ML15348A386

Response to NRC Request for Additional Information Regarding License Amendment Request from the University of California-Davis Mcclellan Nuclear Research Center Per the Letter Dated June 3, 2015

ML15348A386
Person / Time
Site:	University of California-Davis
Issue date:	10/29/2015
From:	Steingass W G McClellan Nuclear Research Center
To:	Tran L N Division of Policy and Rulemaking
References
Download: ML15348A386 (234)

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{{#Wiki_filter:UCDAVISMNRCMcCLELLAN NUCLEAR RESEARCH CENTERU.S. Nuclear Regulatory Commission Attn: Linh N. Tran, Senior Project Manager, NRRMail Stop: 012 D20One White Flint North11555 Rockville PikeRockville, MD 208525335 PRICE AVENUEBUILDING 258McCLELLAN, CA 95652PHONE: (916) 614-6200FAX: (916) 614-6250WEB: http://mn rc.ucdavis.edu October 29, 2015RE: NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUESTFROM THE UNIVERSITY OF CALIFORNIA-DAVIS McCLELLAN NUCLEAR RESEARCH CENTER PER THELETTER DATED JUNE 3, 2015.

Dear Ms. Tran,

In response to your letter dated June 3, 2015, we are submitting the requested documentation per saidletter under Oath and Affirmation. Additionally, we are provided said documentation electronically on a DVD for your convenience. I verify under penalty of perjury that the foregoing is true and correct.Executed on October 29, 2015.Assoca'e Director of Operations Reactor Supervisor McClellan Nuclear Research CenterUniversity of California-Davis Facility Operating License No. R-130.C: B. Klein, UCD/MNRC P:- oNUCLEAR REGULATORY COMMISSIONWASHINGTON, 0.0. zt&o5S5O1Q FACILITY OPERATING LICENSEDOCKET NO. 50-607DEPARTMENT OF THE AIR FORCE AT McCLELLAN AIR FORCE BASELicense No. R-1301. The U.S. Nuclear Regulatory Commission (the Commission) has found that:A." The application for license, filed by the Department of the Air Force atMcClellan Air Force Base, on October 23, 1 996, as supplemented, complies with the standards and requirements of the Atomic Energy Actof 1!954, as =amended (the Act), and the Commission's rules andregulations as set forth in 10 CFR Chapter I;B. Construction. of the facility was completed in substantial conformity withthe provision's of the Act, and the rules and regulations of theCommission; C. The facility Will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;

0. There is reasonable assurance (i) that the activities authorized by thislicense can be conducted

,without endangering the health and safety ofthe public and (ii) that such activities will be conducted in compliance with the Commission's regulations; E. The licensee is technically and financially qualified to engage in theactivities authorized by this operating license in accordance with theregulations of t/he Commission; ..F. The licensee is a Federal agency and will use the facility for defenseprograms and research. The licensee, in accordance with10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements," iis not required to furnish proof of financial protection. The licensee has executed an indemnity agreement that satisfies therequirements o~f 10 CFR Part 140 of the Commission's regulations; G. The issuance of this license will not be inimical to the common defenseand security or to the health and safety of the public;H. The issuance of this license is in accordance with 10 CFR Part 51 of theCommissioh's regulations and all applicable requirements have beensatisfied; andI.The receipt:, possession, and use of the byproduct and special nuclearmaterials as authorized by this license will be in accordance with theCommissioa 's regulations in 10 CFR Parts 30 and 70, including Sections30.33, 70.23, and 70.31.2. Facility Operating License No. R-130 is hereby issued to the Department ofthe Air Force at McClellan Air Force Base as follows:A. The license applies to the training reactor and isotopes production, General Atornics (TRIGA) nuclear reactor (the facility) owned by theDepartment 'of the Air Force at McClellan Air Force Base (the licensee). The facility is located on the licensee's site at McClellan Air Force Easeand is described in the licensee's application for license of October 23,1 996, as supplemented. B. Subject to the conditions and requirements incorporated herein, theCommission ~hereby licenses the Department of the Air Force atMcClellan Air Force Base:(1) Pursuant to Section 104c of the Act and 10 CFR Part 50,"Domestic Licensing of Production and Utilization Facilities," topossess, use, and operate the facility at the designated locationat McClellan Air Force Base, in accordance with the procedures and limitations set forth in this license.(2) Pursuant to the Act and 10 CFR Part 70, "Domestic Licensing of Special Nuclear Material," to receive,

possess, and use up to21 .0: kilograms of contained uranium-235 enriched to less than20 percent in the isotope uranium-235 in the form of reactorfuel;:up to 4 grams of contained uranium-235 of any enrichment in the form of fission chambers; up to 16.1 kilograms ofcontained uranium-235 enriched to less than 20 pecenR[[t in heisotope uranium-235 in the form of plates; and to possess, butnot separate, such special nuclear material as may be producedby the operation of the facility.

3(3) Pursuant to the Act and 1 0 CFR Part 30, "Rules of GeneralApplicability to Domestic Licensing of Byproduct Material," toreceive,

possess, and use a 4-curie sealed americium-beryllium neutron source in connection with operation of the facility; a55-millicurie sealed cesium-1 37 source for instrument calibrations; small instrument calibration and check sources ofless than 0.1 millicurie each; and to possess, use, but notseparate, except for byproduct material produced in reactorexperiments, such byproduct material as may be produced bythe operation of the facility.

C. This license shall be deemed to contain and is subject to the Conditions specified in lParts 20, 30, 50, 51, 55, 70, and 73 of 10 CFR Chapter I;to all applicable provisions of the Act; and to the rules, regulations, andorders of the Commission now or hereafter in effect and to the additional conditions specified below:(1) Maximum Power LevelThe licensee is authorized to operate the facility at steady-state power levels not in excess of 2300 kilowatts (thermal) and inthe pulse mode with reactivity insertions not to exceed $ 1.75(1.23 %Ak/k).(2) Technical Specifications The Technical Specifications contained in Appendix A arehereby incorporated in the license. The licensee shall operatethe !facility in accordance with the Technical Specifications. (3) Physical Security PlanThe licensee shall fully implement and maintain in effect aI.provisions of the Commission-approved physical security plan,including all amendments and revisions made pursuant to theauthority of 10 CFR 50.90 and 1 0 CFR 50.54(p). The approvedplan;i which is exempt from public disclosure pursuant to theprovisions of 10 CFR 2.790, is entitled "Physical Security Planfor the McClellan Nuclear Radiation Center (MNRC) TRIGAReacitor Facility," Revision 3, dated August 1996.

40. This license is effective as of the date of issuance and shall expiretwenty (20) years from its date of issuance.

~FOR THE NUCLEAR REGULATORY COMMISSION Office of Nuclear Reactor Regulation

Enclosure:

Appendix A Technical Specifications Date of Issuance: August 13, 1998 ~UNITED STATES "~NUCLEAR REGULATORY COMMISSION

• .WASHINGTON 1 D.C. 20555-0001" 'December 9, 1998Brigadier General Michael P. Wiedemer, Commander Sacramento Air Logistics CenterSM-ALC/TI-15335 Price AvenueMcClellan AFB, California 95652-2504

SUBJECT:

ISSUANCE OF AMENDMENT NO. 1 TO FACILITY OPERATING LICENSE NO. R-1 30 -DEPARTMENT OF THE AIR FORCE AT McCLELLAN AIR FORCE BASE (TAC NO. MA3477)

Dear General Wiedemer:

The Commission has issued the enclosed Amendment No. 1 to Facility Operating LicenseNo. R-1 30 for the McClellan Nuclear Radiation Center (MNRC) TRIGA Research Reactor.The amendment consists of changes to the technical specifications (TSs) in response toyour submittal of November 18, .1998.The amendment clarifies TS 3.8.3, requirements on the quantity and type of radioactive material allowed in experiments such that experiment failure will not result in airborneradioactivity in the reactor room or the unrestricted area exceeding the applicable doselimits in 10 CFR Part 20.A copy of the safety evaluation supporting Amendment No. 1 is also enclosed. Sincerely, Warren J. Eresian, Project ManagerNon-Power Reactors and Decommissioning

• Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No. 50-607

Enclosures:

1. Amendment No. 12. Safety Evaluation cc w/enclosures:

See next page McClellan AFB TRIGA REACTORDcktN.067 cc:Dr. Wade J. RichardsSM-ALCITI-1 5335 Price Avenue, Bldg. 258McClellan AFB, California 95652-2504 Col. Robert CapellHQ AFMC/SGC4225 Logistics Avenue, Suite 23Wright-Patterson AFB, Ohio 45433-5762 Lt. Col. Catherine ZeringueHQ AFSC/SEW9570 Avenue G, Building 24499Kirtland AFB, New Mexico 87117-5670 Test, Research, and TrainingReactor Newsletter 202 Nuclear Sciences CenterUniversity of FloridaGainesville, Florida 32611 . STATES~NUCLEAR REGULATORY COMMISSION ' WASHINGTON, D.C. 20888-0001 DEPARTMENT OF THE AIR FORCE ATMcCLELLAN AIR FORCE BASEDOCKET NO. 50-607AMENDMENT TO FACILITY OPERATING LICENSEAmendment No. 1License No. R-1301. The U.S. Nuclear Regulatory Commission (the Commission) has found thatA. The application for an amendment to Facility Operating License No. R-1 30 filed bythe Department of the Air Force at McClellan Air Force Base (the licensee) onNovember 1 8, 1 998, conforms to the standards and requirements of the AtomicEnergy Act of 1954, as amended (the Act), and the regulations of theCommission as stated in Chapter I of Title 10 of the Code of Federal Regulations (10 CFR);B. The facility will operate in conformity with the application, the provisions of theAct, and the rules and regulations of the Commission; C. There is reasonable assurance that (I) the activities authorized by this amendment can be conducted without endangering the health and safety of the public and (ii)such activities will be conducted in compliance with the regulations of theCommission; D. The issuance of this amendment will not be inimical to the common defense andsecurity or the health and safety of the public;E. This amendment is issued in accordance with the regulations of the Commission as stated in 10 CFR Part 51, and all applicable requirements have been satisfied; andF. Prior notice of this amendment was not required by 10 CFR 2.105, andpublication of notice for this amendment is not required by 10 CFR 2.106.

22. Accordingly, the license is amended by changes to the Technical Specifications asindicated in the enclosure to this license amendment, and paragraph 2.C.(ii) of FacilityOperating License No. R-1 30 is hereby amended to read as follows:2.C.(ii)

Technical Specifications The Technical Specifications contained in Appendix A, as revised throughAmendment No. 1, are hereby incorporated in the license. The licensee shalloperate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of issuance.

FOR TH.E NUCLEAR REGULATORY COMMISSION Seymour H. Weiss, DirectorNon-Power Reactors and Decommissioning Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation

Enclosure:

Appendix A, Technical Specifications ChangesDate of Issuance: ENCLOSURE TO LICENSE AMENDMENT NO. 1FACILITY OPERATING LICENSE NO? R-1 30DOCKET NO. 50-607Replace the following pages of Appendix A, "Technical Specifications," with the enclosedpages. The revised pages are identified by amendment number and contain vertical linesindicating the areas of change.Remove Insert24 2425 25 .j° ,c. This specification is intended to prevent damage to vital equipment byrestricting the quantity of explosive materials within the 'r~actor tank (SAR Chapter 13,Section 13..2.6.2). .-d. The failure of an experiment involving the irradiation of 3 lbs TNTequivalent or less in any radiography bay external to the reactor tank will not result indamage to the reactor controls or the reactor tank. Safety Analyses have beenperformed (SAR Chapter 13, Section 13.2.6.2) which show that up to six (6) lbs of TNTequivalent can be safely irradiated in any radiography bay. Therefore, the three (3) lblimit gives a safety margin of two (2).3.8.3 Failure and Applicability. This specification applies to experiments installed in thereactor, in-tank experiment facilities, and radiography bays.Objective. The objective is to prevent damage to the reactor or significant releases of radioactive materials in the event of an experiment failure.Specifications.

a. All experiment materials which could off-gas,

sublime, volatilize, or produceaerosols under (1) normal operating conditions of the experiment or reactor, (2) credibleaccident conditions in the reactor, or (3) where the possibility exists that the failure ofan experiment could release radioactive gases or aerosols into the reactor building orinto the unrestricted area, the quantity and type of material in the experiment shall belimited such that the airborne radioactivity in the reactor room or the unrestricted areawill not result in exceeding the applicable dose limits in 10 CFR Part 20, assuming100% of the gases or aerosols escape.b. In calculations pursualtt to (a) above, the following assumptions shall beused:(1) If the effluent from an experiment facility exhausts through a stackwhich is closed on high radiation levels, at least 10% of the gaseous activity or aerosolsproduced will escape.(2) If the effluent from an 'experiment facility exhausts through a filterinstallation designed for greater than 99% efficiency for 0.3 micron and larger particles, at least 10% of these will escape.(3) For materials whose boiling point is above 130°F and where vaporsformed by boiling this material can escape only through an undistributed column ofwater above the core, at least 10% of these vapors 'can escape.24

c. If a capsule fails and releases material which could damage the reactorfuel or structure by corrosion or other means, an evaluation shall be made to.determine the need for corrective action. Insipection and any corrective action takenshall be reviewed by the Facility Director or his designated alternate and determined tobe satisfactory before operation of the reactor is resumed.Basis.a. This specification is intended to reduce the likelihood that airborneradioactivity in the reactor room or the unrestricted area will result in exceeding theapplicable dose limits in 10 CFR Part 20.b. These assumptions are used to evaluate the potential airborneradioactivity release due to an experiment failure (SAR Chapter 13, Section 13.2.6.2).

c. Normal operation of the reactor with damaged reactor fuel orstructural damage is prohibited to avoid release of fission products.

Potential damageto reactor fuel or structure must be brought to the attention of the Facility Director orhis designated alternate for review to assure safe operation of the reactor (SAR Chapter13, Section 13.2.6.2). 4.0 Surveillance Requ~irements: General. The surveillance frequencies denoted herein are based on continuing operation of the reactor. Surveillance activities scheduled to occur during an operating cycle which can not be performed with the re'actor operating may be deferred to the endof the operating cycle. If the reactor is not operated for a reasonable .time, a r'eactorsystem or measuring channel surveillance requirement may be waived during theassociated time period. Prior to reactor system or measuring channel operation, thesurveillance shall be performed for each reactor system or measuring channel for whichsurveillance was waived. A reactor system or measuring channel shall not beconsidered operable until it is successfully tested.4.1 Reactor Core Parameters. 4.1.1 Steady State Operation. Applicability. This specification applies to the surveillance requirement for the power level monitoring channels. Objective. The objective is to verify that the maximum power level of thereactor does not exceed the authorized limit.25

STATES~NUCLEAR REGULATORY COMMISSION ' WASHINGTON," O.C. 20865-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 1 TOFACILITY OPERATING LICENSE NO. R-130DEPARTMENT OF THE AIR FORCE ATMcCLELLAN AIR FORCE BASEDOCKET NO..POO-607

1.0 INTRODUCTION

By letter dated November 18, 1 998, the Department of the Air Force at McClellan AirForce Base (the licensee) submitted a request for amendment of the Technical Specifications (TSs), Appendix A, to Facility Operating License No. R-1 30 for the McClellan Nuclear Radiation Center TRIGA Research Reactor (MNRC). The requested amendment would clarify the quantity and type of material in experiments that could be released in theunlikely event of an experiment failure.2.0 EVALUATION The licensee has requested amendment of TS 3.8.3 concerning limitations on experiments. TS 3.8.3 and the bases of the TS currently read:Aoplicability. This specification applies to experiments installed in the reactor andits experimental facilities. Specifications.

a. All experiment materials which~could off-gas,

sublime, volatilize, orproduce aerosols under (1) normal operating conditions of theexperiment or reactor, (2) credible accident conditions in the reactor, or(3) where the possibility exists that the failure of an experiment couldrelease radioactive gases or aerosols into the reactor building or into theunrestricted area, the quantity and type of material to be irradiated shallbe limited such that the airborne concentration of radioactivity shall notexceed the applicable limits of 10 CFR Part 20 (at the operations boundary),

assuming 100% of the gases or aerosols escape."h. O °" 02Bases.a. This specification is intended to reduce the likelihood that airborneradioactivity in excess of the limits of 10 CFR Part 20 shall be releasedinto the reactor building or to the unrestricted area (SAR Section13.2.6.2). The licensee has proposed that the TS and bases be amended to read:Applicability. This specification applies to experiments installed in the reactor, in-tank experiment facilities, and radiography bays.Specifications.

a. All experiment materials which could off-gas,

sublime, volatilize, orproduce aerosols under (1) normal operating conditions of theexperiment or reactor, (2) credible accident conditions in the reactor, or(3) where the possibility exists that the failure of an experiment couldrelease radioactive gases or aerosols into the reactor building or into theunrestricted area, the quantity and type of material in the experiment shall be limited such that the airborne radioactivity in the reactor room orthe unrestricted area will not result in exceeding the applicable doselimits in 10 CFR Part 20, assuming 100% of the gases or aerosolsescape.Bases.a. This specification is intended to reduce the likelihood that airborneradioactivity in the reactor room or the unrestricted area will result inexceeding the applicable dose limits on 10 CFR 20.The licensee has proposed clarifying the TS by basing the TS on dose instead ofconcentrations of radioactive material.

The purpose of this TS is to limit doses to membersof the public and the MNRC staff to 10 CFR Part 20 limits in the unlikely event that an*experiment were to fail and release airborne radioactive material into the reactorconfinement and subsequently to the environment. Doses to members of the reactor staffand members of the public from accidents at research reactors are limited to the dosesgiven in 10 CFR Part 20 because 10 CFR Part 100 is not applicable to research reactors. The current TS is based on radioac~tivity concentrations. For occupational exposures theannual limit on intake (ALl) is the annual intake which would result in either a committed effective dose equivalent of 5 reins (stochastic ALl) or a committed dose equivalent of 50reins to an organ or tissue (non-stochastic ALl). The derived, air concentration (DAC)values in Table 1 of Appendix B to 10 CFR Part 20 are based on dividing the ALl by 2000working hours per year and is intended to control chronic occupational exposures. For non-occupational exposure (members of the public) the effluent concentrations given in Table 2 3.of Appendix B to 10 CFR Part 20 are equivalent to the radionuclide concentration which ifinhaled continually over the course of a year would produce a total effective doseequivalent of 0.05 rem. The licensee's proposed wording would be based on dose limitsdirectly. The licensee is concerned that the TS as currently written could be interpreted to limitreleases to the instantaneous concentration of airborne radioactive material in the reactorroom and unrestricted areas. This would ignore the time integral aspects of theconcentration limits given in 10 CFR Part 20 as discussed above. For a particular experiment failure event, it is possible to exceed the concentration limits in 10 CFR Part 20while the resulting dose would be a small fraction of the dose limits.The NRC staff notes that the proposed wording of the TS is more encompassing because aTS based on dose would also include consideration of radiation shine from a cloud ofradioactive material. This proposed change to the TSs is acceptable to the staff becausethe dose to members of the reactor staff and members of the public from the accidental failure of experiments will be within the limits given in 10 CFR Part 20 and because the*proposed wording clarifies the TS.3.0 ENVIRONMENTAL CONSIDERATION This amendment involves changes in the installation or use of a facility component locatedwithin the restricted area as defined in 10 CFR Part 20 or changes in inspection andsurveillance requirements. The staff has determined that this amendment involves nosignificant increase in the amounts, and no significant change in the types, of any effluents that may be released off site, and no significant increase in individual or cumulative occupational radiation exposure. Accordingly, this amendment meets the eligibility criteriafor categorical exclusioni set forth in 10 CFR 51.22(c)(9). Pursuant to 1OCFR 51.22(b), noenvironmental impact statement or environmental assessment need be prepared inconnection with the issuance of this amendment.

4.0 CONCLUSION

The staff has concluded, on the basis of the considerations discussed above, that(1) because the amendment does not involve a significant increase in the probability orconsequences of accidents previously evaluated,* or create the possibility of a new ordifferent kind of accident from any accident previously evaluated, and does not involve asignificant reduction in a margin of safety, the amendment does not involve a significant hazards consideration; (2) there is reasonable assurance that the health and safety of thepublic will not be endangered by the proposed activities; and (3) such activities will beconducted in compliance with the Commission's regulations and the issuance of thisamendment will not be inimical to the common defense and security or the health andsafety of the public.Principal Contributor: Warren J. EresianDate: December 9, 1998

STATESo NUCLEAR REGULATORY COMMISSION Z r~oWASHINGTON, D.C. 2055.5-0001 FACILI;TY OPERATING LICENSEDOCKET NO. 50-607DEPARTMENT OF THE AIR FO.RCE.AT McCLELLAN AIR FORCE BASELicense No. R-1 301. The U.S. Nuclear Regulatory Commission (the Commission) has found that:A. The application for license, filed by the Department of the Air Force atMcClellan Air Force Base, on October 23, 1 996, as supplemented, complies with the standards and requirements of the Atomic Energy Actof 1 954, as amended (the Act), and the Commission's rules andregulations as set forth in 10 CFR 'Chapter I;B. Construction of the facility was completed in substantial conformity withthe provisions of the Act, and the rules and regulations of theCommission; C. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; D. There is reasonable assurance (i) that the activities authorized by thislicense can be conducted without' endangering the health and safety ofthe public and (ii) that such activities will be conducted in compliance with the Commission's regulations; E. The licensee is technically and financially qualified to engage in theactivities authorized by this operating license in accordance with theregulations of the Commission;... F. The licensee is a Federal agency and will use the facility for defenseprograms and research. The licensee, in accordance with10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements," is not required to furnish proof of financial protection. The licensee has executed an indemnity agreement that satisfies therequirements of 10 CFR Part 140 of the Commission's regulations; 2G. The issuance of this license will not be inimical to the common defenseand security or to the health and safety of the public;H. The issuance of this license is in accordance with 10 CFR Part 51 of theCommission's regulations and all applicable requirements have beensatisfied; andc]p,1°o s s e-ssio n,-a +n 2. Facility Operating License No. R-130 is hereby issued to the Department ofthe Air Force at McClellan Air Force Base as follows:A. The license applies to the training reactor and isotopes production, General Atomics (TRIGA) nuclear reactor (the facility) owned by theDepartment of the Air Force at McClellan Air Force Base (the licensee). The facility is located on the licensee's site at McClellan Air Force Baseand is described in the licensee's application for license of October 23,1 996, as supplemented. B. Subject to the conditions and requirements incorporated herein, theCommission hereby licenses the Department of the Air Force atMcClellan Air Force Base:(1) Pursuant to Section 1 04c of the Act and 10 CFR Part 50,"Domestic Licensing of Production and Utilization Facilities," topossess, use, and operate the facility at the designated locationat McClellan Air Force Base, in accordance with the procedures and limitations set forth in this license.(2) Pursuant to the Act and 10 CFR Part 70, "Domestic Licensing of Special Nuclear Material," to receive,

possess, and use up to21 .0 kilograms of contained uranium-235 enriched to less than20 percent in the isotope uranium-235 in the form of reactorfuel; up to 4 grams of contained uranium-235 of any enrichment in the form of fission chambers; up to 16.1 kilograms ofcontained Uranium-235 enriched to less than 20 percent in theisotope uranium-235 in the form of plates; and. to possess, butnot separate, such special nuclear material as may be producedby the operation of the facility.

3 4-curie sealed americium-beryllium neutron source in connection with operation of the facility; a55-millicurie sealed cesium-i137 source for instrument calibrations; small instrument calibration and check 'sources ofless than 0.1 millicurie C. This license shall be deemed to contain and is subject to the conditions specified in Parts 20, 30, 50, 51, 55, 70, and 73 of 10 CFR Chapter I;to all applicable provisions of the Act; and to the rules, regulations, andorders of the Commission now or hereafter in effect and to.the additional conditions specified below:(1) Maximum Power LevelThe licensee is authorized to operate the facility at steady-state power levels not in excess of 2300 kilowatts (thermal) and inthe pulse mode with reactivity insertions not to exceed $1.75.(1.23 %Ak/k).(2) Technical Specifications The Technical Specifications contained in Appendix A arehereby incorporated in the license. The licensee shall operatethe facility in accordance with the Technical Specifications. (3) Physical Security PlanThe licensee shall fully implement and maintain in effectel.- provisions of the Commission-approved physical security plan,including all amendments and revisions made pursuant to theauthority of 10 CFR 50.90 and 10 CFR 50.54(p). The approvedplan, which is exempt from public disclosure pursuant to theprovisions of 1 0 CFR 2.790, is entitled "Physical Security Planfor the McClellan Nuclear Radiation Center (MNRC) TRIGAReactor Facility," Revision 3, dated August 1 996. S ..* .:..40. This license is effective as of the date of issuance and shall expiretwenty (20) years from its date of issuance. FOR THE NUCLEAR REGULATORY COMMISSION Office of Nuclear Reactor Regulation

Enclosure:

Appendix A Technical Specifications Date of Issuance: August 13, 1998 ~UNITEDOSTATES SNUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055-.0001 Mrh1, 1999Brigadier General Michael P. Wiedemer, Commander Sacramento Air Logistics CenterSM-ALC/TI-15335 Price AvenueMcClellan AFB, California 95652-2504

SUBJECT:

ISSUANCE OF AMENDMENT NO. 2 TO AMENDED FACILITY OPERATING LICENSENO. R-130 -DEPARTMENT OF THE AIR FORCE AT McCLELLAN AIRFORCE BASE (TAC NO. MA3477)

Dear General Wiedemer:

The Commission has issued enclosed Amendment No. 2 to Facility Operating LicenseNo. R-1 30 for the McClellan Nuclear Radiation Center (MN RC) TRIGA Research Reactor.The amendment consists of changes to the Technical Specifications (TSs) and SafetyAnalysis Report (SAR) to support expanded experimental facilities in response to yoursubmittal of January 11, 1999.The amendment provides for the installation of an Argon-41 Production Facility and aCentral Irradiation Facility. The installation of the Argon-41 Production Facility does notrequire any change or expansion of the TSs since an experiment failure will not result inairborne radioactivity in the reactor room or the unrestricted area exceeding the applicable dose limits already prescribed. The installation of the Central Irradiation Facility requires achange to TS 3.8.1 with regard to the maximum reactivity worth of a moveableexperiment. The change increases the reactivity limit of a moveable experiment in theCentral irradiation Facility to $1.75, corresponding to the pulse limit specified in TS 3.1.2.A copy of the safety evaluation supporting Amendment No. 2 is also enclosed. Si lcerely,Warren J. Iresian, Project ManagerNon-Power Reactors and Decommissioning Project Directorate Division of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation Docket No. 50-607

Enclosures:

1. Amendment No. 22. Safety Evaluation cc w/enclosures:

See next page McClellan AFB TRIGA REACTORDoktN.5-0 cc"Dr. Wade J. RichardsSM-ALC/TI-1 5335 Price Avenue, Bldg. 258McClellan AFB, California 95652-2504 Lt. Col. Marcia ThorntonHQ AFSC/SEW" 9570 Avenue G., Bldg. 24499Kirtland AFB, New Mexico 87117-5670 Col. Robert CapellHQ AFMC/SGC4225 Logistics Avenue, Suite 23Wright-Patterson AFB, Ohio 45433-5762 0* 0UNITED STATES.NucLEAR REGULATORY COMMISSIoN WHNToND.C. 208-o000DEPARTMENT OF THE AIR FORCE ATMc.CLELLAN AIR FORCE BASEDOCKET NO. 50-607AMENDMENT TO FACILITY OPERATING LICENSEAmendmentNo. 2License No. R-1 301.. The U.S. Nuclear Regulatory Commission (the Commission) has found thatA. The application for an amendment to Facility Operating License No. R-1 30 filedby the Department of the Air Force at McClellan Air Force Base (the licensee) onJanuary 11, 1999, conforms to the standards and requirements of the AtomicEnergy Act of 1954, as amended (the Act), and the regulations of theCommission as stated in Chapter I of Title 10 of the Code of Federal Regulations (10 CFR);B. The facility will operate in conformity with the application, the provisions of theAct, and the rules and regulations of the Commission;

• C. There is reasonable assurance that (i) the activities authorized by this amendmentc can be conducted without endangering the health and safety of the public and(ii) such activities will be conducted in compliance with the regulations of theCommission; D. The issuance of this amendment will not be inimical to the common defense andsecurity or to the health and safety of the public;E. This amendment is issued in accordance with the regulations of the Commission as stated in 10 CFR Part 51, and all applicable requirements have been satisfied; andF. Prior notice of this amendment was not required by 10 CFR 2.105, andpublication of notice for this amendment is not required by 10 CFR 2.106. 2. Accordingly, the license is amended by changes to the Safety Analysis Report andTechnical Specifications as indicated in the enclosure to this license amendment, andparagraph 2.C.(ii) of Facility Operating License No. R-130 is hereby amended to readas follows:2.C.(ii)

Technical Specifications The Technical Specifications contained in Appendix A, as revised throughAmendment No. 2, are hereby incorporated in the license. The licenseeshall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION ",i /1f Lt 'Seymour H. Weiss, DirectorNon-Power Reactors and Decommissioning Project Directorate Division of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation

Enclosure:

• Appendix A, Technical Specifications
• and Safety Analysis Report ChangesDate of Issuance:

March 1, 1999 .... 0 .ENCLOSURE TO LICENSE AMENDMENT NO. 2FACILITY OPERATING LICENSE NO. R-130DOCKET NO. 50-607A. Replace the following page of Appendix A, "Technical Specifications," with theenclosed page. The revised page is identified by amendment number and containsvertical lines indicating the areas of change.Remove Insert22 22B. Insert the following sections into the Safety Analysis Report.1. Add new Section 10.5.32. Add new Section 11.1.1.1.6

3. Append to Section 13.2.6.24. Add new Appendix A to Chapter 135. Add new Appendix

.B to Chapter 136. Change Section 10.4.17. Add new Section 10.4.1.48. Append to Section 1 3.2.6.29. Add Reference 13.19 to ChaPter 13

• Sunrestricted area.3.8 Experiments 3.8.1 Reactivity Limits.Applicability.

This specification applies to the reactivity limits on experiments installed in the reactor and in-tank experiment facilities. Obiective. The objective is tQ assure control of the reactor during the irradiation or handling of experiments adjacent to or in the reactor core.Specification. The reactor shall not be operated unless the following conditions governing experiments exist:a. The absolute reactivity worth of any moveable experiment in the CentralIrradiation Facility shall be less than $1.75 (1.23% AK/K); the absolute reactivity worth of anymoveable experiment not in the Central Irradiation Facility shall be less than one (1) dollar(0.7% AK/K).b. The absolute reactivity worth of any single secured experiment shall be lessthan the maximum allowed pulse ($1.75) (1.23% AK/K).c. The absolute total reactivity worth of experiments installed in the reactor andin-tank shall not exceed an absolute value of one dollar and ninety-two cents ($1.92) (1.34%AK/K), including the potential reactivity which might result from malfunction,

flooding, voiding, orremoval and insertion of the experiment.

Basis.*a. A reactivity limit of less than $1.75 specifically for the Central Irradiation Facility is based on the pulsing reactivity insertion limit (section 3.1.2) and on the design of thesample can assembly which allows insertion and withdrawal of experiments in a controlled manner (identical in form, fit, and function to a control rod). See also the MNRC SAR, Chapter13.2.2.2.1 for the maximum reactivity insertion discussion. A reactivity limit of less than one (1)dollar on a single moveable experiment not in the Central Irradiation Facility will precludepulsing if the experiment's fixturing should fail, since the resulting reactivity insertion would notcause prompt criticality if less than one dollar. Given that the reactor will not pulseinadvertently, the additional increase in transient power and temperature will be slow enough sothat the fuel temperature scram will be effective.

b. The absolute worst event which may be considered in conjunction with asingle secured experiment is its sudden accidental or unplanned removal while the reactor isoperating.

This would result in a reactivity increase less than a pulse of $1.92, analyzed in SARChapter 13, Section 13.2.2.2.1.

c. It is conservatively assumed that simultaneous removal of all experiments inthe reactor and in-tank experiment facilities at any given time shall not exceed the maximumreactivity insertion limit. SAR, Chapter .13, Section 13.2.2.2.1 shows that an insertion 22Amendmient No. 2 ADDITION TO MCCLELLAN NUCLEAR RADIATION CENTERSAFETY ANALYSIS REPORT -ARGON-41 PRODUCTION FACILITYNEW SECTION 10,5.310.5.3 Arcqon-41 Production FacilityThe Argon-41 Production Facility will produce 1-2 curies of 41Ar for research andcommercial use. The 41Ar will be produced by introducing argon gas into a stainless steelcontainer located in one of the silicon irradiation positions (adjacent to the graphitereflector and external to the reactor core -Figure 10.11 1A). All the components containing activated 41Ar are located in the reactor room.Argon gas from a commercial argon gas cylinder will supply the irradiation container.

After the irradiation container is pressurized (approximately 500 psig) to the desired level,the gas cylinder will be isolated from the irradiation container. To produce the desiredactivity level of 41Ar the sample will be irradiated for approximately 24 hours.After irradiation, liquid nitrogen is added to a Dewar. A remotely operated solenoid valveis opened to pressurize the cooling coils above the liquid nitrogen bath. The Dewar is thenraised to cover the cooling coils and 41Ar is cryogenically extricated from the irradiation container. After extrication is completed, the solenoid valve from the irradiation container is shut and another remotely operated solenoid valve is opened. This allows diffusion of41Ar gas to the sample container. The liquid nitrogen Dewar is lowered, exposing thecooling coils to room temperature. When that portion of the system between the coolingcoils and the sample container has reached equilibrium the sample container will beisolated and..removed from the room. The coil is surrounded with a lead shield to minimizethe radiation exposure to personnel. A catch tank surrounds the Dewar to contain any liquid nitrogen escaping from'the Dewaror in the unlikely event of a total failure of the Dewar.Over pressure protection of the overall system is provided by several relief valves thatvent to an over pressure tank. The over pressure ta~nk is protected by its own relief valvewhich vents to the reactor room. The tank is located as high as possible in the reactorroom.All piping and valves in the system are stainless steel. Compression fittings or double-ended shut-off quick connectors are used for allconnections normally in contact with the41Ar.The Argon-41 Production Facility consists of several different components, with the majorcomponents listed below. 0COMPONENT Irradiation Container Over PressureRelief ValvesOver PressureRelief TankMATERIAl304 stainless steel304 stainless steelCarbon steel304 stainless steel304 stainless steelDESCRIPTION The irradiation container is a 1000 mlsample cylinder with a working pressureof 1 800 psig and a burst pressure of6000 psig. It conforms to the "Shipping Container Specifications" from the U.S.Code of Federal Regulations, Title 49 orBureau of Explosives Tariff No.BOE6000.The adjustable proportional pressure reliefvalves have a working pressure up to6000 psig. When upstream pressureovercomes the force exerted by thespring, the poppet opens, allowing flowthrough the valve. As the upstreampressure increases, flow through thevalve increases proportionately. Crackingpressure is only sensitive to inlet pressureand is not affected by outlet pressure. 30 gallon tank.ValvesTubingBellows sealed valves.1/4-inch and Y/=-inch. NEW SECTION 11.1.1.1.6 11.1.1.1.6 Araqon-41 from the Argon-41 Production FacilityAr-41 will be produced by the Ar-41 Production Facility (see Chapter 10) as needed. TheAr-41 that is produced by the Ar-41 Argon Production Facility will be contained in thesystem so there should be no increase in the Ar-41 levels in the reactor room or the Ar-41that is released to the unrestricted area. Catastrophic failure of the system will not resultin any 10 CFR 20 limit being exceeded and is further discussed in Chapter 13.APPEND TO SECTION 13.2.6.2The Argon-41 Production Facility (see Chapter 10) can produce argon-41 in excess of theamounts analyzed in Appendix A of the MNRC Safety Analysis Report. However, if thesystem releases argon-41, the gas will be contained in the reactor room and the existing 0reactor room ventilation system will be used in recirculation mode to prevent releasing argon-41 to the environment, recirculating the gas until it decays. The existing StackContinuous Air Monitor will also be used to verify any release outside the MNRCboundary. If the system had a catastrophic failure and 4 curies of argon-41 were released to thevolume of the reactor room, the argon-41 concentration in the reactor room would beapproximately 22 R/hr (based on a semi-infinite cloud; see calculation in Chapter 1 3,Appendix A). Personnel would be evacuated from the reactor room and access would berestricted. The reactor room ventilation system (as described in Chapter 9) would, beoperated in the recirculation mode for approximately one day before the dose rate fromargon-41 decays to less than 1 mR/hr. Therefore, the argon-41 Discharge Limit defined inthe MNRC Technical Specifications will not be exceeded due to the recirculation mode ofthe reactor room ventilation system.Other potential accidents include failure of the irradiation container due tooverpressurization from the argon gas supply cylinder, since a new argon supply cylinderis typically delivered at 2200 psig and the container is rated for 1800 psig. However, thisrequires multiple failures and is considered non-credible: a) the operator would have toviolate an operational procedure; b) the regulator would have to fail, and c) at the sametime the pressure relief valve would have to fail. Also, liquid nitrogen could spill into thereactor tank, causing expansion of the water and expelling a portion of tank water. Toprevent this, a catch basin surrounds the Cold Trap, and the liquid nitrogen is suppliedthrough a pipe in the reactor room wall connecting the trap to a supply container in theequipment room. A third accident could result if the pressure relief valve became chokedwith supersonic flow; however, the flow rates are estimated to be less than sonic (seecalculat~ion in Chapter 13, Appendix A).NEW APPENDIX A TO CHAPTER 13ARGON-41 CONCENTRATION IN REACTOR ROOMGIVEN:1. Reactor room volume =-7.39x10 3 ft3 tReference 1112. 4 curies Ar-41 in argon production system3. D(y)=,2 = O.25Evx [Reference 21Dy= = gamma dose rate from a semi-infinite cloud (rad/sec) Ev = average gamma energy per disintegration (Mev/dis) = 1 .2936 Mev/dis for Ar-41[Rfrne3 0CALCULATIO X)N:*= concentration of gamma emitting isotope in the cloud (Ci/m3)X = (4Ci)/[7.39xl10 3ft3)(1 m3/35.314 ft3) = 1 .91!x 0.2 Gi/m3D(y)=,2 = 0. 25Eyx= (0.25)( 1.2936 Mev/dis)(1 .91 xl 0.2 Cl/rn3)= (0.0062 rads/sec)(3600 seclhr)= 22.24 radslhrD = Doe~xt = -(1/A)In(0D/D) = -(T112Iln2)ln(D/D 0)For 0 = 1 mrad/hrt = -(1 .8hr/In2)ln(1/22,240) = 26 hr

REFERENCES:

1. MNRC Safety Analysis Report, Figure 9.1.1.2. The Health Physics and Radiological Health Handbook (Revised Edition),

editedby Shelein,

p. 4393. Nuclides and Isotopes, 14m' edition, Chart of the Nuclides, GE Nuclear Energy,p. .22 0, .NEW APPENDIX B TO CHAPTER 13*SONIC FLOW FOR ARGON-41 PROJECTAssume: Perfect GasConstants:

Property Value UnitsR 208 N-rn/k g-degKk(c,/c,) 1 .67 dimensionless Problem: determine if the pr~essure relief valve will experience choking due to supersonic flow.Solution: First, calculate the speed of sound in argon at 40 degrees C and -200 degrees C:given c =speed of sound in a medium = (kRTgc)fl c = [1.67(208 N-m/kg-degK)(40+273)K(1 kg-rn/N-sec 2 )]P= 329.7327 rn/s at 40 degrees Cc =[1 .67(208 N-m/kg-degK)(-200 +273)K( 1 kg-rn/N-sec 2 = 159.2397 rn/s at -200 degrees CNext, calculate the velocity of the argon in the tubing at the pressure relief valve:given volumetric flow rate V = (velocity)(area) From tech data on valve, assume V = lft3/min, based on air and relief at 1125 psiV = (1 ft3/min)(12 in/ft)3(2.54 cm/in)3(1 min/60 sec)= 471.9474 cm3/secArea = 2 = 3.14(0.18in/2) 2 = 0.025434 in2 based on 1/4 inch tubing= 0.16409 cm2Velocity = V/Area = 28.7615 rn/secMach Number = Velocity/c = 0.180618 at -200 degrees C= 0.087227 at 40 degrees C

== Conclusion:==

Gas velocity at the relief valve is less than the speed of sound in argon andtherefore should not experience choking at the valve.

Reference:

Fundamentals of Gas Dynamics, Zucker, pp.89, 130-1 33, 375. ADDITION TO MCCLELLAN NUCLEAR RADIATION CENTER SAFETY ANALYSIS REPORTAND TECHNICAL SPECIFICATIONS -CENTRAL IRRADIATION FACILITYCHANGE SECTION 10.4.1The Central Irradiation

Facility, located in the center of the reactor core, may containeither a plug assembly (as described in sections 10.4.1 .1 through 10.4.1 .3 and Figure10.7) or a moveable sample can system (as described in section 10.4.1.4).

All parts areremovable from the reactor using underwater tools.NEW SECTION 10.4.1.410.4.1 .4 Central Irradiation FacilityThe central irradiation facility allows samples to be inserted into the reactor core (i.e.central facility) while operating the reactor at power. The reactor operator controls theinsertion and removal of samples from the central facility through the use of a drivemechanism similar to the control rods.The central thimble is approximately 52 inches in length and 4.22 inches outer diameterwith an inside dimension of approximately 4.0 inches. The central thimble, once in place,passes through the upper grid plate, the lower grid plate and the safety plate. Aluminumshims have been added to the outer periphery of the central thimble in the fuel region.These shims align the central thimble and displace the water from the scallops of the fuelelement locations in the B hex ring 4.25-inch hole. Two captive bolts attach the centralthimble to the upper grid plate. These bolts prevent the accidental removal of the facilitywhen removing samples from the central thimble.An 1100 aluminum slug located inside the central thimble is normally positioned in thereactor core. The aluminum slug is 4 inches in diameter and 24.75 inches in length. Thisvoids the water from the central thimble when the sample can is removed from thethimble.An orifice plate is located on the bottom of the central thimble. In the event the aluminumslug releases from the locating holes and falls to the, bottom of the central thimble, therate of decent will be less than the normal control rod drive speed.The sample can is approximately 30.5 inches long with an outside diameter of 3.99inches and an inside diameter of 3.75 inches. The can could be free flooding or dry, andis used to position samples for irradiation in the reactor core. The positioning of samplescan be accomplished during full power reactor operations (i.e. 2 MW). During insertion into the reactor core and while in the reactor core the assembly has the capability of beingrotated.The drive mechaauism has the same type of drive motor as the control rod drives exceptthe model selected will have more torque. All other aspects of the motor and controller are identical. There are two sets of controls, one in the reactor room and the other in the control room.Normal operational control is from the reactor console where the reactor operators wiBltreat the insertion and removal of the samples as if they were control rods. The reactorroom controls can only be enabled from the reactor console. The normal indicators are asfollows:"A. Power On, switch and indicator (control room only).B. Reactor Room control enable switch and indicator (control room only).C. One set of momentary UP/DOWN switches for 1/22 speed drive.D. One set of momentary UP/DOWN switches for full speed drive.E. Indicators for UP, DOWN, and CLOSE TO DOWN positions. F. Digital indication of the sample can position, scaled 0-1000 units.G. Rotation ON, switch and indicator. Limit switches on the rack are used in the logic design to determine end of travelindications, stop driving limits and start/stop rotation of the carrier.APPEND TO SECTION 13.2.6.2Another potential accident involves the Central Irradiation Facility (see Chapter 10) since itmay be considered similar to a control rod. Therefore, consider three potential scenarios for an uncontrolled reactivity insertion analogous to the Uncontrolled Withdrawal Of aControl Rod (see Section 13.2.2.2.2). First, if the material in the sample can were ofsufficiently different worth than the aluminum

cylinder, the sample can would causereactivity changes in the same fashion as a control rod, and either operator error ormechanical failure could cause an uncontrolled reactivity insertion.

Second, if thealuminum cylinder failed to engage upon the sample can's insertion, a water void wouldbe created in the central facility as the aluminum cylinder descended ahead of the samplecan. Similarly, if the aluminum cylinder failed to replace the can upon removal from thecentral facility a water void would result.All three of the above scenarios can be bounded by the Uncontrolled Withdrawal of aControl Rod analysis (Section 13.2.2.2.2). Specifically, the Central Irradiation Facilitymust have less reactivity and must drive slower than the control rod analyzed ($3.50 and42 inches/minute, respectively). To that end, the reactivity of any material in the samplecan shall be measured at low power to verify it's worth is not only less than $3.50 butalso less than $1.75, the reactivity limit for the Central Irradiation Facility (based on theTechnical Specification limit of $1 .75 for the pulsed reactivity insertion). For example, theworth of a silicon ingot in the previous 1 MW in-core experiment facility was measured at$0.73 positive (vs. water, reference exp. # 96-01, 1/30/96, reactor run #2411.). Theworth of an aluminum cylinder vs. void and vs. water has been analyzed by computersimulation (Reference 13.19). The most positive reactivity effect in the computersimulation is from Case 3 to Case 9, where the voided sample can is lowered 18 inches,resulting in an increase of about $0.06. The most negative reactivity effect is from Case3 to Case 1 2, where in an accident the sample can not only floods but also the aluminumcylinder drops, resulting in a decrease of about $.1.76. Thus, the worth of the sample canor the aluminum cylinder vs. water is less than $3.50, and also less than the most reactive control rod (for example, a typical regulating rod worth is $2.57, measured 6/98).With respect to the drive mechanism, the maximum drive speed is identical to the rodspeed analyzed in the MNRC SAR (Section 13.2.2.2.2). Furthermore, in the event offailure of the aluminum cylinder to engage upon installation of the sample can, the base ofthe Central Thimble is designed (by sizing the hole in the base) to allow the aluminumcylinder to descend at no more than the analyzed 42 inches/minute. Therefore, theaccident analysis in Chapter 13 of the MNRC SAR for Uncontrolled Withdrawal of aControl Rod (Section 13.2.2.2.2) is sufficient to bound any accident associated with theCentral Irradiation Facility since: a) the material in the sample can shall be measured andverified to be less than $1.75 (half of the analyzed $3.50); b) the drive speed cannotexceed the analyzed 42 inches/minute; and c) the aluminum cylinder cannot falluncontrolled faster than the analyzed 42 inches/minute.

Finally, physical impact on the fuel is considered non-credible since the sample can isalways contained in a guide tube or attached to a drive mechanism such that it is unlikelyto drop onto the core (see description in Section 10.4.1.4).

ADD REFERENCE 13.19 TO CHAPTER 1313.1 9 Liu, H. Ben, "Safety Analysis for the Central Irradiation Facility (CIF) at the MNRC",Memorandum to Wade J. Richards, September 22, 1998.CHANGE TECHNICAL SPECIFICATION 3.8.1 AS FOLLOWS:(a) The absolute reactivity worth of any moveable experiment in the CentralIrradiation Facility shall be less than $1 .75 (1 .23% Ak/k); the absolute reactivity worth of any moveable experiment not in the Central Irradiation Facility shall beless than one (1) dollar (0.7% Ak/k).CHANGE TECHNICAL SPECIFICATION 3.8.1 "BASIS" AS FOLLOWS:(a) A reactivity limit of less than $1 .75 specifically for the Central Irradiation Facilityis based on the pulsing reactivity insertion limit (section 3.1.2) and on the designof the sample can assembly which allows insertion and withdrawal ofexperiments in a controlled manner (identical in form, fit, and function to acontrol rod). See also the MNRC SAR, Chapter 13.2.2.2.1 for the maximumreactivity insertion discussion. A reactivity limit of less than one (1) dollar on asingle moveable experiment not in the Central Irradiation Facility will precludepulsing if the experiment's fixturing. should fail, since the resulting reactivity insertion would not cause prompt criticality if less than one dollar. Given that thereactor will not pulse inadvertently, the additional increase in transient power andtemperature will be slow enough so that the fuel temperature scram will beeffective. 0 9 STATES"NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C..20588-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 2 TOAMENDED FACILITY OPERATING LICENSE NO. R-130DEPARTMENT OF THE AIR FORCE ATMcCLELLAN AIR FORCE BASEDOCKET NO. 50-60

71.0 INTRODUCTION

By letter dated January .11, 1999, the Department of the Air Force at McClellan Air ForceBase (the licensee) submitted a request for amendment of the Technical Specifications (TSs), Appendix A, to Facility Operating License No. R-1 30 for the McClellan NuclearRadiation Center TRIGA Research Reactor (MNRC), and changes to the Safety AnalysisReport. The amendment provides for the installation of an Argon-41 Production Facilityand a Central Irradiation Facility. The installation of the Argon-41 Production Facility doesnot require any change or expansion of the TSs since an experiment failure will not resultin airborne radioactivity in the reactor room or the unrestricted area exceeding theapplicable dose limits already prescribed. The installation of the Central Irradiation Facilityrequires a change to TS 3.8.1 with regard to the maximum reactivity worth of a moveable* experiment. The change increases the reactivity limit of a moveable experiment in theCentral Irradiation Facility to $1 .75, corresponding to the pulse limit specified in TS 3.1 .2.2.0 EVALUATION Argon-41 Production FacilityThe licensee has requested amendment of the Safety Analysis Report to provide for theinstallation of an Argon-41 Production Facility. The purpose of the facility is to produceArgon-41 for research and commercial uses. Argon gas from a pressurized argon bottle isintroduced into a stainless steel container located in a position external to the core, but inthe reactor tank. Sufficient argon gas is admitted into the irradiation facility to pressurize it to about 400 psig. The gas is irradiated for 24 hours at full power (48 Megawatt-hours) and is converted to one to two curies of argon-41. The now-radioactive argon-41 isremoved cryogenically and admitted to sample containers. Overpressure protection isprovided by stainless steel relief valves that vent to a 30-gallon carbon steel* overpressure tank which is also protected With a relief valve. The relief valves have a working pressure

• 0-2-of up to 6000 psig. The overpressure tank relief valve vents to the reactor room. Allpiping (1/4 and Y/= inch 304 stainless steel) is anchored to prevent pipe whip in the eventof pipe failure.

The irradiation container has a working pressure of 1 800 psig with a burstpressure of 6000 psig.After the argon gas has been irradiated, the gas is transferred to the sample containers. Acooling coil which has been evacuated with a vacuum pump is immersed in a liquidnitrogen bath. The transfer process is started by opening a valve between the irradiation facility and cooling coil. The argon gas diffuses to the sample containers. When radiation surveys indicate that the transfer process is completed, the sample containers are valvedoff, removed, and placed in.a shipping cask.The licensee has analyzed the case of a catastrophic failure of the irradiation container, which releases 4 curies of argon-41 (about twice as much as is actually produced) into thereactor room resulting in an initial dose rate of approximately 22 rads per hour. Operation of the reactor room ventilation system in the recirculation mode for a period of one daywill result in a dose rate of approximately 1 mrad per hour. The argon-41 discharge limitas defined in the Technical Specifications will not be exceeded. The licensee has considered other potential accidents. These include overpressurization ofthe irradiation container, spilling liquid nitrogen into the reactor tank, and the choking of arelief valve due to supersonic flow. Overpressurization of the irradiation container requiresmultiple mechanical failures and operator violation of the procedure governing the use ofthe production facility. To prevent the spilling of liquid nitrogen into the reactor tank, acatch basin is installed around the liquid nitrogen bath. Finally, the licensee has analyzedthe flow through the relief valves and has determined that the flow remains subsonic, thuspreventing choking at the valve.Central. Irradiation FacilityThe licensee has requested amendment of the 'Technical Specifications and SafetyAnalysis Report to provide for the installation of a Central Irradiation Facility. The facilityallows samples to be inserted into the reactor core while operating the reactor at power.Control of the facility is through use of a drive mechanism similar to that of the normalcontrol rods, and a reactor operator controls the insertion and removal of samples. Drivespeeds are equal to those of the normal control rods.The central thimble is essentially a vertical guide tube which passes through the upper gridplate, the lower grid plate and the safety plate, resting on the tank floor. lA sample canand an aluminum slug move within the central thimble. An aluminum slug normallyoccupies a position in the reactor core. When the sample can is inserted, the aluminumslug moves downward out of the co)re, and its position in the core is replaced by thesample can. Control of the system is only from the reactor c:onsole. The system is provided with*indications

• similar to that of the normal control rods, which include POWER ON, UP,DOWN and CLOSE TO DOWN position indicators, digital indication of sample can position, and UP/DOWN control switches.

From a safety analysis point of viejw, the system can be considered to be an additional control rod and so the analyses in the Safety Analysis Report with respect to control rodmalfunctions are applicable. In particular, the analysiz of an Uncontrolled Withdrawal of aControl Rod (Safety Analysis Report section 13.2.2.2.2) provides a bounding envelope. That analysis showed that an uncontrolled rod withdrawal, at full power of 2 MW, at themaximum withdrawal speed of 42 inches per minute would result in a peak reactivity insertion of $0.25, much lower than the technical specification pulse reactivity insertion limit of $1 .75. Although the maximum single rod worth is approximately $2.65, a rodworth of $3.50 was used to allow for reasonable variations. In order to bound accidents involving the Central Irradiation

Facility, it is required to showthat the worths of the sample can and the aluminum slug are not only less than $3.50,but also less than the pulse limit of $1.75. The licensee has performed a computersimulation (SAR Reference 13.19) of the reactivity changes associated with variousscenarios,-

including normal operations and accidents. The most limiting case, the floodingof the sample can accompanied by a drop of the aluminum slug, results in a reactivity insertion of $1 .76, much less than the most reactive control rod ($3.50) used in the rodwithdrawal accident. Thus the Central Irradiation Facility is bounded by the previously-analyzed rod withdrawal accident. 3.0 ENVIRONMENTAL CONSIDERATION This amendment involves changes in the installation or use of a facility component locatedwithin the restricted area as defined in 10 CFR Part 20 or changes in inspection andsurveillance requirements. The staff has determined that this amendment involves nosignificant increase in the amounts, and no significant change in the types, of anyeffluents that may be released off site, and no significant increase in individual orcumulative occupational radiation exposure. Accordingly, this amendment meets theeligibility criteria for categorical exclusion set forth in 10 CFR 51 .22(c)(9). Pursuant to 10CFR 51 .22(b); no environmental impact statement or environmental assessment need beprepared in connection with the issuance of this amendment.

4.0 CONCLUSION

The staff has concluded, on the basis of the considerations discussed above, that(1) because the amendment does not involve a significant increase in the probability orconsequences of accidents previously evaluated, or create the possibility of a new ordifferent kind of accident from any accident previously evaluated, and does not involve asignificant reduction in a margin of safety, the amendment does not involve a significant hazards consideration; (2) there is reasonable assurance that the health and safety of thepublic will not be endangered by the proposed activities; and (3) such activities will beconducted in compliance with the Commission's regulations and the issuance of thisamendment will not be inimical to the common defense and security or the health andsafety of the public.Principal Contributor: Warren J. EresianDate: MArch 1, 1999 999 9** ** 1~STATES,AUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. Brigadier General Michael P. WledemerCommander Sacramento Air Logistics CenterSM-ALC/TI-1 5335 Price AvenueMcClellan AFB, California 95652-2504 Vice Chancellor Kevin SmithOffice of the Chancellor University of California, DavisOne Shields AvenueDavis, California 95616-8558

SUBJECT:

ORDER APPROVING THE TRANSFER OF THE FACILITY OPERATING LICENSE FOR THE McCLELLAN NUCLEAR RADIATION CENTER FROM THEDEPARTMENT OF THE AIR FORCE TO THE REGENTS OF THE UNIVERSITY OF CALIFORNIA. AND APPROVING CONFORMING AMENDMENT (TAC NO. MA3477)Dear General Wiedemer and Dr. The enclosed Order Is in response to the application dated April 13,.1999, as supplemented onJuly 19 and August 4-,1999, and January 18 and 27, 2000, requesting approval of the transferof Operating License No. R-1 30 for the McClellan Nuclear Radiation Center from theDepartm~ent of the AIr Force to the Regents of the University of California, and approval of aconforming amendment to reflect the transfer. The enclosed Order provides consent to theproposed

transfer, pursuant to Section 50,80 of Title 0oaf the Code of Federal Reaulatlona, and approves Amendment No. 3. Also, enclosed are two copies of the indemnity agreement forthe facility.

The. Vice Chancellor for the University should sIgn one copy and return it to me.The University should keep the other for its records.The Order has been forwarded to the Office of the Federal Register for publication. Sinc~syWarreni~J. Ere fan, Project ManagerEvents Assessment, Generic C~ommunlcations and Non-Power Re~ctom BranchDivIsion of rovement ProgramsOffice of Nuclear Reactor Regulation Docket No. 50-607*

Enclosures:

.1. Order2. Amendment No.3*.3. Safety Evaluation 4, IndemnityAgreement. .*Senextp ge McClellan AFB TRIGA REACTOR Docket No, 50-607cc;Dr. Wade J. RichardsSM-ALC/TI-16335 Price Avenue, Bldg. 258McClellan AFB, CA 95652-2504 Col. Robert CapellHQ AFMC/SGC4225 Logistics Avenue, Suite 23Wright-Patterson AFB, OH 45433-5762 Lt, Cot. Catherine ZeringueHQ AFSC/SEW9570 Avenue G, Building 24499Kircland AFB, New Mexico 871 17-5670Test. Research, and TrainingReactor Newsletter 202 Nuclear Sciences CenterUniversity of FloridaGainesville, FL 32611 7590-01 -PUNITED STATES OF NUCLEAR RIEGU.LATORtY COMMISSION

• In the Matter of ))DEPARTMENT OF THE AIR FORCE ) Docket No, 50-607)(McClellan Nuclear'Radiation Center) )ORDER APPROVING TRANSFER OF LICENSEAND CONFORMING AMENDMENT I,The United States Air Force (USAF) is the owner of the McClellan Nuclear Radiation
• Center (MNRC) and Is authorized to possess, use, and operate thle facility as reflected in*Operating License No, R-130, The Nuclear Regulatory Commission issued Operating License*No. R-1 30 on August 13, 1998, pursuant to Part 50 of Title 10 of the .Code of_ FederalRegufation~s (10 CFR Part 50). The facility is located on McClellan Air Force Base InSacramento, California.

Ii.By letters dated April 13, 1999, the USAF and the Regents of the University of California (University of California) each submitted an application req~uesting approval of the proposedtransfer of Operating License No, R-1 30 from the USAF to the University of California. TheUniversity of Calliornia at Davis (UCD), part of the University of California, was proposed to bethe actual operator of the facility. The application was supplemented by submittals datedJuly 19 and August 4, 1999, and January 18 and 27, 2000, The initial application and thesupplements are hereinafter collectively referred to as "the application" unless otherwise indicated4. ENCLOSURE 1 According to the application, the USAF has agreed to convey the MNRC to the University of California. After completion of the proposed license transfer, UCD would be the soleoperator of the MNRC. The application also sought the approval of a conforming amendment. This conforming amendment is necessary to remove references to the USAF from theoperating license and replace them with references to the UCD, as appropriate, as well as tomake other miscellaneous administrative changes to the operating license to ref lect thetransfer. Under 10 CFR 50.80, no license for a production or utilization

facility, or any rightthereunder, shall be transferred, directly or Indirectly, through transfer of control of the license,unless the Commission shall give Its consent in writing.

Upon review of the information in theapplication and other information before the Commission, the NRC staff has determined thatthe University of California Is qualified to hold the license, and that the transfer of the license tothe University of California is otherwise consistent ~with applicable provisions of law, regulations, and orders issued by the Commission. The NRC staff has further found that the application forthe proposed license amendment complies with the standards and requirements of the AtomicEnergy Act of 1954, as amended, and the Commission's rules and regulations set forth in 10CFR Chapter 1; the facility will operate in conformity with the application, the provisions of theAct, and the rules and regulations of the Commission: there Is. reasonable assurance that theactivities authorized by the proposed license amendment can be conducted withoutendangering the health and safetyof the public and that such activities will be conducted incompliance with theCommission's regulations: the issuance of the proposed licenseamendment will not be inimical to the common defense and security or to the health and safetyof the public; and the issuance of the proposed amendment will be in accordance with 10 CFR r-P ." = 1af1T1 NU.SS5r0 r P.5/) Part 51 of the Commission's regulations and all applicable requirements have been satisfied. The foregoing findings are supported by a Safety Evaluation dated December 2, 1999.Accordingly, IT IS KEREBY ORDERED that the transfer of the license as described hereinto the University of California is approved, subject to the following, condition: Should the transfer of the license not be completed by June 30, 2000, this Order shallbecome null and void, provided,

however, on written application arnd for good causeshown, such date may in writing be extended.

IT IS FURTHER ORDERED that, consistent with 10 CFR 2,1315(b), a license amendment that makes changes, as indicated in Enclosure 2 to the cover letter forwarding this Order, toconform the license to reflect the transfer is approved. This Order is effective upon issuance. Dated at Rock'vilie,

Maryland, this 31't day of ;January 2000,FOR THE= NUCLEAR REGULATORY COMMISSION David B. Matthews, DirectorDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation

4.

STATESWASHINGT"ON, D.C. 20555-0001 DEPARTMENT O T.HE AIR FORCF ATMCCLELLAN. AIR FoRCE BASEDOCKET NO. 50-607AMENDMENT TO AMENDED FACILITY OPERATING LICENSEAmendment No. 3License No. R-1301.The U.$. Nuclear Regulatory Commission (the Commission) has tound thatA. The application for an amendment to Amended Facility Operating License No. R-130filed by tile Department of the Air Force at McClellan Air Force Base and the Regentsof the University of California on April 13, 1999, as supplemented on July 19 andAugust 4, 1999, and January 18 and 27, 2000, conmpiies with the standards andrequirements of the Atomic Energy Act of 1954, as amended (the Act), and theregulations of the Commission as stated In Chapter I of Title 10 of the Code ofFederal R~equlatlons (10 CFR);B. The facility will operate in conformity with the application, the provisions of the Act,and the rules and regulations of the Commission; C. There is reasonable assurance that (i) the activities authorized by this amendment can be conducted without endangering the health and safety of the public and (il)such activties will be conducted in compliance with the regulations of theCommission; D. The issuance of this amendment will not be~inlrmicalto the common defense andsecurity or to the health and safety of the public;and E. This issuance of this amendment is in accordance with the regulations of theCommission as stated in 10 CFR Part 51, and all applicable requirements have beensatisfied.

2. Accordingly, the license is amendedas indicated in the attachment to thisilcense amendment, ENCLOSURE 2

FEB.*1.006

09M N.955 P.7/1.4-2-3. This license amendment is effective as of the date of issuance, FOR THE NUCLEAR REGULATORY COMMISSION Ledyard B. Marsh, ChiefEvents Assessment, Generic Communications and Non-Power Reactors BranchDivsion of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation

Enclosures:

1.2.Amended Facility LicenseAppendix A, Technical Specifications changesDate of Issuance: January 31, 20004 rI" l" NUCLEAR REGULATORY COMMISSION

• ~WASHINGTON, D.C, 20885,=0001FACILITY OPERATING LICENSE~DOCKET NO., 50-607_REGENTS oF THE UNIVERSITY OF ALicense No. R-1301.The U.S. Nuclear Regula.tory Commission (the Commission) has found that:A. The application for license transfer, filed by the Regents of the University of California on April 13, 1999, as supplemented on July 19 and August 4, 1999, and January 18and 27, 2000 comnpiies with the standards and requirements of the Atomic Energy Actof 1954, as amended (the Act), and the Commission's rules and regulations as setforth in 10CFR Chapter I;B. Construction of the facility was completed in substantial conf ormity with the provisions of the Act, and the rules and regulations of the Commission; C. The facility will operate In conformity with the application, the provisions of the Act,and the rules and regulations of the Commission; D. There is reasonable assurance (I) that the activities authorized by this license can beconducted without endangering the health and safety of the public and (II) that suchactivities will be conducted in compliance with the Commission's regulations; E, The licensee is. technically and financially qualifiled to engage in the activities authorized by this operating license in accordan~ce with the regulations of theCommission; F. The licensee is a Nonprofit Educational institution and will use the facility foreducational programs arnd research, and has satief led the applicable provisions of10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements ofthe Commission's regulations; G. The issuance of this license will not be inimical to the common defense and securityor to the health and safety of the public; ."H-. This license is issued in accordance with 10 CFR Part 6.1 of the Commission's regulations, and all applicable requirements have been satisfied; andS.The receipt, possession, and use of the byproduct and special nuclear materials asauthorized by this license will be in accordance with the Commission's regulations in10 CFR Parts 30 and 70, including Sections 30.33. 70,23, and 70.31.Amendment No. 3

2. Facility License No, R-1 30 is hereby issued to the Regents of the University of California as follows:A. The license applies to the TRIGA nuclear reactor (the facility) owned by the Regentsof the University of California (the licensee),

The facility is located on the McClellan IAir Force Base, Sacramento, California, B, Subject to the conditions and requirements Incorporated herein, the Commission hereby licenses the Regents of the University of California at the McClellan Nuclear(i) Pursuant to Section 104o of the Act arid 10 CFR Part 50, "Domestic Licensing ofProduction and Utilization Facilities," to possess, use, and operate the facility atthe designated location at McClellan Air Force Base in accordance with theprocedures and limitations set forth in this license.(Ii) Pursuant to the Act and 10 CFR Part 70, "Domestic Licensing of Special NuclearMaterial,= to receive,

possess, and use up t0 21.0 kilograms of contained uranium-235 enriched to less than 20 percent In the isotope uranium-235 in theformat reactor fuel; up to 4 grams of contained uranium-235 of any enrichment in the~form of fission chambers; u~p to 16,1 kilograms of contained uranium-235 enriched to less than 20 percent in the isotope uranium-235 in the form of plates;and to possess, but not separate, such' special nuclear material as may beproduced by the operation of the facility.

(iii) Pursuant to the Act and 10 CFR Part 30, "Rules of General Applicability toDomestic Licensing of Byproduct Material," to receive,

possess, and use a 4-curie sealed americium-beryllium neutron source in connection with operation ofthe facility; a 55-millicurie sealed cesium-I137 source for instrument calibrations;

• small instrument calibration and check sources of less than 0.1 millicurie each;and to possess, use, but not separate, except for byproduct material produced Inreactor experiments, such byproduct material as may be produced by theape ration of the facility.

C. This license shall be deemed to contain and Is su~bj~ect to the conditions specified inParts 20, 30, 50, 51, 55, 70, and 73 of 10 CFR Chapter I; to all applicable provisions of the Act;and to the rules, regulations, and orders of the Commission now or hereafter in effect and tothe additional conditions specified, below:(i) Maximum Po~wer LevelThe licensee is authorized to operate the facility at steady-state power levels not inexcess of 2300 kilowatts (thermal) arid in the pulse mode with reactivity insertions notto exceed $1.75 (1.23 %/0k/k).Amndent N..h 3 3-3(ii) Technical S~oecfficatlonis The Technical Specifications, as revised through Amendment No. 3, are hereby. fincorporated in the license. The licensee shall operate the facility in accordance withthe Technical Specifications. (lii) Physical Securityv lanThe licensee shall fully implement and maintain in effect all provisions of theCommission-approved physical security plan, including all amendments and revisions made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The approvedplan, which is exempt from public disclosure pursuant to the provisions of 10 CFR2.790, is entitled "Physical Security Plan for the MNRC TRIGA Reactor Facility," Revision 3, and is dated August 1996,D. This license is effective as of the date of issuance and shall expire twenty (20) yearsfrom its date of issuance. FOR THE NUCLEAR REGULATORY COMMISSION Previously signedI byOrigina/ signed bySamuel J, Collins, Directoroffice of Nuclear Reactor Regulation Date of issuance: August 13, 1998Amendment No. 3 Q TO LICENSEAMENDMENT NO.3AMENDED FACILITY OPERATING LI.CENSE NO. R-!30DOCKET NO; 50-807Replace the following pages of Appendix A, "T'echnlcal Specificationts,= with the enclosedpages. The revised pages are identified by amendment number and contain vertical linesindicating the areas of change.*Remove1394041*139404. TECHNICAL SPECIFICATIONS FOR THEU.C. DAVIS MCCLELLAN NUCLEAR RADIATION CENTER (MNRC)GeneralThe McClellan Nuclear Radiation Center (MNRC) reactor is operated by the University ofCalifornia, Davis, CA. The MNRC research reactor Is a TRIGA type reactor. The MNRC iprovides state-of-the-art neutron radiography capabilities. In addition, the MNR~C provides a*wide range of irradiation servic~es far both research and industrial needs. The reactor operatesat a nominal steady start power level up to and including 2 MW. The MNRC reactor is alsocapable of square wave and pulse operational modes. The MNRC reactor fuel Is less than 20%enriched in uranium-235, 1.0 D~efinitions 1.1 ,As Low As Reasonab~ly, Achievable (ALARA), As defined in 10 CFR Part 2.0.1.2 Licens ed DOerators. A MNRC reactor operator is an individual licensed by theNuclear Regulatory Commission (e.g., senior reactor operator or reactor operator) to carry outthe duties and responsibilities associated with the position requiring the license.1.2.1 Senior. Reactor QOerator. An individual who is licensed to direct theactivities of reactor operators and to manipulate the controls of the facility. 1.,2.2 Reactor Onerator. An individual who is licensed to manipulate thecontrols of the facility and perform reactor-related maintenance. 1.3 A channel is the combination of sensor, line amplifier, processor, andoutput devices which are connected for the purpose of measuring the value of a parameter. 1,.3.1 Channel Test. A channel test is the Introduction of a signal into thechannel for verification that it is operable..,.' 1.3.2 Channel Calibratlaon. A channel calibration is an adjustment of thechannel such that its-output corresponds with acceptable accuracy to known values of theparameter which the channel measures. Calibration shall encompass the entire channel,including equipment actuation, alarm or trip, and shall be deemed to include a channel test.1.3.3 Channel Ch~eck. A channel check is a qualit~ative verification ofacceptable performance by observation of channel behavior. This verification, where possible, shall include comparison of the channel with other independent channels or systems measuring the same variable. 1 Amendment No .3. bViCECHACELOR OR ESERCHVICE CHANCELLOR FOR ADMINISTRATION ai.U .D SAFETYK CMMITTEEs IIIsuPERVISOR SUPERVwSOR ...* ------------.[OPERATIONS STAFFI HEALTh- PHYSICS STAFF]~UNIVERSITY MANAGEMENT ORGANIZATION ~Figure 6, !0: Normal Adminisirative Reporting Channel-Technical Review, Communications and Asisisance .

• ~ *I*~C*Lff**J~J

.1. * .LC.[~I.j J.* J.'-tr 7n ----LI.I.VICE OFFzICE OF' RESEARCH II1.* I----I* TUCIEAR SAFETYL AND UCENSINGNUCLEAR SAFETY AND LICENSING

REVIEWS, APPROVALS ANDRECOMMENDATIONS COMMUNICATION LICENSED ACTIVITIES UC. Davis McCleIlan Nuclear Radiatiog Center Lit'easing Organization Figure 6.254An~Iendment Wo. .3

• ,%.UNITED STATESS NUCLEAR REGULATORY COMMISSION' 0, .0. S5-0001Docket No. 50-607This indemnity agreement No. E-40 is entered~into by and between ths University of California at Davis (hereinafter referred to as the licensee) and the United States Nuclear Regulatory Commission (hereinafter referred to as the Commission) pursuant to subsection 170(k) of theAtomic Energy Act of 1954, as amended (hereinafter referred to as the Act).Article IAs used in this agreement,

1. Nuclear reactor, byproduct material,,

person, source material, specIal nuclear material, andprecautionary evacuation shall have the meanings given them in the Atomic Energy Act of 1954,as amended, and the regulations issued by the Commission.

2. (a) Nuclear incident means any occurrence including an extraordinary nuclear occurrence or series of occurrences at the location or in the course of transportation causing bodily injury,sickness,

disease, or death, or loss of use of property, arising out of or resulting from theradioactive, toxic, explosive, or other hazardous properties of the radioactive material.

(b) Any occurrence including an extraordinary nuclear occurrence or series of occurrences causing bodily injury, sickness, disease or death, or loss of or damage to property, or loss of useof property, arising out of or resulting from the radioactive, toxic,.explosive, or other hazardous properties ofi, The radioactive material discharged or dispersed from the location over a period of days,weeks, months or longer and also arising out of such properties of other material defined as theradioactive material in any other agreement or agreements entered into by the Commission under subsection 170(c) or (k) of the Act and so discharged or dispersed from the location asdefined in any such other agreement; orii. The radioactive material in the course of transportation and also arising out of suchproperties of other material defined in any other agreement entered into by the Commission pursuant to subsection 170(c) or (k) of the Act as the radioactive.material and is in thecourse of transportation shall be deemed to be a common octurre.nce. A common occurrence shall be deemed to constitute a single nuclear incident.

3. Extraordinary

nuclear, occurrence mean~s an event which the Commission has determined to be an extraordinary nuclear occurrence as defined in the Atomic Energy Act of 1984, asamended.4. In the course of transportation means In the course of transportation within the UnitedStates, or in the course of transportation outside the United States and any other nation, andmoving from one person licensed by the Commission to another person licensed by theCommission, including handling or temporary storage incidental

thereto, of the radioactive material to the location or from the location provided th~at:ENCLOSURE 4

FEB. ;I.28 5:52PM NO.95? P.2/6(a) With respect to transportationof the radioactive material to the location, suchtransportation is not by predetermination to be interrupted by the removal of the material fromthe transporting conveyance for any purpose other than the continuation of such transportation to the location or temporary storage incidental thereto;(b) The transportation of the radioactive material from the location shall be deemed to endwhen the radioactive material is removed from the transporting conveyance for any purposeother than the continuation of transportation or temporary storage incidental. thereto;(c) In the course of transportation as used in this agreement shall not include transportation ofthe r'adloactive material to the location if the material is also in the course of transportation fromany other location, as defined in any other agreement entered into by the Commission pursuant. to subsection 170(c) or (k) of the Act.5. Person Indemnified means the licensee and any other person who may be liable for public-liability.

6. Public liability means any legal liability arising out of or resulting.from a nuclear incident orprecautionary evacuation (including all reasonable additional costs incurred by a State, or apolitical subdivision of a State, in the course of responding to a nuclear Incident or precautionary evacuation),

except (1) claims under State or Federal Workmnen's Compensation Act ofemployees of persons indemnified who are employed (a) at the location or, if the nuclearIncident occurs in the course of transportation of the radIoactive

material, or the transporting

vehicle, and (b) in connection with the licensee's possession, use, or transfer of the radioaotive material; (2) claims arising out of an act of war;, and (3) claims for loss of, or damage to, or lossof use of (a) property which is located at the location and used in connection with the licensee's possession, use, or transfer of the radioactive

material, and (b) if the nuclear incident occurs Inthe course of transportation of the radioactive

material, the transporting

vehicle, containers used in such transportation, and the radioactive material.

7. The location means the location described in Item 3 of the Attachment hereto.8. The radioactive material means source, special nuclear, and byproduct material which (1)is used or to be used in, or is-irradiated or to be irradiated byl the nuclear reactor or reactorssubject to the license or licenses designated in the Attachment hereto, or (2) which is producedas the result of operation of said reactor(s).

9. United States when used in a geographical sense includes Puerto Rico and all territories and possessions of the united States.Article II1. Any obligations of the licensee under subsection 53e(8.).

of the Act to indemnify the UnitedStates and the Commission from public liability shall not in the aggregate exceed $250,000 withrespe.ct to any nuclear incident.

2. With respect to any extraordinary nuclear occurrence to which this agreement

applies, the,Commission, and the licensee on behalf of itself and other persons indemnified, insofar as theirinterests appear, each agree to waive:(a) Any issue or defense as to the conduct of the claimant or fault of persons indemnified, including, but not limited to (1) Negligence; (2) Contributory negligence; (3) Assumption of the risk;(4) Unforeseeable intervening causes, whether involving the conduct of a third or anact of God.

As used herein, conduct of the claimant includes conduct of persons through whom the claimantderives his cause of action; (b) Any issue or defense as to charitable or governmental immunity: (c) Any Issue or defense based on any statute of limitations if suit is instituted within 3 yearsfrom the date on which the claimant first knew, or reasonably could have known, of his injury ordamage and the cause thereof.*The waiver of any such issue or defense shall be effective regardless of whether such issueor defense may otherwise be deemed jurisdictional or relating to an element in the cause ofaction. The waivers shall be judicially enforceable In accordance with their te.r~ms by the claimantagaInst the person indemnified.

3. The waivers set forth in paragraph 2 of this article:

(a) Shall not preclude a defense basedupon a failure to take reasonable steps to mitigate damages;(b) Shall not apply to injury or damage to a claimant or to a claimant's property which is*intentionally sustained by the claimant or which results from a nuclear incident intentionally andwrongfully caused by the claimant; (c) Shall not apply to injury to a claimant who is employed at the site of and in connection withthe activity where the extraordinary nuclear occurrence takes place if benefits therefor are eitherpayable or required to be provided under any workmen~s compensationi or occupational diseaselaw: Provided,

however, That with respect to an extraordinary nuclear occurrence occurring atthe facility, a claimant who is employed at the facility In connection with the construction of anuclear reactor with respect to which no operating license has been issued by the NuclearRegulatory Commission shall not be considered as employed in connection with the activitywhere the extraordinary nuclear occurrence takes place if:(1) The claimant is employed exclusively in connection with the construction of a nuclearreactor, including all related equipment and installations at the facility, and(2) No operating license has been issued by the NRC with respect to the nuclear reactor, and(3) The claimant is not employed in connection with the possession,

storage, use, or transferof nuclear material at the facility; (d) Shall not apply to anty claim for punitive or exemplary damages.

provided, with respect toany claim for wrongful death under any State law which provides for damages only punitive innature, this exclusion does not apply to the extent that the claimant has sustained actualdamages, measured by the pecuniary injuries resulting from such death but not to exceed themaximum amount otherwise recoverable under such law;(e) Shall be effective only with respect to those obligations set forth in this agreement; (t') Shall not apply to, or prejudice the prosecution or defense of, anty claim or portion of claimwhich is not within the protection afforded under (1) the limit of liability provisions undersubsection 170(e) of the Atomic Energy Act of 1954, as amended, and (b) the terms of thisagreement.

Article Ill1. The Commission undertakes and agrees to Indemnify and hold harmless the licensee andother persons indemnified, as their interest may appear,.from public Bability,

2. With respect to damage caused by a nuclear Incident to property of any person legallyliable for the nuclear incident, the Commission agrees to pay to such person those sums whichsuch person would have been obligated to pay if such property had belonged to another;provided, that the obligation of the Commission under this paragraph 2 does not apply withrespect to: (a) Property which is located at the location and used in connection with thelicensee's possession, use or transfer of the radioactive material; FEB. j..2000 5:53PM NO. .957 P.4/s(b) Property damage due to the neglect of the. person indemnified to use all reasonable means to save and preserve the property after knowledge of a nuclear Incident:,

(C) If the nuclear incident occurs in the course of transportation of the radioactive

material, thetransporting vehicle and containers used-In such transportation; (d) The radioactive material.

3. (Reserved]

4. (a) The obligations of the Commission under this agreement shall apply only with respect tosuch public liability and such damage to property of persons legally liable for the nuclear Incident(other than such property described in the proviso to paragraph 2 of this Article) as in theaggregate exceed $250,000.

(b) With respect to a common occurrence, the obligations of the Commission under this.:agreement shall apply only with respect to such public liability and such damage to property ofpersons legally liable for the nuclear Incident (other than such property described in theproviso to paragraph 2 of this Article) as in the aggregate exceed whichever of the following islower: (1) The sum of the amounts of financial protection established under all applicable agreements: or (2) an amount equal to the sum of $200,000,000 and the amount available assecondary financial protection, As used in this Article applicable agreements means eachagreement entered into by the Commission pursuant to subsection 170(c) or (k) of the Act inwhich agreement the nuclear incident is defined as a common occurrence.

5. The obligations of the Commission under this agreement shall apply only with respect tonuclear incidents occurring during the term of this agreement.

6. The obligations of the Commission Uinder this and all other agreements and contracts towhich the Commission is a party shell not with respect to any nuclear Incident, in the aggregate exceed which ever of the following is the lower. (a) $500,000,000 or (b) With respect to acommon occurrence,

$560,000,000 less the sum of the amounts of financial protection established under all applicable agreements.

7. If the licensee is immune from public liability because It is a state agency, the Commission shall make payments under the agreement in the same manner arnd to the same extent as theCommission would be required to do if the licensee were not such a state agency.8. The obligations of the Commission under this agreement, except to the licensee fordamage to property of the licensee, shall not be affected by any failure on the part of thelicensee to fulfill Its obligations under this agreement.

Bankruptcy or insolvency of tihelicensee or any other person indemnified or of the estate of the licensee or any other personindemnified shall not relieve the Commission of any of its obligations hereunder. Article IV .1. When the Commission determnines that the United States will probably be required to makeindemnity payments under the provisions of this agreement, the Commission shall have the right:to collaborate with the licensee and other persons indemnified in the settlement and defenseof any claim Including such legal costs of the licensee as are approved by the Commission andshall have the right (a) to require the prior approval of the Commission for the settlement orpayment of any claim or action asserted against: the Ilicensee or other person indemnified forpublic liability or damage to property of persons legally liable for the nuclear incident which claimor action the licensee or the Commission may be required to indemnify under this agreement: and (b) to appear through the Attorney General of the United States on behalf of the licensee orother person indemnified, take charge of such action or defend any such action. If the settlement FEB. 1.2B :5P O9 ./or defense of any such action or claim Is undertaken by the Comn'isslon, the licensee shallfurnish all reasonable assistance in effecting a settlement or asserting a defense.2. Neither this agreement nor any interest therein nor claim thereunder may be assigned ortransferred, without the approval of the Commission. Article VThe parties~agree that they will enter into appropriate amendments of this agreement to theextent that such amendments are required pursuant to the Atomic Energy. At of 1954, asamended, or licenses, regulations or orders of the Commission. Article VIThe licensee agrees to pay to the Commission such fees as are established l~y theCommission pursuant to regulations or orders of the Commission,. Article ViiThe term of this agreement shall commence as of the date and time specified in Item 4 of theAttachment and shall terminate at the time of expiration of that license specified in Item 2 of theAttachment, which is the last to expire; provided that, except as may otherwise be provided inapplicable regulations or orders of the Commission, the term of this agreement shall notterminate until all the radioactive material has been removed from the location andtransportation of the radioactive material from the location has ended as defined insubparagraph 4(b), Article I, Termination of the term of this agreement shall not affectany obligation of the licensee or any obligation of the Commission under this agreement withrespect to any nuclear incident occurring during the term of this agreement. 4g. FEB 1200 554p 9NO.957 P.6/6Item 1-Address-- Item 2-Item 3-Item 4-..Attachment to Indemnity Agreement No. E-40LicenseeUniversity of California, DavisOne Shields Avenue, Davis, California 9561648558 License number or numbersR-130LocationThe reactor is located in the McClellan Nuclear Radiation Center Buildingon McClellan AFB, located approximately 8 miles northeast ofSacramento, California. The indemnity agreement designated above, of which this Attachment Isa part of, is effective on the day of , 2000,For the United States Nuclear Regulatory Commission, Cyhit,o,Che Generic Issues, Environmental, Financial, and Rulemaking BranchDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation Dated at Rock'ville, MD, the day of ,2000._________________By Kevin SmithVice C~hancelior University of California, Davis Fz~? 0UNITED STATES%" NUCLEAR REGULATORY COMMISSION / WASHINGTON, D.C. 20555-0001 9, 2001Dr. Kevin Smith, Vice Chancellor Office of the Chancellor University of California, DavisOne Shields AvenueDavis, CA 95616-8558

SUBJECT:

ISSUANCE OF AMENDMENT NO. 4 TO AMENDED FACILITY OPERATING LICENSE NO. R-130 -REGENTS OF THE UNIVERSITY OF CALIFORNIA (TAC NO. 8391)

Dear Dr. Smith:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 4 to FacilityOperating License No. R-130 for the McClellan Nuclear Radiation Center (MNRC) TRIGAResearch Reactor. The amendment consists of changes to the Technical Specifications (TSs)in response to your submittal of May 11, 2001.The amendment reflects the administrative changes to the TSs as a result of the transfer of thelicense from the Department of the Air Force to the Regents of the University of California. There are other, non-administrative

changes, which are also reflected in this amendment andwhich are discussed in the enclosed safety evaluation report.Sincerely, Warren J. Eresian, Project ManagerOperational Experience and Non-Power Reactors BranchDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation Docket No. 50-607

Enclosures:

1. Amendment No. 42. Safety Evaluation cc w/enclosures:

Please see next page University of California -Davis/McClellan MNRC Docket No. 50-607co:Dr. Wade J. Richards5335 Price Avenue, Bldg. 258McClellan AFB, CA 95652-2504 Test, Research, and TrainingReactor Newsletter University of Florida202 Nuclear Sciences CenterGainesville, FL 32611 - STATESNUCLEAR REGULATORY COMMISSIONWASHINGTON, D.C. 20555-0001 REGENTS OF THE UNIVERSITY OF CALIFORNIA ATMcCLELLAN NUCLEAR RADIATION CENTERDOCKET NO. 50-607AMENDMENT TO AMENDED FACILITY OPERATING LICENSEAmendment No. 4License No. R-1301. The U.S. Nuclear Regulatory Commission (the Commission) has found thatA. The application for an amendment to Amended Facility Operating License No.R-1 30 filed by the Regents of the University of California at McClellan NuclearRadiation Center (the licensee) on May 11, 2001, conforms to the standlards andrequirements of the Atomic Energy Act of 1954, as amended (the Act), and theregulations of the Commission as stated in Chapter I of Title 10 of the Code ofFederal Regulations (10 CFR);B. The facility will operate in conformity with the application, the provisions of the Act,and the rules and regulations of the Commission; C. There is reasonable assurance that (i) the activities authorized by this amendment can be conducted without endangering the health and safety of the public and (ii)such activities will be conducted in compliance with the regulations of theCommission; D. The issuance of this amendment will not be inimical to the common defense andsecurity or to the health and safety of the public;E. This amendment is issued in accordance with the regulations of the Commission asstated in 10 CFR Part 51, and all applicable requirements have been satisfied; andF. Prior notice of this amendment was not required by 10 CFR 2.105, and publication of notice for this amendment is not required by 10 CFR2.106. 2. Accordingly, the license is amended by changes to the Technical Specifications asindicated in the enclosure to this license amendment, and paragraph 2.C.(ii) of AmendedFacility Operating License No. R-130 is hereby amended to read as follows:2.c.(ii) Technical Specifications The Technical Specifications contained in Appendix A, as revised throughAmendment No. 4, are hereby incorporated in the license. The licensee shalloperate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Warren J. Eresian, Project ManagerOperational Experience and Non-Power Reactors BranchDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation

Enclosure:

Appendix A, Technical Specification ChangesDate of Issuance: August 9, 2001 S 0ENCLOSURE TO LICENSE AMENDMENT NO. 4AMENDED FACILITY OPERATING LICENSE NO. R-130DOCKET NO. 50-607Replace the following pages of Appendix A, Technical Specifications, with the enclosed pages. Therevised pages are identified by amendment number and contain vertical lines indicating the areas ofchange.Remove Insertii iiiii iiiiv ivV vvi vi1 I2 23 34 46 67 79 913 1314 1415 1516 1617 1718 1819 1925 2526 2627 2728 2829 2930 3031 3132 3233 3334 3435 3536 3639 3940 40 UNITED STATES1"%" NUCLEAR REGULATORY COMMISSION ~WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 4 TOAMENDED FACILITY OPERATING LICENSE NO. R-130REGENTS OF THE UNIVERSITY OF CALIFORNIA ATMcCLELLAN NUCLEAR RADIATION CENTERDOCKET NO. 50-60

71.0 INTRODUCTION

By letter dated May 11, 2001, the Regents of the University of California (the licensee) submitted a request for amendment of the Technical Specifications (TSs), Appendix A, toFacility Operating License No. R-1 30 for the McClellan Nuclear Radiation Center (MNRC)TRIGA research reactor. (On July 9, 2001, the licensee resubmitted the amendment requestunder oath. The resubmittal contained no new information.) The request provides for thefollowing

changes, which if implemented, will result in Revision 11 of the TSs:1, On February 1, 2000, the operating license for MNRC was transferred from theDepartment of the Air Force to the Regents of the University of California.

As a result ofthis transfer, a nUmber of administrative changes simply involving name changes (e.g.,changing references from "Responsible Commander" to "Vice Chancellor of the Office ofResearch" and "Air Force" to "University of California-Davis," etc.) is necessary

2. Section 2.1, Basis b. This section has been expanded to include more detail regarding cladding integrity during pulsing operation.

3. Section 3.3, Table 3.3. A request to increase the alarm setpoint for the heat exchanger outlet temperature from 35 degrees Centigrade to 45 degrees Centigrade.

4. Section 4.7, Specification 4.7.a(3),

4.7.b(3) and 4.7.d(3). A request to allow channelcalibrations to be performed annually rather than semiannually.

5. Section 5.3.1. A request to add the use of 30/20 TRIGA fuel and a new core fuel loadingtermed a 30B core.6. Section 6.0. A request to revise the organization and duties of the Nuclear SafetyCommittee and to clarify the Committee's review and audit functions to reflect the newlicensee. 7. A request for approval of a new Iodine-125 production loop.8. Section 3.8.2. Clarifies reactivity limits for experiments, and adds a new paragraph pertaining to the Iodine-I125 production facility.

2.0 EVALUATION

The staff has considered each of the items 1-8 above. Each item is discussed below.2.1 Administrative changes.As a result of the February 1, 2000, transfer of the Operating License from the Department ofthe Air Force to the Regents of the University of California, the TSs must be modified to takeaccount of administrative changes. These changes will occur in a number of places, andconsist of the substitution of Department of the Air Force organizational and position titles withcorresponding University of California titles. The substitutions are made on a one-for-one basis.These changes are also reflected in Figure 6.1, "UCD/MNRC Organization for Licensing andOperation." The staff concludes that there has been no diminishment of licensee oversight (i.e.,the lines of authority and responsibility have not been weakened) and that these changes areacceptable. 2.2 Section 2.1, Basis b.The previous version of the Technical Specifications addressed the issue of the effect of pulsingon fuel clad integrity and concluded that TRIGA fuel of the type used in the McClellan reactorcould be pulsed up to temperatures of 1150 degrees Centigrade without damage to the clad,provided that the clad temperature was less than 500 degrees Centigrade. The presentanalysis expands the discussion to include more recent measurements of hydrogen pressureresulting from pulses and concludes that the cladding will not rupture if fuel temperatures arenever greater than 1200 to 1250 degrees Centigrade, providing the cladding temperature is lessthan 500 degrees Centigrade. Since the pulse reactivity limit remains at $1.75, the staffconcludes that the bases for Section 2.1 are more conservative and this is acceptable. 2.3 Section 3.3, Table 3.3.A re-evaluation of the thermal and hydraulic analyses and operating limits was performed byResearch Reactor Safety Analysis Services (RRSAS-99-6-1, December 1999) to determine ifthe conservative maximum core inlet temperature (heat exchanger outlet temperature) as set bythe U.S. Air Force in the original design could be raised from 35 degrees Centigrade to 45degrees Centigrade. The effect of the lower limit is that the reactor power is required to bereduced below the license limit of 2 MW whenever ambient local weather conditions prevent thesystem from maintaining the heat exchanger outlet temperature at or below the lower limit.Evaluation of data during 2 MW startup tests as well as data from subsequent steady stateoperations, when compared with previous calculations by Argonne National Laboratory, GeneralAtomics published

reports, and results from power upgrades at the Sandia Annular Core 0-3-Research Reactor facility shows that the maximum core inlet temperature can be raised to45 degrees Centigrade with only a small reduction in Critical Heat Flux ratio (from 2.53 to 2.40).These numbers have been also confirmed by RELAP5 thermal hydraulic calculations.

Thecalculations also show that there is no increase in the maximum fuel temperature or themaximum fuel clad surface temperature, two of the most important parameters which measurefuel integrity. Accordingly, the staff concludes that safety limits will not be reduced and thatthere is no reduction in safety margin.2.4 Section 4.7, Specification 4.7.a(3), 4.7.b(3) and 4.7.d(3). This section of the Technical Specifications addresses channel calibration frequencies for thestack monitor system, the reactor room radiation monitor and the reactor room continuous airmonitor. These systems are presently required to be calibrated semiannually. The licensee hasrequested that they be calibrated annually. The requirement for semiannual calibrations stems from the original Department of the Air Forcelicensing organization, but has no operational safety basis. Research reactors of similar powerlevels currently licensed by the NRC (National Institute of Standards and Technology, RhodeIsland AEC) are permitted to calibrate similar instruments on an annual basis, since there areno operating experience data to suggest that this practice would compromise safety. Inaddition, the American National Standard ANSI/ANS 15.11 "Radiation Protection at ResearchReactor Facilities," states that "Instruments shall be tested at least annually in a performance quality assurance program [i.e., calibration], or more frequently if subject to extreme conditions." The facility is not subject to extreme conditions, and the staff concludes that annual calibrations are acceptable. 2.5 Section 5.3.1.When the McClellan reacto'r was originally licensed by the NRC (August 1998), the reactor wasoperating with a mixed core of 8.5/20 and 20/20 fuel loading (referred to as the MixJ core in theoriginal SAR). At that time it was understood that the reactor would eventually transition to acore consisting of 20/20 and 30/20 fuel, termed a 30B core. The 30B core was analyzed in theoriginal SAR and found to be acceptable by the NRC staff in the SER. In addition, the NRCstaff had previously approved the generic use of TRIGA fuels with uranium loadings of up to30 wt% in licensed TRIGA reactors (NUREG-1282.) The staff concludes that the introduction of 30/20 fuel is consistent with previous analyses and does not create any additional hazards.2.6 Section 6.0.Section 6.0 of the Technical Specifications describes the administrative controls governing theoperation and maintenance of the reactor and associated equipment. There are a number ofminor changes with respect to titles and some changes with .respect to the composition andduties of the Nuclear Safety Committee (NSC). The review and inspection functions of the NSChave been expanded to provide additional oversight. These expanded functions include reviewof the Emergency Plan and Physical Security Plan, review and update of the NSC Charter everytwo years, review of inspections conducted by other agencies, assessment of actions taken tocorrect deficiencies, inspection of currently active experiments, and inspection of future plansfor facility modifications or facility utilization. Since these changes increase oversight of facilityoperations, the staff concludes that they are acceptable. 0-4-2.7 A request for approval of a new Iodine-I125 production loop.The licensee has requested amendment of the Safety Analysis Report to provide for theinstallation of an Iodine-125 production loop. The purpose of the loop is to produce from ten totwenty curies of lodine-I25 for use as a medical radioisotope. The production of Iodine-I25 occurs in five steps:I. Xenon-124 is transferred from a storage tank into an irradiation chamber located in thereactor core.2. The Xenon-I 24 is irradiated over an eight to sixteen hour time span and by neutronactivation results in the production of Xenon-125. The activated Xenon-I124 gascontains up to 4,000 curies of Xenon-125.

3. The Xenon-125 is transferred to a tank, referred to as decay storage I, where it decayswith a 17-hour half-life to Iodine-I125.

After a few days, most of the Xenon-I125 hasdecayed and the Iodine-125 plates out in the tank.4. The Xenon-I 25 remaining in decay storage I is transferred to another tank, referred toas decay storage 2.5. The Iodine-125 in decay storage I is recovered by washing the tank with a NaOHsolution, resulting in a Nal solution which is packaged as a liquid and sent to an off-siteuser in an appropriate DOT container. All equipment used in the production loop is located within a primary containment and asecondary containment. The primary containment houses the irradiation

chamber, tubing,pneumatically operated valves, transfer vessel, decay storage I and decay storage 2. Thesecondary containment is placed around the primary containment to the irradiation chamber andallows for recovering the xenon gas if a leak occurs within the primary containment.

Shielding around the secondary containment reduces radiation levels to below 10 mrem/hr. Both of thesecontainments are within the reactor room, which has a ventilation system withisolation/recirculation capability. There are two other structures within the reactor room which are confinement barriers designedfor the safety of personnel working with the production loop. The first is a glove box whichcontains controls for operation of the Iodine-125 recovery system. The glove box has its ownventilation and filtration system which exhausts into the reactor room ventilation system. Thesecond is a fume hood in which quality assurance of the Iodine-125 is performed. The fumehood also contains its own ventilation and filtration system which exhausts into the reactor roomventilation system.The licensee has analyzed the situation (worst-case) whereby all of the Xenon-I125 from theprimary containment leaks into the secondary containment and subsequently leaks into thereactor room at the design leak rate of the secondary containment. Their analysis shows thatexposures to personnel in the reactor room would result in a deep dose equivalent (DDE) of 17millirem after one hour, about 1.4 millirem for a five-minute occupancy, and about 0.6 millirem for a two-minute occupancy, all well within 10 CFR 20 limits. Exposures to personnel located atthe boundary of the unrestricted area for a full year would be approximately 7 millirem. The Maximum Hypothetical Accident analyzed in the Safety Analysis Report (SAR) is a claddingrupture of one highly irradiated fuel element with no decay followed by instantaneous release offission products into the air. At the closest distance to the site boundary (10 meters), themaximum dose to a member of the general public is 66 millirem, received over an approximately 10-minute period. The dose received at the same location due to a failure of the Iodine-125 production loop is approximately 7 millirem over a period of one year.The staff concludes that the installation of the Iodine-I125 production loop does not reduce themargin of safety with respect to 10 CFR 20 limits and that the installation of the production loopis acceptable. 2.8 Section 3.8.2.This section of the Technical Specifications has been expanded to take account of the Iodine-125 production loop. Sections 3.8.2.c and 3.8.2.d have been added to limit the amount ofIodine-125 present in the reactor room glove box and the reactor room fume hood. Limiting theamount of Iodine-125 in these areas will reduce the occupational dose and dose to personnel inthe unrestricted areas to less than 10 CFR 20 limits if the inventories of Iodine-I125 are totallyreleased within the glove box and fume hood. The staff concludes that this is acceptable. 3.0 ENVIRONMENTAL CONSIDERATION This amendment involves changes in the installation or use of a facility component locatedwithin the restricted area as defined in 10 CFR Part 20 or changes in inspection andsurveillance requirements. The staff has determined that this amendment involves nosignificant increase in the amounts, and no significant change in the types, of any effluents thatmay be released off site, and no significant increase in individual or cumulative occupational radiation exposure. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CER 51 .22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared with the issuance of thisamendment.

