ML20137A916

Environ Assessment Concluding That EIS for Proposed Amend to License R-125 Unnecessary.No Significant Environ Impact Finding Sufficient.Environ Considerations & Licensee Tech Specs Encl

ML20137A916
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Site:	University of Lowell
Issue date:	10/04/1985
From:	NRC
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ML20137A862	List: ML20137A858 ML20137A871 ML20137A890 ML20137A906 ML20137A916
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NUDOCS 8511260138
Download: ML20137A916 (68)

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! o UNITED STATES

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8 n NUCLEAR REGULATORY COMMISSION g E WASHINGTON, D. C. 20655

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ENVIRONMENTAL ASSESSMENT FOR THE

. TRAINING AND RESEARCH REACTOR OF THE UNIVERSITY OF LOWELL LICENSE NO.' R-125 DOCKET NO. 50-223 Description of Proposed Action This Environmental Assessment is written in connection with the proposed rerewal for 30 years of the operating license of the University of Lowell reactor (ULR) located on the North Campus of the University in Lowell, Massachusetts, in response to a timely application from the licensee dated February 14, 1985, as supplemented. The proposed action would authorize continued operation of the reactor in the manner that it has been operated since facility license No. P-125 was issued in 1974. Currently, there are no plans to change any of the structures or operating characteristics associated with the reactor during the renewal period requested by the licensee.

Need for the Proposed Action The operating license for the facility was due to expire in April 1985.

The proposed action is required to authorize continued operation so that the facility can continue to be used in the licensee's mission of education and research.

Alternatives to the Proposed Acticn As required by Section 102(2)(E) of NEPA (42 U.S.C.A. 64332(2)(E)), the

. _. staff has considered possible alternatives te the proposed action. The only reasonable alternative to the proposed action that was considered was not renewing the operating license. This alternative would have led to cessation of operations, with a resulting change in status and a likely small impact on the environment. From the standpoint of environmental impact, there are no appropriate alternatives to the proposed action.

Environmental Impact of Continued Operation The ULR operates in an existing shielded water tank inside an existing multiple-purpose building. No new construction is associated with continued operation of the reactor and there is no change in reactor operating conditions or practices. Therefore, this licensing action would lead to no change in the physical environment.

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! 0511260138 851121 PDR ADOCK 05000223 P PDR l

s Based on the review of the specific facility operating characteristics thatareconsideredforpotentialimpactontgeenvironment,assetforth in the staff's Safety Evaluation Report (SER) for this action, it is concluded that renewal of this operating license will have an insignificant environmental impact.

Argon-41, a product from neutron irradiation of air during operation, is the principal airborne radioactive effluent from the ULR during routine operations. Conservative calculations by the staff, based on the tctal amount of Ar-41 released from the reactor during a year, predict a maximum potential annual whole body dose of less than 1 millirem in unrestricted

areas. Radiation exposure rates measured outside of the reactor facility building are consistent with this computation.

The staff has considered hypothetical credible accidents at the ULR and has concluded that there is reasonable assurance that such accidents will not release a significant quantity of fission products and, therefore, will not cause significant radiological hazard to the environment or the public.

This conclusion is based on the following:

a) the excess reactivity available under the technical specifications is insufficient to support a reactor transient generating enough energy to cause overheating of the fuel or loss of integrity of the cladding,

! b) at a thermal power level of I megawatt, the inventory of fission products in the fuel cannot gererate sufficient radioactive decay heat to cause fuel damage even in the hypothetical event of rapid total less of coolant, and c) the hypothetical loss of integrity of the cladding of the maximum irradiated encapsulated fueled experiment will not lead to radiation exposures in the unrestricted environment that exceed guideline l, values of 10 CFR Part 20.

In addition to the analyses in the SER summarized above, the environmental impact associated with operation of research reactors has been generically evaluated by the staff and is discussed in the attached generic evaluation.

- - This evaluation concludes that there will be no significant environmental impact associated with the operation of research reactors licensed to operate at power levels up to and including 1 MWt and that an Environmental Impact Statement is not required for the issuance of construction permits l

or operating licenses for such facilities. We have deterinined that this generic evaluation is applicable to the centinued operation of the ULR and that there are no special or unique features that would preclude reliance on the generic evaluation.

I NUREG-1139, " Safety Evaluation Report Related to the Renewal of the Operating License for the Training and Research Reactor at the University of Lowell."  !

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Ager.cies and Persons Consulted The staff has obtained technical assistance from the Los Alanos Natinral Leberetory in performing the safety evaluation of continued operation of the ULR facility.

Conclusion and Basis for No Significant Impact Finding Besed en the foregoing considerations, the staff has concluded that there will be no significant environmental impact attributable to this proposed license rerewal. Having reached this conclusion, the staff has further concluded that no Environmental Impact Statement for the preresed action need be prepered and that a No Significant Environcental Impact Finding

- is appropriate.

Dated: October 4, 1985 e

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l ENVIRONMENTAL CONSIDERATIONS REGARDING THE LICENSING OF RESEARCH REACTORS AND CRITICAL FACILITIES Introduction This discussion deals with research reactors and critical facilities which i are designed to operate at low power levels, 2 MWt and lower, and are used primarily for basic research in neutron physics, neutron radiography, isotope production, experiments associated with nuclear engineering, training and as a part of a nuclear physics curriculum. Operation of such facilities will generally not exceed a 5-day week, 8-hour day, or about 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> per year.

