ML100290365

University of California, Irvine - Reactor Response to NRC Request for Additional Information Dated December 3, 2009 Re License Renewal Request

ML100290365
Person / Time
Site:	University of California - Irvine
Issue date:	01/27/2010
From:	Geoffrey Miller University of California - Irvine
To:	Cindy Montgomery Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC ME1579
Download: ML100290365 (66)

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 Departmentof Chemistry and FAX: (949) 824-6082 or (949) 824-8571 Director,Nuclear Reactor Facility Email: gemiller@uci.edu FacultyAdvisor for Science Website: http://chem.ps.uci.edu/-gemiller/

UCI Centerfor EducationalPartnerships January 27, 2010 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 Cindy Montgomery, 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 December 3 rd, 2009 (TAC No. ME 1579) 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 January 27t' 2010 Sincerely yours, George E. Miller Director Nuclear Reactor facility Senior Lecturer Emeritus, Department of Chemistry, University of California, Irvine

University of California, Irvine Reactor License R-116, Docket 50-326 Response to NRC Request for Additional Information (RAI) dated December 3 rd 2009 (TAC NO. ME15 79)

We are pleased to respond as follows to the Technical Questions and Comments. Please note that a number of attachments form an integral part of this response. We will be pleased to respond to any further questions you may have.

Chapter 2 2.1 Chapter 2.6 indicates that at the time of application,new information regardingthe seismicity of the UCI areawas being reviewed and that, when complete, this information would be used to supplement the license application.Provide a more complete descriptionof the seismicity of the UCINRF environs.

Please see Attachment A.

2.2 Confirm that the nearestpermanent residences are located approximately 280 meters (310 yards) southeastof Rowland Hall.

The nearest residences to the reactor are located at the following street addresses: 18 Blake Court, Irvine 92617; 2019A Las Lomas Apartments, Los Trancos Drive, Irvine 92617. This should enable more precise estimation of distances. These residences currently have a "line of sight" to the roof level of Rowland Hall. Both these residences are within the University Hills Community, a campus-owned restricted housing community for faculty and staff operated by the Irvine Campus Housing Authority (see www. icha.uci.edu.)

Chapter 3 3.1 Confirm that the seismic upgrades to the building housing the reactor,as described in Appendix A, were completed If the structure as built differsfrom the Appendix A description,note any differences.

Our information is that these upgrades are fully completed according to original design specifications as submitted.

Chapter 4 4.1 Section 4.2, Fuel-ModeratorElements, Page 4-5. Provide the burnup limits on the fuel and describe how burnup is monitored No bum-up limits have ever been suggested or established for the standard TRIGA fuel elements used at this facility, while experiments indicate bum up to over 50% results in no significant fuel degradation. A statement to this effect has been included in the proposed Technical Specifications. Bum-up is monitored by power calculations based on element core locations and estimated fluxes at each element position.

4.2 Section 4.2, Fuel-ModeratorElements, Page 4-5. Propose Technical Specification (TS) wording that meets the structure of ANSI!ANS-1 5.1. Considermoving the fuel element requirements to Section 3, "Limiting Conditionsfor Operation"and the Page 1 of 4 UCI Response to RAI January 2010

surveillance requirements to Section 4, "SurveillanceRequirements" of the TSfrom their currentlocation in Section 2, "Safety Limits."

Please see Attachment B. The Technical Specifications have been restructured and re-written to meet these and other suggestions.

4.3 Section 4.5.2, ReactorNuclear Parameters,Table 4-2, Page 4-21. Provide a referencefor the prompt temperature coefficient value of13.4 x 1 Q Ak/k! 'C.

The prompt temperature coefficient was provided in the Attachment to Bid document by the manufacturer based on core modeling calculations at that time. [GACP-71-434, Table 2.4, page 2-4]

4.4 Section 4.5.2, ReactorNuclear Parameters.Propose TS wordingfor a surveillance requirementfor excess reactivity and shutdown margin to correspondwith the existing Limiting Conditionfor Operation.

Please see response to 4.3 above. These have been moved to an appropriate surveillance section.

4.5 Thermal-HydraulicAnalysis. Section 5.1, Introduction,Page 5-1. Section 4.6, ThermalHydraulicDesign, Page IV-25. Providejustificationfor the statement that a thermalhydraulicanalysis of the UCINRF reactor is not warranted.This may be done by reference to an acceptable analysisfor afacility with similar or bounding relevant design parameters or by analysis.

A thorough thermal analysis performed for McClellan Research reactor (UC Davis- SAR Section 4,5.2 and 4.6) and further utilized by Oregon State reactor shows that these fuels may be safely employed with convection cooling in similar core designs at power levels up to 2 megawatts, well over the operating power level of the UCI reactor. An absence of film boiling with natural convection is predicted up to 2.3 Mw.

4.6 Provide the Reactor ProtectionSystem setpointsfor fuel element temperature and reactorpower level scrams.

Setpoints currently utilized are 475 degrees C for fuel temperature and just under 110%

(275 kilowatts) for power level scrams.

Chapter 5 5.1 Section 5.2, Water Cooling System, Page 5-1. Providefigures 5.1, 5.2, and 5.3 referencedin this section.

Please see Attachment C for these figures. The complete chapter has been revised and is included for clarity.

5.2 Section 5.2, Water Cooling System, Page 5-1. Propose TS wording on coolant temperature limits.

Please see response to 4.3 above. This has been included in the revised TS proposal (Section 3.3.2) 5.3 Provide a more complete descriptionof the secondary cooling system in accordance with NUREG-1537. The descriptionshould include the provisions to detect leakage through the heat exchangers.

Attachment C shows the cooling system more clearly. The secondary system is part (a branch line) of the campus-wide cooling system operated mostly for air-conditioning. To Page 2 of 4 UCI Response to RAI January 2010

preclude outward leakage, the pressure on the secondary side is maintained higher than pressure on the primary side. Leakage into the reactor pool water is monitored by conductivity measurements as the secondary cooling system is highly loaded with anti-corrosion chemicals that would be easily detected in small amounts by a sudden rise in pool water conductivity.

Chapter 7 7.1 The Figuresfor Chapter 7 were not included in the SAR. Provide the missingfigures.

Please see Attachment D. The complete and updated chapter has been provided to show the current instrumentation.

Chapter 9 9.1 Variousfigures showing the Ventilation system (Figure 3-3 and higher) were not included in the SAR. Provide these missingfigures.

