ML101970039

University of California, Irvine - Response to Request for Additional Information Dated May 25, 2010 Regarding License Renewal Request

ML101970039
Person / Time
Site:	University of California - Irvine
Issue date:	07/14/2010
From:	Geoffrey Miller University of California - Irvine
To:	Linh Tran Document Control Desk, Division of Policy and Rulemaking
References
TAC ME1579
Download: ML101970039 (47)

Text

UNIVERSITY OF CALIFORNIA, IRVINE BERKELEY -

DAVIS

• IRVINE

• LOS ANGELES

• RIVERSIDE

• SAN DIEGO ° SAN FRANCISCO SANTA BARBARA

• SANTA CRUZ George E. Miller IRVINE, CA 92697-2025 Senior Lecturer Emeritus (949) 824-6649 Department of Chemistry and FAX: (949) 824-6082 or (949) 824-8571 Director, Nuclear Reactor Facility Email: gemiller@uci.edu Faculty Advisor for Science Website: httv://chem.ps.uci.edu/-gemiller/

UCI Center for Educational Partnerships US Nuclear Regulatory Commission Attention: Document Control Desk Washington D.C. 20555-0001 FedEx to:

US Nuclear Regulatory Commission Document Control desk 11555 Rockville Pike Rockville, MD 20852 Attention: Ms Linh Tran, Project Manager Ref: Docket 50-326. Licence R-116 University of California, Irvine I am pleased to submit, in an enclosure, a response to the Request for Additional Information dated May 25th, 2010 (TAC No. ME1579) in regard to our license renewal request.

Please contact me if there are further questions in this regard.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on July 14 th 2010 Sincerely yours, George E. Miller Director Nuclear Reactor facility Senior Lecturer Emeritus, Department of Chemistry, University of California, Irvine Muy C

University of California, Irvine Reactor License R-116, Docket 50-326 Response to NRC Request for Additional Information (RAI) dated May 2 6 th 2010 (TAC NO. ME1579)

Please note that any references to Technical Specifications apply only to PROPOSED Technical Specifications (not yet implemented), a complete copy of which is attached to this response.

Please address and provide the requested information to the following:

1. NUREG-1 537, Part 1, Section 4.3, Reactor Tank or Pool states that the applicant should present all information about the pool necessary to ensure its integrity and should assess the possibility of uncontrolled leakage of contaminated primary coolant and should discuss preventive and protective features. Chapter 4, of your 1999 SAR does not provide this information. The following information is required to complete the review.

a. Please discuss the reactor pool water level monitoring system, alarm levels and required responses from operator and/or university personnel, if remote alarm signal is present. What is the minimum detectable amount of leakage?

b. Please discuss potential draining pathways of reactor pool water leakage, operator responses, and radioactivity monitoring. If water enters the uncontrolled environment, what is the radiological impact?

An additional section to the SAR 13.3.4 is attached and addresses these issues. In addition revised Technical Specifications are attached that include proposals regarding pool water monitoring systems and surveillance. (TS 3.3.1.; 3.3.4, 4.3)

2. NUREG-1537, Part 1, Section 10.1 Experimental Facilities and Utilization, states that the applicant should provide sufficient information to demonstrate that no proposed operations involving experimental irradiations or beam utilization will expose reactor operations personnel, experimenters, or the general public to unacceptable radiological consequences. Regulatory Guide 2.2, Development of Technical Specifications for Experiments in Research Reactors, is suggested to the applicant as guidance on technical specifications that may be useful. Regulatory Guide 2.2, Section C.1.c.(3) states that the "materials of construction and fabrication and assembly techniques should be so specified and used that assurance is provided that no stress failure can occur at stresses twice those anticipated in the manipulation and conduct of the experiment or twice those which could occur as a result of unintended but credible changes of, or within, the experiment."

Proposed Technical Specification (TS) 3.8.2(b), Limitations on Experiments, allows that explosive materials in quantities less than 25 mg may be irradiated in the reactor in a container "provided that the pressure produced in the experiment container upon accidental detonation of the explosive has been experimentally determined to be less than the design pressure of the container." The word "half' should be inserted after "less than" in proposed TS 3.8.2(b) or a discussion provide justifying why it is not needed.

TS. 3.8.2(b) has been revised as suggested

3. NUREG-1 537 states, a thermal hydraulic analysis should be performed for the reactor.

In your response dated January 27, 2010 to NRCs request for additional information (RAI) dated December 3, 2009, the analysis is provided through a reference for analyses made at two research reactors. However, the information is incomplete in that no information is provided on the similarity of the research reactors involved.

Please provide a comparison between the thermal hydraulic parameters (i.e., channel dimensions and geometry, linear power, etc.) and characteristics of the UCINRF core vs.

the referenced research reactors core so as to provide validity to using this information for the UCINRF.

A contract has been signed between DOER and General Atomic to provide this information by September 151h 2010.

4. NUREG-1537, Chapter 7 states a description should be provided of the reactor protection system. While this has been provided, there are inconsistencies in the discussion in the response to our RAI. Response to RAI 4.6 mentions "power level scrams", 1999 SAR Fig. 7-9 shows two level scrams and proposed TS 3.2.3, Table 2 shows one level scram. Please provide additional clarification.

Proposed TS 3.2.3 has been revised to remove this inconsistency.

5. NUREG 1537, Part 1, Section 4.5, Nuclear Design, states the applicant should discuss normal operating conditions, reactor core physics parameters and operating limits. The discussion should include a discussion of the complete, operable core; control rod worths; kinetic parameters; excess reactivities; shut down margins; and flux distribution for all planned configurations for the life of the core.

Section 4.5 of the 1999 SAR presents a representation of the flux distribution in the core.

However, based on 1999 SAR Figure 4-13, this flux plot in 1999 SAR Figure 4-16 appears to not be representative in that it shows no flux peaking in the center flux trap and does not portray the flux near the position of the adjustable transient rod (which is important to the Limiting Safety System Setting (LSSS)). Please provide appropriate flux distribution information including how the distribution will affect the peak to average power ratio.

See the response to question 3. This will be addressed by the contractor.

6. NUREG-1537, Part 1, Chapter 14, Technical Specifications, states the applicant needs to establish Technical Specifications that will provide reasonable assurance that the facility will function as analyzed in the SAR without endangering the environment or the health and safety of the public and the facility staff.

Proposed TS 3.2.3, Reactor Safety System. This Limiting Conditions for Operation (LCO) contains the fuel element temperature scram. The listed value of 500 degrees C

(°C) is the same as the LSSS when the instrumented fuel element is located in the Bring position. However, this setpoint is above the LSSS when the instrumented fuel element is located in the C-ring position. Propose appropriate changes in the TS wording for the instrumented fuel element in the C-ring to correct this inconsistency.

TS 3.2.3 specifically Table 2 has been revised to address this.

7. NUREG 1537, Part 1, Chapter 14, Technical Specifications, states the applicant needs to establish Technical Specifications that will provide reasonable assurance that the facility will function as analyzed in the SAR without endangering the environment or the health and safety of the public and the facility staff.

a) Proposed TS 1.11, Experiment. The definition does not include in the definition the distinction of "in or near the core" that is contained in ANSI/ANS-1 5.1.

Propose appropriate TS wording or justify why it is not needed.

TS 1.11 has been revised to meet ANSI/ANS 15.1 language.

b) Proposed TS 1.23, Reactor Secured. The definition uses the term "optimum conditions of configuration or reflection" rather that "optimum conditions of moderation or reflection" that is used in ANSI/ANS-15.1. Propose appropriate TS wording or justify why it is not needed.

TS 1.23 has been revised to meet ANSI/ANS 15.1 language.

c) Proposed TS 1.28, Shall. The definitions do not include the definitions for "should" and "may" that are included in ANSI/ANS-1 5.1. The term "should" was not found to be used in a manner that needs to be included in the TS. However, the term "may" is used in Section 5, Design Features. Propose appropriate TS wording or justify why it is not needed.

TS 1.28 has been revised to be TS 1.30 and to meet ANSL/ANS 15.1 language.

d) Proposed TS 3.1.4. Pulse Operation. Please provide a discussion regarding a reactivity insertion limitation to ensure a fuel temperature of less than 830oC during a pulse. Propose appropriate TS wording or justify why it is not needed.

We believe that the provisions of TS 3.1.4 already are adequate on the basis of the SAR section 13.where it is shown that a $3. 00 pulse from less than 1 kw steady state power level reaches a temperature of only350C and in the event that an interlock is violated, only 5700C. The first of these has been validated by previous actual experience at this facility. These are well below the possible 830TC limit mentioned.

e) Proposed TS 3.1.5 Fuel Burnup. This statement described why there is no limitation on fuel burnup and is not a TS. Propose appropriate TS wording or justify why a TS is not needed.

While there is no clear case for establishing a limit on burn up for 8.5% by weight, stainless steel clad, 20% enriched TRIGA fuel, this issue will be revisited by the contractor (see answer to question 3.).

f) Proposed TS 3.7.2 Effluents. There appear to be a typographical error in the units of the specification. It should be pcuries/ml rather than pc/ml. Propose an appropriate correction to the TS wording for this specification.

TS 3. 7.2 has been revised accordingly.

g) Proposed TS 4.1, Reactor Core Parameters. The requirement to inspect fuel elements is not labeled with a sub-header and should be labeled ("e."). Propose appropriate TS wording changes.

TS 4.1 has been revised h) Proposed TS 4.4 is not a TS. The wording should be changed to indicate that specification is blank or reserved.

TS 4.4 has been declared blank as not needed at this facility.

i) Propose appropriate TS wording for levels of pool radioactivity or justify why they are not needed. The proposed TS do not include LCO or surveillance requirements for levels of radioactivity in the reactor pool as recommended by ANSIIANS-15.1 TS 3.3.4 and TS 4.3. d have been added as suggested j) The requirements for the conditions when fuel elements are removed from service are present in the current UCINRF TS, but have not been carried over to the proposed TS. Reinsert these requirements or justify why they are not needed.