Such reactors are located adjacent to technical service support facilities

.with convenient access for students and faculty.

Sited most frequently on the campuses of large universities, the reactors are usually housed in already existing structures, appropriately modified, or placed in new buildings that are designed and constructed to blend in with existing facilities. However, the environmental considerations discussed herein are not limited to those which are part of universities.

Facility There are no exterior conduits, pipelines, electrical or mechanical. structures or transmission lines attacned to or adjacent to the facility other than for utility services, which are similar to those required in other similar facilities, specifically laboratories. Heat dissipation is generally accom-plished by use of a cooling tower located on the roof of the building. These cooling towers typically are on the order of 10' x 10' x 10' and are comparable to cooling towers associated with the air-conditioning systems of large office buildings.

Make-up for the cooling system is readily available and usually obtained from the local water supply. Radioactive gaseous effluents are limited to Ar-41 and the release of radioactive liquid effluents can be carefully

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monitored and controlled. Liquid wastes are collected in storage tanks to allow for decay and monitoring prior to dilution and release to the sani-tary sewer system. Solid radioactive wastes are packaged and shipped off-site for storage at NRC-approved sites. The transportation of such waste is done in accordance with existing NRC-DOT regulations in approved shipping containers.

Chemical and sanitary waste systems are similar to those existing at other similar laboratories and buildings.

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Environmental Effects of Site Preparation and Facility Construction Construction of such facilities invariably occurs in areas that have already t been disturbed by other building construction and, in some cases, solely within an already existing building. Therefore, construction would not be expected to have any significant effect on the terrain, vegetation, wildlife or nearby waters or aquatic life. The societal, economic and esthetic impacts of construction would be no greater than those associated with the construction of a large office building or similar research facility.

Environmental Effects of Facility Operation Release of thermal effluents from a reactor of less that 2 MWt will not have

_a significant effect on the environment. This small amount of waste heat is generally rejected to the atmosphere by means of small cooling towers. Ex-tensive drift and/or fog will not occur at this low power level.

t Release of routine gaseous effluents can be limited to Ar-41, which is generated by neutron activation of air. Even this will be kept as low as practicable by using gases other than air for supporting experiments. Yearly doses to unre-stricted areas will be at or below established guidelines in 10 CFR 20 limits.

. Routine releases of radioactive liquid effluents can be carefully monitored and controlled in a manner that will ensure compliance with current standards. Solid radioactive wastes will be shipped to an authorized disposal site in approved i

containers. These wastes should not require more than a few shipping containers a year.

Based on experience with other research reactors, specifically TRIGA reactors operating in the 1 to 2 MWt range, the annual release of gaseous and liquid effluents to unrestricted areas should be less than 30 curies and 0.01 curies, respectively.

No-release of potentially hamful chemical substances will occur during nomal operation. Small amounts of chemicals and/or high-solid content water may be released from the facility through the sanitary sewer during periodic blowdown of the cooling tower or from laboratory experiments.

Other potential effects of the facility, such as esthetics, noise, societal or impact on local flora and fauna are expected to be too small to measure.

Environmental Effects of Accidents Accidents ranging from the failure of experiments up to the largest core damage and fission product release considered possible result in doses that i are less than 10 CFR Part 20 guidelines and are considered negligible with

! respect to the environment.

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Unavoidable Effects of Facility Construction and Operation The unavoidable effects of construction and operation involve the m'a terials used in construction that cannot be recovered and the fissionable material used in the reactor. No adverse impact on the environment is expected from either of these unavoidable effects. l Alternatives to Construction and Operation of the Facility To accomplish the objectives associated with research reactors, there are no suitable alternatives. Some of these objectives are training of students in the operation of reactors, production of radioisotopes, and use of neutron

-ard gamma ray beams to conduct experiments.

Long-Term Effects of Facility Construction and Operation The long-term effects of research facilities are considered to be beneficial as a result of the contribution to scientific knowledge and training. Because of the relatively small amount of capital resources involved and the small impact on the environment, very little irreversible and irretrievable commit-ment is associated with such facilities.

Costs and Benefits of Facility Alternatives ,

The costs are on the order of several millions of dollars with very little environmental impact. The benefits include, but are not limited te, some combination of the following: conduct of activation analyses, condunt of neutron radiography, training of operating personnel and education of students.

Some of these activities could be conducted using particle accelerator 3 or radioactive sources which would be more costly and less efficient. There is no.. reasonable alternative to a nuclear research reactor for conducting this spectrum of activites.

Conclusion The staff concludes that there will be no significant environmental impact associated with the licensing of research reactors or critical facilities designed to operate at power levels of 2 MWt or lower and that no? environmental impact statements are required to be written for the issuance of construction permits or operating licenses for such facilities.

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Docket No. 50-??3 DISTRIBUTION:

Docket File DEisenhut i NRC PDR liDenton  !

OCS OELD NSIC EJordan Mr. Leo Beghian, Director Region I JPartlow Nuclear nter PAnderson (3) TParnhart (4)

University f Lowell HBernard CMiles, OPA One Universi v Avenue HBerkow ACPS Lowell, Massa usetts 01854 CThomas JHyder, LANL DCrutchfield PDiggs/LTremper Dear Mr. Beghian. IIThompson DTondi

SUBJECT:

RENEWAL " UNIVERSITY OF LOWELL OPERATING LICENSE NO. P-125 The Nuclear Pogulatory ommission has issued Arrendment No. 9 to Facility Operating License No. R- 25 for the University of Lowell ruclear facility in respense to the applic tion for renewal dated February 14,1985, as supplemented. This amendm .t renews the operating license for thirty years from its date of issua ce.

