ML20198F622

Amend 12 to License R-125,converting from High to low- Enriched U Fuel

ML20198F622
Person / Time
Site:	University of Lowell
Issue date:	07/31/1997
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20198E807	List: ML20198E802 ML20198E808 ML20198E814 ML20198F622
References
R-125-A-012, R-125-A-12, NUDOCS 9708130095
Download: ML20198F622 (60)

Text

_, ---

i ENCLOSURE TO LICENSE AMENDMENT NO. 12 i

FACILITY OPERATING LICENSE NO. R-125_

DOCKET NO. 50-223 REPLACEMENT PAGES FOR TECHNICAL SPECIFICATIONS Replace the Appen enclosed papers. dix A technical specifications in their entirety with the The revised contain vertical lines pages indicating are identified the areas by amendment number and of change, i

9700130095 970731 PDR P ADOCK 05000223 PDR

+.-

ATTACHMENT TO ORDER MODIFYING FACILITY OPERATING LICENSE NO. R-125 A. License Conditions Revised and Added by This Order 2.B.(2) Pursuant to the Act and 10 CFR Part 70, " Domestic Licensing of Special Nuclear Material," to receive, possess, and use at any one time up to 6.0 kilograms of contained uranium 235 at enrichments equal to or less than 20 percent in the form of material test reactor (MTR) type reactor fuel in connection with operation of the reactor and 5 Ci Am Be and 10 Ci Sb Be neutron sources for use in connection with operation of the reactor.

2.B.(4) Pursuant to the Act and 10 CFR Part 70, " Domestic Licensing of Special Nuclear Material," to possess, but not use, up to 4.80 kilograms of contained uranium 235 at greater than 20 percent enrichment in the form of MTR-type reactor fuel until the existing inventory of this fuel is removed from the facility.

2.C.(2) Technical Specifications The technical specificaticas contained in Appendix A, as revised through Amendment No. 12, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the technical specifications.

2.C.(4) The licensee shall submit a startup test report within six months of the initial criticality with low enriched uranium reactor fuel in accordance with Amendment No. 12. This report shall be sent as specified in 10 CFR 50.4, " Written Communications."

8. The technical specifications will be revised by this Order in accordance with the " Enclosure to License Amendment No. 12, Facility Operating License No. R-125, , acket No. 50-223, Replacement Pages for Technical Specifications," and as discussed in the safety evaluation for this Order.

APPENDIX A 10 FACILITY OPERATING LICENSE NO. R-125 Iff'. Ji,,'.

SPECIFICATIONS fc R.ItiE j:1., et TY OF LOWELL l

7 ..

d TABLEOFCONTENTS 1.0 D EFINITI O N S . .. . .. . . . .. .. . . . . . . . . .. . . . .. . . . .. . . . .. .. . . . . . . . . .. . . . . . . .. . . .. . . . . . . . . . . . . .. . . . . . . . . .

1.1 A B NORM AL OCCURR ENCES .......................................................... 1 1.2 CHANNEL............................................................................................1 1.3 Cl IANNEL CALIB R ATION ............................................................. 2 1.4 C HANNEL C HECK ... ... ....... .... ... ..... .. ....... ... .... ... ... . . .. ....... .... ... ....... .. . . . . 2 1.5 C HANNEL TEST . .. .. .. . . . . . . . . . . .. .. . . . . .. . . .. . . . . . .. .. . . . . . . .. . . . . .. . . .. . . . . . . . . . . . . . . . .. . . . .

1.6 CONTAINMENT BUILDING IN'I EGRITY .................................. 2 1.7 CO N'IROL ROD .. .. . . .. . . .. .. . . . . .. . . . . . . . . . . .. .. .. . . .. . .. .. . . . . . . . . .. . . . . . .. . . . . . .. .. .. . . . .. . . . .

1.8 EX CES S REACTIVITY . ... . . . .. .. .... .. .. .. .. .. .... ........ .. .... .. .. .. .... .. .. .. .. ........ . 3 1.9 EXP E RIMENT . .. .. .. .. . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . .. .. .. .. . . . . . . . . . .. . . . . . . . . . . . . . .

