under conditions of natural convection flow:

l. The true value of the reactor thermal power (P) shall not exceed 0.335 MW.

2. The true value of the reactor thermal power (P) shall not exceed 1.33 kW when the true value of the pool TS-9

i TECHNICAL SPECIFICATIONS

} water level (L) is less than 24 feet above the center b line of the core.

l.

3. The reactor shall not be taken critical when the true l value of the pool water level (L) is less than 2 feet- )

i above the center line of the core.

4. The true value of the reactor coolant inlet temperature (pool temperature, Tp) shall not be greater than 1100F.

L Bases 1

l. j The criterion for establishing a safety limit with natural convection flow is the onset of nucleate boiling at the hot I spot on the hot channel. The analysis of natural convection flow given in Section 3.1.2.1 of the FSAR Supplement for Conversion to LEU Fuel shows that ONB occurs at 0.335 MW j with a corresponding fuel clad temperature of 118.60C l L

(245.5 F) which is well below the temperature at which fuel lj clad damage could occur.

Operation of the reactor with less than full water height above the core is limited to 'a power about 250 times lower-1 than the limit with full water height; there is no possibility of fuel clad damage under water immersion at 1.33 kW.

2.2 LIMITING SAFETY SYSTEM SE'ITINGS 2.2.1 Limiting Safety System Settings in the forced convection mode of operation.

l. Applicability i

j TS-10 q i

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TECHNICAL SPECIFICATIONS This specification applies to the setpoints - for the safety channels monitoring reactor thermal power (P), coolant flow rate (W), reactor coolant inlet temperature (Ti), and the height of water above the center line of the core (L).

Objective To assure that automatic protective action is initiated in l order to prevent a Safety Limit from being exce-eded.

Specification i g Under conditions of forced convection flow the values of the 1

Limiting Safety System Settings shall be as follows:

P = 1.25 MWt (max)

W= 1170 GPM (min) l Ti = 1080F (max)

L = 24.25 ft (min)

Bases The Limiting Safety Sys'.em Settings that are given in Specification 2.2.1 represent values of the interrelated variables. which, if exceeded, shall result in automatic protective action that will prevent Safety Limits from being exceeded during the course of the most adverse anticip_ated transient. To determine the LSSS given above, an analysis of l the uncertainties in the instruments and measurements .was -

taken into account. These safety settings are adjusted so that the true value of the measured parameter will not exceed the specified Safety Limits. The results of these adjustments included a flow variation of 4%, a temperature t

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R TECHNICAL SPECIFICATIONS  !

u variation of 20F, a power level variation of 6%, and a pool )

water level variation of three inches. (See Section 3.1.2.5 of the FSAR. Supplement for Conversion to LEU Fuel and Paragraph 9.1.2 of the FSAR).

2.2.2 Limiting Safety System Settings in the natural convection  !

flow mode of operation.

Applicability This specification applies to the setpoints for the safety channels monitoring reactor thermal power (P), reactor pool-temperature (Tp), and the height of water above the center line of the core (L).

Objective To assure that automatic protective action is initiated in-order to prevent undesirable radiation levels on' the surface of. the pool.

Specification Under conditions of natural convection flow the rneasured ,

values of the Limiting Safety System Settings shall' be as follows:  ;

Full Pool Level Low Pool Level P = 125 kW. (max) P = 1.25 kW (max)

Tp = 1080F (max) Tp = 1080F (max)

L = 24.25 ft (min) L = 2.25 ft (min) . -

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TECHNICAL SPECIFICATIONS i i

Bases The Limiting Safety System Settings that are given in Specification 2.2.2 represent values of the interrelated l variables which, if exceeded, shall result in automatic protective action that will prevent undesirable radiation levels on the surface of the pool. due to: a) the production and escape of 16N during the natural convection mode of a operation with full pool level, and b) direct radiation from the core during low pool level operation. The specifications R given above assure that an adequate safety margin exists ,

between the LSSS and the SL for natural convection, because .

the values of the power LSSS would be much higher (335 kW, Section 3.1.2.1 of the FSAR Supplement for Conversion to LEU Fuel) if the specifications were based on Safety Limits rather than on 16N production. The 16N criterion is not related to the ONB which was the criterion used in establishing the Safety Limits (see Section 3.1.2.1 of the FSAR Supplement for Conversion to LEU Fuel).

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TECIINICAL SPECIFICATIONS 3.0 LIMITINO CONDITIONS FOR OPER ATION 3.1 REACTIVITY Applicability These specifications apply to the reactivity condition of the reactor and the reactivity worths of control rods, regulating rod, and experiments.

Obiective To assure that the reactor can be safely shutdown and maintained )

in a safe shutdown condition at all times and that the Safety Limits will not be exceeded.

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

1. The minimum shutdown margin relative to the cold, clean (xenon-free) critical condition, with the most reactive control rod in the fully withdrawn position, is greater than 2.7% delta k/k.

2. The reactor core is loaded so that the excess reactivity in .the cold clean (xenon-free) critical condition does not exceed 4.7% delta k/k.

3. All core grid positions are filled with fuel elements, irradiation baskets, source holders, regulating rod, graphite reflector elements or grid plugs. All but '5 of the peripheral radiation baskets must contain flow restricting devices. This specification will not apply for low power operation, <10 kW, without forced flow.