Please see Attachment E. for these figures.

Chapter 12 12.1 Provide a description of the minimum facility staffingfor operationsand describe the activitiesfor which a SRQ is requiredto be present.Also, propose an appropriateTS for minimum staffing.

Please see Attachment B. These have been included in the TS revision, section 6.

12.2 Describe andprovide revisions to TSfor review and auditfunctions of reactor operationsand activities in accordancewith the guidance ofANSI/ANS-15.1.

Please see Attachment B. These have been clarified in Section 6.

Chapter 13 13.1 Section 13.2, Maximum HypotheticalAccident. Clarify what the anticipated occupationalandpublic doses would befor the MHA. The doses should be presented as TEDE and CDE to the thyroid. The assumptionfor the exposure time offacility personnel should be clearly stated.

Please see Attachment F for comment on this section.

13.2 Section 13.4, Loss of Pool Water, Page 13-15. Provide a descriptionof the seismic analysis including the building, reactorpool, and any equipment consideredessentialfor placing the reactorin a safe shutdown condition. For example, the analysis should show that no binding in the control rod mechanisms or otherfailure mechanisms resulting from a seismic event could preventfull insertion of all rods on a seismic scram initiation.

The analysis should also show that no equipment consideredessentialfor placingthe reactor in a safe shutdown condition could be adversely impactedby other structuralor equipment damage caused by the seismic event, e.g. falling debris.

Please see Attachment G for this analysis.

Page 3 of 4 UCI Response to RAI January 2010

Chapter 14 14.1 TS 3.4, Reactor Safety System, Page 10. For the Reactor Safety System, consider including an interlock to prevent withdrawalof standardrods in pulse mode.

Please see Attachment B. This has been added. (Section 3.2.3.)

14.2 TS 3.6, Ventilation System, Page 11, consider including:

1) a negative differentialpressure requirement

2) a minimum emergency exhaustflow rate

3) a maximum leak

4) all reactor bay external doors closed andpersonneldoors not blocked open

5) no fuel movement during maintenance outage

6) wording changedto clarify that the outage exception applies only to emergency exhaust system, i.e. normalventilation and negative DP are always requiredforreactor operation,fuel movement, etc.

Also, considerproposingcorrespondingSurveillance TS wordingfor parameters1, 4, 5, and 6 above.

Please see Attachment B for proposed revisions to TS. (Section 3) 14.3 TS 3.7, Pool Water Level, Page 12. Providejustificationfor why the pool water level TS does not include a level below which the reactor cannot be operatedor propose appropriate TS wording.

Please see Attachment B, for proposed revision. (Section 3.3.1).

FinancialQuestions and Comments We regret we anticipate considerably more time is needed to compile the necessary information, especially as relating to decommissioning cost estimates in a time of volatile financial expectations in the construction industry, and uncertainty in state and educational budgets.

Page 4 of 4 UCI Response to RAI January 2010

ATTACHMENT A Update on Seismic Background Information for UCI Reactor The following references and discussion are an update to that submitted in 1999. UCI is located in the Tustin Quadrangle Section mapping project of the State of California.

References.

1. The most recent discussion (March 2009) of seismicity of the immediate nearby region is contained in Chapter 5 of the Irvine Business Center Vision Plan and Mixed Use Overlay Zoning Draft EIR available on a web site at:

http://www.cit-yofirvine.o rg/civica/filebank/blobdioad.asp?BlobID=13547

2. An official state of California evaluation updated in 1998/2001 is available at:

http://gmw.consrv.ca.gov/shmp/download/evalrpt/tus eval.pdf

3. The emerging ideas on thrust-faulting as a characteristic of the region are described in an in Geology (1999):abstractavailable at:

http://geolab.seweb.uci.edu/g 99 abstract.pdf To quote from reference 1.

"It is thought that a blind thrust fault, that is, a fault that does not extend to the surface, may exist beneath the San Joaquin Hills, based on indirect evidence. This supposed San Joaquin Hills blind thrust is recognized by the California Geological Survey to be active, although is not in an Alquist-Priolo Earthquake Fault Zone due to its blind nature."

"As with all of Orange County, the project site is in the Uniform Building Code Seismic Zone 4. This is the highest classification of the four zones in the United States, with the most stringent requirements for building design. The project site is also mostly in the City of Irvine Seismic Response Area (SRA) 1, although portions of the project site southwest of Michelson Drive are mostly in SRA 2 (according to Figure D-3 in the City of Irvine General Plan). SRAs describe the different types and magnitudes of potential seismic hazards, making it possible to evaluate the risks of property damage, personal injury, and loss of vital services that may result from an earthquake. In SRA 1 the predominant characteristics are soft soils and high groundwater, and the predominant characteristics in SRA 2 are denser soils and deeper groundwater. In SRA 1, liquefaction is the primary potential seismic hazard. In SRA 2, ground motion is the primary potential seismic hazard.

Peak horizontal ground acceleration (PHGA) is generally used to measure the amplitude of a particular ground motion. The PHGA values for the site were estimated using probabilistic seismic hazard analyses, based on currently available earthquake and fault information.

A probabilistic seismic hazard analysis was performed using the United States Geological Survey Earthquake Hazards website to estimate the PHGA for the site. Various probabilistic density functions were used in the analysis to assess the uncertainty inherent in the calculation with respect to magnitude, distance, and ground motion. The results of the analysis suggest that the PHGA in alluvial conditions, such as those on the site, with a 10 percent probability of exceedance in 50 years (that is, a recurrence interval of 475 years) is approximately 0.34g, or 34 percent of the acceleration of gravity. This level of ground motion is considered the Design Basis Earthquake."

The "project site" referred to in this reference is adjacent to UCI, but at a lower elevation closer to the marshland area.

Attachment A UCI Response to RAI Page 1

Maps from reference 2 and the latest hazard potential mapare attached identifying the location of Rowland Hall. The immediate locale is not identified as being subject to unusual hazard from landslides or liquefaction, though such possibility does apply to other areas on campus that were extensively land-filled. This reference estimates the acceleration 10% probability of exceedance within 50 years of a point near Rowland Hall as 0.30g.

Reference 3 concludes" The San Joaquin Hills have risen at a rate of 0.21 - 0.27 m/k.y.