We believe that criteria for fuel element conditions that require they should be removed from service are adequately addressed in TS 3.1.6. A typographical error in this specification was corrected at this time.

In addition to these issues addressed in this RAI, we have used the opportunity to revise several sections of the proposed Technical Specifications following the suggestions of your staff and on our own discovery of inconsistencies in language and/or punctuation.

These were sufficiently numerous that no attempt was made to identify these changes, but rather to present you with a full new version, including an updated Table of Contents to reflect needed changes in pagination.

Revised SAR July 2010 Section 13.4 13.4 Loss Of Reactor Pool Water (Coolant) 13.4.1 Mechanisms and Rate of Loss of Coolant.

The cooling system and purification system pool outlets to their respective pumps are both situated at less than three (3) feet below the normal water level. Thus "pumping out" water from the pool in the event of a pipe breakage could only lower the water level by less than 3 feet. No significant hazard results from such lowering. All components of the water piping on both systems are above pool level, so no syphoning of water from the pool can occur.

Rapid and complete loss of the water in the reactor pool is considered to be highly improbable:

in fact it is likely that such an event could only arise from an earthquake shock so intense that the reactor hazard would only be one of many serious hazards to persons in Rowland Hall.

However, provision has been made for warning of the loss of pool water as an isolated event and calculations have been made of the fuel temperatures and radiation levels to be anticipated.

Potential slow leakage of water from the pool is dealt with in a following section.

A float switch alarm, operating on a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> basis is provided in the tank to alarm in the control room if the water should reach about 1 foot (30 cm) below the tank lip. If not quickly silenced, this also alarms at the UCI Police Dispatch Desk. As for any other "trouble" alarm report, the Police Department notifies reactor and/or EH&S staff form a call list immediately. Operating Procedures exist to delineate subsequent actions. A radiation area monitor station on the ceiling directly over the pool will also report an alarm should the water falls so far it could result in a potentially dangerous level.

The time it would take for the pool to drain may be conservatively estimated by considering a cataclysm that resulted in the complete opening to the soil of the bottom of the tank. Under these conditions the flow of water through the soil can be described by Darcy's law,' where i is the pressure gradient dx-=Ki dt in meters of water head per meter, and K is a constant depending on the nature of the soil, having the units m/year/mH20)/m. The time to drain the tank is then t = l/K, years/meter(H 20). The permeability K has a maximum value for pure sand of about 3 x 103.1 Thus to drain the tank to the bottom of the core (22 ft) or 6.7 meters the time is 0.0022 years or 19.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. This allows ample time for emergency procedures to be enacted.

Any effect on reactor fuel has been considered in section 13.2 above.

13.4.2 Radiation Levels After Loss of Pool Water Dose rates have been calculated for two locations: 1) at the top of the pool, 2) at the floor level of the classroom over the reactor. The reactor was assumed to have been operating for a long time at 250 kw prior to loss of water.

The dose rates calculated for instantaneous loss of water at shutdown are given in Table 13-2, and are clearly hazardous. However, as shown above, instantaneous loss of all the water above

the core is not conceivable and it is more realistic to estimate the dose rates with partial water loss. These values are given in Table 13 -3.

Thus at least 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> is provided from the alarm before the dose rate approaches the "permissible" 0.24 mr/hr level in the area above the reactor. Such time is ample to clear the area, investigate the damage and initiate remedial action. In addition a rough estimate of the radiation dose at the side of the pool from scattered radiation indicates that an individual who did not expose himself or herself to the direct radiation from the core could work there for at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, 1 day after the accident, without receiving a dose in excess of that permitted by 10 CFR 20.101 for a calendar quarter. This time would be sufficient to view the interior of the pool with a mirror and to continue emergency repairs.

TABLE 13-2 CALCULATED RADIATION DOSE RATES AFTER TOTAL LOSS OF REACTOR POOL WATER Time after shutdown Direct Radiation Levels, R/hr (1)

(2)

Top of Pool Floor of Classroom 0.1 hr 1.1 x 103 2.5 1.0 hr 5.9 x 102 1.4 10 hr 2.8 x 102 0.67 1 day 2.0x 102 0.48 1 week 9.1 x 101 0.22 1 month 16.5 0.04 TABLE 13-3 CALCULATED RADIATION DOSE RATES WITH LOSS OF POOL WATER AT 0.4 METERS PER HOUR Time after shutdown Water Depth over core Direct Radiation at (meters)

Floor of Classroom (mr/hr) 0.l hr.

6.1

< 1010 1.0 hr 5.6* alarm given

< 10-10 5.0 hr 4.1 1.3 x 10"0 10.0 hr 2.1 2.8 x 10-2 15.2 hr 0

ca 550

• Activated by float switch Method of Calculation The following assumptions were made:

1). The core had been operating at 250 kw for a very long time (1000 hr) prior to shutdown.

2). The reactor is shutdown as soon as the leak develops.

3). The core can be approximated to a point source with effective strength determined by the photons escaping from an isotropic volume-equivalent spherical source.

4). The photon energy is only from fission decay and can be averaged to 1.0 MeV.

5). Attenuation by core components other than fuel elements is negligible.

Assumptions 1) and 5) are conservative, and 4) is optimistic. The net effect is conservative.

Effective Source Strength

The fission product decay photon source strength is given by Perkins and King1 and the value of F(totaj) was assumed to be 1.0 Mev. The source strengths obtained are given in Table 13-4.

Each fuel element has a volume of 400 cm 3, thus a 67 element core has a volume of 2.68 x 104 cm 3 which is equivalent to a solid sphere of 18.4 cm radius. The fraction of photons which escape from the core is given2 as a function of gcrc, and is 0.197, where gc is the attenuation coefficient of the core material for 1 Mev photons ( = 0.207 / cm 3) and r. is the radius of the sphere: 18.4 cm. The effective source strength is the product of the source strength and the fraction escaped.

TABLE 13-4 SOURCE STRENGTH FOR SPHERICAL REACTOR CORE OPERATED FOR 1000 HOURS AT 250 KW Time after shutdown Source Strength Effective Source Strength (MeV/sec-250 kw)

(MeV/sec-250 kw) 0.1 hr 1.62x 1016 3.19x 1015 0.5 hr 1.13x 1016 2.22x 1015 1.0 hr 0.9x 1016 1.77x 1015 5.0 hr 5.5x 1015 1.08x 1015 10.0 hr 4.25x 1015 0.84x 1015 24.0 hr 3.Ox 1015 5.9x 1014 lweek 1.37x 1015 2.7x 1014 lmonth 2.5x 1014 4.9x 1014

' J. F. Perkins and R. W. King, Nuel. Sci. Engin., 3, 726 (1958).

2 K. K. Aglintsev, Applied Dosimetry, (Eng. Ed.), Ili, London (1965).

3 GA-5400, p 8-53

Dose Rates-Direct Radiation The dose rate is computed on the basis that the core is a point source, 21 feet below the top of the pool:

D Seff

-B;x K. 47R 2 where D

= the dose rate in rads/hr (= rem/hr for I Mev y)

Seff =

effective source strength (col. 3., Table 13-4)

R

= distance between core and location considered.

K

= conversion factor for Mev/cm2 to rad/hr (= 5.77 x 105 for 1 Mev photons)

B

= build-up factor for water 4, and concrete 5

p.

= linear attenuation coefficient, cm-1 (ref. 1) for 1 Mev photons 9concrete

= 0.149 cm gwater

= 0.067 cm x

= thickness of material, cm. (43 cm for concrete ceiling of reactor room).

Attenuation by air was neglected.

13.4.3 Pool Leakage Consideration needs to be given to the possibility that the reactor pool could suffer loss of integrity and leak at a slow, but possibly constant, rate. This has been observed at several facilities, however, the probability at this facility is considered low in light of the fact that UCINRF has a single below ground tank with an aluminum liner supported by reinforced concrete, as shown in SAR Section 4.1, Figs 4-2 and 4-3. The photograph (Fig 13.1) illustrates the pit with the tank in place before the concrete shell was poured and before backfill. The tank was hot wrapped with tar felts before the concrete surround was poured to obviate any possible corrosion. Backfill was then used surrounding the concrete shell and consolidated to specified standards.

As noted, the tank was installed in an excavated pit and back fill applied. The geology of the excavation is such that layers of shale are embedded in sandstone making a "basin" the sides of which are relatively impermeable. Pre-construction drillings showed complete absence of water to a depth of over 40 feet in any location beneath Rowland Hall. However, it was discovered following occupancy and completion of landscaping that water accumulates in this "basin" as a result of irrigation of landscaping adjacent to Rowland Hall, and the occasional rainfall. To monitor and reduce this accumulation wells have been drilled to a depth of about 35 feet adjacent to Rowland Hall on the north and west sides. Automatic sump pumps are installed in those wells.

However it was found that these do not fully tap into the "basin" area adjacent to the reactor tank, so a third well was drilled within the reactor room approximately 9 feet from the western edge of the aluminum liner, just off its center line. Occasional, manually operated pumping is done from this well to lower the water level which rises after significant rainfall. Samples are 4

H. 0. Wyckoff, "Radiation Attenuation Data", in Radiation Hygiene Handbook, H. Blatz, ed. McGraw Hill, New York, 1959. Chapter 8.

5 Reactor Handbook, Vol. III, Part B, "Shielding", (1962).

taken of the water being removed and are assayed by gamma-ray spectrometry. Significant levels of natural decay products are observed, indicating there is a natural soil content of uranium and thorium. It is a basis for the actions described below that any pool leakage would be into and contained within this "basin" and would not impact any further ground water or human activity/use areas.

There are no penetrations of this tank liner for beam tubes, water circulation, or other access to the core, which have been problematic at some facilities. No water carrying piping for cooling or water purification are below ground, so that any leakage from those would be readily detected and corrected by local action.

Fig 13.1 Reactor tank liner in place during construction (1968).

However provisions have been instituted that would help to establish any significant leakage and monitor any possible consequences in terms of radioactive contamination of the surrounding soils.