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p TECHNICAL SPECIFICATIONS

4. The drop time of each control rod from a fully withdrawn .

s' position is less than 1.0 second. i l

5. The isothermal temperature coefficient of reactivity is negative at temperatures >700F. J

6. The reactivity insertion rates of'the control rods are less i L than 0.025% delta k/k per second.

7. The total reactivity worth of the regulating rod is less than the effective delayed neutron fraction. I

8. The reactivity insertion rate of the regulating rod is less I than 0.054% delta k/k per second.

i 9. The -reactivity worth of experiments shall not exceed the j values indicated in the following table:

e Kind Single Experiment Worth Total Worth -

Movable (including the 0.1% delta k/k 0.5%~ delta k/k j l

pneumatic rabbit) summed L together for all experiments L

l Secured experiments 0.5% delta k/k 2.5% delta k/k

10. The total reactivity worth of all experiments shall not be greater than 2.5% delta k/k. )

1 Bases

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l 1. The shutdown _ margin required .by Specification I assures -

i that the reactor can be shut down from any operating l

condition and will remain shutdown after cooldown and xenon decay, even if the highest worth control rod should be -

in the fully withdrawn position. q TS-15

TECHNICAL SPECIFICATIONS

2. The maximum allowed excess reactivity of 4.7% delta k/k provides sufficient reactivity to accommodate fuel burnup, xenon and samarium poisoning buildup, . experiments, and control requirements, but gives a sufficient shutdown margin even with the highest worth rod fully withdrawn.

3. The requirement that all grid plate positions be filled and the restriction on radiation baskets during reactor operation assures that the quantity of primary coolant which bypasses the heat producing elements will be kept within the limits used in establishing Safety Limits in Section 3.1.2 of the L FSAR Supplement for Conversion to LEU Fuel. This I'

requirement does not apply under natural circulation conditions at low power

4. The control rod drop time required by Specification 4  ;

assures that the Safety Limit will not be exceeded during the flow coast down which occurs upon loss of forced convection coolant flow. The analysis of this. situation, which l

is given in Section 3.1.2.5 of the FSAR Supplement for -1 Conversion to LEU Fuel, assumes a 1 second rod drop time.

5. The requirement for a negative temperature coefficient of l reactivity assures that any temperature rise caused by a i reactor transient will not cause a further increase in reactivity. 1 TS-16 m - -. _ _ _ _ _ . _ _ - - - - _ - _ - _ _ . _ . _ _ _ _ _ _ _ _ _ . - _ _ _ _ _ _ _ _ _ _ _ _ . - - - . - . _ _ . - _ _ _ - - - - _ _ _ _ _ _ . - . _

TECHNICAL SPECIFICATIONS

6. The maximum rate of reactivity insertion by the control rods which is allowed in Specification 6 a'ssures that the Safety _ Limit will not be exceeded during a startup accident due to a continuous linear reactivity insertion. Analysis in Section 3.1.2.9 of the FSAR Supplement for Conversion to LEU Fuel shows that a maximum power of less than 1.3 MW would be reached assuming a continuous linear reactivity insertion rate of 0.035% delta k/k per second, which is greater than the maximum allowed.

7. Limiting the reactivity worth of the regulating rod to a' value le< than the effective delayed neutron fraction assures that-a failure of the automatic servo control system could not result in a prompt critical condition.

8. The maximum rate of reactivity insertion by the regulating rod which is allowed in Specification 8 assures that the Safety Limit on reactor power will not be exceeded during an operational accident involving the continuous withdrawal-of the regulating rod. The analysis, in Section 3.1.2.9 of the FSAR Supplement for Conversion to LEU Fuel, shows that the maximum power reached would be about 1.3 MW.

9. Specification 9 assures that the failure of a single experiment will not result in the exceeding of a Safety Limit; the analysis of the step insertion of 0.5% delta k/k is given -

in Section 3.1.2.8 of the FSAR Supplement for Conversion to LEU Fuel. Limiting a movable experiment such as the pneumatic rabbit to 0.1% delta k/k assures that the prompt TS-17

TECHNICAL SPECIFICATIONS jump, which is about 17%, will result in a power below the power level scram setting, i.e., below 125% of. power.

10. The total reactivity of 2.5% in Specification 10 places a reasonable upper limit on the worth of all . experiments which is compatible with the allowable excess reactivity and the shutdown margin and is consistent with the functional mission of the reactor.

3.2 REACTOR INSTRUMENTATION ,

Applicability This specification applies to the instrumentation which must be available and operable for safe operation of the reactor.

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

Specification The reactor shall not be operated unless the measuring channels ,

listed in the following table are operable:

Minimum Operating Mode in Measurine Channel Required Which Reauired Startup Count Rate 1 All modes- (during reactor startup)

Log N (Period) 1 All modes Power Level (Linear N) 2 All modes 1

Reactor. Coolant Inlet 1 Forced convection l Temperature Coolant Flow Rate 1 Forced convection l TS-18 q

TECHNICAL SPECIFICATIONS Reactor Pool Temperature 1 All modes Bases The neutron detectors assure that measurements . of the reactor power level are adequately displayed during reactor startup and low and high power operation. The temperature and flow detectors give information to the operator to prevent the exceeding of a Safety Limit.

3.3 REACFOR SAFETY SYS'EM Aoplicability This specification applies to the reactor safety system channels.

Obiective  !

To require the minimum number of reactor safety system #

channels that must be operable in ' order to assure safe operation of the reactor.

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

Reactor Safety Minimum Operating Mode System Comnonent/ Channel Reouired Function in Which Reauired Startup Count Rate 1 Prevent ' blade Reactor startup withdrawal in:all modes when N count rate s 2 cps Reactor Period 1 Automatic reactor All modes  ;

scram with s 3 sec ]

period  !

Control blade inhibit j s 15 sec period TS-19 1

TECHNICAL SPECIFICATIONS '

Reactor Safety Minimum Operating Mode System Component / Channel Required Function in Which Required Reactor Power Level 2 Automatic scram All modes when 1125% of ,

range scale  !