In accordance with cur practic we have restated the license in its entirety, incorporating all the changes an' amendments made since the issuance of the original license.

Enclosed with the amended license is copy of the Finding of f:o Significant Enviremental Iripact that has been pub ished in the FEDERAL REGISTER, a copy of the Notice of Penewal that is b ng sent to the Office of the Federal Register for put lication, the Environmen 1 Assessment, and the Safety Evaluation Report (h0FEG-1139) associated ith the rerewal.

Sincerely, Hugh L. Thompson, Director Division of Licens g cc: See next page

Enclosures:

1. Amendment No. 9

2. Notice of Renewal

3. Notice of Finding of f!o Significant Environmental Impact

4. Environmental Assessment

5. Safety Evaluation Report NUREG-1139) - J0 I

DL1 DL: Pf  : LD DL:AD/SA M DCrutchfield DL:DIR Pali rson:ac HBem,ard ow .it/s @ HThompson 10/lb/85 10/c /85 10/n/85 10/ 85 10/7.2/ -5 10/ /85 10/ /85

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l APPENDIX A E

FACILITY OPERATING LICENSE NO. R-125 TECHNICAL SPECIFICATIONS FOR THE UNIVERSITY OF LOWELL Date: October 18, 1985 l

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's 3 s, IV. UNIVERSITY OF LOWELL TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Page No.

1.0 DEFINITIONS 1.1 Abnormal Occurrences................................. IV-1 1.2 Channe1.............................................. IV-1 1.3 Channel Calibration.................................. IV-2 1.4 Channel Check..'...................................... IV-2 1.5 Channel Test......................................... IV-2 1.6 Containment Building Integrity....................... IV-2 ,

, 1.7 Control Rod.......................................... IV-2 1.8 Excess Reactivity.................................... IV-2 1.9 Experiment........................................... IV-3 1.10 Measured Va1ue....................................... IV-3 1.11 Movable Experiment................................... IV-3 1.12 0perable............................................. IV-3 1.13 0perating............................................ IV-3 1.14 Protective Channel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . IV-3 1.15 Reactor Operating Mode............................... IV-3 1.16 Reactor Operating.................................... IV-3 1.17 Reactor Safety System................................ IV-4

_ 1.18 Reactor Secured...................................... IV-4 1.19 Reactor Shutdown..................................... IV-5 1.20 Regulating Rod....................................... IV-5 1.21 Reference Core Condition............................. IV-5 1.22 Safety Channe]....................................... IV-5 1.23 Secured Experiment................................... IV-5 1.24 Shall, Should, and May............................... IV-5 1.25 Shutdown Margin...................................... IV-6 1.26 Surveillance Intervals............................... IV-6 1.27 True Value........................................... IV-6 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SEPTINGS 2.1 Sa f e t y Li mi t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . IV-7 2.1.] Safety Limits in the forced convection mode of ,

operation..................................... IV-7 2.1.2 Safety Limits in the natural convection mode of operation..................................... IV-8 i

TABLE OF CONTENTS (CONT'D)

Pane No.

2.2 Limiting Safety System Settings...................... IV-9 2.2.1 Limiting Safety System Settings in the forced convection mode of operation................. IV-9 2.2.2 Limiting Safety System Settings in the natural convection flow mode of operation............ IV-11 3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Reactivity........................................... IV-13 3.2 Reactor Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . IV-17 3.3 Reactor Safety System................................ IV-18

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3.4 Radiation Monitoring Equipment....................... IV-21 3.5 Containment and Emergency Exhaust System............. IV-22 3.6 Limitations of Experiments........................... IV-24 3.7 Gaseous Effluents.................................... IV-27 3.8 Coolant System....................................... IV-28 4.0 SURVEILLANCE REQUIREMENTS 4.1 Control and Regulating Rods.......................... IV-30 4.2 Reactor Safety System................................ IV-31 4.3 Radiation Monitoring Equipment. . . . . . . . . . . . . . . . . . . . . . . IV-33 4.4 Containment Bu11 ding................................. IV-34 4.5 Pool Water........................................... IV-36

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4.6 Scram by Process Variable Effec t. . . . . . . . . . . . . . . . . . . . . IV-37 i

4.7 Fuel Surveillance.................................... IV-38 5.0 DESIGN FEATURES l 5.1 Reactor Fue1......................................... IV-39 l 5.2 Reactor Core......................................... IV-39 5.3 Reac t or Buil d in g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I V- 40 5.4 Fuel Storage......................................... IV-40 6.0 ADMINSTRATIVE CONTROLS l

i 6.1 Organization and Management.......................... IV-41 l 6.2 Review and Audit..................................... IV-43 l 6.3 Operating Procedures................................. IV-45 6.4 Action to be Taken in the Event of an Abnormal Oc c u rr e n c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I V-4 6 6.5 Action to be Taken in the Event a Safety Limit is Exceeded.......................................... IV-47 ff

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TABLE OF CONTENTS (CONT'D)

Page No.

6.6 Reporting Requirements............................... IV-48 6.7 Plant Operating Records.............................. IV-52 6.8 Approval of Experiments.............................. IV-53 Included in this document are the Technical Specifications and the

" Bases" for the Technical Specifications. These bases, which

. provide the technical support for the individual technical specifications, are included for information purposes only. They are not part of the Technical Specifications, and they do not constitute limitations or requirements to which the licensee must adhere.