13.4.3.1 Pool Water Quantity Monitoring.

Provision is made for recording any make-up water quantity and/or water level changes within the reactor tank. No automatic make-up is provided, so procedures require recording

approximate make up quantity and new level. This is done on the daily start-up checklist assisted by a float level monitor installed in the pool with a read-out at the control console. This latter reports the level in approximately 12 mm increments. For this pool, 25 mm corresponds to 80 gallons of water. Records are examined annually so as to detect any deviation from normal evaporation rates. To enable reasonable comparisons, the pool water temperature is operationally required to be maintained between 18 °C and 24 0C. Average values for atmospheric humidity, which affects evaporation rate, are relatively low in this area, though excursions do occur during certain seasons. The water level is normally maintained between 18 and 30 cm below the upper tank lip. Any unusual drop rate in water level is procedurally required to be reported and understood. Any operation of the low water alarm which implies a drop below 30 cm, will be cause for an alarm and investigation.

13.4.3.2 Pool and Pit Water Quality Monitoring Radionuclides are created in the reactor pool as a result of irradiation of core components and of substances (salt) dissolved in the pool. Many of these, but not all, are removed by the water purification resins. Short-lived radionuclides will be apparent immediately following operation, but will decay before detection. Longer-lived radionuclides may accumulate in the pool. The most noticeable as removed on the ion exchange resin are 60Co and 54Mn. By providing annual comparison of pool water and 'sample' pit water for these nuclides, an additional; check on possible leakage is obtained. In addition measurement of tritium - expected to be below reasonable detection limits for this pool - is done because of public concerns that have arisen at other reactors.

Proposed Technical Specifications for the U. C. Irvine TRIGA Mark I Nuclear Reactor Revised July 14th 2010

TABLE OF CONTENTS DEFINITIONS 1

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limit - Fuel Element Temperature 5

2.2 Limiting Safety System Setting 5

LIMITING CONDITIONS FOR OPERATION 3.1 Reactor Core Parameters 6

3.2 Reactor Control and Safety Systems 8

3.3 Coolant Systems 11 3.4 Confinement 12 3.5 Ventilation Systems 13 3.6 Emergency Power 14 3.7 Radiation Monitoring Systems and Effluents 15 3.8 Limitations on Experiments 16 3.9 Facility-specific Requirements following extended Shutdown.

17 SURVEILLANCE REQUIREMENTS 4.0 General 18 4.1 Reactor Core Parameters 18 4.2 Reactor Control and Safety System 19 4.3 Reactor Pool Water 20 4.4 Containment or Confinement 20 4.5 Ventilation Systems 21 4.6 Emergency Power 21 4.7 Radiation Monitoring Equipment 21 DESIGN FEATURES 5.1 Site and Facility Description 22 5.2 Reactor Coolant System 22 5.3 Reactor Core and Fuel 23 5.4 Fuel Storage 24 ADMINISTRATIVE CONTROLS 6.1 Organization 25 6.2 Review and Audit 27 6.3 Radiation Safety 29 6.4 Operating Procedures 29 6.5 Experiment Review and Approval 30 6.6 Required Actions 30 6.7 Reports 31 6.8 Records 32

1. DEFINITIONS The following frequently used terms are defined to aid in the uniform interpretation of these specifications.

1.1 AUDIT An examination of records, logs, procedures, or other documents to ascertain that appropriate specifications and guidelines are being followed in practice. An audit report is written to detail findings and make recommendations.

1.2 CHANNEL A combination of sensor, lines, amplifier and output device which are connected for the purpose of measuring the value of a parameter.

1.3 CHANNEL CALIBRATION An adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the parameter that the channel measures. Calibration shall include equipment activation, alarm or trip, and shall be deemed to include a CHANNEL TEST.

1.4 CHANNEL CHECK A qualitative verification of acceptable performance by observation of channel behavior, or by comparison of the channel with other independent channels or systems measuring the same parameter.

1.5 CHANNEL TEST An introduction of a signal into the channel to verify that it is operable.

1.6 CONFINEMENT is the enclosure of the overall facility designed to limit release of effluents between the enclosure and the external environment through controlled or defined pathways.

1.7 CONTROL ROD is a device for adjustment of core reactivity through movement of neutron absorbing material or fuel, or both. A control rod may be coupled to its drive unit in a way that allows it to perform its safety function when the coupling is disengaged. Types of control rods include:

a. Regulating (REG): a rod having electromechanical drive and scram capabilities. Its position may be adjustable by manual or electronic control. It may have a fueled follower section.

b. Shim (SHIM): a rod similar to the REG rod but without the possibility of electronic adjustment of position.

c. Transient (ATR or FTR): a rod that can be moved up or down by pneumatic control. It has neutron absorbing material and may have a void follower. The length of movement of the ATR can be adjusted by an electromechanical drive system.

1.8 CORE CONFIGURATION describes a particular arrangement of fuel, control rods, graphite reflector elements, and experimental facilities inserted within the core grid plates.

1.9 CORE POSITION is defined by a series of holes in the top grid plate of the core, designed to hold a standard fuel element. It is specified by a letter, signifying the ring of holes in the grid plate, moving outwards from A through G, and a number indicating the position within the ring.

1.10 EXCESS REACTIVITY is that amount of reactivity that would exist if all reactivity control devices were moved to the maximum reactive condition from the point where the reactor is exactly critical (kff = 1) at reference core conditions or at a specified set of conditions.

UCI Technical Specifications 2010-02 page 1

1.11 EXPERIMENT is any operation that is designed to investigate non-routine reactor characteristics or that is intended for irradiation of items within the pool or in irradiation facilities. Hardware rigidly secured to the core or other reactor structure so as to be part of its design to carry out experiments is not normally considered an experiment. Specific experiments shall include:

a. SECURED EXPERIMENT is any apparatus, device or material held in a stationary position relative to the reactor core by mechanical means. The securing forces must be adequate to maintain a fixed position-in the event of foreseen external forces on the system, or applied as the result of credible malfunctions. Secured experiments are intended to be installed only when the reactor is not operating.

b. UNSECURED EXPERIMENT is any experiment that does not meet the definition of a secured experiment.

c. MOVEABLE EXPERIMENT is any experiment in which it is intended that movement of all or a part of the experiment may be moved in or near the core or into or out of the reactor while the reactor is operating.

1.12 FUEL ELEMENT is a standard TRIGA fuel rod with zirconium hydride/uranium fuel and stainless steel cladding. Maximum enrichment in 235U is 20%, and maximum uranium is less 9% by weight.

1.13 INSTRUMENTED FUEL ELEMENT is an element in which one or more thermocouples are embedded for the purpose of measuring fuel temperature during reactor operation.

1.14 IRRADIATION FACILITIES are pneumatic transfer systems, central tube, rotary specimen rack, and the in-core facilities (including single element positions, three-element positions, and the seven element position) and any other facilities in the tank designed to provide locations for neutron or gamma ray exposure of materials.

1.15 MEASURED VALUE is the value of a parameter as it appears on the output of a channel.

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

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

1.18 OPERATIONAL CORE means a CORE CONFIGURATION that meets all license requirements, including Technical Specifications.

1.19 PULSE MODE means any operation of the reactor with the mode switch in the PULSE position that satisfies all instrumentation and license requirements, including technical specifications, for pulse operation of the reactor.

1.20 REACTOR FACILITY is the physical area defined by rooms B64, B64A, B54, B54A, and B54B in the service level of Rowland Hall on the campus of the University of California Irvine.

1.21 REACTOR OPERATING means any time at which the reactor is not secured.

1.22 REACTOR SAFETY SYSTEMS are those systems, including their associated input channels, that are designed to initiate automatic reactor scram or to provide information for the manual initiation UCI Technical Specifications 2010-02 page 2

of a scram for the purpose of returning the reactor to a shutdown condition.

1.23 REACTOR SECURED. The reactor is secured when

a. Either there is insufficient moderator or fissile materials to attain criticality under optimum conditions of moderation or reflection; or

b. All of the following exist:

(i) all four CONTROL RODS are fully inserted.

(ii) the reactor is SHUTDOWN.

(iii) no experiments or irradiation facilities in the core are being moved or serviced that have, on movement or servicing, a reactivity worth of more than one dollar.

(iv) no work is in progress involving core fuel or core structure, installed control rods, or control rod drive mechanisms unless they are physically decoupled from the control rods.

1.24 REACTOR SHUTDOWN. The reactor is shutdown when:

a. sufficient CONTROL RODS are inserted so as to assure that it is subcritical by at least $1.00 of reactivity with the reactivity worth of all installed EXPERIMENTS and IRRADIATION FACILITIES are included; and the console key switch is in the "off' position and the key is removed.

1.25 REFERENCE CORE CONDITION is when the core is at ambient temperature (cold) and the reactivity worth of any xenon present is negligible (less than $0.20).

1.26 REVIEW means an examination of reports of AUDITS and other records with the purpose of establishing whether or not the facility staff is following appropriate license conditions, procedures and guidelines to operate the reactor in a safe and secure manner.

1.27 SAFETY CHANNEL means a measuring channel whose function is important to reactor and or experiment safety.

1.28 SCRAM TIME is the elapsed time between the initiation of a scram signal from a CHANNEL and a specific movement of a CONTROL ROD to its fully inserted position.

1.29 SEVEN ELEMENT POSITION is a hexagonal section which can be removed from the upper grid plate for insertion of specimens up to 4.4 in. in diameter after relocation of all six B-ring elements and removal of the central tube irradiation facility.

1.30 SHALL, SHOULD and MAY. The word SHALL is used to denote a requirement; the word SHOULD is used to denote a recommendation; and the word MAY is used to denote permission -

neither a requirement nor a recommendation.

1.31 SHUTDOWN MARGIN refers to the minimum shutdown reactivity necessary to provide confidence that the reactor can be made sub-critical by means of the control and safety systems starting from any permissible operating condition and with the most reactive rod in its most reactive position, and will remain subcritical without further operator action.

1.32 STEADY-STATE MODE is whenever the reactor is OPERATING with the mode selector switch in either of the STEADY-STATE positions.