Coolant Flow Rate 1 Automatic scram Forced convec- l at 1170 gpm tion above 0.1 MW Seismic Disturbance 1 Automatic scram at All modes l Modified Mercalli Scale IV Primary Piping Alignment 1 Automatic scram Forced Convec-

• tion above 0.1 MW Pool Water Level 1 Automatic scram at:

(1) 24.25 ft (1) All modes-above core center with full water- 4 line; height; (2) 2.25 ft above (2) operation core center line with limited water - height Pool Temperature 1 Automatic scram All modes ,

2 1080F Coolant Inlet Temperature 1 Automatic scram Forced convec-21080F tion .above 0.1

.M W Bridge Movement 1 Automatic scram All modes if moved >l inch

- TS-20

l

' TECHNICAL SPECIFICATIONS -

1 Reactor Safety Minimum Operating Mode System Component / Channel Reauired Function in Which Reauired Coolant Gate Opens 2 Automatic scram if Forced convec-either the coolant tion above 0.1 riser or coolant MW; down downcomer gate comer flow opens pattern 1 Automatic scram if Forced convec-the coolant riser tion above 0.1 gate opens MW; cross pool flow pattern Detector High Voltage Failure 1 Automatic scrarr if All modes Voltage <500V Thermal Column Door Open 1 Automatic scram All modes Truck Door and/or Air lock 3 Automatic scram All modes Integrity Manual Scram Button 1 Manual scram All modes

" Reactor On" Key-Switch 1 Manual scram All modes if "off" 1

TS-21 I

. 1 l

j

il TECHNICAL SPECIFICATIONS J

Bases The inhibit function on the startup channel assures that the required startup neutron source is sufficient and in a proper location for the reactor startup, such that a minimum source multiplication count rate level is being detected to ensure proper operation of the startup channel.

The automatic protective action initiated by the reactor period channel, high flux channels, flow rate channels, coolant inlet temperature channel, pool temperature channel, and pool water level channel provices the redundant protection to assure that a Safety Limit is not exceeded.

Automatic protection action initiated by the seismic detector, bridge misalignment, opening of coolant gates, high voltage failure, and opening of thermal column do3r assures shutdown of the reactor under conditions that could lead to a safety problem.

The automatic protective action covering the condition of the air lock doors assures that containment capability is maintained.

The manual scram button and the " Reactor On" Key-Switch provide two manual scram methods to the operator if any abnormal condition should occur.

3.4 RADIATION MONITORING EOUIPMENT Applicability This specification applies to the availability of radiation monitoring equipment which must be operable during reactor  ;

-l operation. '

l TS-22 j i

I I

l

1 TECHNICAL SPECIFICATIONS I

Obiective To assure that radiation monitoring equipment is available for evaluation of radiation conditions in restricted and unrestricted areas.

Specification

1. When the reactor is operating, gaseous and particulate sampling of the stack effluent shall be monitored by a stack monitor with readouts in the control room. l t

2. When the reactor is operating, at least one constant air monitoring unit located in the containment building on the reactor pool level and having a readout in the control room shall be operating.

3. The reactor shall not be continuously

• operated with'out a minimum of one radiation monitor on the_ experimental level of the reactor building and one monitor over the reactor pool operating and capable of warning personnel of high radiation levels.

Bases A continuing evaluation of the radiation levels within the reactor building will be made to assure the safety of personnel. This is accomplished by the area monitoring system of the type described-in Chapter 10 of the FSAR.

• In order to continue operation of the reactor, replacement of ' an inoperative monitor must be made within 15 minutes of recognition of failure, except that the reactor may be operated in a steady. state power mode if the installed systems are replaced with portable y gamma-sensitive instruments having their own alarm.

TS-23

TECHNICAL SPECTFICATIONS A continuing evaluation of the stack effluent will.be made using the information recorded from the particulate and gas monitors.

3.5 CONTAINMENT AND EMERGENCY EXHAUST SYSTEM Applicability This specification applies to the operation of the reactor containment and emergency exhaust system.

Obiective To assure that the containment and emergency exhaust system 'is in operation to mitigate the consequences of possible release of radioactive materials resulting from reactor operation.

Specification The reactor shall not be -operated unless the following equipment is operable, and conditions met:

Equipment / Condition Function

1. At least one door in each of the To maintain containment personnel air locks is closed system integrity and the truck door is closed.

2. All isolation valves, except To maintain containment that reactor operation can system. integrity proceed if a failed isolation .

valve is in the closed ~!

(isolated) position.  ;

TS-24 l

1

TECHNICAL SPECIFICATIONS Equioment/ Condition . Function ,

3. Initiation system for containment To maintain containment isolation, system integrity

4. Emergency exhaust system To maintain the ability to tend toward a negative-building pressure without unloading any large fraction of possible airborne activity.

5. Vacuum relief device To ensure that building vacuum will not exceed 0.2 psi.

6. Reactor Alarm system

• To assure that proper emer-gency action is taken.

Bases In the unlikely event of a release of fission products, or other airborne radioactivity, the containment isolation initiation system will secure the normal ventilation exhaust fan, will bypass the normal ventilation supply up the stack, and will close the normal inlet and exhaust valves. In containment,. the emergency exhaust system will tend to maintain a negative building pressure with a combination of controls intended - to prevent unloading any large-  !

i fraction of airborne activity' if the ' internal building pressure is high. The emergency exhaust purges- the building air through charcoal and absolute filters and controls the discharge, which is l

diluted by supply air, through a 100-foot stack on site. Chapter 3 i

l

• The public address system can serve as a temporary substitute for reactor evacuation j and fonnation of the Emergency Team during short periods of maintenance.  ;

1' TS-25

TECHNICAL SPECIFICATIONS of the FSAR describes the system's sequence of operation; Chapter  ;

9 provides the analysis.

I 3.6 LIMITATIONS OF EXPERIMENTS l Applicability This specification applies to experiments to be installed in the reactor and associated experimental facilities.

Obiectives To prevent damage to the reactor or excessive release of radioactive materials in the event of an experirnent failure.

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

1. All materials to be irradiated shall be either corrosion resistant or encapsulated within corrosion resistant containers to prevent interaction with reactor components or pool water. Corrosive materials shall be doubly encapsulated. .