1.10 MEA S URED VA LUE .... . .. .. .. .. . .. .. . . .. . .. . . .. . .. .. .. .... . .. .. .. . .. .. .... . .. .... .. . .. .. .. . 3 1.I 1 MOV A B LE EXPERIMENT .. ... ......................................................... 3 1 ,1 2 O PE RA B LE . . . .. .. .. .... .... .. .. .. .. .. .. . . . . . . . . . . . . . . . .. . . . . . . . . . . .. .. . . .. .. . . .. . . .. .. . . . . .. ... . . . . . 3 1.13 0 P E R ATIN G .... .. .. . . .. .. .. .. .. . . . . .. . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . .. . . .. . . . . .. . . .. .. . .

1.14 PROTEC TIVE CHANNEL ........... . .... ........ ...... ........ .............. ............. 4 1.15 REACTOR OPERATING M ODE ...................................................... 4 1.16 REACTO R O PERATING .. ...... .. .. .. .... .. .. .. .. .. ...... . . . .. ... ... .. .... .. .. .... . .... .. . 4 1.17 REACTOR SAFETY SYSTEM ......... ................................................ 4 1.18 REACTO R S ECURED ........ . .... .... ..... .... ..... .... .... ..... ......... .... .... ..... .... .... 4 i 1.I9 REACFO R S HUTDOWN ........ .. ...... ....... . ... ... ..... ... ... ... .. ...... ........ ... ... .. 5 1.20 REFERENCE CORE CONDITION ..................................................... 5 1.21 REG ULATING RO D .... ...... . . .. .... .... .. .. .. .. .. .. ... .. .. .. .. .... .... .. .... .. ............. . 5 1.22 S AFETY CH ANNEL ... ...... .. ...... .. .. .. .. ... .. . . .. .. .. ...... .... ....... ...... .. .... .. .. .. .. 5 1.23 SECURED EXPERIMENT ................. ...................... ........................... 6 1.24 SHALL, S HOULD, AND MAY ..... ................................................... 6 1.25 S HU TDOWN MA RG IN . . .. .. . . . . .. .. .. . .. . .. . . .. . .. .. . .. . .. .. .. .. . .. . .... ..... .. . .. .. . . . .. . 6 1.26 SURVEILLANCE INTERV ALS ............ ......................................... 6 1.27 TR U E V AL UE . .. .. . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . .. . . . . . . .. . . .. . .. . . . . . . . .

2.0 SAFETY LIMITS AND LIMITING S AFETY SYSTEM ........................

SE'ITINGS.........................................................................................................8 2,1 S AFETY LI M ITS ......... . . .. .... . . .. .. .... .. .. .. .. ... .. .. .. .... .. .. .. .. .. .. . . . . .. .. .. . . ..... .. . 8 2.1.1 Safety limits in the forced convection mode of a p e r a t i o n . . . . . . . . . . . . . . . . . . .. . . .. .. .. .. .. . . .. ... .. .. .. . . .. .. .. .. . . .. . . . . . . .. . . .. .. . . . . . . . . 8 2.1.2 Safety Limits in the natural convection mode o f ope rat i o n . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 2.2 LIMITING S AFETY SYSTEM SETTINGS....................................10 2.2.1 Limiting Safety System Settings in the forced convection mode of operation........................... ............10 2.2.2 Limiting Safety System Settings in the natural convection flow mode of operation. ................ ...........12 3.0 LIMITING CONDITIONS FOR OPERATION................................. ..........I 4

r. -

4 1

t 3.1: R EA CTIV ITY. .. .. . . . . . . .. . . . . . . . .. . .. .. . . . . . . .. . . . .. .. .. . . . .. . .. . . . . . . . . . . . . .. . .. . . . .. . . . .. . . . . . .. .. I 4 '