(meters/l1000 years) during the past 122 k.y. ....movement has the potential to generate a Mw 7.3 earthquake .... An estimated minimum average interval of-1650-31900 myr for moderate sized earthquakes." However this research is still ongoing and has not been totally accepted as final.

In conclusion, recent studies and events indicate that new information, while adding considerably to the understanding of the fault zones and hazards in southern California, do not change the general conclusion regarding level of risk as presented at the time of original licensing, and in the application for relicense in 1999. The building itself has been upgraded to meet later codes to reduce the likelihood of major damage from seismic activity.

Page 2 Attachment A A UCI Response to UCI Response to RAI RAI Page 2

Open-File Report 97-20 O0 OO S *

• S

• 0 D

0 00 Base map enlarged from U.S.G.S. 30 x 60-minute series Plate 2.1 Landslide inventory, Shear Test Sample Locations, Tustin Quadrangle.

0 shear test sample location 1 landslide P areas of significant grading 0~cA L -r IDtJ 0F gofJL J (..L ONE MILE SCALE A.f. 3.

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SCALE 124,000 I-o 0"ýJC 4-*tL-L l

STATE OF CALIFORNIA C

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~o Chaptr7J, D*vton 2 of Y1 C.Wftnt Puba Remvumw Code TUSTIN QUADRANGLE OFFICIAL REVISED MAP Released: January 17, 2001

1998 SEISMIC HAZARD EVALUATION OF THE TUSTIN QUADRANGLE 35 TUSTIN 7.5 MINUTE QUADRANGLE AND PORTIONS OF ADJACENT QUADRANGLES 10% EXCEEDANCE IN 50 YEARS PEAK GROUND ACCELERATION (g) 1998 rIIM Rl"1*ri rnmnITIflN.Q 0 2.5 Kil-ometers 5 Department of Conservation Division of Mines and Geology

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Figure 3.1 WA LL L 0 CA-iON A.P-S

Attachment B Appendix A Proposed Technical Specifications for the U. C. Irvine TRIGA Mark I Nuclear Reactor Submitted January 2010

TABLE OF CONTENTS DEFINITIONS I 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, 4.5 Ventilation Systems 20 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 22 5.4 Fuel Storage 23 ADMINISTRATIVE CONTROLS 6.1 Organization 24 6.2 Review and Audit 25 6.3 Radiation Safety 27 6.4 Operating Procedures 27 6.5 Experiment Review and Approval 27 6.6 Required Actions 28 6.7 Reports 28 6.8 Records 30

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 (kefr = 1) at reference core conditions or at a specified set of conditions.

UCI Technical Specifications 2010-02 page 1

1.11 EXPERIMENT An 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 occur 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 configuration 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 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.

1.28 SHALL The word SHALL implies a specific action required by the license or regulation.

1.29 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.

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

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

1.32 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 UCI Technical Specifications 2010-02 page 3

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.

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 900'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 900 0 C. A conservative value 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 Safety System Setting B-ring 500 0 C C-ring 450 0 C 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 UCI Technical Specifications 2010-02 page 5

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. Maximum recorded temperatures for the UCI reactor B-ring IFE for the period since 1969 are 250TC at steady state, and 350TC for pulse operation. The limiting safety system settings are set to provide a considerable margin between operational 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 275 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 circular grid plate core configuration. Operation at less than 20% of that value will assure that limits are never approached.

• It is intended that normal operations be conducted at 250 kilowatts. Brief periods of operation at power levels up to 10% above will permit direct ("live") testing of a scram level set below that and avoid violations due to fluctuations occurring as a result of cooling water stirring and as a result of sample loading variations in the rotating specimen rack. These latter have been observed in past operations to be within +/- 4% of full power.

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 non-secured 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. These specifications apply to the reactivity condition of the reactor and the reactivity worths of the control rods. They apply for all modes of operation.

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 conditions exist:

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

b. The total reactivity worth of the two transient control rods (ATR + FTR) 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. This specification applies 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, the steady-state power level of the reactor is less than 1 kilowatt.

Basis Inadvertent pulsing from a high steady-state power level could result in a higher final peak temperature approaching the safety limit. TS 3.1.3.b establishes a limit on any planned pulse reactivity insertion to limit temperature rise to anticipated values.

3.1.5 Fuel Burnup. No specification. Bumup tests performed by General Atomic (OSU SAR, page 10.) have shown that TRIGA fuels may successfully be used without significant fuel degradation to burnup in excess of 50% of the contained 235U.

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 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/ 6 th inches (0.0625 in) over the length of the element;

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

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

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

Basis These criteria have been successfully used for hundreds of fuel inspections over many years 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.

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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 for the purpose of diagnosis and/or channel tests or checks on that channel.

Note 2. Any single power level channel that ceases to be operable during reactor operation shall be returned to operating condition within 5 minutes or the reactor shall be shut down.

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.

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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 - 5000 C 1 1 Temperature Reactor Power level Scram - 110% of 250 kw 1 High Voltage loss Scram - loss of HV on any 1 1 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 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 specification applies 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.

Specifications.

a. The reactor shall not be operated unless the pool water level is at least 15 feet above the core (at least 23 feet above the tank floor, or no more than 2 feet below the tank edge.

b. An alarm shall alert personnel 24/7 if the water level in the reactor pool falls below the above limit.

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 significant increase in radiation levels to the surroundings occurs.

UCI Technical Specifications 2010-02 page 11

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 17C and 25°C 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" cooling water system Specification. The pool water conductivity level shall be maintained less than 3 mhos/cm.

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

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.

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

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 UCI Technical Specifications 2010-02 page 12

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.

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.

UCI Technical Specifications 2010-02 page 13

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.

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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.

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 4 1Ar.

Objective. To assure that concentration of 4 1Ar in unrestricted areas shall be below the applicable limits of 10 CFR Part 20.

Specification. The annual average concentration of 4 1Ar discharged into an unrestricted area shall not exceed 4 x 10-6 jtc/ml at the point of discharge.

Basis. It is shown in Chapter 13 of the SAR that the release of 4 1Ar at the above concentration will not result in exposures in unrestricted areas of less than 10 mrem TEDE (Reg Guide 4.20).

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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 be less than $1.00, and

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

c. The sum of absolute values of all experiments shall be less than $3.00 Basis. The limit on 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 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.