2. Irradiation containers to be used in the reactor, in which a static pressure will exist or in which a pressure buildup is H

predicted, shall be designed and tested for a pressure  ;

exceeding the maximum expected by a factor of _2. l

3. Explosive material such as (but not limited to) gunpowder, dynamite, TNT, nitroglycerine, or PETN in quantities <25 mg .

l may be irradiated in the reactor or experimental facilities q provided out-of-core tests indicate that, with the containment provided, no damage to the explosive TS-26 l

1 TECHNICAL SPECIFICATIONS containers, the reactor, the reactor components or the Co-60 l Source shall occur upon detonation of the explosive. ]

4. Explosive materials, in quantities >25 mg shall not be allowed in the reactor or the reactor pool without rigorous safety evaluation, and special authorization from the USNRC.

5. All experiments shall be designed against failure from internal and external heating at the true values -associated with the LSSS for reactor power level and other process variables. ,

6. The outside surface temperature of a submerged experiment-or capsule shall not exceed the saturation temperature of the reactor coolant during operation of the reactor.

7. Experimental apparatus, material or equipment to be . ,

irradiated shall be positioned so as not to cause shadowing of the nuclear instrumentation, interference with control rods, or other perturbations which may interfere with safe operation of the reactor.

8. Cryogenic _ liquids shall not be used in any experiment within '

the reactor pool.

9. The reactor shall not be operated whenever the reactor core is in the same end of the reactor. pool as any portion of the Cobalt-60 Source.

TS-27

TECHNICAL SPECIFICATIONS Bases Specifications 1 through 6 are intended to reduce the likelihood of damage to reactor components and/or radioactivity releases resulting from experiment failure, including any experiment involving the Co-60 Source and, along with the reactivity restriction of pertinent specifications in 3.1, serve as a guide for the review and approval of new and untried experiments by the operations staff as well as the Reactor Safety Subcommittee.

Specification 7 assures that no physical or nuclear interferences compromise the safe operation of the reactor by, for example, tilting the flux in a way that could affect the peaking factor used in the Safety Limit calculations. Review of the experiments using the appropriate LCO's and the Administrative Controls of Section 6 assures that the insertion of experiments will not negate the ,

considerations implicit in the Safety Limits.

Specification 8 prohibits experiments using cryogenic materials.

(Special NRC permission would be required.). Cryogenic liquids present structural and explosive problems which enhance the potential of an experiment failure. Specification 9 assures that there will be no interference, either instrumental or procedural, between the reactor and the cobalt source during reactor operation.

TS-28

TECHNICAL SPECIFICATIONS )

l

'l 3.7 GASEOUS EFFLUENTS Applicability I This specification applies to the routine release of gaseous radioactive effluents from the facility.

Objective The objective is to minimize the release of gaseous radioactive effluents, particularly Argon-41, the effluent most likely to be generated in routine operation.

Specification The release rate of gaseous radioactive material from the reactor stack shall be limited to 8 microcuries per second averaged over a year.

f3ases Calculations based on a very conservative model, allowing for no atmospheric dilution of the gaseous effluent, have- predicted an annual dose of 12 mrem to an individual exposed to the effluent on a continual basis for an Argon-41 release rate of 8 microcuries per second. Allowance for even minimal atmospheric turbulence would reduce this dose number by about a factor of three.

3.8 COOLANT SYSTEM Applicabilitv This specification applies to the reactor pool water requirements-for operation of the reactor.

i TS-29

TECHNICAL SPECIFICATIONS Obiective The objectives are to require that the reactor pool water be of high purity in order to retard corrosion and to monitor the integrity of the fuel cladding and the Cobalt-60.

Spacification

1. The conductivity of the pool water shall be maintained at a value of 5 micromhos per centimeter or less averaged over a ,

month.

2. The pool water shall be analyzed for gross activity and for Cobalt-60. Analyses shall be capable of detecting levels of 10-7 microcuries per milliliter. If a sample analysis reveals. a significant increase of. activity in the water, with respect to the previous samples, or a contamination level greater than 10-6 microcuries of . Cobalt-60 per milliliter of water, prompt action shall be taken to prevent further contamination of the pool water.

If the gross activity of the sample is less than 10-7 microcuries per milliliter, specific analysis for Cobalt-60 need not be performed. If remedial action is required by. this section, notification will be made to the USNRC as required by Section 6.6.2.

Bases Pool water of high purity minimizes the rate of corrosion. .

Radionuclide analysis of the pool water allows early determination of any. significant buildup of radioactivity i om operation of the reactor or the Cobalt-60 source.

TS-30 I

I TECHNICAL SPECIFICATIONS )

l 4.0 SURVEILLANCE REOUIREMENTS 4.1 CONTROL AND REGULATING RODJ Applicability This specification applies to the surveillance requirements for the control and regulating rods.

Objective To assure the operability of the control and regulating rods.

Specifications

1. The reactivity worth of the regulating rod and each control rod shall be determined annually. The reactivity worth of all rods shall. also be determined prior to routine operation of any new fuel configuration in the reactor core.

2. Control rod drop and drive times and regulating rod drive time shall be determired annually, or if maintenance or modification is performed on the mechanism. Nominally, the withdrawl rate of the safety blades is at 3.5 inches per minute and the withdrawl rate of the regulating rod is at 78 inches per minute.

3. The control and regulating rods shall be visually inspected annually.

13ases The reactivity worth of the control and regulating rods is measured to assure that the required shutdown margin is available, and to provide a means for determining the reactivity worths of experiments inserted in the core. Annual measurement.

of reactivity worths provides a correction for the slight variations TS-31

i TECHNICAL SPECIFICATIONS expected because of burnup. The required measurement after any new arrangement of fuel in the core assures that possibly altered rod worths will be known before routine operation.

The visual inspection of the regulating and control rods and the measmements of drive and drop times are made to assure that the rods are capable of performing properly and within the considerations used in transient analyses in the FSAR Supplement ,

for Conversion to LEU Fuel. Appropriate inspection data will be recorded and analyzed for trends. Verification of operability after maintenance or modification of the control system will ensure proper reinstallation or reconnection.

4.2 REACTOR SAFETY SYSTEM Applicability This specification applies to the surveillance requirements for the Reactor Safety System.