UCI Technical Specifications 2010-02 page 3

1.33 SUBSTANTIVE CHANGES are changes in the original intent or safety significance of an action or event.

1.34 SURVEILLANCE INTERVALS that are permitted are established as follows:

a. quinquennial - every five years, not to exceed 66 months

b. biennial - every two years, not to exceed 30 months

c. annual - every year, not to exceed 18 months

d. semi-annual - not to exceed 9 months e.quarterly - not to exceed 5 months

f. monthly, not to exceed 6 weeks g.daily - refers to each day when the reactor is to be operated or before any operation extending more then one day During prolonged periods when the reactor remains shutdown, certain surveillance that needs reactor operation to be accomplished may be deferred. All deferred surveillance shall be completed when the reactor is re-started, before routine operations or experiments are conducted.

1.34 THREE ELEMENT POSITION is one of two generally triangular-shaped sections cut out of the upper grid plate, one encompassing ring holes D5, E6 and E7 and the other D14, El18 and E19. When fuel elements are placed in these locations, a special fixture provides lateral support. With the fixture and fuel removed, an experiment up to 2.4 in. in diameter may be inserted.

UCI Technical Specifications 2010-02 page 4

2. SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limit - Fuel Element Temperature Applicability. This specification applies to the fuel element temperature.

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

Specification. The temperature in a stainless steel clad, high hydride fuel element shall not exceed 750'C under any conditions of operation.

Bases. The important parameter for a TRIGA reactor is the fuel element temperature. This parameter is well suited as a single specification since it can be measured. A loss in the integrity of the fuel element cladding could arise from an excessive build-up of pressure between the fuel moderator and the cladding. The pressure is caused by the presence of fission product gases and hydrogen gas from the dissociation of the zirconium hydride in the fuel moderator. The magnitude of this pressure is determined by the fuel moderator temperature and the ratio of hydrogen atoms to zirconium atoms in the zirconium hydride moderator. (NUREG 1282)

The safety limit for the stainless steel clad, high hydride (Zr/Hi. 7) fuel element is based on analysis (McClellan Nuclear Research Center reactor SAR 4.5.4.1.3 and Oregon State University SAR 4.5.3.1) which indicates that the stress in the cladding due to the hydrogen pressure from the dissociation of the zirconium hydride will remain below the yield stress provided the temperature of the fuel does not exceed 11 50'C and the fuel cladding is water cooled so it remains well below this limit. An incident at Texas A&M under special circumstances have suggested that some effects of temperature may be observable at the lower limit of 830'C. A conservative value of 700 'C is chosen since this facility does not need to approach this limit.

2.2 Limiting Safety System Settings Applicability. This specification applies to the scram setting for the fuel element temperature channel.

Objective. The objective is to prevent the safety limit from being reached.

Specifications. For a core composed entirely of stainless steel clad, high hydride fuel elements, limiting safety system settings apply according to the location of the standard instrumented fuel element (IFE) which shall be located in the B-or C-ring as indicated in the following table:

Location Limiting Safet System Setting B-ring positions B3, B4, B5, or B6 4750 C C-ring positions Cx,Cy, Cz 4250 C UCI Technical Specifications 2010-02 page 5

Basis. For stainless steel clad, high hydride fuel elements, fuel temperature can be measured in a system designed to initiate a reactor scram if a conservative limit is exceeded. Since the fuel element temperature is measured by a fuel element designed for this purpose (IFE), the limiting settings are given for different locations in the fuel array. With the core configuration grid used, the core is always loaded so that the maximum fuel temperature is produced in the B-ring. If the fuel element temperature is measured in the C-ring, the respective temperature is reduced appropriately. Limits are placed on IFE locations in order to assure the measured temperatures are not unduly influenced by control rod positions. Maximum recorded temperatures for the UCI reactor B-ring IFE for the period since 1969 are 250'C at steady state, and 350'C for pulse operation. The limiting safety system settings are set to provide a margin between operational scram temperatures and the safety limit be set based on experience at this and other TRIGA facilities as to maximum operational temperatures reached which are considerably below safety limits. This also allows for any reasonable uncertainties in temperature measurement. There is no evidence to support a need for reduction in fuel temperature limits as fuel ages.

3. LIMITING CONDITIONS FOR OPERATION 3.1 Reactor Core Parameters 3.1.1 Steady-state Operation Applicability. This specification applies to the energy generated in the reactor during steady-state operation Objective. The objective is to assure that the fuel temperature safety limit is not exceeded.

Specification. The reactor power level in steady-state operation shall not exceed 250 kilowatts.

Basis. Experience at other TRIGA reactors and thermal and hydraulic calculations (OSU SAR 4.5.3) indicates that power levels up to 1.9 Mw can be safely used with natural convection cooling of the fuel elements, in a similar circular grid plate core configuration. Operation at less than 20% of that value will be safe.

3.1.2 Shutdown Margin Applicability. These specifications apply to reactivity margins in shut down condition.

Objective. The objective is to assure that the reactor can be shut down at all times and to assure that the fuel temperature safety limit shall not be exceeded.

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

a. The shutdown margin provided by the control rods referred to the reference core condition shall be greater than $0.50 with irradiation facilities and experiments in place and the total worth of all unsecured experiments in'their most reactive state; and the most reactive control rod fully withdrawn;

b. No experiment with a reactivity worth greater than $1.00 is unsecured UCI Technical Specifications 2010-02 page 6

Basis. The value of the shutdown margin and limits on experiments assure that the reactor can be shut down and remain so even if the most reactive control rod should remain fully withdrawn and unsecured experiments are moved.

3.1.3 Core Excess Reactivity Applicability. This specification applies to the reactivity condition of the reactor Objective. The objective is to assure that the reactor fuel safety limit is not exceeded in any mode of operation.

Specification. The reactor shall not be operated unless the following condition exists:

The maximum available core excess reactivity based on the reference core condition shall not exceed $3.00.

Basis. Computations presented in the SAR (Chapter 13.3) establish that a sudden insertion of

$3.00 results in a fuel temperature of approximately 350'C, well below the established safety limit for this fuel (TS 2.1). Such calculations are conservative, being based on a purely adiabatic model. The specifications assure that no insertion of reactivity above this value shall be possible, even under non-normal operating conditions.

3.1.4 Pulse Operation Applicability. These specifications apply to energy generated in the reactor as a result of pulse insertion of reactivity Objective. The objective is to assure that the fuel temperature safety limit shall not be exceeded.

Specifications. The reactor shall not be operated in the pulse mode unless, in addition to the other requirements of Section 3.1,

a. the steady-state power level of the reactor is less than 1 kilowatt; and

b. The total reactivity worth of the two transient control rods (ATR + FTR) shall not exceed

$3.00.

Basis. Inadvertent pulsing from a high steady-state power level or with a large reactivity change could result in a higher final peak temperature approaching the safety limit. This specification establishes a limit on pulsing from an initial high temperature or with excessive reactivity to limit temperature rise to anticipated values. TS 3.1.3 establishes the basis for the

$3.00 limit.

3.1.5 This section intentionally left blank.

UCI Technical Specifications 2010-02 page 7

3.1.6 Fuel Element Inspection Parameters Applicability. The specifications apply to all fuel elements, including fuel follower control rods.

Objective. The objective is to maintain integrity of fuel element cladding.

Specifications. The reactor shall not be operated with any fuel elements that show damage. A fuel element shall be identified as showing damage and be removed from core if:

a. the transverse bend exceeds 1 1 6th inches (0.0625 in) over the length of the element; or

b. the growth in length over original measurements exceeds 1/8h inch (0.125 in); or

c. a cladding defect is suspected by a finding of release of any fission products; or

d. visual inspection identifies unusual pitting, bulging, or corrosion.

Basis. These criteria have been successfully used for hundreds of fuel inspections over many years in similar reactors to successfully identify elements that have cladding issues prior to serious failure.

3.2 Reactor Control and Safety Systems 3.2.1 Control Rods Applicability. This specification applies to the function of all control rods.

Objective. To assure control rods are operable and that prompt reactor shut down following a scram is accomplished.

Specifications. The reactor shall not be operated unless the control rods are operable. Control rods shall not be considered operable if:

a. damage is apparent to rods or drive assemblies that could affect operation; or

b. the scram time exceeds 2 seconds.

Basis. Experience has shown that rod movement is assured in the absence of damage and that scram times of less than 2 seconds are more than adequate to reduce reactivity to assure safety in view of known transient behavior of TRIGA reactors.

UCI Technical Specifications 2010-02 page 8

3.2.2 Reactor Measuring Channels Applicability. This specification applies to the information which shall be available to the reactor operator during reactor operation.

Objective. To specify that minimum number of measuring channels that shall be available to the operator to assure safe operation of the reactor.

Specifications. The reactor shall not be operated in the specified mode unless the measuring channels described in Table 1 are operable.

Table 1. Minimum Measuring Channels Measuring Channel Operating Mode Steady-state Pulse Fuel Element Temperature 1

1 Linear Power Level 1

Log Power Level 1

Power Level (%)

1 1 (peak power)

Nvt circuit 1

Note 1. Any single power level channel may be inoperable while the reactor is operating solely for the purpose of diagnosis and/or channel tests or checks on that channel.

Note 2. Any single power level channel that is not required for safety scram purpose by TS 3.2.3 and ceases to be operable during reactor operation shall be returned to operating condition within 5 minutes or the reactor shall be shut down. For channels required by TS 3.2.3 the reactor shall be shut down immediately if the channel becomes inoperable.

Basis.

The fuel temperature displayed at the control console gives continuous information on the parameter which has a specified safety limit. The power level monitors assure that measurements of the reactor power level are adequately covered at both low and high power ranges in appropriate modes. Notes 1 and 2 allow for necessary tests for brief resolving of problems or recalibration while maintaining sufficient information for safe operation.

UCI Technical Specifications 2010-02 page 9

3.2.3 Reactor Safety System Applicability This specification applies to the reactor safety system channels.

Objective To specify the minimum number of reactor safety system channels that shall be operable in order to assure that the fuel temperature safety limit is not exceeded.