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, -a 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 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 l

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

g. Conditions arising from natural or offsite manmade events i

l that affect or threaten to affect the safe operation of the facility.

1.2 C11ANNEL - A channel is the combination of sensor, line, amplifier, and output devices which are connected for the i

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purpose of measuring the value of a parameter. Such a channel is also refered as a measuring channel.

1.3 CHANNEL CALIBRATION - A channel calibration is an adjustment l 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 '- A channel test is the introduction of a I

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 l

isolating position.

1.7 CONTROL ROD - A control rod is a device fabricated from neutron absorbing raterial which is used to establish neutron

, flux changes and to compensate for routine reactivity losses.

I A control rod is coupled to its drive unit allowing it to perform a safety function when the coupling is disengaged.

l 1.8 EXCESS REACTIVITY - Excess reactivity is that amount of reactivity that would exist if all coatrol rods (control and f

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regulating) were moved to the maximum reactive condition from the point where the reactor is exactly critical (keff=1).

1.9 EXPERIMENT - Any operation, hardware, or target xcluding devices such as detectors, foils, etc.), 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.

i 1.11 MOVABLE EXPERIMENT - A movable experimenc is one where it is intended that the entire experiment may be moved in or near the core or into or out of the reactor while the reactor is operating.

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

, ~ 1.13 OPERATING - Operating means a component or system is performing its intended function.

1.14 PRuiEcu VE CHANNEL - A protective channel is a safety channel in the reactor safety system which is not a measuring channel.

1.15 REACTOR OPERATING MODE - Reactor operating mode refers to i

the method by which the core is cooled, either natural l

convection mode of operation or forced convection mode of operation.

1.16 REACTOR OPERATING - The reactor is operating whenever it is 1

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l not secured or shutdown. j 1.17 REACTOR SAFETY SYSTEM - the reactor safety system are those syrtems, 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 - A reactor is secured when:

(1) It contains insufficient fissile material or moderator present in the reactor, adjacent experiments or control rods, to attain criticality under optimum available conditions or 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 shutdown position, as required by technical

. specifications, and

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b. The console key switch is in the off positon 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 i drives unless they are physically decoupled from the ,

control rods, and

d. No experiments in or near the reactor ar* being l

moved or serviced ,that have, on movement, a reactivity worth exceeding that maximum value allowed for a single experiment or one dollar l

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whichever is smaller.

1.19 REACTOR SHUTDOWN - The reactor is shut down if it is subcritical by at least .7% delta k/k in the Reference Core Condition plus the absolute reactivity worth the reactivity worth of all experiments.

! 1.20 REFERENCE CORE CONDITION - 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 SAFEIT CHANNEL - A safety channel is a measuring or protective channel in the reactor safety system.

1.23 SECURED EXPERIMENT - A secured experiment is an experiment, 1

} experimental facility, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means. The retaining forces must be substantially greater than those to which the experiment 9

might be subjected by hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the ,

experiment, or by forces which can arise as a result of ,

credible malfunctions. r 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 permission, neither a f

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requirement nor a recommendation.

1.25 SHUTDnWN MARGTN - Shutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence that the reactor can be made subcritical by means of control and )

i safety systems starting from any permissible operating i condition although the most reactive rod is in its most reactive position, and that the reactor will remain suberitical without further operator action.

1.26 SURVEILLANCE Ih"rERVALS - Allowable surveillance intervals shall not exceed the following:

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).

1.27 TRUE VALUE - The True Value is the actual value of a parameter.

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTDI SETTINGS 2.1 SAFETY 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:

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 Obiective To assure that the integrity of the fuel cladding is

~- maintained.

Specification Under the conditions of forced convection flow:

1. The true value of the reactor thermal power (P) shall not exceed 4 MW.

2. The true value of the reactor coolant flow rate (W) shall not be less than 1200 gallons per minute (GFM).

3. The true value of the pool water level (L) shall not be less than 24 feet above the center line of IV-7

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the core.

4 The true value of the reactor coolant inlet temperature (pool temperature, T p ) shall not be greater than 110*F.

Bases In the region of full power operation, the criterion used to establish the safety limit was the onset of nucleate boiling at the hot spot in the hot channel. The analysis is given in Paragraph 9.1.1 of the FSAR.

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 natura1 convection flow. These variablea ire:

P = Reactor thermal power Tp = Reactor pool temperature L = Height of water above the center line of the core t Obiective l

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 0.66 MW.

l l 2. The true value of the reactor thermal power (P) shall 1

not exceed 1.33 kW when the true value of the pool IV-8

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water level (L) is less than 2 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 110*F.

Bases The criterion for establishing a safety limit with natural convection flow is established as the fuel clad temperature. The analysis of natural convection flow given in paragraph 9.1.1.7 of the FSAR shows that at 0.66 MW the maximum fuel clad temperature is 250'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 nearly 500

times lower 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

! 2.2.3 Limiting Safety System Settings in the forced convection mode of operation.

Applicability l This specification applies to the setpoints for IV-9

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l the safety channels monitoring reactor thermal power (P), coolant flow rate (W), reactor coolant inlet temperature (T1 ), 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 a Safety Limit from being exceeded.