Obiective To assure that the Reactor Safety System (RSS) will remain operable and will prevent the Safety Limits from being exceeded.

9

.h TS-32

TECIINICAL SPECIFICATIONS Specifications

1. A channel check of each measuring channel in the RSS shall be performed daily when the reactor is in operation.

2. A channel test of each measuring channel in the RSS shall be performed prior to each day's operation, or prior to each operation extending more than one day.

3. A channel calibration (reactor power level) of the Log N and linear safety power level measuring channels shall be made annually.

4. A channel calibration of the following channels shall' be made annually.

a. Pool water temperature

b. Primary coolant flow rate

c. Pool water level

d. Primary coolant inlet and outlet temperature

5. The manual scram shall be verified to be operable prior to each reactor startup.

6. Any RSS instrument channel replacement must have i undergone a channel check prior to installation, and must undergo a channel calibration before routine operation of the reactor after channel installation.

7. Any RSS instrument repaired or replaced while the reactor . l

'l is shutdown must have a channel test prior' to reactor operation.

8. - Each protective channel in the RSS shall be verified to' be operable semi-annually.

TS-33 l

TECHNICAL SPECIFICATIONS I3ases The daily channel tests and checks and periodic verifications will assure that the safety channels are operable. Annual calibrations will assure that long-term drift of the channels is corrected. The calibration of the reactor power level will provide continual reference for the adjustment of the Log N and safety channel L detector positions and current alignments.

4.3 RADIATION MONITORING EOUIPMENT Applicability This specification applies to the surveillance requirements for the area radiation monitoring equipment and systems for monitoring airborne radioactivity.

Objective To assure that the equipment used for monitoring radioactivity is operable and to verify the appropriate alarm settings.

Specification

1. The operation of the area radiation monitoring equipment and systems for monitoring airborne radioactivity, and their 3

associated alarm set points, shall be verified prior to reactor startup.

2. All radiation monitoring systems shall be calibrated semiannually.

TS-34

TECHNICAL SPECIFICATIONS Dases The area radiation monitoring system, described in the Emergency .

Plan, includes _ the stack air monitor, two building constant air monitors, a fission product monitor,12 GM detectors and two ion chamber detectors at selected sites throughout the building. The detectors used have been chosen for stability and operational reliability. The large number of detectors in the area' monitoring system ensures that if a particular monitor should malfunction or-drift out of calibration, sufficient backup monitors are available for reliable information. Calibration of the area monitors semi-annually is sufficient to insure the required reliability. Daily checks (during operating days) of the area monitors ensure that any obvious malfunctions will be d" ected.

4.4 CONTAINMENT BUILDING Anglicability .

This specification applies to the surveillance requirements for the containment building.

Obiective To assure that the containment ' system is operable.

Specification

1. Building pressure will be verified at least every eight hours during reactor operation to ensure that it is less than ambient atmospheric pressure. '!

TS-35 1

TECHNICAL SPECIFICATIONS

2. The containment building isolation system including the  ;

initiating system shall be tested semi-annually. The test shall verify that valve closure is achieved in <2.5 seconds after the initial signal.

3. An integrated leakage rate of the containment building as-is* shall be performed at a pressure of at least 0.5 psig at -

intervals of 5 years to verify leakage rate of-less than 10%

of the building air volume / day at 2 psig.

4. All additions, modifications, or maintenance of the containment building or its penetrations that could affect building containment capability shall be tested to verify containment requirements.

5. The emergency exhaust system including the initiating system shall be verified annually to be operable.

6. At two year intervals, and subsequent to replacement of the facility filters and prior to reactor operation thereafter, the filter trains shall be tested to verify that they are operable.

7. At two year intervals, the air flow rate in the stack exhaust-duct shall be measured.

i

'I y

• Non-routine maintenance or repair for the purpose of reducing containment leakage shall not be performed prior to the leak test.

l TS-36 l

l TECHNICAL SPECIFICATIONS.

l Bases Maintaining a negative pressure ensures that - any leakage in ' the containment is inward.

Valve closure time was chosen to be 1/2 the time required for a given sample of air to travel from the first to the second valve in series in the exhaust line under regular flow conditions. Semi.

annually is considered a reasonable frequency of testing. The containment building was designed to withstand a 2.0. psig internal pressure. An overpressure of less than 0.5 psig would result from an excursion of 538 MWs, which is nearly four times the energy release achieved in the Borax tests. A 0.5 psig test pressure is therefore adequate.

Any additions, modifications or maintenance to the building or its penetrations shall be tested to verify that such work has not adversely affected the integrity of the building.

Surveillance of the. emergency exhaust system and the various filters will verify that these are functioning. See Chapters 3 and 7 of the FSAR.

4.5 POOL WATER Applicability This specification applies to the surveillance requirement of ,

monitoring the ~ quality and the radioactivity _ in the pool water.

Objective -

To assure high quality pool water and to monitor the radic0ctivity in the pool water in order to verify the integrity of th:: fuel cladding.

TS-37 -!

. . - .. ..- .. - -. - - -~ . - . -

TECHNICAL' SPECIFICATIONS .

Soccification

1. The conductivity of the pool water shall be measured weekly.

2. The radioactivity 'in the pool water shall be. analyzed weekly. l Bases Surveillance of water conductivity assures that changes. that could ' l accelerate corrosion have not occurred. Radionuclide analysis of the pool water samples will allow early-. determination - of any; significant buildup of radioactivity from operation of-the reactor or the Co-60 source, l 4.6 SCRAM BY PROCESS VARIABLE EFFECT ' >

Applicability ,

This specification applies to the surveillance requirements' applied '

to process variable scrams.

Obiective  :

To. assure that a Safety Limit is not : exceeded. .

Specification  ;

Following a reactor scram caused by a process variable, 'the  ;

reactor shall not be operated. until n'n evaluation has been made- to

~

determine if a Safety Limit was exceeded, the cause. of the -scram,

-t the effects of operation to the scram point and the ' appropriate - ,

action'. to be 'taken.