Specification. The reactor shall not be operated unless the safety system channels described in Table 2 and the interlocks described in Table 3 are operable in the appropriate operating modes.

Table 2. Minimum Reactor Safety Channels Safety Channel Function and trip level Operating Mode maximum setting Steady-state Pulse Fuel Element Scram - 400'C (B-ring IFE) 1 1

Temperature Scram - 3500C (C-ring IFE)

Reactor Power level Scram - 110% of 250 kw 2

Non-Operate Scram - loss of HV and/or 1

1 signal on any required channel Manual Bar Scram 1

1 Preset Timer Scram pulse rods < 15 seconds 1

after pulse Seismic Switch Scram - Modified Mercalli VI 1

1 Table 3. Minimum Interlocks Operating Mode Interlock Function Steady-Pulse state Wide Range Power Prevent control rod withdrawal when power level is < 1 x 1

Level Channel (Log) 10-7 % of full power REG, SHIM, ATR Prevent application of air to fast transient rod when all 1

Control Rod Drives other rods are not fully inserted REG, SHIM, ATR Prevent simultaneous withdrawal of more than one rod 1

Control Rod Drives REG, SHIM, ATR Prevent movement of rod drives other than by air Control Rod Drives application in pulse mode ATR Cylinder Drive Prevent application of air to adjustable transient rod unless 1

cylinder is fully down Wide Range Linear Prevent ATR or FTR insertion unless power level < 1 I

Power Channel kilowatt UCI Technical Specifications 2010-02 page 10

Bases Scrams. The fuel temperature scram provides the protection to assure that if a condition results in which the LSSS is exceeded, an immediate shutdown will occur to keep the fuel temperature well below the safety limit. The power level scrams are provided as added protection against abnormally high fuel temperature and to assure that reactor operation stays within the licensed limits. The manual scram allows the operator to shut down the system if an unsafe or abnormal condition occurs. A high voltage scram on each channel assures that detector response is operating at all times. The seismic switch will scram the reactor if major earth movement (M.M.

VI or above (equal to 0.06 - 0.07 g) occurs in case the operator is prevented from operating the manual scram at the time. The preset timer scram provides pulse "clipping" to reduce energy production at the tail of a pulse.

Interlocks. The interlock to prevent startup of the reactor with less than 10-7 % power indication assures that indication of neutron multiplication is present as reactivity is inserted. Other interlocks on rod drives are provided to prevent inappropriate multiple simultaneous reactivity insertions by operators. The interlocks which prevent the firing of the transient rods in the steady-state mode or if the power level is greater than 1 kilowatt prevent inadvertent pulses or pulsing when fuel temperature is too high.

3.3 Coolant Systems 3.3.1 Pool Water Level Applicability. These specifications apply to the water level in the reactor pool.

Objective. To assure there is sufficient water in the reactor pool to provide cooling and shielding for radiation from the core, and to check for potential pool leakage.

Specifications.

a. The reactor shall not be operated unless the pool water level is at least 15 feet above the upper grid plate of the core (at least 24 feet above the tank floor, or no more than 1 foot below the tank edge).

b. An audible alarm, with reporting to the UCIPD dispatch desk if not locally silenced by an operator, shall operate 24/7 to alert personnel if the water level in the reactor pool falls below the above limit. Visual checking of water level may be substituted during periods when the alarm is found to be inoperable.

Basis. Facility design calculations and subsequent measurements show that these water levels are sufficient to reduce full power operational radiation levels to acceptable levels within the facility and in any occupied areas above or surrounding the reactor. This is also true for shut down levels. The alarm will notify appropriate responders before any increase in radiation levels to the surroundings occurs. The alarm and the daily operational check will alert operators to the possibility that pool ;leakage might be occurring and resulting in pool level drop. No automatic water make-up is provided so that operators must add water manually to assure meeting this requirement.

UCI Technical Specifications 2010-02 page I11

3.3.2 Pool Water Temperature Applicability. This specification applies to the water temperature in the reactor pool.

Objective. To assure the water in the reactor pool stays within limits that provide sufficient cooling of the fuel and that minimizes stresses to the tank and reactor components.

Specification. The pool water temperature shall be maintained between 17'C and 25TC Basis. These temperature limits are easily maintained using the available cooling system and guard against temperatures that might produce undue stresses on tank components.

3.3.3 Pool Water Conductivity Applicability. This specification applies to the conductivity of water in the reactor pool.

Objective. To assure the water in the reactor pool is maintained at high purity to minimize potential corrosion of reactor components. This also assesses possible leakage from the highly "doped" secondary cooling water system Specification. The pool water conductivity level shall be maintained less than 3 micromhos/cm.

Basis. Experience at other reactor facilities indicates that maintaining the conductivity within 5 micromhos/cm is adequate to provide acceptable control of corrosion (NUREG 1537). An additional margin of assurance is provided by this lower specification.

3.3.4 Pool Water Radioactivity Applicability. This specification applies to the radioactivity of water in the reactor pool.

Obiective. To assure the water in the reactor pool is maintained at high purity to minimize effects of any pool leakage into the surrounding ground. This also can provide indication of leakage.

Specification. The average pool water radioactivity level shall be maintained within limits for sewer disposal as established by 1 OCFRPart 20, Appendix B, Table 3 for radionuclides with half-lives longer than 14 days.

Basis. Maintenance at this level will assure that any disposal of pool water, either planned or inadvertent, will be within appropriate and significant radioactivity limits.

3.4 Confinement 3.4.1 Operations Requiring Confinement Applicability. This specification applies to operations where there is a need to safeguard against release of radioactive materials beyond the facility.

Obiective. To assure there is confinement during certain higher risk operations that might release radioactive materials, especially gases or aerosols.

UCI Technical Specifications 2010-02 page 12

Specification. The following operations shall not be conducted unless confinement is assured by closing doors into the facility and verifying that the facility's ventilation systems and radiation monitors for routine and emergency modes are operable:

a. reactor operation at above 1 kilowatt for more than 10 minutes; or a single pulse; or

b. movement of irradiated fuel or fueled experiments; or

c. any work on core structure, experimental facilities, control rods or control rod drives that could result in a core reactivity change greater than $0.50.

Basis. Release of gaseous or aerosol material is most likely under significant movement operations involving core components, or reactor operations at high fuel energy release. The allowance for a short period of time permits personnel access or withdrawal and movement of un-involved items, which is unlikely to coincide with accidental release.

3.4.2 Equipment to Achieve Confinement This specification is addressed in section 3.5 below.

3.5 Ventilation Systems 3.5.1 Ventilation During Normal Operation.

Applicability. This specification applies to the facility ventilation system.

Objective. To assure there is adequate ventilation and flow control to assure confinement of any released gaseous or aerosol radioactivity.

Specification. The reactor shall not be operated unless the ventilation system is operating and maintaining a minimum of 0.10 inches of water negative pressure within the reactor room (B54) and the control room and between the reactor room and the air outside the building, except for periods of time not to exceed two hours to allow surveillance, maintenance and testing of the ventilation system. During such exception, no reactor pulses shall be fired.

Basis. Through a combination of inflow dampers and outflow exhaust, facility design establishes and exceeds these pressure differentials, which were selected on the recommendation of the reactor installer. Any negative inflow will assist in confinement of released materials. The SAR establishes that normal operation effectively dilutes 41Ar levels well below 10 CFR20 limits and as detailed in facility annual reports.

3.5.2 Ventilation During Emergency Situations Applicability. This specification applies to the ventilation system provided for emergency situations.

Objective. To assure there is confinement of radioactive releases by closing of normal ventilation and establishing a small purge flow to reduce possible exposure to personnel during the emergency.

UCI Technical Specifications 2010-02 page 13

Specification. A signal of high radiation activity alarm from a continuous particulate air monitor (CAM) measuring air from above the pool shall carry out the following functions:

a. close off inflow air by closing dampers; and

b. close off outflow air by closing dampers in exhaust ducts and removing power from relevant exhaust fans and fume hood; and

c. remove power from pneumatic transfer system so it can no longer operate to transfer air through any core region; and

d. apply power to a small exhaust "purge" fan and duct system equipped with a HEPA filter.

Basis. These actions will result in isolation of the main reactor rooms to aid in confinement, while beginning to purge contaminated air through a high grade filter. Experience at other facilities has shown that fission product release from fuel elements is most rapidly detected by a CAM operating in this manner.

The SAR establishes that the emergency purge system will, in the event of a radioactive gas release, be effective in providing personnel with sufficient time to evacuate before experiencing serious exposure. It is shown in Chapter 13 of the SAR that operation of the emergency exhaust system reduces off-site doses to below 10 CFR Part 20 limits in the event of a TRIGA fuel element failure, and that operation of the normal system adequately dilutes the argon 41 released even under unusual experimental operations. The specifications governing operation of the reactor while the ventilation system is undergoing repair preclude the likelihood of fuel element failure during such times. It is shown also that, if the reactor were to be operating at full steady-state power, fuel element failure will not occur even if all the reactor tank water were to be lost immediately.

3.6 Emergency Power Applicability. This specification applies to the use of emergency power systems.

Objective. To assure certain information related to personnel safety is available in the event of main electrical power failures.

Specification. Emergency electrical power, activated rapidly upon main electrical power failure, shall be provided to facility lighting, radiation monitoring and security monitoring systems.

Basis. Provision of power to these systems will assure that personnel present at the time, or responding to an event will have information to assist in monitoring their safety and the safety and security of the facility.

UCI Technical Specifications 2010-02 page 14

3.7 Radiation Monitoring Systems and Effluents 3.7.1 Radiation Monitoring Systems Applicability. This specification applies to monitoring of radiation levels.

Objective. To assure information is available to provide assurance of radiological safety of personnel at the facility, and of the absence of excessive releases beyond the facility.

Specifications.

a. The reactor shall not be operated unless the following minimum radiation monitoring instruments are operating:

Radiation Area Monitors (RAM):

2 Continuous Particulate Radiation Monitor (CAM): 1

b. An environmental monitoring dosimeter pack, exchanged at least quarterly, shall be in place in the primary exhaust duct of the facility at all times, except when undergoing exchange.