Specification

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

P = 1.25 MWt (max) i s

W = 1250 GPM (min)

Ti = 108'F (max)

L = 24.25 ft (min)

Bases _

f The Limiting Safety System Settings that are l 81 ven in Specification 2.2.1 represent values of the interrelated variables which, if exceeded,

, shall result in an automatic protective actions I

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 int IV-10

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account. These cafety 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 variation of 2*F, i a power level variation of 6%, and a pool water level variation of three inches. (See FSAR 1!-

Paragraph 9.1.2.)

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 channel 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.

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

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

Tp = 108'F (max) or Tp = 108'F (max)

L = 24.25 ft (nin) L = 2.25 ft (min)

A Bases The Limiting Safety System Settings that are gg given in Specification 2.2.2 represent values of -

the interrelated variables which, if exceeded, shall result in automatic protective actions that will prevent undesirable radiation levels on the surface of the pool due to: a) the production and escape of 16N during natural convection mode of operation with a 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 s for natural convection, because the values of the power LSSS would be much higher (620 kW, Paragraph 9.1.2.3.2 of the FSAR) if the specifications were based on Safety Limits rather than on 16N production.

The 16N criterion is not related to fuel clad damage which was the criterion used in establishing the Safety Limits (see Specification 2.1.2).

IV-12 l

l l , . _ _ . . ._, , _ . _ _

t : ,

l 3.0 LIMITING CONDITIONS FOR OPERATION l 3.1 REACTIVITY l Applicability These specifications apply to the reactivity condition .

l of the reactor and the reactivity worths of control rods, regulating rod, and experiments.

Obiective To assure that the reactor can be safely shut down 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, l regulating rod, graphite reflector elements or grid plugs. For low power, <10 kW, physics tests without forced flow, this specification will not apply.

4 The drop time of each control rod from a fully 1

1 e

l IV-13

- . _ - - -. _~

withdrawn position is less than 1.0 second.

5. The isothermal temperature coefficient of reactivity is negative at temperatures >70*F.

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.

I 8. The reactivity insertion rate of the regulating is 1ess 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 Sinnie Experiment Worth Total Worth Movable (includes the 0.1% delta k/k 0.5% delta k/k pneumatic rabbit) together Secured 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.

Bases

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.

2. The maximum allowed excess reactivity of 4.7% delta k/k provides sufficient reactivity to accommodate i IV-14 I

I 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 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 Paragraph 9.1 of the FSAR. Under natural circulation conditions at low power, this requirement does not apply.

4. The control rod drop time required by Specification 4

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 j this situation, which is given in Paragraphs

. 9.1.2.3.1 and 9.1.3 of the FSAR, assumes a 1 second rod drop time.

I l 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 reactivity.

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 IV-15

reactivity insertion. Analysis in Paragraph 9.1.11 of the FSAR shows that a maximum power of less than 1.4 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 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 1

not be exceeded during an operational accident involving the continuous withdrawal of the regulating rod.

- 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 Paragraph 9.1.10 of the i FSAR. Limiting a movable experiment such as the I

pneumatic rabbit to 0.1% delta k/k assures that the prompt 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 IV-16

places a reasonable upper limit on the worth of all i

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.

1 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

i. channels listed in the following table are operable:

Minimum Operating Mode in l Measuring Channel Required Which Required Pool Water Level 1 All modes l

i Startup Count Rate 1 All modes (during reactor stG Log N (Period) 1 All modes Power Level (Linear N) 2 All modes Reactor Coolant Inlet 1 Forced convection Temperature Coolant Flow Rate 1 Forced convection l Reactor Pool Temperature 1 Natural convection i

IV-17

--_. ._.y. .

Bases The neutree detectors assure that measurements of the i

reactor ,swer level are adequately displayed during l 1

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 Idmit.

3.3 REACTOR SAFETY SYSTEM Applicability

. 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 Component / Channel Reauired Function in Which Required Startup Count Rate 1 Prevent blade Reactor startup in withdrawal all modes when N count rate j; 2 cps Reactor Period 1 Automatic reactor All modes scram with j; 3 sec period

! Reg rod prohibit i

j[ 15 sec period i 1

Iv-18 l

l

o Reacter Safety Minimum , Operating Mode System Component / Channel Required Function in Which Required Reactor Power Level 2 Automatic scram All modes when 2,125% of range scale Coolant Flow Rate 2 Automatic scram Forced convection at 1250 gpm above 0.1 MW Seismic Disturbance 1 Automatic scram at All mode's Modified Mercalli Scale IV Primary Piping Alignment 1 Automatic scram Forced convection 15 above 0.1 MW Pool Water Level 1 Automatic scram (1) All modes aboy' at: (1) 24.25 ft 1.25 kW (measured :

above core center value) (2) Operatis line (2) 2.25 ft below 1.25 kW above core center (measured value) line

, Pool Temperature 1 Automatic scram All modes 2;108'F Coolant Inlet Temperature 1 Automatic scram Forced convection

L 108'F above 0.1 MW Bridge Movement 1 Automatic scram All modes if moved >l inch Coolant Gates Open 1 Automatic scram if Forced convection either the coolant above 0.1 MW; down

- riser or coolant comer flow pattern downcomer gates open Coolant Gate Opens 1 Automatic scram if Forced convection the coolant riser above 0.1 Mw; crost gate opens pool flow pattern Detector High Voltage 1 Automatic scram if All modes Failure Voltage <500V Thermal Column Door Open 1 Automatic scram All modes Truck Door and/or Air 3 Automatic scram All modes Lock Integrity Manual Scram Button 1 Manual scram All modes

" Reactor On" Key-Switch 1 Manual scram All modes I"-19 V

Bases The inhibit function on the startup channel assures 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 tempert ture channel, and pool water level channel provides the redundant protection to assure that a Safety Limit is not exceeded.