TS-38  !

T 7- F - T M 4 *T 7

TECHNICAL SPECIFICATIONS Bases This specification assures that if a Safety Limit should be exceeded as a result of a malfunction of a process variable, the fact will be' known.

4.7 FUEL SURVEILLANCE i

Applicability This specification applies to the surveillance requirements for reactor fuel.

Objective To assure that reactor fuel is in proper physical condition.

Specification Visual inspection of a representative sample of reactor fuel i

elements shall be performed every two years. l Bases l

The inspection of reactor fuel assures that fuel elements, when i i

used in the core, will perform as designed. 1 i

I TS-39

TECHNICAL SPECIFICATIONS 5.0 DESIGN FEATURES 5.1 REACTOR FUEL The reactor fuel shall be as follows:

1. Standard fuel element: the fuel elements shall be flat plate MTR-type elements. The plates shall be fueled with low enrichment (20% U-235) U3Si2, clad with aluminum. There -

f' shall be 18 plates per element with 16 containing fuel and two outside plates of aluminum. There shall be 200 5.6 grams of Uranium-235 per element.

2. Half-element: same as a standard fuel element except each 1

plate has one half the uranium loading.

l

3. Variable-load element: same as Specification 1 above, but internal plates are removable.  !

5.2 REACIDR CORE

1. The reactor core consists of a 9 x 7 array of 3-inch square modules with the four corners occupied by posts. The j reference core for these Technical Specifications consists of 20 standard fuel elements in a 5 x 5 array with corners removed and the central location filled with a graphite-water aluminum clad flux trap element, as shown in Figure 2.6 of the FSAR Supplement for Conversion to LEU Fuel.

2. Cores from 16 standard elements to 28 elements may be

l used, and cores from 16 elements to 28 elements may contain 2 half-loaded elements.

k TS-40 l

TECHNICAL SPECIFICATIONS

3. Cores with loadings different from 20 standard elements may be operated under forced convection only after analyses using 1) the methods described in the FSAR Supplement for Conversion to LEU Fuel, or 2) flux measurements in natural convection, establish that no alteration of the LSSS's are required to preclude violation of a SL during the transients anticipated in the FSAR.

5.3 REACTOR BUILDING L The reactor shall be housed in the reactor building, designed for containment.

5.4 FUEL STORAGE All reactor fuel element storage facilities shall be designed in geometrical configuration so' that keff is less than 0.85 under quiescent flooding with water.

1 I

i TS-41 1

I

TECHNICAL SPECIFICATIONS h

6.0 ADMINISTRATIVE CONTROLS 6.1 ORGAN 1ZATION AND MANAGEMENT l 1. The reactor facility shall be an integral part of the Radiation Laboratory of the University of Massachusetts Lowell. The 1

reactor shall be related to the University structure as shown in Chart 6-1 TS.

2. The reactor facility shall be under the direction.of the Director of the Radiation Laboratory, who shall be a member L

of the graduate faculty, and it shall be supervised by the Reactor Supervisor who shall be an NRC-licensed senior operator for the facility. The Reactor Supervisor shall be responsible for assuring that all operations are conducted in a safe manner and within the limits prescribed by the facility license and the provisions of the Reactor Safety Subcommittee.

3. There shall be a Radiation Safety Officer responsible for the safety of operations from the standpoint of radiation protection. He does not report formally to the line organization responsible for reactor operations, but rather to the Vice-Chancellor for Academic Affairs (see Chart 6-1 TS).

4. An Operator or Senior Operator licensed pursuant.to .10 CFR

- 55 shall be present at the controls unless the reactor is secured as defined in these specifications. In addition, a -

second individual shall be present in the reactor building or Pinanski building whenever - the reactor is not -secured.- This individiual shall be a Licensed Senior Operator, Licensed TS-42 j

Ch ncolicr UNIVERSITY vice- RADIATION SAFETY

, . Chancellor - COMMITTE E ACADEMIC AFFAIRS AUDIT' REACTOR

__ SAFETY SUB-COMMITTEE

, DIRECTOR REYlEW RADIATION

' LABORATORY RADIATION SAFETY OFFICER MANAGER REACTOR-g RADIATION SUPERVISOR 4 SERYlcES s

STAFF - - - - - - - - - STAFF 3

- Figure 6.1 TS Organizational' Chart For The University cf Massaciiusetts Lowell Radiation Laboratory

_ _ _ .w ..__ *__ -- - __s *-ie<h *

• N".M 7 9 d W '-__ _ _ _-m_ _

TECHNICAL SPECIFICATIONS Operator or an individual who is capable of shutting the reactor down in case of an emergency.

5. A Licensed Senior Operator shall be on the console or readily l available on call whenever the reactor is in operation.

]

6.2 REVIEW AND AUDIT

1. There shall be a Reactor Safety Subcommittee which shall review reactor operations to assure that the facility is operated in a manner consistent with public safety and l within the terms of the facility license. The Subcommittee 1

I shall report to the University Radiation Safety Committee which has overall authority in the use of all radiation sources at the University.

2. The responsibilities of the Subcommittee include, but are not l limited to, the following: '

a. Review and approval of normal, abnormal and emergency operating and maintenance procedures and-records.

b. Review and approval of proposed tests' and experiments utilizing the reactor facilities in accordance with Paragraph 6.8 of these specifications,

c. Review and approval of proposed changes to the facility systems or equipment, procedures, and operations.

d. Determination of whether a proposed change, test, or experiment would constitute an unreviewed safety TS-43

TECHNICAL SPECIFICATIONS question requiring a change to 'the Technical Specifications or facility license,

e. Review of all violations of the Technical Specifications and NRC Regulations, and significant violations of .

internal rules 'or procedures, with recommendations for corrective action to prevent recurrence.

f. Review of the qualifications and competency of the operating organization to assure retention of staff quality.