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 and c reduce the likelihood of damage UCI Technical Specifications 2010-02 page 16

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 rooms or the unrestricted area will not exceed the applicable does limits in 10CFR 20, assuming that:

a. 100% of the gases or aerosols escape form 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 10CFR20 by reducing the likelihood of excessive personnel exposure by gases or aerosols as a result of experiment failure.

3.9. Facility-specific Requirements Following Extended Shutdown.

Applicability. This specification applies during prolonged periods of shutdown when the reactor has not been operated.

Objective. To assure that all reactor systems are fully functional before resuming normal operations.

Specification. Surveillance activities that require reactor operation in order to be accomplished may be suspended during long periods when reactor operation for experiments or training purposes is not required. In this case, all required surveillance shall be satisfactorily completed within 30 days of resuming any operation.

Basis. This specification assures that the' reactor safety systems are fully operational, and that all reactor parameters are within expected specifications before extensive operations are conducted following a prolonged period of shutdown.

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

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

Specification. Any system that has been changed (additions, modifications, maintenance) shall be tested using established surveillance methods before the system or component of a system is declared operable so reactor operation may proceed.

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

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.

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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 shall be visually inspected for deterioration quinquennially.

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

d. 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.

e. 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.

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

g. A calibration of the temperature measuring channels shall be performed annually. This calibration shall consist of introducing electric potentials in place of the thermocouple input to the channels. It shall also include a comparison to early 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.

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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.

This includes 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 and alert set points.

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.

Basis. These verifications will assure that a continued warning system for a 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.

4.4 This surveillance is dealt with in the following section.

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 ventilation system's ability to maintain negative 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 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.

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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.

Specification.

a. A channel test of the radiation monitoring systems required by Section 3.7.1 .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 3.7.1 .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.

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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.

The control room (B62), and an outer office/counting room (B62A) comprise a separate area with a separated ventilation system. The reactor room is readily visible through glass from the control room.

b. The building is equipped with a ventilation system 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 is controlled by a high radioactive particulate count rate alarm signal in the reactor room.

5.2. Reactor Coolant System Specification.

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 alarm is specified in Section 3.3.1 of these specifications.

d. A pool temperature indication is provided at the control console during reactor operation.

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 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.

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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 TRIGA fuel meeting the same specifications as in Section 5.3.1.

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.

5.3.3. Reactor Fuel Specifications. Standard TRIGA fuel elements have the following characteristics:

a. uranium-zirconium hydride, nominally 8.5 % by weight uranium, with a maximum enrichment of 20 percent 235u.

b. 1.55 to 1.65 hydrogen atoms to 1.0 zirconium atom in the zirconium hydride.

c. 304 stainless steel cladding, nominally 0.020 inches thick.

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

5.4. Fuel Storage.

Specifications

a. All fuel elements shall be stored in a geometrical array where the keff is less than 0.9 for all conditions of moderation.

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 1000 C.

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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 Chart I.

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 Chart I.

CHART I UCI Executive Vice Chancellor and -Provost Dean, School of Physical Sciences Vice-Chancellor, Business and Administrative Services Chair, Department of Chemistry Vice-Chair Facilities, Dept of Chemistry Director, Environmental Health and Safety UCI Reactor Operations Committee UCI Radiation Safety Committee Reactor Director Radiation Safety Officer Reactor Supervisor Radiological Safety Technologists Reactor uperators Radiological Safety Technicians UCI Technical Specifications 2010-02 page 24

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 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. This person may be absent for periods not exceeding one hour; 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 in core;

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

4. restart following any scram or other unplanned shutdown, or abnormal occurrence.

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.

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

1. Meeting frequently (at least annually), with provision for additional meetings when circumstances warrant to assure safety at the facility.

2. A quorum shall consist of not less than a majority of the members and shall include the chairperson UCI Technical Specifications 2010-02 page 25

or his/her designee. Votes shall not be taken where a majority of those voting would be directly associated with facility operations.

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

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

6.2.3 ROC Review function The following review functions constitute part of the charter of the ROC.

1. Review and approval of all proposed changes to the facility, its license, procedures, 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 for program adequacy to assure safety;

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; 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 or RSC.

UCI Technical Specifications 2010-02 page 26

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.

6.4 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.

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 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).

UCI Technical Specifications 2010-02 page 27

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:

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

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

c. 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.

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 0 th. 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; UCI Technical Specifications 2010-02 page 28

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;

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.

UCI Technical Specifications 2010-02 page 29

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.

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 30

ATTACHMENT C.

SAR CHAPTER 5. REACTOR COOLANT SYSTEM

5.1 INTRODUCTION

TRIGA reactors up to 1 megawatt power level and beyond are designed, and have operated successfully over many reactor years, with convection cooling. Sufficient water and flow channels exist within the core structure because of the grid plate and fuel element design to permit this. In the case of standard low density fuel such as in UCINRF, there is no concern about cladding temperatures overheating or approaching DNB limits, so no discussion of such is warranted at this facility. Cooling of the primary water is provided in order to enable operations over long periods without placing undue stress on the pool liner or other reactor components. Thus the reactor may be safely operated with no cooling available. Cooling is actuated on an as needed basis at this facility.

Administrative controls are used to limit the pool temperature to below 30'C and in practice operation has never continued above 25 0 C. Maintaining close limits should extend the life of the tank and other components and gives additional assurance that reactor parameters, such as measured neutron fluxes, do not vary as a result of expansion of detector support structures, etc.

To minimize the effects of corrosion and to keep water radioactivity contamination as low as possible, a water purification system is provided. This system operates on a 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day basis.

Contaminants are removed onto removable filters and/or mixed bed ion exchange resin, which are disposed through low level waste disposal procedures on an as needed basis.

5.2 WATER COOLING SYSTEM A diagram of the water cooling circulation is shown in Fig. 5-1. Brief specifications for the components are given below.

Pump Centrifugal, 7.5 hp, close coupled, self priming, stainless steel body and impeller and mechanical shaft seal.

Heat exchanger (Fig. 5-2) All parts that are in contact with the demineralised water are Type 304 stainless steel. The exchanger is designed for a capacity of 880,000 Btulhr. The shell is supplied at < 50°F and

> 50 psi by a chilled water system flowing from UCI Central Plant in black iron piping and heavily chemically treated to reduce corrosion. Pool conductivity is monitored frequently to detect leaks into the pool. The 4 inch pipe connections, bypass, valves and control system are shown in Fig. 5-3.