Automatic protective 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 i

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.

O IV-20

, , . - - y . , .m . . , _ , . - - - --- . -

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

l l

3.4 RADIATION MONITORING EQUIPMENT Applicability This specification applies to the availability of radiation monitoring equipment which must be operable during reactor operation. l Objective 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 a readout 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 shall be operating.

3. The reactor shall not be continuously

• operated without a minimum of one radiation monitor on the expsrimental level of the reactor building and one l monitor over the reactor pool operating and capable l

of warning personnel of high radiation levels.

• In order to continue operation of the reactor, replacement of an l inoperative monitor must be made within 15 minutes of recognition i of failure, except that the reactor may be operated in a l

steady-state power mode if the installed systems are replaced with portable gamma-sensitive instruments having their own alarm.

TV-21

j .

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.

A continuing evaluation of the stack effluent will be rsde 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 j _, possible release of radioactive materials resulting from reactor operation.

l Specification l

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 l

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

I IV-22

. _ - . _ _ _ . _ _ _ __ _ . _ _ . . _ _ . _ _ _ _ . . _ . ~ _ _ _ _ . _ . _ _ _ _ _ . - - _ _ _ _ _ _ . . _ . _ . . _ . . _ _ . _ _ _ _

l 1

Equipment / Condition Function l

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 air-borne 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

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

IV-23 l

l

• 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 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 an'd 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 4

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 1

pressure buildup is predicted, shall be designed and tested for a pressure exceeding the maximum expected IV-24

--. - - m , --., , , , -,e -- _ . , - , _ . + , _ , , -, - , - , , - , - , , ---.yy-, , . - - - - , ,

. l

: l

. l by a factor of 2.

i

3. Explosive material such as (but not limited to) gunpowder, dynamite, TNT, nitroglycerine, or PEIN in I

quantities <25 mg may be irradiated in the reactor i or experimental facilities provided out'-of-core '

tests indicate that, with the containment provided, no damage to the explosive containers, the reactor or the reactor components shall occur upon detonation of the explosive.

4. Explosive materials, in any quantity, shall not be allowed in the reactor pool or experimental facilities without 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 irradfated 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 1

l i IV-25 i

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

operation of the reactor.

8. Cyrogenic liquids shall not be used in any experiment within the reactor pool. l l

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

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 those involving the Co-60 Source and, along with the reactivity restriction of pertinent specification 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 i reactor by, for example, tilting the flux in a way that j could affect the peaking factor used in the Safety Limit calculations. Review of the experiments using the j appropriate LCO's and the Administrative Controls of l Section 6 assures that the insertion of experiments will l

not negate the consideration implicit in the Safety Limits.

Specification 8 insures NRC review of experiments TV-26 i

I

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

i a containing or using cyrogenic materials. Cyrogenic 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 procedur'al between the reactor and the cobalt source during reactor operation.

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

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

Bases Calculations based on a very conservative model, j

allowing for no atmospheric dilution of the gaseous effluent, have predicted an annual dose of 12 mr to an individual exposed to the effluent on a continual basis l

for an argon-41 release rate of 8 microcuries per IV-27

second. Allowance for even minimal atmospheric turbulence would reduce this dose number by about a factor of three.

3.8 COOLANT SYSTDi Applicability This specification applies to the reactor pool water.

Obiective The objective is to require that the reactor pool water

, be of high purity in order to retard corrosion and monitor the integrity of the fuel cladding.

Specification

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. Analysis 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, IV-28

notification will be made to the USNRC as required by l Section 6.6.2.

Bases Pool water of high purity minimizes the rate of corrosion. Radionuclide analysis of the pool water allows for early determination of any significant buildup of radioactivity from operation of the reactor or the Cobalt-60 source.

I t

l IV-29

~

' l i

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

Obiective To assure the operability of the control and regulating i

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 determined annually, or if maintenance or modification is performed on the mechanism.

3. The control IndTegGrating' rods ~ shall be visually

-inspected annually.

Bases l

The reactivity worth of the control and regulating rods

, is measured to assure that the required shutdown margin f is available, and to provide a means for determining the reactivity worths of experiments inserted in the core.

Annual measurement of reactivit y worths provides a l IV-30 I

l

4 I

J correction for the slight variations expected because of burnup, and the required measurement after any new arrangement of fuel in the core assures that possibly i

altered rod worths will be known before routine l

operation.

! The visual $mspection of the regulating and control rods and the measurements of drive and drop times are made to i assure that the rods are capable of performing properly and within the considerations used in transient analyses in Chapter 9 of the FSAR. Appropriate inspection data

' will be recorded and analyzed for trends. Verification I of operability after maintenance or modification of the control system will ensure proper reinstallation or f

reconnection.

4.2 REACTOR SAFETY SYSTEM Applicability ,

- This specification applies to the surveillance

requirements for the Reactor Safety System.

l Obiective To assure that the Reactor Safety System (RSS) will remain operable and will prevent the Safety Limits from i

being exceeded.

l Specifications l

1. A channel check of each measuring channel in the RSS 1

shall be performed daily when the reactor is in l

operation.

i TV-31

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 calibration prior to 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.

Bases

' The daily channel tests and checks and periodic verifications will assure that the safety channels are 1

operable. Annual calibrations will assure that i

l .