4.0 CONCLUSION

The staff has concluded, on the basis of the considerations discussed above, that (1) becausethe amendment does not involve a significant increase in the probability or consequences ofaccidents previously evaluated, or create the possibility of a new or different kind of accidentfrom any accident previously evaluated, and does not involve a significant reduction in a marginof safety, the amendment does not involve a significant hazards consideration; (2) there isreasonable assurance that the health and safety of the public will not be endangered by theproposed changes; and (3) such changes are in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security orthe health and safety of the public.Principal Contributor: Warren J. EresianDate: August 9, 2001 0I 0TECHNICAL SPECIFICATIONS FOR THEUNIVERSITY OF CALIFORNIA -DAVIS MCCLELLAN NUCLEAR RADIATION CENTER(UCD/MNRC) DOCUMENT NUMBER: MNRC-0004-DOC-1 1Rev 11, 12/10/99Amendment No. 4i 0TECHNICAL SPECIFICATIONS APPROVALThese "Technical Specifications" for the University of California at Davis/McClellan Nuclear Radiation Center(UCD/MNRC) Reactor have undergone the following coordination: Reviewed z 'ODteReviewed by: k_,. Q- Reactor Operations S~pervisor Approved by: ,U49UCD/MNI DirectorApproved by:________________

Chairman, UCD/MNRCNuclear Safety Committee (Date)(Date)(Date)Amendment No. 4ii 0TECHNICAL SPECIFICATIONS TABLE OF CONTENTSPage1.0 Definitions...................................................................................................................1 2.0 Safety Limit and Limiting Safety System Setting (LSSS)...............................................................

52.1 Safety 52.2 Limiting Safety System Selling (LSSS)......................................................................... 62.2.1 Fuel Temperature .................................................................................... 63.0 Limiting Conditions for Operations (LCO) ............................................................................... 73.1 Reactor Core Parameters ....................................................................................... 73.1.1 Steady-State Operation.................................................. 73.1.2 Pulse or Square Wave Operation................................................................... 73.1.3 Reactivity Limitations................................................................................. 83.2 Reactor Control and Safety Systems........................................................................... 83.2.1 Control Rods........................................................................................... 83.2.2 Reactor Instrumentation ............................................................................. 93.2.3 Reactor, Scrams and Interlocks .................................................................... 103.2.4 Reactor Fuel Elements ............................................................................. 123.3 Reactor Coolant Systems...................................................................................... 133.4 Reactor Room Exhaust System ............................................................................... 14.3.5 Intentionally Left Blank .................................... ..................................................... 143.6 Intentionally Left Blank ......................................................................................... 143.7 Reactor Radiation Monitoring Systems ....................................................................... 143.7.1 Monitoring Systems................................................................................. 143.7.2 Effluents -Argon-41 Discharge Limit .............................................................. 16Amendment No. 4iii Pane3.8 Experiments ..................................................................................................... 163.8.1 Reactivity Limits ..................................................................................... 163.8.2 Materials Limit ....................................................................................... 173.8.3 Failure and Malfunctions ........................................................................... 184.0 Surveillance Requirements .................................................................. ........................... 194.1 Reactor Core Parameters...................................................................................... 194.1.1 Steady State Operation ............................................................................ 194.1.2 Shutdown Margin and Excess Reactivity.......................................................... 204.2 Reactor Control and Safety Systems ......................................................................... 204.2.1 Control Rods......................................................................................... 204.2.2 Reactor Instrumentation ............................................................................ 214.2.3 Reactor Scrams and Interlocks .................................................................... 224.2.4 Reactor Fuel Elements ............................................................................. 234.3 Reactor Coolant Systems...................................................................................... 244.4 Reactor Room Exhaust Systerm........ ....................................................................... 254.5 Intentionally Left Blank ......................................................................................... 254.6 Intentionally Left Blank ......................................................................................... 254.7 Reactor Radiation Monitoring Systems ....................................................................... 254.8 Experiments ..................................................................................................... 265.0 Design Features.......................................................................................................... 275.1 Site and Facility Description.................................................................................... 275.1.1 Site.................................................................................................... 275.1.2 Facility Exhaust...................................................................................... 285.2 Reactor Coolant System........................................................................................ 28Amendment No. 4iv 0 0Page5.3 Reactor Core and Fuel ........................................................................................... 295.3.1 Reactor Core........................................................................................... 295.3.2 Reactor FueL........................................................................................... 305.3.3 Control Rods and Control Rod Drives .............................................................. 315.4 Fissionable Material Storage .................................................................................... 316.0 Administrative Controls..................................................................................................... 316.1.1 Structure................................................................................................ 326.1.2 Responsibilities........................................................................................ 326.1.3 Staffing.................................................................................................. 326.1.4 Selection and Training of Personnel ................................................................ 326.2 Review, Audit, Recommendation and Approval............................................................... 326.2.1 NSC Composition and Qualifications............................................................... 336.2.2 NSC Charter and Rules` ............................................................................. 336.2.3 Review Functiont...................................................................................... 336.2.4 Audit/Inspection Function ............................................................................ 346.3 Radiation Safety. ............................................. ..................................................... 346.4 Procedures ........................................................................................................ 346.4.1 Reactor Operations Procedures........ .................. ........................................... 346.4.2 Health Physics Procedures .......................................................................... 356.5 Experiment Review and Approval............................................................................... 356.6 Required Actions.................................................................................................. 356.6.1 Actions to be taken in case of a safety limit violation.............................................. 356.6.2 Actions to be taken for reportable occurrences`................................................... 36Amendment No. 4V 6.77Rep Ortstn R po t ..................................................... ........366.7.2 Special Reports........................................................................................ 386.8 Records ............................................................................................................. 39Fig. 6.1 UCD/MNRC Organization for Licensing and 40Amendment No. 4vi 0TECHNICAL SPECIFICATIONS FOR THEUNIVERSITY OF CALIFORNIA -DAVIS/MCCLELLAN NUCLEAR RADIATION CENTER (UCD/MNRC) GeneralThe University of California -Davis/McClellan Nuclear Radiation Center (UCD/MNRC) reactor is operated by theUniversity of California, Davis, California (UCD). The UCD/MNRC research reactor is a TRIGA-type reactor. TheUCD/MNRC provides state-of-the-art neutron radiography capabilities. In addition, the UCD/MNRC provides a widerange of irradiation services for both research and industrial needs. The reactor operates at a nominal steady statepower level up to and including 2 MW. The UCD/MNRC reactor is also capable of square wave and pulseoperational modes. The UCD/MNRC reactor fuel is less than 20% enriched in uranium-235.

1.0 Definitions

1.1 As Low As Reasonably Achievable (ALARA). As defined in 10 CFR, Part 20.1.2 Licensed Operators. A UCD/MNRC licensed operator is an individual licensed by the NuclearRegulatory Commission (e.g., senior reactor operator or reactor operator) to carry out the duties andresponsibilities associated with the position requiring the license.1.2.1 Senior Reactor Operator. An individual who is licensed to direct the activities of reactoroperators and to maniPulate the controls of the facility. 1.2.2 Reactor Operator. An individual who is licensed to manipulate the controls of the facility andperform reactor-related maintenance. 1.3 Channel. A channel is the combination of sensor, line amplifier, processor, and output devices whichare connected for the purpose of measuring the value of a parameter. 1.3.1 Channel Test. A channel test is the introduction of a signal into the channel for verification that it is operable. 1.3.2 Channel Calibration. A channel calibration is an adjustment of the channel such that itsoutput corresponds with acceptable accuracy to known values of the parameter which the channelmeasures. Calibration shall encompass the entire channel, including equipment actuation, alarmor trip, and shall be deemed to include a channel test.1.3.3 Channel Check. A channel check is a qualitative verification of acceptable performance byobservation of channel behavior. This verification, where possible, shall include comparison of thechannel with other independent channels or systems measuring the same variable. 1.4 Confinement. Confinement means isolation of the reactor room air volume such that the movement ofair into and out of the reactor room is through a controlled path.1.5 Experiment. Any operation,

hardware, or target (excluding devices such as detectors, fissionchambers, foils, etc), which is designed to investigate specific reactor characteristics or which is intendedfor irradiation within an experiment facility and which is not rigidly secured to a core or shield structure soas to be a part of their design.1.5.1 Experim~ent.

Moveable. A moveable experiment is one where it is intended that the entireexperiment may be moved in or near the reactor core or into and out of reactor experiment facilities while the reactor is operating. Amendment No. 41

• 01.5.2 Experiment.

Secured. A secured experiment is any experiment, experiment

facility, orcomponent of an experiment that is held in a stationary position relative to the reactor bymechanical means. The restraining force rmust be substantially greater than those to which theexperiment might be subjected by hydraulic, pneumatic,

buoyant, or other forces which are normalto the operating environment of the experiment, or by forces which can arise as a result of credibleconditions.

1.5.3 Experiment Facilities. Experiment facilities shall mean the pneumatic transfer tube,beamtubes, irradiation facilities in the reactor core or in the reactor tank, and radiography bays..1.5.4 Experiment Safety System. Experiment safety systems are those systems, including theirassociated input circuits, which are designed to initiate a scram for the primary purpose ofprotecting an experiment or to provide information which requires manual protective action to beinitiated. 1.6 Fuel Element, Standard. A fuel element is a single TRIGA element. The fuel is U-ZrH clad in stainless steel. The zirconium to hydrogen ratio is nominally 1.65 +/- 0.05. The weight percent (wt%) of uranium canbe either 8.5, 20, or 30 wt%, with an enrichment of less than 20% U-235. A standard fuel element maycontain a burnable poison.1.7 Fuel Element. Instrumented. An instrumented fuel element is a standard fuel element fabricate~d withthermocouples for temperature measurements. An instrumented fuel element shall have at least oneoperable thermocouple embedded in the fuel near the axial and radial midpoints. 1.8 Measured Value. The measured value is the value of a parameter as it appears on the output of achannel.1.9 Mode, Steady-State. Steady-state mode operation shall mean operation of the UCD/MNRC reactor withthe selector switch in the automatic or manual mode position. 1.10 Mode, Square-Wave. Square-wave mode operation shall mean operation of the UCD/MNRC reactorwith the selector switch in the square-wave mode position. 1.11 Mode, Pulse. Pulse mode operation shall mean operation of the UCD/MNRC reactor with the selectorswitch in the pulse mode position. 1.12 Operable. Operable means a component or system is capable of performing its intended function. 1.13 Operatingq. Operating means a component or system is performing its intended function. 1.14 Operatinq Cycle. The period of time starting with reactor startup and ending with reactor shutdown. 1.15 Protective Action. Protective action is the initiation of a signal or the operation of equipment within theUCD/MNRC reactor safety system in response to a variable or condition of the UCD/MNRC reactor facilityhaving reached a specified limit.1.15.1 Channel Level. At the protective instrument channel level, protective action is the generation and transmission of a scram signal indicating that a reactor variable has reached the specified limit.1.15.2 Subsystem Level. At the protective instrument subsystem level, protective action is thegeneration and transmission of a scram signal indicating that a specified limit has been reached.NOTE: Protective action at this level would lead to the operation of the safety shutdown equipment. Amendment No. 42 1.15.3 Instrument System Level. At the protective instrument level, protective action is thegeneration and transmission of the command signal for the safety shutdown equipment to operate.1.15.4 Safety System Level. At the reactor safety system level, protective action is the operation ofsufficient equipment to immediately shut down the reactor.1.16 Pulse QOerational Core. A pulse operational core is a reactor operational core for which the maximumallowable pulse reactivity insertion has been determined. 1.17 .Reactivity, Excess. Excess reactivity is that amount of reactivity that would exist if all control rods(control, regulating, etc.) were moved to the maximum reactive position from the point where the reactor isat ambient temperature and the reactor is critical. = 1)1.18 Reactivity Limits. The reactivity limits are those limits imposed on the reactivity conditions of the reactorcore.1.19 Reactivity Worth of an Experiment. The reactivity worth of an experiment is the maximum value of thereactivity change that could occur as a result of changes that alter experiment position or configuration. 1.20 Reactor Controls. Reactor controls are apparatus and/or mechanisms the manipulation of whichdirectly affect the reactivity or power level of the reactor.1.21 .Reactor Core. Operational. The UCD/MNRC reactor operational core is a core for which theparameters of excess reactivity, shutdown margin, fuel temperature, power calibration and reactivity worthsof control rods and experiments have been determined to satisfy the requirements set forth in theseTechnical Specifications. 1.22 _Reactor Operatingq. The UCD/MNRC reactor is operating whenever it is not shutdown or secured.1.23 Reactor Safety Systems. Reactor safety systems are those systems, including their associated inputchannels, which are designed to initiate automatic reactor protection or to provide information for initiation of manual protective action.1.24 _Reactor Secured. The UCD/MNRC reactor is secured when the console key switch is in the offposition and the key.is removed from the lock and under the control of a licensed

operator, and theconditions of a or b exist:a. (1) The minimum number of control rods are fully inserted to ensure the reactor is shutdown, as requiredby technical specifications; and(2) No work is in progress involving core fuel, core structure, installed control rods, or control rod drives,unless the control rod drives are physically decoupled from the control rods; and(3) No experiments in any reactor experiment

facility, or in any other way near the reactor, are beingmoved or serviced if the experiments have, on movement, a reactivity worth exceeding the maximum valueallowed for a single experiment or $1.00, whichever is smaller, orb. The reactor contains insufficient fissile materials in the reactor core, adjacent experiments or control rodsto attain criticality under optimum available conditions of moderation and reflection.

1.25 _Reactor Shutdown. The UCD/MNRC reactor is shutdown if it is subcritical by at least one dollar ($1.00)both in the Reference Core Condition and for all alloWed ambient conditions with the reactivity worth of allinstalled experiments included. Amendment No. 43

• 01.26 Reference Gore Condition.

The condition of the core when it is at ambient temperature (cold T<28° C),the reactivity worth of xenon is negligible (< $0.30) (i.e., cold and clean), and the central irradiation facilitycontains the graphite thimble plug and the aluminum thimble plug (CIF-1).1.27 Research Reactor. A research reactor is defined as a device designed to support a self-sustaining neutron chain reaction for research development, education, and training, or experimental

purposes, andwhich may have provisions for the production of radioisotopes.

1'.28 Rod, Control. A control rod is a device fabricated from neutron absorbing

material, with or without afuel or air follower, which is used to establish neutron flux changes and to compensate for routine reactivity losses. The follower may be a stainless steel section.

A control rod shall be coupled to its drive unit to allowit to perform its control function, and its safety function when the coupling is disengaged. This safetyfunction is commonly termed a scram.1.28.1 Regqulatingq Rod. A regulating rod is a Control rod'used to maintain an intended power leveland may be varied manually or by a servo-controller. A regulating rod shall have scram capability. 1.28.2 Standard Rod. The regulating and shim rods are standard control rods.1.28.3 Transient Rod. The transient rod is a control rod that is capable of providing rapid reactivity insertion to produce a pulse or square wave.1.29 Safety Channel. A safety channel is a measuring channel in the reactor safety system.1.30 Safety Limit. Safety limits are limits on important process variables, which are found to be necessary to reasonably protect the integrity of the principal barriers which guard against the uncontrolled release ofradioactivity. 1.31 Scram Time. Scram time is the elapsed time between reaching a limiting safety system set point andthe control rods being fully inserted. 1.32 Scram. External. External scrams may arise from the radiography bay doors, radiography bay ripcords, bay shutter interlocks, and any scrams from an experiment. 1.33 Shall. Should and May. The word "shall" is used to denote a requirement; the word "should" to denotea recommendation; the word "may" to denote permission, neither a requirement nor a recommendation. 1.34 Shutdown Margqin. Shutdown margin shall mean the minimum shutdown reactivity necessary toprovide confidence that the reactor can be made subcritical by means of the control and safety systemstarting from any permissible operating condition with the most reactive rod assumed to be in the mostreactive

position, and once this action has been initiated, the reactor will remain subcritical without furtheroperator action.1.35 Shutdown.

Unscheduled. An unscheduled shutdown is any unplanned shutdown of the UCD/MNRC-reactor caused by actuation of the reactor safety system, operator error, equipment malfunction, or a manualshutdown in response to conditions which could adversely affect safe operation, not including shutdowns which occur during testing or check-out operations. 1.36 Surveillance Activities. In general, two types of surveillance activities are specified: operability checksand tests, and calibrations. Operability checks and tests are generally specified as daily, weekly orquarterly. Calibration times are generally specified as quarterly, semi-annually,

annually, or biennially.

1.37 Surveillance Intervals. Maximum intervals are established to provide operational flexibility and not toreduce frequency. Established frequencies shall be maintained over the long term. The allowable surveillance interval is the interval between a check, test, or calibration, whichever is appropriate to the itemAmendment No. 44 0fuel element temperature. This parameter is well suited as it can be measured directly. A loss in theintegrity of the fuel element cladding could arise if the cladding stress exceeds the ultimate strength of thecladding material. The fuel element cladding stress is a function of the element's internal pressure while theultimate strength of the cladding material is a function of its temperature. The cladding stress is a result ofthe internal pressure due to the presence of air, fission product gasses and hydrogen from thedisassociation of hydrogen and zirconium in the fuel moderator. Hydrogen pressure is the most significant. The magnitude of the pressure is determined by the fuel moderator temperature and the ratio of hydrogento zirconium in the alloy. At a fuel temperature of 930°C for ZrH1.7 fuel, the cladding stress due to theinternal pressure is equal to the ultimate strength of the cladding material at the same temperature (SAR Fig4.18). This is a conservative limit since the temperature of the cladding material is always lower than the fueltemperature. (See SAR Chapter 4, Section 4.5.4.)b. This fuel safety limit applies for conditions in which the cladding temperature is less than 50000. Analysis(SAR Chapter 4, Section 4.5.4.1.1), shows that a maximum temperature for the clad during a pulse whichgives a peak adiabatic fuel temperature of 100000 is estimated to be 470°C. Further analysis (SAR Section4.5.4.1.2), shows that the internal pressure for both Zr.65 (at 115000) and Zr1.7 (at 1100°C) increases to apeak value at about 0.3 sec, at which time the pressure is about one-fifth of the equilibrium value or about400 psi (a stress of 14,700 psi). The yield strength of the cladding at 500°C is about 59,000 psi.Calculations for step increases in power to peak ZrH1.65 fuel temperature greater than 115000, over a 200°C range,show that the time to reach the peak pressure and the fraction of equilibrium pressure value achieved wereapproximately the same as for the 115000 case. Similar results were found for fuel with ZrH1 .7. Measurements ofhydrogen pressure in TRIGA fuel elements during transient operations have been made and compared with theresults of analysis similar to that used to make the above prediction. These measurements indicate that in a pulsewhere the maximum temperature in the fuel was greater than 100000, the pressure (ZrH1.65) was only about 6% ofthe equilibrium value evaluated at the peak temperature. Calculations of the pressure gave values about three timesgreater than the measured values. The analysis gives strong indications that the cladding will not rupture if fueltemperatures are never greater than 120000 to 125000, providing the cladding temperature is less than 5000C. For fuel with ZrH1.7,a conservative safety limit is 110000. As a result, at this safety limit temperature, the classpressure is a factor of 4 lower than would be necessary for cladding failure.2.2 Limiting Safety System Setting.q 2.2.1 Fuel Temperature. Applicability -This specification applies to the protective action for the reactor fuel elementtemperature. Obiective -The objective is to prevent the fuel element temperature safety limit from being reached.Specification -The limiting safety system setting shall be 75000 (operationally this may be set moreconservatively) as measured in an instrumented fuel element. One instrumented element shall belocated in the analyzed peak power location of the reactor operational core.Basis -For steady-state operation of the reactor, the limiting safety system setting is a temperature which, if exceeded, shall cause a reactor scram to be initiated preventing the safety limit from beingexceeded. A setting of 75000 provides a safety margin at the point of the measurement of at least13700 for standard TRIGA fuel elements in any condition of operation. A part of the safety marginis used to account for the difference between the true and measured temperatures resulting fromthe actual location of the thermocouple. If the thermocouple element is located in the hottestposition in the core, the difference between the true and measured temperatures will be only a fewdegrees since the thermocouple junction is near the center and mid-plane of the fuel element. Forpulse operation of the reactor, the same limiting safety system setting applies.

However, thetemperature channel will have no effect on limiting

the peak power generated because of itsAmendment No. 46 0 !relatively long time constant (seconds) as compared with the width of the pulse (milliseconds).

Inthis mode, however, the temperature trip will act to limit the energy release after the pulse if thetransient rod should not reinsert and the fuel temperature continues to increase. 3.0 Limiting Conditions For Operation 3.1 Reactor Core Parameters 3.1.1 .Steady-State Operation Applicability -This specification applies to the maximum reactor power attained during steady-state operation. Objective -The objective is to assure that the reactor safety limit (fuel temperature) is not exceeded, and to provide for a setpoint for the high flux limiting safety systems, so that automatic protective action will prevent the safety limit from being reached during steady-state operation. Specification -The nominal reactor steady-state power shall not exceed 2.0 MW. The automatic scram setpoints for the reactor power level safety channels shall be set at 2.2 MW or less. For thepurpose of testing the reactor steady-state power level scram, the power shall not exceed 2.3 MW.Basis -Operational experience and thermal-hydraulic calculations demonstrate that UCD/MNRCTRIGA fuel elements may be safely operated at power levels up to 2.3 MW with natural convection cooling. (SAR Chapter 4, Section 4.6.2.)3.1.2 Pulse or Square Wave Operation Applicability -This specification applies to the peak temperature generated in the fuel as the resultof a step insertion of reactivity. Obiective -The objective is to assure that the fuel temperature safety limit will not be exceeded. Specification -a. For the pulse mode of operation, the maximum insertion of reactivity shall be 1.23% Ak/k ($1.75);b. For the square wave mode of operation, the maximum insertion of reactivity shall be 0.63% Ak/k($0.90).Basis -Standard TRIGA fuel is fabricated with a nominal hydrogen to zirconium ratio of 1.6 to 1.7.This yields delta phase zirconium hydride which has a high creep strength and undergoes no phasechanges at temperatures in excess of 1000C. However, after extensive steady state operation attwo (2) MW the hydrogen will redistribute due to migration from the central high temperature regionsof the fuel to the cooler outer regions. When the fuel is pulsed, the instantaneous temperature distribution is such that the highest values occur at the radial edge of the fuel. The highertemperatures in the outer regions occur in fuel with a hydrogen to zirconium ratio that has nowincreased above the nominal value. This produces hydrogen gas pressures considerably in excessof that expected. If the pulse insertion is such that the temperature of the fuel exceeds about 8750C,then the pressure may be sufficient to cause expansion of microscopic holes in the fuel that growwith each pulse. Analysis (SAR Chapter 13, Section 13.2.2.2.1), shows that the limiting pulse, forthe worst case conditions, is 1.34% Ak/k ($1.92). Therefore, the 1.23% Ak/k ($1.75) limit is belowthe worse case reactivity insertion accident limit. The $0.90 square wave step insertion limit is alsowell below the worse case reactivity insertion accident limit.Amendment No. 47 0Basis -a. The apparent condition of the control rod assemblies shall provide assurance that the rods shallcontinue to perform reliably as designed.

b. This assures that the reactor shall shut down promptly when a scram signal is initiated (SARChapter 13, Section 13.2.2.2.2).

3.2.2 Reactor Instrumentation Applicability -This specification applies to the information which shall be available to the reactoroperator during reactor operations. Objective -The objective is to require that sufficient information is available to the operator to assuresafe operation of the reactor.Specification -The reactor shall not be operated unless the channels described in Table 3.2.2 areoperable and the information is displayed on the reactor console.Table 3.2.2Required Reactor Instrumentation (Minimum Number Operable) Measuring ChannelSteadyStateSquareWaveChannelFunctionSurveillance Requirements* Pulsea. Reactor PowerLevel SafetyChannelb. Linear PowerChannelc. Log PowerChanneld. Fuel Temperature Channele. Pulse Channel201100222Scram at 2.2MW or lessAutomatic Power ControlStartupControlFuelTemperature MeasuresPulse NV & NVTD,M,AD,M,AD,M,ADM,A2IPA(*) Where: D -Channel check during each day's operation M -Channel test monthlyA -Channel calibration annuallyP -Channel test prior to pulsing operation Basis -a. Table 3.2.2. The two reactor power level safety channels assure that the reactor power level isproperly monitored and indicated in the reactor control room (SAR Chapter 7, Sections 7.1.2 &7.1.2.2). Amendment No. 49

• 03.3 Reactor Coolant SystemsApplicability

-These specifications apply to the operation of the reactor water measuring systems.Obiective -The objective is to assure that adequate cooling is provided to maintain fuel temperatures belowthe safety limit, and that the water quality remains high to prevent damage to the reactor fuel.Specification -The reactor shall not be operated unless the systems and instrumentation channelsdescribed in Table 3.3 are operable, and the information is displayed locally or in the control room.Table 3.3REQUIRED WATER SYSTEMS AND INSTRUMENTATION Measuring Channel/System

a. Primary CoolantCore InletTemperature Monitorb. Reactor TankLow Water.Monitorc. Purification**

Inlet Conduc-tivity Monitord. Emergency CoreCooling SystemMinimumNumberOperableSurveillance Requirements* Function: Channel/System For operation of thereactor at 1.5 MW orhigher, alarms on highheat exchanger outlettemperature of 45°C(1 130F)Alarms if water leveldrops below a depth of23 feet in the reactor tankAlarms if the primarycoolant water conductivity is greater than5 micromhos/cm For operation of the reactorat 1.5MW or higher, provideswater to cool fuel in the eventof a Loss of Coolant Accidentfor a minimum of 3.7 hoursat 20 gpm from an appropriate nozzleD,Q,AMD,M,SD,S(*) Where: D -channel check during each day's operation A -channel calibration annuallyQ -channel test quarterly S -channel calibration semiannually M -channel test monthly(**) The purification inlet conductivity monitor can be out-of-service for no more than 3 hours before the reactor shallbe shutdown. Amendment No. 413 Basis -.a. Table 3.3. The primary coolant core inlet temperature alarm assures that large power fluctuations will notoccur (SAR Chapter 4, Section 4.6.2).b. Table 3.3. The minimum height of 23 ft. of water above the reactor tank bottom guarantees that there issufficient water for effective cooling of the fuel and that the radiation levels at the top of the reactor tank arewithin acceptable limits. The reactor tank water level monitor alarms if the water level drops below a heightof 23 ft. (7.01 m) above the tank bottom (SAR Chapter 11, Section 11.1.5.1 ).c. Table 3.3. Maintaining the primary coolant water conductivity below 5 micromhos/cm averaged over aweek will minimize the activation of water impurities and also the corrosion of the reactor structure.

d. Table 3.3. This system will mitigate the Loss of Coolant Accident event analyzed in the SAR Chapter 13,Section 13.2.3.4 Reactor Room Exhaust SystemApplicability

-These specifications apply to the operation of the reactor room exhaust system.Obiective -The objectives of this specification are as follows:a. To reduce concentrations of airborne radioactive material in the reactor room, and maintain the reactorroom pressure negative with respect to surrounding areas.b. To assure continuous air flow through the reactor room in the event of a Loss of Coolant Accident. Specification -a. The reactor shall not be operated unless the reactor room exhaust system is in operation and thepressure in the reactor room is negative relative to surrounding areas.b. The reactor room exhaust system shall be operable within one half hour of the onset of a Loss of CoolantAccident. Basis -Operation of the reactor room exhaust system assures that:a. Concentrations of airborne radioactive material in the reactor room and in air leaving the reactor room willbe reduced due to mixing with exhaust system air (SAR Chapter 9, Section 9.5.1). Pressure in the reactorroom will be negative relative to surrounding areas due to air flow patterns

created, by the reactor roomexhaust system (SAR Chapter 9, Section 6.5.1).b. There will be a timely, adequate and continuous air flow through the reactor room to keep the fueltemperature below the safety limit in the event of a Loss of Coolant Accident.

3.5 This section intentionally left blank.3.6 This section intentionally left blank.3.7 Reactor Radiation Monitorinq Systems3.7.1 Monitorinq SystemsApplicability -This specification applies to the information which shall be available to the reactoroperator during reactor operation. Amendment No. 414 0Objective -The objective is to require that sufficient information regarding radiation levels andradioactive effluents is available to the reactor operator to assure safe operation of the reactor..Specification -The reactor shall not be operated unless the channels described in Table 3.7.1 areoperable, the readings are below the alarm setpoints, and the information is displayed in the controlroom. The stack and reactor room CAMS shall not be shutdown at the same time during reactoroperation. Table 3.7.1REQUIRED RADIATION MONITORING iNSTRUMENTATION Measuring Equipment MinimumNumberOperable** ChannelFunctionSurveillance Requirements*

a. FacilityStack Monitorb. Reactor RoomRadiation Monitorc. Purification System Radia-tion Monitord. Reactor RoomContinuous Air MonitorI111Monitors Argon-41 andradioactive particu-lates, and alarmsMonitors the radiation level in the reactorroom and alarmsMonitors radiation level at the demineral-izer station and alarmsMonitors air from thereactor room for parti-culate and gaseousradioactivity and alarmsD,W,AD,W,AD,W,AD,W,A(*) Where: D -channel check during each day's operation A -channel calibration annuallyW -channel test(**) monitors may be placed out-of-service for up to 2 hours for calibration and maintenance.

During this out-of-service time, no experiment or maintenance activities shall be conducted which could result in alarm conditions (e.g., airborne releases or high radiation levels)Basis -a. Table 3.7.1. The facility stack monitor provides information to operating personnel regarding the releaseof radioactive material to the environment (SAR Chapter 11, Section 11.1.1.1.4). The alarm setpoint on thefacility stack monitor is set to limit Argon-41 concentrations to less than 10 CFR Part 20, Appendix B,Table 2, Column 1 values (averaged over one year) for unrestricted locations outside the operations area.b. Table 3.7.1. The reactor room radiation monitor provides information regarding radiation levels in thereactor room during reactor operation (SAR Chapter 11, Section 11.1.5.1), to limit occupational radiation exposure to less than 10 CFR 20 limits.c. Table 3.7.1. The radiation monitor located next to the purification system resin canisters providesinformation regarding radioactivity in the primary system cooling water (SAR Chapter 11, Section 11.1.5.4.2) Amendment No. 415 and allows assessment of radiation levels in the area to ensure that personnel radiation doses will be below10 CER Part 20 limits.d. Table 3.7.1. The reactor room continuous air monitor provides information regarding airborne radioactivity in the reactor room, (SAR Chapter 11, Sections 11.1.1.1.2 & 11.1.1.1.5), to ensure that occupational exposure to airborne radioactivity will remain below the 10 CFR Part 20 limits.3.7.2 Effluents -Arqon-41 Discharqe Limit.Applicability -This specification applies to the concentration of Argon-41 that may be discharged from the UCD/MNRC reactor facility. Obiective -The objective is to ensure that the health and safety of the public is not endangered bythe discharge of Argon-41 from the UCD/MNRC reactor facility. Specification -The annual average unrestricted area concentration of Argon-41 due to releases ofthis radionuclide from the UCD/MNRC, and the corresponding annual radiation dose from Argon-41in the unrestricted area shall not exceed the applicable levels in 10 CER Part 20.Basis -The annual average concentration limit for Argon-41 in air. in the unrestricted area isspecified in Appendix B, Table 2, Column 1 of 10 CER Part 20. 10 CER 20.1301 specifies doselimitations in the unrestricted area. 10 CFR 20.1101 specifies a constraint on air emissions ofradioactive material to the environment. The SAR Chapter 11, Section 11.1.1.1.4 estimates that theroutine Argon-41 releases and the corresponding doses in the unrestricted area will be below theselimits.3.8 Experiments 3.8.1 Reactivity Limits.Aoplicability -This specification applies to the reactivity limits on experiments installed in specificreactor experiment facilities. Obiective -The objective is to assure control of the reactor during the irradiation or handling ofexperiments in the specifically designated reactor experiment facilities. Specification -The reactor shall not be operated unless the following conditions governing experiments exist:a. The absolute reactivity worth of any single moveable experiment in the pneumatic transfer tube,the central irradiation

facility, the central irradiation fixture 1 (ClF-1),

or any other in-core or in-tankirradiation

facility, shall be less than $1.00 (0.7% Ak/k), except for the automated central irradiation facility (ACIF) (See 3.8.1.c below).b. The absolute reactivity worth of any single secured experiment positioned in a reactor in-core orin-tank irradiation facility shall be less than the maximum allowed pulse ($1.75) (1.23% Ak/k).c. The absolute total reactivity worth of any single experiment or of all experiments collectively positioned in the AClF shall be less than the maximum allowed pulse ($1.75) (1.23% Ak/k).d. The absolute total reactivity Of all experiments positioned in the pneumatic transfer tube, and inany other reactor in-core and in-tank irradiation facilities at any given time shall be less than onedollar and ninety-two cents ($1.92) (1.34% Ak/k), including the potential reactivity which might resultfrom malfunction,

flooding, voiding, or removal and insertion of the experiments.

Amendment No. 416 Basis -a. A limitation of less than one dollar ($1.00) (0.7% Ak/k) on the reactivity worth of a single movableexperiment positioned in the pneumatic transfer tube, the central irradiation facility (SAR, Chapter10, Section 10.4.1), the central irradiation fixture-i (ClF-1) (SAR Chapter 10, Section 10.4.1), orany other in-core or in-tank irradiation

facility, will assure that the pulse limit of $1.75 is notexceeded (SAR Chapter 13, Section 13.2.2.2.1).

In addition, limiting the worth of each movableexperiment to less than $1.00 will assure that the additional increase in transient power andtemperature will be slow enough so that the fuel temperature scram will be effective (SAR Chapter13, Section 13.2.2.2.1).

b. The absolute worst event which may be considered in conjunction with a single securedexperiment is its sudden accidental or unplanned removal while the reactor is operating.

For suchan event, the reactivity limit for fixed experiments ($1.75) would result in a reactivity increase lessthan the $1.92 pulse reactivity insertion needed to reach the fuel temperature safety limit (SARChapter 13, Section 13.2.2.2.1).

c. A reactivity limit of less than $1.75 for any single experiment or for all experiments collectively positioned in the sample can of the automated central irradiation facility (ACIF) (SAR Chapter 10,Section 10.4.2) is based on the pulsing reactivity insertion limit (Technical Specification 3.1.2) (SARChapter 13, Section 13.2.2.2.1) and on the design of the ACIF, which allows control over thepositioning of samples into and out of the central core region in a manner identical in form, fit, andfunction to a control rod.d. it is conservatively assumed that simultaneous removal of all experiments positioned in thepneumatic transfer tube, and in any other reactor in-core and in-tank irradiation facilities at anygiven time shall be less than the maximum reactivity insertion limit of $1.92. The SAR Chapter 13,Section 13.2.2.2.1 indicates that a pulse reactivity insertion of $1.92 would be needed to reach thefuel temperature safety limit.3.8.2 Materials LimitApplicability

-This specification applies to experiments installed in reactor experiment facilities. Obiective -The objective is to prevent damage to the reactor or significant releases of radioactivity by limiting material quantity and the radioactive material inventory of the experiment. Specification -The reactor shall not be operated unless the following conditions governing experiment materials exist:a. Experiments containing materials corrosive to reactor components, compounds highly reactivewith water, potentially explosive materials, and liquid fissionable materials shall be appropriately encapsulated.

b. Each fueled experiment shall be controlled such that the total inventory of iodine isotopes 131through 135 in the experiment is no greater than 1.5 curies and the maximum strontium inventory is no greater than 5 millicuries.

c. Each experiment in the 1-125 production facility shall be controlled such that the total inventory of 1-125 dispensed or stored in the reactor room glove box shall not exceed 20 curies.d. Each experiment in the 1-125 production facility shall be controlled such that the total inventory of 1-125 being processed at any one time in the reactor room fume hood shall not exceed 200millicuries.

An additional 800 millicuries of 1-125 in sealed storage containers may also be presentin the reactor room fume hood.Amendment No. 417 Se. Explosive materials in quantities greater than 25 milligrams of TNT equivalent shall not beirradiated in the reactor tank. Explosive materials in quantities of 25 milligrams of TNT equivalent or less may be irradiated provided the pressure produced upon detonation of the explosive hasbeen calculated and/or experimentally demonstrated to be less than the design pressure of thecontainer.

f. Explosive materials in quantities of three (3) pounds of TNT equivalent or less may be irradiated in any radiography bay. The irradiation of explosives in any bay is limited to those assemblies wherea safety analysis has been performed that shows that there is no damage to the reactor safetysystems upon detonation (SAR Chapter 13, Section 13.2.6.2).

Basis -a. Appropriate encapsulation is required to lessen the experimental hazards of some types ofmaterials.

b. The 1.5 curies limitation on iodine 131 through 135 assures that in the event of failure of a fueledexperiment leading to total release of the iodine, occupational doses and doses to members of thegeneral public in the unrestricted areas shall be within the limits in 10 CFR 20 (SAR Chapter 13,Section 13.2.6.2).

c&d. Limiting the total 1-125 inventory to twenty (20.0) curies in the reactor room glove box and toone (1.0) curie in the reactor room fume hood assures that, if these inventories of 1-125 are totallyreleased into their respective containments, occupational doses and doses to members of thegeneral public in the unrestricted areas shall be within the limits of 10 CFR 20 (SAR Chapter 13,Section 13.2.6.2).

e. This specification is intended to prevent damage to vital equipment by restricting the quantity ofexplosive materials within the reactor tank (SAR Chapter 13, Section 13.2.6.2).

f. The failure of an experiment involving the irradiation of 3 lbs TNT equivalent or less in anyradiography bay external to the reactor tank will not result in damage to the reactor controls or thereactor tank. Safety Analyses have been performed (SAR Chapter 13, Section 13.2.6.2) whichshow that up to six (6) pounds of TNT equivalent can be safely irradiated in any radiogaphy bay.Therefore, the three (3) pound limit gives a safety margin of two (2).3.8.3 Failure and Malfunctions Applicability

-This specification applies to experiments installed in reactor experiment facilities. Objective -The objective is to prevent damage to the reactor or significant releases of radioactive materials in the event of an experiment failure.Specification -a. All experiment materials which could off-gas,

sublime, volatilize, or produce aerosols under:(1) normal operating conditions of the experiment or the reactor,(2) credible accident conditions in the reactor, or(3) where the possibility exists that the failure of an experiment could release radioactive gases or aerosols into the reactor building or into the unrestricted area, the quantity andtype of material in the experiment shall be limited such that the airborne radioactivity in thereactor room will not result in exceeding the applicable dose limits in 10 CFR Part 20 in theunrestricted area, assuming 100% of the gases or aerosols escapes.Amendment No. 418 0b. In calculations pursuant to (a) above, the following assumptions shall be used:(1) If the effluent from an experiment facility exhausts through a stack which is closed onhigh radiation levels, at least 10% of the gaseous activity or aerosols produced will escape.(2) If the effluent from an experiment facility exhausts through a filter installation designedfor greater than 99% efficiency for 0.3 micron and larger particles, at least 10% of these willescape.(3) For materials whose boiling point is above 130°C and where vapors formed by boilingthis material can escape only through an undistributed column of water above the core, atleast 10% of these vapors can escape.c. If a capsule fails and releases material which could damage the reactor fuel or structure bycorrosion or other means, an evaluation shall be made to determine the need for corrective action.Inspection and any corrective action taken shall be reviewed by the UCD/MNRC Director or hisdesignated alternate and determined to be satisfactory before operation of the reactor is resumed.Basis -a. This specification is intended to reduce the likelihood that airborne radioactivity in the reactorroom or the unrestricted area will result in excee~ding the applicable dose limits in 10 CFR Part 20.b. These assumptions are used to evaluate the potential airborne radioactivity release due to anexperiment failure (SAR Chapter 13, Section 13.2.6.2).

c. Normal operation of the reactor with damaged reactor fuel or structural damage is prohibited toavoid release of fission products.

Potential damage to reactor fuel or structure shall be brought tothe attention of the UCD/MNRC Director or his designated alternate for review to assure safeoperation of the reactor (SAR Chapter 13, Section 13.2.6.2). 4.0 Surveillance Requirements General. The surveillance frequencies denoted herein are based on continuing operation of the reactor.Surveillance activities scheduled to occur during an operating cycle which can not be performed with the reactoroperating may be deferred to the end of the operating cycle. If the reactor is not operated for a reasonable time, areactor system or measuring channel surveillance requirement may be waived during the associated time period.Prior to reactor system or measuring channel operation, the surveillance shall be performed for each reactor systemor measuring channel for which surveillance was waived. A reactor system or measuring channel shall not beconsidered operable until it is successfully tested.4.1 Reactor Core Parameters 4.1.1 Steady State Operation Applicability -This specification applies to the surveillance requirement for the power levelmonitoring channels. Obiective -The objective is to verify that the maximum power level of the reactor does not exceedthe authorized limit.Amendment No. 419 0 0Discovery of noncompliance with Technical Specifications 4.3.a-c shall limit operations to that required to performthe surveillance. Noncompliance with Technical Specification 4.3.d shall limit operations to less than 1.5 MW.Basis -a. A channel test quarterly assures the water temperature monitoring system responds correctly to an inputsignal. A channel check during each day's operation assures the channel is operable. A channel calibration annually assures the monitoring system reads properly.

b. A channel test monthly assures that the low water level monitoring system responds correctly to an inputsignal.c. A channel test monthly assures that the purification inlet conductivity monitors respond correctly to aninput signal. A channel check during each day's operation assures that the channel is operable.

A channelcalibration semiannually assures the conductivity monitoring system reads properly.

d. A channel check prior to operation assures that the emergency core cooling system is operable for powerlevels above 1.5 MW. A channel calibration semiannually assures that the Emergency Core Cooling Systemperforms as required for power levels above 1.5 MW.4.4 Reactor Room Exhaust SystemApplicability

-This specification applies to the surveillance requirements for the reactor roomexhaust system.Obiective -The objective is to assure that the reactor room exhaust system is operating properly. Specification -The reactor room exhaust system shall have a channel check during each day'soperation. Discovery of noncompliance with this specification shall limit operations to that required to perform the surveillance. Basis -A channel check during each day's operation of the reactor room exhaust system shall verifythat the exhaust system is maintaining a negative pressure in the reactor room relative to thesurrounding facility areas.4.5 This section intentionally left blank4.6 This section intentionally left blank.4.7 Reactor Radiation Monitorinq SystemsApplicability -This specification applies to the surveillance requirements for the reactor radiation monitoring systems.Obiective -The objective is to assure that the radiation monitoring equipment is operating properly. Specification -a. The facility stack monitor shall have the following: (1) A channel checkduring each day's operation. (2) A channel test weekly.Amendment No. 425 0.!(3) A channel calibration annually.

b. The reactor room radiation monitor shall have the following:

(1) A channel check during each day's operation. (2) A channel test weekly.(3) A channel calibration annually.

c. The purification system radiation monitor shall have the .following:

(1) A channel check during each day's operation. (2) A channel test weekly.(3) A channel calibration annually.

d. The reactor room Continuous Air Monitor (CAM) shall have the following:

(1) A channel check during each day's operation. (2) A channel test weekly.(3) A channel calibration annually. Discovery of noncompliance with Technical Specifications 4.7.a-d shall limit operations to that required to performthe surveillance. Basis -a. A channel check of the facility stack monitor system during each day's operation will assure themonitor is operable. A channel test weekly will assure that the system responds correctly to aknown source. A channel calibration annually will assure that the monitor reads correctly.

b. A channel check of the reactor room radiation monitor during each day's operation will assurethat the monitor is operable.

A channel test weekly will ensure that the system responds to a knownsource. A channel calibration of the monitor annually will assure that the monitor reads correctly.

c. A channel check of the purification system radiation monitor during each day's operation assuresthat the monitor is operable.

A channel test weekly will ensure that the system responds to a knownsource. A channel calibration of the monitor annually will assure that the monitor reads correctly.

d. A channel check of the reactor room Continuous Air Monitor (CAM) during each day's operation will assure that the CAM is operable.

A channel test weekly will assure that the CAM respondscorrectly to a known source. A channel calibration annually will assure that the CAM readscorrectly.

4.8 Experiments

Applicability -This specification applies to the surveillance requirements for experiments installed in any UCD/MNRC reactor experiment facility. Amendment No. 426 0Objective -The objective is to prevent the conduct of experiments or irradiations which may damagethe reactor or release excessive amounts of radioactive materials as a result of experimental failure.Specification -a. A new experiment shall not be installed in any UCD/MNRC reactor experiment facility until awritten safety analysis has been performed and reviewed by the UCD/MNRC

Director, or hisdesignee, to establish compliance with the Limitations on Experiments, (Technical Specifications Section 3.8) and 10 CFR 50.59.b. All experiments performed at the UCD/MNRC shall meet the conditions of an approved FacilityUse Authorization.

Facility Use Authorizations and experiments carried out under theseauthorizations shall be reviewed and approved in accordance with the Utilization of the (UCD)McClellan Nuclear Radiation Center Research Reactor Facility Document (MNRC-0027-DOC). Anexperiment classified as an approved experiment shall not be placed in any UCD/MNRC experiment

facility, until it has been reviewed for compliance with the approved experiment and Facility UseAuthorization by the Reactor Manager and the Health Physics Manager, or their designated alternates.

c. The reactivity worth of any experiment installed in the pneumatic transfer tube, or in any otherUCD/MNRC reactor in-core or in-tank irradiation facility shall be estimated or measured, asappropriate, before reactor operation with said experiment.

Whenever a measurement is done itshall be done at ambient conditions.

d. Experiments shall be identified and a log or other record maintained while experiments are in anyUCD/MNRC reactor experiment facility.

Basis -a & b. Experience at most TRIGA reactor facilities verifies the impo'rtance of reactor staff and safetycommittee reviews of proposed experiments.

c. Measurement of the reactivity worth of an experiment, or estimation of the reactivity worth basedon previous or similar measurements, shall verify that the experiment is within authorized reactivity limits.d. Maintaining a log of experiments while in UCD/MNRC reactor experiment facilities will facilitate maintaining surveillance over such experiments.

5.0 Desigqn Features5.1 Site and Facility Description. 5.1.1 Sit..eeApplicability -This specification applies to the UCD/MNRC site location and specific facility designfeatures. Objective -The objective is to specify those features related to the Safety Analysis evaluation. Specification -Amendment No. 427 0a. The site location is situated approximately 8 miles (13 kin) north-by-northeast of downtownSacramento, California. The site of the UCD/MNRC facility is about 3000 ft. (0.6 mi or 0.9 kin) westof Watt Avenue, and 4500 ft. (0.9 mi or 1.4 kin) south of E Street.b. The restricted area is that area inside the fence surrounding the reactor building. Theunrestricted area is that area outside the fence surrounding the reactor building.

c. The TRIGA reactor is located in Building 258, Room 201 of the UCD/MNRC.

This building hasbeen designed with special safety features.

d. The core is below ground level in a water filled tank and surrounded by a concrete shield.Basis -a. Information on the surrounding population, the hydrology, seismology, and climatography of thesite has been presented in Chapter 2 of the Safety Analysis Report.b. The restricted area is controlled by the UCDIMNRC Director.

c. The room enclosing the reactor has been designed with systems related to the safe operation ofthe facility.

d. The below grade core design is to negate the consequences of an aircraft hitting the reactorbuilding.

This accident was analyzed in Chapter 13 of the Safety Analysis Report, and found to bebeyond a credible accident scenario. 5.1.2 Facility ExhaustApplicability -This specification applies to the facility which houses the reactor.Obiective -The objective is to assure that provisions are made to restrict the amount of radioactivity released into the environment, or during a Loss of Coolant Accident, the system is to assure properremoval of heat from the reactor room.Specification-

a. The UCD/MNRC reactor facility shall be equipped with a system designed to filter and exhaust airfrom the UCD/MNRC facility.

The system shall have an exhaust stack height of a minimum of18.2m (60 feet) above ground level.b. Manually activated shutdown controls for the exhaust system shall be located in the reactorcontrol room.Basis -The UCD/MNRC facility exhaust system is designed such that the reactor room shall bemaintained at a negative pressure with respect to the surrounding areas. The free air volume withinthe UCD/MNRC facility is confined to the facility when there is a shutdown of the exhaust system.Controls for startup, filtering, and normal operation of the exhaust system are located in the reactorcontrol room. Proper handling of airborne radioactive materials (in emergency situations) can beconducted from the reactor control room with a minimum of exposure to operating personnel. 5.2 Reactor Coolant SystemApplicability -This specification applies to the reactor coolant system.Amendment No. 428 0Objective -The objective is to assure that adequate water is available for cooling and shielding duringnormal reactor operation or during a Loss of Coolant Accident. Specification -a. During normal reactor operation the reactor core shall be cooled by a natural convection flow of water.b. The reactor tank water level alarm shall activate if the water level in the reactor tank drops below a depthof 23 ft.c. For operations at 1.5 MW or higher during a Loss of Coolant Accident the reactor core shall be cooled fora minimum of 3.7 hours at 20 gpm by a source of water from the Emergency Core Cooling System.Basis -a. The SAR Chapter 4, Section 4.6, Table 4-19, shows that fuel temperature limit of 930°C will not beexceeded under natural convection flow conditions.

b. A reactor tank water low level alarm sounds when the water level drops significantly.

This alarmannunc~iates in the reactor control room and at a 24 hour monitored location so that appropriate corrective action can be taken to restore water for cooling and shielding.

c. The SAR Chapter 13, Section 13.2, analyzes the requirements for cooling of the reactor fuel and showsthat the fuel safety limit is not exceeded under Loss of Coolant Accident conditions during this water cooling.5.3 Reactor Core and Fuel5.3.1 Reactor CoreApplicability

-This specification applies to the configuration of the fuel.Objective -The objective is to assure that provisions are made to restrict the arrangement of fuelelements so as to provide assurance that excessive power densities will not be produced. Specification -Foroperation at 0.5 MW or greater, the reactor core shall be an arrangement of 96or more fuel elements to include fuel followed control rods. Below 0.5 MW there is no minimumrequired number of fuel elements. In a mixed 20/20, 30/20 and 8.5/20 fuel loading (SAR Chapter 4,Section 4.5.5.6): Mix J Core and Other Variations (1) No fuel shall be loaded into Hex Rings A or B.(2) A fuel followed control rod located in an 8.5 wt% environment shall contain 8.5 wt% fuel.20E Core and Other Variations (1) No fuel shall be loaded into Hex Rings A or B.(2) Fuel followed control rods may contain either 8.5 wt% or 20 wt% fuel.(3) Variations to the 20E core having 20 wt% fuel in Hex Ring C requires the 20 wt% fuel to beloaded into corner positions only, and graphite dummy elements in the flat positions. Theperformance of fuel temperature measurements shall apply to variations to the as-analyzed 20Ecore configurations. Amendment No. 429 0 S308 Core and Other Variations (1) No fuel shall be loaded into Hex Rings A or B.(2) The only fuel types allowed are 20/20 and 30/20.(3) 20/20 fuel may be used in any position in Hex Rings C through G.(4) 30/20 fuel may be used in any position in Hex Rings 0 through G but not in Hex Ring C.(5) An analysis of any irradiation facility installed in the central cavity of this core shall be donebefore it is used with this core.Basis -In order to meet the power density requirements discussed in the SAR Chapter 4, Section4.5.5.6, no less than 96 fuel elements including fuel followed control rods and the above loadingrestrictions will be allowed in an operational 0.5 MW or greater core. Specifications for the 20E coreand for the 308 core allow for variations of the as-analyzed core with the condition that temperature limits are being maintained (SAR Chapter 4, Section 4.5.5.6 and Argonne National Laboratory Report AN L/ED 97-54).5.3.2 Reactor FuelApplicability -These specifications apply to the fuel elements used in the reactor core.Obiective -The objective is to assure that the fuel elements are of such design and fabricated insuch a manner as to permit their use with a high degree of reliability with respect to their physicaland nuclear characteristics. Specification -The individual unirradiated TRIGA fuel elements shall have the following characteristics:

a. Uranium content:

8.5, 20 or 30 wt % uranium enriched nominally to less than 20% U-235.b. Hydrogen to zirconium atom ratio (in the ZrHx): 1.60 to 1.70 (I.65+/- 0.05).c. Cladding: stainless steel, nominal 0.5mm (0.020 inch) thick.Basis -a. The design basis of a TRIGA core loaded with TRIGA fuel demonstrates that limiting operation to2.3 megawatts steady state or to a 36 megawatt-sec pulse assures an ample margin of safetybetween the maximum temperature generated in the fuel and the safety limit for fuel temperature. The fuel temperatures are not expected to exceed 630°C during any condition of normal operation.

b. Analysis shows that the stress in a TRIGA fuel element, H/Zr ratios between 1.6 and 1.7, is equalto the clad yield strength when both fuel and cladding temperature are at the safety limit 930°C.Since the fuel temperatures are not expected to exceed 630°C during any condition of normaloperation, there is a margin between the fuel element clad stress and its ultimate strength.

c. Safety margins in the fuel element design and fabrication allow for normal mill tolerances ofpurchased materials.

Amendment No. 430 0 05.3.3 Control Rods and Control Rod DrivesApplicability -This specification applies to the control rods and control rod drives used in the reactorcore.Obiective -The objective is to assure the control rods and control rod drives are of Such a design asto permit their use with a high degree of reliability with respect to their physical,

nuclear, andmechanical characteristics.

Specification -a. All control rods shall have scram capability and contain a neutron poison such as stainless steel,borated graphite, B4C powder, or boron and its compounds in solid form. The shim and regulating rods shall have fuel followers sealed in stainless steel. The transient rod shall have an air filledfollower and be sealed in an aluminum tube.b. The control rod drives shall be the standard GA rack and pinion type with an electromagnet andarmature attached. Basis -a. The neutron poison requirements for the control rods are satisfied by using stainless steel,neutron absorbing borated graphite, B4C powder, or boron and its compounds. These materials shall be contained in a suitable clad material such as stainless steel or aluminum to assuremechanical stability during movement and to isolate the neutron poison from the tank waterenvironment. Scram capabilities are provided for rapid insertion of the control rods.b. The standard GA TRIGA control rod drive meets the requirements for driving the control rods atthe proper speeds, and the electromagnet and armature provide the requirements for rapid insertion capability. These drives have been tested and proven in many TRIGA reactors. 5.4 Fissionable Material StoraqeApplicability -This specification applies to the storage of reactor fuel at a time when it is not in the reactorcore.Obiective -The objective is to assure that the fuel which is being stored will not become critical and will notreach an unsafe temperature. Specification -a. All fuel elements not in the reactor core shall be stored (wet or dry) in a geometrical array where the keffis less than 0.9 for all conditions of moderation.

b. Irradiated fuel elements shall be stored in an array which shall permit sufficient natural convection coolingby water or air such that the fuel element temperature shall not exceed the safety limit.Basis -The limits imposed by Technical Specifications 5.4.a and 5.4.b assure safe storage.6.0 Administrative Controls6.1 Orqanization.

The Vice Chancellor for Research shall be the licensee for the UCD/MNRC. TheUCD/MNRC facility shall be under the direct control of the UCD/MNRC Director or a licensed senior reactoroperator (SRO) designated by the UCD/MNRC Director to be in direct control. The UCD/MNRC Directorshall be accountable to the Vice Chancellor of the Office of Research for the safe operation andmaintenance of the reactor and its associated equipment. Amendment No. 431 06.1.1 Structure. The management for operation of the UCD/MNRC facility shall consist of theorganizational structure as shown in Figure 6.1.6.1.2 Responsibilities. The UCD/MNRC Director shall be accountable to the Vice Chancellor of theOffice of Research for the safe operation and maintenance of the reactor and its associated equipment. The UCD/MNRC

Director, or his designated alternate, shall review and approve allexperiments and procedures prior to their use in the reactor.

Individuals in themanagement organization (e.g., Reactor Manager, Health Physics Manager, etc.) shall beresponsible for implementing UCD/MNRC policies and for operation of the facility, and shall beresponsible for safeguarding the public and facility personnel from undue radiation exposures andfor adhering to the operating license and technical specifications. The Reactor Manager and HealthPhysics Manager report directly to the UCD/MNRC Director. 6.1.3 Staffing6.1.3.1 The minimum staffing when the reactor is not shutdown shall be:a. A reactor operator in the control room;b. A second person in the facility area who can perform prescribed instructions;

c. A senior reactor operator readily available.

The available senior reactoroperator should be within thirty (30) minutes of the facility and reachable bytelephone, and;d. A senior reactor operator shall be present whenever a reactor startup isperformed, fuel is being moved, or experiments are being placed in the reactortank.6.1.3.2 A list of reactor facility personnel by name and telephone number shall be available to the reactor operator in the control room. The list shall include:a. Management personnel.

b. Health Physics personnel.

c. Reactor Operations personnel.

6.1.4 Selection and Training of Personnel. The selection, training and requalification of operations personnel shall meet or exceed the requirements of the American National Standard for Selection andTraining of Personnel for Research Reactors (ANS 15.4). Qualification and requalification of licensedoperators shall be subject to an approved Nuclear Regulatory Commission (NRC) program.6.2 Review. Audit. Recommendation and ApprovalGeneral Policy. Nuclear facilities shall be designed, constructed,

operated, and maintained in sucha manner that facility personnel, the general public, and both university and non-university propertyare not exposed to undue risk. These activities shall be conducted in accordance with applicable regulatory requirements.

The UCD Vice Chancellor of the Office of Research shall institute the above stated policy as thefacility license holder. The Nuclear Safety COmmittee (NSC) has been chartered to assist inmeeting this responsibility by providing timely, objective, and independent

reviews, audits,recommendations and approvals on matters affecting nuclear safety. The following describes thecomposition and conduct of the NSC.Amendment No. 432 06.2.1 NSC Composition and Qualifications.

The UCD/MNRC Director shall appoint theChairperson of the NSC. The NSC Chairperson shall appoint a Nuclear Safety Committee (NSC)of at least five (5) members knowledgeable in fields which relate to nuclear safety. The NSC shallevaluate and review nuclear safety associated with the operation and use of the UCD/MNRC. 6.2.2 NSC Charter and Rules. The NSC shall conduct its review and audit (inspection) functions in accordance with a written charter. This charter shall include provisions for:a. Meeting frequency (The committee shall meet at least semiannually).

b. Voting rules.c. Quorums (For the full committee, a quorum will be at least five (5) members.d. A committee review function and an audit/inspection function.

e. Use of subcommittees.

f. Review, approval and dissemination of meeting minutes.6.2.3 Review Function.

The responsibilities of the NSC, or a designated subcommittee thereof,shall include but are not limited to the following:

a. Review approved experiments utilizing UCD/MNRC nuclear facilities.

b. Review and approve all proposed changes to the facility

license, the Technical Specifications and the Safety Analysis Report, and any new or changed Facility Use Authorizations and proposedClass I modifications, prior to implementing (Class I) modifications, prior to taking action under thepreceding documents or prior to forwarding any of these documents to the Nuclear Regulatory Commission.

c. Review and determine whether a proposed change, test, or experiment would constitute anunreviewed safety question or require a change to the license, to a Facility Use Authorization, or tothe Technical Specifications.

This determination may be in the form of verifying a decision alreadymade by the UCD/MNRC Director.

d. Review reactor operations and operational maintenance, Class I modification

records, and thehealth physics program and associated records for all UCD/MNRC nuclear facilities.

e. Review the periodic updates of the Emergency Plan and Physical Security Plan for UCDIMNRCnuclear facilities.

f. Review and update the NSC Charter every two (2) years.g. Review abnormal performance of facility equipment and operating anomalies.

h. Review all reportable occurrences and all written reports of such occurrences prior to forwarding the final written report to the Nuclear Regulatory Commission.

i. Review the NSC annual audit/inspection of the UCD/MNRC nuclear facilities and any otherinspections of these facilities conducted by other agencies.

Amendment No. 433 06.2.4 AuditlInspection Function. The NSC or a subcommittee

thereof, shall audit/inspect reactoroperations and health physics annually.

The annual audit/inspection shall include, but not be limitedto the following:

a. Inspection of the reactor operations and operational maintenance, Class I modification records,and the health physics program and associated

records, including the ALARA program, for allUCD/MNRC nuclear facilities.

b. Inspection of the physical facilities at the UCD/MNRC.

c. Examination of reportable events at the UCD/MNRC.

d. Determination of the adequacy of UCD/MNRC standard operating procedures.

e. Assessment of the effectiveness of the training and retraining programs at the UCDIMNRC.

f. Determination of the conformance of operations at the UCD/MNRC with the facility's license andTechnical Specifications, and applicable regulations.

g. Assessment of the results of actions taken to correct deficiencies that have occurred in nuclearsafety related equipment, structures,

systems, or methods of operations.

h. Inspection of the currently active Facility Use Authorizations and associated experiments.

i. Inspection of future plans for facility modifications or facility utilization.

j. Assessment of operating abnormalities.

k. Determination of the status of previous NSC recommendations.

6.3 Radiation Safety. The Health Physics Manager shall be responsible for implementation of theUCD/MNRC Radiation Safety Program. The program should use the guidelines of the American NationalStandard for Radiation Protection at Research Reactor Facilities (ANSI/ANS 15.11). The Health PhysicsManager shall report to the UCD/MNRC Director. 6.4 Procedures. Written procedures shall be prepared and approved prior to initiating any of the activities listed in this section. The procedures shall be approved by the UCD/MNRC Director. A periodic review ofprocedures shall be performed and documented in a timely manner by the UCD/MNRC staff to assure thatprocedures are current. Procedures shall be adequate to assure the safe operation of the reactor, but shallnot preclude the use of independent judgment and action should the situation require. Procedures shall bein effect for the following items:6.4.1 Reactor Operations Procedures

a. Startup, operation, and shutdown of the reactor.b. Fuel loading, unloading, and movement within the reactor.c. Control rod removal or replacement.

d. Routine maintenance of the control rod drives and reactor safety and interlock systems or otherroutine maintenance that could have an effect on reactor safety.e. Testing and calibration of reactor instrumentation and controls, control rods and control roddrives.Amendment No. 434

f. Administrative controls for operations, maintenance, and conduct of irradiations and experiments that could affect reactor safety or core reactivity.

g. Implementation of required plans such as emergency and security plans.h. Actions to be taken to correct potential malfunctions of systems, including responses to alarmsand abnormal reactivity changes.6.4.2 Health Physics Procedures

a. Testing and calibration of area radiation

monitors, facility air monitors, laboratory radiation detection

systems, and portable radiation monitoring instrumentation.

b. Working in laboratories and other areas where radioactive materials are used.c. Facility radiation monitoring program including routine and special surveys, personnel monitoring, monitoring and handling of radioactive waste, and sampling and analysis of solid andliquid waste and gaseous effluents released from the facility.

The program shall include amanagement commitment to maintain exposures and releases as low as reasonably achievable (ALARA).d. Monitoring radioactivity in the environment surrounding the facility.

e. Administrative guidelines for the facility radiation protection program to include personnel orientation and training.

f. Receipt of radioactive materials at the facility, and unrestricted release of materials and itemsfrom the facility which may contain induced radioactivity or radioactive contamination.

g. Leak testing of sealed sources containing radioactive materials.

h. Special nuclear material accountability.

i. Transportation of radioactive materials.

Changes to the above procedures shall require approval of the UCD/MNRC Director. All such changes shall bedocumented. 6.5 Experiment Review and Approval. Experiments having similar characteristics are grouped together forreview and approval under specific Facility Use Authorizations. All specific experiments to be performed under the provisions of an approved Facility Use Authorization shall be approved by the UCD/MNRCDirector, or his designated alternate.

a. Approved experiments shall be carried out in accordance with established and approved procedures.

b. Substantive change to a previously approved experiment shall require the same review and approval asa new experiment.

c. Minor changes to an experiment that do not significantly alter the experiment may be approved by asenior reactor operator.

6.6 Required Actions6.6.1 Action to be taken in case of a safety limit violation. In the event of a safety limit violation (fueltemperature), the following action shall be taken:Amendment No. 435 0.Qa. The reactor shall be shut down and reactor operation shall not be resumed until authorized bythe NRC.b. The safety limit violation shall be promptly reported to the UCD/MNRC Director.

c. The safety limit violation shall be reported to the chairman of the NSC and to the NRC by theUCD/MNRC Director.

d. A safety limit violation report shall be prepared.

The report shall describe the following: (1) Applicable circumstances leading to the violation, including when known, the cause andcontributing factors.(2) Effect of the violation upon reactor facility components,

systems, or structures, and onthe health and safety of personnel and the public.(3) Corrective action to be taken to prevent reoccurrence.

e. The safety limit violation report shall be reviewed by the NSC and then be submitted to the NRCwhen authorization is sought to resume operation of the reactor.6.6.2 Actions to be taken for reportable occurrences.

In the event of reportable occurrences, thefollowing actions shall be taken:a. Reactor conditions shall be returned to normal or the reactor shall be shut down. If it isnecessary to shut down the reactor to correct the occurrence, operations shall not be resumedunless authorized by the UCD/MNRC Director or his designated alternate.

b. The occurrence shall be reported to the UCD/MNRC Director or the designated alternate.

TheUCD/MNRC Director shall report the occurrence to the NRC as required by these Technical Specifications or any applicable regulations.

c. Reportable occurrences should be verbally reported to the Chairman of the NSC and the NRCOperations Center within 24 hours of the occurrence.

A written preliminary report shall be sent tothe NRC, Attn: Document Control Desk, I White Flint North, 11555 Rockville Pike, Rockville MD20852, within 14 days of the occurrence. A final written report shall be sent to the above addresswithin 30 days of the occurrence.

d. Reportable occurrences should be reviewed by the NSC prior to forwarding any written reportto the Vice Chancellor of the Office of Research or to the Nuclear Regulatory Commission.

6.7 Reports. All written reports shall be sent within the prescribed interval to the NRC, Attn: DocumentControl Desk, 1 White Flint North, 11555 Rockville Pike, Rockville MD 20852.6.7.10Operating Reports. An annual report covering the activities of the reactor facility during theprevious calendar year shall be submitted within six months following the end of each calendaryear. Each annual report shall include the following information:

a. A brief summary of operating experiences including experiments performed, changes in facilitydesign, performance characteristics and operating procedures related to reactor safety occurring during the reporting period, and results of surveillance tests and inspections.

b. A tabulation showing the energy generated by the reactor (in megawatt hours), hours the reactorwas critical, and the cumulative total energy output since initial criticality.

Amendment No. 436

• (2) The written report (and, to the extent possible, the preliminary telephone report orreport by similar conveyance) shall describe,

analyze, and evaluate safety implications, andoutline the corrective measures taken or planned to prevent reoccurrence of the event.c. A report within 30 days in writing to the NRC, Document Control Desk, Washington DC.(I) Any significant variation of measured values from a corresponding predicted orpreviously measured value of safety-connected operating characteristics occurring duringoperation of the reactor;(2) Any significant change in the transient or accident analysis as described in the SafetyAnalysis Report (SAR);(3) A personnel change involving the positions of UCD/MNRC Director or UCD ViceChancellor for Research; and(4) Any observed inadequacies in the implementation of administrative or procedural controls such that the inadequacy causes or could have caused the existence ordevelopment of an unsafe condition with regard to reactor operations.

6.8 Records. Records may be in the form of logs, data sheets, or other suitable forms. The requiredinformation may be contained in single or multiple

records, or a combination thereof.

Records and logs shallbe prepared for the following items and retained for a period of at least five years for items a. through f., andindefinitely for items g. through k. (Note: Annual reports, to the extent they contain all of the required,information, may be used as records for items g. through j.)a. Normal reactor operation.

b. Principal maintenance activities.

c. Those events reported as required by Technical Specifications 6.7.1 and 6.7.2.d. Equipment and component surveillance activities required by the Technical Specifications.

e. Experiments performed with the reactor.f. Airborne and liquid radioactive effluents released to the environments and solid radioactive waste shippedoff site.g. Offsite environmental monitoring surveys.h. Fuel inventories and transfers.

i. Facility radiation and contamination surveys.j. Radiation exposures for all personnel.

k. Updated, corrected, and as-built drawings of the facility.

Amendment No. 439 072hZI -I -UNIVERSITY OFCALIFORNIA -DAVISCE CHANCELLOR FORRESEARCH(Licensee) ,TORTIONSNJCHI....................... Formal Licensing ChannelAdministrative Reporting Channel----Communications ChannelUCD/MNRC ORGANIZATION FOR LICENSING AND OPERATION FIGURE 6.140 Dr. Barry M. KleinVice Chancellor for ResearchUniversity of California, DavisOne Shields AvenueDavis, CA 95616-8558

SUBJECT:

ISSUANCE OF AMENDMENT NO. 5 TO AMENDED FACILITY OPERATING LICENSE NO. R-130 -REGENTS OF THE UNIVERSITY OF CALIFORNIA (TAC NO. MB5598)

Dear Dr. Klein:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 5 to FacilityOperating License No. R-1 30 for the McClellan Nuclear Radiation Center (MNRC) TRIGAResearch Reactor. The amendment consists of changes to the Technical Specifications (TSs)in response to your submittal of October 17, 2002, and is discussed in the enclosed SafetyEvaluation Report.Sincerely, Warren J. Eresian, Project ManagerResearch and Test Reactors SectionOperating Reactor Improvements ProgramDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation Docket No. 50-607

Enclosures:

1. Amendment No. 52. Safety Evaluation Report SUniversity of California

-Davis/McClellan MNRC Docket No. 50-607cc:Dr. Wade J. Richards5335 Price Avenue, Bldg. 258McClellan AFB, CA 95652-2504 Test, Research, and TrainingReactor Newsletter University of Florida202 Nuclear Sciences CenterGainesville, FL 32611 Dr. Barry M. KleinVice Chancellor for ResearchUniversity of California, DavisOne Shields AvenueDavis, CA 95616-8558

SUBJECT:

ISSUANCE OF AMENDMENT NO. 5 TO AMENDED FACILITY OPERATING LICENSE NO. R-130 -REGENTS OF THE UNIVERSITY OF CALIFORNIA (TAC NO. MB5598)

Dear Dr. Klein:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 5 to FacilityOperating License No. R-130 for the McClellan Nuclear Radiation Center (MNRC) TRIGAResearch Reactor. The amendment consists of changes to the Technical Specifications (TSs)in response to your submittal of October 17, 2002, and is discussed in the enclosed SafetyEvaluation Report.Sincerely, Warren J. Eresian, Project ManagerResearch and Test Reactors SectionOperating Reactor Improvements ProgramDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation Docket No. 50-607

Enclosures:

1. Amendment No. 52. Safety Evaluation ReportDISTRIBUTION:

PUBLICMMendonca AAdamsEHyltonGHiIl (2) (T5-C3)RORP\R&TR r/fWEresianPDoylePlsaacLBergSHolmesTDragounCBassettDHughesOGCPMaddenDMatthews WBecknerADAMS ACCESSION NO: ML02 TEMPLATE

1. NRR-058NAME WEresian:rdr EHylton SUttal PMadden WBecknerIiDATE 111/ /2002 11/ /2002 11/ /2002 11/ /2002 11/ /2002JOFFICIAL RECORD COPY

• 0REGENTS OF THE UNIVERSITY OF CALIFORNIA ATMcCLELLAN NUCLEAR RADIATION CENTERDOCKET NO. 50-607AMENDMENT TO AMENDED FACILITY OPERATING LICENSEAmendment No. 5License No. R-1301. The U.S. Nuclear Regulatory Commission (the Commission) has found thatA. The application for an amendment to Amended Facility Operating LicenseNo. R-1 30 filed by the Regents of the University of California at McClellan NuclearRadiation Center (the licensee) on October 17, 2002, conforms to the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and theregulations of the Commission as stated in Chapter I of Title 10 of the Code ofFederal Regulations (10 CFR);B. The facility will operate in conformity with the application, the provisions of the Act,and the rules and regulations of the Commission; C. There is reasonable assurance that (i) the activities authorized by this amendment can be conducted without endangering the health and safety of the public and (ii)such activities will be conducted in compliance with the regulations of theCommission; D. The issuance of this amendment will not be inimical to the common defense andsecurity or to the health and safety of the public;E. This amendment is issued in accordance with the regulations of the Commission asstated in 10 CFR Part 51, and all applicable requirements have been satisfied; andF. Prior notice of this amendment was not required by 10 CFR 2.105, and publication of notice for this amendment is not required by 10 CFR 2.106.

D 0-2-2. Accordingly, the license is amended by changes to the Technical Specifications asindicated in the enclosure to this license amendment, and paragraph 2.C.(ii) of AmendedFacility Operating License No. R-130 is hereby amended to read as follows:2.C.(ii) Technical Specifications The Technical Specifications contained in Appendix A, as revised throughAmendment No. 5, are hereby incorporated in the license. The licensee shalloperate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Warren J. Eresian, Project ManagerResearch and Test Reactors Section*Operating Reactor Improvements ProgramDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation

Enclosure:

Appendix A, Technical Specification ChangesDate of Issuance: ENCLOSURE TO LICENSE AMENDMENT NO. 5AMENDED FACILITY OPERATING LICENSE NO. R-130DOCKET NO. 50-607Replace the following pages of Appendix A, Technical Specifications, with the enclosedpages. The revised pages are identified by amendment number and contain vertical linesindicating the areas of change.Remove Insert17 1718 1840 40 Basi.__s-

a. A limitation of less than one dollar ($1 .00)(0.7%Ak/k) on the reactivity worth of a single movable experiment positioned in the pneumatic transfer tube, the central irradiation facility (SAR, Chapter 10, Section 10.4.1 ), thecentral irradiation fixture (CIF-1)(SAR, Chapter 10, Section 10.4.1 ), or any other in-core or in-tank irradiation

facility, will assure that the pulse limit of $1.75 is not exceeded (SAR Chapter 13, Section 13.2.2.2.1).

Inaddition, limiting the worth of each movable experiment to less than $1.00 will assure that the additional increase in transient power and temperature will be slow enough so that the fuel temperature scram will beeffective (SAR Chapter 13, Section 13.2.2.2.1).

b. The absolute worst event which may be considered in conjunction with a single secured experiment is itssudden accidental or unplanned removal while the reactor is operating.

For such an event, the reactivity limitfor fixed experiments ($1.75) would result in a reactivity increase less than the $1.92 pulse reactivity insertion needed to reach the fuel temperature safety limit (SAR Chapter 13, Section 13.2.2.2.1).

c. A reactivity limit of less than $1.75 for any single experiment or for all experiments collectively positioned in the sample can of the automated central irradiation facility (AC1F)(SAR Chapter 10, Section 1.0.4.2) is basedon the pulsing reactivity insertion limit (Technical Specification 3.1 .2)(SAR Chapter 13, Section 13.2.2.2.1) andon the design of the ACIF, which allows control .over the positioning of samples into and out of the central coreregion in a manner identical in form, fit, and function to a control rod.d. It is conservatively assumed that simultaneous removal of all experiments positioned in the pneumatic transfer tube, and in any other reactor in-core and in-tank irradiation facilities at any given time shall be lessthan the maximum reactivity insertion limit of $1.92. The SAR Chapter 13, Section 13.2.2.2.1 indicates thata pulse reactivity insertion of $1.92 would be needed to reach the fuel temperature safety limit.3.8.2 Materials LimitApplicability

-This specification applies to experiments installed in reactor experiment facilities. Objective -The objective is to prevent damage to the reactor or significant releases of radioactivity by limitingmaterial quantity and the radioactive material inventory of the experiment. Specification -The reactor shall rnot be operated unless the following conditions governing experiment materials exist:a. Experiments containing materials corrosive to reactor components, compounds highly reactive with water,potentially explosive materials, and liquid fissionable materials shall be appropriately encapsulated.

b. Each fueled experiment shall be controlled such that the total inventory of iodine isotopes 131 through 135in the experiment is no greater than 1.5 curies and the maximum strontium inventory is no greater than 5millicuries.

c. Each experiment in the 1-125 production facility shall be controlled such that the total inventory of I-125 inthe I-125 glove box shall not exceed 40 curies.d. Each experiment in the 1-125 production facility shall be controlled such that the total inventory of 1-125being handled in the 1-125 fume hood at any one time in preparation for shipping shall not exceed 20 curies.An additional.

1.0 curie of 1-125 (up to 400 millicuries in the form of quality assurance samples and up to 600millicuries in sealed storage containers) may also be present in the 1-125 fume hood.Amendment No. 517

• 0e. Explosive materials in quantities greater than 25 milligrams of TNT equivalent shall not be irradiated in thereactor tank. Explosive materials in quantities of 25 milligrams of TNT equivalent or less may be irradiated provided the pressure produced upon detonation of the explosive has been calculated and/or experimentally demonstrated to be less than the design pressure of the container.

f. Explosive materials in quantities of three (3) pounds of TNT equivalent or less may be irradiated in anyradiography bay. The irradiation of explosives in any bay is limited to those assemblies where a safetyanalysis has been performed that shows that there is no damage to the reactor safety systems upondetonation (SAR Chapter 13, Section 13.2.6.2).

Basis -a. Appropdiate encapsulation is required to lessen the experimental hazards of some types of materials.

b. The 1.5 curies limitation on iodine 131 through 135 assures that in the event of failure of a fueledexperiment leading to total release of the iodine, occupational doses and doses to members of the generalpublic in the unrestricted areas shall be within the limits in 10 CFR 20 (SAR Chapter 13, Section 13.2.6.2).

c&d. Limiting the total 1-125 inventory to forty (40.0) curies in the 1-125 glove box and to twenty-one (21.0)curies in the I-125 fume hood assures that, if either of these inventories of 1-125 is totally released into theirrespective containments, the occupational doses and doses to members of the general public in theunrestricted areas shall be within the limits of 10 CFR 20 (SAR Chapter 13, Section 1 3.2.6.2).

e. This specification is intended to prevent damage to vital equipment by restricting the quantity of explosive materials within the reactor tank (SAR Chapter 13, Section 13.2.6.2).

f. The failure of an experiment involvng the irradiation of 3 lbs TNT equivalent or less in any radiography bayexternal to the reactor tank will not result in damage to the reactor controls or the reactor tank. SafetyAnalyses have been performed (SAR Chapter 13, Section 13.2.6.2) which show that up to six (6) pounds ofTNT equivalent can be safely irradiated in any radiogaphy bay. Therefore, the three (3) pound limit gives asafety margin of two (2).3.8.3 Failure and Malfunctions Applicability

-This specification applies to experiments installed in reactor experiment facilities. Obiective -The objective is to prevent damage to the reactor or significant releases of radioactive materials in the event of an experiment failure.Specification -a. All experiment materials which could. off-gas,

sublime, volatilize, or produce aerosols under:(1) normal operating conditions of the experiment or reactor,(2) credible accident conditions in the reactor, or(3) where the possibility exists that the failure of an experiment could release radioactive gases oraerosols into the reactor building or into the unrestricted area, the quantity and type of material in theexperiment shall be limited such that the airborne radioactivity in the reactor room will not result inexceeding the applicable dose limits in 10 CFR Part 20 in the unrestricted area, assuming 100% ofthe gases or aerosols escapes.Amendment No. 518 S0

• 0SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 5 TOAMENDED FACILITY OPERATING LICENSE NO. R-130REGENTS OF THE UNIVERSITY

.OF CALIFORNIA ATMcCLELLAN NUCLEAR RADIATION CENTERDOCKET NO. 50-60

71.0 INTRODUCTION

By letter dated October 17, 2002, the Regents of the University of California (the licensee) submitted a request for amendment of the Technical Specifications (TSs), Appendix A, to FacilityOperating License No. R-i130 for the McClellan Nuclear Radiation Center (MNRC) TRIGAresearch reactor. The request provides for the following

changes, which if implemented, will resultin Revision 12 of the TSs:1. Incorporate a new management

position, the "Site Manager" into the Technical Specifications.

2. Revise Technical Specification 3.8.2, Materials Limit, to allow an increase in the Iodine-i125 inventory in the Iodine Production Facility from 20 curies to 61 curies.Each of these requests is discussed below.2.0 EVALUATION The current management structure includes an UCD/MNRC Director to whom reports aHealth Physics Manager and Reactor Operations Manager.

The proposed management structure creates a new position of Site Manager, who reports directly to the UCD/MNRC

Director, and towhom reports the Health Physics Manager and the Reactor Operations Manager.

The proposedmanagement structure thus creates an additional layer of oversight. Since this change increases oversight and supervision of facility operations, the staff concludes that it is acceptable. Amendment No. 4 of the Technical Specifications was approved on August 9, 2001. Thisamendment approved the installation of an Iodine-125 production loop. The production loopincluded a reactor room glove box containing up to 20 curies of lodine-125. Technical Specification 3.8.2, which provides materials limits of experiments installed in reactor experiment facilities, was expanded to include limits associated with the production loop and in particular, thereactor room glove box. The justification for the 20 curie limit was provided in Chapter 13,Accident

Analysis, of the facility Safety Analysis Report.Previous calculations supporting the 20 curie limit of Iodine-125 were based on the worst-case assumption that all 20 curies of Iodine-125 volatilized and left the glove box through the glove box 0-2-exhaust system, eventually to make its way to the unrestricted area. The exposure (CEDE to thethyroid) to a person in the unrestricted area for the entire 30 second duration of this event is muchless than 1 millirem.

If the exposure duration is increased to 10 minutes, the estimated CEDE tothe thyroid would still be less than 1 millirem. For those exposed in the reactor room for themaximum assumed occupancy time of 5 minutes the CEDE to the thyroid would be about 67millirem. The results of all of the assumptions and calculations in the accident sequence are directlyproportional to the initial inventory of Iodine-125 in the production system. Increasing the initialassumed inventory from 20 curies to 61 curies will simply result in a tripling of the exposure. Theanalysis in the SAR that supports the increase in iodine inventory shows that the CEDE to thethyroid for a 10-minute exposure in the unrestricted area would be about 3.0 millirem, For thoseexposed in the reactor room for the maximum assumed occupancy time of 5 minutes the CEDE tothe thyroid would be about 205 millirem. In order to assess the potential consequences of the worst-case assumption, the resulting dosesare compared to the doses which are expected for the Maximum Hypothetical Accident (MHA),which serves as the bounding accident for radiological consequences. The MHA has beenanalyzed in the licensee's Safety Analysis Report (SAR), and is a complete cladding rupture of ahighly-irradiated single fuel element, followed by the instantaneous release of fission products intothe air.. The accident analysis calculates the radiological consequences of the MHA with regard todoses to the general public in the unrestricted area, and also calculates occupational doses withinthe site boundary. The MHA results in a CEDE of 53 millirem in the unrestricted area. Since therelease of 61 curies of Iodine-125 through the glovebox exhaust system and eventually to theunrestricted area results in a CEDE of about 3 millirem, the radiological result is significantly less.than that of the MHA, the bounding accident. For those exposed in the reactor room, the MHA results in an exposure (CEDE) of 360 millirem. For the failure analyzed here, the five-minute is about 205 millirem. Again, the exposures are lessthan that of the MHA, the bounding accident. The staff concludes that the consequences of the complete volatilization of 61 curies of Iodine-125 are much less than the consequences of the bounding MHA, and that increasing theallowable activity of lodine-125 in the Iodine Production Facility from 20 curies to 61 curies doesnot significantly reduce the margin of safety with respect to the Maximum Hypothetical Accidentand to 10 CFR Part 20 limits and that the increase is acceptable. 3.0 ENVIRONMENTAL CONSIDERATION This amendment involves changes in the installation or 'use of a facility component located withinthe restricted area as defined ion 10 CFR Part 20 or changes in inspection and surveillance requirements. The staff has determined that this amendment involves no significant increase inthe amounts, and no significant change in the types, of any effluents that may be released off site,and no significant increase in individual or cumulative occupational radiation exposure. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10Amendment No. 5 0 0-3-CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement orenvironmental assessment need be prepared with the issuance of this amendment.

4.0 CONCLUSION

The staff has concluded, on the basis of the considerations discussed above, that (1) because theamendment does not involve a significant increase in the probability or consequences ofaccidents previously evaluated, or create the possibility of a new or different kind of accident fromany accident previously evaluated, and does not involve a significant reduction in a margin ofsafety, the amendment does not involve a significant hazards consideration; (2) there isreasonable assurance that the health and safety of the public will not be endangered by theproposed changes; and (3) such changes are in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security orthe health and safety of the public.Principal Contributor: Warren J. EresianDate:Amendment No. 5

• U REGUALTORY COMMISSION i __ _ _ _ _ _ _ _ _ _ _ _.Ii UNIVERSITY OFl CALIFORNIA

-DAVIS* VICE CHANCELLOR FORI ~RESEARCHi -(Licensee) I II IISDIRECTOR NUCLEAR.H_____SAFETYCO l -COMMITITEE LI A-tC--SITE 1 iMANAGER[ I i-***-*HEALTH PHYSICS REACTORBRANCH OPERATIONS Forml Liensig Chnne___________ Aminstrtie RpotinBCANnel CormmunLicatinsin ChannelUCD/MNRC ORGANIZATION FOR LICENSING AND OPERATION FIGURE 6.1 RE NIE SAENUCLEAR REGULATORY COMMISSION ~WASHJNGTON, D.C. 20555-0001 N~ovemb~er 2_5, 2003Dr. Barry M. KleinVice Chancellor for ResearchUniversity of California, DavisOne Shields AvenueDavis, CA 95616-8558

SUBJECT:

ISSUANCE OF AMENDMENT NO. 6 TO AMENDED FACILITY OPERATING LICENSE NO. R-130 -REGENTS OF THE UNIVERSITY OF CALIFORNIA (TAC NO. MB5598)

Dear Dr. Klein:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 6 toFacility Operating License No. R-130 for the McClellan Nuclear Radiation Center (MNRC)TRIGA Research Reactor. The amendment consists of changes to the Technical Specifications (TSs) in response to your submittal of March 31, 2003, and is discussed in the enclosed SafetyEvaluation Report.Sincerely, 6~)4A,~ .6Warren J. Eresian, Project ManagerResearch and Test Reactors SectionNew, Research and Test Reactors ProgramDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation Docket No. 50-607

Enclosures:

1. Amendment No. 62. Safety Evaluation Report

• 0University of California

-Davis/McClellan MNRC Docket No. 50-607cc:Dr. Wade J. Richards5335 Price Avenue, Bldg. 258McClellan AFB, CA 95652-2504 Test, Research, and TrainingReactor Newsletter University of Florida202 Nuclear Sciences CenterGainesville, FL 32611 e- ~* *~W OUNITED STATES REGULATORY COMMISSION D.C. 20555-0001 REGENTS OF THE UNIVERSITY OF CALIFORNIA ATMcCLELLAN NUCLEAR RADIATION CENTERDOCKET NO. 50-607AMENDMENT TO AMENDED FACILITY OPERATING LICENSEAmendment No. 6License No. R- 1301. The U.S. Nuclear Regulatory Commission (the Commission) has found thatA. The application for an amendment to Amended Facility Operating LicenseNo. R-1 30 filed by the Regents of the University of California at McClellan NuclearRadiation Center (the licensee) on March 31, 2003, conforms to the standards andrequirements of the Atomic Energy Act of 1954, as amended (the Act), and theregulations of the Commission as stated in Chapter I of Title 10 of the Code ofFederal Regulations (10 CFR);B. The facility will operate in conformity with the application, the provisions of the Act,and the rules and regulations of the Commission; C. There is reasonable assurance that (i) the activities authorized by this amendment can be conducted without endangering the health and safety of the public and (ii)such activities will be conducted in compliance with the regulations of theCommission; D. The issuance of this amendment will not be inimical to the common defense andsecurity or to the health and safety of the public;E. This amendment is issued in accordance with the regulations of the Commission asstated in 10 CFR Part 51, and all applicable requirements have been satisfied; andF. Prior notice of this amendment was not required by 10 CFR 2.105, and publication of notice for this amendment is not required by 10 CFR 2.106.

• 0-2-2. Accordingly, the license is amended by changes to the Technical Specifications asindicated in the enclosure to this license amendment, and paragraph 2.C.(ii) of AmendedFacility Operating License No. R-130 is hereby amended to read as follows:2.C.(ii)

Technical Sp~ecifications The Technical Specifications contained in Appendix A, as revised throughAmendment No. 6, are hereby incorporated in the license. The licensee shalloperate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Patrick M. Madder Seto ChiefResearch and Test Reactors SectionNew, Research and Test Reactors ProgramDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation

Enclosure:

Appendix A, Technical Specification ChangesDate of Issuance: November 25, 2003

• 0ENCLOSURE TO LICENSE AMENDMENT NO. 6AMENDED FACILITY OPERATING LICENSE NO. R-130DOCKET NO. 50-607Replace the following pages Of Appendix A, Technical Specifications, with the enclosedpages. The revised pages are identified by amendment number and contain vertical linesindicating the areas of change.Remove Insert.31 3132 3233 33Figure 6.1 Figure 6.1
• 05.4 Fissionable Material StorageAppDlicabilitv

-This specification applies to the storage of reactor fuel at a time when it is not in the reactorcore.Obiective -The objective is to assure that the fuel which is being stored will not become critical and will notreach an unsafe temperature. Specification -a. All fuel elements not In the reactor core shall be stored (wet or diy) in a geometrical array where the keffis less than 0.9 for all conditions of moderation.

b. Irradiated fuel elements shall be stored in an array which shall permit sufficient natural convection cooling by water or air such that the fuel temperature shall not exceed the safety limit.Basis -The limits imlposed by Technical Specifications 5.4.a and 5.4.b assure safe storage.6.0 Administrative Controls6.1 Organization.

The Vice Chancellor for Research shall be the licensee for the UCD1MNRC. The facility shall be under the direct control of the UCD/MNRC Director. The UCD/MNRC Directorshall be accountable to the Vice Chancellor for Research for the safe operation and maintenance of thefacility. 6.1.1 Structure. The management for operation of the UCD/MNRC facility shall consist of theorganizational structure as shown in Figure 6.1.6.1.2 Responsibilities. The UCDIMNRC Director shall be accountable to the Vice Chancellor forResearch for the safe operation and maintenance of the facility. The UCDIMNRC

Director, or hisdesignated alternate, shall review and approve all experiments and experiment procedures prior totheir use in the reactor.

Individuals in the management organization (e.g., Operations Manager,Reactor Supervisor, Health Physics Supervisor) shall be responsible for implementing U CD/MNRCpolicies and for operation of the facility, and shall be responsible for safeguarding the public andfacility personnel from undue radiation exposures and for adhering to the operating license andtechnical specifications. The Operations Manager shall report directly to the UCD/MNRC

Director, and shall immediately report all items involving safety and licensing to the Director for a finaldecision.

The Reactor Supervisor and Health Physics Supervisor report directly to the Operations Manager.. 6.1.3 Staffing6.1.3.1 The minimum staffing when the reactor is not shutdown shall be:a. A reactor operator in the control room;b. A second person in the facility who can perform prescribed instructions;

c. A senior reactor operator readily available.

The available senior reactoroperator should be within thirty (30) minutes of the facility and reachable bytelephone, and;d. A senior reactor operator shall be present whenever a reactor startup isperformed, fuel Is being moved, or experiments are being placed In the reactortank.6.1.3.2 A list of reactor facility personnel by name and telephone number shall be available to the reactor operator in the control room. The list shall include:Amendment No. 631

• 0a. Management personnel.

b. Health Physics personnel.

c. Reactor Operations personnel.

6.1.4 Selection and Training of Personnel. The selection, training and requalification of operations personnel shall meet or exceed the requirements of the American National Standard for Selection and Training of Personnel for Research Reactors (ANS 15.4). Qualification and requalification oflicensed operators shall be subject to an approved Nuclear Regulatory Commission (NRC)program.6.2 Review. Audit. Recommendation and ApprovalGeneral Policy. Nuclear facilities shall be designed, constructed,

operated, and maintained in suchamanner that facility personnel, the general public, and both university and non-university propertyare not exposed to undue risk. These activities shall be conducted in accordance with applicable regulatory requirements.

The UCD Vice Chancellor for Research shall institute the above stated policy as the facility licenseholder. The Nuclear Safety Committee (NSC) has been chartered to assist in meeting thisresponsibility by providing timely, objective, and independent

reviews, audits, recommendations andapprovals on matters affecting nuclear safety. The following describes the composition andconduct of the NSC.6.2.1 NSC Composition and Qualifications.

The UCD Vice Chancellor for Research shall appointthe Chairperson of the NSC. The NSC Chairperson shall a ppoint a Nuclear Safety Committee (NSC) of at least seven (7) members knowledgeable in fields which relate to nuclear safety. TheNSC Shall evaluate and review nuclear safety associated with the operation and use of theUCD/MNRC. 6.2.2 NSC Charter and Rules. The NSC shall conduct its review and audit (inspection) functions inaccordance with a written charter. This charter shall include provisions for:a. Meeting frequency (The committee shall meet at least semiannu'ally.)

b. Voting rules.c. Quorums (For the full committee, a quorum will be at least seven (7) members.d. A committee review function and an audit/inspection function.

e. Use of subcommittees.

f. Review, approval and dissemination of meeting minutes.6.2.3 Review Function.

The responsibilities of the NSC, or a designated subcommittee thereof,shall "but ar--e-n'ot limited to the following:

a. Review approved experiments utilizing UCD/MNRC nuclear facilities.

b. Review and approve all proposed changes to the facility

license, the Technical Specifications and the Safety Analysis Report, and any new or changed Facility Use Authorizations and proposedClass I modifications, prior to implementing (Class I) modifications, prior to taking action under thepreceding documents or prior to forwarding any of these documents to the Nuclear Regulatory Commission for approval.

c. Review and determine whether a proposed change, test, or experiment would constitute anunreviewed safety question or require a change to the license, to a Facility Use Authorization, orto the Technical Specifications.

This determination may be in the form of verifying a decisionalready made by the UCD/MNRC Director. Amendment No. 632

d. Review reactor operations and operational maintenance, Class I modification

records, and thehealth physics program and associated records for all UCDIMNRC nuclear facilities.

e. Review the periodic updates of the Emergency Plan and Physical Security Plan for UCD/MNRCnuclear facilities.

f. Review and update the NSC Charter every two (2) years.g. Review abnormal performance of facility equipment and operating anomalies.

h. Review all reportable occurrences and all written reports of such occurrences prior to forwarding the final written report to the Nuclear Regulatory Commission.

i. Review the NSC annual audit/inspection of the UCD/MNRC nuclear facilities and any otherinspections of these facilities conducted by other agencies.

6.2.4 Audit/Inspection Function. The NSC or a subcommittee

thereof, shall audit/inspect reactoroperations and health physics annually.

The annual audit/inspection shall include, but not belimited to the following:

a. Inspection of the reactor operations and operational maintenance, Class I modification records,and the health physics program and associated

records, including the ALARA program, for allUCDIMNRC nuclear facilities.

b. Inspection of the physical facilities at the UCD/MNRC.

c. Examination of reportable events at the UCDIMNRC.

d. Determination of the adequacy of UCD/MNRC standard operating procedures.

e. Assessment of the effectiveness of the training and refraining programs at the UCD/MNRC.

f. Determination of the conformance of operations at the UCD/MNRC with the facility's license andTechnical Specifications, and applicable regulations.

g. Assessment of the results of actions taken to correct deficiencies that have occurred in nuclear.safety related equipment, structures,

systems, or methods of operations.

h. Inspection of the currently active Facility Use Auhorizations and associated experiments.

i. Inspection of future plans for facility modifications or facility utilization.

j. Assessment of operating abnormalities.

k. Determination of the status of previous NSC recommendations.

6.3 Radiation Safety. The Health Physics Supervisor shall be responsible for implementation of theUCD/MNRC Radiation Safety Program. The program should use the guidelines of the American NationalStandard for Radiation Protection at Research Reactor Facilities (ANSI/ANS 15.11). The Health PhysicsSupervisor shall report to the Operations Manager.6.4 Procedures. Written procedures shall be prepared and approved prior to initiating any of the activities listed nthssction. The procedures shall be approved by the UCD/MNRC Director. A periodic review ofprocedures shall be performed and documented in a timely manner by the UCD/MNRC staff to assure thatprocedures are current. Procedures shall be adequate to assure the safe operation of theAmendment No. 633

• 0I-..... COMMISSION UNVRIYOCAIOM AISAFETYECOITYTEE 1 AIFRI C-MDATEES I VIE MANAGELLRFO I, ISUPERISRECREANCTO AR.. SFTOPERSUPERVISOR M A N A G ER_______________________________________________

iForml Liensig ChnneUCD/NRC ORGAIZAIOSOR REACTOR OETINLICNIGADSFT FIGURE 6.1 1-**R R OUNITED STATES,NUCLEAR REGULATORY COMMISSIONWASHINGTON, D.C. 20555-0001 "SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 6 TOAMENDED FACILITY OPERATING LICENSE NO. R-130REGENTS OF THE UNIVERSITY OF CALIFORNIA ATMcCLELLAN NUCLEAR RADIATION CENTERDOCKET NO. 50-60

71.0 INTRODUCTION

By letter dated March 31, 2003, the Regents of the University of California (the licensee) submitted a request for amendment of the Technical Specifications (TSs), Appendix A, to FacilityOperating License No. R-1 30 for the McClellan Nuclear Radiation Center (MNRC) TRIGAresearch reactor. The request provides for the following

changes, which if implemented, will resultin Revision 13 of the TSs:1. Incorporate a new management

position, the uOperations Manager" into the Technical Specifications and change the UCD/MNRC Organization Chart to reflect this change.2. Change the appointing authority of the Chairperson of the Nuclear Safety Committee (NSC) from the Director of the UCD/MNRC to the UCD Vice Chancellor for Research, andchange the Technical Specifications and UCD/MNRC Organization Chart to reflect thischange.Each of these requests is discussed below.2.0 EVALUATION The current organization structure includes an UCD/MNRC Director to whom reports a SiteManager.

The proposed organization structure, as reflected in Figure 6.1, replaces the Site.Manager position with the position of Operations

Manager, who reports directly to the UCD/MNRCDirector, and to whom reports the Health Physics Branch and the Reactor Operations Branch.Since the proposed organization structure does not alter or reduce lines of authority and oversight, the staff concludes that it is acceptable.

In the current organization structure, the UCD/MNRC Director is responsible for appointing theChairperson of the NSC. In the proposed organization structure, that responsibility is given to theUCD Vice Chancellor for Research, who is also the licensee for the UCD/MNRC. Since thisproposed change increases the level of oversight from the licensee's staff to the licensee, the staffconcludes that it is acceptable. The staff has reviewed the proposed changes to the TSs and concluded that they areadministrative in nature and do not impact the licensee's ability to continue to meet the relevantrequirements of 10 CFR 50.36.3.0 ENVIRONMENTAL CONSIDERATION This amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR51 .22(c)(1 0). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared with the issuance of this amendment.