Controller Watlow, Series 96. Input from platinum resistance sensor (RTD) stainless steel clad in pool. The control is normally set to maintain pool temperature at 20.0 +/- 1.0 °C. Gate valve controls on outflow. (Fig 5-3) 5-1 UCINRF: Reactor Design 1.1, 2009 SAR Rev 1.1, 2009 5-1 UCINRF: Reactor Design

5.3 WATER PURIFICATION SYSTEM Water clarity and purity is maintained by constant circulation through a filter and ion exchange resin.

A pool skimmer helps to remove surface contamination. Purity is assessed by conductivity measurements taken during daily start-up exercises. New resin is installed when the conductivity consistently approaches 2 micromhos/cm. Make up water is added manually, when needed, from the Rowland Hall deionized water supply system. A conductivity probe is provided so make-up water can be checked before a substantial amount is added to the pool. Components are described briefly below.

Pump Centrifugal, self priming type, 1.5 hp, close coupled, plastic impeller and housing, ceramic mechanical shaft seal.

Filters Three replaceable fiber cartridges 25 micron rating.

Pressure gauges Before and after filter gauges measure the drop across the filter as an aid in determining the extent of filter clogging. Pressure gauges are also located at the entry to each demineralizer bed to provide indication of possible clogging.

Demineralizer 2 3 cubic foot tanks, for a total of 6 cf, in a parallel flow system.

Each contains mixed cation and anion resin (initially in H and OH forms) for ion removal and stabilization of pH.

Conductivity meter. Measures the conductivity up and downstream of the demineralizer as a test of its efficiency. Switched sensor also in make-up water delivery line. Temperature compensation adjustment.

Flow meter Range 0 to 30 gallons per minute. Normally 10-20 gpm.

Water surface skimmer Collects foreign particles that float on the surface of main pool.

Water at the surface flows over the top,on the floating portion of the skimmer. Cleaned manually when clogged.

5-2 UCINRF: Reactor Design SAR Rev 1. 1, 2009 Rev 1.1, 2009 5-2 UCINRF: Reactor Design

I FLOW DIAGRAM !--- WATER COOLING  !

AND PURIFICATION SYSTEMS Fig 5-1 Reactor Water Purification and Cooling Systems Schematic SAR Rev 1. 1, 2009 5-3 UCINRF: Reactor Design

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CHILLED WATER CONNECTION TO REACTOR HEAT EXCHANGER POOL Fig 5-3 Heat Exchanger Connections and Controls 5-5 UCINRF: Reactor Design SAR Rev 1.1, 2009

ATTACHMENT D.

SAR 7. REACTOR INSTRUMENTATION AND CONTROL SYSTEMS 7.1 CONTROL CONSOLE AND NEUTRON DETECTORS All the functions essential to the operation of the reactor are controlled by the operator from a desk-type control console in the control room (Fig 7-1) with an auxiliary cabinet (Fig 7-2). The radiation monitors in the cabinet are described elsewhere (chapter x), and the rack also houses part of the security monitoring system. This instrumentation is connected to the control rod drives, the facility interlock system, and various detectors positioned around the reactor core. Three commercial analog neutronics modules (with digital readouts) manufactured by Gamma-metrics, Inc., are incorporated:

(1) Wide Range Linear Monitor (fission chamber signal input). Fig 7-5.

(2) Wide Range Monitor (CIC signal input). Fig 7-6.

(3) Power Range Monitor (UCIC signal input). Fig 7-7.

These channels cover the power ranges indicated in Fig. 7-3.

The individual instruction manuals maintained in the control room should be consulted for details on function and maintenance of all instrumentation in the control systems.

The adjacent cabinet contains control and readout for the remote area monitor and other alarm indicator lights.

7.1.1 Control Console The meters, switches, and recorder used to operate the reactor are mounted in the console as follows (Figs 7-1, 7-4, 7-5, 7-6, 7-7 and 7-8):

In the CENTER panel (Fig 7-4) on the console bench:

1. A MAGNET POWER key switch with OFF: OPERATE: RESET positions;

2. A console power (POWER ON) switch;

3. A three-position OPERATE key switch;

4. Five sets of three control-rod adjustment switches (UP and DOWN) and indicator lights (ON and CONT) for the shim(SHIM), regulating (REG), and adjustable transient (ATR) control rods (two unused sets are provided);

5. A manual SCRAM bar operating the five sets of switches simultaneously;

6. One TRANSIENT ROD FIRE switch; In the CENTER vertical panel:

7. Three LCD ROD POSITION indicators, one each for REG, SHIM, and ATR positions (two slots are vacant, unused);

8. A MODE SELECTOR switch with AUTOMATIC: STEADY STATE: STEADY STATE: PULSE positions; SAR Rev 1. 1, It /09 7-1 UCfNRF: Instrumentation and Control Syaterns

9. A dual-pen strip-chart recorder with one pen (red) recording LOG power (WRM output),

the other (blue) recording LINEAR power (WRLM output);

10. A %-DEMAND control for servo control setting;

11. Annunciators on the LEFT for SCRAMS (7), SOURCE (interlock), and one unused.

12. Annunciators on the RIGHT for COOLING, PURIFICATION, and LAZY SUSAN systems

13. On the RIGHT: ARM switches and indicators (ARM and UP) for the ATR (adjustable transient rod) and FTR (fast transient rod) air solenoid drives.

In the LEFT PANEL:

14. A WIDE RANGE LINEAR (WRLM) Monitor (Fig 7-5) with associated meter, range indicator, MANUAL/AUTO/250 kW mode switch, and REMOTE/LOCAL switch. In the adjacent rack is the main electronics and the test/calibrate controls for this channel. When in REMOTE, the readouts are on the control console face.

15. A WIDE RANGE (WRM) Monitor (Fig 7-6) with digital LEVEL PERCENT and PERIOD SECONDS indicators, bar indicators for these and trip settings, test and calibrate controls.

In the RIGHT PANEL:

16. A POWER RANGE Monitor (Fig 7-7) with digital LEVEL PERCENT, PEAK POWER (MW) and ENERGY (MW-sec) indicators, bar indicators for these and trip settings, and test and calibrate controls.