3. The Reactor Safety Subcommittee shall be composed of at least five members, one of whom shall be the Radiation Safety Officer or his designee and another of whom _shall be the Reactor Supervisor or his designee. The Subcommittee shall be proficient in all areas of reactor operation and reactor safety. The membership of .the Subcommittee shall include at least two senior scientific staff members, and the chairman will not have line responsibility for operation of the reactor.

4. The Subcommittee shall have a written charter defining such matters as the authority of the Subcommittee, the subjects within its purview, and other such administrative provisions as are required for effective functioning of the Subcommittee. Minutes of all meetings of the Subcommittee shall be kept.

5. A quorum of the Subcommittee shall consist of not less than a majority of the full Subcommittee and shall include the i TS-44 ,

1 i

4

-i

a TECHNICAL SPECIFICATIONS Radiation Safety Officer or his designee, and ' the Reactor Supervisor or his designee.

6. The Subcommittee shall meet at least quarterly.

6.3 OPERATING PROCEDURES Written procedures, reviewed and approved by the Reactor Safety Subcomrnittee shall be in effect and followed _for the following items. The procedures shall be adequate to assure the safe operation of the reactor, but should not preclude the use of independent judgment and action should the situation require 1

such.

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

2. Installation or removal of fuel elements, control rods, experiments and experimental facilities.

3. Actions to be taken to correct specific and potential malfunctions of systems or components, including responses to alarms, suspected primary coolant system leaks, and abnormal reactivity changes.

4. Emergency conditions involving potential or actual release of' radioactivity, including provisions for evacuation, re-entry,  ;

recovery, and medical support.

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

6. Periodic surveillance of reactor instrumentation and safety systems, area monitors and continuous air monitors.

7. Civil disturbance on or near campus.

TS-45

TECHNICAL SPECIFICATIONS T. , Radiation control, for which procedures shall be maintained and available to all operations personnel.

9. Receipt, inspection, anJ storage of 'new fuel elements.

10. Handling and storage of irradiated fuel elements.

Substantive changes to the above procedures shall be made only.

with the approval of the Reactor Safety Subcommittee.

Temporary changes to the procedures that do not change their original intent may be made by the Reactor Supervisor.

Temporary changes to procedures shall be documented and subsequently reviewed by the Reactor Safety Subcommittee.

6.4 ACTION TO BE TAKEN IN THE EVENT OF AN ABNORMAL OTURRENCE In the event of an abnormal occurrence:

1. The Reactor Supervisor or his designee shall be notified '

promptly and corrective action shall be taken immediately. ,

to place the facility in a safe condition until the causes of the abnormal occurrence are determined and corrected.

2. The Reactor Supervisor or his designee saall report the occurrence to the Reactor Safety Subcommittee. -The report shall include an analysis of the cause of the occurrence, corrective actions taken, and recommendations for appropriate action to prevent or reduce the probability of a repetition of the occurrence. -

3. The. Reactor Safety Subcommittee shall review the report and the corrective actions taken.

TS-46

TECHNICAL SPECIFICATIONS

4. Notification shall be made to the NRC in accordance with Paragraph 6.6 of these specifications.

6.5 ACTION TO BE TAKEN IN THE EVENT A SAFETY LIMIT IS EX&EDED In the event a Safety Limit has been exceeded:

1. The reactor shall be shut down and reactor operation shall not be resumed until authorization is obtained from the NRC.

2. Immediate notification shall be made to the NRC in accordance with paragraph 6.6 of these specifications and to the. Director of the Radiation Laboratory.

3. A prompt report shall be prepared by the Reactor Supervisor or his designee. The report shall include a complete analysis of the causes of the event and the extent of possible damage together with recommendations to -

prevent or reduce the probability of recurrence. This report shall be submitted to the Reactor Safety Subcommittee for review and appropriate action, and a suitable similar report shall be submitted to the NRC in accordance with Paragraph 6.6 of these specifications and in support of a request for authorization for resumption of operations.

6.6 REPORTING REOUIREMENTS In addition to the requirements of applicable regulations, and in no way substituting therefore, all written reports shall be sent to the U.S. Nuclear Regulatory Commission, Attn: Document Control Desk, Washington, D.C. 20555, with a copy to the Region I adiminstrator.

TS-47  ;

i l

l

e TECHNICAL SPECIFICATIONS

1. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, a report by telephone or telegraph to NRC Region I Administrator of:

a. Any accide.ntal release of radioactivity to unrestricted areas above permissible limits, -whether or . not the release resulted in property damage, personal injury or exposure.

b. Any significant variation of measured values..from a ,

~

corresponding predicted or previously measured value of safety related operating characteristics. occurring during operation of the reactor,

c. Any abnormal' occurrences as defined in ~ Paragraph : 1.1 of these specifications. '

d. Any violation of a Safety Limit.

2. A written report within 14 days in the event of an abnormal occurrence, as defined in Section 1.1. The report shall:

a. Describe, analyze, and evaluate safety implications;

b. Outline the measures taken to assure that the cause of the condition is determined; -

c. Indicate the corrective action including any changes made to the procedures and to the quality assurance program taken to prevent repetition of the occurrence and of similar occurrences involving similar components or systems:

TS-48

TECHNICAL SPECIFICATIONS

d. Evaluate the safety implication of the incident in light of the cumulative experience obtained from . the record P of previous failure and malfunctions of similar systems and components.

3. Unusual Events.

A written report shall be forwarded within 30 days in the event of:

a. Discovery of any substantial ' errors in the transient or accident analyses or in the methods used for such analyses, as described' in the safety analysis or in the bases for the technical specifications;

b. Discovery of any substantial variance from i performance specifications contained in the technical ,

specifications and -safety analysis.