4.0 CONCLUSION

The staff has concluded, on the basis of the considerations discussed above, that (1) because theamendment does not involve a significant increase in the probability or consequences of accidents previously evaluated, or create the possibility of a new or different kind of accident from anyaccident previously evaluated, and does not involve a significant reduction in a margin of safety,the amendment does not involve a significant hazards consideration; (2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposedchanges; and (3) such changes are in compliance with the Commission's regulations and theissuance of this amendment will not be inimical to the common defense and security or the healthand safety of the public.Principal Contributor: Warren J. EresianDate: November 25, 2003Amendment No. 6 0 0TECHNICAL SPECIFICATIONS FOR THEUNIVERSITY OF CALIFORNIA -DAVIS MCCLELLAN NUCLEAR RADIATION CENTER(UCD/MNRC) DOCUMENT NUMBER: MNRC-0004-DOC-1 2Rev 12 09/02 Oct. 16 0? 11:OOa1,JLaJ. rwdIL.Ar t~~r 6. JohnsontS4lJ 753-9743.atDfJh.L.* ~ r. t.wc.p.1TECHNICAL SPECtFICATIONS APPROVALRevision 12 of me "Teclnical Gpo ctifoons* for the Universit of California at DavistlMcCleIlan, NuclearRadiation Cencer (UOI)/MNRG) Reactor have undergone the following coordination: Reviewed Rcvicwcd by'." 'floa rMnae " "10 ~ 02-DaleD~kcIR~eviewed by:Approved by:Site ManagerUCD/MNRc~bir4ctor Date/~zL7z~OZ-DataDateIApprovod by;

• 0Technical Specifications Rev 12 09/2002TtePageRe12 902Titovle Page Rev 12 9/200232 Rev 12 9/2002Figure 6.1 Rev 12 9/2002 S 0TECHNICAL SPECIFICATIONS TABLE OF CONTENTSPaage1.0 Definitions

................................... ............................................................................ 2.0 Safety Limit and Limiting Safety System Setting (LSSS.)............................................................ 52.1 Safety Limits................................................................................................... 52.2 Limiting Safety System Setting (LSSS)...................................................................... 62.2.1 Fuei Temperature.................................................................................. 63.0 Limiting Conditions for Operations (LC.O.0............................................................................. 73.1 Reactor Core Parameters.................................................................................... 73.1 .1 Steady-State Operation ........................................................................... 73.1.2 Pulse or Square Wave Operation ................................................................ 73.1.3 Reactivity Limitations .............................................................................. 83.2 Reactor Control and Safety Systems........................................................................ 83.2.1 Control Rods....................................................................................... 83.2.2 Reactor Instrumentation................................... ........................................ 93.2.3 Reactor Scrams and Interlocks.................................................................. 103.2.4 Reactor Fuel Elemenis........................................................................... 123.3 Reactor Coolant Systems................................................................................... 133.4 Reactor Room Exhaust System ............................................................................ 143.5 Intentionally Left Blank ...................................................................................... 143.6 Intentionally Left Blank ...................................................................................... 143.7 Reactor Radiation Monitoring Systems .................................................................... 143.7.1 Monitoring Systems .............................................................................. 143.7.2 Effluents -Argon-41 Discharge Limit..............,.............................................. 16 9 03.8 Experiments ................................................................................................. 163.8.1 Reactivity Limits................................................................................... 163.8.2 Materials Limit.................................................................................... 173.8.3 Failure and Malfunctions......................................................................... 184.0 Surveillance Requirements .......................................................................................... 194.1 Reactor Core Parameters................................................................................... 194.1.1 Steady State Operation.......................................................................... 194.1.2 Shutdown Margin and Excess Reactivity ....................................................... 204.2 Reactor Control and Safety Systems ...................................................................... 204.2.1 Control Rods ..................................................................................... 204.2.2 Reactor Instrumentation ......................................................................... 214.2.3 Reactor Scrams and Interlocks.................................................................. 224.2.4 Reactor Fuel 234.3 Reactor Coolant .244.4 Reactor Room Exhaust System ............................................................................ 254.5 Intentionally Left Blank...................................................................................... 254.6 Intentionally Left Blank...................................................................................... 254.7 Reactor Radiation Monitoring Systems .................................................................... 254.8 Experiments ................................................................................................. 265.0 Design Features ...................................................................................................... 275.1 Site and Facility Description ................................................................................ 275.1.1 .Site................................................................................................ 275.1.2 Facility Exhaust .................................................................................. 285.2 Reactor Coolant System..................................................................................... 28 0 S5.3 Reactor Core and F~ue]........................................................................................ 295.3.1 Reactor Core ..........................................................................

.............

295.3.2 Reactor .FuelJ........................................................................................ 305.3.3 Control Rods and Control Rod Drives ............................................................ 315.4 Fissionable Material Storage.................................................................................. 316.0 Administrative Controls.................................................................................................. 316.1 Organization.................................................................................................... 316.1.1 Structure............................................................................................. 326.1.2 Responsibilities..................................................................................... 326.1.3 Staffing .............................................................................................. 326.1.4 Selection and Training of Personnel.............................................................. 326.2 Review, Audit, Recommendation and Approval............................................................. 326.2.1 NSC Composition and Qualifications ................................................ i............ 336.2.2 NSC Charter and Rules ........................................................................... 336.2.3 Review Function.................................................................................... 336.2.4 Audit/Inspection Function.......................................................................... 346.3 Radiation Safety................................................................................................ 346.4 Procedures ..................................................................................................... 346.4.1 Reactor Operations Procedures................................................................... 346.4.2 Health Physics Procedures........................................................................ 356.5 Experiment Review and Appro~ial ............................................................................ 356.6 Required Actions............................................................................................... 356.6.1 Actions to be taken in case of a safety limit violation............................................ 356.6.2 Actions to be taken for reportable occurrences ................................................. 36 6.7 Reports .......................................................................................................... 366.7.1 Operating Reports.................................................................................. 366.7.2 Special Reports..................................................................................... 386.8 Records ......................................................................................................... 39 IFig. 6.1 UCD/MNRC Organization for Licensing and Operation......................................................... 40 0 0TECHNICAL SPECIFICATIONS FOR THEUNIVERSITY OF CALIFORNIA -DAVIS/MCCLELLAN NUCLEAR RADIATION CENTER (UCD/MNRC) GeneralThe University of California -Davis/McClellan Nuclear Radiation Center (UCD/MNRC) reactor is operated by theUniversity of California, Davis, California (UCD). The UCD/MNRC research reactor is a TRIGA-type reactor.The UCD/MNRC provides state-of-the-art neutron radiography capabilities. In addition, the UCD/MNRC providesa wide range of irradiation services for both research and industrial needs. The reactor operates at a nominalsteady state power level up to and including 2 MW. The UCD/MNRC reactor is also capable of square waveand pulse operational modes. The UCD/MNRC reactor fuel is less than 20% enriched in uranium-235.

1.0 Definitions

1.1 As Low As Reasonably Achievable (ALARA). As defined in 10 CFR, Part 20.1.2 Licensed Operators. A UCD/MNRC licensed operator is an individual licensed by the NuclearRegulatory Commission (e.g., senior reactor operator or reactor operator) to carry out the duties andresponsibilities associated with the position requiring the license.1.2.1 Senior Reactor Operator. An individual who is licensed to direct the activities of reactoroperators and to manipulate the controls of the facility. 1.2.2 Reactor Operator. An individual who is licensed to manipulate the controls of the facilityand perform reactor-related maintenance. 1.3 Channel. A channel is the combination of sensor, line amplifier, processor, and output deviceswhich are connected for the purpose of measuring the value of a parameter. 1.3.1 Channel Test. A channel test is the introduction of a signal into the channel for verification that it is operable. 1.3.2 Channel Calibration. A channel calibration is an adjustment of the channel such that itsoutput corresponds with acceptable accuracy to known values of the parameter which thechannel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm or trip, and shall be deemed to include a channel test.1.3.3 Channel Check. A channel check is a qualitative verification of acceptable performance byobservation of channel behavior. This verification, where possible, shall include comparison ofthe channel with other independent channels or systems measuring the same variable. 1.4 Confinement. Confinement means isolation of the reactor room air volume such that the movementof air into and out of the reactor room is through a controlled path.1.5 Experiment. Any operation,

hardware, or target (excluding devices such as detectors, fissionchambers, foils, etc), which is designed to investigate specific reactor characteristics or which isintended for irradiation within an experiment facility and which is not rigidly secured to a core or shieldstructure so as to be a part of their design.1.5.1 Experiment.

Moveable. A moveable experiment is one where it is intended that the entireexperiment may be moved in or near the reactor core or into and out of reactor experiment facilities while the reactor is operating. I 0 01.5.2 Experiment. Secured. A secured experiment is any experiment, experiment

facility, orcomponent of an experiment that is held in a stationary position relative to the reactor bymechanical means. The restraining force must be substantially greater than those to which theexperiment might be subjected by hydraulic, pneumatic,

buoyant, or other forces which arenormal to the operating environment of the experiment, or by forces which can arise as a resultof credible conditions.

1.5.3 Experiment Facilities. Experiment facilities shall mean the pneumatic transfer tube,beamtubes, irradiation facilities in the reactor core or in the reactor tank, and radiography bays.1.5.4 Experiment Safety System. Experiment safety systems are those systems, including theirassociated input circuits, which are designed to initiate a scram for the primary purpose ofprotecting an experiment or to provide information which requires manual protective action to beinitiated. 1.6 Fuel Element. Standard. A fuel element is a single TRIGA element. The fuel is U-ZrH clad instainless steel. The zirconium to hydrogen ratio is nominally 1.65 +/- 0.05. The weight percent (wt%) ofuranium can be either 8.5, 20, or 30 wt%, with an enrichment of less than 20% U-235. A standard fuelelement may contain a burnable poison.1.7 _Fuel Element. Instrumented. An instrumented fuel element is a standard fuel element fabricated withthermocouples for temperature measurements. An instrumented fuel element shall have at least oneoperable thermocouple embedded in the fuel near the axial and radial midpoints. 1.8 Measured Value. The measured value is the value of a parameter as it appears on the output of achannel.1.9 Mode, Steady-State. Steady-state mode operation shall mean operation of the UCD/MNRC reactorwith the selector switch in the automatic or manual mode position. 1.10 Mode. Square-Wave. Square-wave mode operation shall mean operation of the UCD/MNRCreactor with the selector switch in the square-wave mode position. 1.11 Mode. Pulse. Pulse mode operation shall mean operation of the UCD/MNRC reactor with theselector switch in the pulse mode position. 1.12 Operable. Operable means a component or system is capable of performing its intended function. 1.13 Operating. Operating means a component or system is performing its intended function. 1.14 Operating Cycle. The period of time starting with reactor startup and ending with reactor shutdown. 1.15 Protective Action. Protective action is the initiation of a signal or the operation of equipment withinthe UCD/MNRC reactor safety system in response to a variable or condition of the UCD/MNRC reactorfacility having reached a specified limit.1.15.1 Channel Level. At the protective instrument channel level, protective action is thegeneration and transmission of a scram signal indicating that a reactor variable has reached thespecified limit.1.15.2 Subsystem Level. At the protective instrument subsystem level, protective action is thegeneration and transmission of a scram signal indicating that a specified limit has been reached.NOTE: Protective action at this level would lead to the operation of the safety shutdownequipment. 2

• 01.15.3 Instrument System Level. At the protective instrument level, protective action is thegeneration and transmission of the command signal for the safety shutdown equipment tooperate.1.15.4 Safety System Level. At the reactor safety system level, protective action is the operation of sufficient equipment to immediately shut down the reactor.1.16 Pulse Operational Core. A pulse operational core is a reactor operational core for which themaximum allowable pulse reactivity insertion has been determined.

1.17 Reactivity. Excess. Excess reactivity is that amount of reactivity that would exist if all control rods(control, regulating, etc.) were moved to the maximum reactive position from the point where the reactoris at ambient temperature and the reactor is critical. (K eff = 1)1.18 Reactivity Limits. The reactivity limits are those limits imposed on the reactivity conditions of thereactor core.1.19 Reactivity Worth of an Experiment. The reactivity worth of an experiment is the maximum value ofthe reactivity change that could occur as a result of changes that alter experiment position orconfiguration. 1.20 Reactor Controls. Reactor controls are apparatus and/or mechanisms the manipulation of whichdirectly affect the reactivity or power level of the reactor.1.21 Reactor Core. Operational. The UCD/MNRC reactor operational core is a core for which theparameters of excess reactivity, shutdown margin, fuel temperature, power calibration and reactivity worths of control rods and experiments have been determined to satisfy the requirements set forth inthese Technical Specifications. 1.22 Reactor Operating. The UCD/MNRC reactor is operating whenever it is not shutdown or secured.1.23 Reactor Safety Systems. Reactor safety systems are those systems, including their associated input channels, which are designed to initiate automatic reactor protection or to provide information forinitiation of manual protective action.1.24 Reactor Secured. The UCD/MNRC reactor is secured when the console key switch is in the offposition and the key is removed from the lock and under the control of a licensed

operator, and theconditions of a or b exist:a. (1) The minimum number of control rods are fully inserted to ensure the reactor is shutdown, asrequired by technical specifications; and(2) No work is in progress involving core fuel, core structure, installed control rods, or control roddrives, unless the control rod drives are physically decoupled from the control rods; and(3) No experiments in any reactor experiment

facility, or in any other way near the reactor, are beingmoved or serviced if the experiments have, on movement, a reactivity worth exceeding the maximumvalue allowed for a single experiment or $1.00, whichever is smaller, orb. The reactor contains insufficient fissile materials in the reactor core, adjacent experiments or controlrods to attain criticality under optimum available conditions of moderation and reflection.

1.25 Reactor Shutdown. The UCDIMNRC reactor is shutdown if it is subcritical by at least one dollar($1.00) both in the Reference Core Condition and for all allowed ambient conditions with the reactivity worth of all installed experiments included. 3 01.26 Reference Core Condition. The condition of the core when it is at ambient temperature (cold T<28°C), the reactivity worth of xenon is negligible (< $0.30) (i.e., cold and clean), and the central irradiation facility contains the graphite thimble plug and the aluminum thimble plug (CIF-1).1.27 Research Reactor. A research reactor is defined as a device designed to support a self-sustaining neutron chain reaction for research development, education, and training, or experimental

purposes, andwhich may have provisions for the production of radioisotopes.

1.28 Rod. Control. A control rod is a device fabricated from neutron absorbing

material, with or without afuel or air follower, which is used to establish neutron flux changes and to compensate for routinereactivity losses. The follower may be a stainless steel section.

A control rod shall be coupled to itsdrive unit to allow it to perform its control function, and its safety function when the coupling isdisengaged. This safety function is commonly termed a scram.1.28.1 Regulating Rod. A regulating rod is a control rod used to maintain an intended powerlevel and may be varied manually or by a servo-controller. A regulating rod shall have scramcapability. 1.28.2 Standard Rod. The regulating and shim rods are standard control rods.1.28.3 Transient Rod. The transient rod is a control rod that is capable of providing rapid-reactivity insertion to produce a pulse or square wave.1.29 Safety Channel. A safety channel is a measuring channel in the reactor safety system.1.30 Safety Limit. Safety limits are limits on important process variables, which are found to benecessary to reasonably protect the integrity of the principal barriers which guard against theuncontrolled release of radioactivity. 1.31 Scram Time. Scram time is the elapsed time between reaching a limiting safety system set pointand the control rods being fully inserted. 1.32 Scram. External. External scrams may arise from the radiography bay doors, radiography bayripcords, bay shutter interlocks, and any scrams from an experiment. 1.33 Shall. Should and May. The word "shall" is used to denote a requirement; the word "should" todenote a recommendation; the word "may" to denote permission, neither a requirement nor arecommendation. 1.34 Shutdown Margin. Shutdown margin shall mean the minimum shutdown reactivity necessary toprovide confidence that the reactor can be made subcritical by means of the control and safety systemstarting from any permissible operating condition with the most reactive rod assumed to be in the mostreactive

position, and once this action has been initiated, the reactor will remain subcritical withoutfurther operator action.1.35 Shutdown.

Unscheduled. An unscheduled shutdown is any unplanned shutdown of theUCD/MNRC reactor caused by actuation of the reactor safety system, operator error, equipment malfunction, or a manual shutdown in response to conditions which could adversely affect safeoperation, not including shutdowns which occur during testing or check-out operations. 1.36 Surveillance Activities. In general, two types of surveillance activities are specified: operability checks and tests, and calibrations. Operability checks and tests are generally specified as djaily, weeklyor quarterly. Calibration times are generally specified as quarterly, semi-annually,

annually, or biennially.

1.37 Surveillance Intervals. Maximum intervals are established to provide operational flexibility and notto reduce frequency. Established frequencies shall be maintained over the long term. The allowable 4 0surveillance interval is the interval between a check, test, or calibration, whichever is appropriate to the itembeing subjected to the surveillance, and is measured from the date of the last surveillance. Allowable surveillance intervals shall not exceed the following: 1.37.1 An nual -interval not to exceed fifteen (15) months.1.37.2 Semiannual -interval not to exceed seven and a half (7.5) months.1.37.3 Quarterly -interval not to exceed four (4) months.1.37.4 Monthly_- interval not to exceed six (6) weeks.1.37.5 Weekly_- interval not to exceed ten (10) days.1.38 Unreviewed Safety Questions. A proposed change, test or experiment shall be deemed to involvean unreviewed safety question:

a. If the probability of occurrence or the consequences of an accident or malfunction ofequipment important to safety previously evaluated in the safety analysis report may beincreased; orb. If a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; orc. If the margin of safety, as defined in the Basis for any technical specification, is reduced.1.39 Value. Measured.

The measured value is the value of a parameter as it appears on the output of achannel.1.40 Value. True. The true value is the actual value of a parameter. 1.41 Watchdog Circuit. The watchdog circuit is a surveillance circuit provided by the Data Acquisition Computer (DAC) and the Control System Computer (CSC) to ensure proper operation of the reactorcomputerized control system.2.0 Safety Limit and Limiting Safety System Setting (LSSS).2.1 Safety Limits.Applicability -This specification applies to the temperature of the reactor fuel in a standard TRIGA fuelelement.Objective -The objective is to define the maximum temperature that can be permitted with confidence that no damage to the fuel element cladding will result..Specification -a. The maximum fuel temperature in a standard TRIGA fuel element shall not exceed 930 °C duringsteady-state operation.

b. The maximum temperature in a standard TRIGA fuel element shall not exceed 1100 0C during pulseoperation.

Basis -a. This fuel safety limit applies for conditions in which the cladding temperature is above 500 °C (SafetyAnalysis Report (SAR), Chapter 4, Section 4.5.4.1.3). The important parameter for a TRIGA reactor is5 0 0the fuel element temperature. This parameter is well suited as it can be measured directly. A loss in theintegrity of the fuel element cladding could arise if the cladding stress exceeds the ultimate strength ofthe cladding material. The fuel element cladding stress is a function of the element's internal pressurewhile the ultimate strength of the cladding material is a function of its temperature. The cladding stressis a result of the internal pressure due to the presence of air, fission product gasses and hydrogen fromthe disassociation of hydrogen and zirconium in the fuel moderator. Hydrogen pressure is the mostsignificant. The magnitude of the pressure is determined by the fuel moderator temperature and theratio of hydrogen to zirconium in the alloy. At a fuel temperature of 930 °C for ZrH1 7 fuel, the claddingstress due to the internal pressure is equal to the ultimate strength of the cladding material at the sametemperature (SAR Fig 4.18). This is a conservative limit since the temperature of the cladding material isalways lower than the fuel temperature. (See SAR Chapter 4, Section 4.5.4.)b. This fuel safety limit applies for conditions in which the cladding temperature is less than 500 °C.Analysis (SAR Chapter 4, Section 4.5.4.1.1), shows that a maximum temperature for the clad during apulse which gives a peak adiabatic fuel temperature of 1000 °C is estimated to be 470 °C. Furtheranalysis (SAR Section 4.5.4.1.2), shows that the internal pressure for both Zr 1.65 (at 11 50°C) and Zr17z (at11 00°C) increases to a peak value at about 0.3 sec, at which time the pressure is about one-fifth of theequilibrium value or about 400 psi (a stress of 14,700 psi). The yield strength of the cladding at 500 °C isabout 59,000 psi.Calculations for step increases in power to peak ZrH 1.65 fuel temperature greater than 1150 °C, over a 200°Crange, show that the time to reach the peak pressure and the fraction of equilibrium pressure value achievedwere approximately the same as for the 1150 °C case. Similar results were found for fuel with ZrH1.7.Measurements of hydrogen pressure in TRIGA fuel elements during transient operations have been made andcompared with the results of analysis similar to that used to make the above prediction. These measurements indicate that in a pulse where the maximum temperature in the fuel was greater than 1000 °C, the pressure(ZrH1.65) was only about 6% of the equilibrium value evaluated at the peak temperature. Calculations of thepressure gave values about three times greater than the measured values. The analysis gives strongindications that the cladding will not rupture if fuel temperatures are never greater than 1200 °C to 1250°C,providing the cladding temperature is less than 500 0 C. For fuel with ZrH 1.7,a conservative safety limit is 1100 °C.As a result, at this safety limit temperature, the class pressure is a factor of 4 lower than would be necessary forcladding failure.2.2 Limiting Safety System Setting.2.2.1 Fuel Temperature. Applicability -This specification applies to the protective action for the reactor fuel elementtemperature. Objective -The objective is to prevent the fuel element temperature safety limit from beingreached.Specification -The limiting safety system setting shall be 750 °C (operationally this may be setmore conservatively) as measured in an instrumented fuel element. One instrumented elementshall be located in the analyzed peak power location of the reactor operational core.Basis -For steady-state operation of the reactor, the limiting safety system setting is atemperature which, if exceeded, shall cause a reactor scram to be initiated preventing the safetylimit from being exceeded. A setting of 750 °C provides a safety margin at the point of themeasurement of at least 137 °C for standard TRIGA fuel elements in any condition of operation. A part of the safety margin is used to account for the difference between the true and measuredtemperatures resulting from the actual location of the thermocouple. If the thermocouple element is located in the hottest position in the core, the difference between the true andmeasured temperatures will be only a few degrees since the thermocouple junction is near thecenter and mid-plane of the fuel element. For pulse operation of the reactor, the same limitingsafety system setting applies.

However, the temperature channel will have no effect on limiting6 0the peak power generated because of its relatively long time constant (seconds) as comparedwith the width of the pulse (milliseconds).

In this mode, however, the temperature trip will act tolimit the energy release after the pulse if the transient rod sho~uld not reinsert and the fueltemperature continues to increase. 3.0 Limiting Conditions For Operation 3.1 Reactor Core Parameters 3.1.1 Steady-State Operation Applicability -This specification applies to the maximum reactor power attained during steady-state operation. Objective -The objective is to assure that the reactor safety limit (fuel temperature) is notexceeded, and to provide for a setpoint for the high flux limiting safety systems, so thatautomatic protective action will prevent the safety limit from being reached during steady-state operation. Specification -The nominal reactor steady-state power shall not exceed 2.0 MW. The automatic scram setpoints for the reactor power level safety channels shall be set at 2.2 MW or less. *Forthe purpose of testing the reactor steady-state power level scram, the power shall not exceed2.3 MW.Basis_- Operational experience and thermal-hydraulic calculations demonstrate that UCD/MNRCTRIGA fuel elements may be safely operated at power levels up to 2.3 MW with naturalconvection cooling. (SAR Chapter 4, Section 4.6.2.)3.1.2 Pulse or Square Wave Operation Applicability -This specification applies to the peak temperature generated in the fuel as theresult of a step insertion of reactivity. Objective -The objective is to assure that the fuel temperature safety limit will not be exceeded. Specification -a. For the pulse mode of operation, the maximum insertion of reactivity shall be 1.23% A k/k($1.75);b. For~the square wave mode of operation, the maximum insertion of reactivity shall be 0.63%Ak/k ($0.90).Basis -Standard TRIGA fuel is .fabricated with a nominal hydrogen to zirconium ratio of 1.6 to1.7. This yields delta phase zirconium hydride which has a high creep strength and undergoes no phase changes at temperatures in excess of 100 °C. However, after extensive steady stateoperation at two (2) MW the hydrogen will redistribute due to migration from the central hightemperature regions of the fuel to the cooler outer regions. When the fuel is pulsed, theinstantaneous temperature distribution is such that the highest values occur at the radial edge ofthe fuel. The higher temperatures in the outer regions occur in fuel with a hydrogen to zirconium ratio that has now increased above the nominal value. This produces hydrogen gas pressures considerably in excess of that expected. If the pulse insertion is such that the temperature of thefuel exceeds about 875 °C, then the pressure may be sufficient to cause expansion ofmicroscopic holes in the fuel that grow with each pulse. Analysis (SAR Chapter 13, Section13.2.2.2.1), shows that the limiting pulse, for the worst case conditions, is 1.34% A k/k ($1.92).Therefore, the 1.23% A k/k ($1.75) limit is below the worse case reactivity insertion accident limit.7 The $0.90 square wave step insertion limit is also well below the worse case reactivity insertion accident limit.3.1.3 Reactivity Limitations Applicability -These specifications apply to the reactivity conditions of the reactor core and thereactivity worths of the control rods and apply to all modes of reactor operation. Objective -The objective is to assure that the reactor can be placed in a shutdown condition atall times and to assure that the safety limit shall not be exceeded. Specification -a. Shutdown Margin -The reactor shall not be operated unless the shutdown margin provided bythe control rods is greater than 0.35% A k/k ($0.50) with:(1) The reactor in any core condition, (2) The most reactive control rod assumed fully withdrawn, and(3) Absolute value of all movable experiments analyzed in their most reactive condition or $1.00 whichever is greater.b. Excess Reactivity -The maximum available excess reactivity (reference core condition) shallnot exceed 6.65% A k/k ($9.50).Basis -a. This specification assures that the reactor can be placed in a shutdown condition from anyoperating condition and remain shutdown, even if the maximum worth control rod should stick inthe fully withdrawn position (SAR Chapter 4, Section 4.5.5).b. This specification sets an overall reactivity limit which provides adequate excess reactivity tooverride the xenon buildup, to overcome the temperature change in going from zero power to 2MW, to permit pulsing at the $1.75 level, to permit irradiation of negative worth experiments andaccount for fuel burnup over time. An adequate shutdown margin exists with an excess of $9.50for the two analyzed cores: (SAR Chapter 4, Section 4.5.5).3.2 Reactor Control and Safety Systems3.2.1 Control RodsApplicability -This specification applies to the function of the control rods.Objective -The objective is to determine that the control rods are operable. Specification -The reactor shall not be operated unless the control rods are operable and,a. Control rods shall not be considered operable if damage is apparent to the rod or driveassemblies.

b. The scram time measured from the instant a signal reaches the value of a limiting safetysystem setting to the instant that the slowest control rod reaches its fully inserted position shallnot exceed one (1) second.8 Basis -a. The apparent condition of the control rod assemblies shall provide assurance that the rodsshall continue to perform reliably as designed.

b. This assures that the reactor shall shut down promptly when a scram signal is initiated (SARChapter 13, Section 13.2.2.2.2).

3.2.2 Reactor Instrumentation Applicability -This specification applies to the information which shall be available to the reactoroperator during reactor operations. Objective -The objective is to require that sufficient information is available to the operator toassure safe operation of the reactor.Specification -The reactor shall not be operated unless the channels described in Table 3.2.2are operable and the information is displayed on the reactor console.Table 3.2.2Required Reactor Instrumentation (Minimum Number Operable) Measuring Steady Square Channel Surveillance Channel State Pulse Wave Function Requirements*

a. Reactor Power 2 0 2 Scram at 2.2 D,M,ALevel Safety MW or lessChannelb. Linear Power 10 1Automatic D,M,AChannel Power Controlc. Log Power 10 1Startup D,M,AChannel Controld. Fuel Temperature 2 2 2Fuel D,M,AChannel Temperature

e. Pulse Channel 0 10Measures P,APulse NV & NVT(*) Where: 0 -Channel check during each day's operation M -Channel test monthlyA -Channel calibration annuallyP -Channel test prior to pulsing operation Basis -a. Table 3.2.2. The two reactor power level safety channels assure that the reactor power levelis properly monitored and indicated in the reactor control room (SAR Chapter 7, Sections 7.1.2 &7.1.2.2).

b. c. & e. Table 3.2.2. The linear power channel, log power channel, and pulse channel assurethat the reactor power level and energy are adequately monitored (SAR Chapter 7, Sections7.1.2 & 7.1.2.2).

9

d. Table 3.2.2. The fuel temperature channels assure that the fuel temperature is properlymonitored and indicated in the reactor control room (SAR Chapter 4, Section 4.5.4.1).

3.2.3 Reactor Scrams and Interlocks Applicability -This specification applies to the scrams and interlocks. Objective -The objective is to assure that the reactor is placed in the shutdown condition promptly and that the scrams and interlocks are operable for safe operation of the reactor.Specification -The reactor shall not be operated unless the scrams and interlocks described inTable 3.2.3 are operable: Table 3.2.3Required Scrams and Interlocks Scram.a. ConsoleManualScramb. Reactor RoomManual Scramc. Radiography Bay ManualScramsd. Reactor PowerLevel SafetyScramse. High VoltagePower SuppliesScramsf. FuelTemperature Scramsg. WatchdogCircuitSteadyState142Pulse1402SquareWave1222Manual Scramand Automatic Scram AlarmManual Scramand Automatic Scram AlarmManual Scramsand Automatic Scram AlarmsAutomatic Scram Alarms & Scramsat 2.2 MW or lessAutomatic Scram Alarms &Scrams on Loss ofHigh Voltage tothe Reactor PowerLevel SafetyChannelsAutomatic ScramAlarms & Scramson indicated fueltemperature of750°C or lessAutomatic ScramAlarms & ScramsMMMMMMMChannelFunctionSurveillance Requirements* 22210

h. ExternalScramsi. One KilowattPulse &Square WaveInterlock

j. Low SourceLevel RodWithdrawal ProhibitInterlock

k. Control RodWithdrawal Interlock I. MagnetPower KeySwitch Scram220IIIAutomatic Scrams and Alarmsif an experiment or radiography scram interlock is activated Prevents initiation of a step reactivity insertion above areactor power levelof I KWPrevents withdrawal of any control rodif the log channelreads less than 1.5times the indicated log channel currentlevel with the neutronsource removed fromthe corePrevents simul-taneous withdrawal of two or more rodsin manual modeDe-energizes thecontrol rodmagnets, scram &alarmMMMMMIIII(*) Where: M -channel test monthlyBasis -a. Table 3.2.3. The console manual scram allows rapid shutdown of the reactor from the controlroom (SAR Chapter 7, Section 7.1.2.5).

b. Table 3.2.3. The reactor room manual scram allows rapid shutdown of the reactor from thereactor room.c. Table 3.2.3. The radiography bay manual scrams allow rapid shutdown of the reactor fromany of the radiography bays (SAR Chapter 9, Section 9.6.3).d. Table 3.2.3. The automatic power level safety scram assures the reactor will be shutdown ifthe power level exceeds 2.2 MW, therefore not exceeding the safety limit (SAR Chapter 4,Section 4.7.2).e. Table 3.2.3. The loss-of-high-voltage scram assures that the reactor power level safetychannels operate within their intended range as required for proper functioning of the powerlevel scrams (SAR Chapter 7, Sections 7.1.2.1 & 7.1.2.2).

f. Table 3.2.3. The fuel temperature scrams assure that the reactor will be shut down if the fueltemperature exceeds 7500° C, therefore ensuring the safety limit will not be exceeded (SARChapter 4, Sections 4.5.4.1 & 4.7.2).11

g. Table 3.2.3. The watchdog circuits assure that the control system computer and the dataacquisition computer are functioning properly (SAR Chapter 7, Section 7.2).h. Table 3.2.3. The external scrams assure that the reactor will be shut down if the radiography bay doors and reactor concrete shutters are not in the proper position for personnel entry intothe bays (SAR Chapter 9, Section 9.6). External scrams from experiments, a subset of theexternal scrams, also assure the integrity of the reactor system, the experiment, the facility, andthe safety of the facility personnel and the public.i. Table 3.2.3. The interlock preventing the initiation of a step reactivity insertion at a level aboveone (1) kilowatt assures that the pulse magnitude will not allow the fuel element temperature toexceed the safety limit (SAR Chapter 7, Section 7.1.2.5).

j.Table 3.2.3. The low source level rod withdrawal prohibit interlock assures an adequate sourceof neutrons is present for safe startup of the reactor (SAR Chapter 7, Section 7.1.2.5).

k. Table 3.2.3. The control rod withdrawal interlock prevents the simultaneous withdrawal of twoor more control rods thus limiting the reactivity-insertion rate from the control rods in manualmode (SAR Chapter 7, Section 7.1.2.5).

I. Table 3.2.3. The magnet current key switch prevents the control rods from being energized without inserting the key. Turning off the magnet current key switch de-energizes the control rodmagnets and results in a scram (SAR Chapter 7, Section 7.1.2.5). 3.2.4 Reactor Fuel ElementsApplicability -This specification applies to the physical dimensions of the fuel elements asmeasured on the last surveillance test.Objective -The objective is to verify the integrity of the fuel-element cladding. Specification -The reactor shall not be used for normal operation with damaged fuel. All fuelelements shall be inspected visually for damage or deterioration as per Technical Specifications Section 4.2.4. A fuel element shall be considered damaged and must be removed from the coreif:a. In measuring the transverse bend, the bend exceeds 0.125 inch (3.175 mm) over the fulllength 23 inches (584 mm) of the cladding, or,b. In measuring the elongation, its length exceeds its initial length by 0.125 inch (3.175 mm), or,c. A cladding failure exists as indicated by measurable release of fission products, or,d. Visual inspection identifies bulges, gross pitting, or corrosion. Basis -The most severe stresses induced in the fuel elements result from pulse operation of thereactor, during which differential expansion between the fuel and the cladding occurs and thepressure of the gases within the elements increases sharply. The above limits on the allowable distortion of a fuel element correspond to strains that are considerably lower than the strainexpected to cause rupturing of a fuel element. Limited operation in the steady state or pulsedmode may be necessary to identify a leaking fuel element especially if the leak is small.12 3.3 Reactor Coolant SystemsApplicability -These specifications apply to the operation of the reactor water measuring systems.Objective -The objective is to assure that adequate cooling is provided to maintain fuel temperatures below the safety limit, and that the water quality remains high to prevent damage to the reactor fuel.Specification -The reactor shall not be operated unless the systems and instrumentation channelsdescribed in Table 3.3 are operable, and the information is displayed locally or in the control room.Table 3.3REQUIRED WATER SYSTEMS AND INSTRUMENTATION Measuring Channel/System

a. Primary CoolantCore InletTemperature Monitorb. Reactor TankLow WaterMonitorc. Purification**

Inlet Conduc-tivity Monitord. Emergency CoreCooling SystemMinimumNumberOperableSurveillance Requirements* 1IIIFunction: Channel/System For operation of thereactor at 1.5 MW orhigher, alarms on highheat exchanger outlettemperature of 45 °C(113°F)Alarms if water leveldrops below a depth of23 feet in the reactor tankAlarms if the primarycoolant water conductivity is greater than5 micromhos/cm For operation of the reactorat 1.5MW or higher, provideswater to cool fuel in the eventof a Loss of Coolant Accidentfor a minimum of 3.7 hoursat 20 gpm from an appropriate nozzleD,Q,AMD,M,SD,S(*) Where: D -channel check during each day's operation A -channel calibration annuallyQ -channel test quarterly S -channel calibration semiannually M -channel test monthly(**) The purification inlet conductivity monitor can be out-of-service for no more than 3 hours before the reactorshall be shutdown. Basis -a. Table 3.3. The primary coolant core inlet temperature alarm assures that large power fluctuations willnot occur (SAR Chapter 4, Section 4.6.2).13

b. Table 3.3. The minimum height of 23 ft. of water above the reactor tank bottom guarantees that thereis sufficient water for effective cooling of the fuel and that the radiation levels at the top of the reactortank are within acceptable limits. The reactor tank water level monitor alarms if the water level dropsbelow a height of 23 ft. (7.01m) above the tank bottom (SAR Chapter 11, Section 11.1.5.1).

c. Table 3.3. Maintaining the primary coolant water conductivity below 5 micromhos/cm averaged over aweek will minimize the activation of water impurities and also the corrosion of the reactor structure.

d. Table 3.3. This system will mitigate the Loss of Coolant Accident event analyzed in the SAR Chapter13, Section 13.2.3.4 Reactor Room Exhaust SystemApplicability

-These specifications apply to the operation of the reactor room exhaust system.Objective -The objectives of this specification are as follows:a. To reduce concentrations of airborne radioactive material in the reactor room, and maintain thereactor room pressure negative with respect to surrounding areas.b. To assure continuous air flow through the reactor room in the event of a Loss of Coolant Accident. S~pecification -a. The reactor shall not be operated unless the reactor room exhaust system is in operation and thepressure in the reactor room is negative relative to surrounding areas.b. The reactor room exhaust system shall be operable within one half hour of the onset of a Loss ofCoolant Accident. Basis -Operation of the reactor room exhaust system assures that:a. Concentrations of airborne radioactive material in the reactor room and in air leaving the reactor roomwill be reduced due to mixing with exhaust system air (SAR Chapter 9, Section 9.5.1). Pressure in thereactor room will be negative relative to surrounding areas due to air flow patterns created by the reactorroom exhaust system (SAR Chapter 9, Section 6.5.1).b. There will be a timely, adequate and continuous air flow through the reactor room to keep the fueltemperature below the safety limit in the event of a Loss of Coolant Accident. 3.5 This section intentionally left blank.3.6 This section intentionally left blank.3.7 Reactor Radiation Monitoring Systems3.7.1 Monitoring SystemsApplicability -This specification applies to the information which shall be available to the reactoroperator during reactor operation. Obiective -The objective is to require that sufficient information regarding radiation levels andradioactive effluents is available to the reactor operator to assure safe operation of the reactor.Specification -The reactor shall not be operated unless the channels described in Table 3.7.1are operable, the readings are below the alarm setpoints, and the information is displayed in the14

• 0control room. The stack and reactor room CAMS shall not be shutdown at the same time duringreactor operation.

Table 3.7.1REQUIRED RADIATION MONITORING INSTRUMENTATION Measuring Equipment MinimumNumberOperable** ChannelFunctionSurveillance Requirements*

a. FacilityStack Monitorb. Reactor RoomRadiation Monitorc. Purification System Radia-tion Monitord. Reactor RoomContinuous Air MonitorIIIIMonitors Argon-41 andradioactive particu-lates, and alarmsMonitors the radiation level in the reactorroom and alarmsMonitors radiation level at the demineral-izer station and alarmsMonitors air from thereactor room for parti-culate and gaseousradioactivity and alarmsD,W,AD,W,AD,W,AD,W,A(*) Where: D -channel check during each day's operation A -channel calibration annuallyW -channel test(**) monitors may be placed out-of-service for up to 2 hours for calibration and maintenance.

During this out-of-service time, no experiment or maintenance activities shall be conducted which could result in alarm conditions (e.g., airborne releases or high radiation levels)Basis -a. Table 3.7.1. The facility stack monitor provides information to operating personnel regarding therelease of radioactive material to the environment (SAR Chapter 11, Section 11.1.1.1.4). The alarmsetpoint on the facility stack monitor is set to limit Argon-41 concentrations to less than 10 CFR Part 20,Appendix B, Table 2, Column 1 values (averaged over one year) for unrestricted locations outside theoperations area.b. Table 3.7.1. The reactor room radiation monitor provides information regarding radiation levels in thereactor room during reactor operation (SAR Chapter 11, Section 11.1.5.1), to limit Occupational radiation exposure to less than 10 CFR 20 limits.c. Table 3.7.1. The radiation monitor located next to the purification system resin cannisters providesinformation regarding radioactivity in the primary system cooling water (SAR Chapter 11, Section11.1.5.4.2) and allows assessment of radiation levels in the area to ensure that personnel radiation doses will be below 10 CFR Part 20 limits.d. Table 3.7.1. The reactor room continuous air monitor provides information regarding airborneradioactivity in the reactor room, (SAR Chapter 11, Sections 11.1.1.1.2 & 11.1.1.1.5), to ensure thatoccupational exposure to airborne radioactivity will remain below the 10 CFR Part 20 limits.15 3.7.2 Effluents -Arqon-41 Dischargle LimitApplicability -This specification applies to the concentration of Argon-41 that may be discharged from the UCD/MNRC reactor facility. Objective -The objective is to ensure that the health and safety of the public is not endangered by the discharge of Argon-4l from the UCD/MNRC reactor facility. Specification -The annual average unrestricted area concentration of Argon-41 due to releasesof this radionuclide from the UCD/MNRC, and the corresponding annual radiation dose fromArgon-4l in the unrestricted area shall not exceed the applicable levels in 10 CFR Part 20.Basis -The annual average concentration limit for Argon-41 in air in the unrestricted area isspecified in Appendix B, Table 2, Column 1 of 10 CFR Part 20. 10 CFR 20.1301 specifies doselimitations in the unrestricted area. 10 CFR 20.1101 specifies a constraint on air emissions ofradioactive material to the environment. The SAR Chapter 11, Section 11.1.1.1.4 estimates thatthe routine Argon-4l releases and the corresponding doses in the unrestricted area will bebelow these limits.3.8 Exp~eriments 3.8.1 Reactivity Limits.Applicability -This specification applies to the reactivity limits on experiments installed in specificreactor experiment facilities. Objective -The objective is to assure control of the reactor during the irradiation or handling ofexperiments in the specifically designated reactor experiment facilities. Specification -The reactor shall not be operated unless the following conditions governing experiments exist:a. The absolute reactivity worth of any single moveable experiment in the pneumatic transfertube, the central irradiation

facility, the central irradiation fixture 1 (CIF-1),

or any other in-core orin-tank irradiation

facility, shall be less than $1.00 (0.7% A k/k), except for the automated centralirradiation facility (ACIF) (See 3.8.1.c below).b. The absolute reactivity worth of any single secured experiment positioned in a reactor in-coreor in-tank irradiation facility shall be less than the maximum allowed pulse ($1.75) (1.23% A k/k).c. The absolute total reactivity worth of any single experiment or of all experiments collectively positioned in the ACIF shall be less than the maximum allowed pulse ($1.75) (1.23% A k/k).d. The absolute total reactivity of all experiments positioned in the pneumatic transfer tube, andin any other reactor in-core and in-tank irradiation facilities at any given time shall be less thanone dollar and ninety-two cents ($1.92) (1.34% A k/k), including the potential reactivity whichmight result from malfunction,

flooding, voiding, or removal and insertion of the experiments.

Basis -a. A limitation of less than one dollar ($1.00) (0.7% A k/k) on the reactivity worth of a singlemovable experiment positioned in the pneumatic transfer tube, the central irradiation facility(SAR, Chapter 10, Section 10.4.1), the central irradiation fixture-I (ClF-1) (SAR Chapter 10,Section 10.4.1), or any other in-core or in-tank irradiation

facility, will assure that the pulse limitof $1.75 is not exceeded (SAR Chapter 13, Section 13.2.2.2.1).

In addition, limiting the worth ofeach movable experiment to less than $1.00 will assure that the additional increase in transient 16

• 0power and temperature will be slow enough so that the fuel temperature scram will be effective (SAR Chapter 13, Section 13.2.2.2.1

).b. The absolute worst event which may be considered in conjunction with a single securedexperiment is its sudden accidental or unplanned removal while the reactor is operating. Forsuch an event, the reactivity limit for fixed experiments ($1.75) would result in a reactivity increase less than the $1.92 pulse reactivity insertion needed to reach the fuel temperature safety limit (SAR Chapter 13, Section 13.2.2.2.1 ).c. A reactivity limit of less than $1.75 for any single experiment or for all experiments collectively positioned in the sample can of the automated central irradiation facility (ACIF) (SAR Chapter10, Section 10.4.2) is based on the pulsing reactivity insertion limit (Technical Specification 3.1.2) (SAR Chapter 13, Section 13.2.2.2.1) and on the design of the ACIF, which allows controlover the positioning of samples into and out of the central core region in a manner identical inform, fit, and function to a control rod.d. It is conservatively assumed that simultaneous removal of all experiments positioned in thepneumatic transfer tube, and in any other reactor in-core and in-tank irradiation facilities at anygiven time shall be less thanthe maximum reactivity insertion limit of $1.92. The SAR Chapter13, Section 13.2.2.2.1 indicates that a pulse reactivity insertion of $1.92 would be needed toreach the fuel temperature safety limit.3.8.2 Materials LimitApplicability -This specification applies to experiments installed in reactor experiment facilities. Objective -The objective is to prevent damage to the reactor or significant releases ofradioactivity by limiting material quantity and the radioactive material inventory of theexperiment. Specification -The reactor shall not be operated unless the following conditions governing experiment materials exist:a. Experiments containing materials corrosive to reactor components, compounds highlyreactive with water, potentially explosive materials, and liquid fissionable materials shall beappropriately encapsulated.

b. Each fueled experiment shall be controlled such that the total inventory of iodine isotopes 131through 135 in the experiment is no greater than 1.5 curies and the maximum strontium inventory is no greater than 5 millicuries.

c. Each experiment in the 1-125 production facility shall be controlled such that the totalinventory of 1-125 in the 1-125 glove box shall not exceed 40 curies.d. Each experiment in the 1-125 production facility shall be controlled such that the totalinventory of 1-125 being handled in the 1-125 fume hood at any one time in preparation forshipping shall not exceed 20 curies. An Additonal 1.0 curie of 1-125 (up to 400 millicuries in theform of quality assurance samples and up to 600 millicuries in sealed storage containers) mayalso be present in the 1-125 fume hood.e. Explosive materials in quantities greater than 25 milligrams of TNT equivalent shall not beirradiated in the reactor tank. Explosive materials in quantities of 25 milligrams of TNTequivalent or less may be irradiated provided the pressure produced upon detonation of theexplosive has been calculated and/or experimentally demonstrated to be less than the designpressure of the container.

f. Explosive materials in quantities of three (3) pounds of TNT equivalent or less may beirradiated in any radiography bay. The irradiation of explosives in any bay is limited to those1"7 assemblies where a safety analysis has been performed that shows that there is no damage tothe reactor safety systems upon detonation (SAR Chapter 13, Section 13.2.6.2).

Basis -a. Appropriate encapsulation is required to lessen the experimental hazards of some types ofmaterials.

b. The 1.5 curies limitation on iodine 131 through 135 assures that in the event of failure of afueled experiment leading to total release of the iodine, occupational doses and doses tomembers of the general public in the unrestricted areas shall be within the limits in 10 CFR 20(SAR Chapter 13, Section 13.2.6.2).

c&d. Limiting the total 1-125 inventory to forty (40.0) curies in the 1-125 glove box and to twenty-one (21.0) curies in the 1-125 fume hood assures that, if either of these inventories of 1-125 istotally released into its respective containment, or if both inventories are simultaneously released into their respective containments, the occupational doses and doses to members ofthe general public in the unrestricted areas will be within the limits of 10 CFR 20 (SAR Chapter13, Section 13.2.6.2).

e. This specification is intended to prevent damage to vital equipment by restricting the quantityof explosive materials within the reactor tank (SAR Chapter 13, Section 13.2.6.2).

f. The failure of an experiment involving the irradiation of 3 lbs TNT equivalent or less in anyradiography bay external to the reactor tank will not result in damage to the reactor controls orthe reactor tank. Safety Analyses have been performed (SAR Chapter 13, Section 13.2.6.2) which show that up to six (6) pounds of TNT equivalent can be safely irradiated in anyradiogaphy bay. Therefore, the three (3) pound limit gives a safety margin of two (2).3.8.3 Failure and Malfunctions Applicability

-This specification applies to experiments installed in reactor experiment facilities. Objective -The objective is to prevent damage to the reactor or significant releases ofradioactive materials in the event of an experiment failure.Specification -a. All experiment materials which could off-gas,

sublime, volatilize, or produce aerosols under:(1) normal operating conditions of the experiment or the reactor,(2) credible accident conditions in the reactor, or(3) where the possibility exists that the failure of an experiment could release radioactive gases or aerosols into the reactor building or into the unrestricted area, the quantity andtype of material in the experiment shall be limited such that the airborne radioactivity inthe reactor room will not result in exceeding the applicable dose limits in 10 CFR Part 20in the unrestricted area, assuming 100% of the gases or aerosols escapes.b. In calculations pursuant to (a) above, the following assumptions shall be used:(1) If the effluent from an experiment facility exhausts through a stack which is closed onhigh radiation levels, at least 10% of the gaseous activity or aerosols produced willescape.18 (2) If the effluent from an experiment facility exhausts through a filter installation designed for greater than 99% efficiency for 0.3 micron and larger particles, at least 10%of these will escape.(3) For materials whose boiling point is above 130 °C and where vapors formed byboiling this material can escape only through an undistributed column of water above thecore, at least 10% of these vapors can escape.c. If a capsule fails and releases material which could damage the reactor fuel or structure bycorrosion or other means, an evaluation shall be made to determine the need for corrective action. Inspection and any corrective action taken shall be reviewed by the UCD/MNRC Directoror his designated alternate and determined to be satisfactory before operation of the reactor isresumed.Basis -a. This specification is intended to reduce the likelihood that airborne radioactivity in the reactorroom or the unrestricted area will result in exceeding the applicable dose limits in 10 CFR Part20.b. These assumptions are used to evaluate the potential airborne radioactivity release due to anexperiment failure (SAR Chapter 13, Section 13.2.6.2).

c. Normal operation of the reactor with damaged reactor fuel or structural damage is prohibited to avoid release of fission products.

Potential damage to reactor fuel or structure shall bebrought to the attention of the UCD/MNRC Director or his designated alternate for review toassure safe operation of the reactor (SAR Chapter 13, Section 13.2.6.2). 4.0 Surveillance Requirements General. The surveillance frequencies denoted herein are based on continuing operation of the reactor.Surveillance activities scheduled to occur during an operating cycle which can not be performed with the reactoroperating may be deferred to the end of the operating cycle. If the reactor is not operated for a reasonable time,a reactor system or measuring channel surveillance requirement may be waived during the associated timeperiod. Prior to reactor system or measuring channel operation, the surveillance shall be performed for eachreactor system or measuring channel for which surveillance was waived. A reactor system or measuring channel shall not be considered operable until it is successfully tested.4.1 Reactor Core Parameters 4.1 .1 Steady State Operation Applicability -This specification applies to the surveillance requirement for the power levelmonitoring channels. Obiective -The objective is to verify that the maximum power level of the reactor does notexceed the authorized limit.Specification -An annual channel calibration shall be made of the power level monitoring channel. If a channel is removed,

replaced, or unscheduled maintenance is performed, or asignificant change in core configuration occurs, a channel calibration shall be required.

Discovery of noncompliance with this specification shall limit reactor operations to that requiredto perform the surveillance. Basis -The annual power level channel calibration will assure that the indicated reactor powerlevel is correct.4.1.2 Shutdown Margin and Excess Reactivity 19

• 0Applicability

-These specifications apply to the surveillance requirements for reactivity control ofthe reactor core.Objective -The objective is to measure and verify the reactivity worth, performance, andoperability of those systems affecting the reactivity of the reactor.Specification -a. The total reactivity worth of each control rod and the shutdown margin shall be determined annually or following any significant change in core or control rod configuration. The shutdownmargin shall be verified by meeting the requirements of Section 3.1.3(a).

b. The core excess reactivity shall be verified:

(1) Prior to each startup operation and,(2) Following any change in core loading or configuration. Discovery of noncompliance with Technical Specifications 4.1 .2.a-b shall limit reactor operations to that requiredto perform the surveillance. Basis -a. The reactivity worth of the control rods is measured to assure that the required shutdownmargin is available and to provide an accurate means for determining the excess reactivity ofthe core. Past experience with similar reactors gives assurance that measurements of thecontrol rod reactivity worth on an annual basis is adequate to assure that there are no significant changes in the shutdown margin, provided no core loading or configuration changes have beenmade.b. Determining the core excess reactivity prior to each reactor startup shall assure that Technical Specifications 3.1.3.b shall be met, and that the critical rod positions do not changeunexpectedly. 4.2 Reactor Control and Safety Systems4.2.1 Control RodsApplicability -This specification applies to the surveillance of the control rods.Objective -The objective is to inspect the physical condition of the reactor control rods andestablish the operable condition of the rods.Specification -Control rod worths shall be determined annually or after physical removal or anysignificant change in core or control rod configuration.

a. Each control rod shall be inspected at annual intervals by visual observation of the fueledsections and absorber sections plus examination of the linkages and drives.b. The scram time of each control rod shall be measured semiannually.

Discovery of noncompliance with Technical Specifications 4.2.1 .a-b shall limit reactor operations to that requiredto perform the surveillance. Basis (Technical Specifications 4.2.1 .a-b) -Annual determination of control rod worths ormeasurements after any physical removal or significant change in core loading or control rod20 configuration provides information about changes in reactor total reactivity and individual rodworths. The frequency of inspection for the control rods shall provide periodic verification of thecondition of the control rod assemblies. The specification intervals for scram time assureoperable performance of the control rods.4.2.2 Reactor Instrumentation Applicability -These specifications apply to the surveillance requirements for measurements, tests, calibration and acceptability of the reactor instrumentation. .Objective -The objective is to ensure that the power level instrumentation and the fueltemperature instrumentation are operable. Specification -a. The reactor power level safety channels shall have the following: (1) A channel test monthly or after any maintenance which could affect their operation. (2) A channel check during each day's operation. (3) A channel calibration annually.

b. The Linear Power Channel sh'all have the following:

(1) A channel test monthly or after any maintenance which could affect the operation. (2) A channel check during each day's operation. (3) A channel calibration annually.

c. The Log Power Channel shall have the following:

(1) A channel test monthly or after any maintenance which could affect its operation. (2) A channel check during each day's operation. (3) A channel calibration annually.

d. The fuel temperature measuring channels shall have the following:

(1) A channel test monthly or after any maintenance which could affect operation. (2) A channel check during each day's operation. (3) A channel calibration annually.

e. The Pulse Energy Integrating Channel shall have the following:

(1) A channel test prior to PUlsing operations. (2) A channel calibration annually. Discovery of noncompliance with Technical Specifications 4.2.2.a-e shall limit reactor operation to that requiredto perform the surveillance. Basis -21

a. A daily channel check and monthly test, plus the annual calibration, will assure that thereactor power level safety channels operate properly.

b. A channel test monthly of the reactor power level multi-range channel will assure that thechannel is operable and responds correctly.

The channel check will assure that the reactorpower level multi-range linear channel is operable on a daily basis. The channel calibration annually of the multi-range linear channel will assure that the reactor power will be accurately measured so the authorized power levels are not exceeded.

c. A channel test monthly will assure that the reactor power level wide range log channel isoperable and responds correctly.

A channel check of the reactor power level wide range logchannel will assure that the channel is operable on a daily basis. A channel calibration willassure that the channel will indicate properly at the corresponding power levels.d. A channel test monthly and check during each day's operation, plus the annual calibration, willassure that the fuel temperature measuring channels operate properly.

e. A channel test prior to pulsing plus the annual channel calibration will assure the pulse energyintegrating channel operates properly.

4.2.3 Reactor Scrams and Interlocks Applicability -These specifications apply to the surveillance requirements for measurements, test, calibration, and acceptability of the reactor scrams and interlocks. Objective -The objective is to ensure that the reactor scrams and interlocks are operable. Specification -a. Console Manual Scram. A channel test shall be performed monthly.b. Reactor Room Manual Scram. A channel test shall be performed monthly.c. Radiography Bay Manual Scrams. A channel test shall be performed monthly.d. Reactor Power Level Safety Scram. A channel test shall be performed monthly.e. High-Voltage-Power Supply Scrams. A channel test shall be performed monthly.f. Fuel Temperature Scram. A channel test shall be performed monthly.g. Watchdog Circuits Scrams. A channel test shall be performed monthly.h. External Scra~ns. A channel test shall be performed monthly.i.The One Kilowatt Pulse Interlock. A channel test shall be performed monthly.j. Low Source Level Rod Withdrawal Prohibit Interlock. A channel test shall be performed monthly.k. Control Rod Withdrawal Interlocks. A channel test shall be performed monthly.I. Magnet Power Key Switch Scram. A channel test shall be performed monthly.Discovery of noncompliance with Specifications 4.2.3.a-I shall limit reactor operation to that required to performthe surveillance. Basis -22 0a. A channel test monthly of the Console Manual Scram will assure that the scram is operable.

b. A channel test monthly of the Reactor Room Manual Scram will assure that the scram isoperable.

c. A channel test monthly of the Radiography Bay Manual Scrams will assure that the scramsare operable.

d. A channel test monthly of the Reactor Power Level Safety Scrams will assure that the scramsare operable.

e. A channel test monthly of the Loss-of-High-Voltage Scram will assure that the high voltagepower supplies are operable and respond correctly.

f. A channel test monthly of the Fuel Temperature Scrams will assure that the scrams areoperable.

g. A channel test monthly of the Watchdog Circuits Scrams will assure that the scram circuits areoperable.

h. A channel test monthly of the External Scrams will assure that the scrams are operable andrespond correctly.

i.A channel test monthly will assure that the One Kilowatt Pulse Interlock works properly.

j. A channel test monthly of the Low Source Level Rod Withdrawal Proh~ibit Interlock will assurethat the interlock is operable.

k. A channel test monthly of the Control Rod Withdrawal Interlock will assure that the interlock isoperable.

I. A channel test monthly of the Magnet Current Key Switch will assure that the scram isoperable. 4.2.4 Reactor Fuel ElementsApplicability -This specification applies to the surveillance requirements for the fuel elements. Objective -The objective is to verify the continuing integrity of the fuel element cladding. Specification -To assure the measurement limitations in Section 3.2.4 are met, the following shall be done:a. The lead elements (i.e., all elements adjacent to the transient rod, with the exception ofinstrumented fuel elements), and all elements adjacent to the central irradiation facility shall beinspected annually.

b. Instrumented fuel elements shall be inspected if any of the elements adjacent to it fail to passthe visual and/or physical measurement requirements of Section 3.2.4. Discovery ofnoncompliance with Technical Specification 4.2.4 shall limit operations to that required toperform the surveillance.

Basis (Technical Specifications 4.2.4.a-b) -The above specifications assure that the lead fuelelements shall be inspected regularly and the integrity of the lead fuel elements shall bemaintained. These are the fuel elements with the highest power density as analyzed in the SARChapter 4, Section 4.5.5.6. The instrumented fuel element is excluded to reduce the risk ofdamage to the thermocouples. 23 .0*4.3 Reactor Coolant SystemsApplicability -This specification applies to the surveillance requirements for the reactor water measuring systems and the emergency core cooling system.Objective -The objective is to assure that the reactor tank water temperature monitoring system, thetank water level alarm, the water conductivity cells and the emergency core cooling system are alloperable. Specification -a. The reactor tank core inlet temperature monitor shall have the following: (1) A channel check during each day's operation. (2) A channel test quarterly. (3) A channel calibration annually.

b. The reactor tank low water level monitoring system shall have the following:

(1) A channel test monthly.c. The purification inlet conductivity monitors shall have the following: (1) A channel check during each day's operation. (2) A channel test monthly.(3) A channel calibration semiannually.

d. The Emergency Core Cooling System shall have the following:

(1) A channel check prior to operation. (2) A channel calibration semiannually. Discovery of noncompliance with Technical Specifications 4.3.a-c shall limit operations to that required toperform the surveillance. Noncompliance with Technical Specification 4.3.d shall limit operations to less than 1.5MW.Basis -a. A channel test quarterly assures the water temperature monitoring system responds correctly to aninput signal. A channel check during each day's operation assures the channel is operable. A channelcalibration annually assures the monitoring system reads properly.

b. A channel test monthly assures that the low water level monitoring system responds correctly to aninput signal.c. A channel test monthly assures that the purification inlet conductivity monitors respond correctly to aninput signal. A channel check during each day's operation assures that the channel is operable.

Achannel calibration semiannually assures the conductivity monitoring system reads properly.

d. A channel check prior to operation assures that the emergency core cooling system is operable forpower levels above 1.5 MW. A channel calibration semiannually assures that the Emergency CoreCooling System performs as required for power levels above 1.5 MW.24 04.4 Reactor Room Exhaust SystemApplicability

-This specification applies to the surveillance requirements for the reactor roomexhaust system.Objective -The objective is to assure that the reactor room exhaust system is operating properly. Specification -The reactor room exhaust system shall have a channel check during each day'soperation. Discovery of noncompliance with this specification shall limit operations to that required to perform thesurveillance. Basis -A channel check during each day's operation of the reactor room exhaust system shallverify that the exhaust system is maintaining a negative pressure in the reactor room relative tothe surrounding facility areas.4.5 This section intentionally left blank4.6 This section intentionally left blank.4.7 Reactor Radiation Monitoring SystemsApplicability -This specification applies to the surveillance requirements for the reactor radiation monitoring systems.Obiective -The objective is to assure that the radiation monitoring equipment is operating properly. Specification -a. The facility stack monitor shall have the following: (1) A channel check during each day's operation. (2) A channel test weekly.(3) A channel calibration annually.

b. The reactor room radiation monitor shall have the following:

(1) A channel check during each day's operation. (2) A channel test weekly.(3) A channel calibration annually.

c. The purification system radiation monitor shall have the following:

(1) A channel check during each day's operation. (2) A channel test weekly.(3) A channel calibration annually.

d. The reactor room Continuous Air Monitor (CAM) shall have the following:

25 (1) A channel check during each day's operation. (2) A channel test weekly.(3) A channel calibration annually. Discovery of noncompliance with Technical Specifications 4.7.a-d shall limit operations to that required toperform the surveillance. Basis -a. A channel check of the facility stack monitor system during each day's operation will assurethe monitor is operable. A channel test weekly will assure that the system responds correctly toa known source. A channel calibration annually will assure that the monitor reads correctly.

b. A channel check of the reactor room radiation monitor during each day's operation will assurethat the monitor is operable.

A channel test weekly will ensure that the system responds to aknown source. A channel calibration of the monitor annually will assure that the monitor readscorrectly.

c. A channel check of the purification system radiation monitor during each day's operation assures that the monitor is operable.

A channel test weekly will ensure that the systemresponds to a known source. A channel calibration of the monitor annually will assure that themonitor reads correctly.

d. A channel check of the reactor room Continuous Air Monitor (CAM) during each day'soperation will assure that the CAM is operable.

A channel test weekly will assure that the CAMresponds correctly to a known source. A channel calibration annually will assure that the CAMreads correctly.

4.8 Experiments

Applicability -This specification applies to the surveillance requirements for experiments installed in any UCD/MNRC reactor experiment facility. Objective -The objective is to prevent the conduct of experiments or irradiations which maydamage the reactor or release excessive amounts of radioactive materials as a result ofexperimental-failure.Specification

a. A new experiment shall not be installed in any UCD/MNRC reactor experiment facility until awritten safety analysis has been performed and reviewed by the UCD/MNRC

Director, or hisdesignee, to establish compliance with the Limitations on Experiments, (Technical Specifications Section 3.8) and 10 CFR 50.59.b. All experiments performed at the UCD/MNRC shall meet the conditions of an approvedFacility Use Authorization.

Facility Use Authorizations and experiments carried out under theseauthorizations shall be reviewed and approved in accordance with the Utilization of the (UCD)McClellan Nuclear Radiation Center Research Reactor Facility Document (MNRC-0027-DOC). An experiment classified as an approved experiment shall not be placed in any UCD/MNRCexperiment facility until it has been reviewed for compliance with the approved experiment andFacility Use Authorization by the Reactor Manager and the Health Physics Manager, or theirdesignated alternates.

c. The reactivity worth of any experiment installed in the pneumatic transfer tube, or in any otherUCD/MNRC reactor in-core or in-tank irradiation facility shall be estimated or measured, as26 0I!appropriate, before reactor operation with said experiment.

Whenever a measurement is done itshall be done at ambient conditions.

d. Experiments shall be identified and a log or other record maintained while experiments are inany UCD/MNRC reactor experiment facility.

Basis -a & b. Experience at most TRIGA reactor facilities verifies the importance of reactor staff andsafety committee reviews of proposed experiments.

c. Measurement of the reactivity worth of an experiment, or estimation of the reactivity worthbased on previous or similar measurements, shall verify that the experiment is within authorized reactivity limits.d. Maintaining a log of experiments while in UCD/MNRC reactor experiment facilities willfacilitate maintaining surveillance over such experiments.

5.0 Design Features5.1 Site and Facility Description. 5.1.1 Sit__eApplicability -This specification applies to the UCD/MNRC site location and specific facilitydesign features. Objective -The objective is to specify those features related to the Safety Analysis evaluation. Specification -a. The site location is situated approximately 8 miles (13 kin) north-by-northeast of downtownSacramento, California. The site of the UCD/MNRC facility is about 3000 ft. (0.6 mi or 0.9 kin)west of Watt Avenue, and 4500 ft. (0.9 mi or 1.4 kin) south of E Street.b. The restricted area is that area inside the fence surrounding the reactor building. Theunrestricted area is that area outside the fence surrounding the reactor building.

c. The TRIGA reactor is located in Building 258, Room 201 of the UCD/MNRC.

This buildinghas been designed with special safety features.

d. The core is below ground level in a water filled tank and surrounded by a concrete shield.Basis -a. Information on the surrounding population, the hydrology, seismology, and climatography ofthe site has been presented in Chapter 2 of the Safety Analysis Report.b. The restricted area is controlled by the UCD/MNRC Director.

c. The room enclosing the reactor has been designed with systems related to the safe operation of the facility.

d. The below grade core design is to negate the consequences of an aircraft hitting the reactorbuilding.

This accident was analyzed in Chapter 13 of the Safety Analysis Report, and found tobe beyond a credible accident scenario. 27 0 05.1.2 Facility ExhaustApplicability -This specification applies to the facility which houses the reactor.Objective -The objective is to assure that provisions are made to restrict the amount ofradioactivity released into the environment, or during a Loss of Coolant Accident, the system isto assure proper removal of heat from the reactor room.Specification -a. The UCD/MNRC reactor facility shall be equipped with a system designed to filter andexhaust air from the UCD/MNRC facility. The system shall have an exhaust stack height of aminimum of 18.2m (60 feet) above ground level.b. Manually activated shutdown controls for the exhaust system shall be located in the reactorcontrol room.Basis -The UCD/MNRC facility exhaust system is designed such that the reactor room shall bemaintained at a negative pressure with respect to the surrounding areas. The free air volumewithin the UCD/MNRC facility is confined to the facility when there is a shutdown of the exhaustsystem. Controls for startup, filtering, and normal operation of the exhaust system are located inthe reactor control room. Proper handling of airborne radioactive materials (in emergency situations) can be conducted from the reactor control room with a minimum of exposure tooperating personnel. 5.2 Reactor Coolant SystemApplicability -This specification applies to the reactor coolant system.Obiective -The objective is to assure that adequate water is available for cooling and shielding duringnormal reactor operation or during a Loss of Coolant Accident. Specification -a. During normal reactor operation the reactor core shall be cooled by a natural convection flow ofwater.b. The reactor tank water level alarm shall activate if the water level in the reactor tank drops below adepth of 23 ft.c. For operations at 1.5 MW or higher during a Loss of Coolant Accident the reactor core shall be cooledfor a minimum of 3.7 hours at 20 gpm by a source of water from the Emergency Core Cooling System.Basis -a. The SAR Chapter 4, Section 4.6, Table 4-19, shows that fuel temperature limit of 930 °C will not beexceeded under natural convection flow conditions.

b. A reactor tank water low level alarm sounds when the water level drops significantly.

This alarmannunciates in the reactor control room and at a 24 hour monitored location so that appropriate corrective action can be taken to restore water for cooling and shielding.

c. The SAR Chapter 13, Section 13.2, analyzes the requirements for cooling of the reactor fuel andshows that the fuel safety limit is not exceeded under Loss of Coolant Accident conditions during thiswater cooling.5.3 Reactor Core and Fuel28

• 05.3.1 Reactor CoreApplicability

-This specification applies to the configuration of the fuel.Objective -The objective is to assure that provisions are made to restrict the arrangement of fuelelements so as to provide assurance that excessive power densities will not be produced. Specification -For operation at 0.5 MW or greater, the reactor core shall be an arrangement of96 or more fuel elements to include fuel followed control rods. Below 0.5 MW there is nominimum required number of fuel elements. In a mixed 20/20, 30/20 and 8.5/20 fuel loading(SAR Chapter 4, Section 4.5.5.6): Mix J Core and Other Variations (1) No fuel shall be loaded into Hex Rings A or B.(2) A fuel followed control rod located in an 8.5 wt% environment shall contain 8.5 wt% fuel.20E Core and Other Variations (1) No fuel shall be loaded into Hex Rings A or B.(2) Fuel followed control rods may contain either 8.5 wt% or 20 wt% fuel.(3) Variations to the 20E core having 20 wt% fuel in Hex Ring C requires the 20 wt% fuel to beloaded into corner positions ony and graphite dummy elements in the flat positions. Theperformance of fuel temperature measurements shall apply to variations to the as-analyzed 20Ecore configurations. 30B Core and Other Variations (1) No fuel shall be loaded into Hex Rings A or B.(2) The only fuel types allowed are 20/20 and 30/20.(3) 20/20 fuel may be used in any position in Hex Rings C through G.(4) 30/20 fuel may be used in any position in Hex Rings 0 through G but not in Hex Ring C.(5) An analysis of any irradiation facility installed in the central cavity of this core shall be donebefore it is used with this core.Basis -In order to meet the power density requirements discussed in the SAR Chapter 4,Section 4.5.5.6, no less than 96 fuel elements including fuel followed control rods and the aboveloading restrictions will be allowed in an operational 0.5 MW or greater core. Specifications forthe 20E core and for the 30B core allow for variations of the as-analyzed core with the condition that temperature limits are being maintained (SAR Chapter 4, Section 4.5.5.6 and ArgonneNational Laboratory Report AN L/ED 97-54).5.3.2 Reactor FuelApplicability -These specifications apply to the fuel elements used in the reactor core.Obiective -The objective is to assure that the fuel elements are of such design and fabricated insuch a manner as to permit their use with a high degree of reliability with respect to theirphysical and nuclear characteristics. 29 0 0Specification -The individual unirradiated TRIGA fuel elements shall have the following characteristics:

a. Uranium content:

8.5, 20 or 30 wt % uranium enriched nominally to less than 20% U-235.b. Hydrogen to zirconium atom ratio (in the ZrH x): 1.60 to 1.70 (I.65+/- 0.05).c. Cladding: stainless steel, nominal 0.5mm (0.020 inch) thick.Basis -a. The design basis of a TRIGA core loaded with TRIGA fuel demonstrates that limitingoperation to 2.3 megawatts steady state or to a 36 megawatt-sec pulse assures an amplemargin of safety between the maximum temperature generated in the fuel and the safety limit forfuel temperature. The fuel temperatures are not expected to exceed 630 00 during any condition of normal operation.

b. Analysis shows that the stress in a TRIGA fuel element, H/Zr ratios between 1.6 and 1.7, isequal to the clad yield strength when both fuel and cladding temperature are at the safety limit9300C. Since the fuel temperatures are not expected to exceed 630 0C during any condition ofnormal operation, there is a margin between the fuel element clad stress and its ultimatestrength.

c. Safety margins in the fuel element design and fabrication allow for normal mill tolerances ofpurchased materials.

5.3.3 Control Rods and Control Rod DrivesApplicability -This Specification applies to the control rods and control rod drives used in thereactor core.Objective -The objective is to assure the control rods and control rod drives are of such adesign as to permit their use with a high degree of reliability with respect to their physical,

nuclear, and mechanical characteristics.

Specification -a. All control rods shall have scram capability and contain a neutron poison such as stainless steel, borated graphite, B 4C powder, or boron and its compounds in solid form. The shim andregulating rods shall have fuel followers sealed in stainless steel. The transient rod shall have anair filled follower and be sealed in an aluminum tube.b. The control rod drives shall be the standard GA rack and pinion type with an electromagnet and armature attached. Basis -a. The neutron poison requirements for the control rods are satisfied by using stainless steel,neutron absorbing borated graphite, B 40 powder, or boron and its compounds. These materials shall be contained in a suitable clad material such as stainless steel or aluminum to assuremechanical stability during movement and to isolate the neutron poison from the tank waterenvironment. Scram capabilities are provided for rapid insertion of the control rods.b. The standard GA TRIGA control rod drive meets the requirements for driving the control rodsat the proper speeds, and the electromagnet and armature provide the requirements for rapidinsertion capability. These drives have been tested and proven in many TRIGA reactors. 30

• 05.4 Fissionable Material StorageApplicability

-This specification applies to the storage of reactor fuel at a time when it is not in thereactor core.Objective -The objective is to assure that the fuel which is being stored will not become critical and willnot reach an unsafe temperature. Specification -a. All fuel elements not in the reactor core shall be stored (wet or dry) in a geometrical array where thekeff is less than 0.9 for all conditions of moderation.

b. Irradiated fuel elements shall be stored in an array which shall permit sufficient natural convection cooling by water or air such that the fuel element temperature shall not exceed the safety limit.Basis -The limits imposed by Technical Specifications 5.4.a and 5.4.b assure safe storage.6.0 Administrative Controls6.1 Organization.

The Vice Chancellor for Research shall be the licensee for the UCD/MNRC. TheUCD/MNRC facility shall be under the direct control of the UCD/MNRC Director or a licensed seniorreactor operator (SRO) designated by the UCD/MNRC Director to be in direct control. The UCD/MNRCDirector shall be accountable to the Vice Chancellor of the Office of Research for the safe operation andmaintenance of the reactor and its associated equipment. 6.1.1 Structure. The management for operation of the UCD/MNRC facility shall consist of theorganizational structure as shown in Figure 6.1.6.1.2 Responsibilities. The UCD/MNRC Director shall be accountable to the Vice Chancellor ofthe Office of Research for the safe operation and maintenance of the reactor and its associated equipment. The UCD/MNRC

Director, or his designated alternate, shall review and approve allexperiments and experiment procedures prior to their use in the reactor.

Individuals in themanagement organization (e.g., Site Manager, Reactor Manager, Health Physics Manager, etc.)shall be responsible for implementing UCD/MNRC policies and for operation of the facility, andshall be responsible for safeguarding the public and facility personnel from undue radiation exposures and for adhering to the operating license and technical specifications. The SiteManager shall report directly to the UCD/MNRC Director. The Reactor Manager and HealthPhysics Manager report directly to the Site Manager.6.1.3 Staffing_ 6.1.3.1 The minimum staffing when the reactor is not shutdown shall be:a. A reactor operator in the control room;b. A second person in the facility area who can perform prescribed instructions;

c. A senior reactor operator readily available.

The available senior reactoroperator should be within thirty (30) minutes of the facility and reachable bytelephone, and;d. A senior reactor operator shall be present whenever a reactor startup isperformed, fuel is being moved, or experiments are being placed in the reactortank.6.1.3.2 A list of reactor facility personnel by name and telephone number shall beavailable to the reactor operator in the control room. The list shall include:31

a. Management personnel.

b. Health Physics personnel.

c. Reactor Operations personnel.

6.1.4 Selection and Training of Personnel. The selection, training and requalification of operations personnel shall meet or exceed the requirements of the American National Standard for Selection andTraining of Personnel for Research Reactors (ANS 15.4). Qualification and requalification of licensedoperators shall be subject to an approved Nuclear Regulatory Commission (NRC) program.6.2 Review. Audit. Recommendation and ApprovalGeneral Policy. Nuclear facilities shall be designed, constructed,

operated, and maintained insuch a manner that facility personnel, the general public, and both university and non-university property are not exposed to undue risk. These activities shall be conducted in accordance withapplicable regulatory requirements.

The UCO Vice Chancellor of the Office of Research shall institute the above stated policy as thefacility license holder. The Nuclear Safety Committee (NSC) has been chartered to assist inmeeting this responsibility by providing timely, objective, and independent

reviews, audits,recommendations and approvals on matters affecting nuclear safety. The following describes the composition and conduct of the NSC.6.2.1 NSC Composition and Qualifications.

The UCD/MNRC Director shall appoint theChairperson of the NSC. The NSC Chairperson shall appoint a Nuclear Safety Committee (NSC) of at least seven (7) members knowledgeable in fields which relate to nuclear safety. TheNSC shall evaluate and review nuclear safety associated with the operation and use of theUCD/MN RC.6.2.2 NSC Charter and Rules. The NSC shall conduct its review and audit (inspection) functions in accordance with a written charter. This charter shall include provisions for:a. Meeting frequency (The committee shall meet at least semiannually).

b. Voting rules.c. Quorums (For the full committee, a quorum will be at least seven (7) members).

d. A committee review function and an audit/inspection function.

e. Use of subcommittees.

f. Review, approval and dissemination of meeting minutes.6.2.3 Review Function.

The responsibilities of the NSC, or a designated subcommittee thereof,shall include but are not limited to the following:

a. Review approved experiments utilizing UCD/MNRC nuclear facilities.

b. Review and approve all proposed changes to the facility

license, the Technical Specifications and the Safety Analysis Report, and any new or changed Facility Use Authorizations andproposed Class 1 modifications, prior to implementing (Class I) modifications, prior to takingaction under the preceding documents or prior to forwarding any of these documents to theNuclear Regulatory Commission for approval.

c. Review and determine whether a proposed change, test, or experiment would constitute anunreviewed safety question or require a change to the license, to a Facility Use Authorization, or32

• 0to the Technical Specifications.

This determination may be in the form of verifying a decisionalready made by the UCD/MNRC Director.

d. Review reactor operations and operational maintenance, Class I modification

records, andthe health physics program and associated records for all UCD/MNRC nuclear facilities.

e. Review the periodic updates of the Emergency Plan and Physical Security Plan forUCD/MNRC nuclear facilities.

f. Review and update the NSC Charter every two (2) years.g. Review abnormal performance of facility equipment and operating anomalies.

h. Review all reportable occurrences and all written reports of such occurrences prior toforwarding the final written report to the Nuclear Regulatory Commission.

i. Review the NSC annual audit/inspection of the UCD/MNRC nuclear facilities and any otherinspections of these facilities conducted by other agencies.

6.2.4 Audit/Inspection Function. The NSC or a subcommittee

thereof, shall audit/inspect reactoroperations and health physics annually.

The annual audit/inspection shall include, but not belimited to the following:

a. Inspection of the reactor operations and operational maintenance, Class I modification

records, and the health physics program and associated

records, including the ALARA program,for all UCD/MNRC nuclear facilities.

b. Inspection of the physical facilities at the UCD/MNRC.

c. Examination of reportable events at the UCD/MNRC.

d. Determination of the adequacy of UCD/MNRC standard operating procedures.

e. Assessment of the effectiveness of the training and retraining programs at the UCD/MNRC.

f. Determination of the conformance of operations at the UCD/MNRC with the facility's licenseand Technical Specifications, and applicable regulations.

g. Assessment of the results of actions taken to correct deficiencies that have occurred innuclear safety related equipment, structures,

systems, or methods of operations.

h. Inspection of the currently active Facility Use Authorizations and associated experiments.

i.Inspection of future plans for facility modifications or facility utilization.

j. Assessment of operating abnormalities.

k. Determination of the status of previous NSC recommendations.

6.3 Radiation Safety. The Health Physics Manager shall be responsible for implementation of theUCD/MNRC Radiation Safety Program. The program should use the guidelines of the AmericanNational Standard for Radiation Protection at Research Reactor Facilities (ANSI/ANS 15.11). TheHealth Physics Manager shall report to the Site Manager.6.4 Procedures. Written procedures shall be prepared and approved prior to initiating any of theactivities listed in this section. The procedures shall be approved by the UCD/MNRC Director. A periodicreview of procedures shall be performed and documented in a timely manner by the UCD/MNRC staff toassure that procedures are current. Procedures shall be adequate to assure the safe operation of the33 0 ireactor, but shall not preclude the use of independent judgment and action should the situation require.Procedures shall be in effect for the following items:6.4.1 Reactor Operations Procedures

a. Startup, operation, and shutdown of the reactor.b. Fuel loading, unloading, and movement within the reactor.c. Control rod removal or replacement.

d. Routine maintenance of the control rod drives and reactor safety and interlock systems orother routine maintenance that could have an effect on reactor safety.e. Testing and calibration of reactor instrumentation and controls, control rods and control roddrives.f. Administrative controls for operations, maintenance, and conduct of irradiations andexperiments that could affect reactor safety or core reactivity.

g. Implementation of required plans such as emergency and security plans.h. Actions to be taken to correct potential malfunctions of systems, including responses toalarms and abnormal reactivity changes.6.4.2 Health Physics Procedures

a. Testing and calibration of area radiation

monitors, facility air monitors, laboratory radiation detection

systems, and portable radiation monitoring instrumentation.

b. Working in laboratories and other areas where radioactive materials are used.c.- Facility radiation monitoring program including routine and special surveys, personnel monitoring, monitoring and handling of radioactive waste, and sampling and analysis of solidand liquid waste and gaseous effluents released from the facility.

The program shall include amanagement commitment to maintain exposures and releases as low as reasonably achievable (ALARA).d. Monitoring radioactivity in the environment surrounding the facility.

e. Administrative guidelines for the facility radiation protection program to include personnel orientation and training.

f. Receipt of radioactive materials at the facility, and unrestricted release of materials and itemsfrom the facility which may contain induced radioactivity or radioactive contamination.

g. Leak testing of sealed sources containing radioactive materials.

h. Special nuclear material accountability.

i. Transportation of radioactive materials.

Changes to the above procedures shall require approval of the UCD/MNRC Director. All such changes shall bedocumented. 6.5 Experiment Review and Approval. Experiments having similar characteristics are grouped togetherfor review and approval under specific Facility Use Authorizations. All specific experiments to be34

• 0performed under the provisions of an approved Facility Use Authorization shall be approved by theUCD/MNRC

Director, or his designated alternate.

a. Approved experiments shall be carried out in accordance with established and approved procedures.

b. Substantive change to a previously approved experiment shall require the same review and approvalas a new experiment.

c. Minor changes to an experiment that do not significantly alter the experiment may be approved by asenior reactor operator.

6.6 Required Actions6.6.1 Action to be taken in case of a safety limit violation. In the event of a safety limit violation (fuel temperature), the following action shall be taken:a. The reactor shall be shut down and reactor operation shall not be resumed until authorized by the NRC.b. The safety limit violation shall be promptly reported to the UCD/MNRC Director.

c. The safety limit violation shall be reported to the chairman of the NSC and to the NRC by theUCD/MNRC Director.

d. A safety limit violation report shall be prepared.

The report shall describe the following: (1) Applicable circumstances leading to the violation, including when known, the causeand contributing factors.(2) Effect of the violation upon reactor facility components,

systems, or structures, andon the health and safety of personnel and the public.(3) Corrective action to be taken to prevent reoccurrence.

e. The safety limit violation report shall be reviewed by the NSC and then be submitted to theNRC when authorization is sought to resume operation of the reactor.6.6.2 Actions to be taken for reportable occurrences.

In the event of reportable occurrences, the following actions shall be taken:a. Reactor conditions shall be returned to normal or the reactor shall be shut down. If it isnecessary to shut down the reactor to correct the occurrence, operations shall not be resumedunless authorized by the UCD/MNRC Director or his designated alternate.

b. The occurrence shall be reported to the UCD/MNRC Director or the designated alternate.

The UCD/MNRC Director shall report the occurrence to the NRC as required by these Technical Specifications or any applicable regulations.

c. Reportable occurrences should be verbally reported to the Chairman of the NSC and theNRC Operations Center within 24 hours of the occurrence.

A written preliminary report shall besent to the NRC, Attn: Document Control Desk, 1 White Flint North, 11555 Rockville Pike,Rockville MD 20852, within 14 days of the occurrence. A final written report shall be sent to theabove address within 30 days of the occurrence.

d. Reportable occurrences should be reviewed by the NSC prior to forwarding any writtenreport to the Vice Chancellor of the Office of Research or to the Nuclear Regulatory Commission.

35

• 06.7 Reports.

All written reports shall be sent within the prescribed interval to the NRC, Attn: DocumentControl Desk, 1 White Flint North, 11555 Rockville Pike, Rockville MD 20852.6.7.1 Operating Reports. An annual report covering the activities of the reactor facility duringthe previous calendar year shall be submitted within six months following the end of eachcalendar year. Each annual report shall include the following information:

a. A brief summary of operating experiences including experiments performed, changes infacility design, performance characteristics and operating procedures related to reactor safetyoccurring during the reporting period, and results of surveillance tests and inspections.

b. A tabulation showing the energy generated by the reactor (in megawatt hours), hours thereactor was critical, and the cumulative total energy output since initial criticality.

c. The number of emergency shutdowns and inadvertent scrams, including reasons for theshutdowns or scrams.d. Discussion of the major maintenance operations performed during the period, including theeffect, if any, on the safety of the operation of the reactor and the reasons for any corrective maintenance required.

e. A brief description, including a summary of the safety evaluations, of changes in the facility orin procedures, and of tests and experiments carried out pursuant to Section 50.59 of 10 CFRPart 50.f. A summary of the nature and amount of radioactive effluents released or discharged to theenvironment beyond the effective control of the licensee as measured at or prior to the point ofsuch release or discharge, including the following:

(1) Liquid Effluents (summarized on a monthly basis).(a) Liquid radioactivity discharged during the reporting period tabluated asfollows:1 The total estimated quantity of radioactivity released (in curies).2 An estimation of the specific activity for each detectable radionuclide present if the specific activity of the released material after dilution isgreater than 1 xl07 microcuries/ml. 3 A summary of the total release in curies of each radionuclide determined in 2_above for the reporting period based on representative isotopic analysis. 4 An estimated average concentration of the released radioactive material at the point of release for each month in which a releaseoccurs, in terms of microcuries/ml and the fraction of the applicable concentration limit in 10 CFR 20.(b) The total volume (in gallons) of effluent water (including diluent) releasedduring each period of liquid effluent release.(2) Airborne Effluents (summarized on a monthly basis):(a) Airborne radioactivity discharged during the reporting period (in curies)tabulated as follows:36 0 0I The totai estimated quantity of radioactivity released (in curies)determined by an appropriate sampling and counting method.2 The total estimated quantity (in curies) of Argon-41 released duringthe reporting period based on data from an appropriate monitoring system.3 The estimated maximum annual average concentration of Argon-41in the unrestricted area (in microcuries/ml), the estimated corresponding annual radiation dose at this location (in millirem), and the fraction of theapplicable 10 CFR 20 limits for these values.4 The total estimated quantity of radioactivity in particulate form withhalf lives greater than eight days (in curies) released during thereporting period as determined by an appropriate particulate monitoring system.5 The average concentration of radioactive particulates with half-lives greater than eight days released (in microcuries/ml) during the reporting period.(3) Solid Waste (summarized on an annual basis)(a) The total amount of solid waste packaged (in cubic feet).(b) The total activity in solid waste (in curies).(c) The dates of shipment and disposition (if shipped off site).g. An annual summary of the radiation exposure received by facility operations personnel, byfacility users, and by visitors in terms of the average radiation exposure per individual and thegreatest exposure per individual in each group.h. An annual summary of the radiation levels and levels of contamination observed duringroutine surveys performed at the facility in terms of average and highest levels.i.An annual summary of any environmental surveys performed outside the facility. 6.7.2. Special Reports. Special reports are used to report unplanned events as well as plannedadministrative changes. The following classifications shall be used to determine the appropriate reporting schedule:

a. A report within 24 hours by telephone or similar conveyance to the NRC operations center of:(1) Any accidental release of radioactivity into unrestricted areas above applicable unrestricted area concentration limits, whether or not the release resulted in propertydamage, personal injury, or exposure; (2) Any violation of a safety limit;(3) Operation with a limiting safety system setting less conservative than specified inSection 2.0, Limiting Safety System Settings; (4) Operation in violation of a Limiting Condition for Operation; 37 0 0(5) Failure of a required reactor or experiment safety system component which couldrender the system incapable of performing its intended safety function unless the failureis discovered during maintenance tests or a period of reactor shutdown; (6) Any unanticipated or uncontrolled change in reactivity greater than $1.00;(7) An observed inadequacy in the implementation of either administrative or procedural

controls, such that the inadequacy could have caused the existence or development of acondition which could have resulted in operation of the reactor outside the specified safety limits; and(8) A measurable release of fission products from a fuel element.b. A report within 14 days in writing to the NRC, Document Control Desk, Washington DC.(1) Those events reported as required by Technical Specifications 6.7.2.a.1 through6.7.2.a.8.

(2) The written report (and, to the extent possible, the preliminary telephone report orreport by similar conveyance) shall describe,

analyze, and evaluate safety implications, and outline the corrective measures taken or planned to prevent reoccurrence of theevent.c. A report within 30 days in writing to the NRC, Document Control Desk, Washington DC.(1) Any significant variation of measured values from a corresponding predicted orpreviously measured value of safety-connected operating characteristics occurring during operation of the reactor;(2) Any significant change in the transient or accident analysis as described in theSafety Analysis Report (SAR);(3) A personnel change involving the positions of UCD/MNRC Director or UCD ViceChancellor for Research; and(4) Any observed inadequacies in the implementation of administrative or procedural controls such that the inadequacy causes or could have caused the existence ordevelopment of an unsafe condition with regard to reactor operations.

6.8 Records. Records may be in the form of logs, data sheets, or other suitable forms. The requiredinformation may be contained in single or multiple

records, or a combination thereof.

Records and logsshall be prepared for the following items and retained for a period of at least five years for items a.through f., and indefinitely for items g. through k. (Note: Annual reports, to the extent they contain all ofthe required information, may be used as records for items g. through j.)a. Normal reactor operation.

b. Principal maintenance activities.

c. Those events reported as required by Technical Specifications 6.7.1 and 6.7.2.d. Equipment and component surveillance activities required by the Technical Specifications.

e. Experiments performed with the reactor.f. Airborne and liquid radioactive effluents released to the environments and solid radioactive wasteshipped off site.38 0 0g. Offsite environmental monitoring surveys.h. Fuel inventories and transfers.

i. Facility radiation and contamination surveys.j. Radiation exposures for all personnel.

k. Updated, corrected, and as-built drawings of the facility.

39 ... NUCLEARcoMSSoREGUALTORY UNIVERSITY OFCALIFORNIA -DAVISVICE CHANCELLOR FORRESEARCH(Licensee) UCD/MNRC UCDIMNRC [DIRECTOR NUCLEAR ._._l COMMITJTEE [ 8At-SITE , 'MANAGERiHEALTH PHYFIGUREAC6.1

• Att EGu _; UNITED STATES ".

REGULATORY COMMISSION20555-0001 March 30, 2004* Dr. Barry M. KleinVice Chancellor for ResearchUniversity of California, DavisOne Shields AvenueDavis, CA 95616-8558

SUBJECT:

REVISION TO SAFETY EVALUATION OF AMENDMENT NO. 7 TO AMENDEDFACILITY OPERATING LICENSE NO. R-130 -REGENTS OF THEUNIVERSITY OF CALIFORNIA (TAC NO. MB5598)

Dear Dr. Klein:

The U.S. Nuclear Regulatory Commission has Issued the enclosed revision to the SafetyEvaluation of Amendment No. 7 to Facility Operating License No. R-1 30 for the McClellan. Nuclear Radiation Center (MNRC) TRIGA Research Reactor. Amendment No. 7 was issuedon December 30, 2003 and is available on the Commission's ADAMS system, Accession Number ML033421339. Sincerely, _ WreJ .EeIn rjc aaeResearch and Test Reactors SectionNew, Research and Test Reactors ProgramDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation Docket No. 50-607

Enclosure:

Revision to Amendment No. 7Safety Evaluation Reportcc w/enclosure: Please see next page 0..University of California -Davis/McClellan MNRC Docket No. 50-607cc:Mr. Jeff Ching5335 Price Avei~ue, Bldg. 258McClellan AFB, CA 95652-2504 Test, Research, and TrainingReactor Newsletter University of Florida202 Nuclear Sciences CenterGainesville, FL 32611 0UNITEb STATESNUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 REVISION TO SAFETY EVALUATION REPORTSUPPORTING AMENDMENT NO. 7 TOAMENDED .FACILITY OPERATING LICENSE NO. R-130REGENTS OF THE UNIVERSITY OF CALIFORNIA ATMcCLELLAN NUCLEAR RADIATION CENTERDOCKET N.O. 50-60

71.0 INTRODUCTION

By letter dated October 21, 2003, the Regents of the University of California (the licensee) submitted a request for amendment of the Facility Operating

• License No. R-130 for theMcClellan Nuclear Radiation Center (MNRC) TRIGA research reactor.

The request provided forthe allowance of radioactive materials not produced by the reactor to be received, possessed and used on the facility site. In particular, it was requested that Section 2.B of the FacilityOperating License be amended to include an additional section 2.B.(4) as follows:2.B.(4) In addition to those items specified in 2.B.(1), 2.B.(2) and 2.B.(3) the following radioactive materials maybe received, possessed, and used at the facility. Radioactive Material(element and mass number)A. Any radioactive materialbetween atomic number I through83, Inclusive B. Any radioactive material withatomic numbers 84 and abovec.. Iodine-125 D. Source material (but only traceamounts of Th-234)E. Special nuclear materialChemical and/orPhysical Form.A. AnyA. Anyc. Iodide/LIquid D. AnyE. AnyMaximum Quantity Licensee MayPossess at Any One TimeA. 20 curies (1 curie each, except asprovided below)A. 4 Curies (100 milllcuries each,except as provided below) or up to 20micrograms C. 40 CuriesD. 4 grams per radionuclide, not toexceed 10 grams totalE. 2 grams per radionuclide, not toexceed 5 grams totalThis amendment request was approved and issued on .December 30, 2003. 0 "-2-2.0 EVALUATION The previous safety evaluation assumed that all of the radioactive materials to be received, possessed and handled in accordance with this amendment request would be located in thereactor room glove box. The significance of this assumption is related to the ability of thereactor room glove box and its associated exhaust system to mitigate the consequences associated with the complete volatilization of the maximum radioactive material inventory contained in the box, a total of 64.4 curies. (The total activity in categories A, B, and C in theabove table is 64 curies. The maximum activity In category D is about 0.1 curie, while themaximum activity in category E is about 0.3 curie.). The staff concluded that the consequences of the complete volatilization of 64.4 curies are much less than the consequences of thebounding MHA, and the amendment request was approved. Instead of locating all of the radioactive materials shown in above table in the reactor roomglove box, some of the materials will be located in the restricted area of the McClellan NuclearRadiation Center. Non-volatile material will be handled in accordance with approvedprocedures. Any unsealed volatile

material, such as Iodine-I125 (the majority of the radioactive materials),

will continue to be handled in areas with filtered ventilation to mitigate theconsequences of complete volatilization of the unsealed material (e.g., the reactor room glovebox and reactor room fume hood), as previously analyzed. The staff has reviewed the proposed change to the Facility Operating License and concluded that it does not impact the licensee's ability to continue to meet the relevant requirements of 10CFR Part 50.38.3.0 ENVIRONMENTAL CONSIDERATION .This amendment does hot Involve changes in the installation or use of a facility component located within the restricted area as defined ion 10 CFR Part 20 or changes in inspection andsurveillance requirements. The staff has determined that this amendment involves nosignificant increase in the amounts, and no significant change in the types, of any effluents thatmay be released off site, and no significant increase in individual or cumulative occupational radiation exposure. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared with the issuance of thisamendment.

4.0 CONCLUSION

The staff has concluded, on the basis of the considerations discussed above, that (1) becausethe amendment does not involve a significant increase in the probability or consequences ofaccidents previously evaluated, or create the possibility of a new or different kind of accidentfrom any accident previously evaluated, and does not involve a significant reduction In a margin* of safety, the amendment does not involve a significant hazards consideration; (2) there isreasonable assurance that the health and safety of the public will not be endangered by the proposed changes; and (3) such changes are in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security orthe health and safety of the public.Principal Contributor: Warren J. EreslanDate: March 30, 2004 '5%. i ./ uNITED STATESNLCLEAR REGULATORY COMMISSION 0 ASIGTNDC.205-00 Deceeiber: 30, 2003Dr. Barry M. KleinVice Chancellor for ResearchUniversity of California, DavisOne Shields AvenueDavis, CA 95616-8558

SUBJECT:

ISSUANCE OF AMENDMENT NO. 7 TO AMENDED FACILITY OPERATING LICENSE NO. R-130 -REGENTS OF THE UNIVERSITY OF CALIFORNIA (TAC NO. MB5598)

Dear Dr. Klein:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 7 toFacility Operating License No. R-1 30 for the McClellan Nuclear Radiation Center (MNRC)TRIGA Research Reactor. The amendment consists of changes to the Facility Operating License in response to your submittals of October 21, 2003 and November 6, 2003, and isdiscussed in the enclosed Safety Evaluation Report. 69~4~tey/Warren J. Eresian, Project ManagerResearch and Test Reactors SectionNew, Research and Test Reactors ProgramDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation Docket No. 50-607

Enclosures:

1. Amendment No. 72. Safety Evaluation Report University of California

-Davis/McClellan MNRC Docket No. 50-607cc:Dr. Wade J. Richards5335 Price Avenue, Bldg. 258McClellan AFB, CA 95652-2504 Test, Research, and TrainingReactor Newsletter University of Florida202 Nuclear Sciences CenterGainesville, FL 32611 UNITED STATESNUCLEAR REGULATORY COMMISSIONWASHINGTON, D.C. 20555-0001 REGENTS OF THE UNIVERSITY OF CALIFORNIA ATMcCLELLAN NUCLEAR RADIATION CENTERDOCKET NO. 50-607AMENDMENT TO AMENDED FACILITY OPERATING LICENSEAmendment No. 7License No. R-1 301. The U.S. Nuclear Regulatory Commission (the Commission) has found thatA. The application for an amendment to Amended Facility Operating LicenseNo. R-1 30 filed by the Regents of the University of California at McClellan NuclearRadiation Center (the licensee) on October 21, 2003 and November 6, 2003,conforms to the standards and requirements of the Atomic Energy Act of 1954:, asamended (the Act), and the regulations of the Commission as stated in Chapter I ofTitle 10 of the Code of Federal Regulations (10 CFR);B. The facility will operate In conformity with the application, the provisions of the Act,and the rules and regulations of the Commission; C. There is reasonable assurance that (i) the activities authorized by this amendment can be conducted without endangering the health and safety of the public and (ii)such activities will be conducted in compliance with the regulations of theCommission; D. The issuance of this amendment will not be Inimical to the common defense andsecurity or to the health and safety of the public;E. This amendment is issued in accordance with the regulations of the Commission asstated in 10 CFR Part 51, and all applicable requirements have been satisfied; andF. Prior notice of this amendment was not required by 10 CFR 2.105, and publication of notice for this amendment is not required by 10 CFR 2.106. O f O0..-2-2. Accordingly, the license is amended by changes to the Facility Operating License asindicated below, and paragraph 2.B of Amended Facility Operating License No. R-130 ishereby amended to read as follows:2.B.(4) In addition to those items specified in 2.B.(1), 2.B.(2) and 2.B.(3) the following radioactive materials may be received, possessed, and used at the facility. Radioactive Material(element and massnumber)A. Any radioactive material betweenatomic number 1through 83, inclusive B. Any radioactive material with atomicnumbers 84 andaboveChemical and/orPhysical FormA. AnyA. AnyMaximum Quantity LicenseeMay Possess at Any One TimeA. A. 20 Curies (I Curie each,except as provided below)A. 4 Curies (100 millicuries each, except as providedbelow) or up to 20micrograms C. 40OCuries D. 4 grams per radionuclide, not to exceed 10 gramstotalE. 2 grams per radionuclide, not to exceed 5 grams totalC. Iodine-125 D. Source material (butonly trace amountsof Th-234)E. Special nuclearmaterialC. Iodide/Liquid

0. AnyE. Any3. This license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Research and Test Reactors SectionNew, Research and Test Reactors ProgramDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation Date of Issuance: December 30, 2003 O 0~UNITED STATESNUCLEAR REGULATORY COMMISSION o~WASHINGTON, D.C. 20555-0001 o#SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 7 TOAMENDED FACILITY OPERATING LICENSE NO. R-130REGENTS OF THE UNIVERSITY OF CALIFORNIA ATMcCLELLAN NUCLEAR RADIATION CENTERDOCKET NO. 50-60

71.0 INTRODUCTION

By letter dated October 21, 2003, the Regents of the University of California (the licensee) submitted a request for amendment of the Facility Operating License No. R-1 30 for the McClellan Nuclear Radiation Center (MNRC) TRIGA research reactor. The request provides for theallowance of radioactive materials not produced by the reactor to be received, possessed andused on the facility site. In particular, it is requested that Section 2.B of the Facility Operating License be amended to include an additional Section 2.B.(4) as follows:2.B.(4) In addition to those items specified in 2.B.(1), 2.B.(2) and 2.B.(3) the following radioactive materials may be received, possessed, and used at the facility. Radioactive Material(element and massnumber)A. Any radioactive material betweenatomic number 1through 83, inclusive B. Any radioactive material with atomicnumbers 84 andaboveChemical and/orPhysical FormA. AnyA. AnyMaximum Quantity Licensee MayPossess at Any One TimeA. A. 20 Curies (1 Curie each,except as provided below)A. 4 Curies (100 mlllicuries each, except as providedbelow) or up to 20micrograms C. 40 CuriesD. 4 grams per radionuclide, not to exceed 10 grams totalE. 2 grams per radionuclide, not to exceed 5 grams totalC. Iodine-I125 D. Source material (butonly trace amounts ofTh-234)E. Special nuclearmaterialC. Iodide/Liquid D. AnyE. Any

• 0-2,-This request is discussed below.2.0 EVALUATION All of the radioactive materials to be received, possessed and handled In accordance with thisamendment request will be located in the reactor room glove box. In November of 2002, the NRCapproved Amendment No. 5 of the Technical Specifications for the McClellan Nuclear Radiation Center. The safety concern addressed in that amendment was related to the ability of the reactorroom glove box and Its associated exhaust system to mitigate the consequences associated withthe complete volatilization of the maximum radioactive material inventory contained in the box, atotal of 61 curies of Iodine-125.