17. A FUEL TEMP module (General Atomics NT1000) with meter (degrees Celsius) and scram function with calibrate and test modes;

18. A digital POOL WATER TEMPERATURE meter indicating in degrees Celsius;

19. MAGNET current supplies and indicator modules (not installed)

Additional details of functions of these control and indicating devices are as follows:

The console POWER ON switch controls primary ac power to all circuits except the +/-25 vdc power supplies, and monitor power supplies (including detector HV supplies). The power supplies, which are left on even though the console is not in operation, are controlled by a circuit breaker on the rear center door.

Current to the magnet power supply is controlled through the three-position OPERATE key switch.

The control-rod adjustment switches (Fig 7-8), UP, DOWN, and combination CONT/ON (contact/on) switches are provided for the shim and regulating rods. UP and DOWN switches are also provided for the adjustable transient rod cylinder. In addition to rod-adjustment switches, a FIRE switch and ARM switches are also provided for the two transient rods.

The %-DEMAND control is to be used in conjunction with a steady-state automatic flux controller (not implemented).

SAR Rev 1. 1, 11/09 7-2 UCINRF: Instrumentation and Control Syatems

20.

FIG 7-1 UCI Nuclear Reactor Facility Control Console.

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, t~jiTLEL FIG 7-2 UCI Nuclear Reactor Facility Auxiliary Control Room Instrumentation Rack.

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1010 2 X 109 W 108

-UNCOMPENSATED ION CHAMBER PEAK POWER (PULSE POWER)

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_j LU od COMPENSATED ION CHAMBER

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- WIDE RANGE LINEAR MONITOR LINEAR RECORDER (STEADY-STATE OPERATION).

inI FISSION COUNTER WIDE RANGE MONITOR LOG RECORDER (STEADY-STATE OPERATION) 10 0

. . SOURCE

. LEVEL 10- 1 m l FIG 7-3 Neutron Detector Power Level Ranges.

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3/ABl* Rev i.l , 1L/09 7-4 UCINRF: Instrumentation and Control Syatems

FIG 7-4 Control Console Center Panel.

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FIG 7-5 Wide Range Linear Monitor (Console Remote Panel).

SAR Rev 1.1, 11 /09 7-5 UCINRF: Instrumentation and Control Syatems

FIG 7-6 Wide Range Monitor.

FIG 7-7 Power Range Monitor.

SAR Rev 1.1, 11/09 7-6 UCINRF: Instrumentation and Control Syatems

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O OWN FIG 7-8 Control Desk Panel 7.2 CONTROL AND SAFETY INSTRUMENTATION Indications of control-rod positions, power levels, fuel temperatures, and rate of power changes are necessary in the operation of the reactor. The individual control and safety channels are described briefly in the following paragraphs.

7.2.1 Control-rod Drive Switches and Circuits (Fig 7-8)

The control-rod circuit consists of the switches and indicating devices used in operating the two standard (rack-and-pinion) control-rod drives. The six illuminated pushbutton switches in the circuit are arranged in the center of the control panel in two vertical rows, one row for each control rod. Each row of switches contains a white DOWN switch, a red UP switch, and a blue and yellow double-pushbutton CONT/ON (contact/on) switch.

Illumination of the various switches signifies the following conditions:

DOWN switch: The DOWN light indicates that the control-rod and drive are at their lower limits.

UP switch: The UP light indicates that the control rod drive is at its upper limit.

CONT/ON switch: The CONT side of the double pushbutton switch indicates contact between the control-rod assembly armature and the control-rod drive electromagnet The ON side -f the switch indicates that the electromagnet power is on The magnet power circuit is energized by a key switch located on the left side of the control panel.

When the double CONT/ON pushbuttons are depressed, magnet current is interrupted and the ON SAR Rev 1. 1, 11/09 7-7 UCINRF: Instrumentation and Control Syatems

lights are extinguished. If a control rod is above the down limit, the rod falls back into the core and the CONT light is extinguished until the automatic magnet-drive-down control lowers the magnet and contact is again made with the rod. Releasing the button closes the magnet circuit and magnet current is restored.

7.2.2. Adjustable Transient Rod Drive Switches and Circuits.

The adjustable transient rod may be used as a steady-state control rod as well as a transient rod.

The pneumatic cylinder can be driven up or down by depressing the appropriate switch on the console. (Fig 7-8) The associated position indicator reads out the position of the pneumatic cylinder. The UP and DOWN pushbuttons become illuminated when the cylinder reaches the respective limits of travel.

To fire the adjustable transient rod the FIRE button must first be armed by depressing the adjustable transient rod ARM switch to the right of the recorder. (Fig7-4) This completes the circuit between the FIRE button (Fig 7-8) and the solenoid-operated air valve. One-half of this ARM button will become illuminated when the FIRE button is armed; the other half is illuminated whenever the control rod is not in the full down position. (After the FIRE button has been armed and pushed.) The control rod may be reinserted into the core by pushing the ARM button a second time.

For steady-state operation, the pneumatic cylinder is driven to its full down position and air pressure is applied under the piston. Thus, when the pneumatic cylinder is driven up, the control rod is also raised. The system is interlocked so that in the steady-state mode, air cannot be applied until the cylinder is in the full down position.

For pulsing operation the cylinder is adjusted to the position required for a particular reactivity insertion and the control rod is ejected by applying air pressure under the piston. In the PULSE mode the solenoid-operated air valve will automatically return to the vent position, permitting the control rod to drop back into the core by gravity. The timing of the automatic return is adjustable up to 10 sec.

7.2.3 Fast Transient Rod Switches and Circuitry The fast transient rod may be operated as a safety rod as well as a transient rod. In either case, the drive must be armed in the same fashion as the adjustable transient drive only using the fast transient rod ARM switch.(Fig 7-4) To use the fast transient rod as a safety rod during steady-state or square wave operation, the adjustable transient rod and all steady-state control rods must first be in the full down position. Then the fast transient rod can be withdrawn by pushing the ARM button (Fig 7-4) and then the FIRE button (Fig7-8). The fast transient rod may be reinserted (without effecting the position of the other rods) by pushing its ARM button a second time.

For pulsing when both transient rods are to be used, the rods are individually armed by depressing their respective ARM buttons. They are then both fired together upon depressing the FIRE button.

The reactivity insertion of the adjustable transient rod takes place in about 100 milliseconds (msec.), whereas the reactivity insertion of the fast transient rod takes place in about 50 msec. To prevent pulse clipping, the end of the reactivity insertions of both rods should occur at the same time. Coordination of the two transient rods is provided by means of a delay circuit. This delay circuit is adjustable between 20 msec. and 100 msec.