3.2 RE ACTOR IN S'IRUMENTATION .................................................... ! 8 3,4 RADIATION 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 S Y STEM ... .. ... .. ... ... ..... ... .. .... .. .. ...... .. ........ ... .. ... .. . ..... ... .. ... .. 2 9 4.0 SURVEILLANCE REQUIREMENTS ........................................................... 3 1 4.1 CONTROL AND REGULATING RODS............................................ 31 4.2 REACTOR SAFETY SYS'EM ........................................................... 3 2 4.3 RADIATION MONITORING EQUIPMENT.................................. 3 4 4.4 CO NTAIN MENT B UILDING . .. . .. .. . .. ... .... . .. .. ... .. . .. ... . . . .. .. . .. .. . .. ... .. .. . . 3 5 4.5 POO L W A'E R .. .. .. .. . . .. . . . . .. . . . . . . . . . . .. . . . . . . . . . . . . . . . .. . . . . . . . . .. . . . . . . . . .. .. .. .. . . .. . . . .. .. . 3 7 4.6 SCRAM BY PROCESS VARIABLE EFFECT................................. 3 8 4.7- FUEL S U R VEILL ANCE ...... .. .... .. .. .. .. .... .. .... ... .. .... .. .. .. .. .. .... .. .. .. .. .. ... .. . 3 9 5.0 DES IG N FEA'IURES .. .. .... .. .. .. .. .. .. .... . . .. . . .. .. .. .. ... .... .. .. . . .. .. .. .. .. .. .... .. .. ....... . ... .. .. . 4 0 5.1 REACIDR FUEL ....... .. .. .. .. .. ...... .. .. .. .... .. .. ... .. .... .. .. .... . . .. .. . . .. ...... .. ..... .... . 4 0 5.2 REA CIDR CORE... ........ .. ... .... .... .... ... ...... ...... .. ..... ........ . ... ..... .... . ........... 4 0 5.3 REACTOR B UILDING ... .. .... .. ........ .. .... .. .. ....... .. .... .... .. .. .. .... .... .. .. ... ..... 4 1 5.4 FUEL STORAG E ... .... .. .. .. .. .. .. .. .. .... .. .. .. .. ..... .... .. .... .... .. .... .. .. .. . . .. .. ....... .. . 4 1 6.0 A D MINISTR ATIV E CONTROLS ................................................................ 4 2 6.1 ORG ANIZATION 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 ACFION TO BE TAKEN IN THE EVENT A SAFETY ................

LI MIT I S EXCEE DED .. .... .. ....... .. .. . .... . ... .. . .. .. .. .. .. . .... .... ... .. .. ..... . . .. .. . . .. . 4 7 6.6 REPORTING REQUIREMENTS ........................................................ 4 7 6.7 PLANT OP" RATING RECORDS ...................................................... 5 2 6.8 APPROVAL OF EXPERIMENTS..................................................... 5 4

i TECHNICAL SPECIFICATIONS 1.0 DEFINITIONS 1.1 ABNORMAL OCCURRENCES -

An abnormal occurrence is any of the following:

a. Any actual safety system setting less conservative than 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 or system malfunction which could, or threatens to, render  ;

o 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;

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 t

of an unsafe condition in connection with the operation of the reactor; 4

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

- facility.

1.2 CH ANNEL -

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

TS-1 AMENDMENT No. 12

)

.~ ,

~.

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 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 -

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

1.6 CONTAINMENT BUILDING INTEGRITY -

Integrity of the containment building is said to be maintained when all isolation system equipment is operable or secured in an isolating position.

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.

TS-2 AMEN MENT No. 12 1

.____._-_____-______.___O .___ _

O TECHNICAL SPECIFICATIONS 1.8 EXCESS REACTIVITY -

Excess reactivity is that amount of reactivity that would exist if all control rods (control and regulating) were moved to the maximum 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 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 -

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 operating.

1.12 OPER ABLE -

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

1.13 OPER ATING -

Operating means a component or system is performing its

-intended function.

TS-3 MIDDIENT NO.12

' TECHNICAL SPECIFICATIONS -'

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

1.16 REACTOR OPERATING -

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

1.17 REACTOR SAFETY 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 RE ACTOR SECURED -

l The reactor is secured when:

-(l) It contains insufficient fissile material or -moderator present in the reactor, adjacent experiments or control rods, to attain criti. slity under optimum available conditions of moderation and reflection, 'or (2) A combination of the following:

a. The minimum number of neutron absorbing control rods are fully inserted or other safety devices are in the shutdown position, as required by technical specifications, and-TS-4 AMENDMENT No. 12 1

7 o

- TECHNICAL SPECIFICATlONS-b, The console key switch is in the off position and the -

key is removed from the lock, and

c. 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

d. No ' experiments in or near the reactor are being moved or serviced that have, on movement, a reactivity worth-exceeding that maximum value allowed for a single experiment or one dollar, whichever is smaller.

l 1.19 REACTOR SHUTDOWN -

The reactor is shut down if it is suberitical 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 CONDrrlON -

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

TS-5 AMENDMENT No. 12'

TECHNICAL SPECIFICATIONS '

l.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 rnalfunctions.