The analysis showed that the CEDE to the thyroid for a 10-minute exposure in the unrestricted area would be about 3 millirem. For those exposed in the reactorroom for the maximum assumed occupancy time of 5 minutes the CEDE to the thyroid would beabout 205 millirem. These doses were compared to the expected doses (CEDE) resulting fromthe Maximum Hypothetical Accident (MHA), which serves as the bounding accident forradiological consequences. The resulting doses from the MHA are 53 millirem in the unrestricted area and 360 millirem in the reactor room. The staff concluded that the consequences of thecomplete volatilization of 61 curies of Iodine-I125 were less than the bounding MHA and therefore there was not a significant reduction of the margin of safety with respect to the MHA.This amendment request will increase the total allowable activity in the reactor room glove boxfrom 61 curies to 64.4 curies. (The total activity in categories A, B, and C In the above table Is 64curles. The maximum activity in category D corresponds to 10 grams of Uranium-233, or about0.1 curie. The maximum activity in category E corresponds to 5 grams of Plutonium-239, or about0.3 curie.) For the complete volatilization of 64.4 curies, doses in the unrestricted area and in thereactor room will scale up proportionally from 61 curies, resulting in a dose in the unrestricted areaof 3.2 mlllirem, and a dose in the reactor room of 216 millirem (i.e., the doses have increased by5.6 percent.) The staff concludes that the consequences of the complete volatilization of 64.4 curies are muchless than the consequences of the bounding MHA, and that increasing the allowable activity in thereactor room glove box from 61 curies to 64.4 curies does not significantly reduce the margin ofsafety with respect to the MHA and to 10 CFR Part 20 limits and that the increase is acceptable. The staff has reviewed the proposed change to the Facility Operating License and concluded thatit does not impact the licensee's ability to continue to meet the relevant requirements of 10 CFRPart 50.36.3.0 ENVIRONMENTAL CONSIDERATION This amendment does not involve changes in the installation or use of a facility component locatedwithin the restricted area as defined in 10 CFR Part 20 or changes in inspection and surveillance requirements. The staff has determined that this amendment involves no significant increase Inthe amounts, and no significant change in the types, of any effluents that may be released off site,and no significant increase in individual or cumulative occupational radiation exposure. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in10 CER 51 .22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement orenvironmental assessment need be prepared with the issuance of this amendment.

4.0 CONCLUSION

The staff has concluded, on the basis of the considerations discussed above, that (1) because theamendment does not involve a significant increase in the probability or consequences of accidents previously evaluated, or create the possibility of a new or different kind of accident from anyaccident previously evaluated, and does not involve a significant reduction in a margin of safety,the amendment does not involve a significant hazards consideration; (2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposedchanges; and (3) such changes are in compliance with the Commission's regulations and theissuance of this amendment will not be inimical to the common defense and security or the healthand safety of the public.Principal Contributor: Warred J. EresianDate: December 30, 2003 UNITED STATES*NUCLEAR REGULATORY COMMISSION ,,.. WASHINGTON, D.C. 20555-0001February 17, 2000*i7/tlJ*Brigadier General Michael P. Wiedemer Vice Chancellor Kevin SmithCommander Office of the Chancellor ..~Sacramento Air Logistics Center University of California, DavisSM-ALCITI-1 One Shields Avenue5335 Price Avenue Davis, California 95616-8558 McClellan AFB, California 95652-2504

SUBJECT:

RE-ISSUANCE OF NOTICE OF CONSIDERATION OF APPROVAL OFTRANSFER OF FACILITY OPERATING LICENSE NO. R-130 FOR THEMcCLELLAN NUCLEAR RADIATION CENTER FROM THE DEPARTMENT OFTHE AIR FORCE TO THE REGENTS OF THE UNIVERSITY OF CALIFORNIA

• AND CONFORMING AMENDMENT, AND OPPORTUNITY FOR A HEARING(TAC NO. MA3477)Dear General Wiedemer and Dr. Smith:The enclosed document has been re-issued in its entirety to correct someadministrative errors. We. apologize for any inconvenience this may have caused.Sincerely, Ledyard B. Marsh, ChiefEvents Assessments, Generic Communications and Non-Power Reactors BranchDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation Docket No. 50-607

Enclosure:

As statedcc: wlenclosures McClellan AFB TRIGA REACTORDcktN.0-7 CC:Dr. Wade J. RichardsSM-ALC/TI-1 5335 Price Avenue, Bldg. 258McClellan AFB, CA 95652-2504 Cot. Robert CapellHQ AFMC/SGC4225 Logistics Avenue, Suite 23Wright-Patterson AFB, OH 45433-5762 Lt. Col. Catherine Ze~ringue HQ AFSCISEW9570 Avenue G, Building 24499Kirtland AFB, New Mexico 871 17-5670Test, Research, and TrainingReactor Newsletter 202 Nuclear Sciences CenterUniversity of FloridaGainesville, FL 3261 1

• L0TECHNICAL SPECIFICATIONS FOR THEUNIVERSITY OF CALIFORNIA

-DAVIS MCCLELLAN NUCLEAR RADIATION CENTER(UCDIMNRC) DOCUMENT NUMBER: MNRC-0004-DOC-13 Rev 13 4/03p.~. ~.1 *>~0 !Revision ,13 of the "Technical Specifications" for the University of California at Davis/McClellan Nuclear Radiation Center (UCD/MNRC) Reactor have undergone the following coordination: Reviewed by: !.0eltPyiSpesoDate Reviewed by: 0.R ator Su~d pervisrDt Approved by:~i'I~toY* ~Ij.e/o3~l1.DateChairman, NRCNuclear Safety Committee I.(K ~/--I 0Technical Specifications Rev 13 412003Title PageApproval Page31Rev 13Rev 13Rev 13Rev 13Rev 13Rev 134/20034/20034/20034/20034/20034/20033233Figure 6.1......................----.--... ~

• 0* " TECHNICAL SPECIFICATIONS TABLE OF CONTENTS1.0 Definitions

.............................................................................................................. 12.0 Safety Limit and Limiting Safety System Setting 2.1 Safety Limits.................................................................................................. 2.2 Limiting Safety System Setting (LSSS)..................................................................... 62.2.1 Fuel Temperature ............................................................ i.................... 63.0 Limiting Conditions for Operations (LC.O.) ........................................................................... 73.1 Reactor Core Parameters................................................................................... 73.1.1 Steady-State Operation .......................................

...................................

73.1.2 Pulse or Square Wave Operation ............................................................... 73.1.3 Reactivity Limitation~s............................................................................. 83.2 Reactor Control and Safety Systems .... .................................................................. 83.2.1 Control Rods...................................................................................... 83.2.2 Reactor Instrumentation.......................................................................... 93.2.3 Reactor Scrams and Interlocks................................................................. 103.2.4 Reactor Fuel Elementts.......................................................................... 123.3 Reactor Coolant Systems.................................................................................. 133.4 Reactor Room Exhaust System ........................................................................... 143.5 Intentionally Left Blank ..................................................................................... 143.6 Intentionally Left Blank..................................................................................... 143.7 Reactor Radiation Monitoring Systems.................................................................... 143.7.1 Monitoring Systems ... ......................................................................... 143.7.2 Effluent~s -Argon-41 Discharge Limit. ..........................................................

16) 0 03.8 Experiments

................................................................................................ 163.8.1 Reactivity Limits ........................................ 163.8.2 Materials Limit................................................................................... 173.8.3 Failure and Malfunctions ...................... ................................................. 184.0 Surveillance Requirements.......................................................................................... 194.1 Reactor Core Parameters ................................................................................. 194.1.1 Steady State Operation......................................................................... 194.1.2 Shutdown Margin and Excess Reactivity ....................................................... 204.2 Reactor Control and Safety Systems...................................................................... 204.2.1 Control Rods ................................................... ................................. 204.2.2 Reactor Instrumentation............................................................... 214.2.3 Reactor Scrams and interlocks.................... ............................................. 224.2.4 Reactor Fuel Elements................................ .......................................... 234.3 Reactor Coolant Systems ................................................................................. 244.4 Reactor Room Exhaust System ........................................................................... 254.5 Intentionally Left Blank..................................................................................... 254.6 Intentionally Left Blank..................................................................................... 254.7 Reactor Radiation Monitoring Systems.................................................................... 254.8 Experiments ................................................................................................ 265.0 Design Features ...........................................................

.........................................

275.1 Site and Facility Description ............................................................................... 275.1.1 .Site............................................................................................... 275.1.2 Facility Exhaust ......................... ....................................................... 285.2 Reactor Coolant system ................................................................................... 28 5.3 Reactor Core and F.uel .................................................................................... 295.3.1 Reactor Care .................................................................................... 295.3.2 Reactor F..u~l..................................................................................... 305.3.3 Control Rods and Control Rod Drives.......................................................... 315.4 Fissionable Material Storage............................................................................... 316.0 Administrative Controls ............................................................................................ .316.1 Organization................................................................................................ 316.1.1 Structure......................................................................................... 326.1.2 Responsibilities ................................................................................. 326.1.3 Staffing .......................................................................................... 326.1.4 Selection and Training of Personnel ........................................................... 326.2 Review, Audit, Recommendation and Approvial........................................................... 326.2.1 NSC Composition and Qualifications........................................................... 336.2.2 NSC Charter and Rules .......................... i.............................................. 33 I6.2.3 Review Function................................................................................. 336.2.4 Audit/Inspection Function ....................................................................... 346.3 Radiation Safety............................................................................................ 34 16.4 Procedures ................................................................................................. 346.4.1 Reactor Operations Procedur~es................................................................. 34 I6.4.2 Health Physics Procedures ..................................................................... 356.5 Experiment Review and Approlval ......................................................................... 356.6 Required Actions ........................................................................................... 356.6.1 Actions to be taken in case of a safety limit violation: ......................................... 35 *6.6.2 Actions to be taken for reportable occurrences ................ .............................. 36 6.7.1 Operating Reports................................................................................. 366.7.2 Special Reports ................................................................................... 386.8 Records........................................................................................................ 39Fig. 6.1 UCD/MNRC Organization for Licensing and Operation........................................................ 40* /

• 0TECHNICAL SPECIFICATIONS FOR THEUNIVERSITY OF CALIFORNIA

-DAVIS/MCCLELLAN NUCLEAR RADIATION CENTER (UCDIMNRC) The University of California -Davis/McClellan Nuclear Radiation Center (UCD/MNRC) reactor is operated by theUniversity of California, Davis, California (UCD). The UCD/MNRC research reactor is a TRIGA-type reactor.The UCD/MNRC provides state-of-the-art neutron radiography capabilities. In addition, the UCD/MNRC providesa wide range of irradiation services for both research and industrial needs. The reactor operates at a nominalsteady state power level up to and including 2 MW. The UCD/MNRC reactor is also capable of square waveand pulse operational modes. The UCD/MNRC reactor fuel is less than 20% enriched in uranium-235.

1.0 Definitions

1.1 As Low As Reasonably Achievable (ALARA)~. As defined in 10 CFR, Part 20.1.2 Licensed Operators. A UCD/MNRC licensed operator is an individual licensed by the NuclearRegulatory Commission (e.g., senior reactor operator or reactor operator) to carry out the duties andresponsibilities associated with the position requiring the license.1.2.1 Sernior Reactor Operator. An individual who is licensed to direct the activities of reactoroperators and to manipulate the controls of the facility. 1.2.2 Rea~ctor Oper~ator. An individual who is licensed to manipulate the controls of the facilityand perform reactor-related maintenance. 1.3 Channel. A channel is the combination of sensor, line amplifier, processor, and output deviceswhich are connected for the purpose of measuring the value of a parameter. 1.3.1 Channel Test. A channel test is the introduction of a signal into the channel for verification that it is operable. 1.3.2 C..hannel Calibration. A channel calibration adjustment of the channel such that itsoutput corresponds with acceptable accuracy to known-'-values of the parameter which thechannel measures. Calibration shall encompass the entire channel,.including equipment actuation, alarm or trip, and shall be deemed to include a channel test. "1.3.3 Channel. Check. A channel check is a qualitative verification of acceptable performance byobservation of channel behavior. This verification, where possible, shall include comparison ofthe channel with other independent channels or systems measuring the same variable. 1.4 Confinement. Confinement means isolation of the reactor room air volume such that the movementof air into and out of the reactor room is through a controlled path.1.5 Experiment. Any operation,

hardware, or target (excluding devices such as detectors, fissionchambers, foils, etc), which is designed to investigate specific reactor characteristics or which isintended for irradiation within an experiment facility and which is not rigidly secured to a core or shieldstructure so as to be a part of their design.1.5.1 E~xperinrlent.

Moveable. A moveable experiment is one where it is intended that the entireexperiment may be moved in or near the reactor core or into and out of reactor experiment facilities while the reactor is operating. I 1.5.2 Exoerdment. Secured. A secured experdment is any experiment, experiment

facility, orcomponent of an experiment that is held in a stationary position relative to the reactor by.... .,mechanical means. The restraining force must be substantially greater than those to which the-.expediment might be subjected by hydraulic, pneumatic,

buoyant, or other forces which arenormal to the operating environment of the experiment, or by forces which can arise as a resultof credible conditions.

1.5.3 Exoeriment Facilities. Experiment facilities shall mean the pneumatic transfer tube,beamtubes, irradiation facilities, in the reactor core or in the reactor tank, and radiography bays.1.5.4 Experiment Safety System. Experiment safety systems are those systems, including theirassociated input circuits, which are designed to initiate a scram for the primary purpose ofprotecting an experiment or to provide information which requires manual protective action to beinitiated. '1.6 .Fuel Element. Standard. A fuel element is a single TRIGA element. The fuel is U-ZrH clad instainless steel. The zirconium to hydrogen ratio is nominally 1.65 +/- 0.05. The weight percent (wt%) ofuranium can be either 8.5, 20, or 30 wt%, with an enrichment of less than 20% U-235. A standard fuelelement may contain a burnable poison.1.7 Fuel Element., Instrumented. An instrumented fuel element is a standard fuel element fabricated withthermocouples for temperature measurements. An instrumented fuel element shell have at least oneoperable thermocouple embedded in the fuel near the axial and radial mnidpoints. 1.8 Measured Valu~e. The measured value is the value of a parameter as it appears on the output of achannel.1.9 Mode. Steady-State. Steady-state mode operation shall mean operation of the UCDIMNRC reactorwith the selector switch in the automatic or manual mode position. " 1.10 Mode. SQuare-Wave. Square-wave mode operation shall mean operation of the UCD/MNRCreactor with the selector switch in the square-wave mode position. 1.11 Mode. Pulse. Pulse mode operation shall mean operation of the UCD/MNRC reactor with theselector switch in the pulse mode position. 1.12 Operable. Operable means a component or system is capable of performing its intended function.. 1.13 Operat~ina. Operating means a component or system is performing its intended function. 1.14.0peratina Cycle. The period of time starting with reactor startup and ending with reactor shutdown. 1.15 Protective Action. Protective action is the initiation of a signal or the operation of equipment withinthe UCDIMNRC reactor safety system in response to a variable or condition of the UCDIMNRC reactorfacility having reached a specified limit.1.15.1 Channel Level. At the protective instrument channel level, protective action is thegeneration and transmission of a scram signal indicating that a reactor variable has reached thespecified limit.1.15.2 Subsystem Level. At the protective instrument subsystem level, protective action is thegeneration and transmission of a scram signal indicating that a specified limit has been reached.NOTE: Protective action at this level would lead to the operation of the safety shutdowni: equipment. 2 1.15.3 Instrument System Level. At the protective instrument level, protective action is thegeneration and transmission of the command signal for the safety shutdown equipment tooperate.1.15.4 Safety System Level. At the reactor safety system level, protective action is the operation of sufficient equipment to immediately shut down the reactor.1.16 Pulse Operation~al Core. A pulse operational core is a reactor operational core for which themaximum allowable pulse reactivity insertion has been determined. 1; 17 Reactivity. Exce~ss, Excess reactivity is that amount of reactivity that would exist if all control rods(control, regulating, etc.) were moved to the maximum reactive position from the point where the reactoris at ambient temperature and the reactor is critical. (K o, = 1)1.18 Reactivity Limit~s. The reactivity limits are those limits imposed on the reactivity conditions of thereactor core.1.19 R~eactivity Worth of an Exoeriment. The reactivity worth of an experiment is the maximum value ofthe reactivity change that could occur as a result of changes that alter experiment position orconfiguration. 1.20 Reactor Controls. Reactor controls are apparatus and/or mechanisms the manipulation of whichdirectly affect the reactivity or power level of the reactor.1.21 R.eac~tor Core. Operational. The UCD/MNRC reactor operational core is a core for which theparameters of excess reactivity, shutdown margin, fuel temperature, power calibration and reactivity worths of control rods and experiments have been determined to satisfy the requirements set forth inthese Technical Specifications. 1.22 Reactor Ooeratingq. The UCO/MNRC reactor is operating whenever it is not shutdown or secured.1.23 R~eactor Safety Systems. Reactor safety systems are those systems, including their associated input channels, which are designed to initiate automatic reactor protection or to provide information forinitiation of manual protective action.1.24 R~eactor Secured. The UCO/MNRC reactor is secured when the console key switch is in the offposition and the key is removed from the lock and under the control of a licensed

operator, and theconditions of a or b exist:a. (1) The minimum number of control rods are fully inserted to ensure the reactor is shutdown, asrequired by technical specifications; and(2) No work is in progress involving core fuel, core structure, installed control rods, or control roddrives, unless the control rod drives are physically decoupled from the control rods; and(3) No experiments in any reactor experiment

facility, or in any other way .near the reactor, are beingmoved or serviced if the experiments have, on movement, a reactivity worth exceeding the maximumvalue allowed for a single experiment or $1.00, whichever is smaller, orb. The reactor contains insufficient fissile materials in the reactor core, a~djacent experiments or controlrods to attain criticality under optimum available conditions of moderation and reflection.

1.25 Reactor Shut~down. The UCD/MNRC reactor is shutdown if it is subcritical by at least one dollar($1.00) both in the Reference Core Condition and for all allowed ambient conditions with the reactivity worth of all installed experiments included. 3

• 01.26 Reference Cpre Condition.

The condition of the core when it is at ambient temperature (cold T<280C), the reactivity worth of xenon is negligible (< $0.30) (i.e., cold and clean), and the central irradiation facility contains the graphite thimble plug and the aluminum thimble plug (CIF-1).1.27 ReserhRatr A research reactor is defined as a device designed to support a self-sustaining neutron chain reaction for research development, education, and training, or experimental

purposes, andwhich may have provisions for the production of radioisotopes.

1.28 Rod. Control, A control rod is a device fabricated from neutron absorbing

material, with or without afuel or air follower, which is used to establish neutron flux changes and to compensate for routinereactivity losses. The follower may be a stainless steel section.

A control rod shall be coupled to itsdrive unit to allow it to perform its control function, and its safety function when the coupling isdisengaged. This safety function is commonly termed a scram.1.28.1 Regulat~ing Rod. A regulating rod is a control rod used to maintain an intended powerlevel and may be varied manually or by a servo-controller. A regulating rod shall have scramcapability. 1.28.2 Standard Rod. The regulating and shim rods are standard control rods.1.28.3 Transient Rod. The transient rod is a control rod that is capable of providing rapidreactivity insertion to produce a pulse or square wave.1.29 Safety Channel. A safety channel is a measuring channel in the reactor safety system.1.30 Safety Limit. Safety limits are limits on important process variables, which are found to benecessary to reasonably protect the integrity of the principal barriers which guard against theuncontrolled release of radioactivity. 1.31 Scram Time. Scram time is the elapsed time between reaching a limiting safety system set pointand the control rods being fully inserted. 1.32 Scram. External. External scrams may arise from the radiography bay doors, radiography bayripcords, bay shutter interlocks, and any scrams from an experiment. 1.33 Shall. Should and May. The word "shall" is used to denote a requirement; the word "should" todenote a recommendation; the word "may" to denote, permission, neither a requirement nor arecommendation. 1.34 Shut~down Margin. Shutdown margin shall mean the minimum shutdown reactivity necessary toprovide confidence that the reactor can be made subcritical by means of the control and safety systemstarting from any permissible operating condition with the most reactive rod assumed to be in the mostreactive

position, and once this action has been initiated, the reactor will remain subcritical withoutfurther operator action.1.35 Shutdown.

Unsched.u.led. An unscheduled shutdown is any unplanned shutdown of theUCD/M NRC reactor caused by actuation of the reactor safety system, operator error, equipment malfunction, or a manual shutdown in response to conditions which could adversely affect safeoperation, not including shutdowns which occur during testing or check-out operations. 1.36 Surveillance Acfivities. In general, two types of surveillance activitiesare specified: operability checks and tests, and calibrations. Operability checks and tests are generally specified as daily, weeklyor quarterly. Calibration times are generally specified as quarterly, semi-annually,

annually, or biennially.

..... 1.37 Surveillance Intervals. Maximum intervals are established to provide operational flexibility and not/) to reduce frequency. Established frequencies shall be maintained over the long term. The allowable 4 0 0***surveillance interval is the interval between a check, test, or calibration, whichever is appropriate to the itembeing subjected to the surveillance, and is measured from the date of the last surveillance. Allowable surveillance intervals shall not exceed the following: 1.37.1 Annual -interval not to exceed fifteen (15) months.1.37.2 Semiannual -interval not to exceed seven and a half (7.5) months.1.37.3 Quarterly -interval not to exceed four (4) months.1.37.4 Mothy- interval not to exceed six (6) weeks.1.37.5 Wee..y- interval not to exceed ten (10) days.1.38 Unreviewed.Safety Questions. A proposed change, test or experiment shall be deemed to involvean unreviewed safety question:

a. If the probability of occurrence or the consequences of an accident or malfunction ofequipment important to safety previously evaluated in the safety analysis report may beincreased; orb. If a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; orc. If the margin of safety, as defined in the Basis for any technical specification, is reduced.1.39'Value.

Measured. The measured value is the value of a parameter as it appears on the output of a* ,. channel." ~1.40 Value, Tr~ue. The true value is the actual value of a parameter. 1.41 Watc.h~doa Circuit. The watchdog circuit is a surveillance circuit provided by the Data Acquisition Computer (DAC) and the Control System Computer (CSC) to ensure proper operation of the reactorcomputerized control system.2.0 Safety Limit an~d Limiting Safety System Setting (LSSS).2.1. Safety Limits.Applicability -This specification applies to the temperature of the reactor fuel in a standard TRIGA fuelelement.Obiective -The objective is to define the maximum temperature that can be- permitted with confidence

• that no damage to the fuel element cladding will result.Specific~ation

-a. The maximum fuel temperature in a standard TRIGA fuel element shall not exceed 930 0C duringsteady-state operation.

b. The maximum ten'perature in a standard TRIGA fuel element shall not exceed 1100 0C during pulseoperation.

....... 'a. This fuel safety limit applies for coniditions in which the cladding temperature is above 500 °C (SafetyAnalysis Report (SAR), Chapter 4, Section 4.5.4.1.3). The important parameter for a TRIGA reactor is5 0 =°the fuel element temperature. This parameter is well suited as it can be measured directly. A loss in theintegrity of the fuel element cladding could arise if the cladding stress exceeds the ultimate strength ofthe cladding material. The fuel element cladding stress is a function of the element's internal pressurewhile the ultimate strength of the cladding material is a function of its temperature. The cladding stressis a result of the internal pressure due to the presence of air, fission product gasses and hydrogen fromthe disassociation of hydrogen and zirconium in the fuel moderator. Hydrogen pressure is the mostsignificant. The magnitude of the pressure is determined by the fuel moderator temperature and theratio of hydrogen to zirconium in the alloy. At a fuel temperature of 930 0C for ZrH 1.7 fuel, the claddingstress due to the internal pressure is equal to the ultimate strength of the cladding material at the sametemperature (SAR Fig 4.18). This is a conservative limit since the temperature of the cladding material isalways lower than the fuel temperature. (See SAR Chapter 4, Section 4.5.4.)b. This fuel safety limit applies for conditions in which the cladding temperature is less than 500 °C.Analysis (SAR Chapter 4, Section 4.5.4.1.1), shows that a maximum temperature for the clad during apulse which gives a peak adiabatic fuel temperature of 1000 °C is estimated to be 470 °C. Furtheranalysis (SAR Section 4.5.4.1.2), shows that the internal pressure for both Zr (at 11500C) and Zr17z (at11 00°C) increases to a peak value at about 0.3 sec, at which time the pressure is about one-fifth of theequilibrium value or about 400 psi (a stress of 14,700 psi). The yield strength of the cladding at 500 0C isabout 59,000 psi.Calculations for step increases in power to peak ZrH 1.85 fuel temperature greater than I115 0 C, over a 200°Crange, show that the time to reach the peak pressure and the fraction of equilibrium pressure value achievedwere approximately the same as for the 1150 °C case. Similar results were found for fuel with ZrH1 .7.Measurements of hydrogen pressure in TRIGA fuel elements during transient operations have been made andcompared with the results of analysis similar to that used to make the above prediction. These measurements indicate that in a pulse where the maximum temperature in the fuel was greater than 1000 0C, the pressure(ZrH1.=) was only about 6% of the equilibrium value evaluated at the peak temperature. Calculations of thepressure gave values about three times greater than the measured values. The analysis gives strongindications that the cladding will not rupture if fuel temperatures are never greater than 1200 °C to 1250°C,providing the cladding temperature is less than 500 0 C. For fuel with ZrH 1.7,a conservative safety limit is 1100 0C.As a result, at this safety limit temperature, the class pressure is a factor of 4 lower than would be necessary forcladding failure.2.2 Limiting Safety System Setting.2.2.1 Fuel Temperature. Applicability -This specification applies to the protective action for the reactor fuel elementtemperature. Obiective -The objective is to prevent the fuel element temperature safety limit from beingreached.Specification -The limiting safety system setting shall be 750 0C (operationally this may be setmore conservatively) as measured in an instrumented fuel element. One instrumented elementshall be located in the analyzed peak power location of the reactor operational core.ass- For steady-state operation of the reactor, the limiting safety system setting is atemperature which, if exceeded, shall cause a reactor scram to be initiated preventing the safetylimit from being exceeded. A setting of 750 °C provides a safety margin at the point of themeasuremenrA of at least 137 °C for standard TRIGA fuel elements in any condition of operation. A part of the safety margin is used to account for the difference between the true and measuredtemperatures resulting from the actual location of the thermocouple. If the thermocouple element is located in the hottest position in the core, the difference between the true and!. / measured temperatures will be only a few degrees since the thermocouple junction is near thecenter and mid-plane of the fuel element. For pulse operation of the reactor, the same limitingsafety system setting applies.

However, the temperature channel will have no effect on limiting6 O ......0the peak power generated because of its relatively long time constant (seconds) as comparedwith the width of the pulse (milliseconds).

in this mode, however, the temperature trip will act to"\ limit the energy release after the pulse if the transient rod should not reinsert and the fuel! temperature continues to increase. 3.0 Limiting Conditions For 3.1 Reactor Core Parameters 3.1.1 Steady-State Ooeration Ajj lcblv- This specification applies to the maximum reactor power attained during steady-state operation. Obiective -The objective is to assure that the reactor safety limit (fuel temperature) is notexceeded, and to provide for a setpoint for the high flux limiting safety systems, so thatautomatic protective action will prevent the safety limit from being reached during steady-state operation. Soecification -The nominal reactor steady-state .power shall not exceed 2.0 MW. The automatic scram setpoints for the reactor power level safety channels shall be set at 2.2 MW or less. Forthe purpose of testing the reactor steady-state power level scram, the power shall not exceed2.3 MW.Basis -Operational experience and thermal-hydraulic calculations demonstrate that UCD/MNRCTRIGA fuel elements may be safely operated at power levels up to 2.3 MW with naturalconvection cooling. (SAR Chapter 4, Section 4.6.2.).3.1.2 Pulse or Square Wave Operation Aoolicabilitv -This specification applies to the peak temperature generated in the fuel as theresult of a step insertion of reactivity. Obiective -The objective is to assure that the fuel temperature safety limit will not be exceeded. Specification -a. For the pulse mode of operation, the maximum insertion of reactivity shall be 1.23% A dkf($1.75);b. For the square wave mode of operation, the maximum insertion of reactivity shall be 0.63%Ak/k ($0.90).Basis -Standard TRIGA fuel is fabricated with a nominal hydrogen to zirconium ratio of 1.6 to1.7. This yields delta phase zirconium hydride which has a high creep strength and undergoes no phase changes at temperatures in excess of 100 0C. However, after extensive steady stateoperation at two (2) MW the hydrogen will redistribute due to migration from the central hightemperature regions of the fuel to the cooler outer regions. When the fuel is pulsed, theinstantaneous temperature distribution is such that the highest values occur at the radial edge ofthe fuel. The higher temperatures in the outer regions occur in fuel with a hydrogen to zirconium ratio that hasnow increased above the nominal value. This produces hydrogen gas pressures considerably in excess of that expected. If the pulse insertion is such that the temperature of thefuel exceeds about 875 0C, then the pressure may be sufficient to cause expansion ofmicroscopic holes in the fuel that grow with each pulse. Analysis (SAR Chapter 13, Section.II 13.2.2.2.1), shows that the limiting pulse, for the worst case conditions, is 1.34% A k/k ($1.92).Therefore, the 1.23% Ak/dk ($1.75) limit is below the worse case reactivity insertion accident limit.7

• ... ... .The $0.90 square wave step insertion limit is also well below the worse case reactivity insertion accident limit.) 3.1.3 Reactivity Limitations Aolcbiiv-These specifications apply to the reactivity conditions of the reactor core and thereactivity worths of the control rods and apply to all modes of reactor operation.

bicie- The objective is to assure that the reactor can be placed in a shutdown condition atall times and to assure that the safety limit shall not be exceeded. Specification -a. Shutdo wn Marginl -The reactor shall not be operated unless the shutdown margin provided bythe control rods is greater than 0.35% k/k ($0.50) with:(1) The reactor in any core condition, (2) The most reactive control rod assumed fully withdrawn, and(3) Absolute value of all movable experiments analyzed in their most reactive condition or $1.00 whichever is greater.b. Excess Reactivity -The maximum available excess reactivity (reference core condition) shallnot exceed 6.65% A k/k ($9.50).Basis -a. This specification assures that the reactor can be placed in a shutdown condition from anyoperating condition and remain shutdown, even if the maximum worth control rod should stick inthe fully withdrawn position (SAR Chapter 4, Section 4.5.5).b. This specification sets an overall reactivity limit which provides adequate excess reactivity tooverride the xenon buildup, to overcome the temperature change in going from zero power to 2MW, to permit pulsing at the $1.75 level, to permit irradiation of negative worth experiments andaccount for fuel burnup over time. An adequate shutdown margin exists with an excess of $9.50for the two analyzed cores: (SAR Chapter 4, Section 4.5.5).3.2 Reactor Control and Safety Systems3.2.1 Control RodsAooljcjili*- This specification applies to the function of the control rods.Obiect~ive -The objective is to determine that the control rods are operable. Specificati~on -The reactor shall not be operated unless the control rods are operable and,a. Control rods shall not be considered operable if damage is apparent to the rod or driveassemblies.

b. The scram' time measured from the instant a signal reaches the value of a limiting safetysystem setting to the instant that the slowest control rod reaches its fully inserted position shall.not exceed one (1) second.8 0 Sa. The apparent condition of the control rod assemblies shall provide assurance that the rodsshall continue to perform reliably as designed.

b. This assures that the reactor shall shut down promptly when a scram signal is initiated (SARChapter 13, Section 13.2.2.2.2).

3.2.2 Reactor Instrumentation Apolicability -This specification applies to the information which shall be available to the reactoroperator during reactor operations. Obiective -The objective is to require that sufficient information is available to the operator toassure safe operation of the reactor.Specification -The reactor shall not be operated unless the channels described in Table 3.2.2are operable and the information is displayed on the reactor console.Table 3.2.2Required Reactor Instrumentation (Minimum Number Operable) Measuring Steady Square Channel Surveillance Channel Stt Pulse Wave Function Requirements*

a. Reactor Power 202Scram at 2.2 D,M,ALevel Safety MW or lessChannelb5. Linear Power 101Automatic D,M,AChannel -Power Controlc. Log Power 101Startup D,M,AChannel Controld. Fuel Temperature 2 2 2 Fuel D,M,AChannel Temperature

e. Pulse Channel 0 10Measures P,APulse NV & NVTr(*) Where: D -Channel check during each day's operation M -Channel test monthlyA -Channel calibration annuallyP -Channel test prior to pulsing operation

a. Table 3,2.2. The two reactor power level safety channels assure that the reactor power levelis properly mdonitored and indicated in the reactor control room (SAR Chapter 7, Sections 7.1.2 &7.1.2.2).

b. c. & e. Table 3.2.2. The linear power channel, log power channel, and pulse channel assurethat the reactor power level and energy are adequately monitored (SAR Chapter 7, Sections.... 7.1.2 & 7.1.2.2).

9 Ia1,.1.2 The fuel temperature channels assure that the fuel temperature is properlymonitored and indicated in the reactor control room (SAR Chapter 4, Section 4.5.4.1). 3.2.3 Reactor Scrams and Interlocks Agllcbjlity-This specification applies to the scrams and interlocks. Obiective -The objective is to assure that the reactor is placed in the shutdown condition promptly and that the scrams and interlocks are operable for safe operation of the reactor.Specification -The reactor shall not be operated unless the scrams and interlocks described inTable 3.2.3 are operable: Table 3.2.3Required Scrams and I!nterlocks SteadyStateScrama. ConsoleManualScramb. Reactor RoomManual ScramI1PulseIISquareWave.1ChannelSurveillance Reouirements* Manual Scramand Automatic Scram AlarmManual Scramand Automatic Scram Alarmc. Radiography Bay ManualScramsd. Reactor PowerLevel SafetyScramse. High VoltagePower SuppliesScramsf. FuelTemperature Scramsg. Watchdog.Circuit4420422Manual Scramsand Automatic Scram AlarmsAutomatic Scram Alarms & Scramsat 2.2 MW or lessAutomatic Scram Alarms &Scrams onLosfHigh Voltage tothe Reactor PowerLevel SafetyChannelsMMMMMMM212222Automatic ScramAlarms & Scramson indicated fueltemperature of750°C or lessAutomatic ScramAlarms & Scrams2210N., 0h. ExternalScrams22\i. One KilowattPulse &Square WaveInterlock

j. Low SourceLevel RodWithdrawal ProhibitInterlock

k. Control RodWithdrawal Interlock I. MagnetPower KeySwitch Scram0111Automatic Scrams and Alarmsif an experiment or radiography scram interlock

• is activated Prevents initiation of a step reactivity insertion above areactor power levelof 1 KWPrevents withdrawal of any control rodif the log channelreads less than 1.5times the indicated log channel currentlevel with the neutronsource removed fromthe corePrevents simul-taneous withdrawal of two or more rodsin manual modeDe-energizes thecontrol rodmagnets, scram &alarmMMMMI1111M(*) Where: M -channel test monthlyBasis -a. Table 3.2.3. The console manual scram allows rapid shutdown of the reactor from the controlroom (SAR Chapter 7, Section 7.1.2.5).

b. Table 3..2.3. The reactor room manual scram allows rapid shutdown of the reactor from thereactor room.p,. Table 3.2.3. The radiography bay manual scrams allow rapid shutdown of the reactor fromany of the radiography bays (SAR Chapter 9, Section 9.6.3).d~i.Tabe32.3.

The automatic power level safety scram assures the reactor will be shutdown ifthe power level exceeds 2.2 MW, therefore not exceeding the safety limit (SAR Chapter 4,Section 4.7.2).e. Table The loss-of-high-voltage scram assures that the reactor power level safetychannels, operate within their intended range as required for proper functioning of the powerlevel scrams (SAR Chapter 7, Sections 7.1.2.1 & 7.1.2.2).

f. Table 3.2.3. The fuel temperature scrams assure that the reactor will be shut down if the fueltemperature exceeds 7500o C, therefore ensuring the safety limit will not be exceeded (SARChapter 4, Sections 4.5.4.1 & 4.7.2).* )11

a. Table 3.2.3, The watchdog circuits assure that the control system computer and the dataacquisition computer are functioning properly (SAR Chapter 7, Section 7.2).h. Table.3.2.3, The external scrams assure that the reactor will be shut down if the radiography bay doors and reactor concrete shutters are not in the proper position for personnel entry intothe bays (SAR Chapter 9, Section 9.6). External scrams from experiments, a subset of theexternal scrams, also assure the integrity of the reactor system, the experiment, the facility, andthe safety of the facility personnel and the public.i. Table 3.2.3. The interlock preventing the initiation of a step reactivity insertion at a level aboveone (1) kilowatt assures that the pulse magnitude will not allow the fuel element temperature toexceed the safety limit (SAR Chapter 7, Section 7.1.2.5).

i. Table 3.2.3. The low source level rod withdrawal prohibit interlock assures an adequate sourceof neutrons is present for safe startup of the reactor (SAR Chapter 7, Section 7.1.2.5).

k. Table 3.2.3. The control rod withdrawal interlock prevents the simultaneous withdrawal of twoor more control rods thus limiting the reactivity-insertion rate from the control rods in manualmode (SAR Chapter 7, Section 7.1.2.5).

I. Table 3.2.3.. The magnet current key switch prevents the control rods frown being energized without inserting the key. Turning off the magnet current key switch de-energizes the control rodmagnets and results in a scram (SAR Chapter 7, Section 7.1.2.5). 3.2.4 Rea~ctor Fuel ElementsAoolicabilitv -This specification applies to the physical dimensions of the fuel elements asmeasured on the last surveillance test.Objective -The objective is to veri{fy the integrity of the fuel-element cladding. Specification -The reactor shall not be used for normal operation with damaged fuel. All fuelelements shall be inspected visually for damage or deterioration as per Technical Specifications Section 4.2.4. A fuel element shall be considered damaged and must be removed from the coreif:a. In measuring the transverse bend, the bend exceeds 0.125 inch (3.175 mam) over the fulllength 23 inches (584 mm) of the cladding, or,b. In measuring the elongation, its length exceeds its initial length by 0.125 inch (3.175 mam), or,c. A cladding failure exists as indicated by measurable release of fission products, or,d. Visual inspection identifies bulges, gross pitting, or corrosion. Basis. -The most severe stresses induced in the fuel elements result from pulse operation of thereactor, during which differential expansion between the fuel and the cladding occurs and thepressure of the gases within the elements increases sharply. The above limits on the allowable distortion of a fuel element correspond to strains that are considerably lower than the strainexpected to ".ause rupturing of a fuel element. Limited operation in the steady state or pulsedmode may be necessary to identify a leaking fuel element especially if the leak is small.12 3.3 Reactor Coolant SystemsAoolicability -These specifications apply to the operation of the reactor water measuring systems.Objective. -The objective is to assure that adequate cooling is provided to maintain fuel temperatures below the safety limit, and that the water quality remains high to prevent damage to the reactor fuel.Specification -The reactor shall not be operated unless the systems and instrumentation channelsdescribed in Table 3.3 are operable, and the information is displayed locally or in the control room.Table 3.3REQUIRED WATERSYSTEMS AND INSTRUMENTATION I I I n I ....Measuring Chann~el/System MinimumNumberOoe~rable. Surveillance Requirements*

a. Primary CoolantCore InletTemperature Monitorb. Reactor TankLow Water*Monitorc. Purification**

Inlet Conduc-tivity Monitord. Emergency CoreCooling SystemI11IFunction: Channel/System For operation of thereactor at 1.5 MW orhigher, alarms on highheat exchanger outlettemperature of 45 °C(113°F)Alarms if water leveldrops below a depth of23 feet in the reactor tankAlarms if the primarycoolant water conductivity is greater than5 micromhos/cm For operation of the reactorat 1.5MW or higher, provideswater to cool fuel in the eventof a Lois of Coolant Accidentfor a minimum of 3.7 hoursat 20 gpm from an appropriate nozzleD,Q,AM(.DM,SD,S(*) Where: D -channel check during each day's operation A -channel calibration annuallyQ -channel test quarterly S -channel calibration semiannually M -channel test monthly(**) The purification inlet conductivity monitor can be out-of-service for no more than 3 hours before the reactorshall be shutdown. Basis -a.Table 3.3, The primary coolant core inlet temperature alarm assures that large power fluctuations willnot occur (S.AR Chapter 4, Section 4.6.2).13

b. Table 3,3, The minimum height of 23 ft. of water above the reactor tank bottom guarantees that thereis sufficient water for effective cooling of the fuel and that the radiation levels at the top of the reactor* .',tank are within acceptable limits. The reactor tank water level monitor alarms if the water level drops) below a height of 23 ft. (7.01m) above the tank bottom (SAR Chapter 11, Section 11.1.5.1).

c. Table 3.3. Maintaining the primary coolant water conductivity below 5 micromhos/cm averaged over aweek will minimize the activation of water impurities and also the corrosion of the reactor structure.

d. Table 3.3. This system will mitigate the.Loss of Coolant Accident event analyzed in the SAR Chapter13, Section 13.2.3.4 Reactor Room Exhaust; SystemAoplicability

-These specifications apply to the operation of the reactor room exhaust system.Obiective -The objectives of this specification are as follows:a. To reduce concentrations of airborne radioactive material in the reactor room, and maintain thereactor room pressure negative with respect to surrounding areas.b. To assure continuous air flow through the reactor room in the event of a Loss of Coolant Accident. Soecification -a. The reactor shall not be operated unless the reactor room exhaust system is in operation and thepressure in the reactor room is negative relative to surrounding areas.~b. The reactor room exhaust system shall be operable within one half hour of the onset of a Loss ofCoolant Accident. Basi__.s -Operation of the reactor room exhaust system assures that:a. Concentrations of airborne radioactive material in the reactor room and in air leaving the reactor roomwill be reduced due to mixing with exhaust system air (SAR Chapter 9, Section 9.5.1). Pressure in thereactor room will be negative relative to surrounding areas due to air flow patterns created by the reactorroom exhaust system (SAR Chapter 9, Section 6.5.1).b. There will be a timely, adequate and continuous air flow through the reactor room to keep the fueltemperature below the safety limit in the event of a Loss of Coolant Accident. 3.5 This section intentionally left blank.3.6 T~his section intentionally left blan~k.3.7 Reactor Radiation Monitoring Systems.3.7.1 Monitoring SystemsAnoDlicability -This specification applies to the information which shall be available to the reactoroperator during reactor operation. Obiective -The objective is to require that sufficient information regarding radiation levels and, radioactive effluents is available to the reactor operator to assure safe operation of the reactor...... ' Specfication .-The reactor shall not be operated unless the channels described in Table 3.7.1are operable, the readings are below the alarm setpoints, and the information is displayed in the14 control room. The stack and reactor room CAMS shall not be shutdown at the same time duringreactor operation. Table 3.7.1REQUIRED RADIATION MONITORING INSTRUMENTATION Measuring Eduioment MinimumNumberO, erable**ChannelFunctionSurveillance Requirements*

a. FacilityStack Monitorb. Reactor RoomRadiation Monitorc. Purification System Radia-tion Monitord. Reactor RoomContinuous Air MonitorI111Monitors Argon-41 andradioactive particu-lates, and alarmsMonitors the radiation level in the reactorroom and alarmsMonitors radiation level at the demineral-izer station and alarmsMonitors air from thereactor room for parti-culate and gaseousradioactivity and alarmsD,W,AD),W,AD:,W,AD ,W.A(*) Where: D -channel check during each day'ls operation A -channel calibration annuallyW -channel test(**) monitors may be placed out-of-service for up to 2 hours for calibration and maintenance.

During this out-of-service time, no experiment or maintenance activities shall be conducted which could result in alarm conditions (e.g., airborne releases or high radiation levels)Basis -a. Thble 3.7.1. The facility stack monitor provides information to operating personnel regarding therelease of radioactive material to the environment (SAR Chapter 11, Section 11.1.1.1.4). The alarmsetpoint on the facility stack monitor is set to limit Argon-41 concentrations to less than 10 CFR Part 20,Appendix B, Table 2, Column 1 values (averaged over one year) for unrestricted locations outside theoperations area.b. Table 3.7.1. The reactor room radiation monitor provides information regarding radiation levels in thereactor room during reactor operation (SAR Chapter 11, Section 11.1.5.1 ), to limit occupational radiation exposure to less than 10 CFR 20 limits.c.; Table 3.7.1. The radiation monitor located next to the purification system resin cannisters providesinformation regarding radioactivity in the primary system cooling water (SAR Chapter 11, Section11.1.5.4.2) and allowS, assessment of radiation levels in the area to ensure that personnel radiation doses will be below 10 CFR Part 20 limits.d. Table 3.7.1. The reactor room continuous air monitor provides information regarding airborneradioactivity in the reactor room, (SAR Chapter 11, Sections 11.1.1.1.2 & 11.1.1.1.5), to ensure thatoccupational exposure to airborne radioactivity will remain below the 10 CFR Part 20 limits..,:15 3.7.2 .Effluents -.Argon-41 Discharge LimitAppJicaiity~- This specification applies to the concentration of Argon-41 that may be discharged from the UCD/MNRC reactor facility, Obiective -The objective is to ensure that the health and safety of the public is not endangered by the discharge of Argon-41 from the UCD/MNRC reactor facility. Soecification -The annual average unrestricted area concentration of Argon-41 due to releasesof this radionuclide from the UCD/MNRG, and the corresponding annual radiation dose fromArgon-41 in the unrestricted area shall not exceed the applicable levels in 10 CFR Part 20.Basis -The annual average concentration limit for Argon-41 in air in the unrestricted area isspecified in Appendix B, Table 2, Column 1 of 10 CFR Part 20.10 CFR 20.1301 specifies doselimitations in the unrestricted area. 10 CFR 20.1101 specifies a constraint on air emissions ofradioactive material to the environment. The SAR Chapter 11, Section 11.1.1.1.4 estimates thatthe routine Argon-41 releases and the corresponding doses in the unrestricted area will bebelow these limits.3.8 Experiments 3.8.1 React~ivity Limnits.Applicability -This specification applies to the reactivity limits on experiments installed in specificreactor experiment facilities. Obiective -The objective is to assure control of the reactor during the irradiation or handling ofexperiments in the specifically designated reactor experiment facilities. Specification -The reactor shall not be operated unless the following conditions governing experiments exist:a. The absolute reactivity worth of any single moveable experiment in the pneumatic transfertube, the central irradiation

facility, the central irradiation fixture 1 (CIF-1),

or any other in-core'or in-tank irradiation

facility, shall be less than $1.00 (0.7% A k/k), except .for the automated centralirradiation facility (ACIF) (See 3.8.1.c below).b. The absolute reactivity worth of any single secured experiment positioned in a reactor in-coreor in-tank irradiation facility shall be less than the maximum allowed pulse ($1.75) (1.23% A k/k).c. The absolute total reactivity worth of any single experiment or of all experiments collectively positioned in the ACIF shall be less than the maximum allowed pulse ($1.75) (1.23% A k/k).d. The absolute total reactivity of all experiments positioned in the pneumatic transfer tube, andin any other reactor in-core and in-tank irradiation facilities at any given time shall be less thanone dollar and ninety-two cents ($1.92) (1.34% A k/k), including the potential reactivity whichmight result from malfunction,

flooding, voiding, or removal and insertion of the experiments.

Basis -a. A limitatiodn of less than one dollar ($1.00) (0.7% A k/k) on the reactivity worth of a singlemovable experiment positioned in the pneumatic transfer tube, the central irradiation facility(SAR, Chapter 10, Section 10.4.1), the central irradiation fixture-I (CIF-1) (SAR Chapter 10,Section 10.4.1), or any other in-core or in-tank irradiation

facility, will assure that the pulse limitof $1 .75 is not exceeded (SAR Chapter 13, Section 13.2.2.2.1

). In addition, limiting the worth ofeach movable experiment to less than $1.00 will assure that the additional increase in transient 16 power and temperature will be slow enough so that the fuel temperature scram will be effective (SAR Chapter 13, Section 13.2.2.2.1).

b. The absolute worst event which may be considered in conjunction with a single securedexperiment is its sudden accidental or unplanned removal while the reactor is operating.

Forsuch an event, the reactivity limit for fixed experiments ($1.75) would result in a reactivity increase less than the $1.92 pulse reactivity insertion needed to reach the fuel temperature safety limit (SAR Chapter 13, Section 13.2.2.2.1).

c. A reactivity limit of less than $1.75 for any single experiment or for all experiments collectively positioned in the sample can of the automated central irradiation facility (ACIP) (SAR Chapter10, Section 10.4.2) is based on the pulsing reactivity insertion limit (Technical Specification 3.1.2) (SAR Chapter 13, Section 13.2.2.2.1) and on the design of the ACIF, which allows controlover the positioning of samples into and out of the central core region in a manner identical inform, fit, and function to a control rod.d. It is conservatively assumed that simultaneous removal of all experiments positioned in thepneumatic transfer tube, and in any other reactor in-core and in-tank irradiation facilities at anygiven time shall be less thanthe maximum reactivity insertion limit of $1.92. The SAR Chapter13, Section 13.2.2.2.1 indicates that a pulse reactivity insertion of $1.92 would be needed toreach the fuel temperature safety limit.3.8.2 .Materials LimitAoplicability

-This specification applies to experiments installed in reactor experiment facilities. Objective -The objective is to prevent damage to the reactor or significant releases of.,. by limiting material quantity and the radioactive material inventory of theexperiment. .Specification -The reactor shall nct be operated unless the following conditions governing experiment materials exist;a. Experiments containing materials corrosive to reactor components, compounds highlyreactive with water, potentially explosive materials, and liquid, fissionable materials shall beappropriately encapsulated.

b. Each fueled experiment shall be controlled such that the total inventory of iodine isotopes 131through 135 in the experiment is no greater than 1.5 curies and the maximum strontium inventory is no greater than 5 millicuries.

c. Each experiment in the 1-125 production facility shall be controlled such that the totalinventory of 1-125 in the 1-125 glove box shall not exceed 40 curies.d. Each experiment in the 1-125 production facility shall be controlled such that the totalinventory of 1-125 being handled in the 1-125 fume hood at any one time in preparation forshipping shall not exceed 20 curies. An Additonal 1.0 curie of 1-125 (up to 400 millicuries in theform of quality assurance samples and up to 600 millicuries in sealed storage containers) mayalso be present in the 1-125 fume hood.e. Explosive

.materials in quantities greater than 25 milligrams of TNT eqluivalent shall not beirradiated in th~e reactor tank. Explosive materials in quantities of 25 milligrams of TNTequivalent or less may be irradiatedl provided the pressure produced upon detonation of theexplosive has been calculated and/or experimentally demonstrated to be less than the designpressure of the container.

f. Explosive materials in quantities of three (3) pounds of TNT equivalent or less may beirradiated in any radiography bay. The irradiation of explosives in any bay is limited to those.17

...... " 0assemblies where a safety analysis has been performed that shows that there is no damage tothe reactor safety systems upon detonation (SAR Chapter 13, Section 13.2.6.2). ") Basis -a. Appropriate encapsulation is required to lessen the experimental hazards of some types ofmaterials.

b. The 1.5 curies limitation on iodine 131 through 135 assures that in the event of failure of afueled experiment leading to total release of the iodine, occupational doses and doses tomembers of the general public in the unrestricted areas shall be within the limits in 10 CFR 20(SAR Chapter 13, Section 13.2.6.2).

c&d. Limiting the total 1-125 inventory to forty (40.0) curies in the 1-125 glove box and to twenty-one (21.0) curies in the 1-125 .fume hood assures that, if either of these inventories of 1-125 istotally released into its respective containment, or if both inventories are simultaneously released into their respective containments, the occupational doses and doses to members ofthe general public in the unrestricted areas will be within the limits of 10 CFR 20 (SAR Chapter13, Section 13.2.6.2).

e. This specification is intended to prevent damage to vital equipment by restricting the quantityof explosive materials within the reactor tank (SAR Chapter 13, Section 13.2.6.2).

f. The failure of an experiment involving the irradiation of 3 lbs TNT equivalent or less in anyradiography bay external to the reactor tank will not result in damage to the reactor controls orthe reactor tank. Safety Analyses have been performed (SAR Chapter 13, Section .13.2.6.2) which show that up to six (6) pounds of TNT equivalent can be safely irradiated in any.*. radiogaphy bay. Therefore, the three (3) pound limit gives a safety margin of two (2).3.8.3 Failure and Malfunctions Applicability

-This specification applies to experiments installed in reactor experiment facilities. Objective -The objective is to prevent damage to the reactor or significant releases ofradioactive materials in the event of an experiment failure.S~ecification -a. All experiment materials which could off-gas,

sublime, volatilize, or produce aerosols under:(1) normal operating conditions of the experiment or the reactor,(2) credible accident conditions in the reactor, or(3) where the possibility exists that the failure of an experiment could release radioactive gases or aerosols into the reactor building or into the unrestricted area, the quantity andtype of material in the experiment shall be limited such that the airborne radioactivity inthe reactor room will not result in exceeding the applicable dose limits in 10 CFR Part 20in the unrestricted area, assuming 100% of the gases or aerosols escapes.b. In calculatio[ns pursuant to (a) above, the following assumptions shall be used:(1) If the effluent from an experiment facility exhausts through a stack which is closed on* ~.* .,high radiation levels, at least 10% of the gaseous activity or aerosols produced will)* escape.

(2) If the effluent from an experiment facility exhausts through a filter installation designed for greater than 9g% efficiency for 0.3 micron and larger particles, at least 10%of these will escape.(3) For materials whose boiling point is above 130 00 and where vapors formed byboiling this material can escape only through an undistributed column of water above thecore, at least 10% of these vapors can escape.c. If a capsule fails and releases material which could damage the reactor fuel or structure bycorrosion or other means, an evaluation shall be made to determine the need for corrective action. Inspection and any corrective action taken shall be reviewed by the UCD/MNRC Directoror his designated alternate and determined to be satisfactory before operation of the reactor isresumed.Basis -a. This specification is intended to reduce the likelihood that airborne radioactivity in the reactorroom or the unrestricted area will result in exceeding the applicable dose limits in 10 CFR Part20.b. These assumptions are used to evaluate the potential airborne radioactivity release due to anexperiment failure (SAR Chapter 13, Section 13.2.6.2).

c. Normal operation of the reactor with damaged reactor fuel or structural damage is prohibited to avoid release of fission products.

Potential damage to reactor fuel or structure shall bebrought to the attention of the UCD/MNRC Director or his designated alternate for review toassure safe operation of the reactor (SAR Chapter 13, Section 13.2.6.2). 4.0 Surveillance Requirements_ General. The surveillance frequencies denoted herein are based on continuing operation of the reactor.Surveillance activities scheduled to occur during an operating cycle which can not be performed with the reactoroperating may be deferred to the end of the operating cycle. If the reactor is not operated for a reasonable time,a reactor system or measuring channel surveillance requirement may be waived during the associated timeperiod. Prior to reactor system or measuring channel operation, the surveillance shall be performed for eachreactor system or measuring channel for which surveillance was waived. A reactor system or measuring channel shall not be considered operable until it is successfully tested.4.1 Reactor Core P~arameters 4.1.1 Steady State Operation Apolicability -This specification applies to the surveillance requirement for the power levelmonitoring channels. O biecltive -The objective is to verify that the maximum power level of the reactor does notexceed the authorized limit.Specifi~cat~ion -An annual channel calibration shall be made of the power level monitoring

• channel.

If a channel is removed,

replaced, or unscheduled maintenance is performed, or a* significant cilfange in core configuration occurs, a channel calibration shall be required.

Discovery of noncompliance with this specification shall limit reactor operations to that requiredto perform the surveillance. Bss-The annual pwrlevel channel calibration will assure that the indicated reactor power.......level is correct.4.1.2 Shutdown.Margin and Exc(;ess Reactivity ................................................................ ~ Aplcbiiy These specifications apply to the surveillance requirements for reactivity control of....... /the reactor core.betve- The objective is to measure and verify the reactivity worth, performance, andoperability of those systems affecting the reactivity of the reactor..Specifica~tion -a. The total reactivity worth of each control rod and the shutdown margin shall be determined annually or following any significant change in core or control rod configuration. The shutdownmargin shall be verified by meeting the requirements of Section 3.1.3(a).

b. The core excess reactivity shall be verified:

(1) Prior to each startup operation and,(2) Following any change in core loading or configuration. Discovery of noncompliance with Technical Specifications 4.1.2.a-b shall limit reactor operations to that requiredto perform the surveillance. Basis -a. The reactivity worth of the control rods is measured to assure that the required shutdownmargin is available and to provide an accurate means for determining the excess reactivity ofthe core. Past experience with similar reactors gives assurance that measurements of th~e,..; control rod reactivity worth on an annual basis is adequate to assure that there are no significant changes in the shutdown margin, provided no core loading or configuration changes have beenmade.b. Determining the core excess reactivity prior to each reactor startup shall assure that Technical Specifications 3.1 .3.b shall be met, and that the critical rod positions do not changeunexpectedly. 4.2 Reactor Control and Sa~fet,! Systems4.2.1 Control RodsApplicability -This specification applies to the surveillance of the control rods.Objective -The objective is to inspect the physical condition of the reactor control rods andestablish the operable condition of the rods.Spoecification -Control rod worths shall be determined annually or after physical removal or anysignificant change in core or control rod configuration.

a. Each control rod shall be inspected at annual intervals by visual observation of the fueledsections and absorber sections plus examination of the linkages and drives.b. The scram time .of each control rod shall be measured semiannually.

I.Discovery of noncompliance with Technical Specifications 4.2.1 .a-b shall limit reactor operations to that requiredto perform the surveillance. ) ~s(ehia pcfctos4.2.1 .b)-Annual determination of control rod worths o". .....measurements after any physical removal or significant change in core loading or control rod/ -z1.

• 0configuration provides information about changes in reactor total reactivity and individual rodworths. The frequency of inspection for the control rods shall provide periodic verification of the* ...%condition of the control rod assemblies.

The specification intervals for scram time assureoperable performance of the control rods.4.2.2 Reactor Instrumentation A~oDlicability -These specifications apply to the surveillance requirements for measurements, tests, calibration and acceptability of the reactor instrumentation. Obie.ctive -The objective is to ensure that the power level instrumentation and the fueltemperature instrumentation are operable. Specification -a. The reactor power level safety channels shall have the following: (1) A channel test monthly or after any maintenance which could affect their operation. (2) A channel check during each day's operation. (3) A channel calibration annually. -b. The Linear Power Channel shall have the following: (1) A channel test monthly or after any maintenance which could affect the operation. (2) A channel check during each day's operation. (3) A channel calibration annually.

c. The Log Power Channel shall have the following:

(1) A channel test monthly or after any maintenance which could affect its operation. (2) A channel check during each day's operation. (3) A channel calibration annually.

d. The fuel temperature measuring channels shall have the following:

(1) A channel test monthly or after any maintenance which could affect operation. (2) A channel check during each day's operation. (3) A channel calibration annually.

e. The Pulse Energy Integrating Channel shall have the following:

-. (1) A channel test prior to pulsing operations. (2) A channel calibration annually.

• Discovery of noncompliance with Technical Specifications 4.2.2.a-e shall limit reactor operation to that required)to perform the surveillance.

Basis -

• a. A daily channel check and monthly test, plus the annual calibration, will assure that thereactor power level safety channels operate properly.

\J b. A channel test monthly of the reactor power level multi-range channel will assure that thechannel is operable and responds correctly. The channel check will assure that the reactorpower level multi-range linear channel is operable on a daily basis. The channel calibration annually of the multi-range linear channel will assure that the reactor power will be accurately measured so the authorized power levels are not exceeded.

c. A channel test monthly will assure that the reactor power level wide range log channel isoperable and responds correctly.

A channel check of the reactor power level wide range logchannel will assure that the channel is operable on a daily basis. A channel calibration willassure that the channel will indicate properly at the corresponding power levels.d. A channel test monthly and check during each day's operation, plus the annual calibration, willassure that the fuel temperature measuring channels operate properly.

e. A channel test prior to pulsing plus the annual channel calibration will assure the pulse energyintegrating channel operates properly.

4.2.3 Rea~ctor Scrams and Interlocks. .Applicability -These specifications apply to the surveillance requirements for measurements, test, calibration, and acceptability of the reactor scrams and interlocks. Obie~ctiyve -The objective is to ensure that the reactor scrams and interlocks are operable.Specification -a. Console Manual Scram. A channel test shall be performed monthly.b. Reactor Room Manual Scram. A channel test shall be performed monthly.c. Radiography Bay Manual Scrams. A channel test shall be performed monthly.d. Reactor Power Level Safety Scram. channel test shall be performed monthly.e. High-Voltage-Power Supply Scrams. A channel test shall be performed monthly.f. Fuel Temperature Scram. A channel test shall be performed monthly.g. Watchdog Circuits Scrams. A channel test shall be performed monthly.h. External Scrams. A channel test shall be performed monthly.i. The One Kilowatt Pulse interlock. A channel test shall be performed monthly.j. Low Source Level Rod Withdrawal prohibit Interlock. A channel test shall be performed monthly.k. Control Rdd Withdrawal Interlocks. A channel test shall be performed monthly.I. Magnet Power Key Switch Scram. A channel test shall be performed monthly.!Discovery of noncompliance with Specifications 4.2.3.a-I shall limit r'eactor operation to that required to perform..... the surveillance. Basis- .--.A channel test monthly of the Console Manual Scram will assure that the scram is operable. ~b. A channel test monthly of the Reactor Room Manual Scram will assure that the scram isoperable.

c. A channel test monthly of the Radiography Bay Manual Scrams will assure that the scramsare operable.

d. A channel test monthly of the Reactor Power Level Safety Scrams will assure that the scramsare operable.

e. A channel test monthly of the Loss-of-High-Voltage Scram will assure that the high voltagepower supplies are operable and respond correctly.

f. A channel test monthly of the Fuel Temperature Scrams will assure that the scrams areoperable.

g. A channel test monthly of the Watchdog Circuits Scrams will assure that the scram circuits areoperable.

h. A channel test monthly of the External Scrams will assure that the scrams are operable andrespond correctly.

i. A channel test monthly will assure that the One Kilowatt Pulse Interlock works properly.

j. A channel test monthly of the Low Source Level Rod Withdrawal Prohibit Interlock will assure*that the interlock is operable.

k. A channel test monthly of the Control Rod Withdrawal Interlock will assure that the interlock isoperable.

I. A channel test monthly of the Magnet Current Key Switch will assure that thescram isoperable. 4.2.4 Reactor Fuel Element~s This specification applies to the surveillance requirements for the fuel elements. Obiective -The objective is to verify the continuing integrity of the fuel element cladding. Sp~ecification -To assure the measurement limitations in Section 3.2.4 are met, the following shall be done:a. The lead elements (i.e., all elements adjacent to the transient rod, with the exception ofinstrumented fuel elements), and all elements adjacent to the central irradiation facility shall beinspected annually.

b. Instrumented fuel elements shall be inspected if any of the elements adjacent to it fail to pass* the visual and/or physical measurement requirements of Section 3.2.4. Discovery of* noncompliantee with Technical Specification 4.2.4 shall limit operations to that required toperform the surveillance.

Basis (Technical Specifications 4,2,4.a-b) -The above specifications assure that the lead fuelelements shall be inspected regularly adteintegrity.o h edfe lmnssalbmaintained. These are the fuel elements with the highest power density as analyzed in the SARChapter 4, Section 4.5.5.6. The instrumented fuel element is excluded to reduce the risk ofdamage to the thermocouples. 4.3 Reactor Coolant SystemsAoolicability -This specification applies to the surveillance requirements for the reactor water measuring systems and the emergency core cooling system..Ojective-The objective is to assure that the reactor tank water temperature monitoring system, thetank water level alarm, the water conductivity cells and the emergency core cooling system are alloperable. .Specification -a. The reactor tank core inlet temperature monitor shall have the following: (1) A channel check during each day's operation. (2) A channel test quarterly. (3) A channel calibration annually.

b. The reactor tank low water level monitoring system shall have the following:

-.-(1) A channel test monthly..

c. The purification inlet conductivity monitors shalt have the following:

(1) A channel check during each day's operation. ) (2) A channel test monthly.(3) A channel calibration semiannually.

d. The Emergency Core Cooling System shall have the following:

(1) A channel check prior to operation. (2) A channel calibration semiannually. Discovery of noncompliance with Technical Specifications 4.3.a-c shall limit operations to that required toperform the surveillance. Noncompliance with Technical Specification 4.3.d shall limit operations to less than 1.5MW.Basis -a. A channel test quarterly assures the water temperature monitoring system responds correctly to aninput signal. A channel check during each day's operation assures the channel is operable. A channelcalibration annually assures the monitoring system reads properly.

b. A channel test monthly assures that the low water level monitoring system responds correctly to aninput signal.c. A channel test monthly assures that the purification inlet conductivity monitors respond correctly to an.input signal. A channel check during each day's operation assures that the channel is operable.

A)channel calibration semiannually assures the conductivity monitoring system reads properly. " d. A channel check prior to operation assures that the emergency core cooling system is operable forpower levels above 1.5 MW. A channel calibration semiannually assures that the Emergency CoreCooling System performs as required for power levels above 1.5 MW.*"L'1 4.4 React lor Room Exha~ust System\ Applicability -This specification applies to the surveillance requirements for the reactor roomexhaust system.Objective -The objective is to assure that the reactor room exhaust system is operating properly. _Soecification -The reactor room exhaust system shall have a channel check during each day'soperation. Discovery of noncompliance with this specification shall limit operations to that required to perform thesurveillance. Basis -A channel check during each day's operation of the reactor room exhaust system shallverify that the exhaust system is maintaining a negative pressure in the reactor room relative tothe surrounding facility areas.4.5 This section intentionally left blank4.6 This section intentionally left blank. .4.7 ,Rea~ctor Radiation Monitoring SystemsApplicability -This specification applies to the surveillance requirements for the reactor radiation monitoring systems.Obiective -The objective is to assure that the radiation monitoring equipment is operating ) properly. Specification -a. The facility stack monitor shall have the following: (1) A channel check during each day's operation. (2) A channel test weekly.(3) A channel calibration annually.

b. The reactor room radiation monitor shall have the following:

(1) A channel check during each day's operation. (2) A channel test weekly.(3) A channel calibration annually.

c. The purification system radiation monitor shall have the following:

(1) A channel check during each day's operation:

• ) (2) A channel test weekly.,...j,(3)

A channel calibration annually.

d. The reactor room Continuous Air Monitor (CAM) shall have the following:

(1) A channel check during each day's operation. (2) A channel test weekly.(3) A channel calibration annually. Discovery of noncompliance with Technical Specifications 4.7.a-d shall limit operations to that required toperform the surveillance. Basis -a. A channel check of the facility stack monitor system during each day's operation will assurethe monitor is operable. A channel test weekly will assure that the system responds correctly toa known source. A channel calibration annually will assure that the monitor reads correctly.

b. A channel check of the reactor room radiation monitor during each day's operation will assurethat the monitor is operable.

A channel test weekly will ensure that the system responds to aknown source. A channel calibration of the monitor annually will assure that the monitor readscorrectly.

c. A channel check of the purification system radiation monitor during each day's operation assures that the monitor is operable.

A channel test weekly will ensure that the systemresponds to a known source. A channel calibration of the monitor annually will assure that themonitor reads correctly.

d. A channel check of the reactor room Continuous Air Monitor (CAM) during each day'soperation will assure that the CAM is operable.

A channel test weekly will assure that the CAMresponds correctly to a known source. A channel calibration annually will assure that the CAMreads correctly.

4.8 Experiments

Aoolicabilitv -This specification applies to the surveillance requirements for experiments installed in any UCD/MNRC reactor~experiment facility. Objective.- The objective is to prevent the conduct of experiments or irradiations which maydamage the reactor or release excessive amounts of radioactive materials as a result of failure.Soecification -a. A new experiment shall not be installed in any UCD/MNRC reactor experiment facility until awritten safety analysis has been performed and reviewed by the UCD/MNRC

Director, or hisdesignee, to establish compliance with the Limitations on Experiments, (Technical Specifications Section 3.8) and 10 CFR 50.59.b. All experiments performed at the UCDIMNRC shall meet the conditions of an approvedFacility Use Authorization.

Facility Use Authorizations and experiments carried out under theseauthorizations shall be reviewed and approved in accordance with the Utilization of the (UCD)McClellan N~zlear Radiation Center Research Reactor Facility Document (MNRC-0027-DOC). An experimenlt classified as an approved experiment shall not be placed in any UCDIMNRCexperiment facility until it has been reviewed for compliance with the approved experiment andFacility Use Authorization by the Reactor Manager and the Health Physics Manager, or theirdesignated alternates.

c. The reactivity worth of any experiment installed in the pneumatic transfer tube, or in any otherUCD/MNRC reactor in-core or in-tank irradiation facility shall be estimated or measured, as.................

~............... Iff_/,/ appropriate, before reactor operation with said experiment. Whenever a measurement is done it---.' shall be done at ambient conditions.

d. Experiments shall be identified and a log or other record maintained while experiments are inany UCD/MNRC reactor experiment facility.

Basis -a & b. Experience at most TRIGA reactor facilities verifies the importance of reactor staff andsafety committee reviews of proposed experiments.

c. Measurement of the reactivity worth of an experiment, or estimation of the reactivity worthbased on previous or similar measurements, shall verify that the experiment is within authorized reactivity limits.d. Maintaining a log of experiments while in UCD/MNRC reactor experiment facilities willfacilitate maintaining surveillance over such experiments.

5.0 Design Feat~ures 5.1 Site and Facility Description!.- 5.1.1 SiteApplicability -This specification applies to the UCD/MNRC site location and specific facilitydesign features. i" Objective. -The objective is to specify those features related to the Safety Analysis evaluation. Specification -a. The site location is situated approximately 8 miles (13 kin) north-by-northeast of downtownSacramento, California. The site of the UCD/MNRC facility is about 3000 ft. (0.6 mi or 0.9 kin)west of Watt Avenue, and 4500 ft. (0.9 mi or 1.4 kin) south of E Street.b. The restricted area is that area inside the fence surrounding the reactor building. Theunrestricted area is that area outside the fence surrounding the reactor building.

c. The TRIGA reactor is located in Building 258, Room 201 of the UCD/MNRC.

This buildinghas been designed with special safety features.

d. The core is below ground level in a water filled tank and surrounded by a concrete shield.Basis -a. Information on the surrounding population, the hydrology, seismology, and cliimatography ofthe site has been presented in Chapter 2 of the Safety Analysis Report.b. The restricted area is controlled by the UCD/MNRC Director.

c. The room b nclosi ng the reactor has been designed with systems related to the safe operation of the facility.

• .}/d. The below grade core design is to negate the consequences of an aircraft hitting the reactor..... building.

This accident was analyzed in Chapter 13 of the Safety Analysis Report, and found tobe beyond a credible accident scenario. 5.1.2 FcltExas,Applicability -This specification applies to the facility which houses the reactor..Obiective -The objective is to assure that provisions are made to restrict the amount ofradioactivity released into the environment, or during a Loss of Coolant Accident, the system isto assure proper removal of heat from the reactor room.Specification -a. The UCD/MNRC reactor facility shall be equipped with a system designed to filter andexhaust air from the UCD/MNRC facility. The system shall have an exhaust stack height of aminimum of 18.2rn (60 feet) above ground level.b. Manually activated shutdown controls for the exhaust system shall be located in the reactorcontrol room.Basis -The UCD/MNRC facility exhaust system is designed such that the reactor room shall bemaintained at a negative pressure with respect to the surrounding areas. The free air volumewithin the UCD/MNRC facility is confined to the facility when there is a shutdown of the exhaustsystem. Controls for startup, filtering, and normal operation of the exhaust system are located inthe reactor control room. Proper handling of airborne radioactive materials (in emergency situations) can be conducted from the reactor control room with a minimum of exposure tooperating personnel. 5.2 Reactor Coolant SystemApplicability -This specification applies to the reactor coolant system..Obiective -The objective is to assure that adequate water is available for cooling and shielding duringnormal reactor operation or during a Loss of Coolant Accident. Specification -a. During normal reactor operation the reactor core shall be cooled by a natural convection flow ofwater.b. The reactor tank water level alarm shall activate if the water level in the reactor tank drops below adepth of 23 ft.c. For operations at 1.5 MW or higher during a Loss of Coolant Accident the reactor core shall be cooledfor a minimum of 3.7 hours at 20 gpm by a source of water from the Emergency Core Cooling System.Basis -a. The SAR Chapter 4, Section 4.6, Table 4-19, shows that fuel temperature limit of 930 °C will not beexceeded under natural convection flow conditions.

b. A reactor tank water low level alarm sounds when the water level drops significantly.

This alarm* annunciates in the reactor control room and at a 24 hour monitored location so that appropriate corrective action can' be taken to restore water for cooling and shielding.

c. The SAR Chapter 13, Section 13.2, analyzes the requirements for cooling of the reactor fuel andi shows that the fuel safety limit is not exceeded under Loss of Coolant Accident conditions during this........

,/water cooling.5.3 Reactor Core and Fuel 5.3.1 RatrCrAorolicalbility -This specification applies to the configuration of the fuel.Obiective -The objective is to assure that provisions are made to restrict the arrangement of fuelelements so as to provide assurance that excessive power densities will not be produced. Soecification -For operation at 0.5 MW or greater, the reactor core shall be an arrangement of96 or more fuel elements to include fuel followed control rods. Below 0.5 MW there is nominimum required number of fuel elements. In a mixed 20/20, 30/20 and 8.5/20 fuel loading(SAR Chapter 4, Section 4.5.5.6): Mix J Core and Other Var'iations (1) No fuel shall be loaded into Hex Rings A or B,(2) A fuel followed control rod located in an 8.5 wt% environment shall contain 8.5 wt% fuel.,2.0E Core and Other Variations (1) No fuel shall be loaded into Hex Rings A or B. .(2) Fuel followed control rods may contain either 8.5 wt% or 20 wt% fuel.(3) Variations to the 20E core having 20 wt% fuel in Hex Ring C requires the 20 wt% fuel to beloaded into corner positions only, and graphite dummy elements in the flat positions. The*performance of fuel temperature measurements shall apply to variations to the as-analyzed 20Ecore configurations. 308 CoQre and Other Variations (1) No fuel shall be loaded into Hex Rings A or B.(2) The only fuel types allowed are 20120 and 30/20.(3) 20/20 fuel may be used in any position in Hex Rings C through G.(4) 30/20 fuel may be used in any position in Hex Rings 0 through G but not in Hex Ring C.(5) An analysis of any irradiation facility installed in the central cavity of this core shall be donebefore it is used with this core.Basis -In order to meet the power density requirements discussed in the SAR Chapter 4,Section 4.5.5.6, no less than 96 fuel elements including fuel followed control rods and the aboveloading restrictions will be allowed in an operational 0.5 MW or greater core. Specifications forthe 202 core and for the 30B core allow for variations of the as-analyze~1 core with the condition that temperature limits are being maintained (SAR Chapter 4, Section 4.5.5.6 and ArgonneNational Laboratory Report ANLIED 97-54).5.3.2 Reactor FuelIApplicability -These specifications apply to the fuel elements used in the reactor core.~Obiective -The objective is to assure that the fuel elements are of such design and fabricated in/ such a manner as to permit their use with a high degree of reliability with respect to theirphysical and nuclear characteristics. Sp~ecification.- The individual unirradiated TRIGA fuel elements shall have the following

• characteristics:

i a. Uranium content: 8.5, 20 or 30 wt % uranium enriched nominally to less than 20% U-235.b. Hydrogen to zirconium atom ratio (in the ZrH ,): 1.60 to 1.70 (1.65+/- 0.05).c. Cladding: stainless steel, nominal 0.5mm (0.020 inch) thick.Basis -a. The design basis of a TRIGA core loaded with TRIGA fuel demonstrates that limitingoperation to 2.3 megawatts steady state or to a 36 megawatt-sec pulse assures an amplemargin of safety between the maximum temperature generated in the fuel and the safety limit forfuel temperature. The fuel temperatures are not expected to exceed 630 °C during any condition of normal operation.

b. Analysis shows that the stress in a TRIGA fuel element, H/Zr ratios between 1.6 and 1.7, isequal to the clad yield strength when both fuel and cladding temperature are at the safety limit930°C. Since the fuel temperatures are not expected to exceed 630 0C during any condition ofnormal operation, there is a margin between the fuel element clad stress and its ultimatestrength.

c. Safety margins in the fuel element design and fabrication allow for normal mill tolerances ofpurchased materials.

5.3.3 Contr~ol .Rods and Control Rod Drives../ A oolicabilitv -This specification applies to the control rods and control rod drives used in thereactor core.Obiective -The objective is to assure the control rods and control rod drives are of such adesign as to permit their use with a high degree of reliability with respect to their physical,

nuclear, and mechanical characteristics.

.Specification -a. All control rods shall have scram capability and contain a neutron poison such as stainless steel, borated graphite, B 4C powder, or boron and its compounds in solid form. The shim andregulating rods shall have fuel followers sealed in stainless steel. The transient rod shall have anair filled follower and be sealed in an aluminum tube.b. The control rod drives shall be the standard GA rack and pinion type with an electromagnet and armature attached.

a. The neutron poison requirements for the control rods are satisfied by using stainless steel,neutron absorbing borated graphite, B 4C powder, or boron and its compounds.

These materials shall be contained in a suitable clad material such as stainless steel or aluminum to assuremechanical st~bility during movement and to isolate the neutron poison from the tank waterenvironment. Scram capabilities are provided for rapid insertion of the control rods.* " \b. The standard GA TRIGA control rod drive meets the requirements for driving the control rods, ...... jat the proper speeds, and the electromagnet and armature provide the requirements for" rapidinsertion capability. These drives have been tested and proven in many TRIGA reactors.

• A5.4 Fissionable Materiall Storaae...." A Dp1icabilitraco coe. This specification applies to the storage of reactor fuel at a time when it is nat in thereacto core* Objective

-The objective is to assure that the fuel which is being stored will not become critical and willnot reach an unsafe temperature.

• -a. All fuel elements not in the reactor core shall be stored (wet or dry) in a geometrical array where thekef is less than 0.9 for all conditions of moderation.

b. Irradiated fuel elements shall be stored in an array which shall permit sufficient natural convection cooling by water or air such that the fuel element temperature shall not exceed the safety limit.Bss- The limits imposed by Technical Specifications 5.4.a and 5.4.b assure safe storage.*6.0 Administrative Controls6.1 Organization.

The Vice Chancellor for Research shall be the licensee for the UCDIMNRC. TheUCD/MNRC facility shall be under the direct control of the UCD/MNRC Director.: The UCD/MNRCDirector shall be accountable to the Vice Chancellor for Research for the safe operation andmaintenance of the fac~ility.

• 6.1.1 'Struciture.

The management for operation of the UCD/MNRC facility shall consist of theorganizational structure as shown in Figure 6.1.6.1.2 R~esoonsibilities. The UCD/MNRC Director shall be accountable to the Vice Chancellor for* ) Research for the safe operation and maintenance of the facility. The UCD/MNRC

Director, or*his designated alternate, shall review and approve all experiments and experiment procedures prior to their use in the reactor.

Individuals irn the management organization (e.g., Operations

Manager, Reactor Supervisor, Health Physics Supervisor) shall be responsible for implementing UCD/MNRC policies and for operation of the facility, and shall be responsible for safeguarding the public and facility personnel from undue radiation exposures and for adhering to theoperating license and technical specifications.

The Operations Manager shall report directly tothe UCDIMNRC

Director, and shall immediately report all items involving safety and licensing tothe Director for a final decision.

The Reactor Supervisor and Health Physics Supervisor reportdirectly to the Operations Manager.6.1.3 Stffn6.1.3.1 The minim~irn staffing when the reactor is not shutdown shall be:a. A reactor operator in the control room; .b. A second person in the facility area who can perform prescribed instructions; c..A senior reactor operator readily available. The available senior reactoroperator should be within thirty (30) minutes of the facility and reachable bytelephone, and;d. A senior reactor operator shall be present whenever a reactor startup isperformed, fuel is being moved, or experiments are being placed in the reactor* tank.... .-...6.1.3.2 A list of reactor facility personnel by name and telephone number shall beavailable to the reactor operator In the control room. ,The list shall include:* 31

a. Management personnel.

., ~b. Health Physics personnel.

• " c. Reactor Operations personnel.

6.1.4 Selectio~n and.Training of Personnel. The selection, training and requalification of. operations

• personnel shall meet or exceed the requirements of the American National Standard for Selection and* Training of Personnel for Research Reactors (ANS 15:4). Qualification and requalification of licensedoperators shall be subject to an approved Nuclear Regulatory Commission (NRC) program.6.2 Review. Audit. Recommendation anld ApprovalGenleral Policy. Nuclear facilities shall be designed, constructed,

operated, and maintained insuch a manner that facility personnel, the general public, and both university and non-university property are not exposed to undue risk. These activities shall be conducted in accordance withapplicable regulatory requirements.

The UCD Vice Chancellor for Research shall institute the above stated policy as the facility--license holder. The Nuclear Safety Committee (NSC) has been chartered to assist in meeting.this responsibility by providing timely, objective, and independent

reviews, audits,* recommendations and approvals on matters affecting nuclear safety. The following describes the composition and conduct of the NSC.6.2.1 N.SC Comoosition and Qualifications,.

The UCD Vice Chancellor for Research shallappoint the Chairperson of the NSC. The NSC Chairperson shall appoint a Nuclear SafetyCommittee (NSC) of at least seven (7) members knowledgeable in fields which relate to nuclearsafety. The NSC shall evaluate and review nuclear safety associated with the operation and usei' of the UCD/MNRC, 6.2.2 NSC Ch~arte~r and Rules. The NSC shall conduct its review and audit (i~nspection) functions in accordance with a written charter. This charter shall include provisions for:a. Meeting frequency (The committee shall meet at least semiarnnually).

b. Voting rules.c. Quorums (For the full committee, a quorum will be at least seven (7) members).

d. A committee review function and an audit/inspection function.

e. Use of subcommittees.

f. Review, approval and dissemination of meeting minutes.6.2;3 Review Eunctio.

The responsibilities of the NSC, or a designated subcommittee thereof,shall include but are not limited to the following:

a. Review approved experiments utilizing UCD/MNRC nuclear facilities.

b. Review and approve all proposed changes to the facility,

license, the Technical Specifications and the Safety Analysis Report, and any new or changed Facility Use Authorizations andproposed Class I modifications, prior to implementing (Class I) modifications, prior to takingaction under the preceding documents or prior to forwarding any of these documents to thei' Nuclear Regulatory Commission for approval.

Review and determine whether a proposed change, test, or experiment would constitute anunreviewed safety question or req.uire a change to the license, to a Facility Use Authorization, or32 to the Technical Specifications. This determination may be in the form of verifying a decisionalready made by the UCD/MNRC Director. .d. Review reactor operations and operational maintenance, Class I modification

records, arid' ! the health physics program and associated records for all UCD/MNRC nuclear facilities.

e. Review the periodic updates of the Emergency Plan and Physical Security Plan forUCD/MNRC nuclear facilities.

f, Review and update t~he NSC Charter every two (2) years.g. Review abnormal performance of facility equipment and operating anomalies.

h. Review all reportable occurrences and all written reports .of such occurrence~s prior to.forwarding the final written report to the Nuclear Regulatory Commission.

i. Review the NSC annual audit/inspection of the UCD/MNRC nuclear facilities and any otherinspectionsof these facilitieg conducted by other agencies.

-*-6.2.4 Audit/Inspection Function. The NSC or a subcommittee

thereof, shall audit/inspect reactoroperations and health physics annually.

The annual audit/inspection shall include, but not be* limited to the following:

a. Inspection of the reactor operations and operational maintenance, Class I modification

• records, and the health physics program and associated

records, including the ALARA program,for all UCD/MNRC nuclear facilities.

b. Inspection of the physical facilities at the UCD/MNRC.

c. Examination of reportable events at the UCDIMNRC.

d. Determination of the adequacy of UCDIMNRC standard operating procedures.

e. Assessment of the effectiveness of the training and retraining progra.ms at the UCD/MNRC.

f. Determination of the conformance of operations at thle UCD/MNRC with the facility's licenseand Technical Specifications, and applicable regulations.

g. Assessment of the results of actions taken to correct deficiencies that have occurred innuclear safety related equipment, structures,

systems, or methods of operations.

h. Inspection of the currently ac~tive Facility Use Authorizations and associated experiments.

i. Inspection of future plans for facility modifications or facility utilization..

j. Assessment of operating abnormalities.

k. Determination of the status of previous NSC recommendations.

• 6.3 Radiation Safety. The Health Physics Supervisor shall be responsible for implementation of the*. ~UCD/MNRC Radiation Safety Program.

T~he program should use the guidelines of the .American National Standard for Radiation Protection at Research Reactor Facilities (ANSI/ANS 15.11). TheHealth Physics Supervisor shall report to the Operations Manager.* 6.4 Procedures. Written .procedures shall be prepared and approved prior to initiating any of the// activities listed in this section. The procedures shall be approved by the UCD/MNRC Director. A periodic.......review of procedures shall be performed and documented in a timely manner by the UCD/MNRC staff toassure that procedures are current, Procedures shall be adequate to assure the safe operation of the33

reactor, but shall not preclude the use of independent judgment and action should the situation require...... *Procedures shall be in effect for the following items:6.4.1 Reactor Operations Procedures

a. Startup, operation, and shutdown of the reactor.b. Fuel loading, unloading, and movement within the reactor.c. Control rod removal or replacement.

d. Routine maintenance of the control rod drives and reactor safety and interlock systems orother routine maintenance that could have an effect on reactor safety.e. Testing and calibration of reactor instrumentation and controls, control rods and control roddrives.f. Administrative controls for operations, maintenance, and conduct of irradiations andexperiments that could affect reactor safety or core reactivity.

g. Implementation of required plans such as emergency and security plans..h. Actions to be taken to correct potential malfunctions of systems, including responses toalarms and abnormal reactivity changes.6.4.2 HealthPhysics Procedures

a. Testing and calibration of area radiation

monitors, facility air monitors, laboratory radiation detection

systems, and portable radiation monitoring instrumentation.

b. Working in laboratories and other areas where radioactive materials are used.c. Facility radiation monitoring program including routine and special surveys, personnel monitoring, monitoring and handling of radioactive waste, and sampling and analysis of solidand liquid waste and gaseous effluents released from the facility.

The program shall include amanagement commitment to maintain exposures and releases as low as reasonably achievable (ALARA).d. Monitoring radioactivity in the environment surrounding the facility.

e. Administrative guidelines for the facility radiation protection program to include personnel orientation and training.

f. Receipt of radioactive materials at the facility, and unrestricted release of materials and itemsfrom the facility which may contain induced radioactivity or radioactive contamination..

g. Leak testing of sealed sources containing

.radioactive materials.

h. Special nuclear material accountability.

i. Transportation of radioactive materials.

Changes to the above procedures shall require approval of the UCD/MNRC Director. All such changes shall be... documented. 6.5 Experiment Review and Aporoval,. Experiments having similar characteristics are grouped togetherfor review and approval under specific Facility Use Authorizations. All specific experiments to be.3Lf performed under the provisions of an approved Facility Use Authorization shall be approved by theUCD/MNRC

Director, or his designated alternate.

a. Approved experiments shall be carried out in accordarnce with established and approved procedures.

b. Substantive change to a previously approved experiment shall require the same review and approvalas a new experiment.

c. Minor changes to an experiment that do not significantly alter the experiment may be approved by asenior reactor operator.

6.6 Req~uired Acti~ons. 6.6.1 Action to be taken in. case. of..a safety limit violation. In the event of a safety limit violation (fuel temperature), the following action shall be taken:a. The reactor shall be shut down and reactor operation shall not be resumed until authorized by the NRC.b. The safety limit violation shall be promptly reported to the UCD/MNRC Director.

c. The safety limit violation shall be reported to the chairman of the NSC and to the NRC by theUCD/MNRC Director.

d. A safety limit violation report shall be prepared.

The report shall describe the following: (1) Applicable circumstances leading to the violation, including when known, the causeand contributing factors.(2) Effect of the violation upon reactor facility components,

systems, or structures, andon the health and safety of personnel and the public.(3) Corrective action to be taken to prevent reoccurrence.

e. The safety limit violation report shall be reviewed by the NSC and then be submitted to theNRC when authorization is sought to resume operation of the reactor.6.6.2 Actions to be taken for reoortable occurrences.

In the event of reportable occurrences. the following actions shall be taken:a. Reactor conditions shall be returned to normal or the reactor shall be shut down. If it isnecessary to shut down the reactor to correct the occurrence, operations shall not be resumedunless authorized by the UCD/MNRC Director or his designated alternate.

b. The occurrence shall be reported to the UCDIMNRC Director or the designated alternate.

The UCD/MNRC Director shall report the occurrence to the NRC as required by these Technical Specifications or any applicable regulations.

c. Reportable occurrences should be verbally reported to the Chairman of the NSC and theNRC Operations Center within 24 hours of the occurrence.

A written preliminary report shall besent to the Attn: Document Control Desk, 1 White Flint North, 11555 Rockville Pike,Rockville MD 20852, within 14 days of the occurrence. A final written report shall be sent to theabove address within 30 days of the occurrence.

d. Reportable occurrences should be reviewed by the NSC prior to forwarding any writtenreport to the Vice Chancellorof the Office of Research or to the Nuclear Regulatory Commission.

6.7 Re..rt. All written reports shall be sent within the prescribed interval to the NRC, Attn: DocumentControl Desk, 1 White Flint North, 11555 Rockville Pike, Rockville MD 20852.6.7.10Operating

Repeorts, An annual report covering the activities of the reactor facility duringthe previous calendar year shall be submitted within six months following the end of eachcalendar year. Each annual report shall include the following~information:

a. A brief summary of operating experiences including experiments performed, changes infacility design, performance characteristics and operating procedures related to reactor safetyoccurring during the reporting period, and results of surveillance tests and inspections.

b. A tabulation showing the energy generated by the reactor (in megawatt hours), hours thereactor was critical, and the cumulative total energy output since initial criticality.

c. The number of emergency shutdowns and inadvertent scrams, including reasons for theshutdowns or scrams.d. Discussion of the major maintenance operations performed during the period, including theeffect, if any, on the safety of the operation of the reactor and the reasons for any corrective maintenance required.

-e. A brief description, including a summary of the safety evaluations, of changes in the facility orin procedures, and of tests and experiments carried out pursuant to Section 50.59 of 10 CFRPart 50.f. A summary of the nature and amount of radioactive effluents released or discharged to the/ ~environment beyond the effective control of the licensee as measured* at or prior to the point of,' such release or discharge, including the following: (1) Liquid Effluents (summarized on a monthly basis).(a) Liquid radioactivity discharged during the reporting period tabluated asfollows:1 The total estimated quantity of radioactivity released (in curies).2 An estimation of the specific activity for each detectable radionuclide present if the specific activity of the released material after dilution isgreater than 1x10"7 microcuries/ml. 3 A summary of the total release in curies of each radionuclide determined in 2_ above for the reporting period based on representative isotopic analysis. 4 An estimated average concentration of the released radioactive material at the point of release for each month in which a releaseoccurs, in terms of microcuries/mi and the fraction of the applicable concentration limit in 10 CFR 20.t: (b) The total volume (in gallons) of effluent water (including diluent) releasedduring each period of liquid effluent release..\}' (2) Airborne Effluents (summarized on a monthly basis):/'(a) Airborne radioactivity discharged during the reporting period (in curies)tabulated as follows: 1 The total estimated quantity of radioactivity released (in curies)determined by an appropriate sampling and counting method.2 The total estimated quantity (in curies) of Argon-41 released duringthe reporting period based on data from an appropriate monitoring system.3 The estimated maximum annual average concentrationof Argon-41in the unrestricted area (in microcuries/mi), the estimated corresponding annual radiation dose at this location (in millirem), and the fraction of theapplicable 10 CFR 20 limits for these values.4 The total estimated quantity of radioactivity in particulate form withhalf lives greater than eight days (in curies) released during thereporting period as determined by an appropriate particulate monitoring system.5l The average concentration of radioactive particulates with half-lives greater than eight days released (in microcuries/mI) during the reporting period. -(3) Solid Waste (summarized on an annual basis)(a) The total amount of solid waste packaged (in cubic feet).(b) The total activity in solid waste (in curies).(c) The dates of shipment and disposition (if shipped off site).g. An annual summary of the radiation exposure received by facility operations personnel, byfacility users. and-by visitors in terms of the average radiation exposure per individual and thegreatest exposure per individual in each group.h. An annual summary of the radiation levels and levels of contamination observed duringroutine surveys performed at the facility in terms of average and highest levels.i. An annual summary of any environmental surveys performed outside the facility. 6.7.2. Special Reports. Special reports are used to report unplanned events as well as plannedadministrative changes. The following classifications shall be used to determine the appropriate reporting schedule:

a. A report within 24 hours by telephone or similar conveyance to the NRC operations center of:(1) Any accidental release of radioactivity into unrestricted areas above applicable unrestricted area concentration limits, whether or not the release resulted in property* damage, personal injury, or exposure;

• (2) Any violation of a safety limit;(3) Operation with a limiting safety system setting less conservative than specified inSection 2.0, Limiting Safety System Settings;

.7 (4) Operation in violation of a Limiting Condition for Operation; (5) Failure of a required reactor or experiment safety system component which couldrender the system incapable of performing its intended safety function unless the failureis discovered during maintenance tests or a period of reactor shutdown; (6) Any unanticipated or uncontrolled change in reactivity greater than $1.00;(7) An observed inadequacy in the implementation of either administrative or procedural

controls, such that the inadequacy could have caused the existence or development of acondition which could have resulted in operation of the reactor outside the specified safety limits; and(8) A measurable release of fission products from a fuel element.b. A report within 14 days in writing to the NRC, Document Control Desk, Washington DC.(1) Those events reported as required by Technical Specifications 6.7.2.a.1 through6.7.2.a.8.

(2) The written report (and. to the extent possible, the preliminary telephone report orreport by similar conveyance) shall describe,

analyze, and evaluate safety implications.

and outline the corrective measures taken or planned to prevent re~occurrence of theevent.c. A report within 30 days in writing to the NRC, Document Control Desk, Washington DC.(1) Any significant variation of measured values from a corresponding predicted orpreviously measured value of safety-connected operating characteristics occurring during operation of the reactor;(2) Any significant change in the transient or accident analysis as described in theSafety Analysis Report (SAR);(3) A personnel change involving the positions of UCD/MNRC Director or UCO ViceChancellor for Research; and(4) Any observed inadequacies in the implementation of administrative or procedural controls such that the inadequacy causes or could have caused the existence ordevelopment of an unsafe condition with regard to reactor operations. 8.8 Records. Records may be in the form of logs, data sheets, or other suitable forms. The requiredinformation may be contained in single or multiple

records, or a combination thereof.

Records and logsshall be prepared for the following items and retained for a period of at least five years for items a.through f., and indefinitely for items g. through k. (Note: Annual reports, to the extent they contain all ofthe required information, may be used as records for items g. through j.)a. Normal reactor operation.

b. Principal maintenance activities.

c. Those events reported as required by Technical Specifications 6.7.1 and 6.7.2.d. Equipment and component surveillance activities required by the Technical Specifications.

,'e. Experiments performed with the reactor.f. Airborne and liquid radioactive effluents released to the environments and solid radioactive wasteshipped off site.

g. Offsite environmental monitoring surveys.i. h. Fuel inventories and transfers.

i. Facility radiation and contamination surveys.1. Radiation exposures for all personnel.

k. Updated, corrected, and as-built drawings of the facility.