Additional packs shall be located in adjacent buildings, and in a more remote control for comparison.

Basis. These instruments and dosimeters will provide adequate notification of abnormal levels that could result in exposures or uncontrolled releases. The environmental dosimeters provide information that can be used to track long term trends that might need attention.

3.7.2 Effluents Applicability. This specification applies to the release rate of 41 Ar.

Objective. To assure that concentration of 41Ar in unrestricted areas shall be below the applicable limits of 10 CFR Part 20.

Specification. The annual average concentration of 41Ar discharged into an unrestricted area shall not exceed 4 x 10-8 ptCi/ml at the point of discharge.

Basis. This is 4 times the limit presented in 10 CFR 20 Appendix B, Table 2 for release to unrestricted areas. The height of the roof location of ventilation release and the restricted access to the roof area provides assurance that no individual will be exposed to such a concentration, and that dilution will always be by greater than a factor of 4.

UCI Technical Specifications 2010-02 page 15

3.8. Limitations on Experiments 3.8.1 Reactivity Limits Applicability. This specification applies to experiments placed in the reactor and its experimental facilities.

Objective. The objective is to prevent damage to the reactor or excessive release of radioactive materials in the event of an experiment failure.

Specifications.

a. The absolute value of any unsecured experiment shall not exceed $1.00, and

b. The reactivity worth of an individual experiment shall not exceed $3.00, and

c. The sum of absolute values of all experiments shall not exceed $3.00.

Basis. The limit on an unsecured experiment is to prevent an inadvertent pulse, and to maintain shutdown margin limitations. The insertion of $3.00 pulses has been analyzed as a safe operating condition for this reactor (SAR Chapter 13) and has been exceeded safely at other reactors with similar fuel core design, so limitation of experiments such that a pulse larger than this value could not occur is prudent and well within safe limits.

3.8.2 Materials Applicability. This specification applies to experiments placed in the reactor and its experimental facilities.

Objective. To assure minimal damage to experimental facilities or core structures as well as to minimize excessive release of radioactive materials in the event of experiment failure.

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

a. Fueled experiments are limited such that the total inventory of iodine isotopes 131 through 135 in the experiment is not greater than 0.3 curies and the Strontium 90 inventory is not greater than 1 microcurie; and

b. The quantity of known explosive materials to be irradiated is less than 25 milligrams and the pressure produced in the experiment container upon accidental detonation of the explosive has been experimentally determined to be less than half the design pressure of the container; and

c. Experiments containing materials corrosive to reactor components, compounds highly reactive with water, potentially explosive materials or liquid fissionable materials are doubly encapsulated.

d. Materials, design, and construction of experiments shall be such that assurance is given that no stress failure can occur at stresses twice those anticipated in the manipulation and conduct of the experiment or twice those that could occur as a result of unintended but credible changes of, or within, the experiment.

UCI Technical Specifications 2010-02 page 16

Basis It is shown in the SAR, Chapter 13, that a release of 0.024 curies of iodine activity will result in a maximum dose to the thyroid of a person in an unrestricted area of less than 1/20 of the permissible dose. The limit on iodine inventory is set at 10 times this value. The limit for Strontium 90 is that which corresponds to the iodine yield of 0.3 curies for a given number of fission events and would be no hazard. Specifications b., c., and d. reduce the likelihood of damage to reactor components resulting from experiment failure.

3.8.3 Failures or Malfunctions Applicability. This specification applies to experiments placed in the reactor and its experimental facilities.

Objective. To assure minimal damage to experimental facilities or core structures as well as to minimize excessive release of radioactive materials in the event of experiment failure.

Specifications. Where the possibility exists that the failure of an experiment under normal operating conditions of the experiment or reactor, credible accident conditions in the reactor, or possible accident conditions in the experiment could release radioactive gases or aerosols to the reactor room or any unrestricted area, the quantity and type of material in the experiment shall be limited such that the airborne radioactivity in the reactor facility or the unrestricted area will not exceed the applicable dose limits in 1 OCFR 20, assuming that:

a. 100% of the gases or aerosols escape from the experiment.

b. If the effluent from an experiment exhausts through a filter system designed for greater than 99% efficiency for 0.3 micron particles, at least 10% of the gases or aerosols will escape.

c. For a material whose boiling point is above 55°C and where vapors formed by boiling this material could escape only through a column of water above the core, at least 10% of those vapors will escape.

Basis. This specification is intended to assist experiment review and design in meeting the goals of 1 OCFR20 by reducing the likelihood of excessive facility personnel or public exposure by gases or aerosols as a result of experiment failure.

3.9. This section deliberately left blank.

UCI Technical Specifications 2010-02 page 17

4. SURVEILLANCE REQUIREMENTS 4.0 General Applicability. This specification applies to surveillance requirements of any system related to reactor safety.

Obiective. To assure the proper operation of any system related to reactor safety.

Specification.

a. Surveillance requirements may be deferred during prolonged reactor shutdown (except for TS 4.3.a, and 4.3.c.), however they shall be completed prior to reactor start-up unless reactor operation is required for performance of the surveillance. This latter surveillance shall be performed as soon as possible after reactor start-up. Scheduled surveillance which cannot be performed while the reactor is operating, may be deferred until the reactor next shuts down.

b. All replacements, modifications, and changes to systems having a safety related function shall be subjected to a quality assurance review. Insofar as possible, the replacement, modification, or change shall be documented as meeting the requirements of the original system or component and have equal or better performance or reliability. The required quality assurance review shall be approved by the ROC as specified in Section 6.2. A system shall not be considered operable until it has been properly tested to meet expected specifications.

Basis. Changes or maintenance can affect reactor operation parameters. This specification will assure that systems function according to established criteria before being utilized during reactor operation.

4.1 Reactor Core Parameters Applicability This specification applies to the surveillance requirements for reactor core parameters Objective. To verify that the reactor does not exceed authorized limits for power, shutdown margin, core excess reactivity, specifications for fuel element condition, and verification of total reactivity worth of each control rod.

Specifications

a. A channel calibration shall be made of the power level monitoring channels by the calorimetric method annually.

b. The total reactivity worth of each control rod shall be measured annually or following any significant change (>$0.25) in core configuration.

c. The core excess reactivity shall be measured using control rod position data prior to each day's operation, or prior to each operation extending more than one day, or following any significant change (>$0.25) in core configuration.

d. The shutdown margin shall be calculated at each day's shutdown, or at the end of any UCI Technical Specifications 2010-02 page 18

operation exceeding one day, or following any significant change (>$0.25) in core configuration.

e. All core fuel elements shall be visually inspected (under water) and measured for length and bend quinquennially, but at intervals separated by not more than 500 pulses of magnitude greater than $1.00 of reactivity. Fuel follower control rods shall be measured for bend at the same time interval. Such surveillance shall also be performed for elements in the B and C rings in the event that there is indication that fuel temperatures greater than the limiting safety system setting on temperature may have been exceeded.

Basis. Experience has shown that the identified frequencies are more than adequate to ensure performance and operability for this reactor. The value of significant change is measureable and will assure sufficient shutdown margin even taking into account decay of poison.

For fuel elements, the most severe stresses induced in the fuel elements result from pulse operation of the reactor, during which differential expansion between the fuel and the cladding occurs and the pressure of the gases within the elements increases sharply. The surveillance interval is selected based on the past history of more frequent, uneventful, inspections for over 40 years at this facility and experience at other TRIGA facilities with similar power levels, fuel type, and operational modes. It is also designed to reduce the possibilities of mechanical failures as a result of handling elements, and to minimize potential radiation exposures to personnel.

4.2 Reactor Control and Safety Systems Applicability. This specification applies to the surveillance requirements for the reactor control and safety systems.

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

Specifications.

a. Control rod drop (scram) times for all four control rods shall be determined annually.

b. Transient (non fuel-follower) control rods and fuel follower control rods shall be visually inspected for deterioration quinquennially. Fuel follower control rods shall be inspected under sufficient water to protect personnel from significant radiation exposure. Transient control rods may be removed from the pool for inspection.

c. The transient (pulse) rod drive cylinders and the associated air supply systems shall be inspected, cleaned, and lubricated if necessary, annually.

e. On each day that pulse mode operation of the reactor is planned, a functional performance check of the transient (pulse) rod system shall be performed.

f. A channel check of each of the reactor safety system channels shall be performed prior to each day's operation or prior to each operation extending more than one day.

UCI Technical Specifications 2010-02 page 19

g. A channel test of each item in Tables 2 and 3 in section 3.2.3, shall be performed semi-annually.

h. A calibration of the temperature measuring channel shall be performed annually. This calibration shall consist of introducing electric potentials in place of the thermocouple input to the channel. It shall also include a comparison to earlier measurements made on a reference core.

Basis. The control rods are inspected and drop times checked to assure safe operations. The surveillance intervals for those and the channel surveillances are selected based on the past history for over 40 years at this facility and are adequate to correct for long term drifts and other instrument problems.

4.3 Reactor Pool Water Applicability. This specification applies to the surveillance requirements for the reactor pool water.

Objective. The objective is to assure that the reactor pool water level channel is operable, that alarm settings are verified and alarm reporting is functional. In addition, that the water purity is being maintained within acceptable limits.

Specifications.

a. A channel check of the pool water level measuring channel shall be performed monthly to include verification of the alarm reporting system.

b. A channel calibration of the pool water level measuring channel shall be performed annually to include verification of alarm set point, and comparison to tabulations of make-up water required in comparison to expected average pool water evaporation rates.

c. The pool water conductivity shall be measured at the end of each operating day, or at shutdown for a period of operation extending more than one day. When the reactor has not been operated for prolonged periods, the conductivity shall be measured at least quarterly.

d.

The pool water radioactivity shall be measured annually in comparison to former data, a tap water sample, and a sample from a pit representing water accumulated in the ground surrounding the reactor tank. As a minimum levels of tritium, cobalt-60, and manganese -54 Basis. These verifications will assure that a continued warning system for a rapid loss of pool water incident is maintained, and any sudden perturbation of pool water quality is noted quickly to allow for corrective action to minimize corrosion, or build-up of radioactivity in the water.