1.24 SHALL. 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 s

permission, neither a requirement nor a recommendation.

1.25 SHUTDOWN MARGIN -

Shutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence that the reactor can be made suberitical 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 -

The average over any extended period for each surveillance time interval shall be closer to the normal surveillance time than 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 following:

TS-6 AMENIS S T No. 12 l

TECilNICAL SPECillCATIONS

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 l 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).

1.27 TRUE VALUE .

The true value is the actual value of a parameter.

TS 7 mmm n 12

TECilNICAL SPECIFICATIONS

• 2.0 SAFETY LlhilTS AND LlhilTING SAFETY SYSTEh1 SE'ITINGS 2.1 SAFETY LlhilTS 2.1.1 Safety limits in the forced convection mode of operation.

Applicabillty This specification applies to the interrelated variables associated with core thermal and hydraulle performance with forced convection flow. These variables are:

P = Reactor thermal power W= Reactor coolant Gow rate Ti = Reactor coolant inlet temperature L = lleight of water above the center line of the core Objective To auure that the integrity of the fuel cladding is maintained.

Specification Under the conditions of forced convection flow:

1. The combination of true values of reactor thermal 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 s

exceeded if the point defined by the true values of P 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 less than 24 feet above the center line of the core.

TS-8 AM1WiaT to.12

0 7

6-5-

g 4-1 l

.5 b

=

l 3-2-

1 O i , , , i-0 500 1000 1500 2000 2300 3000 Flow Rate (gal. / min.)

Figure 2.1 T.S. Power-Flow Safety Limit Curve TS-8A AMENTSEN1' NO.12

TECIINICAL SPECIFICATIONS -

3. The true value of the reactor coolant inlet temperature (pool temperature. Tp) shall not be gitater 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 given in Section 3.1.2.2 of the FSAR Supplement for Conversion to Low Enrichment Uranium (LEU) Fuel.

2.1.2 Safety Limits in the natural convection mode of operation.

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

P = Reactor thermal power Tp = Reactor pool temperature L = lleight of water above the center line of the core Oblective To assure that the integrity of the fuel cladding is maintained.

Specification Under conditions of natural convection flow:

1. 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 AMENDMBT No.12

V T11CHNICAL SPECIFICATIONS water level (L) is less than 24 feet above the center line of the core.

3. The reactor shall not be taken critical when the true value of the pool water level (L) is less than 2 feet 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, Bases The criterion for establishing a safety limit with natural convection flow is the onset of nucleate boiling at the hot 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 with a corresponding fuel clad temperature of 118.60C (245.5'F) which is well below the temperature at which fuel 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 l 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 SETTINGS 2.2.1 Limiting Safety System Settings in the forced convection mode of operation.

Applicability i

TS-10 AMENDMEhT NO.12

\

'ECilNICAL 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 j height of water above the center line of the core (L).

Obiective To assure that automatic protective action is initiated in order to prevent a Safety Limit from being exceeded.

Specification Under conditions of forced convection flow the values of the Limiting Safety System Settings shall be as follows:

P = 1.25 MWt (max)

W= 1170 GPM (min)

Ti = 1080F (max)

L = 24.25 ft (min)

Bases The Limiting Safety System Settings that are given in Specification 2.2.1 represent values of the interrelated variables which, if exceeded, shall result in automatic protu:ive action that will prevent Safety Limits from being exceeded during the course of the most adverse anticipated transient. To determine the LSSS given above, an analysis of 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 TS 11 AM E DMENT No. 12

P TECl1NICAL SPECIFICATIONS l

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).