A magnetic sensor is provided on each transient rod drive at the end of the reactivity insertion.

The signal generated by these sensors, when the drives are fired, can be displayed on a scope SAR Rev 1.1, 11 /09 7-8 UC1NRF: Instrumentation and Control Syatems

during set-up operations. The delay circuit can then be adjusted so that the displayed signals occur at the same time interval after depressing the FIRE button.

7.2.4 Wide Range Monitor (WRM) and Period Channel The WRM receives its input from a fission detector through a special low noise cable into a low-noise preamplifier. The channel converts the signal to logarithmic information over a range of 10-8 to 200 percent of full power. An adjustable trip level is provided to assure sufficient neutrons are present for start-up. This is set at 7 x 10-7 percent, corresponding to about 10 nv. The trip level can be read on a bar graph. A digital (E-format) readout is provided as well as an LCD bar graph indicator over the full range. An output is directed to the console recorder (red pen) to provide a continuous record of Log Power.

The WRM also provides a digital period readout (-199.9 seconds to + 3 seconds) together with a bar graph. An adjustable period trip level is also provided and displayed by bar graph (set at > 3 seconds).

Test and calibrate modes are provided for each output. A NON-OPER trip is initiated during calibration or if detector High Voltage is disabled. Detector High Voltage may be read at an analog test point (0-1.000 volts = 1-1000 volts) in the front panel.

7.2.5 Wide Range Linear Power Channel The Wide Range Linear Monitor (WRLM) is coupled to a compensated ion chamber. The electronic chassis is installed (because of physical limitations in the console) in a cabinet adjacent to the main reactor console. In order to provide clear indications to the reactor operator, the power level displays from this unit are duplicated in the left facing control console panel. This panel is activated by use of a REMOTE/LOCAL switch, normally left in REMOTE. Test and calibration controls are provided on the main unit front panel as are digital readouts of the high voltage and compensation voltage applied to the detector.

During steady-state operation, the WRLM provides linear power level indications from below source level to full power. In AUTO mode the unit switches ranges automatically up at about 80%

of range and down at about 20% of range (except in the top range). A MANUAL mode can be selected in which the operator selects the range. A 250kW mode can be selected if it is desired to use the channel as a percent of full power channel. A lighted mode indicator is provided on the console. A servo output is provided for automatic reactor flux control, but is not presently in use/connected.

An adjustable trip is provided on all ranges to initiate reactor scram if the linear power is allowed to rise beyond a percentage (below 110%) of scale reading. A NON-OPER trip is actuated if the high voltage fails, or the unit is being calibrated. An additional bistable trip is provided that can be adjusted over the entire range. This is used for a "PULSE" interlock which prevents activating the pulse mode circuitry for the reactor if the measured linear power level is too high. This is set to trip at less than 1 kilowatt.

An output signal from the WRLM is used to drive the blue pen of the recorder to provide continuous record of (linear) power level. No range information is transmitted to the recorder.

SAR Rev 1. 1, It /09 7-9 UCINRF: Instrumentation and Control Syatems

7.2.6 Power Range Channel The Power Range Monitor (PRM) is coupled to an uncompensated ion chamber The current from the chamber can be measured from 1.56 x 10-6 amps to 1 x 10-3 amps and provides indication from 0 to up to 125% power on a digital indicator and a bar graph. An adjustable trip is set below 110%

and is also indicated on a bar graph. In the pulse mode, the circuit can accumulate nv over time, determine the maximum (peak power) level (with four operator-selected ranges: 2000, 1000, 500 and 200 MW), and store a value for accumulated (nvt) MW-seconds at 2 secs and/or 10 seconds into the pulse with a 50 MW-sec upper range. Associated calibration and testing circuits are included.

A NONOPER trip actuates if high voltage failure occurs. The detector high voltage can be measured at an analog front panel test point (0-1.000 volts = 1-1000 volts).

7.2.7 Fuel Temperature Thermocouples and Meters A fuel thermocouple is connected to the FUEL TEMPERATURE MODULE mounted at the front of the control panel. This meter indicates fuel temperature during all modes of reactor operation; a scram circuit is associated with this circuit to scram the reactor on high fuel temperature. Calibrate and scram set point test circuits (which place simulated thermocouple signals into the meter circuit) are provided.

7.2.8 Water Temperature Monitoring Channels Reactor pool water temperature is monitored in the reactor tank by means of a stainless steel clad thermistor probe connected to a standard commercial temperature digital readout in degrees Celsius. This value is constantly displayed. The probe measures the water temperature at about 2 feet below the water surface.

An additional readout is provided on the secondary cooling water system controller in the reactor room. This is a platinum resistance probe measuring pool temperature a few inches below the water surface but at the side of the pool near the water cooling and purification piping.

7.3 INTEGRATION OF CONTROL AND SAFETY CIRCUITRY During reactor operation, the functions of the individual control and safety channels are intimately related. In this section, each allowed mode of reactor operation is briefly described, and operator interaction and safety concerns addressed. Block diagrams illustrating the integration of the reactor control and safety channels with regard to specific operating modes are also provided.

7.3.1. Steady-State Operation - Manual and Automatic (Fig 7-9)

Manual steady-state reactor control is used for reactor operations from source levels to power levels up to 250 kw (see Fig. 7-4, center panel switch). This mode is used for reactor startup, change of power level, and steady-state operation. Three power level channels described above provide indication at all levels. Two recorder traces are available indicating log and linear power.

The linear power range is indicated by a digital light display on the Wide Range Linear Monitor console display panel. Trips are enabled on all three channels to scram the reactor if excessive power is indicated on any one. Period indication is available which can alert the operator to SAR Rev 1.12 11/09 7-10 UCINRF: Instrumentation and Control Syatems

unanticipated rapid changes in flux.

Automatic control is not currently implemented. It would adjust the reactor power at any steady condition from 1 w to 250 kw. Automatic control involves the use of a flux comparison signal from the WRM to adjust the position of the regulating rod.

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The uncompensated ion chamber operating with the Power Range Monitor is used. An interlock is provided that prevents firing of the transient rods if the reactor power is above 1 kw.

A block diagram illustrating the integration of the control and safety circuitry for transient operation is shown in Fig. 7-10.

Fuel temperature continues to be monitored in this mode and a fuel temperature scram is obtained if the fuel exceeds a preset (steady state) temperature on the meter.