.1J"39 a ~a ....-.a.'I.Formal Licensing ChannelAdministrative Reporting Channel............. Communications ChannelUCD/MNRC ORGANIZATION FOR LICENSING AND OPERATION FIGURE 6.1 UCDAVISMNRCMcCLELLAN NUCLEAR RESEARCH CENTERU.S. Nuclear Regulatory Commission Attn: Linh N. Tran, Senior Project Manager, NRRMail Stop: 012 D20One White Flint North11555 Rockville PikeRockville, MD 208525335 PRICE AVENUEBUILDING 258McCLELLAN, CA 95652PHONE: (916) 614-6200FAX: (916) 614-6250WEB: http://mn rc.ucdavis.edu October 29, 2015RE: NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUESTFROM THE UNIVERSITY OF CALIFORNIA-DAVIS McCLELLAN NUCLEAR RESEARCH CENTER PER THELETTER DATED JUNE 3, 2015.

Dear Ms. Tran,

In response to your letter dated June 3, 2015, we are submitting the requested documentation per saidletter under Oath and Affirmation. Additionally, we are provided said documentation electronically on a DVD for your convenience. I verify under penalty of perjury that the foregoing is true and correct.Executed on October 29, 2015.Assoca'e Director of Operations Reactor Supervisor McClellan Nuclear Research CenterUniversity of California-Davis Facility Operating License No. R-130.C: B. Klein, UCD/MNRC P:- oNUCLEAR REGULATORY COMMISSIONWASHINGTON, 0.0. zt&o5S5O1Q FACILITY OPERATING LICENSEDOCKET NO. 50-607DEPARTMENT OF THE AIR FORCE AT McCLELLAN AIR FORCE BASELicense No. R-1301. The U.S. Nuclear Regulatory Commission (the Commission) has found that:A." The application for license, filed by the Department of the Air Force atMcClellan Air Force Base, on October 23, 1 996, as supplemented, complies with the standards and requirements of the Atomic Energy Actof 1!954, as =amended (the Act), and the Commission's rules andregulations as set forth in 10 CFR Chapter I;B. Construction. of the facility was completed in substantial conformity withthe provision's of the Act, and the rules and regulations of theCommission; C. The facility Will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;

0. There is reasonable assurance (i) that the activities authorized by thislicense can be conducted

,without endangering the health and safety ofthe public and (ii) that such activities will be conducted in compliance with the Commission's regulations; E. The licensee is technically and financially qualified to engage in theactivities authorized by this operating license in accordance with theregulations of t/he Commission; ..F. The licensee is a Federal agency and will use the facility for defenseprograms and research. The licensee, in accordance with10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements," iis not required to furnish proof of financial protection. The licensee has executed an indemnity agreement that satisfies therequirements o~f 10 CFR Part 140 of the Commission's regulations; G. The issuance of this license will not be inimical to the common defenseand security or to the health and safety of the public;H. The issuance of this license is in accordance with 10 CFR Part 51 of theCommissioh's regulations and all applicable requirements have beensatisfied; andI.The receipt:, possession, and use of the byproduct and special nuclearmaterials as authorized by this license will be in accordance with theCommissioa 's regulations in 10 CFR Parts 30 and 70, including Sections30.33, 70.23, and 70.31.2. Facility Operating License No. R-130 is hereby issued to the Department ofthe Air Force at McClellan Air Force Base as follows:A. The license applies to the training reactor and isotopes production, General Atornics (TRIGA) nuclear reactor (the facility) owned by theDepartment 'of the Air Force at McClellan Air Force Base (the licensee). The facility is located on the licensee's site at McClellan Air Force Easeand is described in the licensee's application for license of October 23,1 996, as supplemented. B. Subject to the conditions and requirements incorporated herein, theCommission ~hereby licenses the Department of the Air Force atMcClellan Air Force Base:(1) Pursuant to Section 104c of the Act and 10 CFR Part 50,"Domestic Licensing of Production and Utilization Facilities," topossess, use, and operate the facility at the designated locationat McClellan Air Force Base, in accordance with the procedures and limitations set forth in this license.(2) Pursuant to the Act and 10 CFR Part 70, "Domestic Licensing of Special Nuclear Material," to receive,

possess, and use up to21 .0: kilograms of contained uranium-235 enriched to less than20 percent in the isotope uranium-235 in the form of reactorfuel;:up to 4 grams of contained uranium-235 of any enrichment in the form of fission chambers; up to 16.1 kilograms ofcontained uranium-235 enriched to less than 20 pecenR[[t in heisotope uranium-235 in the form of plates; and to possess, butnot separate, such special nuclear material as may be producedby the operation of the facility.

3(3) Pursuant to the Act and 1 0 CFR Part 30, "Rules of GeneralApplicability to Domestic Licensing of Byproduct Material," toreceive,

possess, and use a 4-curie sealed americium-beryllium neutron source in connection with operation of the facility; a55-millicurie sealed cesium-1 37 source for instrument calibrations; small instrument calibration and check sources ofless than 0.1 millicurie each; and to possess, use, but notseparate, except for byproduct material produced in reactorexperiments, such byproduct material as may be produced bythe operation of the facility.

C. This license shall be deemed to contain and is subject to the Conditions specified in lParts 20, 30, 50, 51, 55, 70, and 73 of 10 CFR Chapter I;to all applicable provisions of the Act; and to the rules, regulations, andorders of the Commission now or hereafter in effect and to the additional conditions specified below:(1) Maximum Power LevelThe licensee is authorized to operate the facility at steady-state power levels not in excess of 2300 kilowatts (thermal) and inthe pulse mode with reactivity insertions not to exceed $ 1.75(1.23 %Ak/k).(2) Technical Specifications The Technical Specifications contained in Appendix A arehereby incorporated in the license. The licensee shall operatethe !facility in accordance with the Technical Specifications. (3) Physical Security PlanThe licensee shall fully implement and maintain in effect aI.provisions of the Commission-approved physical security plan,including all amendments and revisions made pursuant to theauthority of 10 CFR 50.90 and 1 0 CFR 50.54(p). The approvedplan;i which is exempt from public disclosure pursuant to theprovisions of 10 CFR 2.790, is entitled "Physical Security Planfor the McClellan Nuclear Radiation Center (MNRC) TRIGAReacitor Facility," Revision 3, dated August 1996.

40. This license is effective as of the date of issuance and shall expiretwenty (20) years from its date of issuance.

~FOR THE NUCLEAR REGULATORY COMMISSION Office of Nuclear Reactor Regulation

Enclosure:

Appendix A Technical Specifications Date of Issuance: August 13, 1998 ~UNITED STATES "~NUCLEAR REGULATORY COMMISSION

• .WASHINGTON 1 D.C. 20555-0001" 'December 9, 1998Brigadier General Michael P. Wiedemer, Commander Sacramento Air Logistics CenterSM-ALC/TI-15335 Price AvenueMcClellan AFB, California 95652-2504

SUBJECT:

ISSUANCE OF AMENDMENT NO. 1 TO FACILITY OPERATING LICENSE NO. R-1 30 -DEPARTMENT OF THE AIR FORCE AT McCLELLAN AIR FORCE BASE (TAC NO. MA3477)

Dear General Wiedemer:

The Commission has issued the enclosed Amendment No. 1 to Facility Operating LicenseNo. R-1 30 for the McClellan Nuclear Radiation Center (MNRC) TRIGA Research Reactor.The amendment consists of changes to the technical specifications (TSs) in response toyour submittal of November 18, .1998.The amendment clarifies TS 3.8.3, requirements on the quantity and type of radioactive material allowed in experiments such that experiment failure will not result in airborneradioactivity in the reactor room or the unrestricted area exceeding the applicable doselimits in 10 CFR Part 20.A copy of the safety evaluation supporting Amendment No. 1 is also enclosed. Sincerely, Warren J. Eresian, Project ManagerNon-Power Reactors and Decommissioning

• Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No. 50-607

Enclosures:

1. Amendment No. 12. Safety Evaluation cc w/enclosures:

See next page McClellan AFB TRIGA REACTORDcktN.067 cc:Dr. Wade J. RichardsSM-ALCITI-1 5335 Price Avenue, Bldg. 258McClellan AFB, California 95652-2504 Col. Robert CapellHQ AFMC/SGC4225 Logistics Avenue, Suite 23Wright-Patterson AFB, Ohio 45433-5762 Lt. Col. Catherine ZeringueHQ AFSC/SEW9570 Avenue G, Building 24499Kirtland AFB, New Mexico 87117-5670 Test, Research, and TrainingReactor Newsletter 202 Nuclear Sciences CenterUniversity of FloridaGainesville, Florida 32611 . STATES~NUCLEAR REGULATORY COMMISSION ' WASHINGTON, D.C. 20888-0001 DEPARTMENT OF THE AIR FORCE ATMcCLELLAN AIR FORCE BASEDOCKET NO. 50-607AMENDMENT TO FACILITY OPERATING LICENSEAmendment No. 1License No. R-1301. The U.S. Nuclear Regulatory Commission (the Commission) has found thatA. The application for an amendment to Facility Operating License No. R-1 30 filed bythe Department of the Air Force at McClellan Air Force Base (the licensee) onNovember 1 8, 1 998, conforms to the standards and requirements of the AtomicEnergy Act of 1954, as amended (the Act), and the regulations of theCommission as stated in Chapter I of Title 10 of the Code of Federal Regulations (10 CFR);B. The facility will operate in conformity with the application, the provisions of theAct, and the rules and regulations of the Commission; C. There is reasonable assurance that (I) the activities authorized by this amendment can be conducted without endangering the health and safety of the public and (ii)such activities will be conducted in compliance with the regulations of theCommission; D. The issuance of this amendment will not be inimical to the common defense andsecurity or the health and safety of the public;E. This amendment is issued in accordance with the regulations of the Commission as stated in 10 CFR Part 51, and all applicable requirements have been satisfied; andF. Prior notice of this amendment was not required by 10 CFR 2.105, andpublication of notice for this amendment is not required by 10 CFR 2.106.

22. Accordingly, the license is amended by changes to the Technical Specifications asindicated in the enclosure to this license amendment, and paragraph 2.C.(ii) of FacilityOperating License No. R-1 30 is hereby amended to read as follows:2.C.(ii)

Technical Specifications The Technical Specifications contained in Appendix A, as revised throughAmendment No. 1, are hereby incorporated in the license. The licensee shalloperate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of issuance.

FOR TH.E NUCLEAR REGULATORY COMMISSION Seymour H. Weiss, DirectorNon-Power Reactors and Decommissioning Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation

Enclosure:

Appendix A, Technical Specifications ChangesDate of Issuance: ENCLOSURE TO LICENSE AMENDMENT NO. 1FACILITY OPERATING LICENSE NO? R-1 30DOCKET NO. 50-607Replace the following pages of Appendix A, "Technical Specifications," with the enclosedpages. The revised pages are identified by amendment number and contain vertical linesindicating the areas of change.Remove Insert24 2425 25 .j° ,c. This specification is intended to prevent damage to vital equipment byrestricting the quantity of explosive materials within the 'r~actor tank (SAR Chapter 13,Section 13..2.6.2). .-d. The failure of an experiment involving the irradiation of 3 lbs TNTequivalent or less in any radiography bay external to the reactor tank will not result indamage to the reactor controls or the reactor tank. Safety Analyses have beenperformed (SAR Chapter 13, Section 13.2.6.2) which show that up to six (6) lbs of TNTequivalent can be safely irradiated in any radiography bay. Therefore, the three (3) lblimit gives a safety margin of two (2).3.8.3 Failure and Applicability. This specification applies to experiments installed in thereactor, in-tank experiment facilities, and radiography bays.Objective. The objective is to prevent damage to the reactor or significant releases of radioactive materials in the event of an experiment failure.Specifications.

a. All experiment materials which could off-gas,

sublime, volatilize, or produceaerosols under (1) normal operating conditions of the experiment or reactor, (2) credibleaccident conditions in the reactor, or (3) where the possibility exists that the failure ofan experiment could release radioactive gases or aerosols into the reactor building orinto the unrestricted area, the quantity and type of material in the experiment shall belimited such that the airborne radioactivity in the reactor room or the unrestricted areawill not result in exceeding the applicable dose limits in 10 CFR Part 20, assuming100% of the gases or aerosols escape.b. In calculations pursualtt to (a) above, the following assumptions shall beused:(1) If the effluent from an experiment facility exhausts through a stackwhich is closed on high radiation levels, at least 10% of the gaseous activity or aerosolsproduced will escape.(2) If the effluent from an 'experiment facility exhausts through a filterinstallation designed for greater than 99% efficiency for 0.3 micron and larger particles, at least 10% of these will escape.(3) For materials whose boiling point is above 130°F and where vaporsformed by boiling this material can escape only through an undistributed column ofwater above the core, at least 10% of these vapors 'can escape.24

c. If a capsule fails and releases material which could damage the reactorfuel or structure by corrosion or other means, an evaluation shall be made to.determine the need for corrective action. Insipection and any corrective action takenshall be reviewed by the Facility Director or his designated alternate and determined tobe satisfactory before operation of the reactor is resumed.Basis.a. This specification is intended to reduce the likelihood that airborneradioactivity in the reactor room or the unrestricted area will result in exceeding theapplicable dose limits in 10 CFR Part 20.b. These assumptions are used to evaluate the potential airborneradioactivity release due to an experiment failure (SAR Chapter 13, Section 13.2.6.2).

c. Normal operation of the reactor with damaged reactor fuel orstructural damage is prohibited to avoid release of fission products.

Potential damageto reactor fuel or structure must be brought to the attention of the Facility Director orhis designated alternate for review to assure safe operation of the reactor (SAR Chapter13, Section 13.2.6.2). 4.0 Surveillance Requ~irements: General. The surveillance frequencies denoted herein are based on continuing operation of the reactor. Surveillance activities scheduled to occur during an operating cycle which can not be performed with the re'actor operating may be deferred to the endof the operating cycle. If the reactor is not operated for a reasonable .time, a r'eactorsystem or measuring channel surveillance requirement may be waived during theassociated time period. Prior to reactor system or measuring channel operation, thesurveillance shall be performed for each reactor system or measuring channel for whichsurveillance was waived. A reactor system or measuring channel shall not beconsidered operable until it is successfully tested.4.1 Reactor Core Parameters. 4.1.1 Steady State Operation. Applicability. This specification applies to the surveillance requirement for the power level monitoring channels. Objective. The objective is to verify that the maximum power level of thereactor does not exceed the authorized limit.25

STATES~NUCLEAR REGULATORY COMMISSION ' WASHINGTON," O.C. 20865-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 1 TOFACILITY OPERATING LICENSE NO. R-130DEPARTMENT OF THE AIR FORCE ATMcCLELLAN AIR FORCE BASEDOCKET NO..POO-607

1.0 INTRODUCTION

By letter dated November 18, 1 998, the Department of the Air Force at McClellan AirForce Base (the licensee) submitted a request for amendment of the Technical Specifications (TSs), Appendix A, to Facility Operating License No. R-1 30 for the McClellan Nuclear Radiation Center TRIGA Research Reactor (MNRC). The requested amendment would clarify the quantity and type of material in experiments that could be released in theunlikely event of an experiment failure.2.0 EVALUATION The licensee has requested amendment of TS 3.8.3 concerning limitations on experiments. TS 3.8.3 and the bases of the TS currently read:Aoplicability. This specification applies to experiments installed in the reactor andits experimental facilities. Specifications.

a. All experiment materials which~could off-gas,

sublime, volatilize, orproduce aerosols under (1) normal operating conditions of theexperiment or reactor, (2) credible accident conditions in the reactor, or(3) where the possibility exists that the failure of an experiment couldrelease radioactive gases or aerosols into the reactor building or into theunrestricted area, the quantity and type of material to be irradiated shallbe limited such that the airborne concentration of radioactivity shall notexceed the applicable limits of 10 CFR Part 20 (at the operations boundary),

assuming 100% of the gases or aerosols escape."h. O °" 02Bases.a. This specification is intended to reduce the likelihood that airborneradioactivity in excess of the limits of 10 CFR Part 20 shall be releasedinto the reactor building or to the unrestricted area (SAR Section13.2.6.2). The licensee has proposed that the TS and bases be amended to read:Applicability. This specification applies to experiments installed in the reactor, in-tank experiment facilities, and radiography bays.Specifications.

a. All experiment materials which could off-gas,

sublime, volatilize, orproduce aerosols under (1) normal operating conditions of theexperiment or reactor, (2) credible accident conditions in the reactor, or(3) where the possibility exists that the failure of an experiment couldrelease radioactive gases or aerosols into the reactor building or into theunrestricted area, the quantity and type of material in the experiment shall be limited such that the airborne radioactivity in the reactor room orthe unrestricted area will not result in exceeding the applicable doselimits in 10 CFR Part 20, assuming 100% of the gases or aerosolsescape.Bases.a. This specification is intended to reduce the likelihood that airborneradioactivity in the reactor room or the unrestricted area will result inexceeding the applicable dose limits on 10 CFR 20.The licensee has proposed clarifying the TS by basing the TS on dose instead ofconcentrations of radioactive material.

The purpose of this TS is to limit doses to membersof the public and the MNRC staff to 10 CFR Part 20 limits in the unlikely event that an*experiment were to fail and release airborne radioactive material into the reactorconfinement and subsequently to the environment. Doses to members of the reactor staffand members of the public from accidents at research reactors are limited to the dosesgiven in 10 CFR Part 20 because 10 CFR Part 100 is not applicable to research reactors. The current TS is based on radioac~tivity concentrations. For occupational exposures theannual limit on intake (ALl) is the annual intake which would result in either a committed effective dose equivalent of 5 reins (stochastic ALl) or a committed dose equivalent of 50reins to an organ or tissue (non-stochastic ALl). The derived, air concentration (DAC)values in Table 1 of Appendix B to 10 CFR Part 20 are based on dividing the ALl by 2000working hours per year and is intended to control chronic occupational exposures. For non-occupational exposure (members of the public) the effluent concentrations given in Table 2 3.of Appendix B to 10 CFR Part 20 are equivalent to the radionuclide concentration which ifinhaled continually over the course of a year would produce a total effective doseequivalent of 0.05 rem. The licensee's proposed wording would be based on dose limitsdirectly. The licensee is concerned that the TS as currently written could be interpreted to limitreleases to the instantaneous concentration of airborne radioactive material in the reactorroom and unrestricted areas. This would ignore the time integral aspects of theconcentration limits given in 10 CFR Part 20 as discussed above. For a particular experiment failure event, it is possible to exceed the concentration limits in 10 CFR Part 20while the resulting dose would be a small fraction of the dose limits.The NRC staff notes that the proposed wording of the TS is more encompassing because aTS based on dose would also include consideration of radiation shine from a cloud ofradioactive material. This proposed change to the TSs is acceptable to the staff becausethe dose to members of the reactor staff and members of the public from the accidental failure of experiments will be within the limits given in 10 CFR Part 20 and because the*proposed wording clarifies the TS.3.0 ENVIRONMENTAL CONSIDERATION This amendment involves changes in the installation or use of a facility component locatedwithin the restricted area as defined in 10 CFR Part 20 or changes in inspection andsurveillance requirements. The staff has determined that this amendment involves nosignificant increase in the amounts, and no significant change in the types, of any effluents that may be released off site, and no significant increase in individual or cumulative occupational radiation exposure. Accordingly, this amendment meets the eligibility criteriafor categorical exclusioni set forth in 10 CFR 51.22(c)(9). Pursuant to 1OCFR 51.22(b), noenvironmental impact statement or environmental assessment need be prepared inconnection with the issuance of this amendment.

4.0 CONCLUSION

The staff has concluded, on the basis of the considerations discussed above, that(1) because the amendment does not involve a significant increase in the probability orconsequences of accidents previously evaluated,* or create the possibility of a new ordifferent kind of accident from any accident previously evaluated, and does not involve asignificant reduction in a margin of safety, the amendment does not involve a significant hazards consideration; (2) there is reasonable assurance that the health and safety of thepublic will not be endangered by the proposed activities; and (3) such activities will beconducted in compliance with the Commission's regulations and the issuance of thisamendment will not be inimical to the common defense and security or the health andsafety of the public.Principal Contributor: Warren J. EresianDate: December 9, 1998

STATESo NUCLEAR REGULATORY COMMISSION Z r~oWASHINGTON, D.C. 2055.5-0001 FACILI;TY OPERATING LICENSEDOCKET NO. 50-607DEPARTMENT OF THE AIR FO.RCE.AT McCLELLAN AIR FORCE BASELicense No. R-1 301. The U.S. Nuclear Regulatory Commission (the Commission) has found that:A. The application for license, filed by the Department of the Air Force atMcClellan Air Force Base, on October 23, 1 996, as supplemented, complies with the standards and requirements of the Atomic Energy Actof 1 954, as amended (the Act), and the Commission's rules andregulations as set forth in 10 CFR 'Chapter I;B. Construction of the facility was completed in substantial conformity withthe provisions of the Act, and the rules and regulations of theCommission; C. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; D. There is reasonable assurance (i) that the activities authorized by thislicense can be conducted without' endangering the health and safety ofthe public and (ii) that such activities will be conducted in compliance with the Commission's regulations; E. The licensee is technically and financially qualified to engage in theactivities authorized by this operating license in accordance with theregulations of the Commission;... F. The licensee is a Federal agency and will use the facility for defenseprograms and research. The licensee, in accordance with10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements," is not required to furnish proof of financial protection. The licensee has executed an indemnity agreement that satisfies therequirements of 10 CFR Part 140 of the Commission's regulations; 2G. The issuance of this license will not be inimical to the common defenseand security or to the health and safety of the public;H. The issuance of this license is in accordance with 10 CFR Part 51 of theCommission's regulations and all applicable requirements have beensatisfied; andc]p,1°o s s e-ssio n,-a +n 2. Facility Operating License No. R-130 is hereby issued to the Department ofthe Air Force at McClellan Air Force Base as follows:A. The license applies to the training reactor and isotopes production, General Atomics (TRIGA) nuclear reactor (the facility) owned by theDepartment of the Air Force at McClellan Air Force Base (the licensee). The facility is located on the licensee's site at McClellan Air Force Baseand is described in the licensee's application for license of October 23,1 996, as supplemented. B. Subject to the conditions and requirements incorporated herein, theCommission hereby licenses the Department of the Air Force atMcClellan Air Force Base:(1) Pursuant to Section 1 04c of the Act and 10 CFR Part 50,"Domestic Licensing of Production and Utilization Facilities," topossess, use, and operate the facility at the designated locationat McClellan Air Force Base, in accordance with the procedures and limitations set forth in this license.(2) Pursuant to the Act and 10 CFR Part 70, "Domestic Licensing of Special Nuclear Material," to receive,

possess, and use up to21 .0 kilograms of contained uranium-235 enriched to less than20 percent in the isotope uranium-235 in the form of reactorfuel; up to 4 grams of contained uranium-235 of any enrichment in the form of fission chambers; up to 16.1 kilograms ofcontained Uranium-235 enriched to less than 20 percent in theisotope uranium-235 in the form of plates; and. to possess, butnot separate, such special nuclear material as may be producedby the operation of the facility.

3 4-curie sealed americium-beryllium neutron source in connection with operation of the facility; a55-millicurie sealed cesium-i137 source for instrument calibrations; small instrument calibration and check 'sources ofless than 0.1 millicurie C. This license shall be deemed to contain and is subject to the conditions specified in Parts 20, 30, 50, 51, 55, 70, and 73 of 10 CFR Chapter I;to all applicable provisions of the Act; and to the rules, regulations, andorders of the Commission now or hereafter in effect and to.the additional conditions specified below:(1) Maximum Power LevelThe licensee is authorized to operate the facility at steady-state power levels not in excess of 2300 kilowatts (thermal) and inthe pulse mode with reactivity insertions not to exceed $1.75.(1.23 %Ak/k).(2) Technical Specifications The Technical Specifications contained in Appendix A arehereby incorporated in the license. The licensee shall operatethe facility in accordance with the Technical Specifications. (3) Physical Security PlanThe licensee shall fully implement and maintain in effectel.- provisions of the Commission-approved physical security plan,including all amendments and revisions made pursuant to theauthority of 10 CFR 50.90 and 10 CFR 50.54(p). The approvedplan, which is exempt from public disclosure pursuant to theprovisions of 1 0 CFR 2.790, is entitled "Physical Security Planfor the McClellan Nuclear Radiation Center (MNRC) TRIGAReactor Facility," Revision 3, dated August 1 996. S ..* .:..40. This license is effective as of the date of issuance and shall expiretwenty (20) years from its date of issuance. FOR THE NUCLEAR REGULATORY COMMISSION Office of Nuclear Reactor Regulation

Enclosure:

Appendix A Technical Specifications Date of Issuance: August 13, 1998 ~UNITEDOSTATES SNUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055-.0001 Mrh1, 1999Brigadier General Michael P. Wiedemer, Commander Sacramento Air Logistics CenterSM-ALC/TI-15335 Price AvenueMcClellan AFB, California 95652-2504

SUBJECT:

ISSUANCE OF AMENDMENT NO. 2 TO AMENDED FACILITY OPERATING LICENSENO. R-130 -DEPARTMENT OF THE AIR FORCE AT McCLELLAN AIRFORCE BASE (TAC NO. MA3477)

Dear General Wiedemer:

The Commission has issued enclosed Amendment No. 2 to Facility Operating LicenseNo. R-1 30 for the McClellan Nuclear Radiation Center (MN RC) TRIGA Research Reactor.The amendment consists of changes to the Technical Specifications (TSs) and SafetyAnalysis Report (SAR) to support expanded experimental facilities in response to yoursubmittal of January 11, 1999.The amendment provides for the installation of an Argon-41 Production Facility and aCentral Irradiation Facility. The installation of the Argon-41 Production Facility does notrequire any change or expansion of the TSs since an experiment failure will not result inairborne radioactivity in the reactor room or the unrestricted area exceeding the applicable dose limits already prescribed. The installation of the Central Irradiation Facility requires achange to TS 3.8.1 with regard to the maximum reactivity worth of a moveableexperiment. The change increases the reactivity limit of a moveable experiment in theCentral irradiation Facility to $1.75, corresponding to the pulse limit specified in TS 3.1.2.A copy of the safety evaluation supporting Amendment No. 2 is also enclosed. Si lcerely,Warren J. Iresian, Project ManagerNon-Power Reactors and Decommissioning Project Directorate Division of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation Docket No. 50-607

Enclosures:

1. Amendment No. 22. Safety Evaluation cc w/enclosures:

See next page McClellan AFB TRIGA REACTORDoktN.5-0 cc"Dr. Wade J. RichardsSM-ALC/TI-1 5335 Price Avenue, Bldg. 258McClellan AFB, California 95652-2504 Lt. Col. Marcia ThorntonHQ AFSC/SEW" 9570 Avenue G., Bldg. 24499Kirtland AFB, New Mexico 87117-5670 Col. Robert CapellHQ AFMC/SGC4225 Logistics Avenue, Suite 23Wright-Patterson AFB, Ohio 45433-5762 0* 0UNITED STATES.NucLEAR REGULATORY COMMISSIoN WHNToND.C. 208-o000DEPARTMENT OF THE AIR FORCE ATMc.CLELLAN AIR FORCE BASEDOCKET NO. 50-607AMENDMENT TO FACILITY OPERATING LICENSEAmendmentNo. 2License No. R-1 301.. The U.S. Nuclear Regulatory Commission (the Commission) has found thatA. The application for an amendment to Facility Operating License No. R-1 30 filedby the Department of the Air Force at McClellan Air Force Base (the licensee) onJanuary 11, 1999, conforms to the standards and requirements of the AtomicEnergy Act of 1954, as amended (the Act), and the regulations of theCommission as stated in Chapter I of Title 10 of the Code of Federal Regulations (10 CFR);B. The facility will operate in conformity with the application, the provisions of theAct, and the rules and regulations of the Commission;

• C. There is reasonable assurance that (i) the activities authorized by this amendmentc can be conducted without endangering the health and safety of the public and(ii) such activities will be conducted in compliance with the regulations of theCommission; D. The issuance of this amendment will not be inimical to the common defense andsecurity or to the health and safety of the public;E. This amendment is issued in accordance with the regulations of the Commission as stated in 10 CFR Part 51, and all applicable requirements have been satisfied; andF. Prior notice of this amendment was not required by 10 CFR 2.105, andpublication of notice for this amendment is not required by 10 CFR 2.106. 2. Accordingly, the license is amended by changes to the Safety Analysis Report andTechnical Specifications as indicated in the enclosure to this license amendment, andparagraph 2.C.(ii) of Facility Operating License No. R-130 is hereby amended to readas follows:2.C.(ii)

Technical Specifications The Technical Specifications contained in Appendix A, as revised throughAmendment No. 2, are hereby incorporated in the license. The licenseeshall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION ",i /1f Lt 'Seymour H. Weiss, DirectorNon-Power Reactors and Decommissioning Project Directorate Division of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation

Enclosure:

• Appendix A, Technical Specifications
• and Safety Analysis Report ChangesDate of Issuance:

March 1, 1999 .... 0 .ENCLOSURE TO LICENSE AMENDMENT NO. 2FACILITY OPERATING LICENSE NO. R-130DOCKET NO. 50-607A. Replace the following page of Appendix A, "Technical Specifications," with theenclosed page. The revised page is identified by amendment number and containsvertical lines indicating the areas of change.Remove Insert22 22B. Insert the following sections into the Safety Analysis Report.1. Add new Section 10.5.32. Add new Section 11.1.1.1.6

3. Append to Section 13.2.6.24. Add new Appendix A to Chapter 135. Add new Appendix

.B to Chapter 136. Change Section 10.4.17. Add new Section 10.4.1.48. Append to Section 1 3.2.6.29. Add Reference 13.19 to ChaPter 13

• Sunrestricted area.3.8 Experiments 3.8.1 Reactivity Limits.Applicability.

This specification applies to the reactivity limits on experiments installed in the reactor and in-tank experiment facilities. Obiective. The objective is tQ assure control of the reactor during the irradiation or handling of experiments adjacent to or in the reactor core.Specification. The reactor shall not be operated unless the following conditions governing experiments exist:a. The absolute reactivity worth of any moveable experiment in the CentralIrradiation Facility shall be less than $1.75 (1.23% AK/K); the absolute reactivity worth of anymoveable experiment not in the Central Irradiation Facility shall be less than one (1) dollar(0.7% AK/K).b. The absolute reactivity worth of any single secured experiment shall be lessthan the maximum allowed pulse ($1.75) (1.23% AK/K).c. The absolute total reactivity worth of experiments installed in the reactor andin-tank shall not exceed an absolute value of one dollar and ninety-two cents ($1.92) (1.34%AK/K), including the potential reactivity which might result from malfunction,

flooding, voiding, orremoval and insertion of the experiment.

Basis.*a. A reactivity limit of less than $1.75 specifically for the Central Irradiation Facility is based on the pulsing reactivity insertion limit (section 3.1.2) and on the design of thesample can assembly which allows insertion and withdrawal of experiments in a controlled manner (identical in form, fit, and function to a control rod). See also the MNRC SAR, Chapter13.2.2.2.1 for the maximum reactivity insertion discussion. A reactivity limit of less than one (1)dollar on a single moveable experiment not in the Central Irradiation Facility will precludepulsing if the experiment's fixturing should fail, since the resulting reactivity insertion would notcause prompt criticality if less than one dollar. Given that the reactor will not pulseinadvertently, the additional increase in transient power and temperature will be slow enough sothat the fuel temperature scram will be effective.

b. The absolute worst event which may be considered in conjunction with asingle secured experiment is its sudden accidental or unplanned removal while the reactor isoperating.

This would result in a reactivity increase less than a pulse of $1.92, analyzed in SARChapter 13, Section 13.2.2.2.1.

c. It is conservatively assumed that simultaneous removal of all experiments inthe reactor and in-tank experiment facilities at any given time shall not exceed the maximumreactivity insertion limit. SAR, Chapter .13, Section 13.2.2.2.1 shows that an insertion 22Amendmient No. 2 ADDITION TO MCCLELLAN NUCLEAR RADIATION CENTERSAFETY ANALYSIS REPORT -ARGON-41 PRODUCTION FACILITYNEW SECTION 10,5.310.5.3 Arcqon-41 Production FacilityThe Argon-41 Production Facility will produce 1-2 curies of 41Ar for research andcommercial use. The 41Ar will be produced by introducing argon gas into a stainless steelcontainer located in one of the silicon irradiation positions (adjacent to the graphitereflector and external to the reactor core -Figure 10.11 1A). All the components containing activated 41Ar are located in the reactor room.Argon gas from a commercial argon gas cylinder will supply the irradiation container.

After the irradiation container is pressurized (approximately 500 psig) to the desired level,the gas cylinder will be isolated from the irradiation container. To produce the desiredactivity level of 41Ar the sample will be irradiated for approximately 24 hours.After irradiation, liquid nitrogen is added to a Dewar. A remotely operated solenoid valveis opened to pressurize the cooling coils above the liquid nitrogen bath. The Dewar is thenraised to cover the cooling coils and 41Ar is cryogenically extricated from the irradiation container. After extrication is completed, the solenoid valve from the irradiation container is shut and another remotely operated solenoid valve is opened. This allows diffusion of41Ar gas to the sample container. The liquid nitrogen Dewar is lowered, exposing thecooling coils to room temperature. When that portion of the system between the coolingcoils and the sample container has reached equilibrium the sample container will beisolated and..removed from the room. The coil is surrounded with a lead shield to minimizethe radiation exposure to personnel. A catch tank surrounds the Dewar to contain any liquid nitrogen escaping from'the Dewaror in the unlikely event of a total failure of the Dewar.Over pressure protection of the overall system is provided by several relief valves thatvent to an over pressure tank. The over pressure ta~nk is protected by its own relief valvewhich vents to the reactor room. The tank is located as high as possible in the reactorroom.All piping and valves in the system are stainless steel. Compression fittings or double-ended shut-off quick connectors are used for allconnections normally in contact with the41Ar.The Argon-41 Production Facility consists of several different components, with the majorcomponents listed below. 0COMPONENT Irradiation Container Over PressureRelief ValvesOver PressureRelief TankMATERIAl304 stainless steel304 stainless steelCarbon steel304 stainless steel304 stainless steelDESCRIPTION The irradiation container is a 1000 mlsample cylinder with a working pressureof 1 800 psig and a burst pressure of6000 psig. It conforms to the "Shipping Container Specifications" from the U.S.Code of Federal Regulations, Title 49 orBureau of Explosives Tariff No.BOE6000.The adjustable proportional pressure reliefvalves have a working pressure up to6000 psig. When upstream pressureovercomes the force exerted by thespring, the poppet opens, allowing flowthrough the valve. As the upstreampressure increases, flow through thevalve increases proportionately. Crackingpressure is only sensitive to inlet pressureand is not affected by outlet pressure. 30 gallon tank.ValvesTubingBellows sealed valves.1/4-inch and Y/=-inch. NEW SECTION 11.1.1.1.6 11.1.1.1.6 Araqon-41 from the Argon-41 Production FacilityAr-41 will be produced by the Ar-41 Production Facility (see Chapter 10) as needed. TheAr-41 that is produced by the Ar-41 Argon Production Facility will be contained in thesystem so there should be no increase in the Ar-41 levels in the reactor room or the Ar-41that is released to the unrestricted area. Catastrophic failure of the system will not resultin any 10 CFR 20 limit being exceeded and is further discussed in Chapter 13.APPEND TO SECTION 13.2.6.2The Argon-41 Production Facility (see Chapter 10) can produce argon-41 in excess of theamounts analyzed in Appendix A of the MNRC Safety Analysis Report. However, if thesystem releases argon-41, the gas will be contained in the reactor room and the existing 0reactor room ventilation system will be used in recirculation mode to prevent releasing argon-41 to the environment, recirculating the gas until it decays. The existing StackContinuous Air Monitor will also be used to verify any release outside the MNRCboundary. If the system had a catastrophic failure and 4 curies of argon-41 were released to thevolume of the reactor room, the argon-41 concentration in the reactor room would beapproximately 22 R/hr (based on a semi-infinite cloud; see calculation in Chapter 1 3,Appendix A). Personnel would be evacuated from the reactor room and access would berestricted. The reactor room ventilation system (as described in Chapter 9) would, beoperated in the recirculation mode for approximately one day before the dose rate fromargon-41 decays to less than 1 mR/hr. Therefore, the argon-41 Discharge Limit defined inthe MNRC Technical Specifications will not be exceeded due to the recirculation mode ofthe reactor room ventilation system.Other potential accidents include failure of the irradiation container due tooverpressurization from the argon gas supply cylinder, since a new argon supply cylinderis typically delivered at 2200 psig and the container is rated for 1800 psig. However, thisrequires multiple failures and is considered non-credible: a) the operator would have toviolate an operational procedure; b) the regulator would have to fail, and c) at the sametime the pressure relief valve would have to fail. Also, liquid nitrogen could spill into thereactor tank, causing expansion of the water and expelling a portion of tank water. Toprevent this, a catch basin surrounds the Cold Trap, and the liquid nitrogen is suppliedthrough a pipe in the reactor room wall connecting the trap to a supply container in theequipment room. A third accident could result if the pressure relief valve became chokedwith supersonic flow; however, the flow rates are estimated to be less than sonic (seecalculat~ion in Chapter 13, Appendix A).NEW APPENDIX A TO CHAPTER 13ARGON-41 CONCENTRATION IN REACTOR ROOMGIVEN:1. Reactor room volume =-7.39x10 3 ft3 tReference 1112. 4 curies Ar-41 in argon production system3. D(y)=,2 = O.25Evx [Reference 21Dy= = gamma dose rate from a semi-infinite cloud (rad/sec) Ev = average gamma energy per disintegration (Mev/dis) = 1 .2936 Mev/dis for Ar-41[Rfrne3 0CALCULATIO X)N:*= concentration of gamma emitting isotope in the cloud (Ci/m3)X = (4Ci)/[7.39xl10 3ft3)(1 m3/35.314 ft3) = 1 .91!x 0.2 Gi/m3D(y)=,2 = 0. 25Eyx= (0.25)( 1.2936 Mev/dis)(1 .91 xl 0.2 Cl/rn3)= (0.0062 rads/sec)(3600 seclhr)= 22.24 radslhrD = Doe~xt = -(1/A)In(0D/D) = -(T112Iln2)ln(D/D 0)For 0 = 1 mrad/hrt = -(1 .8hr/In2)ln(1/22,240) = 26 hr

REFERENCES:

1. MNRC Safety Analysis Report, Figure 9.1.1.2. The Health Physics and Radiological Health Handbook (Revised Edition),

editedby Shelein,

p. 4393. Nuclides and Isotopes, 14m' edition, Chart of the Nuclides, GE Nuclear Energy,p. .22 0, .NEW APPENDIX B TO CHAPTER 13*SONIC FLOW FOR ARGON-41 PROJECTAssume: Perfect GasConstants:

Property Value UnitsR 208 N-rn/k g-degKk(c,/c,) 1 .67 dimensionless Problem: determine if the pr~essure relief valve will experience choking due to supersonic flow.Solution: First, calculate the speed of sound in argon at 40 degrees C and -200 degrees C:given c =speed of sound in a medium = (kRTgc)fl c = [1.67(208 N-m/kg-degK)(40+273)K(1 kg-rn/N-sec 2 )]P= 329.7327 rn/s at 40 degrees Cc =[1 .67(208 N-m/kg-degK)(-200 +273)K( 1 kg-rn/N-sec 2 = 159.2397 rn/s at -200 degrees CNext, calculate the velocity of the argon in the tubing at the pressure relief valve:given volumetric flow rate V = (velocity)(area) From tech data on valve, assume V = lft3/min, based on air and relief at 1125 psiV = (1 ft3/min)(12 in/ft)3(2.54 cm/in)3(1 min/60 sec)= 471.9474 cm3/secArea = 2 = 3.14(0.18in/2) 2 = 0.025434 in2 based on 1/4 inch tubing= 0.16409 cm2Velocity = V/Area = 28.7615 rn/secMach Number = Velocity/c = 0.180618 at -200 degrees C= 0.087227 at 40 degrees C

== Conclusion:==

Gas velocity at the relief valve is less than the speed of sound in argon andtherefore should not experience choking at the valve.

Reference:

Fundamentals of Gas Dynamics, Zucker, pp.89, 130-1 33, 375. ADDITION TO MCCLELLAN NUCLEAR RADIATION CENTER SAFETY ANALYSIS REPORTAND TECHNICAL SPECIFICATIONS -CENTRAL IRRADIATION FACILITYCHANGE SECTION 10.4.1The Central Irradiation

Facility, located in the center of the reactor core, may containeither a plug assembly (as described in sections 10.4.1 .1 through 10.4.1 .3 and Figure10.7) or a moveable sample can system (as described in section 10.4.1.4).

All parts areremovable from the reactor using underwater tools.NEW SECTION 10.4.1.410.4.1 .4 Central Irradiation FacilityThe central irradiation facility allows samples to be inserted into the reactor core (i.e.central facility) while operating the reactor at power. The reactor operator controls theinsertion and removal of samples from the central facility through the use of a drivemechanism similar to the control rods.The central thimble is approximately 52 inches in length and 4.22 inches outer diameterwith an inside dimension of approximately 4.0 inches. The central thimble, once in place,passes through the upper grid plate, the lower grid plate and the safety plate. Aluminumshims have been added to the outer periphery of the central thimble in the fuel region.These shims align the central thimble and displace the water from the scallops of the fuelelement locations in the B hex ring 4.25-inch hole. Two captive bolts attach the centralthimble to the upper grid plate. These bolts prevent the accidental removal of the facilitywhen removing samples from the central thimble.An 1100 aluminum slug located inside the central thimble is normally positioned in thereactor core. The aluminum slug is 4 inches in diameter and 24.75 inches in length. Thisvoids the water from the central thimble when the sample can is removed from thethimble.An orifice plate is located on the bottom of the central thimble. In the event the aluminumslug releases from the locating holes and falls to the, bottom of the central thimble, therate of decent will be less than the normal control rod drive speed.The sample can is approximately 30.5 inches long with an outside diameter of 3.99inches and an inside diameter of 3.75 inches. The can could be free flooding or dry, andis used to position samples for irradiation in the reactor core. The positioning of samplescan be accomplished during full power reactor operations (i.e. 2 MW). During insertion into the reactor core and while in the reactor core the assembly has the capability of beingrotated.The drive mechaauism has the same type of drive motor as the control rod drives exceptthe model selected will have more torque. All other aspects of the motor and controller are identical. There are two sets of controls, one in the reactor room and the other in the control room.Normal operational control is from the reactor console where the reactor operators wiBltreat the insertion and removal of the samples as if they were control rods. The reactorroom controls can only be enabled from the reactor console. The normal indicators are asfollows:"A. Power On, switch and indicator (control room only).B. Reactor Room control enable switch and indicator (control room only).C. One set of momentary UP/DOWN switches for 1/22 speed drive.D. One set of momentary UP/DOWN switches for full speed drive.E. Indicators for UP, DOWN, and CLOSE TO DOWN positions. F. Digital indication of the sample can position, scaled 0-1000 units.G. Rotation ON, switch and indicator. Limit switches on the rack are used in the logic design to determine end of travelindications, stop driving limits and start/stop rotation of the carrier.APPEND TO SECTION 13.2.6.2Another potential accident involves the Central Irradiation Facility (see Chapter 10) since itmay be considered similar to a control rod. Therefore, consider three potential scenarios for an uncontrolled reactivity insertion analogous to the Uncontrolled Withdrawal Of aControl Rod (see Section 13.2.2.2.2). First, if the material in the sample can were ofsufficiently different worth than the aluminum

cylinder, the sample can would causereactivity changes in the same fashion as a control rod, and either operator error ormechanical failure could cause an uncontrolled reactivity insertion.

Second, if thealuminum cylinder failed to engage upon the sample can's insertion, a water void wouldbe created in the central facility as the aluminum cylinder descended ahead of the samplecan. Similarly, if the aluminum cylinder failed to replace the can upon removal from thecentral facility a water void would result.All three of the above scenarios can be bounded by the Uncontrolled Withdrawal of aControl Rod analysis (Section 13.2.2.2.2). Specifically, the Central Irradiation Facilitymust have less reactivity and must drive slower than the control rod analyzed ($3.50 and42 inches/minute, respectively). To that end, the reactivity of any material in the samplecan shall be measured at low power to verify it's worth is not only less than $3.50 butalso less than $1.75, the reactivity limit for the Central Irradiation Facility (based on theTechnical Specification limit of $1 .75 for the pulsed reactivity insertion). For example, theworth of a silicon ingot in the previous 1 MW in-core experiment facility was measured at$0.73 positive (vs. water, reference exp. # 96-01, 1/30/96, reactor run #2411.). Theworth of an aluminum cylinder vs. void and vs. water has been analyzed by computersimulation (Reference 13.19). The most positive reactivity effect in the computersimulation is from Case 3 to Case 9, where the voided sample can is lowered 18 inches,resulting in an increase of about $0.06. The most negative reactivity effect is from Case3 to Case 1 2, where in an accident the sample can not only floods but also the aluminumcylinder drops, resulting in a decrease of about $.1.76. Thus, the worth of the sample canor the aluminum cylinder vs. water is less than $3.50, and also less than the most reactive control rod (for example, a typical regulating rod worth is $2.57, measured 6/98).With respect to the drive mechanism, the maximum drive speed is identical to the rodspeed analyzed in the MNRC SAR (Section 13.2.2.2.2). Furthermore, in the event offailure of the aluminum cylinder to engage upon installation of the sample can, the base ofthe Central Thimble is designed (by sizing the hole in the base) to allow the aluminumcylinder to descend at no more than the analyzed 42 inches/minute. Therefore, theaccident analysis in Chapter 13 of the MNRC SAR for Uncontrolled Withdrawal of aControl Rod (Section 13.2.2.2.2) is sufficient to bound any accident associated with theCentral Irradiation Facility since: a) the material in the sample can shall be measured andverified to be less than $1.75 (half of the analyzed $3.50); b) the drive speed cannotexceed the analyzed 42 inches/minute; and c) the aluminum cylinder cannot falluncontrolled faster than the analyzed 42 inches/minute.

Finally, physical impact on the fuel is considered non-credible since the sample can isalways contained in a guide tube or attached to a drive mechanism such that it is unlikelyto drop onto the core (see description in Section 10.4.1.4).

ADD REFERENCE 13.19 TO CHAPTER 1313.1 9 Liu, H. Ben, "Safety Analysis for the Central Irradiation Facility (CIF) at the MNRC",Memorandum to Wade J. Richards, September 22, 1998.CHANGE TECHNICAL SPECIFICATION 3.8.1 AS FOLLOWS:(a) The absolute reactivity worth of any moveable experiment in the CentralIrradiation Facility shall be less than $1 .75 (1 .23% Ak/k); the absolute reactivity worth of any moveable experiment not in the Central Irradiation Facility shall beless than one (1) dollar (0.7% Ak/k).CHANGE TECHNICAL SPECIFICATION 3.8.1 "BASIS" AS FOLLOWS:(a) A reactivity limit of less than $1 .75 specifically for the Central Irradiation Facilityis based on the pulsing reactivity insertion limit (section 3.1.2) and on the designof the sample can assembly which allows insertion and withdrawal ofexperiments in a controlled manner (identical in form, fit, and function to acontrol rod). See also the MNRC SAR, Chapter 13.2.2.2.1 for the maximumreactivity insertion discussion. A reactivity limit of less than one (1) dollar on asingle moveable experiment not in the Central Irradiation Facility will precludepulsing if the experiment's fixturing. should fail, since the resulting reactivity insertion would not cause prompt criticality if less than one dollar. Given that thereactor will not pulse inadvertently, the additional increase in transient power andtemperature will be slow enough so that the fuel temperature scram will beeffective. 0 9 STATES"NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C..20588-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 2 TOAMENDED FACILITY OPERATING LICENSE NO. R-130DEPARTMENT OF THE AIR FORCE ATMcCLELLAN AIR FORCE BASEDOCKET NO. 50-60

71.0 INTRODUCTION

By letter dated January .11, 1999, the Department of the Air Force at McClellan Air ForceBase (the licensee) submitted a request for amendment of the Technical Specifications (TSs), Appendix A, to Facility Operating License No. R-1 30 for the McClellan NuclearRadiation Center TRIGA Research Reactor (MNRC), and changes to the Safety AnalysisReport. The amendment provides for the installation of an Argon-41 Production Facilityand a Central Irradiation Facility. The installation of the Argon-41 Production Facility doesnot require any change or expansion of the TSs since an experiment failure will not resultin airborne radioactivity in the reactor room or the unrestricted area exceeding theapplicable dose limits already prescribed. The installation of the Central Irradiation Facilityrequires a change to TS 3.8.1 with regard to the maximum reactivity worth of a moveable* experiment. The change increases the reactivity limit of a moveable experiment in theCentral Irradiation Facility to $1 .75, corresponding to the pulse limit specified in TS 3.1 .2.2.0 EVALUATION Argon-41 Production FacilityThe licensee has requested amendment of the Safety Analysis Report to provide for theinstallation of an Argon-41 Production Facility. The purpose of the facility is to produceArgon-41 for research and commercial uses. Argon gas from a pressurized argon bottle isintroduced into a stainless steel container located in a position external to the core, but inthe reactor tank. Sufficient argon gas is admitted into the irradiation facility to pressurize it to about 400 psig. The gas is irradiated for 24 hours at full power (48 Megawatt-hours) and is converted to one to two curies of argon-41. The now-radioactive argon-41 isremoved cryogenically and admitted to sample containers. Overpressure protection isprovided by stainless steel relief valves that vent to a 30-gallon carbon steel* overpressure tank which is also protected With a relief valve. The relief valves have a working pressure

• 0-2-of up to 6000 psig. The overpressure tank relief valve vents to the reactor room. Allpiping (1/4 and Y/= inch 304 stainless steel) is anchored to prevent pipe whip in the eventof pipe failure.

The irradiation container has a working pressure of 1 800 psig with a burstpressure of 6000 psig.After the argon gas has been irradiated, the gas is transferred to the sample containers. Acooling coil which has been evacuated with a vacuum pump is immersed in a liquidnitrogen bath. The transfer process is started by opening a valve between the irradiation facility and cooling coil. The argon gas diffuses to the sample containers. When radiation surveys indicate that the transfer process is completed, the sample containers are valvedoff, removed, and placed in.a shipping cask.The licensee has analyzed the case of a catastrophic failure of the irradiation container, which releases 4 curies of argon-41 (about twice as much as is actually produced) into thereactor room resulting in an initial dose rate of approximately 22 rads per hour. Operation of the reactor room ventilation system in the recirculation mode for a period of one daywill result in a dose rate of approximately 1 mrad per hour. The argon-41 discharge limitas defined in the Technical Specifications will not be exceeded. The licensee has considered other potential accidents. These include overpressurization ofthe irradiation container, spilling liquid nitrogen into the reactor tank, and the choking of arelief valve due to supersonic flow. Overpressurization of the irradiation container requiresmultiple mechanical failures and operator violation of the procedure governing the use ofthe production facility. To prevent the spilling of liquid nitrogen into the reactor tank, acatch basin is installed around the liquid nitrogen bath. Finally, the licensee has analyzedthe flow through the relief valves and has determined that the flow remains subsonic, thuspreventing choking at the valve.Central. Irradiation FacilityThe licensee has requested amendment of the 'Technical Specifications and SafetyAnalysis Report to provide for the installation of a Central Irradiation Facility. The facilityallows samples to be inserted into the reactor core while operating the reactor at power.Control of the facility is through use of a drive mechanism similar to that of the normalcontrol rods, and a reactor operator controls the insertion and removal of samples. Drivespeeds are equal to those of the normal control rods.The central thimble is essentially a vertical guide tube which passes through the upper gridplate, the lower grid plate and the safety plate, resting on the tank floor. lA sample canand an aluminum slug move within the central thimble. An aluminum slug normallyoccupies a position in the reactor core. When the sample can is inserted, the aluminumslug moves downward out of the co)re, and its position in the core is replaced by thesample can. Control of the system is only from the reactor c:onsole. The system is provided with*indications

• similar to that of the normal control rods, which include POWER ON, UP,DOWN and CLOSE TO DOWN position indicators, digital indication of sample can position, and UP/DOWN control switches.

From a safety analysis point of viejw, the system can be considered to be an additional control rod and so the analyses in the Safety Analysis Report with respect to control rodmalfunctions are applicable. In particular, the analysiz of an Uncontrolled Withdrawal of aControl Rod (Safety Analysis Report section 13.2.2.2.2) provides a bounding envelope. That analysis showed that an uncontrolled rod withdrawal, at full power of 2 MW, at themaximum withdrawal speed of 42 inches per minute would result in a peak reactivity insertion of $0.25, much lower than the technical specification pulse reactivity insertion limit of $1 .75. Although the maximum single rod worth is approximately $2.65, a rodworth of $3.50 was used to allow for reasonable variations. In order to bound accidents involving the Central Irradiation

Facility, it is required to showthat the worths of the sample can and the aluminum slug are not only less than $3.50,but also less than the pulse limit of $1.75. The licensee has performed a computersimulation (SAR Reference 13.19) of the reactivity changes associated with variousscenarios,-

including normal operations and accidents. The most limiting case, the floodingof the sample can accompanied by a drop of the aluminum slug, results in a reactivity insertion of $1 .76, much less than the most reactive control rod ($3.50) used in the rodwithdrawal accident. Thus the Central Irradiation Facility is bounded by the previously-analyzed rod withdrawal accident. 3.0 ENVIRONMENTAL CONSIDERATION This amendment involves changes in the installation or use of a facility component locatedwithin the restricted area as defined in 10 CFR Part 20 or changes in inspection andsurveillance requirements. The staff has determined that this amendment involves nosignificant increase in the amounts, and no significant change in the types, of anyeffluents that may be released off site, and no significant increase in individual orcumulative occupational radiation exposure. Accordingly, this amendment meets theeligibility criteria for categorical exclusion set forth in 10 CFR 51 .22(c)(9). Pursuant to 10CFR 51 .22(b); no environmental impact statement or environmental assessment need beprepared in connection with the issuance of this amendment.

4.0 CONCLUSION

The staff has concluded, on the basis of the considerations discussed above, that(1) because the amendment does not involve a significant increase in the probability orconsequences of accidents previously evaluated, or create the possibility of a new ordifferent kind of accident from any accident previously evaluated, and does not involve asignificant reduction in a margin of safety, the amendment does not involve a significant hazards consideration; (2) there is reasonable assurance that the health and safety of thepublic will not be endangered by the proposed activities; and (3) such activities will beconducted in compliance with the Commission's regulations and the issuance of thisamendment will not be inimical to the common defense and security or the health andsafety of the public.Principal Contributor: Warren J. EresianDate: MArch 1, 1999 999 9** ** 1~STATES,AUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. Brigadier General Michael P. WledemerCommander Sacramento Air Logistics CenterSM-ALC/TI-1 5335 Price AvenueMcClellan AFB, California 95652-2504 Vice Chancellor Kevin SmithOffice of the Chancellor University of California, DavisOne Shields AvenueDavis, California 95616-8558

SUBJECT:

ORDER APPROVING THE TRANSFER OF THE FACILITY OPERATING LICENSE FOR THE McCLELLAN NUCLEAR RADIATION CENTER FROM THEDEPARTMENT OF THE AIR FORCE TO THE REGENTS OF THE UNIVERSITY OF CALIFORNIA. AND APPROVING CONFORMING AMENDMENT (TAC NO. MA3477)Dear General Wiedemer and Dr. The enclosed Order Is in response to the application dated April 13,.1999, as supplemented onJuly 19 and August 4-,1999, and January 18 and 27, 2000, requesting approval of the transferof Operating License No. R-1 30 for the McClellan Nuclear Radiation Center from theDepartm~ent of the AIr Force to the Regents of the University of California, and approval of aconforming amendment to reflect the transfer. The enclosed Order provides consent to theproposed

transfer, pursuant to Section 50,80 of Title 0oaf the Code of Federal Reaulatlona, and approves Amendment No. 3. Also, enclosed are two copies of the indemnity agreement forthe facility.

The. Vice Chancellor for the University should sIgn one copy and return it to me.The University should keep the other for its records.The Order has been forwarded to the Office of the Federal Register for publication. Sinc~syWarreni~J. Ere fan, Project ManagerEvents Assessment, Generic C~ommunlcations and Non-Power Re~ctom BranchDivIsion of rovement ProgramsOffice of Nuclear Reactor Regulation Docket No. 50-607*

Enclosures:

.1. Order2. Amendment No.3*.3. Safety Evaluation 4, IndemnityAgreement. .*Senextp ge McClellan AFB TRIGA REACTOR Docket No, 50-607cc;Dr. Wade J. RichardsSM-ALC/TI-16335 Price Avenue, Bldg. 258McClellan AFB, CA 95652-2504 Col. Robert CapellHQ AFMC/SGC4225 Logistics Avenue, Suite 23Wright-Patterson AFB, OH 45433-5762 Lt, Cot. Catherine ZeringueHQ AFSC/SEW9570 Avenue G, Building 24499Kircland AFB, New Mexico 871 17-5670Test. Research, and TrainingReactor Newsletter 202 Nuclear Sciences CenterUniversity of FloridaGainesville, FL 32611 7590-01 -PUNITED STATES OF NUCLEAR RIEGU.LATORtY COMMISSION

• In the Matter of ))DEPARTMENT OF THE AIR FORCE ) Docket No, 50-607)(McClellan Nuclear'Radiation Center) )ORDER APPROVING TRANSFER OF LICENSEAND CONFORMING AMENDMENT I,The United States Air Force (USAF) is the owner of the McClellan Nuclear Radiation
• Center (MNRC) and Is authorized to possess, use, and operate thle facility as reflected in*Operating License No, R-130, The Nuclear Regulatory Commission issued Operating License*No. R-1 30 on August 13, 1998, pursuant to Part 50 of Title 10 of the .Code of_ FederalRegufation~s (10 CFR Part 50). The facility is located on McClellan Air Force Base InSacramento, California.

Ii.By letters dated April 13, 1999, the USAF and the Regents of the University of California (University of California) each submitted an application req~uesting approval of the proposedtransfer of Operating License No, R-1 30 from the USAF to the University of California. TheUniversity of Calliornia at Davis (UCD), part of the University of California, was proposed to bethe actual operator of the facility. The application was supplemented by submittals datedJuly 19 and August 4, 1999, and January 18 and 27, 2000, The initial application and thesupplements are hereinafter collectively referred to as "the application" unless otherwise indicated4. ENCLOSURE 1 According to the application, the USAF has agreed to convey the MNRC to the University of California. After completion of the proposed license transfer, UCD would be the soleoperator of the MNRC. The application also sought the approval of a conforming amendment. This conforming amendment is necessary to remove references to the USAF from theoperating license and replace them with references to the UCD, as appropriate, as well as tomake other miscellaneous administrative changes to the operating license to ref lect thetransfer. Under 10 CFR 50.80, no license for a production or utilization

facility, or any rightthereunder, shall be transferred, directly or Indirectly, through transfer of control of the license,unless the Commission shall give Its consent in writing.

Upon review of the information in theapplication and other information before the Commission, the NRC staff has determined thatthe University of California Is qualified to hold the license, and that the transfer of the license tothe University of California is otherwise consistent ~with applicable provisions of law, regulations, and orders issued by the Commission. The NRC staff has further found that the application forthe proposed license amendment complies with the standards and requirements of the AtomicEnergy Act of 1954, as amended, and the Commission's rules and regulations set forth in 10CFR Chapter 1; the facility will operate in conformity with the application, the provisions of theAct, and the rules and regulations of the Commission: there Is. reasonable assurance that theactivities authorized by the proposed license amendment can be conducted withoutendangering the health and safetyof the public and that such activities will be conducted incompliance with theCommission's regulations: the issuance of the proposed licenseamendment will not be inimical to the common defense and security or to the health and safetyof the public; and the issuance of the proposed amendment will be in accordance with 10 CFR r-P ." = 1af1T1 NU.SS5r0 r P.5/) Part 51 of the Commission's regulations and all applicable requirements have been satisfied. The foregoing findings are supported by a Safety Evaluation dated December 2, 1999.Accordingly, IT IS KEREBY ORDERED that the transfer of the license as described hereinto the University of California is approved, subject to the following, condition: Should the transfer of the license not be completed by June 30, 2000, this Order shallbecome null and void, provided,

however, on written application arnd for good causeshown, such date may in writing be extended.

IT IS FURTHER ORDERED that, consistent with 10 CFR 2,1315(b), a license amendment that makes changes, as indicated in Enclosure 2 to the cover letter forwarding this Order, toconform the license to reflect the transfer is approved. This Order is effective upon issuance. Dated at Rock'vilie,

Maryland, this 31't day of ;January 2000,FOR THE= NUCLEAR REGULATORY COMMISSION David B. Matthews, DirectorDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation

4.

STATESWASHINGT"ON, D.C. 20555-0001 DEPARTMENT O T.HE AIR FORCF ATMCCLELLAN. AIR FoRCE BASEDOCKET NO. 50-607AMENDMENT TO AMENDED FACILITY OPERATING LICENSEAmendment No. 3License No. R-1301.The U.$. Nuclear Regulatory Commission (the Commission) has tound thatA. The application for an amendment to Amended Facility Operating License No. R-130filed by tile Department of the Air Force at McClellan Air Force Base and the Regentsof the University of California on April 13, 1999, as supplemented on July 19 andAugust 4, 1999, and January 18 and 27, 2000, conmpiies with the standards andrequirements of the Atomic Energy Act of 1954, as amended (the Act), and theregulations of the Commission as stated In Chapter I of Title 10 of the Code ofFederal R~equlatlons (10 CFR);B. The facility will operate in conformity with the application, the provisions of the Act,and the rules and regulations of the Commission; C. There is reasonable assurance that (i) the activities authorized by this amendment can be conducted without endangering the health and safety of the public and (il)such activties will be conducted in compliance with the regulations of theCommission; D. The issuance of this amendment will not be~inlrmicalto the common defense andsecurity or to the health and safety of the public;and E. This issuance of this amendment is in accordance with the regulations of theCommission as stated in 10 CFR Part 51, and all applicable requirements have beensatisfied.

2. Accordingly, the license is amendedas indicated in the attachment to thisilcense amendment, ENCLOSURE 2

FEB.*1.006

09M N.955 P.7/1.4-2-3. This license amendment is effective as of the date of issuance, FOR THE NUCLEAR REGULATORY COMMISSION Ledyard B. Marsh, ChiefEvents Assessment, Generic Communications and Non-Power Reactors BranchDivsion of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation

Enclosures:

1.2.Amended Facility LicenseAppendix A, Technical Specifications changesDate of Issuance: January 31, 20004 rI" l" NUCLEAR REGULATORY COMMISSION

• ~WASHINGTON, D.C, 20885,=0001FACILITY OPERATING LICENSE~DOCKET NO., 50-607_REGENTS oF THE UNIVERSITY OF ALicense No. R-1301.The U.S. Nuclear Regula.tory Commission (the Commission) has found that:A. The application for license transfer, filed by the Regents of the University of California on April 13, 1999, as supplemented on July 19 and August 4, 1999, and January 18and 27, 2000 comnpiies with the standards and requirements of the Atomic Energy Actof 1954, as amended (the Act), and the Commission's rules and regulations as setforth in 10CFR Chapter I;B. Construction of the facility was completed in substantial conf ormity with the provisions of the Act, and the rules and regulations of the Commission; C. The facility will operate In conformity with the application, the provisions of the Act,and the rules and regulations of the Commission; D. There is reasonable assurance (I) that the activities authorized by this license can beconducted without endangering the health and safety of the public and (II) that suchactivities will be conducted in compliance with the Commission's regulations; E, The licensee is. technically and financially qualifiled to engage in the activities authorized by this operating license in accordan~ce with the regulations of theCommission; F. The licensee is a Nonprofit Educational institution and will use the facility foreducational programs arnd research, and has satief led the applicable provisions of10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements ofthe Commission's regulations; G. The issuance of this license will not be inimical to the common defense and securityor to the health and safety of the public; ."H-. This license is issued in accordance with 10 CFR Part 6.1 of the Commission's regulations, and all applicable requirements have been satisfied; andS.The receipt, possession, and use of the byproduct and special nuclear materials asauthorized by this license will be in accordance with the Commission's regulations in10 CFR Parts 30 and 70, including Sections 30.33. 70,23, and 70.31.Amendment No. 3

2. Facility License No, R-1 30 is hereby issued to the Regents of the University of California as follows:A. The license applies to the TRIGA nuclear reactor (the facility) owned by the Regentsof the University of California (the licensee),

The facility is located on the McClellan IAir Force Base, Sacramento, California, B, Subject to the conditions and requirements Incorporated herein, the Commission hereby licenses the Regents of the University of California at the McClellan Nuclear(i) Pursuant to Section 104o of the Act arid 10 CFR Part 50, "Domestic Licensing ofProduction and Utilization Facilities," to possess, use, and operate the facility atthe designated location at McClellan Air Force Base in accordance with theprocedures and limitations set forth in this license.(Ii) Pursuant to the Act and 10 CFR Part 70, "Domestic Licensing of Special NuclearMaterial,= to receive,

possess, and use up t0 21.0 kilograms of contained uranium-235 enriched to less than 20 percent In the isotope uranium-235 in theformat reactor fuel; up to 4 grams of contained uranium-235 of any enrichment in the~form of fission chambers; u~p to 16,1 kilograms of contained uranium-235 enriched to less than 20 percent in the isotope uranium-235 in the form of plates;and to possess, but not separate, such' special nuclear material as may beproduced by the operation of the facility.

(iii) Pursuant to the Act and 10 CFR Part 30, "Rules of General Applicability toDomestic Licensing of Byproduct Material," to receive,

possess, and use a 4-curie sealed americium-beryllium neutron source in connection with operation ofthe facility; a 55-millicurie sealed cesium-I137 source for instrument calibrations;

• small instrument calibration and check sources of less than 0.1 millicurie each;and to possess, use, but not separate, except for byproduct material produced Inreactor experiments, such byproduct material as may be produced by theape ration of the facility.

C. This license shall be deemed to contain and Is su~bj~ect to the conditions specified inParts 20, 30, 50, 51, 55, 70, and 73 of 10 CFR Chapter I; to all applicable provisions of the Act;and to the rules, regulations, and orders of the Commission now or hereafter in effect and tothe additional conditions specified, below:(i) Maximum Po~wer LevelThe licensee is authorized to operate the facility at steady-state power levels not inexcess of 2300 kilowatts (thermal) arid in the pulse mode with reactivity insertions notto exceed $1.75 (1.23 %/0k/k).Amndent N..h 3 3-3(ii) Technical S~oecfficatlonis The Technical Specifications, as revised through Amendment No. 3, are hereby. fincorporated in the license. The licensee shall operate the facility in accordance withthe Technical Specifications. (lii) Physical Securityv lanThe licensee shall fully implement and maintain in effect all provisions of theCommission-approved physical security plan, including all amendments and revisions made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The approvedplan, which is exempt from public disclosure pursuant to the provisions of 10 CFR2.790, is entitled "Physical Security Plan for the MNRC TRIGA Reactor Facility," Revision 3, and is dated August 1996,D. This license is effective as of the date of issuance and shall expire twenty (20) yearsfrom its date of issuance. FOR THE NUCLEAR REGULATORY COMMISSION Previously signedI byOrigina/ signed bySamuel J, Collins, Directoroffice of Nuclear Reactor Regulation Date of issuance: August 13, 1998Amendment No. 3 Q TO LICENSEAMENDMENT NO.3AMENDED FACILITY OPERATING LI.CENSE NO. R-!30DOCKET NO; 50-807Replace the following pages of Appendix A, "T'echnlcal Specificationts,= with the enclosedpages. The revised pages are identified by amendment number and contain vertical linesindicating the areas of change.*Remove1394041*139404. TECHNICAL SPECIFICATIONS FOR THEU.C. DAVIS MCCLELLAN NUCLEAR RADIATION CENTER (MNRC)GeneralThe McClellan Nuclear Radiation Center (MNRC) reactor is operated by the University ofCalifornia, Davis, CA. The MNRC research reactor Is a TRIGA type reactor. The MNRC iprovides state-of-the-art neutron radiography capabilities. In addition, the MNR~C provides a*wide range of irradiation servic~es far both research and industrial needs. The reactor operatesat a nominal steady start power level up to and including 2 MW. The MNRC reactor is alsocapable of square wave and pulse operational modes. The MNRC reactor fuel Is less than 20%enriched in uranium-235, 1.0 D~efinitions 1.1 ,As Low As Reasonab~ly, Achievable (ALARA), As defined in 10 CFR Part 2.0.1.2 Licens ed DOerators. A MNRC reactor operator is an individual licensed by theNuclear Regulatory Commission (e.g., senior reactor operator or reactor operator) to carry outthe duties and responsibilities associated with the position requiring the license.1.2.1 Senior. Reactor QOerator. An individual who is licensed to direct theactivities of reactor operators and to manipulate the controls of the facility. 1.,2.2 Reactor Onerator. An individual who is licensed to manipulate thecontrols of the facility and perform reactor-related maintenance. 1.3 A channel is the combination of sensor, line amplifier, processor, andoutput devices which are connected for the purpose of measuring the value of a parameter. 1,.3.1 Channel Test. A channel test is the Introduction of a signal into thechannel for verification that it is operable..,.' 1.3.2 Channel Calibratlaon. A channel calibration is an adjustment of thechannel such that its-output corresponds with acceptable accuracy to known values of theparameter which the channel measures. Calibration shall encompass the entire channel,including equipment actuation, alarm or trip, and shall be deemed to include a channel test.1.3.3 Channel Ch~eck. A channel check is a qualit~ative verification ofacceptable performance by observation of channel behavior. This verification, where possible, shall include comparison of the channel with other independent channels or systems measuring the same variable. 1 Amendment No .3. bViCECHACELOR OR ESERCHVICE CHANCELLOR FOR ADMINISTRATION ai.U .D SAFETYK CMMITTEEs IIIsuPERVISOR SUPERVwSOR ...* ------------.[OPERATIONS STAFFI HEALTh- PHYSICS STAFF]~UNIVERSITY MANAGEMENT ORGANIZATION ~Figure 6, !0: Normal Adminisirative Reporting Channel-Technical Review, Communications and Asisisance .

• ~ *I*~C*Lff**J~J

.1. * .LC.[~I.j J.* J.'-tr 7n ----LI.I.VICE OFFzICE OF' RESEARCH II1.* I----I* TUCIEAR SAFETYL AND UCENSINGNUCLEAR SAFETY AND LICENSING

REVIEWS, APPROVALS ANDRECOMMENDATIONS COMMUNICATION LICENSED ACTIVITIES UC. Davis McCleIlan Nuclear Radiatiog Center Lit'easing Organization Figure 6.254An~Iendment Wo. .3

• ,%.UNITED STATESS NUCLEAR REGULATORY COMMISSION' 0, .0. S5-0001Docket No. 50-607This indemnity agreement No. E-40 is entered~into by and between ths University of California at Davis (hereinafter referred to as the licensee) and the United States Nuclear Regulatory Commission (hereinafter referred to as the Commission) pursuant to subsection 170(k) of theAtomic Energy Act of 1954, as amended (hereinafter referred to as the Act).Article IAs used in this agreement,

1. Nuclear reactor, byproduct material,,

person, source material, specIal nuclear material, andprecautionary evacuation shall have the meanings given them in the Atomic Energy Act of 1954,as amended, and the regulations issued by the Commission.

2. (a) Nuclear incident means any occurrence including an extraordinary nuclear occurrence or series of occurrences at the location or in the course of transportation causing bodily injury,sickness,

disease, or death, or loss of use of property, arising out of or resulting from theradioactive, toxic, explosive, or other hazardous properties of the radioactive material.

(b) Any occurrence including an extraordinary nuclear occurrence or series of occurrences causing bodily injury, sickness, disease or death, or loss of or damage to property, or loss of useof property, arising out of or resulting from the radioactive, toxic,.explosive, or other hazardous properties ofi, The radioactive material discharged or dispersed from the location over a period of days,weeks, months or longer and also arising out of such properties of other material defined as theradioactive material in any other agreement or agreements entered into by the Commission under subsection 170(c) or (k) of the Act and so discharged or dispersed from the location asdefined in any such other agreement; orii. The radioactive material in the course of transportation and also arising out of suchproperties of other material defined in any other agreement entered into by the Commission pursuant to subsection 170(c) or (k) of the Act as the radioactive.material and is in thecourse of transportation shall be deemed to be a common octurre.nce. A common occurrence shall be deemed to constitute a single nuclear incident.

3. Extraordinary

nuclear, occurrence mean~s an event which the Commission has determined to be an extraordinary nuclear occurrence as defined in the Atomic Energy Act of 1984, asamended.4. In the course of transportation means In the course of transportation within the UnitedStates, or in the course of transportation outside the United States and any other nation, andmoving from one person licensed by the Commission to another person licensed by theCommission, including handling or temporary storage incidental

thereto, of the radioactive material to the location or from the location provided th~at:ENCLOSURE 4

FEB. ;I.28 5:52PM NO.95? P.2/6(a) With respect to transportationof the radioactive material to the location, suchtransportation is not by predetermination to be interrupted by the removal of the material fromthe transporting conveyance for any purpose other than the continuation of such transportation to the location or temporary storage incidental thereto;(b) The transportation of the radioactive material from the location shall be deemed to endwhen the radioactive material is removed from the transporting conveyance for any purposeother than the continuation of transportation or temporary storage incidental. thereto;(c) In the course of transportation as used in this agreement shall not include transportation ofthe r'adloactive material to the location if the material is also in the course of transportation fromany other location, as defined in any other agreement entered into by the Commission pursuant. to subsection 170(c) or (k) of the Act.5. Person Indemnified means the licensee and any other person who may be liable for public-liability.

6. Public liability means any legal liability arising out of or resulting.from a nuclear incident orprecautionary evacuation (including all reasonable additional costs incurred by a State, or apolitical subdivision of a State, in the course of responding to a nuclear Incident or precautionary evacuation),

except (1) claims under State or Federal Workmnen's Compensation Act ofemployees of persons indemnified who are employed (a) at the location or, if the nuclearIncident occurs in the course of transportation of the radIoactive

material, or the transporting

vehicle, and (b) in connection with the licensee's possession, use, or transfer of the radioaotive material; (2) claims arising out of an act of war;, and (3) claims for loss of, or damage to, or lossof use of (a) property which is located at the location and used in connection with the licensee's possession, use, or transfer of the radioactive

material, and (b) if the nuclear incident occurs Inthe course of transportation of the radioactive

material, the transporting

vehicle, containers used in such transportation, and the radioactive material.

7. The location means the location described in Item 3 of the Attachment hereto.8. The radioactive material means source, special nuclear, and byproduct material which (1)is used or to be used in, or is-irradiated or to be irradiated byl the nuclear reactor or reactorssubject to the license or licenses designated in the Attachment hereto, or (2) which is producedas the result of operation of said reactor(s).

9. United States when used in a geographical sense includes Puerto Rico and all territories and possessions of the united States.Article II1. Any obligations of the licensee under subsection 53e(8.).

of the Act to indemnify the UnitedStates and the Commission from public liability shall not in the aggregate exceed $250,000 withrespe.ct to any nuclear incident.

2. With respect to any extraordinary nuclear occurrence to which this agreement

applies, the,Commission, and the licensee on behalf of itself and other persons indemnified, insofar as theirinterests appear, each agree to waive:(a) Any issue or defense as to the conduct of the claimant or fault of persons indemnified, including, but not limited to (1) Negligence; (2) Contributory negligence; (3) Assumption of the risk;(4) Unforeseeable intervening causes, whether involving the conduct of a third or anact of God.

As used herein, conduct of the claimant includes conduct of persons through whom the claimantderives his cause of action; (b) Any issue or defense as to charitable or governmental immunity: (c) Any Issue or defense based on any statute of limitations if suit is instituted within 3 yearsfrom the date on which the claimant first knew, or reasonably could have known, of his injury ordamage and the cause thereof.*The waiver of any such issue or defense shall be effective regardless of whether such issueor defense may otherwise be deemed jurisdictional or relating to an element in the cause ofaction. The waivers shall be judicially enforceable In accordance with their te.r~ms by the claimantagaInst the person indemnified.

3. The waivers set forth in paragraph 2 of this article:

(a) Shall not preclude a defense basedupon a failure to take reasonable steps to mitigate damages;(b) Shall not apply to injury or damage to a claimant or to a claimant's property which is*intentionally sustained by the claimant or which results from a nuclear incident intentionally andwrongfully caused by the claimant; (c) Shall not apply to injury to a claimant who is employed at the site of and in connection withthe activity where the extraordinary nuclear occurrence takes place if benefits therefor are eitherpayable or required to be provided under any workmen~s compensationi or occupational diseaselaw: Provided,

however, That with respect to an extraordinary nuclear occurrence occurring atthe facility, a claimant who is employed at the facility In connection with the construction of anuclear reactor with respect to which no operating license has been issued by the NuclearRegulatory Commission shall not be considered as employed in connection with the activitywhere the extraordinary nuclear occurrence takes place if:(1) The claimant is employed exclusively in connection with the construction of a nuclearreactor, including all related equipment and installations at the facility, and(2) No operating license has been issued by the NRC with respect to the nuclear reactor, and(3) The claimant is not employed in connection with the possession,

storage, use, or transferof nuclear material at the facility; (d) Shall not apply to anty claim for punitive or exemplary damages.

provided, with respect toany claim for wrongful death under any State law which provides for damages only punitive innature, this exclusion does not apply to the extent that the claimant has sustained actualdamages, measured by the pecuniary injuries resulting from such death but not to exceed themaximum amount otherwise recoverable under such law;(e) Shall be effective only with respect to those obligations set forth in this agreement; (t') Shall not apply to, or prejudice the prosecution or defense of, anty claim or portion of claimwhich is not within the protection afforded under (1) the limit of liability provisions undersubsection 170(e) of the Atomic Energy Act of 1954, as amended, and (b) the terms of thisagreement.

Article Ill1. The Commission undertakes and agrees to Indemnify and hold harmless the licensee andother persons indemnified, as their interest may appear,.from public Bability,

2. With respect to damage caused by a nuclear Incident to property of any person legallyliable for the nuclear incident, the Commission agrees to pay to such person those sums whichsuch person would have been obligated to pay if such property had belonged to another;provided, that the obligation of the Commission under this paragraph 2 does not apply withrespect to: (a) Property which is located at the location and used in connection with thelicensee's possession, use or transfer of the radioactive material; FEB. j..2000 5:53PM NO. .957 P.4/s(b) Property damage due to the neglect of the. person indemnified to use all reasonable means to save and preserve the property after knowledge of a nuclear Incident:,

(C) If the nuclear incident occurs in the course of transportation of the radioactive

material, thetransporting vehicle and containers used-In such transportation; (d) The radioactive material.

3. (Reserved]

4. (a) The obligations of the Commission under this agreement shall apply only with respect tosuch public liability and such damage to property of persons legally liable for the nuclear Incident(other than such property described in the proviso to paragraph 2 of this Article) as in theaggregate exceed $250,000.

(b) With respect to a common occurrence, the obligations of the Commission under this.:agreement shall apply only with respect to such public liability and such damage to property ofpersons legally liable for the nuclear Incident (other than such property described in theproviso to paragraph 2 of this Article) as in the aggregate exceed whichever of the following islower: (1) The sum of the amounts of financial protection established under all applicable agreements: or (2) an amount equal to the sum of $200,000,000 and the amount available assecondary financial protection, As used in this Article applicable agreements means eachagreement entered into by the Commission pursuant to subsection 170(c) or (k) of the Act inwhich agreement the nuclear incident is defined as a common occurrence.

5. The obligations of the Commission under this agreement shall apply only with respect tonuclear incidents occurring during the term of this agreement.

6. The obligations of the Commission Uinder this and all other agreements and contracts towhich the Commission is a party shell not with respect to any nuclear Incident, in the aggregate exceed which ever of the following is the lower. (a) $500,000,000 or (b) With respect to acommon occurrence,

$560,000,000 less the sum of the amounts of financial protection established under all applicable agreements.

7. If the licensee is immune from public liability because It is a state agency, the Commission shall make payments under the agreement in the same manner arnd to the same extent as theCommission would be required to do if the licensee were not such a state agency.8. The obligations of the Commission under this agreement, except to the licensee fordamage to property of the licensee, shall not be affected by any failure on the part of thelicensee to fulfill Its obligations under this agreement.

Bankruptcy or insolvency of tihelicensee or any other person indemnified or of the estate of the licensee or any other personindemnified shall not relieve the Commission of any of its obligations hereunder. Article IV .1. When the Commission determnines that the United States will probably be required to makeindemnity payments under the provisions of this agreement, the Commission shall have the right:to collaborate with the licensee and other persons indemnified in the settlement and defenseof any claim Including such legal costs of the licensee as are approved by the Commission andshall have the right (a) to require the prior approval of the Commission for the settlement orpayment of any claim or action asserted against: the Ilicensee or other person indemnified forpublic liability or damage to property of persons legally liable for the nuclear incident which claimor action the licensee or the Commission may be required to indemnify under this agreement: and (b) to appear through the Attorney General of the United States on behalf of the licensee orother person indemnified, take charge of such action or defend any such action. If the settlement FEB. 1.2B :5P O9 ./or defense of any such action or claim Is undertaken by the Comn'isslon, the licensee shallfurnish all reasonable assistance in effecting a settlement or asserting a defense.2. Neither this agreement nor any interest therein nor claim thereunder may be assigned ortransferred, without the approval of the Commission. Article VThe parties~agree that they will enter into appropriate amendments of this agreement to theextent that such amendments are required pursuant to the Atomic Energy. At of 1954, asamended, or licenses, regulations or orders of the Commission. Article VIThe licensee agrees to pay to the Commission such fees as are established l~y theCommission pursuant to regulations or orders of the Commission,. Article ViiThe term of this agreement shall commence as of the date and time specified in Item 4 of theAttachment and shall terminate at the time of expiration of that license specified in Item 2 of theAttachment, which is the last to expire; provided that, except as may otherwise be provided inapplicable regulations or orders of the Commission, the term of this agreement shall notterminate until all the radioactive material has been removed from the location andtransportation of the radioactive material from the location has ended as defined insubparagraph 4(b), Article I, Termination of the term of this agreement shall not affectany obligation of the licensee or any obligation of the Commission under this agreement withrespect to any nuclear incident occurring during the term of this agreement. 4g. FEB 1200 554p 9NO.957 P.6/6Item 1-Address-- Item 2-Item 3-Item 4-..Attachment to Indemnity Agreement No. E-40LicenseeUniversity of California, DavisOne Shields Avenue, Davis, California 9561648558 License number or numbersR-130LocationThe reactor is located in the McClellan Nuclear Radiation Center Buildingon McClellan AFB, located approximately 8 miles northeast ofSacramento, California. The indemnity agreement designated above, of which this Attachment Isa part of, is effective on the day of , 2000,For the United States Nuclear Regulatory Commission, Cyhit,o,Che Generic Issues, Environmental, Financial, and Rulemaking BranchDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation Dated at Rock'ville, MD, the day of ,2000._________________By Kevin SmithVice C~hancelior University of California, Davis Fz~? 0UNITED STATES%" NUCLEAR REGULATORY COMMISSION / WASHINGTON, D.C. 20555-0001 9, 2001Dr. Kevin Smith, Vice Chancellor Office of the Chancellor University of California, DavisOne Shields AvenueDavis, CA 95616-8558

SUBJECT:

ISSUANCE OF AMENDMENT NO. 4 TO AMENDED FACILITY OPERATING LICENSE NO. R-130 -REGENTS OF THE UNIVERSITY OF CALIFORNIA (TAC NO. 8391)

Dear Dr. Smith:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 4 to FacilityOperating License No. R-130 for the McClellan Nuclear Radiation Center (MNRC) TRIGAResearch Reactor. The amendment consists of changes to the Technical Specifications (TSs)in response to your submittal of May 11, 2001.The amendment reflects the administrative changes to the TSs as a result of the transfer of thelicense from the Department of the Air Force to the Regents of the University of California. There are other, non-administrative

changes, which are also reflected in this amendment andwhich are discussed in the enclosed safety evaluation report.Sincerely, Warren J. Eresian, Project ManagerOperational Experience and Non-Power Reactors BranchDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation Docket No. 50-607

Enclosures:

1. Amendment No. 42. Safety Evaluation cc w/enclosures:

Please see next page University of California -Davis/McClellan MNRC Docket No. 50-607co:Dr. Wade J. Richards5335 Price Avenue, Bldg. 258McClellan AFB, CA 95652-2504 Test, Research, and TrainingReactor Newsletter University of Florida202 Nuclear Sciences CenterGainesville, FL 32611 - STATESNUCLEAR REGULATORY COMMISSIONWASHINGTON, D.C. 20555-0001 REGENTS OF THE UNIVERSITY OF CALIFORNIA ATMcCLELLAN NUCLEAR RADIATION CENTERDOCKET NO. 50-607AMENDMENT TO AMENDED FACILITY OPERATING LICENSEAmendment No. 4License No. R-1301. The U.S. Nuclear Regulatory Commission (the Commission) has found thatA. The application for an amendment to Amended Facility Operating License No.R-1 30 filed by the Regents of the University of California at McClellan NuclearRadiation Center (the licensee) on May 11, 2001, conforms to the standlards andrequirements of the Atomic Energy Act of 1954, as amended (the Act), and theregulations of the Commission as stated in Chapter I of Title 10 of the Code ofFederal Regulations (10 CFR);B. The facility will operate in conformity with the application, the provisions of the Act,and the rules and regulations of the Commission; C. There is reasonable assurance that (i) the activities authorized by this amendment can be conducted without endangering the health and safety of the public and (ii)such activities will be conducted in compliance with the regulations of theCommission; D. The issuance of this amendment will not be inimical to the common defense andsecurity or to the health and safety of the public;E. This amendment is issued in accordance with the regulations of the Commission asstated in 10 CFR Part 51, and all applicable requirements have been satisfied; andF. Prior notice of this amendment was not required by 10 CFR 2.105, and publication of notice for this amendment is not required by 10 CFR2.106. 2. Accordingly, the license is amended by changes to the Technical Specifications asindicated in the enclosure to this license amendment, and paragraph 2.C.(ii) of AmendedFacility Operating License No. R-130 is hereby amended to read as follows:2.c.(ii) Technical Specifications The Technical Specifications contained in Appendix A, as revised throughAmendment No. 4, are hereby incorporated in the license. The licensee shalloperate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Warren J. Eresian, Project ManagerOperational Experience and Non-Power Reactors BranchDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation

Enclosure:

Appendix A, Technical Specification ChangesDate of Issuance: August 9, 2001 S 0ENCLOSURE TO LICENSE AMENDMENT NO. 4AMENDED FACILITY OPERATING LICENSE NO. R-130DOCKET NO. 50-607Replace the following pages of Appendix A, Technical Specifications, with the enclosed pages. Therevised pages are identified by amendment number and contain vertical lines indicating the areas ofchange.Remove Insertii iiiii iiiiv ivV vvi vi1 I2 23 34 46 67 79 913 1314 1415 1516 1617 1718 1819 1925 2526 2627 2728 2829 2930 3031 3132 3233 3334 3435 3536 3639 3940 40 UNITED STATES1"%" NUCLEAR REGULATORY COMMISSION ~WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 4 TOAMENDED FACILITY OPERATING LICENSE NO. R-130REGENTS OF THE UNIVERSITY OF CALIFORNIA ATMcCLELLAN NUCLEAR RADIATION CENTERDOCKET NO. 50-60

71.0 INTRODUCTION

By letter dated May 11, 2001, the Regents of the University of California (the licensee) submitted a request for amendment of the Technical Specifications (TSs), Appendix A, toFacility Operating License No. R-1 30 for the McClellan Nuclear Radiation Center (MNRC)TRIGA research reactor. (On July 9, 2001, the licensee resubmitted the amendment requestunder oath. The resubmittal contained no new information.) The request provides for thefollowing

changes, which if implemented, will result in Revision 11 of the TSs:1, On February 1, 2000, the operating license for MNRC was transferred from theDepartment of the Air Force to the Regents of the University of California.

As a result ofthis transfer, a nUmber of administrative changes simply involving name changes (e.g.,changing references from "Responsible Commander" to "Vice Chancellor of the Office ofResearch" and "Air Force" to "University of California-Davis," etc.) is necessary

2. Section 2.1, Basis b. This section has been expanded to include more detail regarding cladding integrity during pulsing operation.

3. Section 3.3, Table 3.3. A request to increase the alarm setpoint for the heat exchanger outlet temperature from 35 degrees Centigrade to 45 degrees Centigrade.

4. Section 4.7, Specification 4.7.a(3),

4.7.b(3) and 4.7.d(3). A request to allow channelcalibrations to be performed annually rather than semiannually.

5. Section 5.3.1. A request to add the use of 30/20 TRIGA fuel and a new core fuel loadingtermed a 30B core.6. Section 6.0. A request to revise the organization and duties of the Nuclear SafetyCommittee and to clarify the Committee's review and audit functions to reflect the newlicensee. 7. A request for approval of a new Iodine-125 production loop.8. Section 3.8.2. Clarifies reactivity limits for experiments, and adds a new paragraph pertaining to the Iodine-I125 production facility.

2.0 EVALUATION

The staff has considered each of the items 1-8 above. Each item is discussed below.2.1 Administrative changes.As a result of the February 1, 2000, transfer of the Operating License from the Department ofthe Air Force to the Regents of the University of California, the TSs must be modified to takeaccount of administrative changes. These changes will occur in a number of places, andconsist of the substitution of Department of the Air Force organizational and position titles withcorresponding University of California titles. The substitutions are made on a one-for-one basis.These changes are also reflected in Figure 6.1, "UCD/MNRC Organization for Licensing andOperation." The staff concludes that there has been no diminishment of licensee oversight (i.e.,the lines of authority and responsibility have not been weakened) and that these changes areacceptable. 2.2 Section 2.1, Basis b.The previous version of the Technical Specifications addressed the issue of the effect of pulsingon fuel clad integrity and concluded that TRIGA fuel of the type used in the McClellan reactorcould be pulsed up to temperatures of 1150 degrees Centigrade without damage to the clad,provided that the clad temperature was less than 500 degrees Centigrade. The presentanalysis expands the discussion to include more recent measurements of hydrogen pressureresulting from pulses and concludes that the cladding will not rupture if fuel temperatures arenever greater than 1200 to 1250 degrees Centigrade, providing the cladding temperature is lessthan 500 degrees Centigrade. Since the pulse reactivity limit remains at $1.75, the staffconcludes that the bases for Section 2.1 are more conservative and this is acceptable. 2.3 Section 3.3, Table 3.3.A re-evaluation of the thermal and hydraulic analyses and operating limits was performed byResearch Reactor Safety Analysis Services (RRSAS-99-6-1, December 1999) to determine ifthe conservative maximum core inlet temperature (heat exchanger outlet temperature) as set bythe U.S. Air Force in the original design could be raised from 35 degrees Centigrade to 45degrees Centigrade. The effect of the lower limit is that the reactor power is required to bereduced below the license limit of 2 MW whenever ambient local weather conditions prevent thesystem from maintaining the heat exchanger outlet temperature at or below the lower limit.Evaluation of data during 2 MW startup tests as well as data from subsequent steady stateoperations, when compared with previous calculations by Argonne National Laboratory, GeneralAtomics published

reports, and results from power upgrades at the Sandia Annular Core 0-3-Research Reactor facility shows that the maximum core inlet temperature can be raised to45 degrees Centigrade with only a small reduction in Critical Heat Flux ratio (from 2.53 to 2.40).These numbers have been also confirmed by RELAP5 thermal hydraulic calculations.

Thecalculations also show that there is no increase in the maximum fuel temperature or themaximum fuel clad surface temperature, two of the most important parameters which measurefuel integrity. Accordingly, the staff concludes that safety limits will not be reduced and thatthere is no reduction in safety margin.2.4 Section 4.7, Specification 4.7.a(3), 4.7.b(3) and 4.7.d(3). This section of the Technical Specifications addresses channel calibration frequencies for thestack monitor system, the reactor room radiation monitor and the reactor room continuous airmonitor. These systems are presently required to be calibrated semiannually. The licensee hasrequested that they be calibrated annually. The requirement for semiannual calibrations stems from the original Department of the Air Forcelicensing organization, but has no operational safety basis. Research reactors of similar powerlevels currently licensed by the NRC (National Institute of Standards and Technology, RhodeIsland AEC) are permitted to calibrate similar instruments on an annual basis, since there areno operating experience data to suggest that this practice would compromise safety. Inaddition, the American National Standard ANSI/ANS 15.11 "Radiation Protection at ResearchReactor Facilities," states that "Instruments shall be tested at least annually in a performance quality assurance program [i.e., calibration], or more frequently if subject to extreme conditions." The facility is not subject to extreme conditions, and the staff concludes that annual calibrations are acceptable. 2.5 Section 5.3.1.When the McClellan reacto'r was originally licensed by the NRC (August 1998), the reactor wasoperating with a mixed core of 8.5/20 and 20/20 fuel loading (referred to as the MixJ core in theoriginal SAR). At that time it was understood that the reactor would eventually transition to acore consisting of 20/20 and 30/20 fuel, termed a 30B core. The 30B core was analyzed in theoriginal SAR and found to be acceptable by the NRC staff in the SER. In addition, the NRCstaff had previously approved the generic use of TRIGA fuels with uranium loadings of up to30 wt% in licensed TRIGA reactors (NUREG-1282.) The staff concludes that the introduction of 30/20 fuel is consistent with previous analyses and does not create any additional hazards.2.6 Section 6.0.Section 6.0 of the Technical Specifications describes the administrative controls governing theoperation and maintenance of the reactor and associated equipment. There are a number ofminor changes with respect to titles and some changes with .respect to the composition andduties of the Nuclear Safety Committee (NSC). The review and inspection functions of the NSChave been expanded to provide additional oversight. These expanded functions include reviewof the Emergency Plan and Physical Security Plan, review and update of the NSC Charter everytwo years, review of inspections conducted by other agencies, assessment of actions taken tocorrect deficiencies, inspection of currently active experiments, and inspection of future plansfor facility modifications or facility utilization. Since these changes increase oversight of facilityoperations, the staff concludes that they are acceptable. 0-4-2.7 A request for approval of a new Iodine-I125 production loop.The licensee has requested amendment of the Safety Analysis Report to provide for theinstallation of an Iodine-125 production loop. The purpose of the loop is to produce from ten totwenty curies of lodine-I25 for use as a medical radioisotope. The production of Iodine-I25 occurs in five steps:I. Xenon-124 is transferred from a storage tank into an irradiation chamber located in thereactor core.2. The Xenon-I 24 is irradiated over an eight to sixteen hour time span and by neutronactivation results in the production of Xenon-125. The activated Xenon-I124 gascontains up to 4,000 curies of Xenon-125.

3. The Xenon-125 is transferred to a tank, referred to as decay storage I, where it decayswith a 17-hour half-life to Iodine-I125.

After a few days, most of the Xenon-I125 hasdecayed and the Iodine-125 plates out in the tank.4. The Xenon-I 25 remaining in decay storage I is transferred to another tank, referred toas decay storage 2.5. The Iodine-125 in decay storage I is recovered by washing the tank with a NaOHsolution, resulting in a Nal solution which is packaged as a liquid and sent to an off-siteuser in an appropriate DOT container. All equipment used in the production loop is located within a primary containment and asecondary containment. The primary containment houses the irradiation

chamber, tubing,pneumatically operated valves, transfer vessel, decay storage I and decay storage 2. Thesecondary containment is placed around the primary containment to the irradiation chamber andallows for recovering the xenon gas if a leak occurs within the primary containment.

Shielding around the secondary containment reduces radiation levels to below 10 mrem/hr. Both of thesecontainments are within the reactor room, which has a ventilation system withisolation/recirculation capability. There are two other structures within the reactor room which are confinement barriers designedfor the safety of personnel working with the production loop. The first is a glove box whichcontains controls for operation of the Iodine-125 recovery system. The glove box has its ownventilation and filtration system which exhausts into the reactor room ventilation system. Thesecond is a fume hood in which quality assurance of the Iodine-125 is performed. The fumehood also contains its own ventilation and filtration system which exhausts into the reactor roomventilation system.The licensee has analyzed the situation (worst-case) whereby all of the Xenon-I125 from theprimary containment leaks into the secondary containment and subsequently leaks into thereactor room at the design leak rate of the secondary containment. Their analysis shows thatexposures to personnel in the reactor room would result in a deep dose equivalent (DDE) of 17millirem after one hour, about 1.4 millirem for a five-minute occupancy, and about 0.6 millirem for a two-minute occupancy, all well within 10 CFR 20 limits. Exposures to personnel located atthe boundary of the unrestricted area for a full year would be approximately 7 millirem. The Maximum Hypothetical Accident analyzed in the Safety Analysis Report (SAR) is a claddingrupture of one highly irradiated fuel element with no decay followed by instantaneous release offission products into the air. At the closest distance to the site boundary (10 meters), themaximum dose to a member of the general public is 66 millirem, received over an approximately 10-minute period. The dose received at the same location due to a failure of the Iodine-125 production loop is approximately 7 millirem over a period of one year.The staff concludes that the installation of the Iodine-I125 production loop does not reduce themargin of safety with respect to 10 CFR 20 limits and that the installation of the production loopis acceptable. 2.8 Section 3.8.2.This section of the Technical Specifications has been expanded to take account of the Iodine-125 production loop. Sections 3.8.2.c and 3.8.2.d have been added to limit the amount ofIodine-125 present in the reactor room glove box and the reactor room fume hood. Limiting theamount of Iodine-125 in these areas will reduce the occupational dose and dose to personnel inthe unrestricted areas to less than 10 CFR 20 limits if the inventories of Iodine-I125 are totallyreleased within the glove box and fume hood. The staff concludes that this is acceptable. 3.0 ENVIRONMENTAL CONSIDERATION This amendment involves changes in the installation or use of a facility component locatedwithin the restricted area as defined in 10 CFR Part 20 or changes in inspection andsurveillance requirements. The staff has determined that this amendment involves nosignificant increase in the amounts, and no significant change in the types, of any effluents thatmay be released off site, and no significant increase in individual or cumulative occupational radiation exposure. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CER 51 .22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared with the issuance of thisamendment.