Obiective -

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

Soecification Under conditions of natural convection flow the measured 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)

TS-12 AuNatt to.12

_____7 ,

11?CilNICAL SPECIFICATIONS 'l Bases The Limiting Safety System Settings that are given in  !

l Specification 2.2.2 represent values of the interrelated  ;

vanables 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 operation with full pool level, and b) direct radiation from  !

the core during low pool level operation. The specifications 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).

TS-13 MENDMEhT No.12

i TECilNICAL SPECIFICATIONS 3.0 LIMITING CONDITIONS FOR OPER ATION t 3.1 REACTIVITY Applicability TFese specifications apply to the reactivity condition of the reactor and the reactivity worths of control rods, regulating rod, and experiments.

Oti lective 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. 'lne 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.

TS- 14 MirNtimT to.12

TECilNICAL SPECIFICATIONS *

4. The drop time of each control rod from a fully withdrawn position is less than 1.0 second.

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

6. The reactivity insertion rates of the control rods are less 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. l

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

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

Kind Single Experiment Worth Total Worth Movable (including the 0.1% delta k/k pneumatic rabbit) summed 0.5% delta k/k together for all experiments -

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

10. The total reactivity worth of all experiments shall not be

'l greater than 2.5% della k/k.

Al u n

1. The shutdown margin required by Specification 1 assures that the reactor can be shut down from any operating condition and will remain shutdown after cooldown and xenon decay, even if the highest worth control rod should be in the fully withdrawn position.

TS-15 AMEmim to.12

d TECilNICAL 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 j 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 '

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

5. The requirement for a negative temperature coefficient of reactivity assures that any temperature rise caused by a reactor transient will not cause a further increase in re ac t,ivity.

TS 16 AMEMtHT NO.12 l

_-__-___y ,

y TECilNICAL SPECIFICAT!ONS -

6. The maximum rate of reactivity insertion by the control rods which is allowed in Specification 6 assures 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% delt a k/k per seconJ. which is greater than the maximum .illowed.

7. Limiting the reactivity worth of the regulating rod to a value less 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 Limitt 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 experkr. cat such as the pneumatic rabbit to 0.1% delta k/k assures that the prompt TS- 17 etmDM N NO. 12

o TECIINICAL SPECIFICATlONS l

Jump, which is about 17%, will result in a power below the i 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 l

which is compatible with the allowable excess reactivity and t

the shutdown margin and is consistent with the functional mission of the reactor.

3.2 REACTOR INSTRUMENTATION Applicability l

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

Obiective 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 Reauired Which Required Startup Count Rate 1 All modes (during reactor startup)

Log N (Period) 1 All modes Power Level (Linear N) 2 All modes Reactor Coolant inlet 1 Forced consection Temperature Coolant Flow Rate 1 Forced convection TS 18 MtEtOIENT No.12 l

TECl!NICAL 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 REACIDR SAFETY SYS111M Applicability This specifi:ation applies to the reactor safety system channels.

Objective 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 Reauired Function in Which Ilequired 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

, s 15 see period TS. I 9 MlFM!ENT !O.12

0 TECliNICAL SPECIFICATIONS Reactor Safety Minimum Operating Mode System Component / Channel Required Function in Which Requits.d Reactor Power Level 2 Automatic scram All modes when 2125% of range scale Coolant Flow Rate 1 Automatic scram Forced convec-at 1170 gpm tion above 0.1 MW Seismic Disturbance 1 Automatic scram at All modes 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 line; height; j (2) 2.25 ft above (2) operation core center line with limited water height Pool Temperature 1 Automatic scram All modes 1 1080F Coolant inlet Temperature 1 Automatic scram Forced convec-21080F tion above 0.1 MW Bridge Movement 1 Automatic scram All modes if moved >l inch TS 20 MtLNtMNr to.12

Reactor Safety hiinimum Operating Mode System Component / Channel Required Function in Which Regnited Coolant Gate Opens 2 Automatic scram if Forced cor vec.

either the coolant tion above 0.1 riser or coolant hiW; 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 liigh Voltage Failure i Automatic scram 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" TS 21 AMEMtBT No.12

o .. -

0 TECIINICAL SPECIFICATIONS Ilaits 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 I

temperature channel, pool temperature channel, and pool water level channel provides 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 door 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 EO_UlpMENT Applicability This specification applies to the availability of radiation monitoring equipment which must be operable during reactor operation.