IV-32

l l

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 detectors positions and current

! alignments.

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

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

2. All radiation monitoring systems shall be calibrated semiannually.

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 det ectors and two don chamber detectors l IV-33

~ -

at selected sites throughout the building. The detectors used have been chosen for stability and operational reliability. The large number of detectors I

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 corrected.

4.4 CONTAINMENT BUILDING Applicability 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 i it is less than ambient atmospheric pressure.

2. The containment building isolation system including the initiating system shall be tested semi-annually.

The test shall verify that valve closure is achieved I

in <2.5 seconds after the initial signal.

3. An integrated leakage rate of the containment IV-34

m building as-is* ahall be performed at a pressure of at least 0.5 psig at intervals of 5 years to verify l 1eakage 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.

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

Valve closure time was chosen to te 1/2 the time required for a given sample of air to travel from the l

• Non-routine maintenance or repair for the purpose of l reducing containmenc leakage shall not be performed price en the leak test.

l I

l IV-35 r

_ _ _ _ _ _ _ . . - - . _ _ . _ _ . _ _ _ . . _ , __ .. . _ . _ _ _ ~ - _ _ _ _ . _ - _ _ . _ _ _

i first to the second valve in series in the exhaust line under regular flow conditions. Semi-annually is considered a reasonable frequency of testing.

I' 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, 4

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 i

j building or its penetrations shall be tested to verify that such work has not adversely affected the leaktightness of the building.

j Surveillance of the emergency exhaust system and the various filters will verify that these are functioning.

l i

See Chapters 3 and 7 of the FSAR.

i- 4.5 POOL WATER Applicability This specification applies to the surveillance

! requirement for monitoring the quality and the radioactivity in the pool water.

Obiective To assure high quality pool water an'd to monitor the radioactivity in the pool water in order to verify the integrity of the fuel cladding.

IV-36 I. _ _ _ - - -

_ - . - _ _ _ - - - . - . - - - . ~ . . _ - - . . ._ - -

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 significant buildup of radioactivity from operation of the reactor or the Co-60 source.

4.6 SCRAM BY PROCESS VARIABLE EFFECT Applicability This specification applies to the surveillance requirements applied to process variable scrams.

Obiective 1

To assure that a Safety Limit is not exceeded.

i Specification 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 i

to be taken.

[

l 0

IV-37

Bases This specification assures that if a safety limit should be exceeded as a result of a malfunction of process variable, the fact will be known.

4.7 FUEL SURVEILLANCE Apolicability 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 performed every two years.

Bases The inspection of reactor fuel assures that fuel elements, when used in the core, will perform as designed.

j l

r TV-38

i 5.0 DESIGN FEATURES J

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 highly enriched (93%) uranium-aluminum alloy or powder mixture fuel, clad with aluminum. There shall be 135 ( f,4) grams of uranium-235 per element. There shall be 18 plates per fuel element.

2. Half-element: same as a standard fuel element i

except each plate has one half the uranium loading. ,

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

5.2 REACTOR 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 26 standard fuel elements arranged symmetrically around four safety control blades as shown in Figure 4.23 of the FSAR.

2. Cores from 23 standard elements to 30 elements may be used, and cores from 24 elements to 30 elements may contain 2 half-loaded elements.

l 3. Cores with an internal fuel element replaced by a I radiation basket may be operated under forced i

convection only af ter flux measurements made under TV-39

forced 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

1. The reactor shall be housed in the reactor building, designed fc: containment.

5.4 FUEL STORAGE All reactor fiel element storage facilities shall be designed zu a geometrical configuration where ke rf is less than 0.8 under quiescent flooding with water.

i.,

l I

TV-40

6.0 ADMINISTRATIVE CONTROLS

6.1 ORGANIZATION AND MANAGEMENT

l. The reactor facility shall be an integral part of the Radiation Laboratory of the University of Lowell. The reactor shall be related to the

, University structure as shown in Chart 6-1.

2. The reactor facility shall be under the direction of the Director of the Radiation laboratory, who shall f, be a member 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 1

Subcommittee.

3. There shall be a Radiation Safety Officer 4.

responsible for the safety of operations from the i standpoint of radiation protection. He does not report to the line organization responsible for c reactor operations, but rather to the Vice-President i

for Academic Services and Technical Research (see I

Chart 6-1).

4 An operator or senior operator licensed pursuant to 10 CFR 55 shall Fe present at the controls unless the reactor is secured as defined in these l

l IV-41

1 i

I

o-u 4

University l Associate Vice Radiation Safety

}

1- President for -

Consmittee r Research I I i . i '

softetten Reactor Safety i Audit ll neview Sub-cossnittee h I.eberatory strecter * ,

l Radiation n,.ca.,

j Safety superviser i

i w .

i < I l i E

" i Operating i

' Staff 1

i I

l 1

i .

Figure 6.1 ULR Organizational Structure I

j ,  :

I l l i i ,

! i

i l ,

! t

specifications.

5. A Licensed Senior Operator shall be on the console or readily available on call within the Pinanski Building whenever the reactor is in cperation.

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 4

- 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

- 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 i

., emergency operating and maintenance procedures and records.

i

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, f

I l

IV-43 i

_ - _ _ . . _ _ _ _ . ~ . . . _ . _ _

test, or experiment would constitute an unreviewed safety question requiring a change to

the Technical Specifications or facility license.

e. Review of all violations of the Technical 4

Specifications and NRC Reguistions, and significant violations of internal rules or 1

procedures, with recommendations for corrective

. , action to prevent recurrence.

f. Revi_ew 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 and another of whom shall i

i be the Reactor Supervisor. The Subcommittee shall f [, 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 function $ng of the Subcommittee.