4.0 CONCLUSION

The staff has concluded, on the basis of the considerations discussed above, that (1) becausethe amendment does not involve a significant increase in the probability or consequences ofaccidents previously evaluated, or create the possibility of a new or different kind of accidentfrom any accident previously evaluated, and does not involve a significant reduction in a marginof safety, the amendment does not involve a significant hazards consideration; (2) there isreasonable assurance that the health and safety of the public will not be endangered by theproposed changes; and (3) such changes are in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security orthe health and safety of the public.Principal Contributor: Warren J. EresianDate: August 9, 2001 0I 0TECHNICAL SPECIFICATIONS FOR THEUNIVERSITY OF CALIFORNIA -DAVIS MCCLELLAN NUCLEAR RADIATION CENTER(UCD/MNRC) DOCUMENT NUMBER: MNRC-0004-DOC-1 1Rev 11, 12/10/99Amendment No. 4i 0TECHNICAL SPECIFICATIONS APPROVALThese "Technical Specifications" for the University of California at Davis/McClellan Nuclear Radiation Center(UCD/MNRC) Reactor have undergone the following coordination: Reviewed z 'ODteReviewed by: k_,. Q- Reactor Operations S~pervisor Approved by: ,U49UCD/MNI DirectorApproved by:________________

Chairman, UCD/MNRCNuclear Safety Committee (Date)(Date)(Date)Amendment No. 4ii 0TECHNICAL SPECIFICATIONS TABLE OF CONTENTSPage1.0 Definitions...................................................................................................................1 2.0 Safety Limit and Limiting Safety System Setting (LSSS)...............................................................

52.1 Safety 52.2 Limiting Safety System Selling (LSSS)......................................................................... 62.2.1 Fuel Temperature .................................................................................... 63.0 Limiting Conditions for Operations (LCO) ............................................................................... 73.1 Reactor Core Parameters ....................................................................................... 73.1.1 Steady-State Operation.................................................. 73.1.2 Pulse or Square Wave Operation................................................................... 73.1.3 Reactivity Limitations................................................................................. 83.2 Reactor Control and Safety Systems........................................................................... 83.2.1 Control Rods........................................................................................... 83.2.2 Reactor Instrumentation ............................................................................. 93.2.3 Reactor, Scrams and Interlocks .................................................................... 103.2.4 Reactor Fuel Elements ............................................................................. 123.3 Reactor Coolant Systems...................................................................................... 133.4 Reactor Room Exhaust System ............................................................................... 14.3.5 Intentionally Left Blank .................................... ..................................................... 143.6 Intentionally Left Blank ......................................................................................... 143.7 Reactor Radiation Monitoring Systems ....................................................................... 143.7.1 Monitoring Systems................................................................................. 143.7.2 Effluents -Argon-41 Discharge Limit .............................................................. 16Amendment No. 4iii Pane3.8 Experiments ..................................................................................................... 163.8.1 Reactivity Limits ..................................................................................... 163.8.2 Materials Limit ....................................................................................... 173.8.3 Failure and Malfunctions ........................................................................... 184.0 Surveillance Requirements .................................................................. ........................... 194.1 Reactor Core Parameters...................................................................................... 194.1.1 Steady State Operation ............................................................................ 194.1.2 Shutdown Margin and Excess Reactivity.......................................................... 204.2 Reactor Control and Safety Systems ......................................................................... 204.2.1 Control Rods......................................................................................... 204.2.2 Reactor Instrumentation ............................................................................ 214.2.3 Reactor Scrams and Interlocks .................................................................... 224.2.4 Reactor Fuel Elements ............................................................................. 234.3 Reactor Coolant Systems...................................................................................... 244.4 Reactor Room Exhaust Systerm........ ....................................................................... 254.5 Intentionally Left Blank ......................................................................................... 254.6 Intentionally Left Blank ......................................................................................... 254.7 Reactor Radiation Monitoring Systems ....................................................................... 254.8 Experiments ..................................................................................................... 265.0 Design Features.......................................................................................................... 275.1 Site and Facility Description.................................................................................... 275.1.1 Site.................................................................................................... 275.1.2 Facility Exhaust...................................................................................... 285.2 Reactor Coolant System........................................................................................ 28Amendment No. 4iv 0 0Page5.3 Reactor Core and Fuel ........................................................................................... 295.3.1 Reactor Core........................................................................................... 295.3.2 Reactor FueL........................................................................................... 305.3.3 Control Rods and Control Rod Drives .............................................................. 315.4 Fissionable Material Storage .................................................................................... 316.0 Administrative Controls..................................................................................................... 316.1.1 Structure................................................................................................ 326.1.2 Responsibilities........................................................................................ 326.1.3 Staffing.................................................................................................. 326.1.4 Selection and Training of Personnel ................................................................ 326.2 Review, Audit, Recommendation and Approval............................................................... 326.2.1 NSC Composition and Qualifications............................................................... 336.2.2 NSC Charter and Rules` ............................................................................. 336.2.3 Review Functiont...................................................................................... 336.2.4 Audit/Inspection Function ............................................................................ 346.3 Radiation Safety. ............................................. ..................................................... 346.4 Procedures ........................................................................................................ 346.4.1 Reactor Operations Procedures........ .................. ........................................... 346.4.2 Health Physics Procedures .......................................................................... 356.5 Experiment Review and Approval............................................................................... 356.6 Required Actions.................................................................................................. 356.6.1 Actions to be taken in case of a safety limit violation.............................................. 356.6.2 Actions to be taken for reportable occurrences`................................................... 36Amendment No. 4V 6.77Rep Ortstn R po t ..................................................... ........366.7.2 Special Reports........................................................................................ 386.8 Records ............................................................................................................. 39Fig. 6.1 UCD/MNRC Organization for Licensing and 40Amendment No. 4vi 0TECHNICAL SPECIFICATIONS FOR THEUNIVERSITY OF CALIFORNIA -DAVIS/MCCLELLAN NUCLEAR RADIATION CENTER (UCD/MNRC) GeneralThe University of California -Davis/McClellan Nuclear Radiation Center (UCD/MNRC) reactor is operated by theUniversity of California, Davis, California (UCD). The UCD/MNRC research reactor is a TRIGA-type reactor. TheUCD/MNRC provides state-of-the-art neutron radiography capabilities. In addition, the UCD/MNRC provides a widerange of irradiation services for both research and industrial needs. The reactor operates at a nominal steady statepower level up to and including 2 MW. The UCD/MNRC reactor is also capable of square wave and pulseoperational modes. The UCD/MNRC reactor fuel is less than 20% enriched in uranium-235.

1.0 Definitions

1.1 As Low As Reasonably Achievable (ALARA). As defined in 10 CFR, Part 20.1.2 Licensed Operators. A UCD/MNRC licensed operator is an individual licensed by the NuclearRegulatory Commission (e.g., senior reactor operator or reactor operator) to carry out the duties andresponsibilities associated with the position requiring the license.1.2.1 Senior Reactor Operator. An individual who is licensed to direct the activities of reactoroperators and to maniPulate the controls of the facility. 1.2.2 Reactor Operator. An individual who is licensed to manipulate the controls of the facility andperform reactor-related maintenance. 1.3 Channel. A channel is the combination of sensor, line amplifier, processor, and output devices whichare connected for the purpose of measuring the value of a parameter. 1.3.1 Channel Test. A channel test is the introduction of a signal into the channel for verification that it is operable. 1.3.2 Channel Calibration. A channel calibration is an adjustment of the channel such that itsoutput corresponds with acceptable accuracy to known values of the parameter which the channelmeasures. Calibration shall encompass the entire channel, including equipment actuation, alarmor trip, and shall be deemed to include a channel test.1.3.3 Channel Check. A channel check is a qualitative verification of acceptable performance byobservation of channel behavior. This verification, where possible, shall include comparison of thechannel with other independent channels or systems measuring the same variable. 1.4 Confinement. Confinement means isolation of the reactor room air volume such that the movement ofair into and out of the reactor room is through a controlled path.1.5 Experiment. Any operation,

hardware, or target (excluding devices such as detectors, fissionchambers, foils, etc), which is designed to investigate specific reactor characteristics or which is intendedfor irradiation within an experiment facility and which is not rigidly secured to a core or shield structure soas to be a part of their design.1.5.1 Experim~ent.

Moveable. A moveable experiment is one where it is intended that the entireexperiment may be moved in or near the reactor core or into and out of reactor experiment facilities while the reactor is operating. Amendment No. 41

• 01.5.2 Experiment.

Secured. A secured experiment is any experiment, experiment

facility, orcomponent of an experiment that is held in a stationary position relative to the reactor bymechanical means. The restraining force rmust be substantially greater than those to which theexperiment might be subjected by hydraulic, pneumatic,

buoyant, or other forces which are normalto the operating environment of the experiment, or by forces which can arise as a result of credibleconditions.

1.5.3 Experiment Facilities. Experiment facilities shall mean the pneumatic transfer tube,beamtubes, irradiation facilities in the reactor core or in the reactor tank, and radiography bays..1.5.4 Experiment Safety System. Experiment safety systems are those systems, including theirassociated input circuits, which are designed to initiate a scram for the primary purpose ofprotecting an experiment or to provide information which requires manual protective action to beinitiated. 1.6 Fuel Element, Standard. A fuel element is a single TRIGA element. The fuel is U-ZrH clad in stainless steel. The zirconium to hydrogen ratio is nominally 1.65 +/- 0.05. The weight percent (wt%) of uranium canbe either 8.5, 20, or 30 wt%, with an enrichment of less than 20% U-235. A standard fuel element maycontain a burnable poison.1.7 Fuel Element. Instrumented. An instrumented fuel element is a standard fuel element fabricate~d withthermocouples for temperature measurements. An instrumented fuel element shall have at least oneoperable thermocouple embedded in the fuel near the axial and radial midpoints. 1.8 Measured Value. The measured value is the value of a parameter as it appears on the output of achannel.1.9 Mode, Steady-State. Steady-state mode operation shall mean operation of the UCD/MNRC reactor withthe selector switch in the automatic or manual mode position. 1.10 Mode, Square-Wave. Square-wave mode operation shall mean operation of the UCD/MNRC reactorwith the selector switch in the square-wave mode position. 1.11 Mode, Pulse. Pulse mode operation shall mean operation of the UCD/MNRC reactor with the selectorswitch in the pulse mode position. 1.12 Operable. Operable means a component or system is capable of performing its intended function. 1.13 Operatingq. Operating means a component or system is performing its intended function. 1.14 Operatinq Cycle. The period of time starting with reactor startup and ending with reactor shutdown. 1.15 Protective Action. Protective action is the initiation of a signal or the operation of equipment within theUCD/MNRC reactor safety system in response to a variable or condition of the UCD/MNRC reactor facilityhaving reached a specified limit.1.15.1 Channel Level. At the protective instrument channel level, protective action is the generation and transmission of a scram signal indicating that a reactor variable has reached the specified limit.1.15.2 Subsystem Level. At the protective instrument subsystem level, protective action is thegeneration and transmission of a scram signal indicating that a specified limit has been reached.NOTE: Protective action at this level would lead to the operation of the safety shutdown equipment. Amendment No. 42 1.15.3 Instrument System Level. At the protective instrument level, protective action is thegeneration and transmission of the command signal for the safety shutdown equipment to operate.1.15.4 Safety System Level. At the reactor safety system level, protective action is the operation ofsufficient equipment to immediately shut down the reactor.1.16 Pulse QOerational Core. A pulse operational core is a reactor operational core for which the maximumallowable pulse reactivity insertion has been determined. 1.17 .Reactivity, Excess. Excess reactivity is that amount of reactivity that would exist if all control rods(control, regulating, etc.) were moved to the maximum reactive position from the point where the reactor isat ambient temperature and the reactor is critical. = 1)1.18 Reactivity Limits. The reactivity limits are those limits imposed on the reactivity conditions of the reactorcore.1.19 Reactivity Worth of an Experiment. The reactivity worth of an experiment is the maximum value of thereactivity change that could occur as a result of changes that alter experiment position or configuration. 1.20 Reactor Controls. Reactor controls are apparatus and/or mechanisms the manipulation of whichdirectly affect the reactivity or power level of the reactor.1.21 .Reactor Core. Operational. The UCD/MNRC reactor operational core is a core for which theparameters of excess reactivity, shutdown margin, fuel temperature, power calibration and reactivity worthsof control rods and experiments have been determined to satisfy the requirements set forth in theseTechnical Specifications. 1.22 _Reactor Operatingq. The UCD/MNRC reactor is operating whenever it is not shutdown or secured.1.23 Reactor Safety Systems. Reactor safety systems are those systems, including their associated inputchannels, which are designed to initiate automatic reactor protection or to provide information for initiation of manual protective action.1.24 _Reactor Secured. The UCD/MNRC reactor is secured when the console key switch is in the offposition and the key.is removed from the lock and under the control of a licensed

operator, and theconditions of a or b exist:a. (1) The minimum number of control rods are fully inserted to ensure the reactor is shutdown, as requiredby technical specifications; and(2) No work is in progress involving core fuel, core structure, installed control rods, or control rod drives,unless the control rod drives are physically decoupled from the control rods; and(3) No experiments in any reactor experiment

facility, or in any other way near the reactor, are beingmoved or serviced if the experiments have, on movement, a reactivity worth exceeding the maximum valueallowed for a single experiment or $1.00, whichever is smaller, orb. The reactor contains insufficient fissile materials in the reactor core, adjacent experiments or control rodsto attain criticality under optimum available conditions of moderation and reflection.

1.25 _Reactor Shutdown. The UCD/MNRC reactor is shutdown if it is subcritical by at least one dollar ($1.00)both in the Reference Core Condition and for all alloWed ambient conditions with the reactivity worth of allinstalled experiments included. Amendment No. 43

• 01.26 Reference Gore Condition.

The condition of the core when it is at ambient temperature (cold T<28° C),the reactivity worth of xenon is negligible (< $0.30) (i.e., cold and clean), and the central irradiation facilitycontains the graphite thimble plug and the aluminum thimble plug (CIF-1).1.27 Research Reactor. A research reactor is defined as a device designed to support a self-sustaining neutron chain reaction for research development, education, and training, or experimental

purposes, andwhich may have provisions for the production of radioisotopes.

1'.28 Rod, Control. A control rod is a device fabricated from neutron absorbing

material, with or without afuel or air follower, which is used to establish neutron flux changes and to compensate for routine reactivity losses. The follower may be a stainless steel section.

A control rod shall be coupled to its drive unit to allowit to perform its control function, and its safety function when the coupling is disengaged. This safetyfunction is commonly termed a scram.1.28.1 Regqulatingq Rod. A regulating rod is a Control rod'used to maintain an intended power leveland may be varied manually or by a servo-controller. A regulating rod shall have scram capability. 1.28.2 Standard Rod. The regulating and shim rods are standard control rods.1.28.3 Transient Rod. The transient rod is a control rod that is capable of providing rapid reactivity insertion to produce a pulse or square wave.1.29 Safety Channel. A safety channel is a measuring channel in the reactor safety system.1.30 Safety Limit. Safety limits are limits on important process variables, which are found to be necessary to reasonably protect the integrity of the principal barriers which guard against the uncontrolled release ofradioactivity. 1.31 Scram Time. Scram time is the elapsed time between reaching a limiting safety system set point andthe control rods being fully inserted. 1.32 Scram. External. External scrams may arise from the radiography bay doors, radiography bay ripcords, bay shutter interlocks, and any scrams from an experiment. 1.33 Shall. Should and May. The word "shall" is used to denote a requirement; the word "should" to denotea recommendation; the word "may" to denote permission, neither a requirement nor a recommendation. 1.34 Shutdown Margqin. Shutdown margin shall mean the minimum shutdown reactivity necessary toprovide confidence that the reactor can be made subcritical by means of the control and safety systemstarting from any permissible operating condition with the most reactive rod assumed to be in the mostreactive

position, and once this action has been initiated, the reactor will remain subcritical without furtheroperator action.1.35 Shutdown.

Unscheduled. An unscheduled shutdown is any unplanned shutdown of the UCD/MNRC-reactor caused by actuation of the reactor safety system, operator error, equipment malfunction, or a manualshutdown in response to conditions which could adversely affect safe operation, not including shutdowns which occur during testing or check-out operations. 1.36 Surveillance Activities. In general, two types of surveillance activities are specified: operability checksand tests, and calibrations. Operability checks and tests are generally specified as daily, weekly orquarterly. Calibration times are generally specified as quarterly, semi-annually,

annually, or biennially.

1.37 Surveillance Intervals. Maximum intervals are established to provide operational flexibility and not toreduce frequency. Established frequencies shall be maintained over the long term. The allowable surveillance interval is the interval between a check, test, or calibration, whichever is appropriate to the itemAmendment No. 44 0fuel element temperature. This parameter is well suited as it can be measured directly. A loss in theintegrity of the fuel element cladding could arise if the cladding stress exceeds the ultimate strength of thecladding material. The fuel element cladding stress is a function of the element's internal pressure while theultimate strength of the cladding material is a function of its temperature. The cladding stress is a result ofthe internal pressure due to the presence of air, fission product gasses and hydrogen from thedisassociation of hydrogen and zirconium in the fuel moderator. Hydrogen pressure is the most significant. The magnitude of the pressure is determined by the fuel moderator temperature and the ratio of hydrogento zirconium in the alloy. At a fuel temperature of 930°C for ZrH1.7 fuel, the cladding stress due to theinternal pressure is equal to the ultimate strength of the cladding material at the same temperature (SAR Fig4.18). This is a conservative limit since the temperature of the cladding material is always lower than the fueltemperature. (See SAR Chapter 4, Section 4.5.4.)b. This fuel safety limit applies for conditions in which the cladding temperature is less than 50000. Analysis(SAR Chapter 4, Section 4.5.4.1.1), shows that a maximum temperature for the clad during a pulse whichgives a peak adiabatic fuel temperature of 100000 is estimated to be 470°C. Further analysis (SAR Section4.5.4.1.2), shows that the internal pressure for both Zr.65 (at 115000) and Zr1.7 (at 1100°C) increases to apeak value at about 0.3 sec, at which time the pressure is about one-fifth of the equilibrium value or about400 psi (a stress of 14,700 psi). The yield strength of the cladding at 500°C is about 59,000 psi.Calculations for step increases in power to peak ZrH1.65 fuel temperature greater than 115000, over a 200°C range,show that the time to reach the peak pressure and the fraction of equilibrium pressure value achieved wereapproximately the same as for the 115000 case. Similar results were found for fuel with ZrH1 .7. Measurements ofhydrogen pressure in TRIGA fuel elements during transient operations have been made and compared with theresults of analysis similar to that used to make the above prediction. These measurements indicate that in a pulsewhere the maximum temperature in the fuel was greater than 100000, the pressure (ZrH1.65) was only about 6% ofthe equilibrium value evaluated at the peak temperature. Calculations of the pressure gave values about three timesgreater than the measured values. The analysis gives strong indications that the cladding will not rupture if fueltemperatures are never greater than 120000 to 125000, providing the cladding temperature is less than 5000C. For fuel with ZrH1.7,a conservative safety limit is 110000. As a result, at this safety limit temperature, the classpressure is a factor of 4 lower than would be necessary for cladding failure.2.2 Limiting Safety System Setting.q 2.2.1 Fuel Temperature. Applicability -This specification applies to the protective action for the reactor fuel elementtemperature. Obiective -The objective is to prevent the fuel element temperature safety limit from being reached.Specification -The limiting safety system setting shall be 75000 (operationally this may be set moreconservatively) as measured in an instrumented fuel element. One instrumented element shall belocated in the analyzed peak power location of the reactor operational core.Basis -For steady-state operation of the reactor, the limiting safety system setting is a temperature which, if exceeded, shall cause a reactor scram to be initiated preventing the safety limit from beingexceeded. A setting of 75000 provides a safety margin at the point of the measurement of at least13700 for standard TRIGA fuel elements in any condition of operation. A part of the safety marginis used to account for the difference between the true and measured temperatures resulting fromthe actual location of the thermocouple. If the thermocouple element is located in the hottestposition in the core, the difference between the true and measured temperatures will be only a fewdegrees since the thermocouple junction is near the center and mid-plane of the fuel element. Forpulse operation of the reactor, the same limiting safety system setting applies.

However, thetemperature channel will have no effect on limiting

the peak power generated because of itsAmendment No. 46 0 !relatively long time constant (seconds) as compared with the width of the pulse (milliseconds).

Inthis mode, however, the temperature trip will act to limit the energy release after the pulse if thetransient rod should not reinsert and the fuel temperature continues to increase. 3.0 Limiting Conditions For Operation 3.1 Reactor Core Parameters 3.1.1 .Steady-State Operation Applicability -This specification applies to the maximum reactor power attained during steady-state operation. Objective -The objective is to assure that the reactor safety limit (fuel temperature) is not exceeded, and to provide for a setpoint for the high flux limiting safety systems, so that automatic protective action will prevent the safety limit from being reached during steady-state operation. Specification -The nominal reactor steady-state power shall not exceed 2.0 MW. The automatic scram setpoints for the reactor power level safety channels shall be set at 2.2 MW or less. For thepurpose of testing the reactor steady-state power level scram, the power shall not exceed 2.3 MW.Basis -Operational experience and thermal-hydraulic calculations demonstrate that UCD/MNRCTRIGA fuel elements may be safely operated at power levels up to 2.3 MW with natural convection cooling. (SAR Chapter 4, Section 4.6.2.)3.1.2 Pulse or Square Wave Operation Applicability -This specification applies to the peak temperature generated in the fuel as the resultof a step insertion of reactivity. Obiective -The objective is to assure that the fuel temperature safety limit will not be exceeded. Specification -a. For the pulse mode of operation, the maximum insertion of reactivity shall be 1.23% Ak/k ($1.75);b. For the square wave mode of operation, the maximum insertion of reactivity shall be 0.63% Ak/k($0.90).Basis -Standard TRIGA fuel is fabricated with a nominal hydrogen to zirconium ratio of 1.6 to 1.7.This yields delta phase zirconium hydride which has a high creep strength and undergoes no phasechanges at temperatures in excess of 1000C. However, after extensive steady state operation attwo (2) MW the hydrogen will redistribute due to migration from the central high temperature regionsof the fuel to the cooler outer regions. When the fuel is pulsed, the instantaneous temperature distribution is such that the highest values occur at the radial edge of the fuel. The highertemperatures in the outer regions occur in fuel with a hydrogen to zirconium ratio that has nowincreased above the nominal value. This produces hydrogen gas pressures considerably in excessof that expected. If the pulse insertion is such that the temperature of the fuel exceeds about 8750C,then the pressure may be sufficient to cause expansion of microscopic holes in the fuel that growwith each pulse. Analysis (SAR Chapter 13, Section 13.2.2.2.1), shows that the limiting pulse, forthe worst case conditions, is 1.34% Ak/k ($1.92). Therefore, the 1.23% Ak/k ($1.75) limit is belowthe worse case reactivity insertion accident limit. The $0.90 square wave step insertion limit is alsowell below the worse case reactivity insertion accident limit.Amendment No. 47 0Basis -a. The apparent condition of the control rod assemblies shall provide assurance that the rods shallcontinue to perform reliably as designed.

b. This assures that the reactor shall shut down promptly when a scram signal is initiated (SARChapter 13, Section 13.2.2.2.2).

3.2.2 Reactor Instrumentation Applicability -This specification applies to the information which shall be available to the reactoroperator during reactor operations. Objective -The objective is to require that sufficient information is available to the operator to assuresafe operation of the reactor.Specification -The reactor shall not be operated unless the channels described in Table 3.2.2 areoperable and the information is displayed on the reactor console.Table 3.2.2Required Reactor Instrumentation (Minimum Number Operable) Measuring ChannelSteadyStateSquareWaveChannelFunctionSurveillance Requirements* Pulsea. Reactor PowerLevel SafetyChannelb. Linear PowerChannelc. Log PowerChanneld. Fuel Temperature Channele. Pulse Channel201100222Scram at 2.2MW or lessAutomatic Power ControlStartupControlFuelTemperature MeasuresPulse NV & NVTD,M,AD,M,AD,M,ADM,A2IPA(*) Where: D -Channel check during each day's operation M -Channel test monthlyA -Channel calibration annuallyP -Channel test prior to pulsing operation Basis -a. Table 3.2.2. The two reactor power level safety channels assure that the reactor power level isproperly monitored and indicated in the reactor control room (SAR Chapter 7, Sections 7.1.2 &7.1.2.2). Amendment No. 49

• 03.3 Reactor Coolant SystemsApplicability

-These specifications apply to the operation of the reactor water measuring systems.Obiective -The objective is to assure that adequate cooling is provided to maintain fuel temperatures belowthe safety limit, and that the water quality remains high to prevent damage to the reactor fuel.Specification -The reactor shall not be operated unless the systems and instrumentation channelsdescribed in Table 3.3 are operable, and the information is displayed locally or in the control room.Table 3.3REQUIRED WATER SYSTEMS AND INSTRUMENTATION Measuring Channel/System

a. Primary CoolantCore InletTemperature Monitorb. Reactor TankLow Water.Monitorc. Purification**

Inlet Conduc-tivity Monitord. Emergency CoreCooling SystemMinimumNumberOperableSurveillance Requirements* Function: Channel/System For operation of thereactor at 1.5 MW orhigher, alarms on highheat exchanger outlettemperature of 45°C(1 130F)Alarms if water leveldrops below a depth of23 feet in the reactor tankAlarms if the primarycoolant water conductivity is greater than5 micromhos/cm For operation of the reactorat 1.5MW or higher, provideswater to cool fuel in the eventof a Loss of Coolant Accidentfor a minimum of 3.7 hoursat 20 gpm from an appropriate nozzleD,Q,AMD,M,SD,S(*) Where: D -channel check during each day's operation A -channel calibration annuallyQ -channel test quarterly S -channel calibration semiannually M -channel test monthly(**) The purification inlet conductivity monitor can be out-of-service for no more than 3 hours before the reactor shallbe shutdown. Amendment No. 413 Basis -.a. Table 3.3. The primary coolant core inlet temperature alarm assures that large power fluctuations will notoccur (SAR Chapter 4, Section 4.6.2).b. Table 3.3. The minimum height of 23 ft. of water above the reactor tank bottom guarantees that there issufficient water for effective cooling of the fuel and that the radiation levels at the top of the reactor tank arewithin acceptable limits. The reactor tank water level monitor alarms if the water level drops below a heightof 23 ft. (7.01 m) above the tank bottom (SAR Chapter 11, Section 11.1.5.1 ).c. Table 3.3. Maintaining the primary coolant water conductivity below 5 micromhos/cm averaged over aweek will minimize the activation of water impurities and also the corrosion of the reactor structure.

d. Table 3.3. This system will mitigate the Loss of Coolant Accident event analyzed in the SAR Chapter 13,Section 13.2.3.4 Reactor Room Exhaust SystemApplicability

-These specifications apply to the operation of the reactor room exhaust system.Obiective -The objectives of this specification are as follows:a. To reduce concentrations of airborne radioactive material in the reactor room, and maintain the reactorroom pressure negative with respect to surrounding areas.b. To assure continuous air flow through the reactor room in the event of a Loss of Coolant Accident. Specification -a. The reactor shall not be operated unless the reactor room exhaust system is in operation and thepressure in the reactor room is negative relative to surrounding areas.b. The reactor room exhaust system shall be operable within one half hour of the onset of a Loss of CoolantAccident. Basis -Operation of the reactor room exhaust system assures that:a. Concentrations of airborne radioactive material in the reactor room and in air leaving the reactor room willbe reduced due to mixing with exhaust system air (SAR Chapter 9, Section 9.5.1). Pressure in the reactorroom will be negative relative to surrounding areas due to air flow patterns

created, by the reactor roomexhaust system (SAR Chapter 9, Section 6.5.1).b. There will be a timely, adequate and continuous air flow through the reactor room to keep the fueltemperature below the safety limit in the event of a Loss of Coolant Accident.

3.5 This section intentionally left blank.3.6 This section intentionally left blank.3.7 Reactor Radiation Monitorinq Systems3.7.1 Monitorinq SystemsApplicability -This specification applies to the information which shall be available to the reactoroperator during reactor operation. Amendment No. 414 0Objective -The objective is to require that sufficient information regarding radiation levels andradioactive effluents is available to the reactor operator to assure safe operation of the reactor..Specification -The reactor shall not be operated unless the channels described in Table 3.7.1 areoperable, the readings are below the alarm setpoints, and the information is displayed in the controlroom. The stack and reactor room CAMS shall not be shutdown at the same time during reactoroperation. Table 3.7.1REQUIRED RADIATION MONITORING iNSTRUMENTATION Measuring Equipment MinimumNumberOperable** ChannelFunctionSurveillance Requirements*

a. FacilityStack Monitorb. Reactor RoomRadiation Monitorc. Purification System Radia-tion Monitord. Reactor RoomContinuous Air MonitorI111Monitors Argon-41 andradioactive particu-lates, and alarmsMonitors the radiation level in the reactorroom and alarmsMonitors radiation level at the demineral-izer station and alarmsMonitors air from thereactor room for parti-culate and gaseousradioactivity and alarmsD,W,AD,W,AD,W,AD,W,A(*) Where: D -channel check during each day's operation A -channel calibration annuallyW -channel test(**) monitors may be placed out-of-service for up to 2 hours for calibration and maintenance.

During this out-of-service time, no experiment or maintenance activities shall be conducted which could result in alarm conditions (e.g., airborne releases or high radiation levels)Basis -a. Table 3.7.1. The facility stack monitor provides information to operating personnel regarding the releaseof radioactive material to the environment (SAR Chapter 11, Section 11.1.1.1.4). The alarm setpoint on thefacility stack monitor is set to limit Argon-41 concentrations to less than 10 CFR Part 20, Appendix B,Table 2, Column 1 values (averaged over one year) for unrestricted locations outside the operations area.b. Table 3.7.1. The reactor room radiation monitor provides information regarding radiation levels in thereactor room during reactor operation (SAR Chapter 11, Section 11.1.5.1), to limit occupational radiation exposure to less than 10 CFR 20 limits.c. Table 3.7.1. The radiation monitor located next to the purification system resin canisters providesinformation regarding radioactivity in the primary system cooling water (SAR Chapter 11, Section 11.1.5.4.2) Amendment No. 415 and allows assessment of radiation levels in the area to ensure that personnel radiation doses will be below10 CER Part 20 limits.d. Table 3.7.1. The reactor room continuous air monitor provides information regarding airborne radioactivity in the reactor room, (SAR Chapter 11, Sections 11.1.1.1.2 & 11.1.1.1.5), to ensure that occupational exposure to airborne radioactivity will remain below the 10 CFR Part 20 limits.3.7.2 Effluents -Arqon-41 Discharqe Limit.Applicability -This specification applies to the concentration of Argon-41 that may be discharged from the UCD/MNRC reactor facility. Obiective -The objective is to ensure that the health and safety of the public is not endangered bythe discharge of Argon-41 from the UCD/MNRC reactor facility. Specification -The annual average unrestricted area concentration of Argon-41 due to releases ofthis radionuclide from the UCD/MNRC, and the corresponding annual radiation dose from Argon-41in the unrestricted area shall not exceed the applicable levels in 10 CER Part 20.Basis -The annual average concentration limit for Argon-41 in air. in the unrestricted area isspecified in Appendix B, Table 2, Column 1 of 10 CER Part 20. 10 CER 20.1301 specifies doselimitations in the unrestricted area. 10 CFR 20.1101 specifies a constraint on air emissions ofradioactive material to the environment. The SAR Chapter 11, Section 11.1.1.1.4 estimates that theroutine Argon-41 releases and the corresponding doses in the unrestricted area will be below theselimits.3.8 Experiments 3.8.1 Reactivity Limits.Aoplicability -This specification applies to the reactivity limits on experiments installed in specificreactor experiment facilities. Obiective -The objective is to assure control of the reactor during the irradiation or handling ofexperiments in the specifically designated reactor experiment facilities. Specification -The reactor shall not be operated unless the following conditions governing experiments exist:a. The absolute reactivity worth of any single moveable experiment in the pneumatic transfer tube,the central irradiation

facility, the central irradiation fixture 1 (ClF-1),

or any other in-core or in-tankirradiation

facility, shall be less than $1.00 (0.7% Ak/k), except for the automated central irradiation facility (ACIF) (See 3.8.1.c below).b. The absolute reactivity worth of any single secured experiment positioned in a reactor in-core orin-tank irradiation facility shall be less than the maximum allowed pulse ($1.75) (1.23% Ak/k).c. The absolute total reactivity worth of any single experiment or of all experiments collectively positioned in the AClF shall be less than the maximum allowed pulse ($1.75) (1.23% Ak/k).d. The absolute total reactivity Of all experiments positioned in the pneumatic transfer tube, and inany other reactor in-core and in-tank irradiation facilities at any given time shall be less than onedollar and ninety-two cents ($1.92) (1.34% Ak/k), including the potential reactivity which might resultfrom malfunction,

flooding, voiding, or removal and insertion of the experiments.

Amendment No. 416 Basis -a. A limitation of less than one dollar ($1.00) (0.7% Ak/k) on the reactivity worth of a single movableexperiment positioned in the pneumatic transfer tube, the central irradiation facility (SAR, Chapter10, Section 10.4.1), the central irradiation fixture-i (ClF-1) (SAR Chapter 10, Section 10.4.1), orany other in-core or in-tank irradiation

facility, will assure that the pulse limit of $1.75 is notexceeded (SAR Chapter 13, Section 13.2.2.2.1).

In addition, limiting the worth of each movableexperiment to less than $1.00 will assure that the additional increase in transient power andtemperature will be slow enough so that the fuel temperature scram will be effective (SAR Chapter13, Section 13.2.2.2.1).

b. The absolute worst event which may be considered in conjunction with a single securedexperiment is its sudden accidental or unplanned removal while the reactor is operating.

For suchan event, the reactivity limit for fixed experiments ($1.75) would result in a reactivity increase lessthan the $1.92 pulse reactivity insertion needed to reach the fuel temperature safety limit (SARChapter 13, Section 13.2.2.2.1).

c. A reactivity limit of less than $1.75 for any single experiment or for all experiments collectively positioned in the sample can of the automated central irradiation facility (ACIF) (SAR Chapter 10,Section 10.4.2) is based on the pulsing reactivity insertion limit (Technical Specification 3.1.2) (SARChapter 13, Section 13.2.2.2.1) and on the design of the ACIF, which allows control over thepositioning of samples into and out of the central core region in a manner identical in form, fit, andfunction to a control rod.d. it is conservatively assumed that simultaneous removal of all experiments positioned in thepneumatic transfer tube, and in any other reactor in-core and in-tank irradiation facilities at anygiven time shall be less than the maximum reactivity insertion limit of $1.92. The SAR Chapter 13,Section 13.2.2.2.1 indicates that a pulse reactivity insertion of $1.92 would be needed to reach thefuel temperature safety limit.3.8.2 Materials LimitApplicability

-This specification applies to experiments installed in reactor experiment facilities. Obiective -The objective is to prevent damage to the reactor or significant releases of radioactivity by limiting material quantity and the radioactive material inventory of the experiment. Specification -The reactor shall not be operated unless the following conditions governing experiment materials exist:a. Experiments containing materials corrosive to reactor components, compounds highly reactivewith water, potentially explosive materials, and liquid fissionable materials shall be appropriately encapsulated.

b. Each fueled experiment shall be controlled such that the total inventory of iodine isotopes 131through 135 in the experiment is no greater than 1.5 curies and the maximum strontium inventory is no greater than 5 millicuries.

c. Each experiment in the 1-125 production facility shall be controlled such that the total inventory of 1-125 dispensed or stored in the reactor room glove box shall not exceed 20 curies.d. Each experiment in the 1-125 production facility shall be controlled such that the total inventory of 1-125 being processed at any one time in the reactor room fume hood shall not exceed 200millicuries.

An additional 800 millicuries of 1-125 in sealed storage containers may also be presentin the reactor room fume hood.Amendment No. 417 Se. Explosive materials in quantities greater than 25 milligrams of TNT equivalent shall not beirradiated in the reactor tank. Explosive materials in quantities of 25 milligrams of TNT equivalent or less may be irradiated provided the pressure produced upon detonation of the explosive hasbeen calculated and/or experimentally demonstrated to be less than the design pressure of thecontainer.

f. Explosive materials in quantities of three (3) pounds of TNT equivalent or less may be irradiated in any radiography bay. The irradiation of explosives in any bay is limited to those assemblies wherea safety analysis has been performed that shows that there is no damage to the reactor safetysystems upon detonation (SAR Chapter 13, Section 13.2.6.2).

Basis -a. Appropriate encapsulation is required to lessen the experimental hazards of some types ofmaterials.

b. The 1.5 curies limitation on iodine 131 through 135 assures that in the event of failure of a fueledexperiment leading to total release of the iodine, occupational doses and doses to members of thegeneral public in the unrestricted areas shall be within the limits in 10 CFR 20 (SAR Chapter 13,Section 13.2.6.2).

c&d. Limiting the total 1-125 inventory to twenty (20.0) curies in the reactor room glove box and toone (1.0) curie in the reactor room fume hood assures that, if these inventories of 1-125 are totallyreleased into their respective containments, occupational doses and doses to members of thegeneral public in the unrestricted areas shall be within the limits of 10 CFR 20 (SAR Chapter 13,Section 13.2.6.2).

e. This specification is intended to prevent damage to vital equipment by restricting the quantity ofexplosive materials within the reactor tank (SAR Chapter 13, Section 13.2.6.2).

f. The failure of an experiment involving the irradiation of 3 lbs TNT equivalent or less in anyradiography bay external to the reactor tank will not result in damage to the reactor controls or thereactor tank. Safety Analyses have been performed (SAR Chapter 13, Section 13.2.6.2) whichshow that up to six (6) pounds of TNT equivalent can be safely irradiated in any radiogaphy bay.Therefore, the three (3) pound limit gives a safety margin of two (2).3.8.3 Failure and Malfunctions Applicability

-This specification applies to experiments installed in reactor experiment facilities. Objective -The objective is to prevent damage to the reactor or significant releases of radioactive materials in the event of an experiment failure.Specification -a. All experiment materials which could off-gas,

sublime, volatilize, or produce aerosols under:(1) normal operating conditions of the experiment or the reactor,(2) credible accident conditions in the reactor, or(3) where the possibility exists that the failure of an experiment could release radioactive gases or aerosols into the reactor building or into the unrestricted area, the quantity andtype of material in the experiment shall be limited such that the airborne radioactivity in thereactor room will not result in exceeding the applicable dose limits in 10 CFR Part 20 in theunrestricted area, assuming 100% of the gases or aerosols escapes.Amendment No. 418 0b. In calculations pursuant to (a) above, the following assumptions shall be used:(1) If the effluent from an experiment facility exhausts through a stack which is closed onhigh radiation levels, at least 10% of the gaseous activity or aerosols produced will escape.(2) If the effluent from an experiment facility exhausts through a filter installation designedfor greater than 99% efficiency for 0.3 micron and larger particles, at least 10% of these willescape.(3) For materials whose boiling point is above 130°C and where vapors formed by boilingthis material can escape only through an undistributed column of water above the core, atleast 10% of these vapors can escape.c. If a capsule fails and releases material which could damage the reactor fuel or structure bycorrosion or other means, an evaluation shall be made to determine the need for corrective action.Inspection and any corrective action taken shall be reviewed by the UCD/MNRC Director or hisdesignated alternate and determined to be satisfactory before operation of the reactor is resumed.Basis -a. This specification is intended to reduce the likelihood that airborne radioactivity in the reactorroom or the unrestricted area will result in excee~ding the applicable dose limits in 10 CFR Part 20.b. These assumptions are used to evaluate the potential airborne radioactivity release due to anexperiment failure (SAR Chapter 13, Section 13.2.6.2).

c. Normal operation of the reactor with damaged reactor fuel or structural damage is prohibited toavoid release of fission products.

Potential damage to reactor fuel or structure shall be brought tothe attention of the UCD/MNRC Director or his designated alternate for review to assure safeoperation of the reactor (SAR Chapter 13, Section 13.2.6.2). 4.0 Surveillance Requirements General. The surveillance frequencies denoted herein are based on continuing operation of the reactor.Surveillance activities scheduled to occur during an operating cycle which can not be performed with the reactoroperating may be deferred to the end of the operating cycle. If the reactor is not operated for a reasonable time, areactor system or measuring channel surveillance requirement may be waived during the associated time period.Prior to reactor system or measuring channel operation, the surveillance shall be performed for each reactor systemor measuring channel for which surveillance was waived. A reactor system or measuring channel shall not beconsidered operable until it is successfully tested.4.1 Reactor Core Parameters 4.1.1 Steady State Operation Applicability -This specification applies to the surveillance requirement for the power levelmonitoring channels. Obiective -The objective is to verify that the maximum power level of the reactor does not exceedthe authorized limit.Amendment No. 419 0 0Discovery of noncompliance with Technical Specifications 4.3.a-c shall limit operations to that required to performthe surveillance. Noncompliance with Technical Specification 4.3.d shall limit operations to less than 1.5 MW.Basis -a. A channel test quarterly assures the water temperature monitoring system responds correctly to an inputsignal. A channel check during each day's operation assures the channel is operable. A channel calibration annually assures the monitoring system reads properly.

b. A channel test monthly assures that the low water level monitoring system responds correctly to an inputsignal.c. A channel test monthly assures that the purification inlet conductivity monitors respond correctly to aninput signal. A channel check during each day's operation assures that the channel is operable.

A channelcalibration semiannually assures the conductivity monitoring system reads properly.

d. A channel check prior to operation assures that the emergency core cooling system is operable for powerlevels above 1.5 MW. A channel calibration semiannually assures that the Emergency Core Cooling Systemperforms as required for power levels above 1.5 MW.4.4 Reactor Room Exhaust SystemApplicability

-This specification applies to the surveillance requirements for the reactor roomexhaust system.Obiective -The objective is to assure that the reactor room exhaust system is operating properly. Specification -The reactor room exhaust system shall have a channel check during each day'soperation. Discovery of noncompliance with this specification shall limit operations to that required to perform the surveillance. Basis -A channel check during each day's operation of the reactor room exhaust system shall verifythat the exhaust system is maintaining a negative pressure in the reactor room relative to thesurrounding facility areas.4.5 This section intentionally left blank4.6 This section intentionally left blank.4.7 Reactor Radiation Monitorinq SystemsApplicability -This specification applies to the surveillance requirements for the reactor radiation monitoring systems.Obiective -The objective is to assure that the radiation monitoring equipment is operating properly. Specification -a. The facility stack monitor shall have the following: (1) A channel checkduring each day's operation. (2) A channel test weekly.Amendment No. 425 0.!(3) A channel calibration annually.

b. The reactor room radiation monitor shall have the following:

(1) A channel check during each day's operation. (2) A channel test weekly.(3) A channel calibration annually.

c. The purification system radiation monitor shall have the .following:

(1) A channel check during each day's operation. (2) A channel test weekly.(3) A channel calibration annually.

d. The reactor room Continuous Air Monitor (CAM) shall have the following:

(1) A channel check during each day's operation. (2) A channel test weekly.(3) A channel calibration annually. Discovery of noncompliance with Technical Specifications 4.7.a-d shall limit operations to that required to performthe surveillance. Basis -a. A channel check of the facility stack monitor system during each day's operation will assure themonitor is operable. A channel test weekly will assure that the system responds correctly to aknown source. A channel calibration annually will assure that the monitor reads correctly.

b. A channel check of the reactor room radiation monitor during each day's operation will assurethat the monitor is operable.

A channel test weekly will ensure that the system responds to a knownsource. A channel calibration of the monitor annually will assure that the monitor reads correctly.

c. A channel check of the purification system radiation monitor during each day's operation assuresthat the monitor is operable.

A channel test weekly will ensure that the system responds to a knownsource. A channel calibration of the monitor annually will assure that the monitor reads correctly.

d. A channel check of the reactor room Continuous Air Monitor (CAM) during each day's operation will assure that the CAM is operable.

A channel test weekly will assure that the CAM respondscorrectly to a known source. A channel calibration annually will assure that the CAM readscorrectly.

4.8 Experiments

Applicability -This specification applies to the surveillance requirements for experiments installed in any UCD/MNRC reactor experiment facility. Amendment No. 426 0Objective -The objective is to prevent the conduct of experiments or irradiations which may damagethe reactor or release excessive amounts of radioactive materials as a result of experimental failure.Specification -a. A new experiment shall not be installed in any UCD/MNRC reactor experiment facility until awritten safety analysis has been performed and reviewed by the UCD/MNRC

Director, or hisdesignee, to establish compliance with the Limitations on Experiments, (Technical Specifications Section 3.8) and 10 CFR 50.59.b. All experiments performed at the UCD/MNRC shall meet the conditions of an approved FacilityUse Authorization.

Facility Use Authorizations and experiments carried out under theseauthorizations shall be reviewed and approved in accordance with the Utilization of the (UCD)McClellan Nuclear Radiation Center Research Reactor Facility Document (MNRC-0027-DOC). Anexperiment classified as an approved experiment shall not be placed in any UCD/MNRC experiment

facility, until it has been reviewed for compliance with the approved experiment and Facility UseAuthorization by the Reactor Manager and the Health Physics Manager, or their designated alternates.

c. The reactivity worth of any experiment installed in the pneumatic transfer tube, or in any otherUCD/MNRC reactor in-core or in-tank irradiation facility shall be estimated or measured, asappropriate, before reactor operation with said experiment.

Whenever a measurement is done itshall be done at ambient conditions.

d. Experiments shall be identified and a log or other record maintained while experiments are in anyUCD/MNRC reactor experiment facility.

Basis -a & b. Experience at most TRIGA reactor facilities verifies the impo'rtance of reactor staff and safetycommittee reviews of proposed experiments.

c. Measurement of the reactivity worth of an experiment, or estimation of the reactivity worth basedon previous or similar measurements, shall verify that the experiment is within authorized reactivity limits.d. Maintaining a log of experiments while in UCD/MNRC reactor experiment facilities will facilitate maintaining surveillance over such experiments.

5.0 Desigqn Features5.1 Site and Facility Description. 5.1.1 Sit..eeApplicability -This specification applies to the UCD/MNRC site location and specific facility designfeatures. Objective -The objective is to specify those features related to the Safety Analysis evaluation. Specification -Amendment No. 427 0a. The site location is situated approximately 8 miles (13 kin) north-by-northeast of downtownSacramento, California. The site of the UCD/MNRC facility is about 3000 ft. (0.6 mi or 0.9 kin) westof Watt Avenue, and 4500 ft. (0.9 mi or 1.4 kin) south of E Street.b. The restricted area is that area inside the fence surrounding the reactor building. Theunrestricted area is that area outside the fence surrounding the reactor building.

c. The TRIGA reactor is located in Building 258, Room 201 of the UCD/MNRC.

This building hasbeen designed with special safety features.

d. The core is below ground level in a water filled tank and surrounded by a concrete shield.Basis -a. Information on the surrounding population, the hydrology, seismology, and climatography of thesite has been presented in Chapter 2 of the Safety Analysis Report.b. The restricted area is controlled by the UCDIMNRC Director.

c. The room enclosing the reactor has been designed with systems related to the safe operation ofthe facility.

d. The below grade core design is to negate the consequences of an aircraft hitting the reactorbuilding.

This accident was analyzed in Chapter 13 of the Safety Analysis Report, and found to bebeyond a credible accident scenario. 5.1.2 Facility ExhaustApplicability -This specification applies to the facility which houses the reactor.Obiective -The objective is to assure that provisions are made to restrict the amount of radioactivity released into the environment, or during a Loss of Coolant Accident, the system is to assure properremoval of heat from the reactor room.Specification-

a. The UCD/MNRC reactor facility shall be equipped with a system designed to filter and exhaust airfrom the UCD/MNRC facility.

The system shall have an exhaust stack height of a minimum of18.2m (60 feet) above ground level.b. Manually activated shutdown controls for the exhaust system shall be located in the reactorcontrol room.Basis -The UCD/MNRC facility exhaust system is designed such that the reactor room shall bemaintained at a negative pressure with respect to the surrounding areas. The free air volume withinthe UCD/MNRC facility is confined to the facility when there is a shutdown of the exhaust system.Controls for startup, filtering, and normal operation of the exhaust system are located in the reactorcontrol room. Proper handling of airborne radioactive materials (in emergency situations) can beconducted from the reactor control room with a minimum of exposure to operating personnel. 5.2 Reactor Coolant SystemApplicability -This specification applies to the reactor coolant system.Amendment No. 428 0Objective -The objective is to assure that adequate water is available for cooling and shielding duringnormal reactor operation or during a Loss of Coolant Accident. Specification -a. During normal reactor operation the reactor core shall be cooled by a natural convection flow of water.b. The reactor tank water level alarm shall activate if the water level in the reactor tank drops below a depthof 23 ft.c. For operations at 1.5 MW or higher during a Loss of Coolant Accident the reactor core shall be cooled fora minimum of 3.7 hours at 20 gpm by a source of water from the Emergency Core Cooling System.Basis -a. The SAR Chapter 4, Section 4.6, Table 4-19, shows that fuel temperature limit of 930°C will not beexceeded under natural convection flow conditions.

b. A reactor tank water low level alarm sounds when the water level drops significantly.

This alarmannunc~iates in the reactor control room and at a 24 hour monitored location so that appropriate corrective action can be taken to restore water for cooling and shielding.

c. The SAR Chapter 13, Section 13.2, analyzes the requirements for cooling of the reactor fuel and showsthat the fuel safety limit is not exceeded under Loss of Coolant Accident conditions during this water cooling.5.3 Reactor Core and Fuel5.3.1 Reactor CoreApplicability

-This specification applies to the configuration of the fuel.Objective -The objective is to assure that provisions are made to restrict the arrangement of fuelelements so as to provide assurance that excessive power densities will not be produced. Specification -Foroperation at 0.5 MW or greater, the reactor core shall be an arrangement of 96or more fuel elements to include fuel followed control rods. Below 0.5 MW there is no minimumrequired number of fuel elements. In a mixed 20/20, 30/20 and 8.5/20 fuel loading (SAR Chapter 4,Section 4.5.5.6): Mix J Core and Other Variations (1) No fuel shall be loaded into Hex Rings A or B.(2) A fuel followed control rod located in an 8.5 wt% environment shall contain 8.5 wt% fuel.20E Core and Other Variations (1) No fuel shall be loaded into Hex Rings A or B.(2) Fuel followed control rods may contain either 8.5 wt% or 20 wt% fuel.(3) Variations to the 20E core having 20 wt% fuel in Hex Ring C requires the 20 wt% fuel to beloaded into corner positions only, and graphite dummy elements in the flat positions. Theperformance of fuel temperature measurements shall apply to variations to the as-analyzed 20Ecore configurations. Amendment No. 429 0 S308 Core and Other Variations (1) No fuel shall be loaded into Hex Rings A or B.(2) The only fuel types allowed are 20/20 and 30/20.(3) 20/20 fuel may be used in any position in Hex Rings C through G.(4) 30/20 fuel may be used in any position in Hex Rings 0 through G but not in Hex Ring C.(5) An analysis of any irradiation facility installed in the central cavity of this core shall be donebefore it is used with this core.Basis -In order to meet the power density requirements discussed in the SAR Chapter 4, Section4.5.5.6, no less than 96 fuel elements including fuel followed control rods and the above loadingrestrictions will be allowed in an operational 0.5 MW or greater core. Specifications for the 20E coreand for the 308 core allow for variations of the as-analyzed core with the condition that temperature limits are being maintained (SAR Chapter 4, Section 4.5.5.6 and Argonne National Laboratory Report AN L/ED 97-54).5.3.2 Reactor FuelApplicability -These specifications apply to the fuel elements used in the reactor core.Obiective -The objective is to assure that the fuel elements are of such design and fabricated insuch a manner as to permit their use with a high degree of reliability with respect to their physicaland nuclear characteristics. Specification -The individual unirradiated TRIGA fuel elements shall have the following characteristics:

a. Uranium content:

8.5, 20 or 30 wt % uranium enriched nominally to less than 20% U-235.b. Hydrogen to zirconium atom ratio (in the ZrHx): 1.60 to 1.70 (I.65+/- 0.05).c. Cladding: stainless steel, nominal 0.5mm (0.020 inch) thick.Basis -a. The design basis of a TRIGA core loaded with TRIGA fuel demonstrates that limiting operation to2.3 megawatts steady state or to a 36 megawatt-sec pulse assures an ample margin of safetybetween the maximum temperature generated in the fuel and the safety limit for fuel temperature. The fuel temperatures are not expected to exceed 630°C during any condition of normal operation.

b. Analysis shows that the stress in a TRIGA fuel element, H/Zr ratios between 1.6 and 1.7, is equalto the clad yield strength when both fuel and cladding temperature are at the safety limit 930°C.Since the fuel temperatures are not expected to exceed 630°C during any condition of normaloperation, there is a margin between the fuel element clad stress and its ultimate strength.

c. Safety margins in the fuel element design and fabrication allow for normal mill tolerances ofpurchased materials.

Amendment No. 430 0 05.3.3 Control Rods and Control Rod DrivesApplicability -This specification applies to the control rods and control rod drives used in the reactorcore.Obiective -The objective is to assure the control rods and control rod drives are of Such a design asto permit their use with a high degree of reliability with respect to their physical,

nuclear, andmechanical characteristics.

Specification -a. All control rods shall have scram capability and contain a neutron poison such as stainless steel,borated graphite, B4C powder, or boron and its compounds in solid form. The shim and regulating rods shall have fuel followers sealed in stainless steel. The transient rod shall have an air filledfollower and be sealed in an aluminum tube.b. The control rod drives shall be the standard GA rack and pinion type with an electromagnet andarmature attached. Basis -a. The neutron poison requirements for the control rods are satisfied by using stainless steel,neutron absorbing borated graphite, B4C powder, or boron and its compounds. These materials shall be contained in a suitable clad material such as stainless steel or aluminum to assuremechanical stability during movement and to isolate the neutron poison from the tank waterenvironment. Scram capabilities are provided for rapid insertion of the control rods.b. The standard GA TRIGA control rod drive meets the requirements for driving the control rods atthe proper speeds, and the electromagnet and armature provide the requirements for rapid insertion capability. These drives have been tested and proven in many TRIGA reactors. 5.4 Fissionable Material StoraqeApplicability -This specification applies to the storage of reactor fuel at a time when it is not in the reactorcore.Obiective -The objective is to assure that the fuel which is being stored will not become critical and will notreach an unsafe temperature. Specification -a. All fuel elements not in the reactor core shall be stored (wet or dry) in a geometrical array where the keffis less than 0.9 for all conditions of moderation.

b. Irradiated fuel elements shall be stored in an array which shall permit sufficient natural convection coolingby water or air such that the fuel element temperature shall not exceed the safety limit.Basis -The limits imposed by Technical Specifications 5.4.a and 5.4.b assure safe storage.6.0 Administrative Controls6.1 Orqanization.

The Vice Chancellor for Research shall be the licensee for the UCD/MNRC. TheUCD/MNRC facility shall be under the direct control of the UCD/MNRC Director or a licensed senior reactoroperator (SRO) designated by the UCD/MNRC Director to be in direct control. The UCD/MNRC Directorshall be accountable to the Vice Chancellor of the Office of Research for the safe operation andmaintenance of the reactor and its associated equipment. Amendment No. 431 06.1.1 Structure. The management for operation of the UCD/MNRC facility shall consist of theorganizational structure as shown in Figure 6.1.6.1.2 Responsibilities. The UCD/MNRC Director shall be accountable to the Vice Chancellor of theOffice of Research for the safe operation and maintenance of the reactor and its associated equipment. The UCD/MNRC

Director, or his designated alternate, shall review and approve allexperiments and procedures prior to their use in the reactor.

Individuals in themanagement organization (e.g., Reactor Manager, Health Physics Manager, etc.) shall beresponsible for implementing UCD/MNRC policies and for operation of the facility, and shall beresponsible for safeguarding the public and facility personnel from undue radiation exposures andfor adhering to the operating license and technical specifications. The Reactor Manager and HealthPhysics Manager report directly to the UCD/MNRC Director. 6.1.3 Staffing6.1.3.1 The minimum staffing when the reactor is not shutdown shall be:a. A reactor operator in the control room;b. A second person in the facility area who can perform prescribed instructions;

c. A senior reactor operator readily available.

The available senior reactoroperator should be within thirty (30) minutes of the facility and reachable bytelephone, and;d. A senior reactor operator shall be present whenever a reactor startup isperformed, fuel is being moved, or experiments are being placed in the reactortank.6.1.3.2 A list of reactor facility personnel by name and telephone number shall be available to the reactor operator in the control room. The list shall include:a. Management personnel.

b. Health Physics personnel.

c. Reactor Operations personnel.

6.1.4 Selection and Training of Personnel. The selection, training and requalification of operations personnel shall meet or exceed the requirements of the American National Standard for Selection andTraining of Personnel for Research Reactors (ANS 15.4). Qualification and requalification of licensedoperators shall be subject to an approved Nuclear Regulatory Commission (NRC) program.6.2 Review. Audit. Recommendation and ApprovalGeneral Policy. Nuclear facilities shall be designed, constructed,

operated, and maintained in sucha manner that facility personnel, the general public, and both university and non-university propertyare not exposed to undue risk. These activities shall be conducted in accordance with applicable regulatory requirements.

The UCD Vice Chancellor of the Office of Research shall institute the above stated policy as thefacility license holder. The Nuclear Safety COmmittee (NSC) has been chartered to assist inmeeting this responsibility by providing timely, objective, and independent

reviews, audits,recommendations and approvals on matters affecting nuclear safety. The following describes thecomposition and conduct of the NSC.Amendment No. 432 06.2.1 NSC Composition and Qualifications.

The UCD/MNRC Director shall appoint theChairperson of the NSC. The NSC Chairperson shall appoint a Nuclear Safety Committee (NSC)of at least five (5) members knowledgeable in fields which relate to nuclear safety. The NSC shallevaluate and review nuclear safety associated with the operation and use of the UCD/MNRC. 6.2.2 NSC Charter and Rules. The NSC shall conduct its review and audit (inspection) functions in accordance with a written charter. This charter shall include provisions for:a. Meeting frequency (The committee shall meet at least semiannually).

b. Voting rules.c. Quorums (For the full committee, a quorum will be at least five (5) members.d. A committee review function and an audit/inspection function.

e. Use of subcommittees.

f. Review, approval and dissemination of meeting minutes.6.2.3 Review Function.

The responsibilities of the NSC, or a designated subcommittee thereof,shall include but are not limited to the following:

a. Review approved experiments utilizing UCD/MNRC nuclear facilities.

b. Review and approve all proposed changes to the facility

license, the Technical Specifications and the Safety Analysis Report, and any new or changed Facility Use Authorizations and proposedClass I modifications, prior to implementing (Class I) modifications, prior to taking action under thepreceding documents or prior to forwarding any of these documents to the Nuclear Regulatory Commission.

c. Review and determine whether a proposed change, test, or experiment would constitute anunreviewed safety question or require a change to the license, to a Facility Use Authorization, or tothe Technical Specifications.

This determination may be in the form of verifying a decision alreadymade by the UCD/MNRC Director.

d. Review reactor operations and operational maintenance, Class I modification

records, and thehealth physics program and associated records for all UCD/MNRC nuclear facilities.

e. Review the periodic updates of the Emergency Plan and Physical Security Plan for UCDIMNRCnuclear facilities.

f. Review and update the NSC Charter every two (2) years.g. Review abnormal performance of facility equipment and operating anomalies.

h. Review all reportable occurrences and all written reports of such occurrences prior to forwarding the final written report to the Nuclear Regulatory Commission.

i. Review the NSC annual audit/inspection of the UCD/MNRC nuclear facilities and any otherinspections of these facilities conducted by other agencies.

Amendment No. 433 06.2.4 AuditlInspection Function. The NSC or a subcommittee

thereof, shall audit/inspect reactoroperations and health physics annually.

The annual audit/inspection shall include, but not be limitedto the following:

a. Inspection of the reactor operations and operational maintenance, Class I modification records,and the health physics program and associated

records, including the ALARA program, for allUCD/MNRC nuclear facilities.

b. Inspection of the physical facilities at the UCD/MNRC.

c. Examination of reportable events at the UCD/MNRC.

d. Determination of the adequacy of UCD/MNRC standard operating procedures.

e. Assessment of the effectiveness of the training and retraining programs at the UCDIMNRC.

f. Determination of the conformance of operations at the UCD/MNRC with the facility's license andTechnical Specifications, and applicable regulations.

g. Assessment of the results of actions taken to correct deficiencies that have occurred in nuclearsafety related equipment, structures,

systems, or methods of operations.

h. Inspection of the currently active Facility Use Authorizations and associated experiments.

i. Inspection of future plans for facility modifications or facility utilization.

j. Assessment of operating abnormalities.

k. Determination of the status of previous NSC recommendations.

6.3 Radiation Safety. The Health Physics Manager shall be responsible for implementation of theUCD/MNRC Radiation Safety Program. The program should use the guidelines of the American NationalStandard for Radiation Protection at Research Reactor Facilities (ANSI/ANS 15.11). The Health PhysicsManager shall report to the UCD/MNRC Director. 6.4 Procedures. Written procedures shall be prepared and approved prior to initiating any of the activities listed in this section. The procedures shall be approved by the UCD/MNRC Director. A periodic review ofprocedures shall be performed and documented in a timely manner by the UCD/MNRC staff to assure thatprocedures are current. Procedures shall be adequate to assure the safe operation of the reactor, but shallnot preclude the use of independent judgment and action should the situation require. Procedures shall bein effect for the following items:6.4.1 Reactor Operations Procedures

a. Startup, operation, and shutdown of the reactor.b. Fuel loading, unloading, and movement within the reactor.c. Control rod removal or replacement.

d. Routine maintenance of the control rod drives and reactor safety and interlock systems or otherroutine maintenance that could have an effect on reactor safety.e. Testing and calibration of reactor instrumentation and controls, control rods and control roddrives.Amendment No. 434

f. Administrative controls for operations, maintenance, and conduct of irradiations and experiments that could affect reactor safety or core reactivity.

g. Implementation of required plans such as emergency and security plans.h. Actions to be taken to correct potential malfunctions of systems, including responses to alarmsand abnormal reactivity changes.6.4.2 Health Physics Procedures

a. Testing and calibration of area radiation

monitors, facility air monitors, laboratory radiation detection

systems, and portable radiation monitoring instrumentation.

b. Working in laboratories and other areas where radioactive materials are used.c. Facility radiation monitoring program including routine and special surveys, personnel monitoring, monitoring and handling of radioactive waste, and sampling and analysis of solid andliquid waste and gaseous effluents released from the facility.

The program shall include amanagement commitment to maintain exposures and releases as low as reasonably achievable (ALARA).d. Monitoring radioactivity in the environment surrounding the facility.

e. Administrative guidelines for the facility radiation protection program to include personnel orientation and training.

f. Receipt of radioactive materials at the facility, and unrestricted release of materials and itemsfrom the facility which may contain induced radioactivity or radioactive contamination.

g. Leak testing of sealed sources containing radioactive materials.

h. Special nuclear material accountability.

i. Transportation of radioactive materials.

Changes to the above procedures shall require approval of the UCD/MNRC Director. All such changes shall bedocumented. 6.5 Experiment Review and Approval. Experiments having similar characteristics are grouped together forreview and approval under specific Facility Use Authorizations. All specific experiments to be performed under the provisions of an approved Facility Use Authorization shall be approved by the UCD/MNRCDirector, or his designated alternate.

a. Approved experiments shall be carried out in accordance with established and approved procedures.

b. Substantive change to a previously approved experiment shall require the same review and approval asa new experiment.

c. Minor changes to an experiment that do not significantly alter the experiment may be approved by asenior reactor operator.

6.6 Required Actions6.6.1 Action to be taken in case of a safety limit violation. In the event of a safety limit violation (fueltemperature), the following action shall be taken:Amendment No. 435 0.Qa. The reactor shall be shut down and reactor operation shall not be resumed until authorized bythe NRC.b. The safety limit violation shall be promptly reported to the UCD/MNRC Director.

c. The safety limit violation shall be reported to the chairman of the NSC and to the NRC by theUCD/MNRC Director.

d. A safety limit violation report shall be prepared.

The report shall describe the following: (1) Applicable circumstances leading to the violation, including when known, the cause andcontributing factors.(2) Effect of the violation upon reactor facility components,

systems, or structures, and onthe health and safety of personnel and the public.(3) Corrective action to be taken to prevent reoccurrence.

e. The safety limit violation report shall be reviewed by the NSC and then be submitted to the NRCwhen authorization is sought to resume operation of the reactor.6.6.2 Actions to be taken for reportable occurrences.

In the event of reportable occurrences, thefollowing actions shall be taken:a. Reactor conditions shall be returned to normal or the reactor shall be shut down. If it isnecessary to shut down the reactor to correct the occurrence, operations shall not be resumedunless authorized by the UCD/MNRC Director or his designated alternate.

b. The occurrence shall be reported to the UCD/MNRC Director or the designated alternate.

TheUCD/MNRC Director shall report the occurrence to the NRC as required by these Technical Specifications or any applicable regulations.

c. Reportable occurrences should be verbally reported to the Chairman of the NSC and the NRCOperations Center within 24 hours of the occurrence.

A written preliminary report shall be sent tothe NRC, Attn: Document Control Desk, I White Flint North, 11555 Rockville Pike, Rockville MD20852, within 14 days of the occurrence. A final written report shall be sent to the above addresswithin 30 days of the occurrence.

d. Reportable occurrences should be reviewed by the NSC prior to forwarding any written reportto the Vice Chancellor of the Office of Research or to the Nuclear Regulatory Commission.

6.7 Reports. All written reports shall be sent within the prescribed interval to the NRC, Attn: DocumentControl Desk, 1 White Flint North, 11555 Rockville Pike, Rockville MD 20852.6.7.10Operating Reports. An annual report covering the activities of the reactor facility during theprevious calendar year shall be submitted within six months following the end of each calendaryear. Each annual report shall include the following information:

a. A brief summary of operating experiences including experiments performed, changes in facilitydesign, performance characteristics and operating procedures related to reactor safety occurring during the reporting period, and results of surveillance tests and inspections.

b. A tabulation showing the energy generated by the reactor (in megawatt hours), hours the reactorwas critical, and the cumulative total energy output since initial criticality.

Amendment No. 436

• (2) The written report (and, to the extent possible, the preliminary telephone report orreport by similar conveyance) shall describe,

analyze, and evaluate safety implications, andoutline the corrective measures taken or planned to prevent reoccurrence of the event.c. A report within 30 days in writing to the NRC, Document Control Desk, Washington DC.(I) Any significant variation of measured values from a corresponding predicted orpreviously measured value of safety-connected operating characteristics occurring duringoperation of the reactor;(2) Any significant change in the transient or accident analysis as described in the SafetyAnalysis Report (SAR);(3) A personnel change involving the positions of UCD/MNRC Director or UCD ViceChancellor for Research; and(4) Any observed inadequacies in the implementation of administrative or procedural controls such that the inadequacy causes or could have caused the existence ordevelopment of an unsafe condition with regard to reactor operations.

6.8 Records. Records may be in the form of logs, data sheets, or other suitable forms. The requiredinformation may be contained in single or multiple

records, or a combination thereof.

Records and logs shallbe prepared for the following items and retained for a period of at least five years for items a. through f., andindefinitely for items g. through k. (Note: Annual reports, to the extent they contain all of the required,information, may be used as records for items g. through j.)a. Normal reactor operation.

b. Principal maintenance activities.

c. Those events reported as required by Technical Specifications 6.7.1 and 6.7.2.d. Equipment and component surveillance activities required by the Technical Specifications.

e. Experiments performed with the reactor.f. Airborne and liquid radioactive effluents released to the environments and solid radioactive waste shippedoff site.g. Offsite environmental monitoring surveys.h. Fuel inventories and transfers.

i. Facility radiation and contamination surveys.j. Radiation exposures for all personnel.

k. Updated, corrected, and as-built drawings of the facility.

Amendment No. 439 072hZI -I -UNIVERSITY OFCALIFORNIA -DAVISCE CHANCELLOR FORRESEARCH(Licensee) ,TORTIONSNJCHI....................... Formal Licensing ChannelAdministrative Reporting Channel----Communications ChannelUCD/MNRC ORGANIZATION FOR LICENSING AND OPERATION FIGURE 6.140 Dr. Barry M. KleinVice Chancellor for ResearchUniversity of California, DavisOne Shields AvenueDavis, CA 95616-8558

SUBJECT:

ISSUANCE OF AMENDMENT NO. 5 TO AMENDED FACILITY OPERATING LICENSE NO. R-130 -REGENTS OF THE UNIVERSITY OF CALIFORNIA (TAC NO. MB5598)

Dear Dr. Klein:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 5 to FacilityOperating License No. R-1 30 for the McClellan Nuclear Radiation Center (MNRC) TRIGAResearch Reactor. The amendment consists of changes to the Technical Specifications (TSs)in response to your submittal of October 17, 2002, and is discussed in the enclosed SafetyEvaluation Report.Sincerely, Warren J. Eresian, Project ManagerResearch and Test Reactors SectionOperating Reactor Improvements ProgramDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation Docket No. 50-607

Enclosures:

1. Amendment No. 52. Safety Evaluation Report SUniversity of California

-Davis/McClellan MNRC Docket No. 50-607cc:Dr. Wade J. Richards5335 Price Avenue, Bldg. 258McClellan AFB, CA 95652-2504 Test, Research, and TrainingReactor Newsletter University of Florida202 Nuclear Sciences CenterGainesville, FL 32611 Dr. Barry M. KleinVice Chancellor for ResearchUniversity of California, DavisOne Shields AvenueDavis, CA 95616-8558

SUBJECT:

ISSUANCE OF AMENDMENT NO. 5 TO AMENDED FACILITY OPERATING LICENSE NO. R-130 -REGENTS OF THE UNIVERSITY OF CALIFORNIA (TAC NO. MB5598)

Dear Dr. Klein:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 5 to FacilityOperating License No. R-130 for the McClellan Nuclear Radiation Center (MNRC) TRIGAResearch Reactor. The amendment consists of changes to the Technical Specifications (TSs)in response to your submittal of October 17, 2002, and is discussed in the enclosed SafetyEvaluation Report.Sincerely, Warren J. Eresian, Project ManagerResearch and Test Reactors SectionOperating Reactor Improvements ProgramDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation Docket No. 50-607

Enclosures:

1. Amendment No. 52. Safety Evaluation ReportDISTRIBUTION:

PUBLICMMendonca AAdamsEHyltonGHiIl (2) (T5-C3)RORP\R&TR r/fWEresianPDoylePlsaacLBergSHolmesTDragounCBassettDHughesOGCPMaddenDMatthews WBecknerADAMS ACCESSION NO: ML02 TEMPLATE

1. NRR-058NAME WEresian:rdr EHylton SUttal PMadden WBecknerIiDATE 111/ /2002 11/ /2002 11/ /2002 11/ /2002 11/ /2002JOFFICIAL RECORD COPY

• 0REGENTS OF THE UNIVERSITY OF CALIFORNIA ATMcCLELLAN NUCLEAR RADIATION CENTERDOCKET NO. 50-607AMENDMENT TO AMENDED FACILITY OPERATING LICENSEAmendment No. 5License No. R-1301. The U.S. Nuclear Regulatory Commission (the Commission) has found thatA. The application for an amendment to Amended Facility Operating LicenseNo. R-1 30 filed by the Regents of the University of California at McClellan NuclearRadiation Center (the licensee) on October 17, 2002, conforms to the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and theregulations of the Commission as stated in Chapter I of Title 10 of the Code ofFederal Regulations (10 CFR);B. The facility will operate in conformity with the application, the provisions of the Act,and the rules and regulations of the Commission; C. There is reasonable assurance that (i) the activities authorized by this amendment can be conducted without endangering the health and safety of the public and (ii)such activities will be conducted in compliance with the regulations of theCommission; D. The issuance of this amendment will not be inimical to the common defense andsecurity or to the health and safety of the public;E. This amendment is issued in accordance with the regulations of the Commission asstated in 10 CFR Part 51, and all applicable requirements have been satisfied; andF. Prior notice of this amendment was not required by 10 CFR 2.105, and publication of notice for this amendment is not required by 10 CFR 2.106.

D 0-2-2. Accordingly, the license is amended by changes to the Technical Specifications asindicated in the enclosure to this license amendment, and paragraph 2.C.(ii) of AmendedFacility Operating License No. R-130 is hereby amended to read as follows:2.C.(ii) Technical Specifications The Technical Specifications contained in Appendix A, as revised throughAmendment No. 5, are hereby incorporated in the license. The licensee shalloperate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Warren J. Eresian, Project ManagerResearch and Test Reactors Section*Operating Reactor Improvements ProgramDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation

Enclosure:

Appendix A, Technical Specification ChangesDate of Issuance: ENCLOSURE TO LICENSE AMENDMENT NO. 5AMENDED FACILITY OPERATING LICENSE NO. R-130DOCKET NO. 50-607Replace the following pages of Appendix A, Technical Specifications, with the enclosedpages. The revised pages are identified by amendment number and contain vertical linesindicating the areas of change.Remove Insert17 1718 1840 40 Basi.__s-

a. A limitation of less than one dollar ($1 .00)(0.7%Ak/k) on the reactivity worth of a single movable experiment positioned in the pneumatic transfer tube, the central irradiation facility (SAR, Chapter 10, Section 10.4.1 ), thecentral irradiation fixture (CIF-1)(SAR, Chapter 10, Section 10.4.1 ), or any other in-core or in-tank irradiation

facility, will assure that the pulse limit of $1.75 is not exceeded (SAR Chapter 13, Section 13.2.2.2.1).

Inaddition, limiting the worth of each movable experiment to less than $1.00 will assure that the additional increase in transient power and temperature will be slow enough so that the fuel temperature scram will beeffective (SAR Chapter 13, Section 13.2.2.2.1).

b. The absolute worst event which may be considered in conjunction with a single secured experiment is itssudden accidental or unplanned removal while the reactor is operating.

For such an event, the reactivity limitfor fixed experiments ($1.75) would result in a reactivity increase less than the $1.92 pulse reactivity insertion needed to reach the fuel temperature safety limit (SAR Chapter 13, Section 13.2.2.2.1).

c. A reactivity limit of less than $1.75 for any single experiment or for all experiments collectively positioned in the sample can of the automated central irradiation facility (AC1F)(SAR Chapter 10, Section 1.0.4.2) is basedon the pulsing reactivity insertion limit (Technical Specification 3.1 .2)(SAR Chapter 13, Section 13.2.2.2.1) andon the design of the ACIF, which allows control .over the positioning of samples into and out of the central coreregion in a manner identical in form, fit, and function to a control rod.d. It is conservatively assumed that simultaneous removal of all experiments positioned in the pneumatic transfer tube, and in any other reactor in-core and in-tank irradiation facilities at any given time shall be lessthan the maximum reactivity insertion limit of $1.92. The SAR Chapter 13, Section 13.2.2.2.1 indicates thata pulse reactivity insertion of $1.92 would be needed to reach the fuel temperature safety limit.3.8.2 Materials LimitApplicability

-This specification applies to experiments installed in reactor experiment facilities. Objective -The objective is to prevent damage to the reactor or significant releases of radioactivity by limitingmaterial quantity and the radioactive material inventory of the experiment. Specification -The reactor shall rnot be operated unless the following conditions governing experiment materials exist:a. Experiments containing materials corrosive to reactor components, compounds highly reactive with water,potentially explosive materials, and liquid fissionable materials shall be appropriately encapsulated.

b. Each fueled experiment shall be controlled such that the total inventory of iodine isotopes 131 through 135in the experiment is no greater than 1.5 curies and the maximum strontium inventory is no greater than 5millicuries.

c. Each experiment in the 1-125 production facility shall be controlled such that the total inventory of I-125 inthe I-125 glove box shall not exceed 40 curies.d. Each experiment in the 1-125 production facility shall be controlled such that the total inventory of 1-125being handled in the 1-125 fume hood at any one time in preparation for shipping shall not exceed 20 curies.An additional.

1.0 curie of 1-125 (up to 400 millicuries in the form of quality assurance samples and up to 600millicuries in sealed storage containers) may also be present in the 1-125 fume hood.Amendment No. 517

• 0e. Explosive materials in quantities greater than 25 milligrams of TNT equivalent shall not be irradiated in thereactor tank. Explosive materials in quantities of 25 milligrams of TNT equivalent or less may be irradiated provided the pressure produced upon detonation of the explosive has been calculated and/or experimentally demonstrated to be less than the design pressure of the container.

f. Explosive materials in quantities of three (3) pounds of TNT equivalent or less may be irradiated in anyradiography bay. The irradiation of explosives in any bay is limited to those assemblies where a safetyanalysis has been performed that shows that there is no damage to the reactor safety systems upondetonation (SAR Chapter 13, Section 13.2.6.2).

Basis -a. Appropdiate encapsulation is required to lessen the experimental hazards of some types of materials.

b. The 1.5 curies limitation on iodine 131 through 135 assures that in the event of failure of a fueledexperiment leading to total release of the iodine, occupational doses and doses to members of the generalpublic in the unrestricted areas shall be within the limits in 10 CFR 20 (SAR Chapter 13, Section 13.2.6.2).

c&d. Limiting the total 1-125 inventory to forty (40.0) curies in the 1-125 glove box and to twenty-one (21.0)curies in the I-125 fume hood assures that, if either of these inventories of 1-125 is totally released into theirrespective containments, the occupational doses and doses to members of the general public in theunrestricted areas shall be within the limits of 10 CFR 20 (SAR Chapter 13, Section 1 3.2.6.2).

e. This specification is intended to prevent damage to vital equipment by restricting the quantity of explosive materials within the reactor tank (SAR Chapter 13, Section 13.2.6.2).

f. The failure of an experiment involvng the irradiation of 3 lbs TNT equivalent or less in any radiography bayexternal to the reactor tank will not result in damage to the reactor controls or the reactor tank. SafetyAnalyses have been performed (SAR Chapter 13, Section 13.2.6.2) which show that up to six (6) pounds ofTNT equivalent can be safely irradiated in any radiogaphy bay. Therefore, the three (3) pound limit gives asafety margin of two (2).3.8.3 Failure and Malfunctions Applicability

-This specification applies to experiments installed in reactor experiment facilities. Obiective -The objective is to prevent damage to the reactor or significant releases of radioactive materials in the event of an experiment failure.Specification -a. All experiment materials which could. off-gas,

sublime, volatilize, or produce aerosols under:(1) normal operating conditions of the experiment or reactor,(2) credible accident conditions in the reactor, or(3) where the possibility exists that the failure of an experiment could release radioactive gases oraerosols into the reactor building or into the unrestricted area, the quantity and type of material in theexperiment shall be limited such that the airborne radioactivity in the reactor room will not result inexceeding the applicable dose limits in 10 CFR Part 20 in the unrestricted area, assuming 100% ofthe gases or aerosols escapes.Amendment No. 518 S0

• 0SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 5 TOAMENDED FACILITY OPERATING LICENSE NO. R-130REGENTS OF THE UNIVERSITY

.OF CALIFORNIA ATMcCLELLAN NUCLEAR RADIATION CENTERDOCKET NO. 50-60

71.0 INTRODUCTION

By letter dated October 17, 2002, the Regents of the University of California (the licensee) submitted a request for amendment of the Technical Specifications (TSs), Appendix A, to FacilityOperating License No. R-i130 for the McClellan Nuclear Radiation Center (MNRC) TRIGAresearch reactor. The request provides for the following

changes, which if implemented, will resultin Revision 12 of the TSs:1. Incorporate a new management

position, the "Site Manager" into the Technical Specifications.

2. Revise Technical Specification 3.8.2, Materials Limit, to allow an increase in the Iodine-i125 inventory in the Iodine Production Facility from 20 curies to 61 curies.Each of these requests is discussed below.2.0 EVALUATION The current management structure includes an UCD/MNRC Director to whom reports aHealth Physics Manager and Reactor Operations Manager.

The proposed management structure creates a new position of Site Manager, who reports directly to the UCD/MNRC

Director, and towhom reports the Health Physics Manager and the Reactor Operations Manager.

The proposedmanagement structure thus creates an additional layer of oversight. Since this change increases oversight and supervision of facility operations, the staff concludes that it is acceptable. Amendment No. 4 of the Technical Specifications was approved on August 9, 2001. Thisamendment approved the installation of an Iodine-125 production loop. The production loopincluded a reactor room glove box containing up to 20 curies of lodine-125. Technical Specification 3.8.2, which provides materials limits of experiments installed in reactor experiment facilities, was expanded to include limits associated with the production loop and in particular, thereactor room glove box. The justification for the 20 curie limit was provided in Chapter 13,Accident

Analysis, of the facility Safety Analysis Report.Previous calculations supporting the 20 curie limit of Iodine-125 were based on the worst-case assumption that all 20 curies of Iodine-125 volatilized and left the glove box through the glove box 0-2-exhaust system, eventually to make its way to the unrestricted area. The exposure (CEDE to thethyroid) to a person in the unrestricted area for the entire 30 second duration of this event is muchless than 1 millirem.

If the exposure duration is increased to 10 minutes, the estimated CEDE tothe thyroid would still be less than 1 millirem. For those exposed in the reactor room for themaximum assumed occupancy time of 5 minutes the CEDE to the thyroid would be about 67millirem. The results of all of the assumptions and calculations in the accident sequence are directlyproportional to the initial inventory of Iodine-125 in the production system. Increasing the initialassumed inventory from 20 curies to 61 curies will simply result in a tripling of the exposure. Theanalysis in the SAR that supports the increase in iodine inventory shows that the CEDE to thethyroid for a 10-minute exposure in the unrestricted area would be about 3.0 millirem, For thoseexposed in the reactor room for the maximum assumed occupancy time of 5 minutes the CEDE tothe thyroid would be about 205 millirem. In order to assess the potential consequences of the worst-case assumption, the resulting dosesare compared to the doses which are expected for the Maximum Hypothetical Accident (MHA),which serves as the bounding accident for radiological consequences. The MHA has beenanalyzed in the licensee's Safety Analysis Report (SAR), and is a complete cladding rupture of ahighly-irradiated single fuel element, followed by the instantaneous release of fission products intothe air.. The accident analysis calculates the radiological consequences of the MHA with regard todoses to the general public in the unrestricted area, and also calculates occupational doses withinthe site boundary. The MHA results in a CEDE of 53 millirem in the unrestricted area. Since therelease of 61 curies of Iodine-125 through the glovebox exhaust system and eventually to theunrestricted area results in a CEDE of about 3 millirem, the radiological result is significantly less.than that of the MHA, the bounding accident. For those exposed in the reactor room, the MHA results in an exposure (CEDE) of 360 millirem. For the failure analyzed here, the five-minute is about 205 millirem. Again, the exposures are lessthan that of the MHA, the bounding accident. The staff concludes that the consequences of the complete volatilization of 61 curies of Iodine-125 are much less than the consequences of the bounding MHA, and that increasing theallowable activity of lodine-125 in the Iodine Production Facility from 20 curies to 61 curies doesnot significantly reduce the margin of safety with respect to the Maximum Hypothetical Accidentand to 10 CFR Part 20 limits and that the increase is acceptable. 3.0 ENVIRONMENTAL CONSIDERATION This amendment involves changes in the installation or 'use of a facility component located withinthe restricted area as defined ion 10 CFR Part 20 or changes in inspection and surveillance requirements. The staff has determined that this amendment involves no significant increase inthe amounts, and no significant change in the types, of any effluents that may be released off site,and no significant increase in individual or cumulative occupational radiation exposure. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10Amendment No. 5 0 0-3-CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement orenvironmental assessment need be prepared with the issuance of this amendment.

4.0 CONCLUSION

The staff has concluded, on the basis of the considerations discussed above, that (1) because theamendment does not involve a significant increase in the probability or consequences ofaccidents previously evaluated, or create the possibility of a new or different kind of accident fromany accident previously evaluated, and does not involve a significant reduction in a margin ofsafety, the amendment does not involve a significant hazards consideration; (2) there isreasonable assurance that the health and safety of the public will not be endangered by theproposed changes; and (3) such changes are in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security orthe health and safety of the public.Principal Contributor: Warren J. EresianDate:Amendment No. 5

• U REGUALTORY COMMISSION i __ _ _ _ _ _ _ _ _ _ _ _.Ii UNIVERSITY OFl CALIFORNIA

-DAVIS* VICE CHANCELLOR FORI ~RESEARCHi -(Licensee) I II IISDIRECTOR NUCLEAR.H_____SAFETYCO l -COMMITITEE LI A-tC--SITE 1 iMANAGER[ I i-***-*HEALTH PHYSICS REACTORBRANCH OPERATIONS Forml Liensig Chnne___________ Aminstrtie RpotinBCANnel CormmunLicatinsin ChannelUCD/MNRC ORGANIZATION FOR LICENSING AND OPERATION FIGURE 6.1 RE NIE SAENUCLEAR REGULATORY COMMISSION ~WASHJNGTON, D.C. 20555-0001 N~ovemb~er 2_5, 2003Dr. Barry M. KleinVice Chancellor for ResearchUniversity of California, DavisOne Shields AvenueDavis, CA 95616-8558

SUBJECT:

ISSUANCE OF AMENDMENT NO. 6 TO AMENDED FACILITY OPERATING LICENSE NO. R-130 -REGENTS OF THE UNIVERSITY OF CALIFORNIA (TAC NO. MB5598)

Dear Dr. Klein:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 6 toFacility Operating License No. R-130 for the McClellan Nuclear Radiation Center (MNRC)TRIGA Research Reactor. The amendment consists of changes to the Technical Specifications (TSs) in response to your submittal of March 31, 2003, and is discussed in the enclosed SafetyEvaluation Report.Sincerely, 6~)4A,~ .6Warren J. Eresian, Project ManagerResearch and Test Reactors SectionNew, Research and Test Reactors ProgramDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation Docket No. 50-607

Enclosures:

1. Amendment No. 62. Safety Evaluation Report

• 0University of California

-Davis/McClellan MNRC Docket No. 50-607cc:Dr. Wade J. Richards5335 Price Avenue, Bldg. 258McClellan AFB, CA 95652-2504 Test, Research, and TrainingReactor Newsletter University of Florida202 Nuclear Sciences CenterGainesville, FL 32611 e- ~* *~W OUNITED STATES REGULATORY COMMISSION D.C. 20555-0001 REGENTS OF THE UNIVERSITY OF CALIFORNIA ATMcCLELLAN NUCLEAR RADIATION CENTERDOCKET NO. 50-607AMENDMENT TO AMENDED FACILITY OPERATING LICENSEAmendment No. 6License No. R- 1301. The U.S. Nuclear Regulatory Commission (the Commission) has found thatA. The application for an amendment to Amended Facility Operating LicenseNo. R-1 30 filed by the Regents of the University of California at McClellan NuclearRadiation Center (the licensee) on March 31, 2003, conforms to the standards andrequirements of the Atomic Energy Act of 1954, as amended (the Act), and theregulations of the Commission as stated in Chapter I of Title 10 of the Code ofFederal Regulations (10 CFR);B. The facility will operate in conformity with the application, the provisions of the Act,and the rules and regulations of the Commission; C. There is reasonable assurance that (i) the activities authorized by this amendment can be conducted without endangering the health and safety of the public and (ii)such activities will be conducted in compliance with the regulations of theCommission; D. The issuance of this amendment will not be inimical to the common defense andsecurity or to the health and safety of the public;E. This amendment is issued in accordance with the regulations of the Commission asstated in 10 CFR Part 51, and all applicable requirements have been satisfied; andF. Prior notice of this amendment was not required by 10 CFR 2.105, and publication of notice for this amendment is not required by 10 CFR 2.106.

• 0-2-2. Accordingly, the license is amended by changes to the Technical Specifications asindicated in the enclosure to this license amendment, and paragraph 2.C.(ii) of AmendedFacility Operating License No. R-130 is hereby amended to read as follows:2.C.(ii)

Technical Sp~ecifications The Technical Specifications contained in Appendix A, as revised throughAmendment No. 6, are hereby incorporated in the license. The licensee shalloperate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Patrick M. Madder Seto ChiefResearch and Test Reactors SectionNew, Research and Test Reactors ProgramDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation

Enclosure:

Appendix A, Technical Specification ChangesDate of Issuance: November 25, 2003

• 0ENCLOSURE TO LICENSE AMENDMENT NO. 6AMENDED FACILITY OPERATING LICENSE NO. R-130DOCKET NO. 50-607Replace the following pages Of Appendix A, Technical Specifications, with the enclosedpages. The revised pages are identified by amendment number and contain vertical linesindicating the areas of change.Remove Insert.31 3132 3233 33Figure 6.1 Figure 6.1
• 05.4 Fissionable Material StorageAppDlicabilitv

-This specification applies to the storage of reactor fuel at a time when it is not in the reactorcore.Obiective -The objective is to assure that the fuel which is being stored will not become critical and will notreach an unsafe temperature. Specification -a. All fuel elements not In the reactor core shall be stored (wet or diy) in a geometrical array where the keffis less than 0.9 for all conditions of moderation.

b. Irradiated fuel elements shall be stored in an array which shall permit sufficient natural convection cooling by water or air such that the fuel temperature shall not exceed the safety limit.Basis -The limits imlposed by Technical Specifications 5.4.a and 5.4.b assure safe storage.6.0 Administrative Controls6.1 Organization.

The Vice Chancellor for Research shall be the licensee for the UCD1MNRC. The facility shall be under the direct control of the UCD/MNRC Director. The UCD/MNRC Directorshall be accountable to the Vice Chancellor for Research for the safe operation and maintenance of thefacility. 6.1.1 Structure. The management for operation of the UCD/MNRC facility shall consist of theorganizational structure as shown in Figure 6.1.6.1.2 Responsibilities. The UCDIMNRC Director shall be accountable to the Vice Chancellor forResearch for the safe operation and maintenance of the facility. The UCDIMNRC

Director, or hisdesignated alternate, shall review and approve all experiments and experiment procedures prior totheir use in the reactor.

Individuals in the management organization (e.g., Operations Manager,Reactor Supervisor, Health Physics Supervisor) shall be responsible for implementing U CD/MNRCpolicies and for operation of the facility, and shall be responsible for safeguarding the public andfacility personnel from undue radiation exposures and for adhering to the operating license andtechnical specifications. The Operations Manager shall report directly to the UCD/MNRC

Director, and shall immediately report all items involving safety and licensing to the Director for a finaldecision.

The Reactor Supervisor and Health Physics Supervisor report directly to the Operations Manager.. 6.1.3 Staffing6.1.3.1 The minimum staffing when the reactor is not shutdown shall be:a. A reactor operator in the control room;b. A second person in the facility who can perform prescribed instructions;

c. A senior reactor operator readily available.

The available senior reactoroperator should be within thirty (30) minutes of the facility and reachable bytelephone, and;d. A senior reactor operator shall be present whenever a reactor startup isperformed, fuel Is being moved, or experiments are being placed In the reactortank.6.1.3.2 A list of reactor facility personnel by name and telephone number shall be available to the reactor operator in the control room. The list shall include:Amendment No. 631

• 0a. Management personnel.

b. Health Physics personnel.

c. Reactor Operations personnel.

6.1.4 Selection and Training of Personnel. The selection, training and requalification of operations personnel shall meet or exceed the requirements of the American National Standard for Selection and Training of Personnel for Research Reactors (ANS 15.4). Qualification and requalification oflicensed operators shall be subject to an approved Nuclear Regulatory Commission (NRC)program.6.2 Review. Audit. Recommendation and ApprovalGeneral Policy. Nuclear facilities shall be designed, constructed,

operated, and maintained in suchamanner that facility personnel, the general public, and both university and non-university propertyare not exposed to undue risk. These activities shall be conducted in accordance with applicable regulatory requirements.

The UCD Vice Chancellor for Research shall institute the above stated policy as the facility licenseholder. The Nuclear Safety Committee (NSC) has been chartered to assist in meeting thisresponsibility by providing timely, objective, and independent

reviews, audits, recommendations andapprovals on matters affecting nuclear safety. The following describes the composition andconduct of the NSC.6.2.1 NSC Composition and Qualifications.

The UCD Vice Chancellor for Research shall appointthe Chairperson of the NSC. The NSC Chairperson shall a ppoint a Nuclear Safety Committee (NSC) of at least seven (7) members knowledgeable in fields which relate to nuclear safety. TheNSC Shall evaluate and review nuclear safety associated with the operation and use of theUCD/MNRC. 6.2.2 NSC Charter and Rules. The NSC shall conduct its review and audit (inspection) functions inaccordance with a written charter. This charter shall include provisions for:a. Meeting frequency (The committee shall meet at least semiannu'ally.)

b. Voting rules.c. Quorums (For the full committee, a quorum will be at least seven (7) members.d. A committee review function and an audit/inspection function.

e. Use of subcommittees.

f. Review, approval and dissemination of meeting minutes.6.2.3 Review Function.

The responsibilities of the NSC, or a designated subcommittee thereof,shall "but ar--e-n'ot limited to the following:

a. Review approved experiments utilizing UCD/MNRC nuclear facilities.

b. Review and approve all proposed changes to the facility

license, the Technical Specifications and the Safety Analysis Report, and any new or changed Facility Use Authorizations and proposedClass I modifications, prior to implementing (Class I) modifications, prior to taking action under thepreceding documents or prior to forwarding any of these documents to the Nuclear Regulatory Commission for approval.

c. Review and determine whether a proposed change, test, or experiment would constitute anunreviewed safety question or require a change to the license, to a Facility Use Authorization, orto the Technical Specifications.

This determination may be in the form of verifying a decisionalready made by the UCD/MNRC Director. Amendment No. 632

d. Review reactor operations and operational maintenance, Class I modification

records, and thehealth physics program and associated records for all UCDIMNRC nuclear facilities.

e. Review the periodic updates of the Emergency Plan and Physical Security Plan for UCD/MNRCnuclear facilities.

f. Review and update the NSC Charter every two (2) years.g. Review abnormal performance of facility equipment and operating anomalies.

h. Review all reportable occurrences and all written reports of such occurrences prior to forwarding the final written report to the Nuclear Regulatory Commission.

i. Review the NSC annual audit/inspection of the UCD/MNRC nuclear facilities and any otherinspections of these facilities conducted by other agencies.

6.2.4 Audit/Inspection Function. The NSC or a subcommittee

thereof, shall audit/inspect reactoroperations and health physics annually.

The annual audit/inspection shall include, but not belimited to the following:

a. Inspection of the reactor operations and operational maintenance, Class I modification records,and the health physics program and associated

records, including the ALARA program, for allUCDIMNRC nuclear facilities.

b. Inspection of the physical facilities at the UCD/MNRC.

c. Examination of reportable events at the UCDIMNRC.

d. Determination of the adequacy of UCD/MNRC standard operating procedures.

e. Assessment of the effectiveness of the training and refraining programs at the UCD/MNRC.

f. Determination of the conformance of operations at the UCD/MNRC with the facility's license andTechnical Specifications, and applicable regulations.

g. Assessment of the results of actions taken to correct deficiencies that have occurred in nuclear.safety related equipment, structures,

systems, or methods of operations.

h. Inspection of the currently active Facility Use Auhorizations and associated experiments.

i. Inspection of future plans for facility modifications or facility utilization.

j. Assessment of operating abnormalities.

k. Determination of the status of previous NSC recommendations.

6.3 Radiation Safety. The Health Physics Supervisor shall be responsible for implementation of theUCD/MNRC Radiation Safety Program. The program should use the guidelines of the American NationalStandard for Radiation Protection at Research Reactor Facilities (ANSI/ANS 15.11). The Health PhysicsSupervisor shall report to the Operations Manager.6.4 Procedures. Written procedures shall be prepared and approved prior to initiating any of the activities listed nthssction. The procedures shall be approved by the UCD/MNRC Director. A periodic review ofprocedures shall be performed and documented in a timely manner by the UCD/MNRC staff to assure thatprocedures are current. Procedures shall be adequate to assure the safe operation of theAmendment No. 633

• 0I-..... COMMISSION UNVRIYOCAIOM AISAFETYECOITYTEE 1 AIFRI C-MDATEES I VIE MANAGELLRFO I, ISUPERISRECREANCTO AR.. SFTOPERSUPERVISOR M A N A G ER_______________________________________________

iForml Liensig ChnneUCD/NRC ORGAIZAIOSOR REACTOR OETINLICNIGADSFT FIGURE 6.1 1-**R R OUNITED STATES,NUCLEAR REGULATORY COMMISSIONWASHINGTON, D.C. 20555-0001 "SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 6 TOAMENDED FACILITY OPERATING LICENSE NO. R-130REGENTS OF THE UNIVERSITY OF CALIFORNIA ATMcCLELLAN NUCLEAR RADIATION CENTERDOCKET NO. 50-60

71.0 INTRODUCTION

By letter dated March 31, 2003, the Regents of the University of California (the licensee) submitted a request for amendment of the Technical Specifications (TSs), Appendix A, to FacilityOperating License No. R-1 30 for the McClellan Nuclear Radiation Center (MNRC) TRIGAresearch reactor. The request provides for the following

changes, which if implemented, will resultin Revision 13 of the TSs:1. Incorporate a new management

position, the uOperations Manager" into the Technical Specifications and change the UCD/MNRC Organization Chart to reflect this change.2. Change the appointing authority of the Chairperson of the Nuclear Safety Committee (NSC) from the Director of the UCD/MNRC to the UCD Vice Chancellor for Research, andchange the Technical Specifications and UCD/MNRC Organization Chart to reflect thischange.Each of these requests is discussed below.2.0 EVALUATION The current organization structure includes an UCD/MNRC Director to whom reports a SiteManager.

The proposed organization structure, as reflected in Figure 6.1, replaces the Site.Manager position with the position of Operations

Manager, who reports directly to the UCD/MNRCDirector, and to whom reports the Health Physics Branch and the Reactor Operations Branch.Since the proposed organization structure does not alter or reduce lines of authority and oversight, the staff concludes that it is acceptable.

In the current organization structure, the UCD/MNRC Director is responsible for appointing theChairperson of the NSC. In the proposed organization structure, that responsibility is given to theUCD Vice Chancellor for Research, who is also the licensee for the UCD/MNRC. Since thisproposed change increases the level of oversight from the licensee's staff to the licensee, the staffconcludes that it is acceptable. The staff has reviewed the proposed changes to the TSs and concluded that they areadministrative in nature and do not impact the licensee's ability to continue to meet the relevantrequirements of 10 CFR 50.36.3.0 ENVIRONMENTAL CONSIDERATION This amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR51 .22(c)(1 0). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared with the issuance of this amendment.