TS 22 MtLw m m. 12

'IT:CilNICAL SPECIFICATIONS '

Oldective 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 l monitor with readouts in the control room.

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 l shall be operating.

3. The reactor shall not be continuously

• operated without 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 continutug 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 wnhin 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 ponable gamma sensitive instruments having their own alarm.

TS 23 AMENN NO.12

'!11 CLINICAL SPECIFICATIONS A continuing evaluation of the stack efnuent will be made using the information recorded from the particulate and gas monitors.

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

Oblective To assure that the containtnent 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 . Fu nc tion

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

, hat reactor operation can system integrity proceed if a failed isolation valve is in the closed (isolated) position.

i TS 24 AMINI' No.12

TECilNICAL SPECIFICATIONS -

Eaulpment/ 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 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 diluted by supply air, through a 100 foot stack on site. Chapter 3

• The public address system can serve as a temporary substitute for reactor evacuation  !*

[ p:

and formation of the Emergency Team during shon periods of maintenance. #

TECliNICAL SPECIFICATIONS of the FSAR describes the system's sequence of operation; Chapter 9 provides the analysis.

3.6 LIMITATIONS OF EXPERIMENTS 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 experiment failure.

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

1. All materials to be irradiated shall be either corrosiori 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 oe used in the reactor, in which a static pressure will exist or in which a pressure buildup is predicted, shall be designed and tested for a pressure exceeding the maximum expected by a factor of 2.

3. Explosive material such as (but not limited to) gunpowder, dynamite, TNT, nitroglycerine, or PETN in quantities <25 mg may be irradiated in the reactor or experimental facilities provided out of core tests indicate that, with the containment provided, no damage to the explosive TS-26 #ptmT No.12

TECilNICAL SPECIFICATIONS '

containers, the reactor, the reactor components or the Co 60 Source shall occur upon detonation of the explosive.

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

1 S. 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 tem,..m ature 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 Mi m m No. 12

i

'lliCIINICAL SPECIFICATIONS Bases Specifications I 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 und 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 MIMtET No.12

..______..____.________._.________.___,7,._

i

'o t

TECHNICAL SPECIFICATIONS l '

3.7 GASEOUS EITLUENTS I

Anolicability

~ '

1 i This specification applies to the routine release of gaseous i

radioactive effluents from the facility.

abjective 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 froni the reactor stack shall be limited to 8 microcuries per second averaged over a year.

Bases Calculations based on a very conservative model, allowing for no atmospheric diludon 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 SYSTF_' '

Applienbility This specification applies to the reactor pool water requirements for operation of the reactor.

TS-29 A M M to. 12

'IliCHNICAL 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.

Specification

1. The conductivity of the pool water shall be maintained at a value of 5 micrombos 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 railliliter, 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.

Radionuclica analysis of the pool water allows early determination of any signift: ant buildup of radioactivity from operation of the reactor or the Cobalt 60 source.

TS 30 N N ! INT No. 12

TECilNICAL SPECIFICATIONS $

4.0 SURVEILLANCE REOUIREMENTS 4.1 CONTROL AND REGULA'nNO RODS Applicability This specification applies to the surveillance requirements for the control and regulating rods.

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

Sascifications

1. The reactivity worth of the regulating rod and each control 1

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  !

t!me shall be determined annually, or if maintenance or  !

j 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 j minute.

3.

The control and regulating rods shall be visually inspected annually, lhic.s 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 M e m No. 12

1 TECl{NICAL 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 measurements 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 SYSTEh1 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.

TS 32 MimIt1EhT NO 12

_a

e .

TECliN1 CAL 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 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 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 M M !ELT No. 12

m-.r - - - - - - - - - - - - - - - - - - - - - - - - -

TECllNICAL SPECIFICATIONS Bases 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 snd safety channel detector positions and current alignments.

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

Obiective I

To assure that the equipment used for menitoring radioactivity is 4 operable and to verit'y the appropriate alarm settings.