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

4. An annual report shall be submitted in writing within ~ 60 days following the 30th of June of each year. The report F shall include th'e following information:

TS-49

TECHNICAL SPECIFICATIONS

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

results of surveillance tests and inspections.

b. Tabulation showing the energy generated by the reactor (in megawatt days), 'the number--of hours the reactor was critical, and the cumulative total energy output since initial criticality. '

c. The number of emergency shutdowns and inadvertent scrams, including the reasons therefore.

d. Discussion of the major maintenance operations performed during the period, including the effect, if any, on the safe operation of the reactor, and the reasons for any corrective maintenance required.

e. A description of each change to the facility or procedures, tests, and experiments carried out under the conditions of Section 50.59 of 10 CFR 50 including a summary of the safety evaluation of each,

f. A description of any environmental surveys performed outside the facility.

TS-50 4

TECHNICAL SPECIFICATIONS

g. A summary of radiation exposures received by facility personnel and visitors, including the dates' and times of significant exposures, and a summary of the results of radiation and contamination surveys performed within the facility.

h. A summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the licensee as measured at or prior to the point of such release or discharge.

Liould Waste (Summarized on a monthly basis)

(1) Total gross beta radioactivity released '(in curies) during the reporting period.

(2) Total radioactivity released (in curies) for specific nuclides, if the gross beta radioactivity-exceeds 3 x 10-6 Ci/cm3 at point of release, during the reporting period.

(3) Average concentration (pCi/cm3) of release as diluted by sewage system flow of 2.7 x 108 cm3/d ay.

Gaseous Waste (Summarized on a monthly basis)-

(1) Radioactivity discharged during the reported period (in curies) for: a) gases, b) particulates with half lives greater than eight days.

(2) The MPC used and the estimm ' .ntivity (in curies) . discharged during the reported . period, TS-51

TECHNICAL SPECIFICATIONS by nuclide,. based on representative isotopic analysis.

Solid Waste (Summarized on a monthly basis)

(1)' The total amount of solid waste packaged (in cubic feet).

(2) The total . activity and type of activity involved (in curies).

(3) -The dates of shipment and disposition (if shipped off-site).

6.7 PLANTOPERATING RECORDS In addition to the requirements of applicable regulations and in no way substituting therefore, records and logs of the following items, as a minimum, shall be kept in a manner convenien't for review and shall be retained as indicated:

I o

TS-52

TECHNICAL SPECIFICATIONS

1. Records to be retained for a period- of at least five years:

a. Reactor operations;

b. Principal maintenance activities;

c. Experiments performed including aspects of the experiments which could affect the safety of reactor operation or have radiological safety implications;

d. Abnormal occurrences; and

e. Equipment and component surveillance activities.

2. Records to be retained for the life of the facility:

a. Gaseous and liquid radioactive effluents released to the environs;

b. Off-site environmental monitoring surveys; ,

c. Facility radiation and monitoring surveys;

d. Personnel radiation exposures;

e. Fuel inventories and transfers;

f. Changes to procedures, systems, components, and equipment;

g. Updated, "as-built" drawings of the facility; and

h. Minutes of the Reactor Safety Subcommittee meetings.

TS-53

b

(

TECHNICAL SPECIFICATIONS 6.8 APPROVAL OF EXPERIMENTS

-1. All proposed experiments using the reactor shall be evaluated by the experimenter and a staff member who has been approved by the Reactor Safety Subcommittee. The-evaluation shall be reviewed by the Reactor Supervisor and the Radiation Safety Officer to ensure compliance with the provisions of the facility license, these Technical Specifications, and 10 CFR 20. If the experiment complies with the above provisions, it shall be submitted by the -

Reactor Supervisor to the Reactor Safety Subcommittee 1 -

approval if it is a new experiment, as indicated in -4. below.

The experimenter evaluation shall include:

a. The reactivity worth _ of the experiment;

b. The integrity of the experiment, including the effect of changes in . temperature, pressure, chemical composition, or radiolytic decomposition;

c. Any physical or chemical interaction that could occur with the reactor components;

d. Any radiation hazard that may result from the activation of materials or from external beams; and

e. An estimate of the amount of radioactive materials produced.

2. Prior to performing any new reactor experiment, an evaluation of the experiment shall be made by the Reactor Safety Subcommittee. The Subcommittee evaluation shall l l

consider:

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TECHNICAL SPECIFICATIONS

a. The purpose of the experiment;

b. The effect of the experiment on reactor operation and the possibility and consequences of failure of some aspect of the experiment, including, where significant, chemical reactions, physical integrity, design life, proper cooling interaction with core components, and reactivity effects;

c. Whether or not the experiment, by virtue of its nature and/or design, includes an unreviewed safety - question or. constitutes a significant threat to the integrity of the core, the integrity of the reactor, or to the safety of personnel; and

d. A procedure for the performance of the experiment. JA.

favorable Subcommittee' ev,aluation' will not lead to direct failure of any reactor component 'or .of other experiments. An experiment shall not be conducted until a favorable evaluation indicated in writing is rendered by the Reactor Safety Subcommittee.

3. In evaluating experiments, the following assumptions shall be used for.the purpose' of determining that failure of the experiment would not cause the appropriate limits of.10 CFR 20 to be exceeded:

a. If the possibility exists that airborne concentrations of radioactive gases or aerosols may be released within the containment TS-55

L' TECHNICAL SPECIFICATIONS ~

i building,100% of the gases or aerosols will escape; F

b. If the effluent exhausts through a filter installation designed for greater than 90%'

efficiency for 0.3 micron particles, at 'least 10% of gases or aerosols will escape; and

c. For a material whose boiling point is above 1300F and where vapors formed by boiling.this material could escape only through a volume. of water above the core, at least 10% of these vapors will escape.

4. An experiment that has had prior Subcommittee approval and has been performed safely shall be a routine experiment and requires only the approval of the Reactor Supervisor or.his designee and the Radiation Safety Officer or his designee .to be repeated.

An experiment' that represents a minor variation from a routine experiment not. involving safety considerations of a different kind nor of a magnitude greater than a routine experiment shall be considered the equivalent of a routine experiment and may be approved for the Subcommittee by agreement of the Reactor Supervisor or his designee and the Radiation. -

Safety Officer or his designee.

4 TS-56