4.0 CONCLUSION

The staff has concluded, on the basis of the considerations discussed above, that (1) because theamendment does not involve a significant increase in the probability or consequences of accidents previously evaluated, or create the possibility of a new or different kind of accident from anyaccident previously evaluated, and does not involve a significant reduction in a margin of safety,the amendment does not involve a significant hazards consideration; (2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposedchanges; and (3) such changes are in compliance with the Commission's regulations and theissuance of this amendment will not be inimical to the common defense and security or the healthand safety of the public.Principal Contributor: Warren J. EresianDate: November 25, 2003Amendment No. 6 0 0TECHNICAL SPECIFICATIONS FOR THEUNIVERSITY OF CALIFORNIA -DAVIS MCCLELLAN NUCLEAR RADIATION CENTER(UCD/MNRC) DOCUMENT NUMBER: MNRC-0004-DOC-1 2Rev 12 09/02 Oct. 16 0? 11:OOa1,JLaJ. rwdIL.Ar t~~r 6. JohnsontS4lJ 753-9743.atDfJh.L.* ~ r. t.wc.p.1TECHNICAL SPECtFICATIONS APPROVALRevision 12 of me "Teclnical Gpo ctifoons* for the Universit of California at DavistlMcCleIlan, NuclearRadiation Cencer (UOI)/MNRG) Reactor have undergone the following coordination: Reviewed Rcvicwcd by'." 'floa rMnae " "10 ~ 02-DaleD~kcIR~eviewed by:Approved by:Site ManagerUCD/MNRc~bir4ctor Date/~zL7z~OZ-DataDateIApprovod by;

• 0Technical Specifications Rev 12 09/2002TtePageRe12 902Titovle Page Rev 12 9/200232 Rev 12 9/2002Figure 6.1 Rev 12 9/2002 S 0TECHNICAL SPECIFICATIONS TABLE OF CONTENTSPaage1.0 Definitions

................................... ............................................................................ 2.0 Safety Limit and Limiting Safety System Setting (LSSS.)............................................................ 52.1 Safety Limits................................................................................................... 52.2 Limiting Safety System Setting (LSSS)...................................................................... 62.2.1 Fuei Temperature.................................................................................. 63.0 Limiting Conditions for Operations (LC.O.0............................................................................. 73.1 Reactor Core Parameters.................................................................................... 73.1 .1 Steady-State Operation ........................................................................... 73.1.2 Pulse or Square Wave Operation ................................................................ 73.1.3 Reactivity Limitations .............................................................................. 83.2 Reactor Control and Safety Systems........................................................................ 83.2.1 Control Rods....................................................................................... 83.2.2 Reactor Instrumentation................................... ........................................ 93.2.3 Reactor Scrams and Interlocks.................................................................. 103.2.4 Reactor Fuel Elemenis........................................................................... 123.3 Reactor Coolant Systems................................................................................... 133.4 Reactor Room Exhaust System ............................................................................ 143.5 Intentionally Left Blank ...................................................................................... 143.6 Intentionally Left Blank ...................................................................................... 143.7 Reactor Radiation Monitoring Systems .................................................................... 143.7.1 Monitoring Systems .............................................................................. 143.7.2 Effluents -Argon-41 Discharge Limit..............,.............................................. 16 9 03.8 Experiments ................................................................................................. 163.8.1 Reactivity Limits................................................................................... 163.8.2 Materials Limit.................................................................................... 173.8.3 Failure and Malfunctions......................................................................... 184.0 Surveillance Requirements .......................................................................................... 194.1 Reactor Core Parameters................................................................................... 194.1.1 Steady State Operation.......................................................................... 194.1.2 Shutdown Margin and Excess Reactivity ....................................................... 204.2 Reactor Control and Safety Systems ...................................................................... 204.2.1 Control Rods ..................................................................................... 204.2.2 Reactor Instrumentation ......................................................................... 214.2.3 Reactor Scrams and Interlocks.................................................................. 224.2.4 Reactor Fuel 234.3 Reactor Coolant .244.4 Reactor Room Exhaust System ............................................................................ 254.5 Intentionally Left Blank...................................................................................... 254.6 Intentionally Left Blank...................................................................................... 254.7 Reactor Radiation Monitoring Systems .................................................................... 254.8 Experiments ................................................................................................. 265.0 Design Features ...................................................................................................... 275.1 Site and Facility Description ................................................................................ 275.1.1 .Site................................................................................................ 275.1.2 Facility Exhaust .................................................................................. 285.2 Reactor Coolant System..................................................................................... 28 0 S5.3 Reactor Core and F~ue]........................................................................................ 295.3.1 Reactor Core ..........................................................................

.............

295.3.2 Reactor .FuelJ........................................................................................ 305.3.3 Control Rods and Control Rod Drives ............................................................ 315.4 Fissionable Material Storage.................................................................................. 316.0 Administrative Controls.................................................................................................. 316.1 Organization.................................................................................................... 316.1.1 Structure............................................................................................. 326.1.2 Responsibilities..................................................................................... 326.1.3 Staffing .............................................................................................. 326.1.4 Selection and Training of Personnel.............................................................. 326.2 Review, Audit, Recommendation and Approval............................................................. 326.2.1 NSC Composition and Qualifications ................................................ i............ 336.2.2 NSC Charter and Rules ........................................................................... 336.2.3 Review Function.................................................................................... 336.2.4 Audit/Inspection Function.......................................................................... 346.3 Radiation Safety................................................................................................ 346.4 Procedures ..................................................................................................... 346.4.1 Reactor Operations Procedures................................................................... 346.4.2 Health Physics Procedures........................................................................ 356.5 Experiment Review and Appro~ial ............................................................................ 356.6 Required Actions............................................................................................... 356.6.1 Actions to be taken in case of a safety limit violation............................................ 356.6.2 Actions to be taken for reportable occurrences ................................................. 36 6.7 Reports .......................................................................................................... 366.7.1 Operating Reports.................................................................................. 366.7.2 Special Reports..................................................................................... 386.8 Records ......................................................................................................... 39 IFig. 6.1 UCD/MNRC Organization for Licensing and Operation......................................................... 40 0 0TECHNICAL SPECIFICATIONS FOR THEUNIVERSITY OF CALIFORNIA -DAVIS/MCCLELLAN NUCLEAR RADIATION CENTER (UCD/MNRC) GeneralThe University of California -Davis/McClellan Nuclear Radiation Center (UCD/MNRC) reactor is operated by theUniversity of California, Davis, California (UCD). The UCD/MNRC research reactor is a TRIGA-type reactor.The UCD/MNRC provides state-of-the-art neutron radiography capabilities. In addition, the UCD/MNRC providesa wide range of irradiation services for both research and industrial needs. The reactor operates at a nominalsteady state power level up to and including 2 MW. The UCD/MNRC reactor is also capable of square waveand pulse operational modes. The UCD/MNRC reactor fuel is less than 20% enriched in uranium-235.

1.0 Definitions

1.1 As Low As Reasonably Achievable (ALARA). As defined in 10 CFR, Part 20.1.2 Licensed Operators. A UCD/MNRC licensed operator is an individual licensed by the NuclearRegulatory Commission (e.g., senior reactor operator or reactor operator) to carry out the duties andresponsibilities associated with the position requiring the license.1.2.1 Senior Reactor Operator. An individual who is licensed to direct the activities of reactoroperators and to manipulate the controls of the facility. 1.2.2 Reactor Operator. An individual who is licensed to manipulate the controls of the facilityand perform reactor-related maintenance. 1.3 Channel. A channel is the combination of sensor, line amplifier, processor, and output deviceswhich are connected for the purpose of measuring the value of a parameter. 1.3.1 Channel Test. A channel test is the introduction of a signal into the channel for verification that it is operable. 1.3.2 Channel Calibration. A channel calibration is an adjustment of the channel such that itsoutput corresponds with acceptable accuracy to known values of the parameter which thechannel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm or trip, and shall be deemed to include a channel test.1.3.3 Channel Check. A channel check is a qualitative verification of acceptable performance byobservation of channel behavior. This verification, where possible, shall include comparison ofthe channel with other independent channels or systems measuring the same variable. 1.4 Confinement. Confinement means isolation of the reactor room air volume such that the movementof air into and out of the reactor room is through a controlled path.1.5 Experiment. Any operation,

hardware, or target (excluding devices such as detectors, fissionchambers, foils, etc), which is designed to investigate specific reactor characteristics or which isintended for irradiation within an experiment facility and which is not rigidly secured to a core or shieldstructure so as to be a part of their design.1.5.1 Experiment.

Moveable. A moveable experiment is one where it is intended that the entireexperiment may be moved in or near the reactor core or into and out of reactor experiment facilities while the reactor is operating. I 0 01.5.2 Experiment. Secured. A secured experiment is any experiment, experiment

facility, orcomponent of an experiment that is held in a stationary position relative to the reactor bymechanical means. The restraining force must be substantially greater than those to which theexperiment might be subjected by hydraulic, pneumatic,

buoyant, or other forces which arenormal to the operating environment of the experiment, or by forces which can arise as a resultof credible conditions.

1.5.3 Experiment Facilities. Experiment facilities shall mean the pneumatic transfer tube,beamtubes, irradiation facilities in the reactor core or in the reactor tank, and radiography bays.1.5.4 Experiment Safety System. Experiment safety systems are those systems, including theirassociated input circuits, which are designed to initiate a scram for the primary purpose ofprotecting an experiment or to provide information which requires manual protective action to beinitiated. 1.6 Fuel Element. Standard. A fuel element is a single TRIGA element. The fuel is U-ZrH clad instainless steel. The zirconium to hydrogen ratio is nominally 1.65 +/- 0.05. The weight percent (wt%) ofuranium can be either 8.5, 20, or 30 wt%, with an enrichment of less than 20% U-235. A standard fuelelement may contain a burnable poison.1.7 _Fuel Element. Instrumented. An instrumented fuel element is a standard fuel element fabricated withthermocouples for temperature measurements. An instrumented fuel element shall have at least oneoperable thermocouple embedded in the fuel near the axial and radial midpoints. 1.8 Measured Value. The measured value is the value of a parameter as it appears on the output of achannel.1.9 Mode, Steady-State. Steady-state mode operation shall mean operation of the UCD/MNRC reactorwith the selector switch in the automatic or manual mode position. 1.10 Mode. Square-Wave. Square-wave mode operation shall mean operation of the UCD/MNRCreactor with the selector switch in the square-wave mode position. 1.11 Mode. Pulse. Pulse mode operation shall mean operation of the UCD/MNRC reactor with theselector switch in the pulse mode position. 1.12 Operable. Operable means a component or system is capable of performing its intended function. 1.13 Operating. Operating means a component or system is performing its intended function. 1.14 Operating Cycle. The period of time starting with reactor startup and ending with reactor shutdown. 1.15 Protective Action. Protective action is the initiation of a signal or the operation of equipment withinthe UCD/MNRC reactor safety system in response to a variable or condition of the UCD/MNRC reactorfacility having reached a specified limit.1.15.1 Channel Level. At the protective instrument channel level, protective action is thegeneration and transmission of a scram signal indicating that a reactor variable has reached thespecified limit.1.15.2 Subsystem Level. At the protective instrument subsystem level, protective action is thegeneration and transmission of a scram signal indicating that a specified limit has been reached.NOTE: Protective action at this level would lead to the operation of the safety shutdownequipment. 2

• 01.15.3 Instrument System Level. At the protective instrument level, protective action is thegeneration and transmission of the command signal for the safety shutdown equipment tooperate.1.15.4 Safety System Level. At the reactor safety system level, protective action is the operation of sufficient equipment to immediately shut down the reactor.1.16 Pulse Operational Core. A pulse operational core is a reactor operational core for which themaximum allowable pulse reactivity insertion has been determined.

1.17 Reactivity. Excess. Excess reactivity is that amount of reactivity that would exist if all control rods(control, regulating, etc.) were moved to the maximum reactive position from the point where the reactoris at ambient temperature and the reactor is critical. (K eff = 1)1.18 Reactivity Limits. The reactivity limits are those limits imposed on the reactivity conditions of thereactor core.1.19 Reactivity Worth of an Experiment. The reactivity worth of an experiment is the maximum value ofthe reactivity change that could occur as a result of changes that alter experiment position orconfiguration. 1.20 Reactor Controls. Reactor controls are apparatus and/or mechanisms the manipulation of whichdirectly affect the reactivity or power level of the reactor.1.21 Reactor Core. Operational. The UCD/MNRC reactor operational core is a core for which theparameters of excess reactivity, shutdown margin, fuel temperature, power calibration and reactivity worths of control rods and experiments have been determined to satisfy the requirements set forth inthese Technical Specifications. 1.22 Reactor Operating. The UCD/MNRC reactor is operating whenever it is not shutdown or secured.1.23 Reactor Safety Systems. Reactor safety systems are those systems, including their associated input channels, which are designed to initiate automatic reactor protection or to provide information forinitiation of manual protective action.1.24 Reactor Secured. The UCD/MNRC reactor is secured when the console key switch is in the offposition and the key is removed from the lock and under the control of a licensed

operator, and theconditions of a or b exist:a. (1) The minimum number of control rods are fully inserted to ensure the reactor is shutdown, asrequired by technical specifications; and(2) No work is in progress involving core fuel, core structure, installed control rods, or control roddrives, unless the control rod drives are physically decoupled from the control rods; and(3) No experiments in any reactor experiment

facility, or in any other way near the reactor, are beingmoved or serviced if the experiments have, on movement, a reactivity worth exceeding the maximumvalue allowed for a single experiment or $1.00, whichever is smaller, orb. The reactor contains insufficient fissile materials in the reactor core, adjacent experiments or controlrods to attain criticality under optimum available conditions of moderation and reflection.

1.25 Reactor Shutdown. The UCDIMNRC reactor is shutdown if it is subcritical by at least one dollar($1.00) both in the Reference Core Condition and for all allowed ambient conditions with the reactivity worth of all installed experiments included. 3 01.26 Reference Core Condition. The condition of the core when it is at ambient temperature (cold T<28°C), the reactivity worth of xenon is negligible (< $0.30) (i.e., cold and clean), and the central irradiation facility contains the graphite thimble plug and the aluminum thimble plug (CIF-1).1.27 Research Reactor. A research reactor is defined as a device designed to support a self-sustaining neutron chain reaction for research development, education, and training, or experimental

purposes, andwhich may have provisions for the production of radioisotopes.

1.28 Rod. Control. A control rod is a device fabricated from neutron absorbing

material, with or without afuel or air follower, which is used to establish neutron flux changes and to compensate for routinereactivity losses. The follower may be a stainless steel section.

A control rod shall be coupled to itsdrive unit to allow it to perform its control function, and its safety function when the coupling isdisengaged. This safety function is commonly termed a scram.1.28.1 Regulating Rod. A regulating rod is a control rod used to maintain an intended powerlevel and may be varied manually or by a servo-controller. A regulating rod shall have scramcapability. 1.28.2 Standard Rod. The regulating and shim rods are standard control rods.1.28.3 Transient Rod. The transient rod is a control rod that is capable of providing rapid-reactivity insertion to produce a pulse or square wave.1.29 Safety Channel. A safety channel is a measuring channel in the reactor safety system.1.30 Safety Limit. Safety limits are limits on important process variables, which are found to benecessary to reasonably protect the integrity of the principal barriers which guard against theuncontrolled release of radioactivity. 1.31 Scram Time. Scram time is the elapsed time between reaching a limiting safety system set pointand the control rods being fully inserted. 1.32 Scram. External. External scrams may arise from the radiography bay doors, radiography bayripcords, bay shutter interlocks, and any scrams from an experiment. 1.33 Shall. Should and May. The word "shall" is used to denote a requirement; the word "should" todenote a recommendation; the word "may" to denote permission, neither a requirement nor arecommendation. 1.34 Shutdown Margin. Shutdown margin shall mean the minimum shutdown reactivity necessary toprovide confidence that the reactor can be made subcritical by means of the control and safety systemstarting from any permissible operating condition with the most reactive rod assumed to be in the mostreactive

position, and once this action has been initiated, the reactor will remain subcritical withoutfurther operator action.1.35 Shutdown.

Unscheduled. An unscheduled shutdown is any unplanned shutdown of theUCD/MNRC reactor caused by actuation of the reactor safety system, operator error, equipment malfunction, or a manual shutdown in response to conditions which could adversely affect safeoperation, not including shutdowns which occur during testing or check-out operations. 1.36 Surveillance Activities. In general, two types of surveillance activities are specified: operability checks and tests, and calibrations. Operability checks and tests are generally specified as djaily, weeklyor quarterly. Calibration times are generally specified as quarterly, semi-annually,

annually, or biennially.

1.37 Surveillance Intervals. Maximum intervals are established to provide operational flexibility and notto reduce frequency. Established frequencies shall be maintained over the long term. The allowable 4 0surveillance interval is the interval between a check, test, or calibration, whichever is appropriate to the itembeing subjected to the surveillance, and is measured from the date of the last surveillance. Allowable surveillance intervals shall not exceed the following: 1.37.1 An nual -interval not to exceed fifteen (15) months.1.37.2 Semiannual -interval not to exceed seven and a half (7.5) months.1.37.3 Quarterly -interval not to exceed four (4) months.1.37.4 Monthly_- interval not to exceed six (6) weeks.1.37.5 Weekly_- interval not to exceed ten (10) days.1.38 Unreviewed Safety Questions. A proposed change, test or experiment shall be deemed to involvean unreviewed safety question:

a. If the probability of occurrence or the consequences of an accident or malfunction ofequipment important to safety previously evaluated in the safety analysis report may beincreased; orb. If a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; orc. If the margin of safety, as defined in the Basis for any technical specification, is reduced.1.39 Value. Measured.

The measured value is the value of a parameter as it appears on the output of achannel.1.40 Value. True. The true value is the actual value of a parameter. 1.41 Watchdog Circuit. The watchdog circuit is a surveillance circuit provided by the Data Acquisition Computer (DAC) and the Control System Computer (CSC) to ensure proper operation of the reactorcomputerized control system.2.0 Safety Limit and Limiting Safety System Setting (LSSS).2.1 Safety Limits.Applicability -This specification applies to the temperature of the reactor fuel in a standard TRIGA fuelelement.Objective -The objective is to define the maximum temperature that can be permitted with confidence that no damage to the fuel element cladding will result..Specification -a. The maximum fuel temperature in a standard TRIGA fuel element shall not exceed 930 °C duringsteady-state operation.

b. The maximum temperature in a standard TRIGA fuel element shall not exceed 1100 0C during pulseoperation.

Basis -a. This fuel safety limit applies for conditions in which the cladding temperature is above 500 °C (SafetyAnalysis Report (SAR), Chapter 4, Section 4.5.4.1.3). The important parameter for a TRIGA reactor is5 0 0the fuel element temperature. This parameter is well suited as it can be measured directly. A loss in theintegrity of the fuel element cladding could arise if the cladding stress exceeds the ultimate strength ofthe cladding material. The fuel element cladding stress is a function of the element's internal pressurewhile the ultimate strength of the cladding material is a function of its temperature. The cladding stressis a result of the internal pressure due to the presence of air, fission product gasses and hydrogen fromthe disassociation of hydrogen and zirconium in the fuel moderator. Hydrogen pressure is the mostsignificant. The magnitude of the pressure is determined by the fuel moderator temperature and theratio of hydrogen to zirconium in the alloy. At a fuel temperature of 930 °C for ZrH1 7 fuel, the claddingstress due to the internal pressure is equal to the ultimate strength of the cladding material at the sametemperature (SAR Fig 4.18). This is a conservative limit since the temperature of the cladding material isalways lower than the fuel temperature. (See SAR Chapter 4, Section 4.5.4.)b. This fuel safety limit applies for conditions in which the cladding temperature is less than 500 °C.Analysis (SAR Chapter 4, Section 4.5.4.1.1), shows that a maximum temperature for the clad during apulse which gives a peak adiabatic fuel temperature of 1000 °C is estimated to be 470 °C. Furtheranalysis (SAR Section 4.5.4.1.2), shows that the internal pressure for both Zr 1.65 (at 11 50°C) and Zr17z (at11 00°C) increases to a peak value at about 0.3 sec, at which time the pressure is about one-fifth of theequilibrium value or about 400 psi (a stress of 14,700 psi). The yield strength of the cladding at 500 °C isabout 59,000 psi.Calculations for step increases in power to peak ZrH 1.65 fuel temperature greater than 1150 °C, over a 200°Crange, show that the time to reach the peak pressure and the fraction of equilibrium pressure value achievedwere approximately the same as for the 1150 °C case. Similar results were found for fuel with ZrH1.7.Measurements of hydrogen pressure in TRIGA fuel elements during transient operations have been made andcompared with the results of analysis similar to that used to make the above prediction. These measurements indicate that in a pulse where the maximum temperature in the fuel was greater than 1000 °C, the pressure(ZrH1.65) was only about 6% of the equilibrium value evaluated at the peak temperature. Calculations of thepressure gave values about three times greater than the measured values. The analysis gives strongindications that the cladding will not rupture if fuel temperatures are never greater than 1200 °C to 1250°C,providing the cladding temperature is less than 500 0 C. For fuel with ZrH 1.7,a conservative safety limit is 1100 °C.As a result, at this safety limit temperature, the class pressure is a factor of 4 lower than would be necessary forcladding failure.2.2 Limiting Safety System Setting.2.2.1 Fuel Temperature. Applicability -This specification applies to the protective action for the reactor fuel elementtemperature. Objective -The objective is to prevent the fuel element temperature safety limit from beingreached.Specification -The limiting safety system setting shall be 750 °C (operationally this may be setmore conservatively) as measured in an instrumented fuel element. One instrumented elementshall be located in the analyzed peak power location of the reactor operational core.Basis -For steady-state operation of the reactor, the limiting safety system setting is atemperature which, if exceeded, shall cause a reactor scram to be initiated preventing the safetylimit from being exceeded. A setting of 750 °C provides a safety margin at the point of themeasurement of at least 137 °C for standard TRIGA fuel elements in any condition of operation. A part of the safety margin is used to account for the difference between the true and measuredtemperatures resulting from the actual location of the thermocouple. If the thermocouple element is located in the hottest position in the core, the difference between the true andmeasured temperatures will be only a few degrees since the thermocouple junction is near thecenter and mid-plane of the fuel element. For pulse operation of the reactor, the same limitingsafety system setting applies.

However, the temperature channel will have no effect on limiting6 0the peak power generated because of its relatively long time constant (seconds) as comparedwith the width of the pulse (milliseconds).

In this mode, however, the temperature trip will act tolimit the energy release after the pulse if the transient rod sho~uld not reinsert and the fueltemperature continues to increase. 3.0 Limiting Conditions For Operation 3.1 Reactor Core Parameters 3.1.1 Steady-State Operation Applicability -This specification applies to the maximum reactor power attained during steady-state operation. Objective -The objective is to assure that the reactor safety limit (fuel temperature) is notexceeded, and to provide for a setpoint for the high flux limiting safety systems, so thatautomatic protective action will prevent the safety limit from being reached during steady-state operation. Specification -The nominal reactor steady-state power shall not exceed 2.0 MW. The automatic scram setpoints for the reactor power level safety channels shall be set at 2.2 MW or less. *Forthe purpose of testing the reactor steady-state power level scram, the power shall not exceed2.3 MW.Basis_- Operational experience and thermal-hydraulic calculations demonstrate that UCD/MNRCTRIGA fuel elements may be safely operated at power levels up to 2.3 MW with naturalconvection cooling. (SAR Chapter 4, Section 4.6.2.)3.1.2 Pulse or Square Wave Operation Applicability -This specification applies to the peak temperature generated in the fuel as theresult of a step insertion of reactivity. Objective -The objective is to assure that the fuel temperature safety limit will not be exceeded. Specification -a. For the pulse mode of operation, the maximum insertion of reactivity shall be 1.23% A k/k($1.75);b. For~the square wave mode of operation, the maximum insertion of reactivity shall be 0.63%Ak/k ($0.90).Basis -Standard TRIGA fuel is .fabricated with a nominal hydrogen to zirconium ratio of 1.6 to1.7. This yields delta phase zirconium hydride which has a high creep strength and undergoes no phase changes at temperatures in excess of 100 °C. However, after extensive steady stateoperation at two (2) MW the hydrogen will redistribute due to migration from the central hightemperature regions of the fuel to the cooler outer regions. When the fuel is pulsed, theinstantaneous temperature distribution is such that the highest values occur at the radial edge ofthe fuel. The higher temperatures in the outer regions occur in fuel with a hydrogen to zirconium ratio that has now increased above the nominal value. This produces hydrogen gas pressures considerably in excess of that expected. If the pulse insertion is such that the temperature of thefuel exceeds about 875 °C, then the pressure may be sufficient to cause expansion ofmicroscopic holes in the fuel that grow with each pulse. Analysis (SAR Chapter 13, Section13.2.2.2.1), shows that the limiting pulse, for the worst case conditions, is 1.34% A k/k ($1.92).Therefore, the 1.23% A k/k ($1.75) limit is below the worse case reactivity insertion accident limit.7 The $0.90 square wave step insertion limit is also well below the worse case reactivity insertion accident limit.3.1.3 Reactivity Limitations Applicability -These specifications apply to the reactivity conditions of the reactor core and thereactivity worths of the control rods and apply to all modes of reactor operation. Objective -The objective is to assure that the reactor can be placed in a shutdown condition atall times and to assure that the safety limit shall not be exceeded. Specification -a. Shutdown Margin -The reactor shall not be operated unless the shutdown margin provided bythe control rods is greater than 0.35% A k/k ($0.50) with:(1) The reactor in any core condition, (2) The most reactive control rod assumed fully withdrawn, and(3) Absolute value of all movable experiments analyzed in their most reactive condition or $1.00 whichever is greater.b. Excess Reactivity -The maximum available excess reactivity (reference core condition) shallnot exceed 6.65% A k/k ($9.50).Basis -a. This specification assures that the reactor can be placed in a shutdown condition from anyoperating condition and remain shutdown, even if the maximum worth control rod should stick inthe fully withdrawn position (SAR Chapter 4, Section 4.5.5).b. This specification sets an overall reactivity limit which provides adequate excess reactivity tooverride the xenon buildup, to overcome the temperature change in going from zero power to 2MW, to permit pulsing at the $1.75 level, to permit irradiation of negative worth experiments andaccount for fuel burnup over time. An adequate shutdown margin exists with an excess of $9.50for the two analyzed cores: (SAR Chapter 4, Section 4.5.5).3.2 Reactor Control and Safety Systems3.2.1 Control RodsApplicability -This specification applies to the function of the control rods.Objective -The objective is to determine that the control rods are operable. Specification -The reactor shall not be operated unless the control rods are operable and,a. Control rods shall not be considered operable if damage is apparent to the rod or driveassemblies.

b. The scram time measured from the instant a signal reaches the value of a limiting safetysystem setting to the instant that the slowest control rod reaches its fully inserted position shallnot exceed one (1) second.8 Basis -a. The apparent condition of the control rod assemblies shall provide assurance that the rodsshall continue to perform reliably as designed.

b. This assures that the reactor shall shut down promptly when a scram signal is initiated (SARChapter 13, Section 13.2.2.2.2).

3.2.2 Reactor Instrumentation Applicability -This specification applies to the information which shall be available to the reactoroperator during reactor operations. Objective -The objective is to require that sufficient information is available to the operator toassure safe operation of the reactor.Specification -The reactor shall not be operated unless the channels described in Table 3.2.2are operable and the information is displayed on the reactor console.Table 3.2.2Required Reactor Instrumentation (Minimum Number Operable) Measuring Steady Square Channel Surveillance Channel State Pulse Wave Function Requirements*

a. Reactor Power 2 0 2 Scram at 2.2 D,M,ALevel Safety MW or lessChannelb. Linear Power 10 1Automatic D,M,AChannel Power Controlc. Log Power 10 1Startup D,M,AChannel Controld. Fuel Temperature 2 2 2Fuel D,M,AChannel Temperature

e. Pulse Channel 0 10Measures P,APulse NV & NVT(*) Where: 0 -Channel check during each day's operation M -Channel test monthlyA -Channel calibration annuallyP -Channel test prior to pulsing operation Basis -a. Table 3.2.2. The two reactor power level safety channels assure that the reactor power levelis properly monitored and indicated in the reactor control room (SAR Chapter 7, Sections 7.1.2 &7.1.2.2).

b. c. & e. Table 3.2.2. The linear power channel, log power channel, and pulse channel assurethat the reactor power level and energy are adequately monitored (SAR Chapter 7, Sections7.1.2 & 7.1.2.2).

9

d. Table 3.2.2. The fuel temperature channels assure that the fuel temperature is properlymonitored and indicated in the reactor control room (SAR Chapter 4, Section 4.5.4.1).

3.2.3 Reactor Scrams and Interlocks Applicability -This specification applies to the scrams and interlocks. Objective -The objective is to assure that the reactor is placed in the shutdown condition promptly and that the scrams and interlocks are operable for safe operation of the reactor.Specification -The reactor shall not be operated unless the scrams and interlocks described inTable 3.2.3 are operable: Table 3.2.3Required Scrams and Interlocks Scram.a. ConsoleManualScramb. Reactor RoomManual Scramc. Radiography Bay ManualScramsd. Reactor PowerLevel SafetyScramse. High VoltagePower SuppliesScramsf. FuelTemperature Scramsg. WatchdogCircuitSteadyState142Pulse1402SquareWave1222Manual Scramand Automatic Scram AlarmManual Scramand Automatic Scram AlarmManual Scramsand Automatic Scram AlarmsAutomatic Scram Alarms & Scramsat 2.2 MW or lessAutomatic Scram Alarms &Scrams on Loss ofHigh Voltage tothe Reactor PowerLevel SafetyChannelsAutomatic ScramAlarms & Scramson indicated fueltemperature of750°C or lessAutomatic ScramAlarms & ScramsMMMMMMMChannelFunctionSurveillance Requirements* 22210

h. ExternalScramsi. One KilowattPulse &Square WaveInterlock

j. Low SourceLevel RodWithdrawal ProhibitInterlock

k. Control RodWithdrawal Interlock I. MagnetPower KeySwitch Scram220IIIAutomatic Scrams and Alarmsif an experiment or radiography scram interlock is activated Prevents initiation of a step reactivity insertion above areactor power levelof I KWPrevents withdrawal of any control rodif the log channelreads less than 1.5times the indicated log channel currentlevel with the neutronsource removed fromthe corePrevents simul-taneous withdrawal of two or more rodsin manual modeDe-energizes thecontrol rodmagnets, scram &alarmMMMMMIIII(*) Where: M -channel test monthlyBasis -a. Table 3.2.3. The console manual scram allows rapid shutdown of the reactor from the controlroom (SAR Chapter 7, Section 7.1.2.5).

b. Table 3.2.3. The reactor room manual scram allows rapid shutdown of the reactor from thereactor room.c. Table 3.2.3. The radiography bay manual scrams allow rapid shutdown of the reactor fromany of the radiography bays (SAR Chapter 9, Section 9.6.3).d. Table 3.2.3. The automatic power level safety scram assures the reactor will be shutdown ifthe power level exceeds 2.2 MW, therefore not exceeding the safety limit (SAR Chapter 4,Section 4.7.2).e. Table 3.2.3. The loss-of-high-voltage scram assures that the reactor power level safetychannels operate within their intended range as required for proper functioning of the powerlevel scrams (SAR Chapter 7, Sections 7.1.2.1 & 7.1.2.2).

f. Table 3.2.3. The fuel temperature scrams assure that the reactor will be shut down if the fueltemperature exceeds 7500° C, therefore ensuring the safety limit will not be exceeded (SARChapter 4, Sections 4.5.4.1 & 4.7.2).11

g. Table 3.2.3. The watchdog circuits assure that the control system computer and the dataacquisition computer are functioning properly (SAR Chapter 7, Section 7.2).h. Table 3.2.3. The external scrams assure that the reactor will be shut down if the radiography bay doors and reactor concrete shutters are not in the proper position for personnel entry intothe bays (SAR Chapter 9, Section 9.6). External scrams from experiments, a subset of theexternal scrams, also assure the integrity of the reactor system, the experiment, the facility, andthe safety of the facility personnel and the public.i. Table 3.2.3. The interlock preventing the initiation of a step reactivity insertion at a level aboveone (1) kilowatt assures that the pulse magnitude will not allow the fuel element temperature toexceed the safety limit (SAR Chapter 7, Section 7.1.2.5).

j.Table 3.2.3. The low source level rod withdrawal prohibit interlock assures an adequate sourceof neutrons is present for safe startup of the reactor (SAR Chapter 7, Section 7.1.2.5).

k. Table 3.2.3. The control rod withdrawal interlock prevents the simultaneous withdrawal of twoor more control rods thus limiting the reactivity-insertion rate from the control rods in manualmode (SAR Chapter 7, Section 7.1.2.5).

I. Table 3.2.3. The magnet current key switch prevents the control rods from being energized without inserting the key. Turning off the magnet current key switch de-energizes the control rodmagnets and results in a scram (SAR Chapter 7, Section 7.1.2.5). 3.2.4 Reactor Fuel ElementsApplicability -This specification applies to the physical dimensions of the fuel elements asmeasured on the last surveillance test.Objective -The objective is to verify the integrity of the fuel-element cladding. Specification -The reactor shall not be used for normal operation with damaged fuel. All fuelelements shall be inspected visually for damage or deterioration as per Technical Specifications Section 4.2.4. A fuel element shall be considered damaged and must be removed from the coreif:a. In measuring the transverse bend, the bend exceeds 0.125 inch (3.175 mm) over the fulllength 23 inches (584 mm) of the cladding, or,b. In measuring the elongation, its length exceeds its initial length by 0.125 inch (3.175 mm), or,c. A cladding failure exists as indicated by measurable release of fission products, or,d. Visual inspection identifies bulges, gross pitting, or corrosion. Basis -The most severe stresses induced in the fuel elements result from pulse operation of thereactor, during which differential expansion between the fuel and the cladding occurs and thepressure of the gases within the elements increases sharply. The above limits on the allowable distortion of a fuel element correspond to strains that are considerably lower than the strainexpected to cause rupturing of a fuel element. Limited operation in the steady state or pulsedmode may be necessary to identify a leaking fuel element especially if the leak is small.12 3.3 Reactor Coolant SystemsApplicability -These specifications apply to the operation of the reactor water measuring systems.Objective -The objective is to assure that adequate cooling is provided to maintain fuel temperatures below the safety limit, and that the water quality remains high to prevent damage to the reactor fuel.Specification -The reactor shall not be operated unless the systems and instrumentation channelsdescribed in Table 3.3 are operable, and the information is displayed locally or in the control room.Table 3.3REQUIRED WATER SYSTEMS AND INSTRUMENTATION Measuring Channel/System

a. Primary CoolantCore InletTemperature Monitorb. Reactor TankLow WaterMonitorc. Purification**

Inlet Conduc-tivity Monitord. Emergency CoreCooling SystemMinimumNumberOperableSurveillance Requirements* 1IIIFunction: Channel/System For operation of thereactor at 1.5 MW orhigher, alarms on highheat exchanger outlettemperature of 45 °C(113°F)Alarms if water leveldrops below a depth of23 feet in the reactor tankAlarms if the primarycoolant water conductivity is greater than5 micromhos/cm For operation of the reactorat 1.5MW or higher, provideswater to cool fuel in the eventof a Loss of Coolant Accidentfor a minimum of 3.7 hoursat 20 gpm from an appropriate nozzleD,Q,AMD,M,SD,S(*) Where: D -channel check during each day's operation A -channel calibration annuallyQ -channel test quarterly S -channel calibration semiannually M -channel test monthly(**) The purification inlet conductivity monitor can be out-of-service for no more than 3 hours before the reactorshall be shutdown. Basis -a. Table 3.3. The primary coolant core inlet temperature alarm assures that large power fluctuations willnot occur (SAR Chapter 4, Section 4.6.2).13

b. Table 3.3. The minimum height of 23 ft. of water above the reactor tank bottom guarantees that thereis sufficient water for effective cooling of the fuel and that the radiation levels at the top of the reactortank are within acceptable limits. The reactor tank water level monitor alarms if the water level dropsbelow a height of 23 ft. (7.01m) above the tank bottom (SAR Chapter 11, Section 11.1.5.1).

c. Table 3.3. Maintaining the primary coolant water conductivity below 5 micromhos/cm averaged over aweek will minimize the activation of water impurities and also the corrosion of the reactor structure.

d. Table 3.3. This system will mitigate the Loss of Coolant Accident event analyzed in the SAR Chapter13, Section 13.2.3.4 Reactor Room Exhaust SystemApplicability

-These specifications apply to the operation of the reactor room exhaust system.Objective -The objectives of this specification are as follows:a. To reduce concentrations of airborne radioactive material in the reactor room, and maintain thereactor room pressure negative with respect to surrounding areas.b. To assure continuous air flow through the reactor room in the event of a Loss of Coolant Accident. S~pecification -a. The reactor shall not be operated unless the reactor room exhaust system is in operation and thepressure in the reactor room is negative relative to surrounding areas.b. The reactor room exhaust system shall be operable within one half hour of the onset of a Loss ofCoolant Accident. Basis -Operation of the reactor room exhaust system assures that:a. Concentrations of airborne radioactive material in the reactor room and in air leaving the reactor roomwill be reduced due to mixing with exhaust system air (SAR Chapter 9, Section 9.5.1). Pressure in thereactor room will be negative relative to surrounding areas due to air flow patterns created by the reactorroom exhaust system (SAR Chapter 9, Section 6.5.1).b. There will be a timely, adequate and continuous air flow through the reactor room to keep the fueltemperature below the safety limit in the event of a Loss of Coolant Accident. 3.5 This section intentionally left blank.3.6 This section intentionally left blank.3.7 Reactor Radiation Monitoring Systems3.7.1 Monitoring SystemsApplicability -This specification applies to the information which shall be available to the reactoroperator during reactor operation. Obiective -The objective is to require that sufficient information regarding radiation levels andradioactive effluents is available to the reactor operator to assure safe operation of the reactor.Specification -The reactor shall not be operated unless the channels described in Table 3.7.1are operable, the readings are below the alarm setpoints, and the information is displayed in the14

• 0control room. The stack and reactor room CAMS shall not be shutdown at the same time duringreactor operation.

Table 3.7.1REQUIRED RADIATION MONITORING INSTRUMENTATION Measuring Equipment MinimumNumberOperable** ChannelFunctionSurveillance Requirements*

a. FacilityStack Monitorb. Reactor RoomRadiation Monitorc. Purification System Radia-tion Monitord. Reactor RoomContinuous Air MonitorIIIIMonitors Argon-41 andradioactive particu-lates, and alarmsMonitors the radiation level in the reactorroom and alarmsMonitors radiation level at the demineral-izer station and alarmsMonitors air from thereactor room for parti-culate and gaseousradioactivity and alarmsD,W,AD,W,AD,W,AD,W,A(*) Where: D -channel check during each day's operation A -channel calibration annuallyW -channel test(**) monitors may be placed out-of-service for up to 2 hours for calibration and maintenance.

During this out-of-service time, no experiment or maintenance activities shall be conducted which could result in alarm conditions (e.g., airborne releases or high radiation levels)Basis -a. Table 3.7.1. The facility stack monitor provides information to operating personnel regarding therelease of radioactive material to the environment (SAR Chapter 11, Section 11.1.1.1.4). The alarmsetpoint on the facility stack monitor is set to limit Argon-41 concentrations to less than 10 CFR Part 20,Appendix B, Table 2, Column 1 values (averaged over one year) for unrestricted locations outside theoperations area.b. Table 3.7.1. The reactor room radiation monitor provides information regarding radiation levels in thereactor room during reactor operation (SAR Chapter 11, Section 11.1.5.1), to limit Occupational radiation exposure to less than 10 CFR 20 limits.c. Table 3.7.1. The radiation monitor located next to the purification system resin cannisters providesinformation regarding radioactivity in the primary system cooling water (SAR Chapter 11, Section11.1.5.4.2) and allows assessment of radiation levels in the area to ensure that personnel radiation doses will be below 10 CFR Part 20 limits.d. Table 3.7.1. The reactor room continuous air monitor provides information regarding airborneradioactivity in the reactor room, (SAR Chapter 11, Sections 11.1.1.1.2 & 11.1.1.1.5), to ensure thatoccupational exposure to airborne radioactivity will remain below the 10 CFR Part 20 limits.15 3.7.2 Effluents -Arqon-41 Dischargle LimitApplicability -This specification applies to the concentration of Argon-41 that may be discharged from the UCD/MNRC reactor facility. Objective -The objective is to ensure that the health and safety of the public is not endangered by the discharge of Argon-4l from the UCD/MNRC reactor facility. Specification -The annual average unrestricted area concentration of Argon-41 due to releasesof this radionuclide from the UCD/MNRC, and the corresponding annual radiation dose fromArgon-4l in the unrestricted area shall not exceed the applicable levels in 10 CFR Part 20.Basis -The annual average concentration limit for Argon-41 in air in the unrestricted area isspecified in Appendix B, Table 2, Column 1 of 10 CFR Part 20. 10 CFR 20.1301 specifies doselimitations in the unrestricted area. 10 CFR 20.1101 specifies a constraint on air emissions ofradioactive material to the environment. The SAR Chapter 11, Section 11.1.1.1.4 estimates thatthe routine Argon-4l releases and the corresponding doses in the unrestricted area will bebelow these limits.3.8 Exp~eriments 3.8.1 Reactivity Limits.Applicability -This specification applies to the reactivity limits on experiments installed in specificreactor experiment facilities. Objective -The objective is to assure control of the reactor during the irradiation or handling ofexperiments in the specifically designated reactor experiment facilities. Specification -The reactor shall not be operated unless the following conditions governing experiments exist:a. The absolute reactivity worth of any single moveable experiment in the pneumatic transfertube, the central irradiation

facility, the central irradiation fixture 1 (CIF-1),

or any other in-core orin-tank irradiation

facility, shall be less than $1.00 (0.7% A k/k), except for the automated centralirradiation facility (ACIF) (See 3.8.1.c below).b. The absolute reactivity worth of any single secured experiment positioned in a reactor in-coreor in-tank irradiation facility shall be less than the maximum allowed pulse ($1.75) (1.23% A k/k).c. The absolute total reactivity worth of any single experiment or of all experiments collectively positioned in the ACIF shall be less than the maximum allowed pulse ($1.75) (1.23% A k/k).d. The absolute total reactivity of all experiments positioned in the pneumatic transfer tube, andin any other reactor in-core and in-tank irradiation facilities at any given time shall be less thanone dollar and ninety-two cents ($1.92) (1.34% A k/k), including the potential reactivity whichmight result from malfunction,

flooding, voiding, or removal and insertion of the experiments.

Basis -a. A limitation of less than one dollar ($1.00) (0.7% A k/k) on the reactivity worth of a singlemovable experiment positioned in the pneumatic transfer tube, the central irradiation facility(SAR, Chapter 10, Section 10.4.1), the central irradiation fixture-I (ClF-1) (SAR Chapter 10,Section 10.4.1), or any other in-core or in-tank irradiation

facility, will assure that the pulse limitof $1.75 is not exceeded (SAR Chapter 13, Section 13.2.2.2.1).

In addition, limiting the worth ofeach movable experiment to less than $1.00 will assure that the additional increase in transient 16

• 0power and temperature will be slow enough so that the fuel temperature scram will be effective (SAR Chapter 13, Section 13.2.2.2.1

).b. The absolute worst event which may be considered in conjunction with a single securedexperiment is its sudden accidental or unplanned removal while the reactor is operating. Forsuch an event, the reactivity limit for fixed experiments ($1.75) would result in a reactivity increase less than the $1.92 pulse reactivity insertion needed to reach the fuel temperature safety limit (SAR Chapter 13, Section 13.2.2.2.1 ).c. A reactivity limit of less than $1.75 for any single experiment or for all experiments collectively positioned in the sample can of the automated central irradiation facility (ACIF) (SAR Chapter10, Section 10.4.2) is based on the pulsing reactivity insertion limit (Technical Specification 3.1.2) (SAR Chapter 13, Section 13.2.2.2.1) and on the design of the ACIF, which allows controlover the positioning of samples into and out of the central core region in a manner identical inform, fit, and function to a control rod.d. It is conservatively assumed that simultaneous removal of all experiments positioned in thepneumatic transfer tube, and in any other reactor in-core and in-tank irradiation facilities at anygiven time shall be less thanthe maximum reactivity insertion limit of $1.92. The SAR Chapter13, Section 13.2.2.2.1 indicates that a pulse reactivity insertion of $1.92 would be needed toreach the fuel temperature safety limit.3.8.2 Materials LimitApplicability -This specification applies to experiments installed in reactor experiment facilities. Objective -The objective is to prevent damage to the reactor or significant releases ofradioactivity by limiting material quantity and the radioactive material inventory of theexperiment. Specification -The reactor shall not be operated unless the following conditions governing experiment materials exist:a. Experiments containing materials corrosive to reactor components, compounds highlyreactive with water, potentially explosive materials, and liquid fissionable materials shall beappropriately encapsulated.

b. Each fueled experiment shall be controlled such that the total inventory of iodine isotopes 131through 135 in the experiment is no greater than 1.5 curies and the maximum strontium inventory is no greater than 5 millicuries.

c. Each experiment in the 1-125 production facility shall be controlled such that the totalinventory of 1-125 in the 1-125 glove box shall not exceed 40 curies.d. Each experiment in the 1-125 production facility shall be controlled such that the totalinventory of 1-125 being handled in the 1-125 fume hood at any one time in preparation forshipping shall not exceed 20 curies. An Additonal 1.0 curie of 1-125 (up to 400 millicuries in theform of quality assurance samples and up to 600 millicuries in sealed storage containers) mayalso be present in the 1-125 fume hood.e. Explosive materials in quantities greater than 25 milligrams of TNT equivalent shall not beirradiated in the reactor tank. Explosive materials in quantities of 25 milligrams of TNTequivalent or less may be irradiated provided the pressure produced upon detonation of theexplosive has been calculated and/or experimentally demonstrated to be less than the designpressure of the container.

f. Explosive materials in quantities of three (3) pounds of TNT equivalent or less may beirradiated in any radiography bay. The irradiation of explosives in any bay is limited to those1"7 assemblies where a safety analysis has been performed that shows that there is no damage tothe reactor safety systems upon detonation (SAR Chapter 13, Section 13.2.6.2).

Basis -a. Appropriate encapsulation is required to lessen the experimental hazards of some types ofmaterials.

b. The 1.5 curies limitation on iodine 131 through 135 assures that in the event of failure of afueled experiment leading to total release of the iodine, occupational doses and doses tomembers of the general public in the unrestricted areas shall be within the limits in 10 CFR 20(SAR Chapter 13, Section 13.2.6.2).

c&d. Limiting the total 1-125 inventory to forty (40.0) curies in the 1-125 glove box and to twenty-one (21.0) curies in the 1-125 fume hood assures that, if either of these inventories of 1-125 istotally released into its respective containment, or if both inventories are simultaneously released into their respective containments, the occupational doses and doses to members ofthe general public in the unrestricted areas will be within the limits of 10 CFR 20 (SAR Chapter13, Section 13.2.6.2).

e. This specification is intended to prevent damage to vital equipment by restricting the quantityof explosive materials within the reactor tank (SAR Chapter 13, Section 13.2.6.2).

f. The failure of an experiment involving the irradiation of 3 lbs TNT equivalent or less in anyradiography bay external to the reactor tank will not result in damage to the reactor controls orthe reactor tank. Safety Analyses have been performed (SAR Chapter 13, Section 13.2.6.2) which show that up to six (6) pounds of TNT equivalent can be safely irradiated in anyradiogaphy bay. Therefore, the three (3) pound limit gives a safety margin of two (2).3.8.3 Failure and Malfunctions Applicability

-This specification applies to experiments installed in reactor experiment facilities. Objective -The objective is to prevent damage to the reactor or significant releases ofradioactive materials in the event of an experiment failure.Specification -a. All experiment materials which could off-gas,

sublime, volatilize, or produce aerosols under:(1) normal operating conditions of the experiment or the reactor,(2) credible accident conditions in the reactor, or(3) where the possibility exists that the failure of an experiment could release radioactive gases or aerosols into the reactor building or into the unrestricted area, the quantity andtype of material in the experiment shall be limited such that the airborne radioactivity inthe reactor room will not result in exceeding the applicable dose limits in 10 CFR Part 20in the unrestricted area, assuming 100% of the gases or aerosols escapes.b. In calculations pursuant to (a) above, the following assumptions shall be used:(1) If the effluent from an experiment facility exhausts through a stack which is closed onhigh radiation levels, at least 10% of the gaseous activity or aerosols produced willescape.18 (2) If the effluent from an experiment facility exhausts through a filter installation designed for greater than 99% efficiency for 0.3 micron and larger particles, at least 10%of these will escape.(3) For materials whose boiling point is above 130 °C and where vapors formed byboiling this material can escape only through an undistributed column of water above thecore, at least 10% of these vapors can escape.c. If a capsule fails and releases material which could damage the reactor fuel or structure bycorrosion or other means, an evaluation shall be made to determine the need for corrective action. Inspection and any corrective action taken shall be reviewed by the UCD/MNRC Directoror his designated alternate and determined to be satisfactory before operation of the reactor isresumed.Basis -a. This specification is intended to reduce the likelihood that airborne radioactivity in the reactorroom or the unrestricted area will result in exceeding the applicable dose limits in 10 CFR Part20.b. These assumptions are used to evaluate the potential airborne radioactivity release due to anexperiment failure (SAR Chapter 13, Section 13.2.6.2).

c. Normal operation of the reactor with damaged reactor fuel or structural damage is prohibited to avoid release of fission products.

Potential damage to reactor fuel or structure shall bebrought to the attention of the UCD/MNRC Director or his designated alternate for review toassure safe operation of the reactor (SAR Chapter 13, Section 13.2.6.2). 4.0 Surveillance Requirements General. The surveillance frequencies denoted herein are based on continuing operation of the reactor.Surveillance activities scheduled to occur during an operating cycle which can not be performed with the reactoroperating may be deferred to the end of the operating cycle. If the reactor is not operated for a reasonable time,a reactor system or measuring channel surveillance requirement may be waived during the associated timeperiod. Prior to reactor system or measuring channel operation, the surveillance shall be performed for eachreactor system or measuring channel for which surveillance was waived. A reactor system or measuring channel shall not be considered operable until it is successfully tested.4.1 Reactor Core Parameters 4.1 .1 Steady State Operation Applicability -This specification applies to the surveillance requirement for the power levelmonitoring channels. Obiective -The objective is to verify that the maximum power level of the reactor does notexceed the authorized limit.Specification -An annual channel calibration shall be made of the power level monitoring channel. If a channel is removed,

replaced, or unscheduled maintenance is performed, or asignificant change in core configuration occurs, a channel calibration shall be required.

Discovery of noncompliance with this specification shall limit reactor operations to that requiredto perform the surveillance. Basis -The annual power level channel calibration will assure that the indicated reactor powerlevel is correct.4.1.2 Shutdown Margin and Excess Reactivity 19

• 0Applicability

-These specifications apply to the surveillance requirements for reactivity control ofthe reactor core.Objective -The objective is to measure and verify the reactivity worth, performance, andoperability of those systems affecting the reactivity of the reactor.Specification -a. The total reactivity worth of each control rod and the shutdown margin shall be determined annually or following any significant change in core or control rod configuration. The shutdownmargin shall be verified by meeting the requirements of Section 3.1.3(a).

b. The core excess reactivity shall be verified:

(1) Prior to each startup operation and,(2) Following any change in core loading or configuration. Discovery of noncompliance with Technical Specifications 4.1 .2.a-b shall limit reactor operations to that requiredto perform the surveillance. Basis -a. The reactivity worth of the control rods is measured to assure that the required shutdownmargin is available and to provide an accurate means for determining the excess reactivity ofthe core. Past experience with similar reactors gives assurance that measurements of thecontrol rod reactivity worth on an annual basis is adequate to assure that there are no significant changes in the shutdown margin, provided no core loading or configuration changes have beenmade.b. Determining the core excess reactivity prior to each reactor startup shall assure that Technical Specifications 3.1.3.b shall be met, and that the critical rod positions do not changeunexpectedly. 4.2 Reactor Control and Safety Systems4.2.1 Control RodsApplicability -This specification applies to the surveillance of the control rods.Objective -The objective is to inspect the physical condition of the reactor control rods andestablish the operable condition of the rods.Specification -Control rod worths shall be determined annually or after physical removal or anysignificant change in core or control rod configuration.

a. Each control rod shall be inspected at annual intervals by visual observation of the fueledsections and absorber sections plus examination of the linkages and drives.b. The scram time of each control rod shall be measured semiannually.

Discovery of noncompliance with Technical Specifications 4.2.1 .a-b shall limit reactor operations to that requiredto perform the surveillance. Basis (Technical Specifications 4.2.1 .a-b) -Annual determination of control rod worths ormeasurements after any physical removal or significant change in core loading or control rod20 configuration provides information about changes in reactor total reactivity and individual rodworths. The frequency of inspection for the control rods shall provide periodic verification of thecondition of the control rod assemblies. The specification intervals for scram time assureoperable performance of the control rods.4.2.2 Reactor Instrumentation Applicability -These specifications apply to the surveillance requirements for measurements, tests, calibration and acceptability of the reactor instrumentation. .Objective -The objective is to ensure that the power level instrumentation and the fueltemperature instrumentation are operable. Specification -a. The reactor power level safety channels shall have the following: (1) A channel test monthly or after any maintenance which could affect their operation. (2) A channel check during each day's operation. (3) A channel calibration annually.

b. The Linear Power Channel sh'all have the following:

(1) A channel test monthly or after any maintenance which could affect the operation. (2) A channel check during each day's operation. (3) A channel calibration annually.

c. The Log Power Channel shall have the following:

(1) A channel test monthly or after any maintenance which could affect its operation. (2) A channel check during each day's operation. (3) A channel calibration annually.

d. The fuel temperature measuring channels shall have the following:

(1) A channel test monthly or after any maintenance which could affect operation. (2) A channel check during each day's operation. (3) A channel calibration annually.

e. The Pulse Energy Integrating Channel shall have the following:

(1) A channel test prior to PUlsing operations. (2) A channel calibration annually. Discovery of noncompliance with Technical Specifications 4.2.2.a-e shall limit reactor operation to that requiredto perform the surveillance. Basis -21

a. A daily channel check and monthly test, plus the annual calibration, will assure that thereactor power level safety channels operate properly.

b. A channel test monthly of the reactor power level multi-range channel will assure that thechannel is operable and responds correctly.

The channel check will assure that the reactorpower level multi-range linear channel is operable on a daily basis. The channel calibration annually of the multi-range linear channel will assure that the reactor power will be accurately measured so the authorized power levels are not exceeded.

c. A channel test monthly will assure that the reactor power level wide range log channel isoperable and responds correctly.

A channel check of the reactor power level wide range logchannel will assure that the channel is operable on a daily basis. A channel calibration willassure that the channel will indicate properly at the corresponding power levels.d. A channel test monthly and check during each day's operation, plus the annual calibration, willassure that the fuel temperature measuring channels operate properly.

e. A channel test prior to pulsing plus the annual channel calibration will assure the pulse energyintegrating channel operates properly.

4.2.3 Reactor Scrams and Interlocks Applicability -These specifications apply to the surveillance requirements for measurements, test, calibration, and acceptability of the reactor scrams and interlocks. Objective -The objective is to ensure that the reactor scrams and interlocks are operable. Specification -a. Console Manual Scram. A channel test shall be performed monthly.b. Reactor Room Manual Scram. A channel test shall be performed monthly.c. Radiography Bay Manual Scrams. A channel test shall be performed monthly.d. Reactor Power Level Safety Scram. A channel test shall be performed monthly.e. High-Voltage-Power Supply Scrams. A channel test shall be performed monthly.f. Fuel Temperature Scram. A channel test shall be performed monthly.g. Watchdog Circuits Scrams. A channel test shall be performed monthly.h. External Scra~ns. A channel test shall be performed monthly.i.The One Kilowatt Pulse Interlock. A channel test shall be performed monthly.j. Low Source Level Rod Withdrawal Prohibit Interlock. A channel test shall be performed monthly.k. Control Rod Withdrawal Interlocks. A channel test shall be performed monthly.I. Magnet Power Key Switch Scram. A channel test shall be performed monthly.Discovery of noncompliance with Specifications 4.2.3.a-I shall limit reactor operation to that required to performthe surveillance. Basis -22 0a. A channel test monthly of the Console Manual Scram will assure that the scram is operable.

b. A channel test monthly of the Reactor Room Manual Scram will assure that the scram isoperable.

c. A channel test monthly of the Radiography Bay Manual Scrams will assure that the scramsare operable.

d. A channel test monthly of the Reactor Power Level Safety Scrams will assure that the scramsare operable.

e. A channel test monthly of the Loss-of-High-Voltage Scram will assure that the high voltagepower supplies are operable and respond correctly.

f. A channel test monthly of the Fuel Temperature Scrams will assure that the scrams areoperable.

g. A channel test monthly of the Watchdog Circuits Scrams will assure that the scram circuits areoperable.

h. A channel test monthly of the External Scrams will assure that the scrams are operable andrespond correctly.

i.A channel test monthly will assure that the One Kilowatt Pulse Interlock works properly.

j. A channel test monthly of the Low Source Level Rod Withdrawal Proh~ibit Interlock will assurethat the interlock is operable.

k. A channel test monthly of the Control Rod Withdrawal Interlock will assure that the interlock isoperable.

I. A channel test monthly of the Magnet Current Key Switch will assure that the scram isoperable. 4.2.4 Reactor Fuel ElementsApplicability -This specification applies to the surveillance requirements for the fuel elements. Objective -The objective is to verify the continuing integrity of the fuel element cladding. Specification -To assure the measurement limitations in Section 3.2.4 are met, the following shall be done:a. The lead elements (i.e., all elements adjacent to the transient rod, with the exception ofinstrumented fuel elements), and all elements adjacent to the central irradiation facility shall beinspected annually.

b. Instrumented fuel elements shall be inspected if any of the elements adjacent to it fail to passthe visual and/or physical measurement requirements of Section 3.2.4. Discovery ofnoncompliance with Technical Specification 4.2.4 shall limit operations to that required toperform the surveillance.

Basis (Technical Specifications 4.2.4.a-b) -The above specifications assure that the lead fuelelements shall be inspected regularly and the integrity of the lead fuel elements shall bemaintained. These are the fuel elements with the highest power density as analyzed in the SARChapter 4, Section 4.5.5.6. The instrumented fuel element is excluded to reduce the risk ofdamage to the thermocouples. 23 .0*4.3 Reactor Coolant SystemsApplicability -This specification applies to the surveillance requirements for the reactor water measuring systems and the emergency core cooling system.Objective -The objective is to assure that the reactor tank water temperature monitoring system, thetank water level alarm, the water conductivity cells and the emergency core cooling system are alloperable. Specification -a. The reactor tank core inlet temperature monitor shall have the following: (1) A channel check during each day's operation. (2) A channel test quarterly. (3) A channel calibration annually.

b. The reactor tank low water level monitoring system shall have the following:

(1) A channel test monthly.c. The purification inlet conductivity monitors shall have the following: (1) A channel check during each day's operation. (2) A channel test monthly.(3) A channel calibration semiannually.

d. The Emergency Core Cooling System shall have the following:

(1) A channel check prior to operation. (2) A channel calibration semiannually. Discovery of noncompliance with Technical Specifications 4.3.a-c shall limit operations to that required toperform the surveillance. Noncompliance with Technical Specification 4.3.d shall limit operations to less than 1.5MW.Basis -a. A channel test quarterly assures the water temperature monitoring system responds correctly to aninput signal. A channel check during each day's operation assures the channel is operable. A channelcalibration annually assures the monitoring system reads properly.

b. A channel test monthly assures that the low water level monitoring system responds correctly to aninput signal.c. A channel test monthly assures that the purification inlet conductivity monitors respond correctly to aninput signal. A channel check during each day's operation assures that the channel is operable.

Achannel calibration semiannually assures the conductivity monitoring system reads properly.

d. A channel check prior to operation assures that the emergency core cooling system is operable forpower levels above 1.5 MW. A channel calibration semiannually assures that the Emergency CoreCooling System performs as required for power levels above 1.5 MW.24 04.4 Reactor Room Exhaust SystemApplicability

-This specification applies to the surveillance requirements for the reactor roomexhaust system.Objective -The objective is to assure that the reactor room exhaust system is operating properly. Specification -The reactor room exhaust system shall have a channel check during each day'soperation. Discovery of noncompliance with this specification shall limit operations to that required to perform thesurveillance. Basis -A channel check during each day's operation of the reactor room exhaust system shallverify that the exhaust system is maintaining a negative pressure in the reactor room relative tothe surrounding facility areas.4.5 This section intentionally left blank4.6 This section intentionally left blank.4.7 Reactor Radiation Monitoring SystemsApplicability -This specification applies to the surveillance requirements for the reactor radiation monitoring systems.Obiective -The objective is to assure that the radiation monitoring equipment is operating properly. Specification -a. The facility stack monitor shall have the following: (1) A channel check during each day's operation. (2) A channel test weekly.(3) A channel calibration annually.

b. The reactor room radiation monitor shall have the following:

(1) A channel check during each day's operation. (2) A channel test weekly.(3) A channel calibration annually.

c. The purification system radiation monitor shall have the following:

(1) A channel check during each day's operation. (2) A channel test weekly.(3) A channel calibration annually.

d. The reactor room Continuous Air Monitor (CAM) shall have the following:

25 (1) A channel check during each day's operation. (2) A channel test weekly.(3) A channel calibration annually. Discovery of noncompliance with Technical Specifications 4.7.a-d shall limit operations to that required toperform the surveillance. Basis -a. A channel check of the facility stack monitor system during each day's operation will assurethe monitor is operable. A channel test weekly will assure that the system responds correctly toa known source. A channel calibration annually will assure that the monitor reads correctly.

b. A channel check of the reactor room radiation monitor during each day's operation will assurethat the monitor is operable.

A channel test weekly will ensure that the system responds to aknown source. A channel calibration of the monitor annually will assure that the monitor readscorrectly.

c. A channel check of the purification system radiation monitor during each day's operation assures that the monitor is operable.

A channel test weekly will ensure that the systemresponds to a known source. A channel calibration of the monitor annually will assure that themonitor reads correctly.

d. A channel check of the reactor room Continuous Air Monitor (CAM) during each day'soperation will assure that the CAM is operable.

A channel test weekly will assure that the CAMresponds correctly to a known source. A channel calibration annually will assure that the CAMreads correctly.

4.8 Experiments

Applicability -This specification applies to the surveillance requirements for experiments installed in any UCD/MNRC reactor experiment facility. Objective -The objective is to prevent the conduct of experiments or irradiations which maydamage the reactor or release excessive amounts of radioactive materials as a result ofexperimental-failure.Specification

a. A new experiment shall not be installed in any UCD/MNRC reactor experiment facility until awritten safety analysis has been performed and reviewed by the UCD/MNRC

Director, or hisdesignee, to establish compliance with the Limitations on Experiments, (Technical Specifications Section 3.8) and 10 CFR 50.59.b. All experiments performed at the UCD/MNRC shall meet the conditions of an approvedFacility Use Authorization.

Facility Use Authorizations and experiments carried out under theseauthorizations shall be reviewed and approved in accordance with the Utilization of the (UCD)McClellan Nuclear Radiation Center Research Reactor Facility Document (MNRC-0027-DOC). An experiment classified as an approved experiment shall not be placed in any UCD/MNRCexperiment facility until it has been reviewed for compliance with the approved experiment andFacility Use Authorization by the Reactor Manager and the Health Physics Manager, or theirdesignated alternates.

c. The reactivity worth of any experiment installed in the pneumatic transfer tube, or in any otherUCD/MNRC reactor in-core or in-tank irradiation facility shall be estimated or measured, as26 0I!appropriate, before reactor operation with said experiment.

Whenever a measurement is done itshall be done at ambient conditions.

d. Experiments shall be identified and a log or other record maintained while experiments are inany UCD/MNRC reactor experiment facility.

Basis -a & b. Experience at most TRIGA reactor facilities verifies the importance of reactor staff andsafety committee reviews of proposed experiments.

c. Measurement of the reactivity worth of an experiment, or estimation of the reactivity worthbased on previous or similar measurements, shall verify that the experiment is within authorized reactivity limits.d. Maintaining a log of experiments while in UCD/MNRC reactor experiment facilities willfacilitate maintaining surveillance over such experiments.

5.0 Design Features5.1 Site and Facility Description. 5.1.1 Sit__eApplicability -This specification applies to the UCD/MNRC site location and specific facilitydesign features. Objective -The objective is to specify those features related to the Safety Analysis evaluation. Specification -a. The site location is situated approximately 8 miles (13 kin) north-by-northeast of downtownSacramento, California. The site of the UCD/MNRC facility is about 3000 ft. (0.6 mi or 0.9 kin)west of Watt Avenue, and 4500 ft. (0.9 mi or 1.4 kin) south of E Street.b. The restricted area is that area inside the fence surrounding the reactor building. Theunrestricted area is that area outside the fence surrounding the reactor building.

c. The TRIGA reactor is located in Building 258, Room 201 of the UCD/MNRC.

This buildinghas been designed with special safety features.

d. The core is below ground level in a water filled tank and surrounded by a concrete shield.Basis -a. Information on the surrounding population, the hydrology, seismology, and climatography ofthe site has been presented in Chapter 2 of the Safety Analysis Report.b. The restricted area is controlled by the UCD/MNRC Director.

c. The room enclosing the reactor has been designed with systems related to the safe operation of the facility.

d. The below grade core design is to negate the consequences of an aircraft hitting the reactorbuilding.

This accident was analyzed in Chapter 13 of the Safety Analysis Report, and found tobe beyond a credible accident scenario. 27 0 05.1.2 Facility ExhaustApplicability -This specification applies to the facility which houses the reactor.Objective -The objective is to assure that provisions are made to restrict the amount ofradioactivity released into the environment, or during a Loss of Coolant Accident, the system isto assure proper removal of heat from the reactor room.Specification -a. The UCD/MNRC reactor facility shall be equipped with a system designed to filter andexhaust air from the UCD/MNRC facility. The system shall have an exhaust stack height of aminimum of 18.2m (60 feet) above ground level.b. Manually activated shutdown controls for the exhaust system shall be located in the reactorcontrol room.Basis -The UCD/MNRC facility exhaust system is designed such that the reactor room shall bemaintained at a negative pressure with respect to the surrounding areas. The free air volumewithin the UCD/MNRC facility is confined to the facility when there is a shutdown of the exhaustsystem. Controls for startup, filtering, and normal operation of the exhaust system are located inthe reactor control room. Proper handling of airborne radioactive materials (in emergency situations) can be conducted from the reactor control room with a minimum of exposure tooperating personnel. 5.2 Reactor Coolant SystemApplicability -This specification applies to the reactor coolant system.Obiective -The objective is to assure that adequate water is available for cooling and shielding duringnormal reactor operation or during a Loss of Coolant Accident. Specification -a. During normal reactor operation the reactor core shall be cooled by a natural convection flow ofwater.b. The reactor tank water level alarm shall activate if the water level in the reactor tank drops below adepth of 23 ft.c. For operations at 1.5 MW or higher during a Loss of Coolant Accident the reactor core shall be cooledfor a minimum of 3.7 hours at 20 gpm by a source of water from the Emergency Core Cooling System.Basis -a. The SAR Chapter 4, Section 4.6, Table 4-19, shows that fuel temperature limit of 930 °C will not beexceeded under natural convection flow conditions.

b. A reactor tank water low level alarm sounds when the water level drops significantly.

This alarmannunciates in the reactor control room and at a 24 hour monitored location so that appropriate corrective action can be taken to restore water for cooling and shielding.

c. The SAR Chapter 13, Section 13.2, analyzes the requirements for cooling of the reactor fuel andshows that the fuel safety limit is not exceeded under Loss of Coolant Accident conditions during thiswater cooling.5.3 Reactor Core and Fuel28

• 05.3.1 Reactor CoreApplicability

-This specification applies to the configuration of the fuel.Objective -The objective is to assure that provisions are made to restrict the arrangement of fuelelements so as to provide assurance that excessive power densities will not be produced. Specification -For operation at 0.5 MW or greater, the reactor core shall be an arrangement of96 or more fuel elements to include fuel followed control rods. Below 0.5 MW there is nominimum required number of fuel elements. In a mixed 20/20, 30/20 and 8.5/20 fuel loading(SAR Chapter 4, Section 4.5.5.6): Mix J Core and Other Variations (1) No fuel shall be loaded into Hex Rings A or B.(2) A fuel followed control rod located in an 8.5 wt% environment shall contain 8.5 wt% fuel.20E Core and Other Variations (1) No fuel shall be loaded into Hex Rings A or B.(2) Fuel followed control rods may contain either 8.5 wt% or 20 wt% fuel.(3) Variations to the 20E core having 20 wt% fuel in Hex Ring C requires the 20 wt% fuel to beloaded into corner positions ony and graphite dummy elements in the flat positions. Theperformance of fuel temperature measurements shall apply to variations to the as-analyzed 20Ecore configurations. 30B Core and Other Variations (1) No fuel shall be loaded into Hex Rings A or B.(2) The only fuel types allowed are 20/20 and 30/20.(3) 20/20 fuel may be used in any position in Hex Rings C through G.(4) 30/20 fuel may be used in any position in Hex Rings 0 through G but not in Hex Ring C.(5) An analysis of any irradiation facility installed in the central cavity of this core shall be donebefore it is used with this core.Basis -In order to meet the power density requirements discussed in the SAR Chapter 4,Section 4.5.5.6, no less than 96 fuel elements including fuel followed control rods and the aboveloading restrictions will be allowed in an operational 0.5 MW or greater core. Specifications forthe 20E core and for the 30B core allow for variations of the as-analyzed core with the condition that temperature limits are being maintained (SAR Chapter 4, Section 4.5.5.6 and ArgonneNational Laboratory Report AN L/ED 97-54).5.3.2 Reactor FuelApplicability -These specifications apply to the fuel elements used in the reactor core.Obiective -The objective is to assure that the fuel elements are of such design and fabricated insuch a manner as to permit their use with a high degree of reliability with respect to theirphysical and nuclear characteristics. 29 0 0Specification -The individual unirradiated TRIGA fuel elements shall have the following characteristics:

a. Uranium content:

8.5, 20 or 30 wt % uranium enriched nominally to less than 20% U-235.b. Hydrogen to zirconium atom ratio (in the ZrH x): 1.60 to 1.70 (I.65+/- 0.05).c. Cladding: stainless steel, nominal 0.5mm (0.020 inch) thick.Basis -a. The design basis of a TRIGA core loaded with TRIGA fuel demonstrates that limitingoperation to 2.3 megawatts steady state or to a 36 megawatt-sec pulse assures an amplemargin of safety between the maximum temperature generated in the fuel and the safety limit forfuel temperature. The fuel temperatures are not expected to exceed 630 00 during any condition of normal operation.

b. Analysis shows that the stress in a TRIGA fuel element, H/Zr ratios between 1.6 and 1.7, isequal to the clad yield strength when both fuel and cladding temperature are at the safety limit9300C. Since the fuel temperatures are not expected to exceed 630 0C during any condition ofnormal operation, there is a margin between the fuel element clad stress and its ultimatestrength.

c. Safety margins in the fuel element design and fabrication allow for normal mill tolerances ofpurchased materials.

5.3.3 Control Rods and Control Rod DrivesApplicability -This Specification applies to the control rods and control rod drives used in thereactor core.Objective -The objective is to assure the control rods and control rod drives are of such adesign as to permit their use with a high degree of reliability with respect to their physical,

nuclear, and mechanical characteristics.

Specification -a. All control rods shall have scram capability and contain a neutron poison such as stainless steel, borated graphite, B 4C powder, or boron and its compounds in solid form. The shim andregulating rods shall have fuel followers sealed in stainless steel. The transient rod shall have anair filled follower and be sealed in an aluminum tube.b. The control rod drives shall be the standard GA rack and pinion type with an electromagnet and armature attached. Basis -a. The neutron poison requirements for the control rods are satisfied by using stainless steel,neutron absorbing borated graphite, B 40 powder, or boron and its compounds. These materials shall be contained in a suitable clad material such as stainless steel or aluminum to assuremechanical stability during movement and to isolate the neutron poison from the tank waterenvironment. Scram capabilities are provided for rapid insertion of the control rods.b. The standard GA TRIGA control rod drive meets the requirements for driving the control rodsat the proper speeds, and the electromagnet and armature provide the requirements for rapidinsertion capability. These drives have been tested and proven in many TRIGA reactors. 30

• 05.4 Fissionable Material StorageApplicability

-This specification applies to the storage of reactor fuel at a time when it is not in thereactor core.Objective -The objective is to assure that the fuel which is being stored will not become critical and willnot reach an unsafe temperature. Specification -a. All fuel elements not in the reactor core shall be stored (wet or dry) in a geometrical array where thekeff is less than 0.9 for all conditions of moderation.

b. Irradiated fuel elements shall be stored in an array which shall permit sufficient natural convection cooling by water or air such that the fuel element temperature shall not exceed the safety limit.Basis -The limits imposed by Technical Specifications 5.4.a and 5.4.b assure safe storage.6.0 Administrative Controls6.1 Organization.

The Vice Chancellor for Research shall be the licensee for the UCD/MNRC. TheUCD/MNRC facility shall be under the direct control of the UCD/MNRC Director or a licensed seniorreactor operator (SRO) designated by the UCD/MNRC Director to be in direct control. The UCD/MNRCDirector shall be accountable to the Vice Chancellor of the Office of Research for the safe operation andmaintenance of the reactor and its associated equipment. 6.1.1 Structure. The management for operation of the UCD/MNRC facility shall consist of theorganizational structure as shown in Figure 6.1.6.1.2 Responsibilities. The UCD/MNRC Director shall be accountable to the Vice Chancellor ofthe Office of Research for the safe operation and maintenance of the reactor and its associated equipment. The UCD/MNRC

Director, or his designated alternate, shall review and approve allexperiments and experiment procedures prior to their use in the reactor.

Individuals in themanagement organization (e.g., Site Manager, Reactor Manager, Health Physics Manager, etc.)shall be responsible for implementing UCD/MNRC policies and for operation of the facility, andshall be responsible for safeguarding the public and facility personnel from undue radiation exposures and for adhering to the operating license and technical specifications. The SiteManager shall report directly to the UCD/MNRC Director. The Reactor Manager and HealthPhysics Manager report directly to the Site Manager.6.1.3 Staffing_ 6.1.3.1 The minimum staffing when the reactor is not shutdown shall be:a. A reactor operator in the control room;b. A second person in the facility area who can perform prescribed instructions;

c. A senior reactor operator readily available.

The available senior reactoroperator should be within thirty (30) minutes of the facility and reachable bytelephone, and;d. A senior reactor operator shall be present whenever a reactor startup isperformed, fuel is being moved, or experiments are being placed in the reactortank.6.1.3.2 A list of reactor facility personnel by name and telephone number shall beavailable to the reactor operator in the control room. The list shall include:31

a. Management personnel.

b. Health Physics personnel.

c. Reactor Operations personnel.

6.1.4 Selection and Training of Personnel. The selection, training and requalification of operations personnel shall meet or exceed the requirements of the American National Standard for Selection andTraining of Personnel for Research Reactors (ANS 15.4). Qualification and requalification of licensedoperators shall be subject to an approved Nuclear Regulatory Commission (NRC) program.6.2 Review. Audit. Recommendation and ApprovalGeneral Policy. Nuclear facilities shall be designed, constructed,

operated, and maintained insuch a manner that facility personnel, the general public, and both university and non-university property are not exposed to undue risk. These activities shall be conducted in accordance withapplicable regulatory requirements.

The UCO Vice Chancellor of the Office of Research shall institute the above stated policy as thefacility license holder. The Nuclear Safety Committee (NSC) has been chartered to assist inmeeting this responsibility by providing timely, objective, and independent

reviews, audits,recommendations and approvals on matters affecting nuclear safety. The following describes the composition and conduct of the NSC.6.2.1 NSC Composition and Qualifications.

The UCD/MNRC Director shall appoint theChairperson of the NSC. The NSC Chairperson shall appoint a Nuclear Safety Committee (NSC) of at least seven (7) members knowledgeable in fields which relate to nuclear safety. TheNSC shall evaluate and review nuclear safety associated with the operation and use of theUCD/MN RC.6.2.2 NSC Charter and Rules. The NSC shall conduct its review and audit (inspection) functions in accordance with a written charter. This charter shall include provisions for:a. Meeting frequency (The committee shall meet at least semiannually).

b. Voting rules.c. Quorums (For the full committee, a quorum will be at least seven (7) members).

d. A committee review function and an audit/inspection function.

e. Use of subcommittees.

f. Review, approval and dissemination of meeting minutes.6.2.3 Review Function.

The responsibilities of the NSC, or a designated subcommittee thereof,shall include but are not limited to the following:

a. Review approved experiments utilizing UCD/MNRC nuclear facilities.

b. Review and approve all proposed changes to the facility

license, the Technical Specifications and the Safety Analysis Report, and any new or changed Facility Use Authorizations andproposed Class 1 modifications, prior to implementing (Class I) modifications, prior to takingaction under the preceding documents or prior to forwarding any of these documents to theNuclear Regulatory Commission for approval.

c. Review and determine whether a proposed change, test, or experiment would constitute anunreviewed safety question or require a change to the license, to a Facility Use Authorization, or32

• 0to the Technical Specifications.

This determination may be in the form of verifying a decisionalready made by the UCD/MNRC Director.

d. Review reactor operations and operational maintenance, Class I modification

records, andthe health physics program and associated records for all UCD/MNRC nuclear facilities.

e. Review the periodic updates of the Emergency Plan and Physical Security Plan forUCD/MNRC nuclear facilities.

f. Review and update the NSC Charter every two (2) years.g. Review abnormal performance of facility equipment and operating anomalies.

h. Review all reportable occurrences and all written reports of such occurrences prior toforwarding the final written report to the Nuclear Regulatory Commission.

i. Review the NSC annual audit/inspection of the UCD/MNRC nuclear facilities and any otherinspections of these facilities conducted by other agencies.

6.2.4 Audit/Inspection Function. The NSC or a subcommittee

thereof, shall audit/inspect reactoroperations and health physics annually.

The annual audit/inspection shall include, but not belimited to the following:

a. Inspection of the reactor operations and operational maintenance, Class I modification

records, and the health physics program and associated

records, including the ALARA program,for all UCD/MNRC nuclear facilities.

b. Inspection of the physical facilities at the UCD/MNRC.

c. Examination of reportable events at the UCD/MNRC.

d. Determination of the adequacy of UCD/MNRC standard operating procedures.

e. Assessment of the effectiveness of the training and retraining programs at the UCD/MNRC.

f. Determination of the conformance of operations at the UCD/MNRC with the facility's licenseand Technical Specifications, and applicable regulations.

g. Assessment of the results of actions taken to correct deficiencies that have occurred innuclear safety related equipment, structures,

systems, or methods of operations.

h. Inspection of the currently active Facility Use Authorizations and associated experiments.

i.Inspection of future plans for facility modifications or facility utilization.

j. Assessment of operating abnormalities.

k. Determination of the status of previous NSC recommendations.

6.3 Radiation Safety. The Health Physics Manager shall be responsible for implementation of theUCD/MNRC Radiation Safety Program. The program should use the guidelines of the AmericanNational Standard for Radiation Protection at Research Reactor Facilities (ANSI/ANS 15.11). TheHealth Physics Manager shall report to the Site Manager.6.4 Procedures. Written procedures shall be prepared and approved prior to initiating any of theactivities listed in this section. The procedures shall be approved by the UCD/MNRC Director. A periodicreview of procedures shall be performed and documented in a timely manner by the UCD/MNRC staff toassure that procedures are current. Procedures shall be adequate to assure the safe operation of the33 0 ireactor, but shall not preclude the use of independent judgment and action should the situation require.Procedures shall be in effect for the following items:6.4.1 Reactor Operations Procedures

a. Startup, operation, and shutdown of the reactor.b. Fuel loading, unloading, and movement within the reactor.c. Control rod removal or replacement.

d. Routine maintenance of the control rod drives and reactor safety and interlock systems orother routine maintenance that could have an effect on reactor safety.e. Testing and calibration of reactor instrumentation and controls, control rods and control roddrives.f. Administrative controls for operations, maintenance, and conduct of irradiations andexperiments that could affect reactor safety or core reactivity.

g. Implementation of required plans such as emergency and security plans.h. Actions to be taken to correct potential malfunctions of systems, including responses toalarms and abnormal reactivity changes.6.4.2 Health Physics Procedures

a. Testing and calibration of area radiation

monitors, facility air monitors, laboratory radiation detection

systems, and portable radiation monitoring instrumentation.

b. Working in laboratories and other areas where radioactive materials are used.c.- Facility radiation monitoring program including routine and special surveys, personnel monitoring, monitoring and handling of radioactive waste, and sampling and analysis of solidand liquid waste and gaseous effluents released from the facility.

The program shall include amanagement commitment to maintain exposures and releases as low as reasonably achievable (ALARA).d. Monitoring radioactivity in the environment surrounding the facility.

e. Administrative guidelines for the facility radiation protection program to include personnel orientation and training.

f. Receipt of radioactive materials at the facility, and unrestricted release of materials and itemsfrom the facility which may contain induced radioactivity or radioactive contamination.

g. Leak testing of sealed sources containing radioactive materials.

h. Special nuclear material accountability.

i. Transportation of radioactive materials.

Changes to the above procedures shall require approval of the UCD/MNRC Director. All such changes shall bedocumented. 6.5 Experiment Review and Approval. Experiments having similar characteristics are grouped togetherfor review and approval under specific Facility Use Authorizations. All specific experiments to be34

• 0performed under the provisions of an approved Facility Use Authorization shall be approved by theUCD/MNRC

Director, or his designated alternate.

a. Approved experiments shall be carried out in accordance with established and approved procedures.

b. Substantive change to a previously approved experiment shall require the same review and approvalas a new experiment.

c. Minor changes to an experiment that do not significantly alter the experiment may be approved by asenior reactor operator.

6.6 Required Actions6.6.1 Action to be taken in case of a safety limit violation. In the event of a safety limit violation (fuel temperature), the following action shall be taken:a. The reactor shall be shut down and reactor operation shall not be resumed until authorized by the NRC.b. The safety limit violation shall be promptly reported to the UCD/MNRC Director.

c. The safety limit violation shall be reported to the chairman of the NSC and to the NRC by theUCD/MNRC Director.

d. A safety limit violation report shall be prepared.

The report shall describe the following: (1) Applicable circumstances leading to the violation, including when known, the causeand contributing factors.(2) Effect of the violation upon reactor facility components,

systems, or structures, andon the health and safety of personnel and the public.(3) Corrective action to be taken to prevent reoccurrence.

e. The safety limit violation report shall be reviewed by the NSC and then be submitted to theNRC when authorization is sought to resume operation of the reactor.6.6.2 Actions to be taken for reportable occurrences.

In the event of reportable occurrences, the following actions shall be taken:a. Reactor conditions shall be returned to normal or the reactor shall be shut down. If it isnecessary to shut down the reactor to correct the occurrence, operations shall not be resumedunless authorized by the UCD/MNRC Director or his designated alternate.

b. The occurrence shall be reported to the UCD/MNRC Director or the designated alternate.

The UCD/MNRC Director shall report the occurrence to the NRC as required by these Technical Specifications or any applicable regulations.

c. Reportable occurrences should be verbally reported to the Chairman of the NSC and theNRC Operations Center within 24 hours of the occurrence.

A written preliminary report shall besent to the NRC, Attn: Document Control Desk, 1 White Flint North, 11555 Rockville Pike,Rockville MD 20852, within 14 days of the occurrence. A final written report shall be sent to theabove address within 30 days of the occurrence.

d. Reportable occurrences should be reviewed by the NSC prior to forwarding any writtenreport to the Vice Chancellor of the Office of Research or to the Nuclear Regulatory Commission.

35

• 06.7 Reports.

All written reports shall be sent within the prescribed interval to the NRC, Attn: DocumentControl Desk, 1 White Flint North, 11555 Rockville Pike, Rockville MD 20852.6.7.1 Operating Reports. An annual report covering the activities of the reactor facility duringthe previous calendar year shall be submitted within six months following the end of eachcalendar year. Each annual report shall include the following information:

a. A brief summary of operating experiences including experiments performed, changes infacility design, performance characteristics and operating procedures related to reactor safetyoccurring during the reporting period, and results of surveillance tests and inspections.

b. A tabulation showing the energy generated by the reactor (in megawatt hours), hours thereactor was critical, and the cumulative total energy output since initial criticality.

c. The number of emergency shutdowns and inadvertent scrams, including reasons for theshutdowns or scrams.d. Discussion of the major maintenance operations performed during the period, including theeffect, if any, on the safety of the operation of the reactor and the reasons for any corrective maintenance required.

e. A brief description, including a summary of the safety evaluations, of changes in the facility orin procedures, and of tests and experiments carried out pursuant to Section 50.59 of 10 CFRPart 50.f. A summary of the nature and amount of radioactive effluents released or discharged to theenvironment beyond the effective control of the licensee as measured at or prior to the point ofsuch release or discharge, including the following:

(1) Liquid Effluents (summarized on a monthly basis).(a) Liquid radioactivity discharged during the reporting period tabluated asfollows:1 The total estimated quantity of radioactivity released (in curies).2 An estimation of the specific activity for each detectable radionuclide present if the specific activity of the released material after dilution isgreater than 1 xl07 microcuries/ml. 3 A summary of the total release in curies of each radionuclide determined in 2_above for the reporting period based on representative isotopic analysis. 4 An estimated average concentration of the released radioactive material at the point of release for each month in which a releaseoccurs, in terms of microcuries/ml and the fraction of the applicable concentration limit in 10 CFR 20.(b) The total volume (in gallons) of effluent water (including diluent) releasedduring each period of liquid effluent release.(2) Airborne Effluents (summarized on a monthly basis):(a) Airborne radioactivity discharged during the reporting period (in curies)tabulated as follows:36 0 0I The totai estimated quantity of radioactivity released (in curies)determined by an appropriate sampling and counting method.2 The total estimated quantity (in curies) of Argon-41 released duringthe reporting period based on data from an appropriate monitoring system.3 The estimated maximum annual average concentration of Argon-41in the unrestricted area (in microcuries/ml), the estimated corresponding annual radiation dose at this location (in millirem), and the fraction of theapplicable 10 CFR 20 limits for these values.4 The total estimated quantity of radioactivity in particulate form withhalf lives greater than eight days (in curies) released during thereporting period as determined by an appropriate particulate monitoring system.5 The average concentration of radioactive particulates with half-lives greater than eight days released (in microcuries/ml) during the reporting period.(3) Solid Waste (summarized on an annual basis)(a) The total amount of solid waste packaged (in cubic feet).(b) The total activity in solid waste (in curies).(c) The dates of shipment and disposition (if shipped off site).g. An annual summary of the radiation exposure received by facility operations personnel, byfacility users, and by visitors in terms of the average radiation exposure per individual and thegreatest exposure per individual in each group.h. An annual summary of the radiation levels and levels of contamination observed duringroutine surveys performed at the facility in terms of average and highest levels.i.An annual summary of any environmental surveys performed outside the facility. 6.7.2. Special Reports. Special reports are used to report unplanned events as well as plannedadministrative changes. The following classifications shall be used to determine the appropriate reporting schedule:

a. A report within 24 hours by telephone or similar conveyance to the NRC operations center of:(1) Any accidental release of radioactivity into unrestricted areas above applicable unrestricted area concentration limits, whether or not the release resulted in propertydamage, personal injury, or exposure; (2) Any violation of a safety limit;(3) Operation with a limiting safety system setting less conservative than specified inSection 2.0, Limiting Safety System Settings; (4) Operation in violation of a Limiting Condition for Operation; 37 0 0(5) Failure of a required reactor or experiment safety system component which couldrender the system incapable of performing its intended safety function unless the failureis discovered during maintenance tests or a period of reactor shutdown; (6) Any unanticipated or uncontrolled change in reactivity greater than $1.00;(7) An observed inadequacy in the implementation of either administrative or procedural

controls, such that the inadequacy could have caused the existence or development of acondition which could have resulted in operation of the reactor outside the specified safety limits; and(8) A measurable release of fission products from a fuel element.b. A report within 14 days in writing to the NRC, Document Control Desk, Washington DC.(1) Those events reported as required by Technical Specifications 6.7.2.a.1 through6.7.2.a.8.

(2) The written report (and, to the extent possible, the preliminary telephone report orreport by similar conveyance) shall describe,

analyze, and evaluate safety implications, and outline the corrective measures taken or planned to prevent reoccurrence of theevent.c. A report within 30 days in writing to the NRC, Document Control Desk, Washington DC.(1) Any significant variation of measured values from a corresponding predicted orpreviously measured value of safety-connected operating characteristics occurring during operation of the reactor;(2) Any significant change in the transient or accident analysis as described in theSafety Analysis Report (SAR);(3) A personnel change involving the positions of UCD/MNRC Director or UCD ViceChancellor for Research; and(4) Any observed inadequacies in the implementation of administrative or procedural controls such that the inadequacy causes or could have caused the existence ordevelopment of an unsafe condition with regard to reactor operations.

6.8 Records. Records may be in the form of logs, data sheets, or other suitable forms. The requiredinformation may be contained in single or multiple

records, or a combination thereof.

Records and logsshall be prepared for the following items and retained for a period of at least five years for items a.through f., and indefinitely for items g. through k. (Note: Annual reports, to the extent they contain all ofthe required information, may be used as records for items g. through j.)a. Normal reactor operation.

b. Principal maintenance activities.

c. Those events reported as required by Technical Specifications 6.7.1 and 6.7.2.d. Equipment and component surveillance activities required by the Technical Specifications.

e. Experiments performed with the reactor.f. Airborne and liquid radioactive effluents released to the environments and solid radioactive wasteshipped off site.38 0 0g. Offsite environmental monitoring surveys.h. Fuel inventories and transfers.

i. Facility radiation and contamination surveys.j. Radiation exposures for all personnel.

k. Updated, corrected, and as-built drawings of the facility.

39 ... NUCLEARcoMSSoREGUALTORY UNIVERSITY OFCALIFORNIA -DAVISVICE CHANCELLOR FORRESEARCH(Licensee) UCD/MNRC UCDIMNRC [DIRECTOR NUCLEAR ._._l COMMITJTEE [ 8At-SITE , 'MANAGERiHEALTH PHYFIGUREAC6.1

• Att EGu _; UNITED STATES ".

REGULATORY COMMISSION20555-0001 March 30, 2004* Dr. Barry M. KleinVice Chancellor for ResearchUniversity of California, DavisOne Shields AvenueDavis, CA 95616-8558

SUBJECT:

REVISION TO SAFETY EVALUATION OF AMENDMENT NO. 7 TO AMENDEDFACILITY OPERATING LICENSE NO. R-130 -REGENTS OF THEUNIVERSITY OF CALIFORNIA (TAC NO. MB5598)

Dear Dr. Klein:

The U.S. Nuclear Regulatory Commission has Issued the enclosed revision to the SafetyEvaluation of Amendment No. 7 to Facility Operating License No. R-1 30 for the McClellan. Nuclear Radiation Center (MNRC) TRIGA Research Reactor. Amendment No. 7 was issuedon December 30, 2003 and is available on the Commission's ADAMS system, Accession Number ML033421339. Sincerely, _ WreJ .EeIn rjc aaeResearch and Test Reactors SectionNew, Research and Test Reactors ProgramDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation Docket No. 50-607

Enclosure:

Revision to Amendment No. 7Safety Evaluation Reportcc w/enclosure: Please see next page 0..University of California -Davis/McClellan MNRC Docket No. 50-607cc:Mr. Jeff Ching5335 Price Avei~ue, Bldg. 258McClellan AFB, CA 95652-2504 Test, Research, and TrainingReactor Newsletter University of Florida202 Nuclear Sciences CenterGainesville, FL 32611 0UNITEb STATESNUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 REVISION TO SAFETY EVALUATION REPORTSUPPORTING AMENDMENT NO. 7 TOAMENDED .FACILITY OPERATING LICENSE NO. R-130REGENTS OF THE UNIVERSITY OF CALIFORNIA ATMcCLELLAN NUCLEAR RADIATION CENTERDOCKET N.O. 50-60

71.0 INTRODUCTION

By letter dated October 21, 2003, the Regents of the University of California (the licensee) submitted a request for amendment of the Facility Operating

• License No. R-130 for theMcClellan Nuclear Radiation Center (MNRC) TRIGA research reactor.

The request provided forthe allowance of radioactive materials not produced by the reactor to be received, possessed and used on the facility site. In particular, it was requested that Section 2.B of the FacilityOperating License be amended to include an additional section 2.B.(4) as follows:2.B.(4) In addition to those items specified in 2.B.(1), 2.B.(2) and 2.B.(3) the following radioactive materials maybe received, possessed, and used at the facility. Radioactive Material(element and mass number)A. Any radioactive materialbetween atomic number I through83, Inclusive B. Any radioactive material withatomic numbers 84 and abovec.. Iodine-125 D. Source material (but only traceamounts of Th-234)E. Special nuclear materialChemical and/orPhysical Form.A. AnyA. Anyc. Iodide/LIquid D. AnyE. AnyMaximum Quantity Licensee MayPossess at Any One TimeA. 20 curies (1 curie each, except asprovided below)A. 4 Curies (100 milllcuries each,except as provided below) or up to 20micrograms C. 40 CuriesD. 4 grams per radionuclide, not toexceed 10 grams totalE. 2 grams per radionuclide, not toexceed 5 grams totalThis amendment request was approved and issued on .December 30, 2003. 0 "-2-2.0 EVALUATION The previous safety evaluation assumed that all of the radioactive materials to be received, possessed and handled in accordance with this amendment request would be located in thereactor room glove box. The significance of this assumption is related to the ability of thereactor room glove box and its associated exhaust system to mitigate the consequences associated with the complete volatilization of the maximum radioactive material inventory contained in the box, a total of 64.4 curies. (The total activity in categories A, B, and C in theabove table is 64 curies. The maximum activity In category D is about 0.1 curie, while themaximum activity in category E is about 0.3 curie.). The staff concluded that the consequences of the complete volatilization of 64.4 curies are much less than the consequences of thebounding MHA, and the amendment request was approved. Instead of locating all of the radioactive materials shown in above table in the reactor roomglove box, some of the materials will be located in the restricted area of the McClellan NuclearRadiation Center. Non-volatile material will be handled in accordance with approvedprocedures. Any unsealed volatile

material, such as Iodine-I125 (the majority of the radioactive materials),

will continue to be handled in areas with filtered ventilation to mitigate theconsequences of complete volatilization of the unsealed material (e.g., the reactor room glovebox and reactor room fume hood), as previously analyzed. The staff has reviewed the proposed change to the Facility Operating License and concluded that it does not impact the licensee's ability to continue to meet the relevant requirements of 10CFR Part 50.38.3.0 ENVIRONMENTAL CONSIDERATION .This amendment does hot Involve changes in the installation or use of a facility component located within the restricted area as defined ion 10 CFR Part 20 or changes in inspection andsurveillance requirements. The staff has determined that this amendment involves nosignificant increase in the amounts, and no significant change in the types, of any effluents thatmay be released off site, and no significant increase in individual or cumulative occupational radiation exposure. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared with the issuance of thisamendment.

4.0 CONCLUSION

The staff has concluded, on the basis of the considerations discussed above, that (1) becausethe amendment does not involve a significant increase in the probability or consequences ofaccidents previously evaluated, or create the possibility of a new or different kind of accidentfrom any accident previously evaluated, and does not involve a significant reduction In a margin* of safety, the amendment does not involve a significant hazards consideration; (2) there isreasonable assurance that the health and safety of the public will not be endangered by the proposed changes; and (3) such changes are in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security orthe health and safety of the public.Principal Contributor: Warren J. EreslanDate: March 30, 2004 '5%. i ./ uNITED STATESNLCLEAR REGULATORY COMMISSION 0 ASIGTNDC.205-00 Deceeiber: 30, 2003Dr. Barry M. KleinVice Chancellor for ResearchUniversity of California, DavisOne Shields AvenueDavis, CA 95616-8558

SUBJECT:

ISSUANCE OF AMENDMENT NO. 7 TO AMENDED FACILITY OPERATING LICENSE NO. R-130 -REGENTS OF THE UNIVERSITY OF CALIFORNIA (TAC NO. MB5598)

Dear Dr. Klein:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 7 toFacility Operating License No. R-1 30 for the McClellan Nuclear Radiation Center (MNRC)TRIGA Research Reactor. The amendment consists of changes to the Facility Operating License in response to your submittals of October 21, 2003 and November 6, 2003, and isdiscussed in the enclosed Safety Evaluation Report. 69~4~tey/Warren J. Eresian, Project ManagerResearch and Test Reactors SectionNew, Research and Test Reactors ProgramDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation Docket No. 50-607

Enclosures:

1. Amendment No. 72. Safety Evaluation Report University of California

-Davis/McClellan MNRC Docket No. 50-607cc:Dr. Wade J. Richards5335 Price Avenue, Bldg. 258McClellan AFB, CA 95652-2504 Test, Research, and TrainingReactor Newsletter University of Florida202 Nuclear Sciences CenterGainesville, FL 32611 UNITED STATESNUCLEAR REGULATORY COMMISSIONWASHINGTON, D.C. 20555-0001 REGENTS OF THE UNIVERSITY OF CALIFORNIA ATMcCLELLAN NUCLEAR RADIATION CENTERDOCKET NO. 50-607AMENDMENT TO AMENDED FACILITY OPERATING LICENSEAmendment No. 7License No. R-1 301. The U.S. Nuclear Regulatory Commission (the Commission) has found thatA. The application for an amendment to Amended Facility Operating LicenseNo. R-1 30 filed by the Regents of the University of California at McClellan NuclearRadiation Center (the licensee) on October 21, 2003 and November 6, 2003,conforms to the standards and requirements of the Atomic Energy Act of 1954:, asamended (the Act), and the regulations of the Commission as stated in Chapter I ofTitle 10 of the Code of Federal Regulations (10 CFR);B. The facility will operate In conformity with the application, the provisions of the Act,and the rules and regulations of the Commission; C. There is reasonable assurance that (i) the activities authorized by this amendment can be conducted without endangering the health and safety of the public and (ii)such activities will be conducted in compliance with the regulations of theCommission; D. The issuance of this amendment will not be Inimical to the common defense andsecurity or to the health and safety of the public;E. This amendment is issued in accordance with the regulations of the Commission asstated in 10 CFR Part 51, and all applicable requirements have been satisfied; andF. Prior notice of this amendment was not required by 10 CFR 2.105, and publication of notice for this amendment is not required by 10 CFR 2.106. O f O0..-2-2. Accordingly, the license is amended by changes to the Facility Operating License asindicated below, and paragraph 2.B of Amended Facility Operating License No. R-130 ishereby amended to read as follows:2.B.(4) In addition to those items specified in 2.B.(1), 2.B.(2) and 2.B.(3) the following radioactive materials may be received, possessed, and used at the facility. Radioactive Material(element and massnumber)A. Any radioactive material betweenatomic number 1through 83, inclusive B. Any radioactive material with atomicnumbers 84 andaboveChemical and/orPhysical FormA. AnyA. AnyMaximum Quantity LicenseeMay Possess at Any One TimeA. A. 20 Curies (I Curie each,except as provided below)A. 4 Curies (100 millicuries each, except as providedbelow) or up to 20micrograms C. 40OCuries D. 4 grams per radionuclide, not to exceed 10 gramstotalE. 2 grams per radionuclide, not to exceed 5 grams totalC. Iodine-125 D. Source material (butonly trace amountsof Th-234)E. Special nuclearmaterialC. Iodide/Liquid

0. AnyE. Any3. This license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Research and Test Reactors SectionNew, Research and Test Reactors ProgramDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation Date of Issuance: December 30, 2003 O 0~UNITED STATESNUCLEAR REGULATORY COMMISSION o~WASHINGTON, D.C. 20555-0001 o#SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 7 TOAMENDED FACILITY OPERATING LICENSE NO. R-130REGENTS OF THE UNIVERSITY OF CALIFORNIA ATMcCLELLAN NUCLEAR RADIATION CENTERDOCKET NO. 50-60

71.0 INTRODUCTION

By letter dated October 21, 2003, the Regents of the University of California (the licensee) submitted a request for amendment of the Facility Operating License No. R-1 30 for the McClellan Nuclear Radiation Center (MNRC) TRIGA research reactor. The request provides for theallowance of radioactive materials not produced by the reactor to be received, possessed andused on the facility site. In particular, it is requested that Section 2.B of the Facility Operating License be amended to include an additional Section 2.B.(4) as follows:2.B.(4) In addition to those items specified in 2.B.(1), 2.B.(2) and 2.B.(3) the following radioactive materials may be received, possessed, and used at the facility. Radioactive Material(element and massnumber)A. Any radioactive material betweenatomic number 1through 83, inclusive B. Any radioactive material with atomicnumbers 84 andaboveChemical and/orPhysical FormA. AnyA. AnyMaximum Quantity Licensee MayPossess at Any One TimeA. A. 20 Curies (1 Curie each,except as provided below)A. 4 Curies (100 mlllicuries each, except as providedbelow) or up to 20micrograms C. 40 CuriesD. 4 grams per radionuclide, not to exceed 10 grams totalE. 2 grams per radionuclide, not to exceed 5 grams totalC. Iodine-I125 D. Source material (butonly trace amounts ofTh-234)E. Special nuclearmaterialC. Iodide/Liquid D. AnyE. Any

• 0-2,-This request is discussed below.2.0 EVALUATION All of the radioactive materials to be received, possessed and handled In accordance with thisamendment request will be located in the reactor room glove box. In November of 2002, the NRCapproved Amendment No. 5 of the Technical Specifications for the McClellan Nuclear Radiation Center. The safety concern addressed in that amendment was related to the ability of the reactorroom glove box and Its associated exhaust system to mitigate the consequences associated withthe complete volatilization of the maximum radioactive material inventory contained in the box, atotal of 61 curies of Iodine-125.