Specification

1. The operation of the area radiation monitoring equipment and systems for monitoring airborne radioactivity, and their associated alarm set points, shall be verified prior to reactor startup.

2. All radiation monitoring systems shall be calibrated semiannually.

TS-34 AM m mim T No. 12

~.

TECHNICAL SPECIFICATIONS t

Bases 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 lon 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 detected.

4.4 CONTAINMENT BUILD 1NG Applicability .

This specification applies to the surveillance requirements for the containment bui' ding.

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 MIENRET No.12

_y,____ _ _ _ - - - -

c.

17.CHNICAL 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 L containment building or its penetrations that could affect building containment capability' shall be tested to verify containment requirements.

5. The emeyncy exhaust system including the initiating I

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.

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

TS-36 #!ENRIDR NO.12

TECHNICAL SPECIFICATIONS *

'?Als.1 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. i Any additi'ons, 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.

Obiective-To assure high quality pool water and to monitor the radioactivity in the pool water in order to verify the integrity-of the fuel

- -- - cladding. - -- --

TS-37 #ENRIENT to.12 I

TECHNICAL SPECIFICATIONS..

Specification

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

2. The radioactivity in the pool- water shall be analyzed weekly.

Bases Surveillance of water conductivity assures that changes that could accelerate corrosion have not occurred. Radionuclide analysis of the pool water samples will allow early determination of any--

sign'ificant buildup of radioactivity from operation of the reactor or the Co 60 source.

'4.6 SCRAM BY PROCESS VARIABLE EFFECT l

L Anolicability-This specification. applies to the surveillance requirements applied to process variable scrams.

Ob_lective

-To assure that a Safety Limit is noi exceeded.

Soccification Following a reactor scram caused by a process variable, the reactor shall not be operated until an evaluation -has 'been made to

' determine if a Safety Limit was exceeded, the cause of the scram, the effects of-operation to the -scram point and the appropriate action to be taken.

_=

TS-38 M E D M NO. 12 ,

i

, _ _a

.~ .

y 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 SURVEII I ANCE Applicability This specification applies to the surveillance requirements for reactor fuel.

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

Specification Visual inspection of a representative sample of reactor fuel elements shall be pcrformed every two years.

Bases

-The--inspection of reactor fuel assures that fuel elements; -when used in the core, will perform as ' designed, i

TS-39 ens m.12

, m --,v.r ,_ . -

TECHNICAL SPECIFICATIONS 5.0 DESIGN FEARJRES 5,I REACIDR RJEL 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) U 3Si2, clad with aluminum. There shall be 18 plates per element with 16 containing fuel and two outside plates of aluminum. There shall be 200 . 5.6 1 grams of Uranium-235 per element.

2. Half-element: same as a standard fuel element except each plate has one half the uranium loading.

3. Variable-load element: same as Specification I 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 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 used, and cores from 16 elements to 28 elements may contain 2 half-loaded elements.

TS-40 #!ENRI N No. 12 1

TF.ClINICAI. 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 l 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 iri the FSAR. The analysis results and flus measurements for LEU corcs with fewer than 20 standard elements to be operated above 100Kw must be resiewed and approved by the NRC prior to operation with the smaller cores.

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

5.4 FUEL. STOR AGE All reactor luel element storage facilitics shall be designed in geometrical configuration so that keff is less than 0.85 under quiescent flooding with water.

TS-41 MIENTIENT E 12 1

t TECHNICAL SPECIFICATIONS 6.0 ADMINISTRATIVE CONTROLS 6.1 : ORGANIZATION AND MANAGEMENT

1. The reactor facility shall be an integral part of the Radiation Laboratory of the University of Massachusetts Lowell. The 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 of the graduate faculty, and it shall be superviked 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 A>1EN MENT NO. 12

1

, ~

. w,  !

Ch ccollor 1

UNIVERSITY vice- RADI ATION SA FETY  ;

. Chancellor COMMgTTEE  !

ACADEMIC AFFAIRS AUDIT REACTOR

___ SAFETY

-""~~~

SUB-COMMiiTEE DIRECTOR REYlEW

.......... ......... ... RADIATION LABORATORY RADIATION SAFETY OFFICER MANAGER REACTOR g RADIATION SUPERYlSOR 4 SERVICES V

STAFF - - - . . . . . . STAFF g Figure 6.1 TS Organizational Chart For The University of Massachusetts g Lowell Radiation Laboratory 52 a

8

• 4 a

':-h-....