The check on conductivity monitors possible leakage into the pool from the secondary water system. Radioactivity measurements will enable assessments of impacts of pool leaks. The long term comparisons enable information on possible slow leaks to be acquired.

4.4 This section deliberately left blank.

UCI Technical Specifications 2010-02 page 20

4.5 Ventilation Systems Applicability. This specification applies to the surveillance requirements for the reactor room ventilation system.

Objective. To verify performance is adequate to provide for normal and emergency mode ventilation for the facility to control and confine releases of airborne radioactive materials.

Specification.

a. A channel check of the existence of negative air pressure between the reactor room and the control room, and the reactor room and the outside air shall be performed daily.

b. A channel test of the function of the particulate high radiation (CAM) alarm to properly set the ventilation system into "emergency" mode shall be performed daily.

Basis. These checks will assure that any reactor operation that results in release of airborne radioactivity will result in appropriate confinement of that activity to the reactor room.

4.6 Emergency Power Applicability. This specification applies to the provision of emergency electrical power to room lighting, radiological safety, and security instrumentation.

Objective. To assure proper connection and function of the emergency electrical power so that personnel are provided lighting and information relating to radiological safety in the event of main electrical power failure.

Specification. A verification check that the instruments relating to radiological safety are attached to the correct circuit for emergency electric power provision shall be performed annually. Verification shall also be sought from the campus Facilities Management operation, that the emergency power generator has been successfully tested for operation and "switch-over" during the previous year.

Basis. It is important for safety that verification of emergency power functions be carried out.

Past experience has shown that this frequency is adequate to assure continuity of this service.

4.7 Radiation Monitoring System Applicability. This specification applies to the surveillance requirements for the radiation monitoring instrumentation required by Section 3.7.1.a of these specifications.

Objective. The objective is to assure that the radiation monitoring system is operating properly and to verify the appropriate alarm settings.

Specifications.

a. A channel test of the radiation monitoring systems required by Section 3.7.l.a. shall be performed daily. This shall include verification of the alarm set points.

b. A channel calibration of the radiation monitoring systems required by Section 3.7.l.a. shall be performed annually.

Basis. Surveillance of the equipment will assure that sufficient protection against radiation is available. Past experience has shown that these frequencies are adequate to assure proper operation.

UCI Technical Specifications 2010-02 page 21

5.0 DESIGN FEATURES 5.1 Site and Facility Description Specifications

a. The reactor facility is housed in a closed five room suite on a single level in the basement of Rowland Hall, on the University of California Irvine campus. Three interconnected rooms (designated B54, B54A, B54B comprise the "reactor room" area and shall be controlled for radiological work by prohibiting eating and drinking. The control room (B62), and an outer office/counting room (B62A) comprise a separate area with a separated ventilation system where eating and drinking is permitted. Access to the control room (B62),which controls access to the reactor room is security controlled. The reactor room is readily visible through glass from the control room. The minimum free air volume of the reactor room shall be 10,000 cubic feet.

b. A ventilation system shall be capable of exhausting air or other gases from the reactor room at a minimum of 70 feet above ground level.

c. Emergency shutdown of the specific reactor area exhaust and air input ducts shall be controlled by a high radioactive particulate count rate alarm signal originating in the reactor room.

Basis. The extent of the facility is specified to define the controlled areas so as to establish ventilation and other controls needed to assure minimal radiation exposure to facility personnel and the community.

5.2. Reactor Coolant System Specifications.

a. The reactor core is cooled by natural convection water flow.

b. All piping and other equipment for pool cooling systems is above pool level and inlet and outlet pipes to the heat exchanger and demineralizer are equipped with siphon breaks not less than 14 feet above the upper core grid plate.

c. A pool water level indication is provided at the control console.

d. A pool water temperature indication is provided at the control console.

e. A pool water conductivity measurement instrument is provided in the reactor room.

f. Gamma and beta radiation spectrometry equipment is provided for water sample radioactivity assay.

Basis. Pool water quantity and quality is controlled so as to limit radiation and/or radioactivity release, corrosion of components. Information is necessary to provide staff with indications of change in pool water characteristics.

UCI Technical Specifications 2010-02 page 22

5.3. Reactor Core and Fuel 5.3.1 Reactor Core.

Specifications.

a. The core assembly consists of TRIGA fuel elements.

b. The core fuel shall be kept in a close-packed array except for control rods, single-or three-element or seven-element positions occupied by in-core experiments, irradiation facilities (including transfer system termini), graphite dummy elements, and a central dry tube.

c. Reflection of neutrons is provided by combinations of graphite and water, with the graphite in sealed containment with aluminum cladding, either in the form of rods occupying grid positions, or in a larger reflector structure surrounding the core.

d. An Am-Be neutron source is provided in a specific location in the upper grid plate to provide start-up neutrons. It may be removed for maintenance purposes.

Basis. TRIGAO fuel and reactor design has a long and successful history of use when used under specific conditions. The Am-Be source is in a sealed capsule and has a long useful life.

5.3.2. Control Rods.

Specifications.

a. The SHIM and REG rods are motor driven with scram capability and solid boron compounds in a poison section, with fuel followers of standard TRIGAO fuel meeting the same specifications as in Section 5.3.3.

b. The transient rods (ATR and FTR) are pneumatically driven, have scram capability and contain solid boron compounds in a poison section. They incorporate air filled followers.

The ATR has an adjustable upper limit to provide variable pulse insertion capability.

Basis. These control rods have a long successful history of use.

5.3.3. Reactor Fuel.

Specifications. Standard unirradiated TRIGA fuel elements have the following characteristics:

a. The total uranium content shall be nominally 8.5 % by weight, enriched to less than 20%

235U.

b. The hydrogen to zirconium atom ratio in the zirconium hydride shall be a nominal 1.65 hydrogen atoms to 1.0 zirconium atom.

c. The cladding shall be 304 stainless steel, nominally 0.020 inches thick.

d. An upper fitting designed to fit a latching tool for fuel movement, with engraved unique serial numbers.

UCI Technical Specifications 2010-02 page 23

Basis. TRIGA fuel elements meeting these manufacturer's specifications have a long history of successful use with minimal failures. Minor deviations about these levels due to manufacturing variations are not to be considered violations of this specification.

5.4. Fuel Storage Specifications.

a. All fuel elements shall be stored in a geometrical array where the keff is less than 0.90 for all conditions of moderation and reflection.

b. Irradiated fuel elements and fuel devices shall be stored in an array which will permit sufficient natural convection cooling by water or air such that the fuel element or fueled device temperature will not exceed 80'C.

c. Fuel showing evidence of damage (see TS 3.1.6) shall be stored separately from fuel not suspected to be damaged, and shall be checked thoroughly for fission product leakage.

Basis. These specifications establish a sufficient reactivity margin to guard against accidental criticality of elements in storage, and that heat dissipation does not create excess corrosion or other problems. Damaged fuel is more likely to have or develop fission product leakage and so must be monitored and kept separately.

UCI Technical Specifications 2010-02 page 24

6 0 ADMINISTRATIVE CONTROLS 6.1 Organization The UCI Nuclear Reactor operations involve no shift work and mostly short operating schedules, only a few times a week at most. This necessitates only a small staff, not necessarily full-time.

6.1.1 Structure.

The reactor facility is housed in the School of Physical Sciences of the University of California, Irvine.

The official licensee of the reactor is the Board of Regents of the University of California, who has delegated authority for license matters to the Executive Vice Chancellor and Provost of the University of California, Irvine. The reactor is related to the University structure of positions shown in the organization chart.

6.1.2 Responsibilities.

a. The reactor facility is under the direction of a Reactor Director who shall be a tenure member of the Chemistry Department faculty. Operations are supervised by the Reactor Supervisor who shall hold a valid senior operator's license for the facility. This position shall be responsible for assuring that all operations are conducted in a safe manner and within the limits prescribed by the facility license, the provisions of the Reactor Operations Committee and the provisions of the UCI Radiation Safety Committee. The Reactor Director and Reactor Supervisor positions may be occupied by the same individual.

b. There is a UCI Radiation Safety Officer responsible for the safety of operations from the standpoint of radiation protection. This position reports to the Office of Environmental Health and Safety which is an organization independent of the reactor operations organization as shown in Chart I. An independent campus-wide Radiation Safety Committee (RSC) is responsible for establishment and review of all policies involving radiation and radioactivity.

c. In the event of absence of specific individuals, temporary duties and responsibilities may be assumed by the person next in line in the organization chart.

UCI Technical Specifications 2010-02 page 25

Executive Vice Chancellor and Provost Vice Chancellor Administration and Business Services Associate Vice Chancellor, ABS I

I I

I I

I L -

I UCI Technical Specifications 2010-02 page 26

6.1.3 Staffing.

a. The minimum staffing when the reactor is operating shall include:

1. A licensed operator with direct access to the reactor controls;

2. A second designated individual present within Rowland Hall with the ability to check on the safety of the licensed operator and to act in the event of emergency; and

3. A licensed Senior Operator (SRO) on call and expected to be available at the facility within 30 minutes, if called.

b. A list of reactor facility personnel, and other persons responsible for radiological safety and security on campus shall be kept in the reactor control room for use by an operator. The list shall include the Reactor Director, the Reactor Supervisor, the Radiation Safety Officer and other back-up radiological safety personnel, senior or other licensed operators, and facilities management personnel with responsibilities for maintenance of Rowland Hall.

c. The following events require the presence in the facility of a licensed Senior Reactor Operator:

1. initial daily start-up checkouts, including approach to critical;

2. fuel or control-rod relocation within the core region;

3. insertion, removal, or relocation of any experiment worth more than $1.00; and

4. restart following any unplanned or unscheduled shutdown, or significant power reduction.

6.1.4 Selection and Training of Personnel.

The selection, training, and requalification of operations personnel should, where applicable, meet the requirements of ANSI/ANS-15.4, latest revision.