The analysis showed that the CEDE to the thyroid for a 10-minute exposure in the unrestricted area would be about 3 millirem. For those exposed in the reactorroom for the maximum assumed occupancy time of 5 minutes the CEDE to the thyroid would beabout 205 millirem. These doses were compared to the expected doses (CEDE) resulting fromthe Maximum Hypothetical Accident (MHA), which serves as the bounding accident forradiological consequences. The resulting doses from the MHA are 53 millirem in the unrestricted area and 360 millirem in the reactor room. The staff concluded that the consequences of thecomplete volatilization of 61 curies of Iodine-I125 were less than the bounding MHA and therefore there was not a significant reduction of the margin of safety with respect to the MHA.This amendment request will increase the total allowable activity in the reactor room glove boxfrom 61 curies to 64.4 curies. (The total activity in categories A, B, and C In the above table Is 64curles. The maximum activity in category D corresponds to 10 grams of Uranium-233, or about0.1 curie. The maximum activity in category E corresponds to 5 grams of Plutonium-239, or about0.3 curie.) For the complete volatilization of 64.4 curies, doses in the unrestricted area and in thereactor room will scale up proportionally from 61 curies, resulting in a dose in the unrestricted areaof 3.2 mlllirem, and a dose in the reactor room of 216 millirem (i.e., the doses have increased by5.6 percent.) The staff concludes that the consequences of the complete volatilization of 64.4 curies are muchless than the consequences of the bounding MHA, and that increasing the allowable activity in thereactor room glove box from 61 curies to 64.4 curies does not significantly reduce the margin ofsafety with respect to the MHA and to 10 CFR Part 20 limits and that the increase is acceptable. The staff has reviewed the proposed change to the Facility Operating License and concluded thatit does not impact the licensee's ability to continue to meet the relevant requirements of 10 CFRPart 50.36.3.0 ENVIRONMENTAL CONSIDERATION This amendment does not involve changes in the installation or use of a facility component locatedwithin the restricted area as defined in 10 CFR Part 20 or changes in inspection and surveillance requirements. The staff has determined that this amendment involves no significant increase Inthe amounts, and no significant change in the types, of any effluents that may be released off site,and no significant increase in individual or cumulative occupational radiation exposure. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in10 CER 51 .22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement orenvironmental assessment need be prepared with the issuance of this amendment.

4.0 CONCLUSION

The staff has concluded, on the basis of the considerations discussed above, that (1) because theamendment does not involve a significant increase in the probability or consequences of accidents previously evaluated, or create the possibility of a new or different kind of accident from anyaccident previously evaluated, and does not involve a significant reduction in a margin of safety,the amendment does not involve a significant hazards consideration; (2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposedchanges; and (3) such changes are in compliance with the Commission's regulations and theissuance of this amendment will not be inimical to the common defense and security or the healthand safety of the public.Principal Contributor: Warred J. EresianDate: December 30, 2003 UNITED STATES*NUCLEAR REGULATORY COMMISSION ,,.. WASHINGTON, D.C. 20555-0001February 17, 2000*i7/tlJ*Brigadier General Michael P. Wiedemer Vice Chancellor Kevin SmithCommander Office of the Chancellor ..~Sacramento Air Logistics Center University of California, DavisSM-ALCITI-1 One Shields Avenue5335 Price Avenue Davis, California 95616-8558 McClellan AFB, California 95652-2504

SUBJECT:

RE-ISSUANCE OF NOTICE OF CONSIDERATION OF APPROVAL OFTRANSFER OF FACILITY OPERATING LICENSE NO. R-130 FOR THEMcCLELLAN NUCLEAR RADIATION CENTER FROM THE DEPARTMENT OFTHE AIR FORCE TO THE REGENTS OF THE UNIVERSITY OF CALIFORNIA

• AND CONFORMING AMENDMENT, AND OPPORTUNITY FOR A HEARING(TAC NO. MA3477)Dear General Wiedemer and Dr. Smith:The enclosed document has been re-issued in its entirety to correct someadministrative errors. We. apologize for any inconvenience this may have caused.Sincerely, Ledyard B. Marsh, ChiefEvents Assessments, Generic Communications and Non-Power Reactors BranchDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor Regulation Docket No. 50-607

Enclosure:

As statedcc: wlenclosures McClellan AFB TRIGA REACTORDcktN.0-7 CC:Dr. Wade J. RichardsSM-ALC/TI-1 5335 Price Avenue, Bldg. 258McClellan AFB, CA 95652-2504 Cot. Robert CapellHQ AFMC/SGC4225 Logistics Avenue, Suite 23Wright-Patterson AFB, OH 45433-5762 Lt. Col. Catherine Ze~ringue HQ AFSCISEW9570 Avenue G, Building 24499Kirtland AFB, New Mexico 871 17-5670Test, Research, and TrainingReactor Newsletter 202 Nuclear Sciences CenterUniversity of FloridaGainesville, FL 3261 1

• L0TECHNICAL SPECIFICATIONS FOR THEUNIVERSITY OF CALIFORNIA

-DAVIS MCCLELLAN NUCLEAR RADIATION CENTER(UCDIMNRC) DOCUMENT NUMBER: MNRC-0004-DOC-13 Rev 13 4/03p.~. ~.1 *>~0 !Revision ,13 of the "Technical Specifications" for the University of California at Davis/McClellan Nuclear Radiation Center (UCD/MNRC) Reactor have undergone the following coordination: Reviewed by: !.0eltPyiSpesoDate Reviewed by: 0.R ator Su~d pervisrDt Approved by:~i'I~toY* ~Ij.e/o3~l1.DateChairman, NRCNuclear Safety Committee I.(K ~/--I 0Technical Specifications Rev 13 412003Title PageApproval Page31Rev 13Rev 13Rev 13Rev 13Rev 13Rev 134/20034/20034/20034/20034/20034/20033233Figure 6.1......................----.--... ~

• 0* " TECHNICAL SPECIFICATIONS TABLE OF CONTENTS1.0 Definitions

.............................................................................................................. 12.0 Safety Limit and Limiting Safety System Setting 2.1 Safety Limits.................................................................................................. 2.2 Limiting Safety System Setting (LSSS)..................................................................... 62.2.1 Fuel Temperature ............................................................ i.................... 63.0 Limiting Conditions for Operations (LC.O.) ........................................................................... 73.1 Reactor Core Parameters................................................................................... 73.1.1 Steady-State Operation .......................................

...................................

73.1.2 Pulse or Square Wave Operation ............................................................... 73.1.3 Reactivity Limitation~s............................................................................. 83.2 Reactor Control and Safety Systems .... .................................................................. 83.2.1 Control Rods...................................................................................... 83.2.2 Reactor Instrumentation.......................................................................... 93.2.3 Reactor Scrams and Interlocks................................................................. 103.2.4 Reactor Fuel Elementts.......................................................................... 123.3 Reactor Coolant Systems.................................................................................. 133.4 Reactor Room Exhaust System ........................................................................... 143.5 Intentionally Left Blank ..................................................................................... 143.6 Intentionally Left Blank..................................................................................... 143.7 Reactor Radiation Monitoring Systems.................................................................... 143.7.1 Monitoring Systems ... ......................................................................... 143.7.2 Effluent~s -Argon-41 Discharge Limit. ..........................................................

16) 0 03.8 Experiments

................................................................................................ 163.8.1 Reactivity Limits ........................................ 163.8.2 Materials Limit................................................................................... 173.8.3 Failure and Malfunctions ...................... ................................................. 184.0 Surveillance Requirements.......................................................................................... 194.1 Reactor Core Parameters ................................................................................. 194.1.1 Steady State Operation......................................................................... 194.1.2 Shutdown Margin and Excess Reactivity ....................................................... 204.2 Reactor Control and Safety Systems...................................................................... 204.2.1 Control Rods ................................................... ................................. 204.2.2 Reactor Instrumentation............................................................... 214.2.3 Reactor Scrams and interlocks.................... ............................................. 224.2.4 Reactor Fuel Elements................................ .......................................... 234.3 Reactor Coolant Systems ................................................................................. 244.4 Reactor Room Exhaust System ........................................................................... 254.5 Intentionally Left Blank..................................................................................... 254.6 Intentionally Left Blank..................................................................................... 254.7 Reactor Radiation Monitoring Systems.................................................................... 254.8 Experiments ................................................................................................ 265.0 Design Features ...........................................................

.........................................

275.1 Site and Facility Description ............................................................................... 275.1.1 .Site............................................................................................... 275.1.2 Facility Exhaust ......................... ....................................................... 285.2 Reactor Coolant system ................................................................................... 28 5.3 Reactor Core and F.uel .................................................................................... 295.3.1 Reactor Care .................................................................................... 295.3.2 Reactor F..u~l..................................................................................... 305.3.3 Control Rods and Control Rod Drives.......................................................... 315.4 Fissionable Material Storage............................................................................... 316.0 Administrative Controls ............................................................................................ .316.1 Organization................................................................................................ 316.1.1 Structure......................................................................................... 326.1.2 Responsibilities ................................................................................. 326.1.3 Staffing .......................................................................................... 326.1.4 Selection and Training of Personnel ........................................................... 326.2 Review, Audit, Recommendation and Approvial........................................................... 326.2.1 NSC Composition and Qualifications........................................................... 336.2.2 NSC Charter and Rules .......................... i.............................................. 33 I6.2.3 Review Function................................................................................. 336.2.4 Audit/Inspection Function ....................................................................... 346.3 Radiation Safety............................................................................................ 34 16.4 Procedures ................................................................................................. 346.4.1 Reactor Operations Procedur~es................................................................. 34 I6.4.2 Health Physics Procedures ..................................................................... 356.5 Experiment Review and Approlval ......................................................................... 356.6 Required Actions ........................................................................................... 356.6.1 Actions to be taken in case of a safety limit violation: ......................................... 35 *6.6.2 Actions to be taken for reportable occurrences ................ .............................. 36 6.7.1 Operating Reports................................................................................. 366.7.2 Special Reports ................................................................................... 386.8 Records........................................................................................................ 39Fig. 6.1 UCD/MNRC Organization for Licensing and Operation........................................................ 40* /

• 0TECHNICAL SPECIFICATIONS FOR THEUNIVERSITY OF CALIFORNIA

-DAVIS/MCCLELLAN NUCLEAR RADIATION CENTER (UCDIMNRC) The University of California -Davis/McClellan Nuclear Radiation Center (UCD/MNRC) reactor is operated by theUniversity of California, Davis, California (UCD). The UCD/MNRC research reactor is a TRIGA-type reactor.The UCD/MNRC provides state-of-the-art neutron radiography capabilities. In addition, the UCD/MNRC providesa wide range of irradiation services for both research and industrial needs. The reactor operates at a nominalsteady state power level up to and including 2 MW. The UCD/MNRC reactor is also capable of square waveand pulse operational modes. The UCD/MNRC reactor fuel is less than 20% enriched in uranium-235.

1.0 Definitions

1.1 As Low As Reasonably Achievable (ALARA)~. As defined in 10 CFR, Part 20.1.2 Licensed Operators. A UCD/MNRC licensed operator is an individual licensed by the NuclearRegulatory Commission (e.g., senior reactor operator or reactor operator) to carry out the duties andresponsibilities associated with the position requiring the license.1.2.1 Sernior Reactor Operator. An individual who is licensed to direct the activities of reactoroperators and to manipulate the controls of the facility. 1.2.2 Rea~ctor Oper~ator. An individual who is licensed to manipulate the controls of the facilityand perform reactor-related maintenance. 1.3 Channel. A channel is the combination of sensor, line amplifier, processor, and output deviceswhich are connected for the purpose of measuring the value of a parameter. 1.3.1 Channel Test. A channel test is the introduction of a signal into the channel for verification that it is operable. 1.3.2 C..hannel Calibration. A channel calibration adjustment of the channel such that itsoutput corresponds with acceptable accuracy to known-'-values of the parameter which thechannel measures. Calibration shall encompass the entire channel,.including equipment actuation, alarm or trip, and shall be deemed to include a channel test. "1.3.3 Channel. Check. A channel check is a qualitative verification of acceptable performance byobservation of channel behavior. This verification, where possible, shall include comparison ofthe channel with other independent channels or systems measuring the same variable. 1.4 Confinement. Confinement means isolation of the reactor room air volume such that the movementof air into and out of the reactor room is through a controlled path.1.5 Experiment. Any operation,

hardware, or target (excluding devices such as detectors, fissionchambers, foils, etc), which is designed to investigate specific reactor characteristics or which isintended for irradiation within an experiment facility and which is not rigidly secured to a core or shieldstructure so as to be a part of their design.1.5.1 E~xperinrlent.

Moveable. A moveable experiment is one where it is intended that the entireexperiment may be moved in or near the reactor core or into and out of reactor experiment facilities while the reactor is operating. I 1.5.2 Exoerdment. Secured. A secured experdment is any experiment, experiment

facility, orcomponent of an experiment that is held in a stationary position relative to the reactor by.... .,mechanical means. The restraining force must be substantially greater than those to which the-.expediment might be subjected by hydraulic, pneumatic,

buoyant, or other forces which arenormal to the operating environment of the experiment, or by forces which can arise as a resultof credible conditions.

1.5.3 Exoeriment Facilities. Experiment facilities shall mean the pneumatic transfer tube,beamtubes, irradiation facilities, in the reactor core or in the reactor tank, and radiography bays.1.5.4 Experiment Safety System. Experiment safety systems are those systems, including theirassociated input circuits, which are designed to initiate a scram for the primary purpose ofprotecting an experiment or to provide information which requires manual protective action to beinitiated. '1.6 .Fuel Element. Standard. A fuel element is a single TRIGA element. The fuel is U-ZrH clad instainless steel. The zirconium to hydrogen ratio is nominally 1.65 +/- 0.05. The weight percent (wt%) ofuranium can be either 8.5, 20, or 30 wt%, with an enrichment of less than 20% U-235. A standard fuelelement may contain a burnable poison.1.7 Fuel Element., Instrumented. An instrumented fuel element is a standard fuel element fabricated withthermocouples for temperature measurements. An instrumented fuel element shell have at least oneoperable thermocouple embedded in the fuel near the axial and radial mnidpoints. 1.8 Measured Valu~e. The measured value is the value of a parameter as it appears on the output of achannel.1.9 Mode. Steady-State. Steady-state mode operation shall mean operation of the UCDIMNRC reactorwith the selector switch in the automatic or manual mode position. " 1.10 Mode. SQuare-Wave. Square-wave mode operation shall mean operation of the UCD/MNRCreactor with the selector switch in the square-wave mode position. 1.11 Mode. Pulse. Pulse mode operation shall mean operation of the UCD/MNRC reactor with theselector switch in the pulse mode position. 1.12 Operable. Operable means a component or system is capable of performing its intended function.. 1.13 Operat~ina. Operating means a component or system is performing its intended function. 1.14.0peratina Cycle. The period of time starting with reactor startup and ending with reactor shutdown. 1.15 Protective Action. Protective action is the initiation of a signal or the operation of equipment withinthe UCDIMNRC reactor safety system in response to a variable or condition of the UCDIMNRC reactorfacility having reached a specified limit.1.15.1 Channel Level. At the protective instrument channel level, protective action is thegeneration and transmission of a scram signal indicating that a reactor variable has reached thespecified limit.1.15.2 Subsystem Level. At the protective instrument subsystem level, protective action is thegeneration and transmission of a scram signal indicating that a specified limit has been reached.NOTE: Protective action at this level would lead to the operation of the safety shutdowni: equipment. 2 1.15.3 Instrument System Level. At the protective instrument level, protective action is thegeneration and transmission of the command signal for the safety shutdown equipment tooperate.1.15.4 Safety System Level. At the reactor safety system level, protective action is the operation of sufficient equipment to immediately shut down the reactor.1.16 Pulse Operation~al Core. A pulse operational core is a reactor operational core for which themaximum allowable pulse reactivity insertion has been determined. 1; 17 Reactivity. Exce~ss, Excess reactivity is that amount of reactivity that would exist if all control rods(control, regulating, etc.) were moved to the maximum reactive position from the point where the reactoris at ambient temperature and the reactor is critical. (K o, = 1)1.18 Reactivity Limit~s. The reactivity limits are those limits imposed on the reactivity conditions of thereactor core.1.19 R~eactivity Worth of an Exoeriment. The reactivity worth of an experiment is the maximum value ofthe reactivity change that could occur as a result of changes that alter experiment position orconfiguration. 1.20 Reactor Controls. Reactor controls are apparatus and/or mechanisms the manipulation of whichdirectly affect the reactivity or power level of the reactor.1.21 R.eac~tor Core. Operational. The UCD/MNRC reactor operational core is a core for which theparameters of excess reactivity, shutdown margin, fuel temperature, power calibration and reactivity worths of control rods and experiments have been determined to satisfy the requirements set forth inthese Technical Specifications. 1.22 Reactor Ooeratingq. The UCO/MNRC reactor is operating whenever it is not shutdown or secured.1.23 R~eactor Safety Systems. Reactor safety systems are those systems, including their associated input channels, which are designed to initiate automatic reactor protection or to provide information forinitiation of manual protective action.1.24 R~eactor Secured. The UCO/MNRC reactor is secured when the console key switch is in the offposition and the key is removed from the lock and under the control of a licensed

operator, and theconditions of a or b exist:a. (1) The minimum number of control rods are fully inserted to ensure the reactor is shutdown, asrequired by technical specifications; and(2) No work is in progress involving core fuel, core structure, installed control rods, or control roddrives, unless the control rod drives are physically decoupled from the control rods; and(3) No experiments in any reactor experiment

facility, or in any other way .near the reactor, are beingmoved or serviced if the experiments have, on movement, a reactivity worth exceeding the maximumvalue allowed for a single experiment or $1.00, whichever is smaller, orb. The reactor contains insufficient fissile materials in the reactor core, a~djacent experiments or controlrods to attain criticality under optimum available conditions of moderation and reflection.

1.25 Reactor Shut~down. The UCD/MNRC reactor is shutdown if it is subcritical by at least one dollar($1.00) both in the Reference Core Condition and for all allowed ambient conditions with the reactivity worth of all installed experiments included. 3

• 01.26 Reference Cpre Condition.

The condition of the core when it is at ambient temperature (cold T<280C), the reactivity worth of xenon is negligible (< $0.30) (i.e., cold and clean), and the central irradiation facility contains the graphite thimble plug and the aluminum thimble plug (CIF-1).1.27 ReserhRatr A research reactor is defined as a device designed to support a self-sustaining neutron chain reaction for research development, education, and training, or experimental

purposes, andwhich may have provisions for the production of radioisotopes.

1.28 Rod. Control, A control rod is a device fabricated from neutron absorbing

material, with or without afuel or air follower, which is used to establish neutron flux changes and to compensate for routinereactivity losses. The follower may be a stainless steel section.

A control rod shall be coupled to itsdrive unit to allow it to perform its control function, and its safety function when the coupling isdisengaged. This safety function is commonly termed a scram.1.28.1 Regulat~ing Rod. A regulating rod is a control rod used to maintain an intended powerlevel and may be varied manually or by a servo-controller. A regulating rod shall have scramcapability. 1.28.2 Standard Rod. The regulating and shim rods are standard control rods.1.28.3 Transient Rod. The transient rod is a control rod that is capable of providing rapidreactivity insertion to produce a pulse or square wave.1.29 Safety Channel. A safety channel is a measuring channel in the reactor safety system.1.30 Safety Limit. Safety limits are limits on important process variables, which are found to benecessary to reasonably protect the integrity of the principal barriers which guard against theuncontrolled release of radioactivity. 1.31 Scram Time. Scram time is the elapsed time between reaching a limiting safety system set pointand the control rods being fully inserted. 1.32 Scram. External. External scrams may arise from the radiography bay doors, radiography bayripcords, bay shutter interlocks, and any scrams from an experiment. 1.33 Shall. Should and May. The word "shall" is used to denote a requirement; the word "should" todenote a recommendation; the word "may" to denote, permission, neither a requirement nor arecommendation. 1.34 Shut~down Margin. Shutdown margin shall mean the minimum shutdown reactivity necessary toprovide confidence that the reactor can be made subcritical by means of the control and safety systemstarting from any permissible operating condition with the most reactive rod assumed to be in the mostreactive

position, and once this action has been initiated, the reactor will remain subcritical withoutfurther operator action.1.35 Shutdown.

Unsched.u.led. An unscheduled shutdown is any unplanned shutdown of theUCD/M NRC reactor caused by actuation of the reactor safety system, operator error, equipment malfunction, or a manual shutdown in response to conditions which could adversely affect safeoperation, not including shutdowns which occur during testing or check-out operations. 1.36 Surveillance Acfivities. In general, two types of surveillance activitiesare specified: operability checks and tests, and calibrations. Operability checks and tests are generally specified as daily, weeklyor quarterly. Calibration times are generally specified as quarterly, semi-annually,

annually, or biennially.

..... 1.37 Surveillance Intervals. Maximum intervals are established to provide operational flexibility and not/) to reduce frequency. Established frequencies shall be maintained over the long term. The allowable 4 0 0***surveillance interval is the interval between a check, test, or calibration, whichever is appropriate to the itembeing subjected to the surveillance, and is measured from the date of the last surveillance. Allowable surveillance intervals shall not exceed the following: 1.37.1 Annual -interval not to exceed fifteen (15) months.1.37.2 Semiannual -interval not to exceed seven and a half (7.5) months.1.37.3 Quarterly -interval not to exceed four (4) months.1.37.4 Mothy- interval not to exceed six (6) weeks.1.37.5 Wee..y- interval not to exceed ten (10) days.1.38 Unreviewed.Safety Questions. A proposed change, test or experiment shall be deemed to involvean unreviewed safety question:

a. If the probability of occurrence or the consequences of an accident or malfunction ofequipment important to safety previously evaluated in the safety analysis report may beincreased; orb. If a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; orc. If the margin of safety, as defined in the Basis for any technical specification, is reduced.1.39'Value.

Measured. The measured value is the value of a parameter as it appears on the output of a* ,. channel." ~1.40 Value, Tr~ue. The true value is the actual value of a parameter. 1.41 Watc.h~doa Circuit. The watchdog circuit is a surveillance circuit provided by the Data Acquisition Computer (DAC) and the Control System Computer (CSC) to ensure proper operation of the reactorcomputerized control system.2.0 Safety Limit an~d Limiting Safety System Setting (LSSS).2.1. Safety Limits.Applicability -This specification applies to the temperature of the reactor fuel in a standard TRIGA fuelelement.Obiective -The objective is to define the maximum temperature that can be- permitted with confidence

• that no damage to the fuel element cladding will result.Specific~ation

-a. The maximum fuel temperature in a standard TRIGA fuel element shall not exceed 930 0C duringsteady-state operation.

b. The maximum ten'perature in a standard TRIGA fuel element shall not exceed 1100 0C during pulseoperation.

....... 'a. This fuel safety limit applies for coniditions in which the cladding temperature is above 500 °C (SafetyAnalysis Report (SAR), Chapter 4, Section 4.5.4.1.3). The important parameter for a TRIGA reactor is5 0 =°the fuel element temperature. This parameter is well suited as it can be measured directly. A loss in theintegrity of the fuel element cladding could arise if the cladding stress exceeds the ultimate strength ofthe cladding material. The fuel element cladding stress is a function of the element's internal pressurewhile the ultimate strength of the cladding material is a function of its temperature. The cladding stressis a result of the internal pressure due to the presence of air, fission product gasses and hydrogen fromthe disassociation of hydrogen and zirconium in the fuel moderator. Hydrogen pressure is the mostsignificant. The magnitude of the pressure is determined by the fuel moderator temperature and theratio of hydrogen to zirconium in the alloy. At a fuel temperature of 930 0C for ZrH 1.7 fuel, the claddingstress due to the internal pressure is equal to the ultimate strength of the cladding material at the sametemperature (SAR Fig 4.18). This is a conservative limit since the temperature of the cladding material isalways lower than the fuel temperature. (See SAR Chapter 4, Section 4.5.4.)b. This fuel safety limit applies for conditions in which the cladding temperature is less than 500 °C.Analysis (SAR Chapter 4, Section 4.5.4.1.1), shows that a maximum temperature for the clad during apulse which gives a peak adiabatic fuel temperature of 1000 °C is estimated to be 470 °C. Furtheranalysis (SAR Section 4.5.4.1.2), shows that the internal pressure for both Zr (at 11500C) and Zr17z (at11 00°C) increases to a peak value at about 0.3 sec, at which time the pressure is about one-fifth of theequilibrium value or about 400 psi (a stress of 14,700 psi). The yield strength of the cladding at 500 0C isabout 59,000 psi.Calculations for step increases in power to peak ZrH 1.85 fuel temperature greater than I115 0 C, over a 200°Crange, show that the time to reach the peak pressure and the fraction of equilibrium pressure value achievedwere approximately the same as for the 1150 °C case. Similar results were found for fuel with ZrH1 .7.Measurements of hydrogen pressure in TRIGA fuel elements during transient operations have been made andcompared with the results of analysis similar to that used to make the above prediction. These measurements indicate that in a pulse where the maximum temperature in the fuel was greater than 1000 0C, the pressure(ZrH1.=) was only about 6% of the equilibrium value evaluated at the peak temperature. Calculations of thepressure gave values about three times greater than the measured values. The analysis gives strongindications that the cladding will not rupture if fuel temperatures are never greater than 1200 °C to 1250°C,providing the cladding temperature is less than 500 0 C. For fuel with ZrH 1.7,a conservative safety limit is 1100 0C.As a result, at this safety limit temperature, the class pressure is a factor of 4 lower than would be necessary forcladding failure.2.2 Limiting Safety System Setting.2.2.1 Fuel Temperature. Applicability -This specification applies to the protective action for the reactor fuel elementtemperature. Obiective -The objective is to prevent the fuel element temperature safety limit from beingreached.Specification -The limiting safety system setting shall be 750 0C (operationally this may be setmore conservatively) as measured in an instrumented fuel element. One instrumented elementshall be located in the analyzed peak power location of the reactor operational core.ass- For steady-state operation of the reactor, the limiting safety system setting is atemperature which, if exceeded, shall cause a reactor scram to be initiated preventing the safetylimit from being exceeded. A setting of 750 °C provides a safety margin at the point of themeasuremenrA of at least 137 °C for standard TRIGA fuel elements in any condition of operation. A part of the safety margin is used to account for the difference between the true and measuredtemperatures resulting from the actual location of the thermocouple. If the thermocouple element is located in the hottest position in the core, the difference between the true and!. / measured temperatures will be only a few degrees since the thermocouple junction is near thecenter and mid-plane of the fuel element. For pulse operation of the reactor, the same limitingsafety system setting applies.

However, the temperature channel will have no effect on limiting6 O ......0the peak power generated because of its relatively long time constant (seconds) as comparedwith the width of the pulse (milliseconds).

in this mode, however, the temperature trip will act to"\ limit the energy release after the pulse if the transient rod should not reinsert and the fuel! temperature continues to increase. 3.0 Limiting Conditions For 3.1 Reactor Core Parameters 3.1.1 Steady-State Ooeration Ajj lcblv- This specification applies to the maximum reactor power attained during steady-state operation. Obiective -The objective is to assure that the reactor safety limit (fuel temperature) is notexceeded, and to provide for a setpoint for the high flux limiting safety systems, so thatautomatic protective action will prevent the safety limit from being reached during steady-state operation. Soecification -The nominal reactor steady-state .power shall not exceed 2.0 MW. The automatic scram setpoints for the reactor power level safety channels shall be set at 2.2 MW or less. Forthe purpose of testing the reactor steady-state power level scram, the power shall not exceed2.3 MW.Basis -Operational experience and thermal-hydraulic calculations demonstrate that UCD/MNRCTRIGA fuel elements may be safely operated at power levels up to 2.3 MW with naturalconvection cooling. (SAR Chapter 4, Section 4.6.2.).3.1.2 Pulse or Square Wave Operation Aoolicabilitv -This specification applies to the peak temperature generated in the fuel as theresult of a step insertion of reactivity. Obiective -The objective is to assure that the fuel temperature safety limit will not be exceeded. Specification -a. For the pulse mode of operation, the maximum insertion of reactivity shall be 1.23% A dkf($1.75);b. For the square wave mode of operation, the maximum insertion of reactivity shall be 0.63%Ak/k ($0.90).Basis -Standard TRIGA fuel is fabricated with a nominal hydrogen to zirconium ratio of 1.6 to1.7. This yields delta phase zirconium hydride which has a high creep strength and undergoes no phase changes at temperatures in excess of 100 0C. However, after extensive steady stateoperation at two (2) MW the hydrogen will redistribute due to migration from the central hightemperature regions of the fuel to the cooler outer regions. When the fuel is pulsed, theinstantaneous temperature distribution is such that the highest values occur at the radial edge ofthe fuel. The higher temperatures in the outer regions occur in fuel with a hydrogen to zirconium ratio that hasnow increased above the nominal value. This produces hydrogen gas pressures considerably in excess of that expected. If the pulse insertion is such that the temperature of thefuel exceeds about 875 0C, then the pressure may be sufficient to cause expansion ofmicroscopic holes in the fuel that grow with each pulse. Analysis (SAR Chapter 13, Section.II 13.2.2.2.1), shows that the limiting pulse, for the worst case conditions, is 1.34% A k/k ($1.92).Therefore, the 1.23% Ak/dk ($1.75) limit is below the worse case reactivity insertion accident limit.7

• ... ... .The $0.90 square wave step insertion limit is also well below the worse case reactivity insertion accident limit.) 3.1.3 Reactivity Limitations Aolcbiiv-These specifications apply to the reactivity conditions of the reactor core and thereactivity worths of the control rods and apply to all modes of reactor operation.

bicie- The objective is to assure that the reactor can be placed in a shutdown condition atall times and to assure that the safety limit shall not be exceeded. Specification -a. Shutdo wn Marginl -The reactor shall not be operated unless the shutdown margin provided bythe control rods is greater than 0.35% k/k ($0.50) with:(1) The reactor in any core condition, (2) The most reactive control rod assumed fully withdrawn, and(3) Absolute value of all movable experiments analyzed in their most reactive condition or $1.00 whichever is greater.b. Excess Reactivity -The maximum available excess reactivity (reference core condition) shallnot exceed 6.65% A k/k ($9.50).Basis -a. This specification assures that the reactor can be placed in a shutdown condition from anyoperating condition and remain shutdown, even if the maximum worth control rod should stick inthe fully withdrawn position (SAR Chapter 4, Section 4.5.5).b. This specification sets an overall reactivity limit which provides adequate excess reactivity tooverride the xenon buildup, to overcome the temperature change in going from zero power to 2MW, to permit pulsing at the $1.75 level, to permit irradiation of negative worth experiments andaccount for fuel burnup over time. An adequate shutdown margin exists with an excess of $9.50for the two analyzed cores: (SAR Chapter 4, Section 4.5.5).3.2 Reactor Control and Safety Systems3.2.1 Control RodsAooljcjili*- This specification applies to the function of the control rods.Obiect~ive -The objective is to determine that the control rods are operable. Specificati~on -The reactor shall not be operated unless the control rods are operable and,a. Control rods shall not be considered operable if damage is apparent to the rod or driveassemblies.

b. The scram' time measured from the instant a signal reaches the value of a limiting safetysystem setting to the instant that the slowest control rod reaches its fully inserted position shall.not exceed one (1) second.8 0 Sa. The apparent condition of the control rod assemblies shall provide assurance that the rodsshall continue to perform reliably as designed.

b. This assures that the reactor shall shut down promptly when a scram signal is initiated (SARChapter 13, Section 13.2.2.2.2).

3.2.2 Reactor Instrumentation Apolicability -This specification applies to the information which shall be available to the reactoroperator during reactor operations. Obiective -The objective is to require that sufficient information is available to the operator toassure safe operation of the reactor.Specification -The reactor shall not be operated unless the channels described in Table 3.2.2are operable and the information is displayed on the reactor console.Table 3.2.2Required Reactor Instrumentation (Minimum Number Operable) Measuring Steady Square Channel Surveillance Channel Stt Pulse Wave Function Requirements*

a. Reactor Power 202Scram at 2.2 D,M,ALevel Safety MW or lessChannelb5. Linear Power 101Automatic D,M,AChannel -Power Controlc. Log Power 101Startup D,M,AChannel Controld. Fuel Temperature 2 2 2 Fuel D,M,AChannel Temperature

e. Pulse Channel 0 10Measures P,APulse NV & NVTr(*) Where: D -Channel check during each day's operation M -Channel test monthlyA -Channel calibration annuallyP -Channel test prior to pulsing operation

a. Table 3,2.2. The two reactor power level safety channels assure that the reactor power levelis properly mdonitored and indicated in the reactor control room (SAR Chapter 7, Sections 7.1.2 &7.1.2.2).

b. c. & e. Table 3.2.2. The linear power channel, log power channel, and pulse channel assurethat the reactor power level and energy are adequately monitored (SAR Chapter 7, Sections.... 7.1.2 & 7.1.2.2).

9 Ia1,.1.2 The fuel temperature channels assure that the fuel temperature is properlymonitored and indicated in the reactor control room (SAR Chapter 4, Section 4.5.4.1). 3.2.3 Reactor Scrams and Interlocks Agllcbjlity-This specification applies to the scrams and interlocks. Obiective -The objective is to assure that the reactor is placed in the shutdown condition promptly and that the scrams and interlocks are operable for safe operation of the reactor.Specification -The reactor shall not be operated unless the scrams and interlocks described inTable 3.2.3 are operable: Table 3.2.3Required Scrams and I!nterlocks SteadyStateScrama. ConsoleManualScramb. Reactor RoomManual ScramI1PulseIISquareWave.1ChannelSurveillance Reouirements* Manual Scramand Automatic Scram AlarmManual Scramand Automatic Scram Alarmc. Radiography Bay ManualScramsd. Reactor PowerLevel SafetyScramse. High VoltagePower SuppliesScramsf. FuelTemperature Scramsg. Watchdog.Circuit4420422Manual Scramsand Automatic Scram AlarmsAutomatic Scram Alarms & Scramsat 2.2 MW or lessAutomatic Scram Alarms &Scrams onLosfHigh Voltage tothe Reactor PowerLevel SafetyChannelsMMMMMMM212222Automatic ScramAlarms & Scramson indicated fueltemperature of750°C or lessAutomatic ScramAlarms & Scrams2210N., 0h. ExternalScrams22\i. One KilowattPulse &Square WaveInterlock

j. Low SourceLevel RodWithdrawal ProhibitInterlock

k. Control RodWithdrawal Interlock I. MagnetPower KeySwitch Scram0111Automatic Scrams and Alarmsif an experiment or radiography scram interlock

• is activated Prevents initiation of a step reactivity insertion above areactor power levelof 1 KWPrevents withdrawal of any control rodif the log channelreads less than 1.5times the indicated log channel currentlevel with the neutronsource removed fromthe corePrevents simul-taneous withdrawal of two or more rodsin manual modeDe-energizes thecontrol rodmagnets, scram &alarmMMMMI1111M(*) Where: M -channel test monthlyBasis -a. Table 3.2.3. The console manual scram allows rapid shutdown of the reactor from the controlroom (SAR Chapter 7, Section 7.1.2.5).

b. Table 3..2.3. The reactor room manual scram allows rapid shutdown of the reactor from thereactor room.p,. Table 3.2.3. The radiography bay manual scrams allow rapid shutdown of the reactor fromany of the radiography bays (SAR Chapter 9, Section 9.6.3).d~i.Tabe32.3.

The automatic power level safety scram assures the reactor will be shutdown ifthe power level exceeds 2.2 MW, therefore not exceeding the safety limit (SAR Chapter 4,Section 4.7.2).e. Table The loss-of-high-voltage scram assures that the reactor power level safetychannels, operate within their intended range as required for proper functioning of the powerlevel scrams (SAR Chapter 7, Sections 7.1.2.1 & 7.1.2.2).

f. Table 3.2.3. The fuel temperature scrams assure that the reactor will be shut down if the fueltemperature exceeds 7500o C, therefore ensuring the safety limit will not be exceeded (SARChapter 4, Sections 4.5.4.1 & 4.7.2).* )11

a. Table 3.2.3, The watchdog circuits assure that the control system computer and the dataacquisition computer are functioning properly (SAR Chapter 7, Section 7.2).h. Table.3.2.3, The external scrams assure that the reactor will be shut down if the radiography bay doors and reactor concrete shutters are not in the proper position for personnel entry intothe bays (SAR Chapter 9, Section 9.6). External scrams from experiments, a subset of theexternal scrams, also assure the integrity of the reactor system, the experiment, the facility, andthe safety of the facility personnel and the public.i. Table 3.2.3. The interlock preventing the initiation of a step reactivity insertion at a level aboveone (1) kilowatt assures that the pulse magnitude will not allow the fuel element temperature toexceed the safety limit (SAR Chapter 7, Section 7.1.2.5).

i. Table 3.2.3. The low source level rod withdrawal prohibit interlock assures an adequate sourceof neutrons is present for safe startup of the reactor (SAR Chapter 7, Section 7.1.2.5).

k. Table 3.2.3. The control rod withdrawal interlock prevents the simultaneous withdrawal of twoor more control rods thus limiting the reactivity-insertion rate from the control rods in manualmode (SAR Chapter 7, Section 7.1.2.5).

I. Table 3.2.3.. The magnet current key switch prevents the control rods frown being energized without inserting the key. Turning off the magnet current key switch de-energizes the control rodmagnets and results in a scram (SAR Chapter 7, Section 7.1.2.5). 3.2.4 Rea~ctor Fuel ElementsAoolicabilitv -This specification applies to the physical dimensions of the fuel elements asmeasured on the last surveillance test.Objective -The objective is to veri{fy the integrity of the fuel-element cladding. Specification -The reactor shall not be used for normal operation with damaged fuel. All fuelelements shall be inspected visually for damage or deterioration as per Technical Specifications Section 4.2.4. A fuel element shall be considered damaged and must be removed from the coreif:a. In measuring the transverse bend, the bend exceeds 0.125 inch (3.175 mam) over the fulllength 23 inches (584 mm) of the cladding, or,b. In measuring the elongation, its length exceeds its initial length by 0.125 inch (3.175 mam), or,c. A cladding failure exists as indicated by measurable release of fission products, or,d. Visual inspection identifies bulges, gross pitting, or corrosion. Basis. -The most severe stresses induced in the fuel elements result from pulse operation of thereactor, during which differential expansion between the fuel and the cladding occurs and thepressure of the gases within the elements increases sharply. The above limits on the allowable distortion of a fuel element correspond to strains that are considerably lower than the strainexpected to ".ause rupturing of a fuel element. Limited operation in the steady state or pulsedmode may be necessary to identify a leaking fuel element especially if the leak is small.12 3.3 Reactor Coolant SystemsAoolicability -These specifications apply to the operation of the reactor water measuring systems.Objective. -The objective is to assure that adequate cooling is provided to maintain fuel temperatures below the safety limit, and that the water quality remains high to prevent damage to the reactor fuel.Specification -The reactor shall not be operated unless the systems and instrumentation channelsdescribed in Table 3.3 are operable, and the information is displayed locally or in the control room.Table 3.3REQUIRED WATERSYSTEMS AND INSTRUMENTATION I I I n I ....Measuring Chann~el/System MinimumNumberOoe~rable. Surveillance Requirements*

a. Primary CoolantCore InletTemperature Monitorb. Reactor TankLow Water*Monitorc. Purification**

Inlet Conduc-tivity Monitord. Emergency CoreCooling SystemI11IFunction: Channel/System For operation of thereactor at 1.5 MW orhigher, alarms on highheat exchanger outlettemperature of 45 °C(113°F)Alarms if water leveldrops below a depth of23 feet in the reactor tankAlarms if the primarycoolant water conductivity is greater than5 micromhos/cm For operation of the reactorat 1.5MW or higher, provideswater to cool fuel in the eventof a Lois of Coolant Accidentfor a minimum of 3.7 hoursat 20 gpm from an appropriate nozzleD,Q,AM(.DM,SD,S(*) Where: D -channel check during each day's operation A -channel calibration annuallyQ -channel test quarterly S -channel calibration semiannually M -channel test monthly(**) The purification inlet conductivity monitor can be out-of-service for no more than 3 hours before the reactorshall be shutdown. Basis -a.Table 3.3, The primary coolant core inlet temperature alarm assures that large power fluctuations willnot occur (S.AR Chapter 4, Section 4.6.2).13

b. Table 3,3, The minimum height of 23 ft. of water above the reactor tank bottom guarantees that thereis sufficient water for effective cooling of the fuel and that the radiation levels at the top of the reactor* .',tank are within acceptable limits. The reactor tank water level monitor alarms if the water level drops) below a height of 23 ft. (7.01m) above the tank bottom (SAR Chapter 11, Section 11.1.5.1).

c. Table 3.3. Maintaining the primary coolant water conductivity below 5 micromhos/cm averaged over aweek will minimize the activation of water impurities and also the corrosion of the reactor structure.

d. Table 3.3. This system will mitigate the.Loss of Coolant Accident event analyzed in the SAR Chapter13, Section 13.2.3.4 Reactor Room Exhaust; SystemAoplicability

-These specifications apply to the operation of the reactor room exhaust system.Obiective -The objectives of this specification are as follows:a. To reduce concentrations of airborne radioactive material in the reactor room, and maintain thereactor room pressure negative with respect to surrounding areas.b. To assure continuous air flow through the reactor room in the event of a Loss of Coolant Accident. Soecification -a. The reactor shall not be operated unless the reactor room exhaust system is in operation and thepressure in the reactor room is negative relative to surrounding areas.~b. The reactor room exhaust system shall be operable within one half hour of the onset of a Loss ofCoolant Accident. Basi__.s -Operation of the reactor room exhaust system assures that:a. Concentrations of airborne radioactive material in the reactor room and in air leaving the reactor roomwill be reduced due to mixing with exhaust system air (SAR Chapter 9, Section 9.5.1). Pressure in thereactor room will be negative relative to surrounding areas due to air flow patterns created by the reactorroom exhaust system (SAR Chapter 9, Section 6.5.1).b. There will be a timely, adequate and continuous air flow through the reactor room to keep the fueltemperature below the safety limit in the event of a Loss of Coolant Accident. 3.5 This section intentionally left blank.3.6 T~his section intentionally left blan~k.3.7 Reactor Radiation Monitoring Systems.3.7.1 Monitoring SystemsAnoDlicability -This specification applies to the information which shall be available to the reactoroperator during reactor operation. Obiective -The objective is to require that sufficient information regarding radiation levels and, radioactive effluents is available to the reactor operator to assure safe operation of the reactor...... ' Specfication .-The reactor shall not be operated unless the channels described in Table 3.7.1are operable, the readings are below the alarm setpoints, and the information is displayed in the14 control room. The stack and reactor room CAMS shall not be shutdown at the same time duringreactor operation. Table 3.7.1REQUIRED RADIATION MONITORING INSTRUMENTATION Measuring Eduioment MinimumNumberO, erable**ChannelFunctionSurveillance Requirements*

a. FacilityStack Monitorb. Reactor RoomRadiation Monitorc. Purification System Radia-tion Monitord. Reactor RoomContinuous Air MonitorI111Monitors Argon-41 andradioactive particu-lates, and alarmsMonitors the radiation level in the reactorroom and alarmsMonitors radiation level at the demineral-izer station and alarmsMonitors air from thereactor room for parti-culate and gaseousradioactivity and alarmsD,W,AD),W,AD:,W,AD ,W.A(*) Where: D -channel check during each day'ls operation A -channel calibration annuallyW -channel test(**) monitors may be placed out-of-service for up to 2 hours for calibration and maintenance.

During this out-of-service time, no experiment or maintenance activities shall be conducted which could result in alarm conditions (e.g., airborne releases or high radiation levels)Basis -a. Thble 3.7.1. The facility stack monitor provides information to operating personnel regarding therelease of radioactive material to the environment (SAR Chapter 11, Section 11.1.1.1.4). The alarmsetpoint on the facility stack monitor is set to limit Argon-41 concentrations to less than 10 CFR Part 20,Appendix B, Table 2, Column 1 values (averaged over one year) for unrestricted locations outside theoperations area.b. Table 3.7.1. The reactor room radiation monitor provides information regarding radiation levels in thereactor room during reactor operation (SAR Chapter 11, Section 11.1.5.1 ), to limit occupational radiation exposure to less than 10 CFR 20 limits.c.; Table 3.7.1. The radiation monitor located next to the purification system resin cannisters providesinformation regarding radioactivity in the primary system cooling water (SAR Chapter 11, Section11.1.5.4.2) and allowS, assessment of radiation levels in the area to ensure that personnel radiation doses will be below 10 CFR Part 20 limits.d. Table 3.7.1. The reactor room continuous air monitor provides information regarding airborneradioactivity in the reactor room, (SAR Chapter 11, Sections 11.1.1.1.2 & 11.1.1.1.5), to ensure thatoccupational exposure to airborne radioactivity will remain below the 10 CFR Part 20 limits..,:15 3.7.2 .Effluents -.Argon-41 Discharge LimitAppJicaiity~- This specification applies to the concentration of Argon-41 that may be discharged from the UCD/MNRC reactor facility, Obiective -The objective is to ensure that the health and safety of the public is not endangered by the discharge of Argon-41 from the UCD/MNRC reactor facility. Soecification -The annual average unrestricted area concentration of Argon-41 due to releasesof this radionuclide from the UCD/MNRG, and the corresponding annual radiation dose fromArgon-41 in the unrestricted area shall not exceed the applicable levels in 10 CFR Part 20.Basis -The annual average concentration limit for Argon-41 in air in the unrestricted area isspecified in Appendix B, Table 2, Column 1 of 10 CFR Part 20.10 CFR 20.1301 specifies doselimitations in the unrestricted area. 10 CFR 20.1101 specifies a constraint on air emissions ofradioactive material to the environment. The SAR Chapter 11, Section 11.1.1.1.4 estimates thatthe routine Argon-41 releases and the corresponding doses in the unrestricted area will bebelow these limits.3.8 Experiments 3.8.1 React~ivity Limnits.Applicability -This specification applies to the reactivity limits on experiments installed in specificreactor experiment facilities. Obiective -The objective is to assure control of the reactor during the irradiation or handling ofexperiments in the specifically designated reactor experiment facilities. Specification -The reactor shall not be operated unless the following conditions governing experiments exist:a. The absolute reactivity worth of any single moveable experiment in the pneumatic transfertube, the central irradiation

facility, the central irradiation fixture 1 (CIF-1),

or any other in-core'or in-tank irradiation

facility, shall be less than $1.00 (0.7% A k/k), except .for the automated centralirradiation facility (ACIF) (See 3.8.1.c below).b. The absolute reactivity worth of any single secured experiment positioned in a reactor in-coreor in-tank irradiation facility shall be less than the maximum allowed pulse ($1.75) (1.23% A k/k).c. The absolute total reactivity worth of any single experiment or of all experiments collectively positioned in the ACIF shall be less than the maximum allowed pulse ($1.75) (1.23% A k/k).d. The absolute total reactivity of all experiments positioned in the pneumatic transfer tube, andin any other reactor in-core and in-tank irradiation facilities at any given time shall be less thanone dollar and ninety-two cents ($1.92) (1.34% A k/k), including the potential reactivity whichmight result from malfunction,

flooding, voiding, or removal and insertion of the experiments.

Basis -a. A limitatiodn of less than one dollar ($1.00) (0.7% A k/k) on the reactivity worth of a singlemovable experiment positioned in the pneumatic transfer tube, the central irradiation facility(SAR, Chapter 10, Section 10.4.1), the central irradiation fixture-I (CIF-1) (SAR Chapter 10,Section 10.4.1), or any other in-core or in-tank irradiation

facility, will assure that the pulse limitof $1 .75 is not exceeded (SAR Chapter 13, Section 13.2.2.2.1

). In addition, limiting the worth ofeach movable experiment to less than $1.00 will assure that the additional increase in transient 16 power and temperature will be slow enough so that the fuel temperature scram will be effective (SAR Chapter 13, Section 13.2.2.2.1).

b. The absolute worst event which may be considered in conjunction with a single securedexperiment is its sudden accidental or unplanned removal while the reactor is operating.

Forsuch an event, the reactivity limit for fixed experiments ($1.75) would result in a reactivity increase less than the $1.92 pulse reactivity insertion needed to reach the fuel temperature safety limit (SAR Chapter 13, Section 13.2.2.2.1).

c. A reactivity limit of less than $1.75 for any single experiment or for all experiments collectively positioned in the sample can of the automated central irradiation facility (ACIP) (SAR Chapter10, Section 10.4.2) is based on the pulsing reactivity insertion limit (Technical Specification 3.1.2) (SAR Chapter 13, Section 13.2.2.2.1) and on the design of the ACIF, which allows controlover the positioning of samples into and out of the central core region in a manner identical inform, fit, and function to a control rod.d. It is conservatively assumed that simultaneous removal of all experiments positioned in thepneumatic transfer tube, and in any other reactor in-core and in-tank irradiation facilities at anygiven time shall be less thanthe maximum reactivity insertion limit of $1.92. The SAR Chapter13, Section 13.2.2.2.1 indicates that a pulse reactivity insertion of $1.92 would be needed toreach the fuel temperature safety limit.3.8.2 .Materials LimitAoplicability

-This specification applies to experiments installed in reactor experiment facilities. Objective -The objective is to prevent damage to the reactor or significant releases of.,. by limiting material quantity and the radioactive material inventory of theexperiment. .Specification -The reactor shall nct be operated unless the following conditions governing experiment materials exist;a. Experiments containing materials corrosive to reactor components, compounds highlyreactive with water, potentially explosive materials, and liquid, fissionable materials shall beappropriately encapsulated.

b. Each fueled experiment shall be controlled such that the total inventory of iodine isotopes 131through 135 in the experiment is no greater than 1.5 curies and the maximum strontium inventory is no greater than 5 millicuries.

c. Each experiment in the 1-125 production facility shall be controlled such that the totalinventory of 1-125 in the 1-125 glove box shall not exceed 40 curies.d. Each experiment in the 1-125 production facility shall be controlled such that the totalinventory of 1-125 being handled in the 1-125 fume hood at any one time in preparation forshipping shall not exceed 20 curies. An Additonal 1.0 curie of 1-125 (up to 400 millicuries in theform of quality assurance samples and up to 600 millicuries in sealed storage containers) mayalso be present in the 1-125 fume hood.e. Explosive

.materials in quantities greater than 25 milligrams of TNT eqluivalent shall not beirradiated in th~e reactor tank. Explosive materials in quantities of 25 milligrams of TNTequivalent or less may be irradiatedl provided the pressure produced upon detonation of theexplosive has been calculated and/or experimentally demonstrated to be less than the designpressure of the container.

f. Explosive materials in quantities of three (3) pounds of TNT equivalent or less may beirradiated in any radiography bay. The irradiation of explosives in any bay is limited to those.17

...... " 0assemblies where a safety analysis has been performed that shows that there is no damage tothe reactor safety systems upon detonation (SAR Chapter 13, Section 13.2.6.2). ") Basis -a. Appropriate encapsulation is required to lessen the experimental hazards of some types ofmaterials.

b. The 1.5 curies limitation on iodine 131 through 135 assures that in the event of failure of afueled experiment leading to total release of the iodine, occupational doses and doses tomembers of the general public in the unrestricted areas shall be within the limits in 10 CFR 20(SAR Chapter 13, Section 13.2.6.2).

c&d. Limiting the total 1-125 inventory to forty (40.0) curies in the 1-125 glove box and to twenty-one (21.0) curies in the 1-125 .fume hood assures that, if either of these inventories of 1-125 istotally released into its respective containment, or if both inventories are simultaneously released into their respective containments, the occupational doses and doses to members ofthe general public in the unrestricted areas will be within the limits of 10 CFR 20 (SAR Chapter13, Section 13.2.6.2).

e. This specification is intended to prevent damage to vital equipment by restricting the quantityof explosive materials within the reactor tank (SAR Chapter 13, Section 13.2.6.2).

f. The failure of an experiment involving the irradiation of 3 lbs TNT equivalent or less in anyradiography bay external to the reactor tank will not result in damage to the reactor controls orthe reactor tank. Safety Analyses have been performed (SAR Chapter 13, Section .13.2.6.2) which show that up to six (6) pounds of TNT equivalent can be safely irradiated in any.*. radiogaphy bay. Therefore, the three (3) pound limit gives a safety margin of two (2).3.8.3 Failure and Malfunctions Applicability

-This specification applies to experiments installed in reactor experiment facilities. Objective -The objective is to prevent damage to the reactor or significant releases ofradioactive materials in the event of an experiment failure.S~ecification -a. All experiment materials which could off-gas,

sublime, volatilize, or produce aerosols under:(1) normal operating conditions of the experiment or the reactor,(2) credible accident conditions in the reactor, or(3) where the possibility exists that the failure of an experiment could release radioactive gases or aerosols into the reactor building or into the unrestricted area, the quantity andtype of material in the experiment shall be limited such that the airborne radioactivity inthe reactor room will not result in exceeding the applicable dose limits in 10 CFR Part 20in the unrestricted area, assuming 100% of the gases or aerosols escapes.b. In calculatio[ns pursuant to (a) above, the following assumptions shall be used:(1) If the effluent from an experiment facility exhausts through a stack which is closed on* ~.* .,high radiation levels, at least 10% of the gaseous activity or aerosols produced will)* escape.

(2) If the effluent from an experiment facility exhausts through a filter installation designed for greater than 9g% efficiency for 0.3 micron and larger particles, at least 10%of these will escape.(3) For materials whose boiling point is above 130 00 and where vapors formed byboiling this material can escape only through an undistributed column of water above thecore, at least 10% of these vapors can escape.c. If a capsule fails and releases material which could damage the reactor fuel or structure bycorrosion or other means, an evaluation shall be made to determine the need for corrective action. Inspection and any corrective action taken shall be reviewed by the UCD/MNRC Directoror his designated alternate and determined to be satisfactory before operation of the reactor isresumed.Basis -a. This specification is intended to reduce the likelihood that airborne radioactivity in the reactorroom or the unrestricted area will result in exceeding the applicable dose limits in 10 CFR Part20.b. These assumptions are used to evaluate the potential airborne radioactivity release due to anexperiment failure (SAR Chapter 13, Section 13.2.6.2).

c. Normal operation of the reactor with damaged reactor fuel or structural damage is prohibited to avoid release of fission products.

Potential damage to reactor fuel or structure shall bebrought to the attention of the UCD/MNRC Director or his designated alternate for review toassure safe operation of the reactor (SAR Chapter 13, Section 13.2.6.2). 4.0 Surveillance Requirements_ General. The surveillance frequencies denoted herein are based on continuing operation of the reactor.Surveillance activities scheduled to occur during an operating cycle which can not be performed with the reactoroperating may be deferred to the end of the operating cycle. If the reactor is not operated for a reasonable time,a reactor system or measuring channel surveillance requirement may be waived during the associated timeperiod. Prior to reactor system or measuring channel operation, the surveillance shall be performed for eachreactor system or measuring channel for which surveillance was waived. A reactor system or measuring channel shall not be considered operable until it is successfully tested.4.1 Reactor Core P~arameters 4.1.1 Steady State Operation Apolicability -This specification applies to the surveillance requirement for the power levelmonitoring channels. O biecltive -The objective is to verify that the maximum power level of the reactor does notexceed the authorized limit.Specifi~cat~ion -An annual channel calibration shall be made of the power level monitoring

• channel.

If a channel is removed,

replaced, or unscheduled maintenance is performed, or a* significant cilfange in core configuration occurs, a channel calibration shall be required.

Discovery of noncompliance with this specification shall limit reactor operations to that requiredto perform the surveillance. Bss-The annual pwrlevel channel calibration will assure that the indicated reactor power.......level is correct.4.1.2 Shutdown.Margin and Exc(;ess Reactivity ................................................................ ~ Aplcbiiy These specifications apply to the surveillance requirements for reactivity control of....... /the reactor core.betve- The objective is to measure and verify the reactivity worth, performance, andoperability of those systems affecting the reactivity of the reactor..Specifica~tion -a. The total reactivity worth of each control rod and the shutdown margin shall be determined annually or following any significant change in core or control rod configuration. The shutdownmargin shall be verified by meeting the requirements of Section 3.1.3(a).

b. The core excess reactivity shall be verified:

(1) Prior to each startup operation and,(2) Following any change in core loading or configuration. Discovery of noncompliance with Technical Specifications 4.1.2.a-b shall limit reactor operations to that requiredto perform the surveillance. Basis -a. The reactivity worth of the control rods is measured to assure that the required shutdownmargin is available and to provide an accurate means for determining the excess reactivity ofthe core. Past experience with similar reactors gives assurance that measurements of th~e,..; control rod reactivity worth on an annual basis is adequate to assure that there are no significant changes in the shutdown margin, provided no core loading or configuration changes have beenmade.b. Determining the core excess reactivity prior to each reactor startup shall assure that Technical Specifications 3.1 .3.b shall be met, and that the critical rod positions do not changeunexpectedly. 4.2 Reactor Control and Sa~fet,! Systems4.2.1 Control RodsApplicability -This specification applies to the surveillance of the control rods.Objective -The objective is to inspect the physical condition of the reactor control rods andestablish the operable condition of the rods.Spoecification -Control rod worths shall be determined annually or after physical removal or anysignificant change in core or control rod configuration.

a. Each control rod shall be inspected at annual intervals by visual observation of the fueledsections and absorber sections plus examination of the linkages and drives.b. The scram time .of each control rod shall be measured semiannually.

I.Discovery of noncompliance with Technical Specifications 4.2.1 .a-b shall limit reactor operations to that requiredto perform the surveillance. ) ~s(ehia pcfctos4.2.1 .b)-Annual determination of control rod worths o". .....measurements after any physical removal or significant change in core loading or control rod/ -z1.

• 0configuration provides information about changes in reactor total reactivity and individual rodworths. The frequency of inspection for the control rods shall provide periodic verification of the* ...%condition of the control rod assemblies.

The specification intervals for scram time assureoperable performance of the control rods.4.2.2 Reactor Instrumentation A~oDlicability -These specifications apply to the surveillance requirements for measurements, tests, calibration and acceptability of the reactor instrumentation. Obie.ctive -The objective is to ensure that the power level instrumentation and the fueltemperature instrumentation are operable. Specification -a. The reactor power level safety channels shall have the following: (1) A channel test monthly or after any maintenance which could affect their operation. (2) A channel check during each day's operation. (3) A channel calibration annually. -b. The Linear Power Channel shall have the following: (1) A channel test monthly or after any maintenance which could affect the operation. (2) A channel check during each day's operation. (3) A channel calibration annually.

c. The Log Power Channel shall have the following:

(1) A channel test monthly or after any maintenance which could affect its operation. (2) A channel check during each day's operation. (3) A channel calibration annually.

d. The fuel temperature measuring channels shall have the following:

(1) A channel test monthly or after any maintenance which could affect operation. (2) A channel check during each day's operation. (3) A channel calibration annually.

e. The Pulse Energy Integrating Channel shall have the following:

-. (1) A channel test prior to pulsing operations. (2) A channel calibration annually.

• Discovery of noncompliance with Technical Specifications 4.2.2.a-e shall limit reactor operation to that required)to perform the surveillance.

Basis -

• a. A daily channel check and monthly test, plus the annual calibration, will assure that thereactor power level safety channels operate properly.

\J b. A channel test monthly of the reactor power level multi-range channel will assure that thechannel is operable and responds correctly. The channel check will assure that the reactorpower level multi-range linear channel is operable on a daily basis. The channel calibration annually of the multi-range linear channel will assure that the reactor power will be accurately measured so the authorized power levels are not exceeded.

c. A channel test monthly will assure that the reactor power level wide range log channel isoperable and responds correctly.

A channel check of the reactor power level wide range logchannel will assure that the channel is operable on a daily basis. A channel calibration willassure that the channel will indicate properly at the corresponding power levels.d. A channel test monthly and check during each day's operation, plus the annual calibration, willassure that the fuel temperature measuring channels operate properly.

e. A channel test prior to pulsing plus the annual channel calibration will assure the pulse energyintegrating channel operates properly.

4.2.3 Rea~ctor Scrams and Interlocks. .Applicability -These specifications apply to the surveillance requirements for measurements, test, calibration, and acceptability of the reactor scrams and interlocks. Obie~ctiyve -The objective is to ensure that the reactor scrams and interlocks are operable.Specification -a. Console Manual Scram. A channel test shall be performed monthly.b. Reactor Room Manual Scram. A channel test shall be performed monthly.c. Radiography Bay Manual Scrams. A channel test shall be performed monthly.d. Reactor Power Level Safety Scram. channel test shall be performed monthly.e. High-Voltage-Power Supply Scrams. A channel test shall be performed monthly.f. Fuel Temperature Scram. A channel test shall be performed monthly.g. Watchdog Circuits Scrams. A channel test shall be performed monthly.h. External Scrams. A channel test shall be performed monthly.i. The One Kilowatt Pulse interlock. A channel test shall be performed monthly.j. Low Source Level Rod Withdrawal prohibit Interlock. A channel test shall be performed monthly.k. Control Rdd Withdrawal Interlocks. A channel test shall be performed monthly.I. Magnet Power Key Switch Scram. A channel test shall be performed monthly.!Discovery of noncompliance with Specifications 4.2.3.a-I shall limit r'eactor operation to that required to perform..... the surveillance. Basis- .--.A channel test monthly of the Console Manual Scram will assure that the scram is operable. ~b. A channel test monthly of the Reactor Room Manual Scram will assure that the scram isoperable.

c. A channel test monthly of the Radiography Bay Manual Scrams will assure that the scramsare operable.

d. A channel test monthly of the Reactor Power Level Safety Scrams will assure that the scramsare operable.

e. A channel test monthly of the Loss-of-High-Voltage Scram will assure that the high voltagepower supplies are operable and respond correctly.

f. A channel test monthly of the Fuel Temperature Scrams will assure that the scrams areoperable.

g. A channel test monthly of the Watchdog Circuits Scrams will assure that the scram circuits areoperable.

h. A channel test monthly of the External Scrams will assure that the scrams are operable andrespond correctly.

i. A channel test monthly will assure that the One Kilowatt Pulse Interlock works properly.

j. A channel test monthly of the Low Source Level Rod Withdrawal Prohibit Interlock will assure*that the interlock is operable.

k. A channel test monthly of the Control Rod Withdrawal Interlock will assure that the interlock isoperable.

I. A channel test monthly of the Magnet Current Key Switch will assure that thescram isoperable. 4.2.4 Reactor Fuel Element~s This specification applies to the surveillance requirements for the fuel elements. Obiective -The objective is to verify the continuing integrity of the fuel element cladding. Sp~ecification -To assure the measurement limitations in Section 3.2.4 are met, the following shall be done:a. The lead elements (i.e., all elements adjacent to the transient rod, with the exception ofinstrumented fuel elements), and all elements adjacent to the central irradiation facility shall beinspected annually.

b. Instrumented fuel elements shall be inspected if any of the elements adjacent to it fail to pass* the visual and/or physical measurement requirements of Section 3.2.4. Discovery of* noncompliantee with Technical Specification 4.2.4 shall limit operations to that required toperform the surveillance.

Basis (Technical Specifications 4,2,4.a-b) -The above specifications assure that the lead fuelelements shall be inspected regularly adteintegrity.o h edfe lmnssalbmaintained. These are the fuel elements with the highest power density as analyzed in the SARChapter 4, Section 4.5.5.6. The instrumented fuel element is excluded to reduce the risk ofdamage to the thermocouples. 4.3 Reactor Coolant SystemsAoolicability -This specification applies to the surveillance requirements for the reactor water measuring systems and the emergency core cooling system..Ojective-The objective is to assure that the reactor tank water temperature monitoring system, thetank water level alarm, the water conductivity cells and the emergency core cooling system are alloperable. .Specification -a. The reactor tank core inlet temperature monitor shall have the following: (1) A channel check during each day's operation. (2) A channel test quarterly. (3) A channel calibration annually.

b. The reactor tank low water level monitoring system shall have the following:

-.-(1) A channel test monthly..

c. The purification inlet conductivity monitors shalt have the following:

(1) A channel check during each day's operation. ) (2) A channel test monthly.(3) A channel calibration semiannually.

d. The Emergency Core Cooling System shall have the following:

(1) A channel check prior to operation. (2) A channel calibration semiannually. Discovery of noncompliance with Technical Specifications 4.3.a-c shall limit operations to that required toperform the surveillance. Noncompliance with Technical Specification 4.3.d shall limit operations to less than 1.5MW.Basis -a. A channel test quarterly assures the water temperature monitoring system responds correctly to aninput signal. A channel check during each day's operation assures the channel is operable. A channelcalibration annually assures the monitoring system reads properly.

b. A channel test monthly assures that the low water level monitoring system responds correctly to aninput signal.c. A channel test monthly assures that the purification inlet conductivity monitors respond correctly to an.input signal. A channel check during each day's operation assures that the channel is operable.

A)channel calibration semiannually assures the conductivity monitoring system reads properly. " d. A channel check prior to operation assures that the emergency core cooling system is operable forpower levels above 1.5 MW. A channel calibration semiannually assures that the Emergency CoreCooling System performs as required for power levels above 1.5 MW.*"L'1 4.4 React lor Room Exha~ust System\ Applicability -This specification applies to the surveillance requirements for the reactor roomexhaust system.Objective -The objective is to assure that the reactor room exhaust system is operating properly. _Soecification -The reactor room exhaust system shall have a channel check during each day'soperation. Discovery of noncompliance with this specification shall limit operations to that required to perform thesurveillance. Basis -A channel check during each day's operation of the reactor room exhaust system shallverify that the exhaust system is maintaining a negative pressure in the reactor room relative tothe surrounding facility areas.4.5 This section intentionally left blank4.6 This section intentionally left blank. .4.7 ,Rea~ctor Radiation Monitoring SystemsApplicability -This specification applies to the surveillance requirements for the reactor radiation monitoring systems.Obiective -The objective is to assure that the radiation monitoring equipment is operating ) properly. Specification -a. The facility stack monitor shall have the following: (1) A channel check during each day's operation. (2) A channel test weekly.(3) A channel calibration annually.

b. The reactor room radiation monitor shall have the following:

(1) A channel check during each day's operation. (2) A channel test weekly.(3) A channel calibration annually.

c. The purification system radiation monitor shall have the following:

(1) A channel check during each day's operation:

• ) (2) A channel test weekly.,...j,(3)

A channel calibration annually.

d. The reactor room Continuous Air Monitor (CAM) shall have the following:

(1) A channel check during each day's operation. (2) A channel test weekly.(3) A channel calibration annually. Discovery of noncompliance with Technical Specifications 4.7.a-d shall limit operations to that required toperform the surveillance. Basis -a. A channel check of the facility stack monitor system during each day's operation will assurethe monitor is operable. A channel test weekly will assure that the system responds correctly toa known source. A channel calibration annually will assure that the monitor reads correctly.

b. A channel check of the reactor room radiation monitor during each day's operation will assurethat the monitor is operable.

A channel test weekly will ensure that the system responds to aknown source. A channel calibration of the monitor annually will assure that the monitor readscorrectly.

c. A channel check of the purification system radiation monitor during each day's operation assures that the monitor is operable.

A channel test weekly will ensure that the systemresponds to a known source. A channel calibration of the monitor annually will assure that themonitor reads correctly.

d. A channel check of the reactor room Continuous Air Monitor (CAM) during each day'soperation will assure that the CAM is operable.

A channel test weekly will assure that the CAMresponds correctly to a known source. A channel calibration annually will assure that the CAMreads correctly.

4.8 Experiments

Aoolicabilitv -This specification applies to the surveillance requirements for experiments installed in any UCD/MNRC reactor~experiment facility. Objective.- The objective is to prevent the conduct of experiments or irradiations which maydamage the reactor or release excessive amounts of radioactive materials as a result of failure.Soecification -a. A new experiment shall not be installed in any UCD/MNRC reactor experiment facility until awritten safety analysis has been performed and reviewed by the UCD/MNRC

Director, or hisdesignee, to establish compliance with the Limitations on Experiments, (Technical Specifications Section 3.8) and 10 CFR 50.59.b. All experiments performed at the UCDIMNRC shall meet the conditions of an approvedFacility Use Authorization.

Facility Use Authorizations and experiments carried out under theseauthorizations shall be reviewed and approved in accordance with the Utilization of the (UCD)McClellan N~zlear Radiation Center Research Reactor Facility Document (MNRC-0027-DOC). An experimenlt classified as an approved experiment shall not be placed in any UCDIMNRCexperiment facility until it has been reviewed for compliance with the approved experiment andFacility Use Authorization by the Reactor Manager and the Health Physics Manager, or theirdesignated alternates.

c. The reactivity worth of any experiment installed in the pneumatic transfer tube, or in any otherUCD/MNRC reactor in-core or in-tank irradiation facility shall be estimated or measured, as.................

~............... Iff_/,/ appropriate, before reactor operation with said experiment. Whenever a measurement is done it---.' shall be done at ambient conditions.

d. Experiments shall be identified and a log or other record maintained while experiments are inany UCD/MNRC reactor experiment facility.

Basis -a & b. Experience at most TRIGA reactor facilities verifies the importance of reactor staff andsafety committee reviews of proposed experiments.

c. Measurement of the reactivity worth of an experiment, or estimation of the reactivity worthbased on previous or similar measurements, shall verify that the experiment is within authorized reactivity limits.d. Maintaining a log of experiments while in UCD/MNRC reactor experiment facilities willfacilitate maintaining surveillance over such experiments.

5.0 Design Feat~ures 5.1 Site and Facility Description!.- 5.1.1 SiteApplicability -This specification applies to the UCD/MNRC site location and specific facilitydesign features. i" Objective. -The objective is to specify those features related to the Safety Analysis evaluation. Specification -a. The site location is situated approximately 8 miles (13 kin) north-by-northeast of downtownSacramento, California. The site of the UCD/MNRC facility is about 3000 ft. (0.6 mi or 0.9 kin)west of Watt Avenue, and 4500 ft. (0.9 mi or 1.4 kin) south of E Street.b. The restricted area is that area inside the fence surrounding the reactor building. Theunrestricted area is that area outside the fence surrounding the reactor building.

c. The TRIGA reactor is located in Building 258, Room 201 of the UCD/MNRC.

This buildinghas been designed with special safety features.

d. The core is below ground level in a water filled tank and surrounded by a concrete shield.Basis -a. Information on the surrounding population, the hydrology, seismology, and cliimatography ofthe site has been presented in Chapter 2 of the Safety Analysis Report.b. The restricted area is controlled by the UCD/MNRC Director.

c. The room b nclosi ng the reactor has been designed with systems related to the safe operation of the facility.

• .}/d. The below grade core design is to negate the consequences of an aircraft hitting the reactor..... building.

This accident was analyzed in Chapter 13 of the Safety Analysis Report, and found tobe beyond a credible accident scenario. 5.1.2 FcltExas,Applicability -This specification applies to the facility which houses the reactor..Obiective -The objective is to assure that provisions are made to restrict the amount ofradioactivity released into the environment, or during a Loss of Coolant Accident, the system isto assure proper removal of heat from the reactor room.Specification -a. The UCD/MNRC reactor facility shall be equipped with a system designed to filter andexhaust air from the UCD/MNRC facility. The system shall have an exhaust stack height of aminimum of 18.2rn (60 feet) above ground level.b. Manually activated shutdown controls for the exhaust system shall be located in the reactorcontrol room.Basis -The UCD/MNRC facility exhaust system is designed such that the reactor room shall bemaintained at a negative pressure with respect to the surrounding areas. The free air volumewithin the UCD/MNRC facility is confined to the facility when there is a shutdown of the exhaustsystem. Controls for startup, filtering, and normal operation of the exhaust system are located inthe reactor control room. Proper handling of airborne radioactive materials (in emergency situations) can be conducted from the reactor control room with a minimum of exposure tooperating personnel. 5.2 Reactor Coolant SystemApplicability -This specification applies to the reactor coolant system..Obiective -The objective is to assure that adequate water is available for cooling and shielding duringnormal reactor operation or during a Loss of Coolant Accident. Specification -a. During normal reactor operation the reactor core shall be cooled by a natural convection flow ofwater.b. The reactor tank water level alarm shall activate if the water level in the reactor tank drops below adepth of 23 ft.c. For operations at 1.5 MW or higher during a Loss of Coolant Accident the reactor core shall be cooledfor a minimum of 3.7 hours at 20 gpm by a source of water from the Emergency Core Cooling System.Basis -a. The SAR Chapter 4, Section 4.6, Table 4-19, shows that fuel temperature limit of 930 °C will not beexceeded under natural convection flow conditions.

b. A reactor tank water low level alarm sounds when the water level drops significantly.

This alarm* annunciates in the reactor control room and at a 24 hour monitored location so that appropriate corrective action can' be taken to restore water for cooling and shielding.

c. The SAR Chapter 13, Section 13.2, analyzes the requirements for cooling of the reactor fuel andi shows that the fuel safety limit is not exceeded under Loss of Coolant Accident conditions during this........

,/water cooling.5.3 Reactor Core and Fuel 5.3.1 RatrCrAorolicalbility -This specification applies to the configuration of the fuel.Obiective -The objective is to assure that provisions are made to restrict the arrangement of fuelelements so as to provide assurance that excessive power densities will not be produced. Soecification -For operation at 0.5 MW or greater, the reactor core shall be an arrangement of96 or more fuel elements to include fuel followed control rods. Below 0.5 MW there is nominimum required number of fuel elements. In a mixed 20/20, 30/20 and 8.5/20 fuel loading(SAR Chapter 4, Section 4.5.5.6): Mix J Core and Other Var'iations (1) No fuel shall be loaded into Hex Rings A or B,(2) A fuel followed control rod located in an 8.5 wt% environment shall contain 8.5 wt% fuel.,2.0E Core and Other Variations (1) No fuel shall be loaded into Hex Rings A or B. .(2) Fuel followed control rods may contain either 8.5 wt% or 20 wt% fuel.(3) Variations to the 20E core having 20 wt% fuel in Hex Ring C requires the 20 wt% fuel to beloaded into corner positions only, and graphite dummy elements in the flat positions. The*performance of fuel temperature measurements shall apply to variations to the as-analyzed 20Ecore configurations. 308 CoQre and Other Variations (1) No fuel shall be loaded into Hex Rings A or B.(2) The only fuel types allowed are 20120 and 30/20.(3) 20/20 fuel may be used in any position in Hex Rings C through G.(4) 30/20 fuel may be used in any position in Hex Rings 0 through G but not in Hex Ring C.(5) An analysis of any irradiation facility installed in the central cavity of this core shall be donebefore it is used with this core.Basis -In order to meet the power density requirements discussed in the SAR Chapter 4,Section 4.5.5.6, no less than 96 fuel elements including fuel followed control rods and the aboveloading restrictions will be allowed in an operational 0.5 MW or greater core. Specifications forthe 202 core and for the 30B core allow for variations of the as-analyze~1 core with the condition that temperature limits are being maintained (SAR Chapter 4, Section 4.5.5.6 and ArgonneNational Laboratory Report ANLIED 97-54).5.3.2 Reactor FuelIApplicability -These specifications apply to the fuel elements used in the reactor core.~Obiective -The objective is to assure that the fuel elements are of such design and fabricated in/ such a manner as to permit their use with a high degree of reliability with respect to theirphysical and nuclear characteristics. Sp~ecification.- The individual unirradiated TRIGA fuel elements shall have the following

• characteristics:

i a. Uranium content: 8.5, 20 or 30 wt % uranium enriched nominally to less than 20% U-235.b. Hydrogen to zirconium atom ratio (in the ZrH ,): 1.60 to 1.70 (1.65+/- 0.05).c. Cladding: stainless steel, nominal 0.5mm (0.020 inch) thick.Basis -a. The design basis of a TRIGA core loaded with TRIGA fuel demonstrates that limitingoperation to 2.3 megawatts steady state or to a 36 megawatt-sec pulse assures an amplemargin of safety between the maximum temperature generated in the fuel and the safety limit forfuel temperature. The fuel temperatures are not expected to exceed 630 °C during any condition of normal operation.

b. Analysis shows that the stress in a TRIGA fuel element, H/Zr ratios between 1.6 and 1.7, isequal to the clad yield strength when both fuel and cladding temperature are at the safety limit930°C. Since the fuel temperatures are not expected to exceed 630 0C during any condition ofnormal operation, there is a margin between the fuel element clad stress and its ultimatestrength.

c. Safety margins in the fuel element design and fabrication allow for normal mill tolerances ofpurchased materials.

5.3.3 Contr~ol .Rods and Control Rod Drives../ A oolicabilitv -This specification applies to the control rods and control rod drives used in thereactor core.Obiective -The objective is to assure the control rods and control rod drives are of such adesign as to permit their use with a high degree of reliability with respect to their physical,

nuclear, and mechanical characteristics.

.Specification -a. All control rods shall have scram capability and contain a neutron poison such as stainless steel, borated graphite, B 4C powder, or boron and its compounds in solid form. The shim andregulating rods shall have fuel followers sealed in stainless steel. The transient rod shall have anair filled follower and be sealed in an aluminum tube.b. The control rod drives shall be the standard GA rack and pinion type with an electromagnet and armature attached.

a. The neutron poison requirements for the control rods are satisfied by using stainless steel,neutron absorbing borated graphite, B 4C powder, or boron and its compounds.

These materials shall be contained in a suitable clad material such as stainless steel or aluminum to assuremechanical st~bility during movement and to isolate the neutron poison from the tank waterenvironment. Scram capabilities are provided for rapid insertion of the control rods.* " \b. The standard GA TRIGA control rod drive meets the requirements for driving the control rods, ...... jat the proper speeds, and the electromagnet and armature provide the requirements for" rapidinsertion capability. These drives have been tested and proven in many TRIGA reactors.

• A5.4 Fissionable Materiall Storaae...." A Dp1icabilitraco coe. This specification applies to the storage of reactor fuel at a time when it is nat in thereacto core* Objective

-The objective is to assure that the fuel which is being stored will not become critical and willnot reach an unsafe temperature.

• -a. All fuel elements not in the reactor core shall be stored (wet or dry) in a geometrical array where thekef is less than 0.9 for all conditions of moderation.

b. Irradiated fuel elements shall be stored in an array which shall permit sufficient natural convection cooling by water or air such that the fuel element temperature shall not exceed the safety limit.Bss- The limits imposed by Technical Specifications 5.4.a and 5.4.b assure safe storage.*6.0 Administrative Controls6.1 Organization.

The Vice Chancellor for Research shall be the licensee for the UCDIMNRC. TheUCD/MNRC facility shall be under the direct control of the UCD/MNRC Director.: The UCD/MNRCDirector shall be accountable to the Vice Chancellor for Research for the safe operation andmaintenance of the fac~ility.

• 6.1.1 'Struciture.

The management for operation of the UCD/MNRC facility shall consist of theorganizational structure as shown in Figure 6.1.6.1.2 R~esoonsibilities. The UCD/MNRC Director shall be accountable to the Vice Chancellor for* ) Research for the safe operation and maintenance of the facility. The UCD/MNRC

Director, or*his designated alternate, shall review and approve all experiments and experiment procedures prior to their use in the reactor.

Individuals irn the management organization (e.g., Operations

Manager, Reactor Supervisor, Health Physics Supervisor) shall be responsible for implementing UCD/MNRC policies and for operation of the facility, and shall be responsible for safeguarding the public and facility personnel from undue radiation exposures and for adhering to theoperating license and technical specifications.

The Operations Manager shall report directly tothe UCDIMNRC

Director, and shall immediately report all items involving safety and licensing tothe Director for a final decision.

The Reactor Supervisor and Health Physics Supervisor reportdirectly to the Operations Manager.6.1.3 Stffn6.1.3.1 The minim~irn staffing when the reactor is not shutdown shall be:a. A reactor operator in the control room; .b. A second person in the facility area who can perform prescribed instructions; c..A senior reactor operator readily available. The available senior reactoroperator should be within thirty (30) minutes of the facility and reachable bytelephone, and;d. A senior reactor operator shall be present whenever a reactor startup isperformed, fuel is being moved, or experiments are being placed in the reactor* tank.... .-...6.1.3.2 A list of reactor facility personnel by name and telephone number shall beavailable to the reactor operator In the control room. ,The list shall include:* 31

a. Management personnel.

., ~b. Health Physics personnel.

• " c. Reactor Operations personnel.

6.1.4 Selectio~n and.Training of Personnel. The selection, training and requalification of. operations

• personnel shall meet or exceed the requirements of the American National Standard for Selection and* Training of Personnel for Research Reactors (ANS 15:4). Qualification and requalification of licensedoperators shall be subject to an approved Nuclear Regulatory Commission (NRC) program.6.2 Review. Audit. Recommendation anld ApprovalGenleral Policy. Nuclear facilities shall be designed, constructed,

operated, and maintained insuch a manner that facility personnel, the general public, and both university and non-university property are not exposed to undue risk. These activities shall be conducted in accordance withapplicable regulatory requirements.

The UCD Vice Chancellor for Research shall institute the above stated policy as the facility--license holder. The Nuclear Safety Committee (NSC) has been chartered to assist in meeting.this responsibility by providing timely, objective, and independent

reviews, audits,* recommendations and approvals on matters affecting nuclear safety. The following describes the composition and conduct of the NSC.6.2.1 N.SC Comoosition and Qualifications,.

The UCD Vice Chancellor for Research shallappoint the Chairperson of the NSC. The NSC Chairperson shall appoint a Nuclear SafetyCommittee (NSC) of at least seven (7) members knowledgeable in fields which relate to nuclearsafety. The NSC shall evaluate and review nuclear safety associated with the operation and usei' of the UCD/MNRC, 6.2.2 NSC Ch~arte~r and Rules. The NSC shall conduct its review and audit (i~nspection) functions in accordance with a written charter. This charter shall include provisions for:a. Meeting frequency (The committee shall meet at least semiarnnually).

b. Voting rules.c. Quorums (For the full committee, a quorum will be at least seven (7) members).

d. A committee review function and an audit/inspection function.

e. Use of subcommittees.

f. Review, approval and dissemination of meeting minutes.6.2;3 Review Eunctio.

The responsibilities of the NSC, or a designated subcommittee thereof,shall include but are not limited to the following:

a. Review approved experiments utilizing UCD/MNRC nuclear facilities.

b. Review and approve all proposed changes to the facility,

license, the Technical Specifications and the Safety Analysis Report, and any new or changed Facility Use Authorizations andproposed Class I modifications, prior to implementing (Class I) modifications, prior to takingaction under the preceding documents or prior to forwarding any of these documents to thei' Nuclear Regulatory Commission for approval.

Review and determine whether a proposed change, test, or experiment would constitute anunreviewed safety question or req.uire a change to the license, to a Facility Use Authorization, or32 to the Technical Specifications. This determination may be in the form of verifying a decisionalready made by the UCD/MNRC Director. .d. Review reactor operations and operational maintenance, Class I modification

records, arid' ! the health physics program and associated records for all UCD/MNRC nuclear facilities.

e. Review the periodic updates of the Emergency Plan and Physical Security Plan forUCD/MNRC nuclear facilities.

f, Review and update t~he NSC Charter every two (2) years.g. Review abnormal performance of facility equipment and operating anomalies.

h. Review all reportable occurrences and all written reports .of such occurrence~s prior to.forwarding the final written report to the Nuclear Regulatory Commission.

i. Review the NSC annual audit/inspection of the UCD/MNRC nuclear facilities and any otherinspectionsof these facilitieg conducted by other agencies.

-*-6.2.4 Audit/Inspection Function. The NSC or a subcommittee

thereof, shall audit/inspect reactoroperations and health physics annually.

The annual audit/inspection shall include, but not be* limited to the following:

a. Inspection of the reactor operations and operational maintenance, Class I modification

• records, and the health physics program and associated

records, including the ALARA program,for all UCD/MNRC nuclear facilities.

b. Inspection of the physical facilities at the UCD/MNRC.

c. Examination of reportable events at the UCDIMNRC.

d. Determination of the adequacy of UCDIMNRC standard operating procedures.

e. Assessment of the effectiveness of the training and retraining progra.ms at the UCD/MNRC.

f. Determination of the conformance of operations at thle UCD/MNRC with the facility's licenseand Technical Specifications, and applicable regulations.

g. Assessment of the results of actions taken to correct deficiencies that have occurred innuclear safety related equipment, structures,

systems, or methods of operations.

h. Inspection of the currently ac~tive Facility Use Authorizations and associated experiments.

i. Inspection of future plans for facility modifications or facility utilization..

j. Assessment of operating abnormalities.

k. Determination of the status of previous NSC recommendations.

• 6.3 Radiation Safety. The Health Physics Supervisor shall be responsible for implementation of the*. ~UCD/MNRC Radiation Safety Program.

T~he program should use the guidelines of the .American National Standard for Radiation Protection at Research Reactor Facilities (ANSI/ANS 15.11). TheHealth Physics Supervisor shall report to the Operations Manager.* 6.4 Procedures. Written .procedures shall be prepared and approved prior to initiating any of the// activities listed in this section. The procedures shall be approved by the UCD/MNRC Director. A periodic.......review of procedures shall be performed and documented in a timely manner by the UCD/MNRC staff toassure that procedures are current, Procedures shall be adequate to assure the safe operation of the33

reactor, but shall not preclude the use of independent judgment and action should the situation require...... *Procedures shall be in effect for the following items:6.4.1 Reactor Operations Procedures

a. Startup, operation, and shutdown of the reactor.b. Fuel loading, unloading, and movement within the reactor.c. Control rod removal or replacement.

d. Routine maintenance of the control rod drives and reactor safety and interlock systems orother routine maintenance that could have an effect on reactor safety.e. Testing and calibration of reactor instrumentation and controls, control rods and control roddrives.f. Administrative controls for operations, maintenance, and conduct of irradiations andexperiments that could affect reactor safety or core reactivity.

g. Implementation of required plans such as emergency and security plans..h. Actions to be taken to correct potential malfunctions of systems, including responses toalarms and abnormal reactivity changes.6.4.2 HealthPhysics Procedures

a. Testing and calibration of area radiation

monitors, facility air monitors, laboratory radiation detection

systems, and portable radiation monitoring instrumentation.

b. Working in laboratories and other areas where radioactive materials are used.c. Facility radiation monitoring program including routine and special surveys, personnel monitoring, monitoring and handling of radioactive waste, and sampling and analysis of solidand liquid waste and gaseous effluents released from the facility.

The program shall include amanagement commitment to maintain exposures and releases as low as reasonably achievable (ALARA).d. Monitoring radioactivity in the environment surrounding the facility.

e. Administrative guidelines for the facility radiation protection program to include personnel orientation and training.

f. Receipt of radioactive materials at the facility, and unrestricted release of materials and itemsfrom the facility which may contain induced radioactivity or radioactive contamination..

g. Leak testing of sealed sources containing

.radioactive materials.

h. Special nuclear material accountability.

i. Transportation of radioactive materials.

Changes to the above procedures shall require approval of the UCD/MNRC Director. All such changes shall be... documented. 6.5 Experiment Review and Aporoval,. Experiments having similar characteristics are grouped togetherfor review and approval under specific Facility Use Authorizations. All specific experiments to be.3Lf performed under the provisions of an approved Facility Use Authorization shall be approved by theUCD/MNRC

Director, or his designated alternate.

a. Approved experiments shall be carried out in accordarnce with established and approved procedures.

b. Substantive change to a previously approved experiment shall require the same review and approvalas a new experiment.

c. Minor changes to an experiment that do not significantly alter the experiment may be approved by asenior reactor operator.

6.6 Req~uired Acti~ons. 6.6.1 Action to be taken in. case. of..a safety limit violation. In the event of a safety limit violation (fuel temperature), the following action shall be taken:a. The reactor shall be shut down and reactor operation shall not be resumed until authorized by the NRC.b. The safety limit violation shall be promptly reported to the UCD/MNRC Director.

c. The safety limit violation shall be reported to the chairman of the NSC and to the NRC by theUCD/MNRC Director.

d. A safety limit violation report shall be prepared.

The report shall describe the following: (1) Applicable circumstances leading to the violation, including when known, the causeand contributing factors.(2) Effect of the violation upon reactor facility components,

systems, or structures, andon the health and safety of personnel and the public.(3) Corrective action to be taken to prevent reoccurrence.

e. The safety limit violation report shall be reviewed by the NSC and then be submitted to theNRC when authorization is sought to resume operation of the reactor.6.6.2 Actions to be taken for reoortable occurrences.

In the event of reportable occurrences. the following actions shall be taken:a. Reactor conditions shall be returned to normal or the reactor shall be shut down. If it isnecessary to shut down the reactor to correct the occurrence, operations shall not be resumedunless authorized by the UCD/MNRC Director or his designated alternate.

b. The occurrence shall be reported to the UCDIMNRC Director or the designated alternate.

The UCD/MNRC Director shall report the occurrence to the NRC as required by these Technical Specifications or any applicable regulations.

c. Reportable occurrences should be verbally reported to the Chairman of the NSC and theNRC Operations Center within 24 hours of the occurrence.

A written preliminary report shall besent to the Attn: Document Control Desk, 1 White Flint North, 11555 Rockville Pike,Rockville MD 20852, within 14 days of the occurrence. A final written report shall be sent to theabove address within 30 days of the occurrence.

d. Reportable occurrences should be reviewed by the NSC prior to forwarding any writtenreport to the Vice Chancellorof the Office of Research or to the Nuclear Regulatory Commission.

6.7 Re..rt. All written reports shall be sent within the prescribed interval to the NRC, Attn: DocumentControl Desk, 1 White Flint North, 11555 Rockville Pike, Rockville MD 20852.6.7.10Operating

Repeorts, An annual report covering the activities of the reactor facility duringthe previous calendar year shall be submitted within six months following the end of eachcalendar year. Each annual report shall include the following~information:

a. A brief summary of operating experiences including experiments performed, changes infacility design, performance characteristics and operating procedures related to reactor safetyoccurring during the reporting period, and results of surveillance tests and inspections.

b. A tabulation showing the energy generated by the reactor (in megawatt hours), hours thereactor was critical, and the cumulative total energy output since initial criticality.

c. The number of emergency shutdowns and inadvertent scrams, including reasons for theshutdowns or scrams.d. Discussion of the major maintenance operations performed during the period, including theeffect, if any, on the safety of the operation of the reactor and the reasons for any corrective maintenance required.

-e. A brief description, including a summary of the safety evaluations, of changes in the facility orin procedures, and of tests and experiments carried out pursuant to Section 50.59 of 10 CFRPart 50.f. A summary of the nature and amount of radioactive effluents released or discharged to the/ ~environment beyond the effective control of the licensee as measured* at or prior to the point of,' such release or discharge, including the following: (1) Liquid Effluents (summarized on a monthly basis).(a) Liquid radioactivity discharged during the reporting period tabluated asfollows:1 The total estimated quantity of radioactivity released (in curies).2 An estimation of the specific activity for each detectable radionuclide present if the specific activity of the released material after dilution isgreater than 1x10"7 microcuries/ml. 3 A summary of the total release in curies of each radionuclide determined in 2_ above for the reporting period based on representative isotopic analysis. 4 An estimated average concentration of the released radioactive material at the point of release for each month in which a releaseoccurs, in terms of microcuries/mi and the fraction of the applicable concentration limit in 10 CFR 20.t: (b) The total volume (in gallons) of effluent water (including diluent) releasedduring each period of liquid effluent release..\}' (2) Airborne Effluents (summarized on a monthly basis):/'(a) Airborne radioactivity discharged during the reporting period (in curies)tabulated as follows: 1 The total estimated quantity of radioactivity released (in curies)determined by an appropriate sampling and counting method.2 The total estimated quantity (in curies) of Argon-41 released duringthe reporting period based on data from an appropriate monitoring system.3 The estimated maximum annual average concentrationof Argon-41in the unrestricted area (in microcuries/mi), the estimated corresponding annual radiation dose at this location (in millirem), and the fraction of theapplicable 10 CFR 20 limits for these values.4 The total estimated quantity of radioactivity in particulate form withhalf lives greater than eight days (in curies) released during thereporting period as determined by an appropriate particulate monitoring system.5l The average concentration of radioactive particulates with half-lives greater than eight days released (in microcuries/mI) during the reporting period. -(3) Solid Waste (summarized on an annual basis)(a) The total amount of solid waste packaged (in cubic feet).(b) The total activity in solid waste (in curies).(c) The dates of shipment and disposition (if shipped off site).g. An annual summary of the radiation exposure received by facility operations personnel, byfacility users. and-by visitors in terms of the average radiation exposure per individual and thegreatest exposure per individual in each group.h. An annual summary of the radiation levels and levels of contamination observed duringroutine surveys performed at the facility in terms of average and highest levels.i. An annual summary of any environmental surveys performed outside the facility. 6.7.2. Special Reports. Special reports are used to report unplanned events as well as plannedadministrative changes. The following classifications shall be used to determine the appropriate reporting schedule:

a. A report within 24 hours by telephone or similar conveyance to the NRC operations center of:(1) Any accidental release of radioactivity into unrestricted areas above applicable unrestricted area concentration limits, whether or not the release resulted in property* damage, personal injury, or exposure;

• (2) Any violation of a safety limit;(3) Operation with a limiting safety system setting less conservative than specified inSection 2.0, Limiting Safety System Settings;

.7 (4) Operation in violation of a Limiting Condition for Operation; (5) Failure of a required reactor or experiment safety system component which couldrender the system incapable of performing its intended safety function unless the failureis discovered during maintenance tests or a period of reactor shutdown; (6) Any unanticipated or uncontrolled change in reactivity greater than $1.00;(7) An observed inadequacy in the implementation of either administrative or procedural

controls, such that the inadequacy could have caused the existence or development of acondition which could have resulted in operation of the reactor outside the specified safety limits; and(8) A measurable release of fission products from a fuel element.b. A report within 14 days in writing to the NRC, Document Control Desk, Washington DC.(1) Those events reported as required by Technical Specifications 6.7.2.a.1 through6.7.2.a.8.

(2) The written report (and. to the extent possible, the preliminary telephone report orreport by similar conveyance) shall describe,

analyze, and evaluate safety implications.

and outline the corrective measures taken or planned to prevent re~occurrence of theevent.c. A report within 30 days in writing to the NRC, Document Control Desk, Washington DC.(1) Any significant variation of measured values from a corresponding predicted orpreviously measured value of safety-connected operating characteristics occurring during operation of the reactor;(2) Any significant change in the transient or accident analysis as described in theSafety Analysis Report (SAR);(3) A personnel change involving the positions of UCD/MNRC Director or UCO ViceChancellor for Research; and(4) Any observed inadequacies in the implementation of administrative or procedural controls such that the inadequacy causes or could have caused the existence ordevelopment of an unsafe condition with regard to reactor operations. 8.8 Records. Records may be in the form of logs, data sheets, or other suitable forms. The requiredinformation may be contained in single or multiple

records, or a combination thereof.

Records and logsshall be prepared for the following items and retained for a period of at least five years for items a.through f., and indefinitely for items g. through k. (Note: Annual reports, to the extent they contain all ofthe required information, may be used as records for items g. through j.)a. Normal reactor operation.

b. Principal maintenance activities.

c. Those events reported as required by Technical Specifications 6.7.1 and 6.7.2.d. Equipment and component surveillance activities required by the Technical Specifications.

,'e. Experiments performed with the reactor.f. Airborne and liquid radioactive effluents released to the environments and solid radioactive wasteshipped off site.

g. Offsite environmental monitoring surveys.i. h. Fuel inventories and transfers.

i. Facility radiation and contamination surveys.1. Radiation exposures for all personnel.

k. Updated, corrected, and as-built drawings of the facility.

.1J"39 a ~a ....-.a.'I.Formal Licensing ChannelAdministrative Reporting Channel............. Communications ChannelUCD/MNRC ORGANIZATION FOR LICENSING AND OPERATION FIGURE 6.1}}