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 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 cc,asistent with public safety and within the terms of the facility license. The Subcommittee shall report to the University Radiation Safety Committee which has overall authority in the use of all radiation l- sources at the University.

2. The responsibilities of the Subcommittee include, but are not 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 l

facility systems or equipment, - procedures, and-operations.

d. Determination of whether a proposed change, test, or ,

experiment would constitute an unreviewed safety TS-43 M a a s T in 12 l

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

c. 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 TS 44 etENRIENT NO. 12

111CHNICAL 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 PROGDURES Written procedures, reviewed and approved- by the Reactor Safety Subcommittee 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 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 ch'anges.

4. Emergency conditions involving potential or actual release of j 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 M1ENDMD4T No.12

TECHNICAL SPECIFICATIONS 8.- Radiation control, for which procedures shall be maintained and available to all operations personnel.

9. Receipt, inspection, and 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 l

subsequently reviewed by the Reactor Safety Subcommittee.

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

1. The Reactor Supervisor or his designec 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 shall report the occu.Tence 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-4 6 - AMmmmT No.12

b TECHNICAL SPECIFICATIONS e

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 1 adiminstrator.

TS-47 MimmmT No.12

_]

5 u TEC11NICAL SPECIFICATIONS

l. 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 i

Region i Administrator of:

a. Aray accidental 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 d: fined 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 M N IENT No. 12

t i

TECHNICAL SPECIFICATIONS -

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

3. Unusual - Events.

L_

A written report shall be forwarded within 30 days in the cvent 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-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 shall include. the following information:

TS-49 #1mmmT No.12

,1 TECliNICAL 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 s 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 FR 50 including a summary of the safety evaluation of each,

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

TS-50 MIENDMaT NO.12 1

TECIINICAL 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 effectiw; control of the licensee as measured at or prior to the point of such release or discharge.

Liquid Waste (Summarized on a monthly basis)

(1) Total gross beta radioactivity released (in curies) during the reporting period. I (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 Gow of 2.7 x 108 cm3/ day.

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 estimated activity (in curies) discharged during the reported period, TS-51 AMENINENT No.12 i

i

._ . _.- . ___ . . . - -. _. _ ..~ . . _ - - - _ . . . . . - _ - , ,

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 (i_f shipped off-site).

6,7 . PLANT OPERATING RECORpl 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 convenient for review and shall be retained as indicated:'

TS-52 AMENRIENT NO.12

's

- ~ - , , , . , 'h.

,, _ -- , ,- ,. - , , - _ - 7-.- -.-, - ,

/

TECHNICAL SPECIFICATIONS

1. Records to be retained for a period of at least five yearsi

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 l

l e. Equipment and component surveillance activities.

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

I

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 AMDDDT NO.12

7, _

TECHNICAL SPECIFICATIONS a

4 y

6,8 - APPROVAL OF EXPERIMENTS 1,

- All. proposed experiments using the reactor shall be  ;

evaluated by the experimenter and a staff member who- has; been1 approved by the Reactor Safety Subcommittee.- The j 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 4

Specifications, and 10 CFR 20. If the experiment complies

^

with the above provisions -it shall be submitted by the c Reactor Supervisor to the Reactor Safety Subcommittee for 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 ,

j changes in. temperature, pressure, chemical composition. or radiolytic decomposition;

c. Any- physical .or chemical interaction that could occur.

4 with the reactor components;

~

d. Any1 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 i

evaluation of the experiment--shall be made by the Reactor-Safety Subcommittee. The -Subcommittee evaluation shall 1

i consider:

o 4

TS-54 MIENIMNT No.12

a 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 i

d. A procedure for the performance of the experiment. A favorable Subcommittee evaluation 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 assumptwns 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 AMEMBT NO.12 a

TECHNICAL SPECIFICATIONS building,100% of the gases or aerosols will escape;

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 magniturie 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.

TS-56 MIENINENT No.12

. _ _ _ _ _ _ _ _ _ _ _ _