6.2 Review and audit A Reactor Operations Committee (ROC) shall review reactor operations to assure that the facility is operated in a manner consistent with public safety and within the terms of the facility license. Review and audit of radiological safety at the facility is also carried out by the UCI Radiation Safety Committee.

6.2.1 ROC Composition and Qualifications The ROC shall have at least five members, at least one of whom shall be a health physicist designated by the Office of Environmental Health and Safety of the University. The Committee as a whole shall be knowledgeable in nuclear science and issues related to reactor and/or radiological safety. The membership shall include at least two members who are not associated with the Department of Chemistry. Approved alternates may serve in the absence of regular members. Members and alternates and a chairperson for the committee shall be appointed by the Chair of the UC Irvine Department of Chemistry or higher authority.

6.2.2 ROC Charter and rules The following responsibilities constitute the charter of the ROC.

1. Meeting semi-annually, with provision for additional meetings when circumstances warrant to assure UCI Technical Specifications 2010-02 page 27

safety at the facility.

2. A quorum shall consist of not less than a majority of the members and shall include the chairperson or his/her designee. Votes shall not be taken where a majority of those voting would be directly associated with facility operations.

3. Review and audit of facility staff and operations as indicated further below.

4. Designation of individuals to perform audits of facility operations and records.

5. Preparation, approval, and dissemination of minutes of meetings.

6. Preparation and dissemination of findings and other reports as needed to assure safe operatios of the reactor.

6.2.3 ROC Review function The following review functions shall be the responsibility of the ROC.

1. Review and approval of all proposed changes to the facility, its license, procedures, ROC chater, and Technical Specifications, including those made under provisions of 10 CFR 50.59, and the determinations leading to decisions relating to 50.59 approvals;

2. Review and approval of new or changed procedures, experiments, or instruments having safety significance;

3. Review of new experiments or changes in experiments that could have reactivity or safety significance;

4. Review of violations of technical specifications, license, or violations of procedures or instructions having safety significance;

5. Review of operating abnormalities that have safety significance.

6. Review of reportable occurrences listed in Sections 6.6.1 or 6.7.2;

7. Review of audit reports.

6.2.4 ROC Audit function The ROC shall audit or review audits performed by designated individuals on its behalf at least annually. The audit shall include, but not be limited to:

1. Facility operations for conformance to the technical specifications and applicable license or other conditions;

2. Retraining and requalification of operators according to the Requalification Plan;

3. The result of action taken to correct those deficiencies that may occur in the reactor facility equipment, systems, structures, procedures or methods of operation that affect reactor safety; UCI Technical Specifications 2010-02 page 28

and

4. The facility Emergency Plan (EP) and implementing procedures including written reports of any drills or exercises carried out. A full audit of the EP should be conducted biennially by the ROC and/or RSC.

6.3 Radiation Safety As delineated in section 6.1.2.b, the UCI Radiation Safety Officer (RSO) is responsible for implementation of the radiological safety program at the reactor facility in accordance with applicable federal and state of California standards and regulations. The program should use the guidelines of ANSI/ANS 15.11.

6.4Property "ANSI code" (as page type) with input value "ANSI/ANS 15.11.</br></br>6.4" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process. Operating Procedures Written procedures, reviewed and approved by the ROC, shall be in effect and implemented for the following items. The procedures shall be adequate to assure the safety 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. Maintenance of major components of systems that could have an effect on reactor safety.

4. Surveillance checks, calibrations and inspections required by the technical specifications or that could have an effect on reactor safety;

5. Personnel radiation protection, including provisions to maintain personnel exposures as low as reasonably achievable (ALARA);

6. Administrative controls for operations and maintenance, and for the conduct of irradiations or experiments that could affect reactor safety;

7. Implementation of required plans including Emergency (EP) and Physical Security (PSP) plans;

8. Shipping and/or transfer of radioactive materials.

Substantive changes to the above procedures shall be made only with the approval of the ROC. Temporary changes to the procedures that do not change their original intent may be made by the Reactor Supervisor.

All such temporary changes to procedures shall be documented and subsequently reviewed by the Reactor Director and the ROC. Substantive changes affecting radiological safety should be made only with the approval of the RSO. Temporary, minor, changes in radiological safety procedures may be made by the Reactor Supervisor, but should be reported to the RSO as soon as possible.

UCI Technical Specifications 2010-02 page 29

6.5 Experiment Review and Approval Approved experiments shall be carried out in accordance with established and approved procedures.

Procedures for experiment review and approval shall include:

1. All new experiments or class of experiment shall be reviewed and approved by the ROC and approved in writing by the Reactor Director. The review shall include analysis by the RSO or other designated radiation safety personnel.

2. Substantive changes to existing experiments or classes shall be made only after review by the ROC and RSO or their designees. Minor changes that do not significantly alter the experiment may be approved by a senior reactor operator (SRO).

6.6 Required Actions 6.6.1 Actions To Be Taken In Case of a Safety Limit Violation.

In the event the safety limit on fuel temperature is exceeded:

1. The reactor shall be shut down and reactor operation shall not be resumed until authorized by the NRC.

2. The event shall be reported immediately to the Reactor Director, the ROC chairperson, and the RSO.

3. The event shall be reported to the NRC Operations Center within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />;

4. A report, and any applicable follow-up report, shall be prepared and reviewed by the ROC, for submission to NRC, describing:

c. applicable circumstances leading to the violation including, where known, the cause and contributing factors;

d. effects of the violation upon reactor facility components, systems, or structures, and on the health and safety of personnel and the public; and

e. corrective action to prevent occurrence.

6.6.2 Actions to be taken in the event of an occurrence of the type identified in Section 6.7.2, other than a safety limit violation.

1. The reactor shall be secured and the Reactor Director and/or Supervisor notified.

2. Operation shall not be resumed until authorized by the Reactor Director and/or Supervisor.

3. The occurrence shall be reported to NRC as required in Section 6.7.2 of these specifications, and reviewed by the ROC at their next meeting.

UCI Technical Specifications 2010-02 page 30

6.7 Reports In addition to the requirements of applicable regulations, and in no way substituting for them, reports shall be made to the NRC as follows:

6.7.1. Annual Operating Report.

A routine annual report shall be submitted by the Reactor Director to NRC at the end of each 12-month period for operations for the preceding year's activities between July 1st through June 3 0th. The report shall include:

1. a brief narrative summary of operating experience (including experiments performed) and a tabulation showing the energy generated by the reactor (in megawatt hours), the amount of pulse operation, and the number of hours the reactor was critical;

2. the number of unplanned shutdowns and inadvertent scrams, including the reasons therefore, and corrective actions taken (if any) to reduce recurrence;

3. a tabulation of major preventive and corrective maintenance operations having safety significance;

4. a tabulation of major changes in the reactor facility and procedures, and tabulations of new experiments that are significantly different from those performed previously, including a summary of safety evaluations performed to assess that they do not require prior NRC approval and are authorized by 10CFR 50.59;

5. a summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the facility as measured at or prior to the point of such release or discharge. The summary shall include, to the extent practicable, an estimate of individual radionuclides present in the effluent. If the estimated average release after dilution or diffusion is less than 25% of the concentration allowed, a statement to this effect is sufficient;

6. a summarized result of environmental surveys performed outside the facility; and

7. a summary of radiation exposures received by facility personnel and visitors, where such exposures are greater than 25% of that allowed.

6.7.2 Special Reports.

1. A report not later than the following working day by telephone to the NRC Operations Center, and confirmed in writing by FAX, to be followed by a written report that describes the circumstances of the event within 14 days to the USNRC Document Control Desk, of any of the following:

a.. violation of a safety limit (fuel temperature);

b. release of radioactivity from the site above allowed limits;

c. operation with actual safety system settings for required systems less conservative than the limiting safety system settings in these specifications;

d. operation in violation of limiting conditions for operation unless prompt remedial action is taken as permitted in section 3; UCI Technical Specifications 2010-02 page 31

e. a required reactor safety system component malfunction that renders or could render the safety system incapable of performing its intended safety function. If the malfunction or condition is caused by maintenance, then no report is required;

f. an unanticipated or uncontrolled change in reactivity greater than one dollar. Reactor trips resulting from known cause are excluded;

g. abnormal or significant degradation in reactor fuel or cladding, or both, coolant boundary, or confinement boundary (excluding minor leaks) where applicable; or

h. an observed inadequacy in implementation of administrative or procedural controls such that the inadequacy causes or could have caused the existence or development of an unsafe condition with regard to reactor operations.

2. A report within 30 days (in writing to the Document Control Desk, USNRC, Washington, D. C. 20555) of:

1. permanent significant changes in facility organization; and

2. significant changes in the transient or accident analyses as described in the SAR.

6.8 Records In addition to the requirements of applicable regulations, and in no way substituting therefore, records and logs shall be prepared and retained for periods as described here. Records may be in a variety of formats.

6.8.1 Records to be retained for a period of at least 5 years or for the life of the component involved if less than 5 years.

1. Normal reactor facility operation, but not including supporting documentation such as checklists, log sheets, etc., which shall be retained for one year.

2. principal maintenance activities;

3. reportable occurrences;

4. surveillance activities required by the Technical Specifications;

5. reactor facility radiation and contamination surveys;

6. experiments performed with the reactor;

7. fuel inventories, receipts and shipments;

8. approved changes in operating procedures; and

9. ROC records of meetings and audit reports.

UCI Technical Specifications 2010-02 page 32

6.8.2 Records to be retained for at least one certification cycle.

Records of retraining and requalification of licensed operators (and SRO's) shall be retained at all times the individual has duties as an operator or his or her license is renewed.

6.8.3 Records to be retained for the lifetime of the reactor facility.

Applicable annual reports containing this information may also be used as records for the following:

1. gaseous and liquid radioactive effluents released to the environs;

2. off-site environmental monitoring surveys;

3. radiation exposures for all personnel monitored; and

4. drawings of the reactor facility and safety related components.

UCI Technical Specifications 2010-02 page 33