IV-44 i

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. . - . - -- , - _ . , . . . _ . -____., ..- __ _-,._---____. _ _. ~ . - - . ~ _ _ - - _ . ~ . - . .

. s t

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 Radiation Safety Officer or his designee, the Reactor Supervisor or his designee, and the chairman or his designee.

) 6. The Subcommittee shall meet at least quarterly.

1

- 6.3 OPERATING PROCEDURES 4

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 i

j 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.

i i,

2. Installation or removal of fuel elements, control

\

l rods, experiments and experimental facilities.

3. Actions to be taken to correct specific and I

potential malfunctions of systems or components, l

, including responses to alarms, suspected primary i coolant system leaks, and abnormal reactivity i

l changes.

4 Emergency conditions involving potential or actual l

l release of radioactivity, including provisions for evacuation, re-entry, recovery, and medical support.

IV-45 l

. _ . . __ _ . _ _ _ . _ ___ . _ . . _ _ . _ _ _ _ _ . _ . _ _ _ . _ _ . _ . _ _ . ~ . . . _ . -

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

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

7. Civil disturbance on or near campus.

8. Radiation control 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.

j 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 J

shall be documented and subsequently reviewed by the Reactor Safety Subcommittee.

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

The Reactor Supervisor shall be notified promptly and

1. corrective action shall be taken immediately to place the facility in a safe condition until the

! causes of the abnormal occurrence are determined and I

corrected.

2. The Reactor Supervisor shall report the occurrence i IV-46 I _ _ _ _ .- _

. - _ ~ . - . - - . - _ . - .

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.

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 EXCEEDED 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. 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 l IV-47 l

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 REQUIREMEhTS 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 ComEission, Attn: Document Control Desk, Washington, D.C. 20555, with a copy to the Region I Administrator.

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 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 l

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 IV-48

abnormal occurrence, as defined in Section 1.1. The report shall:

1) Describe, analyze, and evaluate safety implications;

2) Outline the measures taken to assure that the cause of the condition is determined;

3) 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;

4) Evaluate the safety implication of the incident in light of the cumulative experience obtained ,

from the record of previous failure and malfunctions of similar systems and components.

,, 3. Unusual Events

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

1) Discovery of any substantial errors in the transient or accident analyses or in the methods used for such analyses, as described in the i

safety analysis or in the bases for the technical specifications;

2) Discovery of any substantial variance from performance specifications contained in the IV-49

~

technical specifications and safety analysis.

3) Discovery of any condition involving a possible t

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: -_

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.

4

d. Discussion of the major maintenance operations performed during the period, including the IV-50

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 j procedures, tests, and experiments carried out i

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.

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 disch~arged to the environs beyond the effective 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.

(2) Total radioactivity released (in curies) for specific nuclides, if the gross beta IV-51 l

. - - - , - - - - - - , . , , . , , , , - - - - - - . -, - - - , - - - ----n - - - , . -. , .-,------ - . . . - - . -

~

1 radioactivity exceeds 3 x 10-4uci/cm3 at point of release, during the reporting period.

3 (3) Average concentration (#Ci/cm ) og release as diluted by sewage system flow of

- 2.7 x 108 cm3/ day.

Gaseous Weste (summarized on a monthly basis)

(1) Radioactivity discharged during the reporting period (in curies) for:

(a)- Gase's (b) Particulates, with half lives greater than eight days.

(2) The MPC used and the estimated activity (in I

curies) discharged during the reporting period, by nuclide, based on representative isotopic analysis.

Solid Waste (1) The total amount of solid waste packaged I (in cubic feet).

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

(3) The dates of shipment and disposition (if l

shipped off-site).

! 6.7 PLANT OPERATING RECORDS i

In addition to the requirements of applicable regulations and in no way substituting therefore, I

IV-52 l

I l

l :. . .

l l

l 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:

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 facili,ty:

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; and

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

h. Minutes of Safety Review Committee meetings.

i 6.8 APPROVAL OF EXPERIMENTS

1. All proposed experiments using the reactor shall be TV-53

evaluated by the experimenter and a staff member who i

has been approved by the Reactor Safety Subcommittee. The evaluation shall be reviewed by the Reactor Supervisor and the Radiation Safety l Officer to ensure compliance with the provisions of i the facility license, these Technical

]

Specifications, and 10 CFR 20. If the experiment i complies with the above provisions, it shall be submitted by the Recctor 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 changes in temperature, pressure, chemical composition, or radiolytic

. decomposition;

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

d. Any radfation 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. Irior to performing any new reactor experiment, an evaluation of the experiment shall be made by the IV-54

+. .

Reactor Safety Subcommittee. The subcommittee evaluation shall consider:

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 i 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.

A favorable subcommittee evaluation shall conclude that failure of the experiment 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 IV-55

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  ;

concentration of radioactive gases or aerosols may be released within the containment 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 130*F 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 perforced safely shall be a routine experiment and requires only the approval of the Reactor Supervisor and the Radiation Safety Officer 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

! IV-56

' s and may be approved for the subcommittee by agreement of the Reactor Supervisor and the Radiation Safety Officer.

I E

i.

4 IV-57