SAR Rev 1.1, 11/09 7-11 UCINRF: Instrumentation and Control Syatems

FUEL ELEMN co. 4 LECOM*

we I UNCOMPENSATED ION CHAP TE N I KW INTERLOCK PEAK DETECT, NVT Fig 7-10 Instrumentation Arranged for Pulse Mode Operation.

1. 1, 11/09 Rev 1.1, SAR Rev 11 /09 7-12 UCINRF: Instrumentation and Control Syatems 7-12 UCINRF: Instrumentation and Control Syatems

7.4 SAFETY DEVICES (

SUMMARY

)

7.4.1 Scrams

1. Wide Range Monitor channel, fission chamber, <110% of full power.

2. Wide Range Linear Monitor, compensated ion chamber, <110% of full power.

3. Power Range Monitor, uncompensated ion chamber, <110% of full power.

4. Manual SCRAM bar.

5. Detector high voltage supply failure, provided on each channel.

6. Console power failure.

7. Seismic switch, set approx MM V motion.

7.4.2 Interlocks

1. To assure minimum source strength before control rods can be withdrawn.

2. To prevent withdrawal of two control rods simultaneously.

3. To ensure that pulsing cannot occur with reactor power greater than 1 kw.

4. To prevent application of air to the fast transient rod in the steady state mode unless all other rods are fully inserted.

5. To prevent application of air to the adjustable transient rod in the steady state mode unless the cylinder is in the 'down' position.

6. To prevent movement of any control rod except transient rods in pulsing mode.

7-13 UCIINRF: Instrumentation and Control Syatems 1.1, 11/09 SAR Rev 1.1, SARRev 1] /09 7-13 UCINRF: Instrumentation and Control Syatems

ATTACHMENT E.

Ventilation System Diagrams omitted from Chapter 3. (Figures 3-3, 3-5, 3-6)

Note: In error, there was no diagram 3-4 referenced.

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ATTACHMENT F Exposure Estimates From Maximum Hypothetical Accident The computations submitted were those performed in 1968 to represent a worst possible scenario. Some were for other facilities to provide some realistic estimates. They should be done for this facility using modem computer codes to provide more realistic estimates of source terms, dispersions, and exposures for future revision.

The estimates computed were done based on the highly conservative assumption for the source term of infinite irradiation time prior to accident. Since operating hours are rarely in excess of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per week, the FP inventory prior to a hypothesized release will realistically be significantly lower. The other assumption made at most facilities is that element cladding is completely removed to expose the entire fuel meat. Actual incidents so far at TRIGA reactors worldwide have only involved pin-hole leaks vastly below such MHA scenario.

For exposure to personnel within the facility, where all iodine isotopes released became airborne (element cladding removal in air) no credit was taken for ventilation and a dose rate was computed for instantaneous dose assuming all iodine radioisotopes were contributing and concentrated in the thyroid. Thus an instantaneous rate at time of release was 0.45 rads/second.

This would mean an exposure of only 2 minutes would exceed the annual limit of 50 Rads. So immediate evacuation would be essential. Exposure from noble gases would be minimal in such event.

No further computations were submitted as to actual TEDE following continued presence in the facility, so no actual time of exposure was assumed beyond the 2 minutes suggested.

For exposures beyond the facility, similar assumptions were made with no credit for absorption by the installed HEPA filter for the unlikely scenario of a person on the roof, breathing the average air with some dilution from other exhausts included. Both infinite time and 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> exposures were assumed. No allowance was made for isotope decay, but lifetime commitments were not included. Comparisons were then introduced from other reactor facilities to show the level of predicted exposures that would not be exceeded at this facility. No specific atmospheric dispersion calculations have been carried out for this facility. Normal operation estimates assumed the worst case scenario of an un-dispersed plume, as this gave a result showing little concern.

As a footnote, since 1311 is the worst offender for the thyroid, using a value of 1.5 x 10-3 Ci released of 131, (infinite fuel irradiation inventory from 1 element in Table 13-5) or 5.5 x 107 Bq, and the factor in ICRP 123 of 1.47 x 10- Sv/Bq for inhalation gives 0.8 Sv (80 Rem) EDE, in excess of the 50 Rem limit. However this unrealistically assumes all the FP released from the element is completely inhaled by one person, but does represent an upper limit.

ATTACHMENT G Qualitative Seismic Considerations for Emergency Shutdown UCI TRIGA Reactor is a Mark I TRIGA installation. The core is contained in a below ground-level tank containing 23,000 gallons of water. As described in the SAR the tank is reinforced concrete with an aluminum liner to make a pool. The reactor core structure is bolted to the aluminum floor of the pool. The four control rods are held vertically by straight aluminum rods extending through a steel bridge to the motor or piston drives. The rods are held up by either magnets or air pistons that will release on reactor scram. Rods are tested to fall normally within 2 seconds, even though water damping is occurring.

The fuel follower rods (REG and SHIM) simply hang through holes in the upper and lower grid plates with no additional guidance. The ATR has a short guide tube in which it slides. FTR is completely shrouded by a heavy aluminum pipe extending to the top of a guide tube within which the FTR slides freely. A strong "safety plate" welded into the core structure precludes any rods falling though the core.

The bridge across the tank is constructed of heavy steel girders supporting solid steel plates across the entire width of the tank. The bridge is fastened securely at its ends to the concrete surround and hence to the concrete tank. It covers much of the central part of the core structure and all of the area in which control rods are located.

In a seismic event, it is anticipated that the tank, the fastened core structure and the steel bridge are unlikely to experience differential forces - in other words they will move together as a result of the reinforced concrete shell surrounding the pool liner to which they are all secured. Thus ir t is extremely unlikely that the rods could not fall to effect shut-down. Since either of the fuel follower rods is worth more than normal core excess, it would only take one of these to fall to effect non-criticality, with additional safety provided by either of the transient rods falling.

It is also anticipated that the heavy steel bridge is sufficiently robust to withstand any falling "debris" should the ceiling or other overhead structures partly collapse. The control rod area is thus unlikely to be blocked by material that would impede rod drop.

In support of the above scenario, some local seismic experts have suggested that thrust fault movement is most likely in this immediate area, with lateral movement on the San Andreas and other Southern California faults, at a considerable distance. The reactor would "scram" even more readily following an "up-down" motion of the tank and its structures. In the past 40 years, we have experienced a few small seismic events of such a vertical nature.