ML15348A386
| ML15348A386 | |
| Person / Time | |
|---|---|
| Site: | University of California-Davis |
| Issue date: | 10/29/2015 |
| From: | Steingass W McClellan Nuclear Research Center |
| To: | Linh Tran Division of Policy and Rulemaking |
| References | |
| Download: ML15348A386 (234) | |
Text
{{#Wiki_filter:UCDAVIS 5335 PRICE AVENUE MNRC McCLELLAN NUCLEAR RESEARCH CENTER BUILDING 258 McCLELLAN, CA 95652 PHONE: (916) 614-6200 FAX: (916) 614-6250 WEB: http://mn rc.ucdavis.edu U.S. Nuclear Regulatory Commission October 29, 2015 Attn: Linh N. Tran, Senior Project Manager, NRR Mail Stop: 012 D20 One White Flint North 11555 Rockville Pike Rockville, MD 20852 RE: 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.
Dear Ms. Tran,
In response to your letter dated June 3, 2015, we are submitting the requested documentation per said letter 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 October29, 2015. Assoca'e Director of Operations Reactor Supervisor McClellan Nuclear Research Center University of California-Davis Facility Operating License No. R-130. C: B. Klein, UCD/MNRC
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- - *.*oNUCLEAR REGULATORY COMMISSION
- WASHINGTON, 0.0. zt&o5S5O1Q FACILITY OPERATING LICENSE DOCKET NO. 50-607 DEPARTMENT OF THE AIR FORCE AT McCLELLAN AIR FORCE BASE
- , License No. R-130
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A." The application for license, filed by the Department of the Air Force at McClellan Air Force Base, on October 23, 1 996, as supplemented, complies with the standards and requirements of the Atomic Energy Act of 1!954, as =amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. Construction. of the facility was completed in substantial conformity with the provision's 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;
- 0. There is reasonable assurance (i) that the activities authorized by this license can be conducted ,without endangering the health and safety of the 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 the activities authorized by this operating license in accordance with the regulations of t/he Commission; ..
F. The licensee is a Federal agency and will use the facility for defense programs and research. The licensee, in accordance with 10 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 the requirements o~f 10 CFR Part 140 of the Commission's regulations;
G. The issuance of this license will not be inimical to the common defense and 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 the Commissioh's regulations and all applicable requirements have been satisfied; and I.The receipt:, possession, and use of the byproduct and special nuclear materials as authorized by this license will be in accordance with the Commissioa 's regulations in 10 CFR Parts 30 and 70, including Sections 30.33, 70.23, and 70.31.
- 2. Facility Operating License No. R-130 is hereby issued to the Department of the 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 the Department 'of the Air Force at McClellan Air Force Base (the licensee). The facility is located on the licensee's site at McClellan Air Force Ease and 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, the Commission ~hereby licenses the Department of the Air Force at McClellan Air Force Base: (1) Pursuant to Section 104c of the Act and 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," to possess, use, and operate the facility at the designated location at 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 to 21 .0: kilograms of contained uranium-235 enriched to less than 20 percent in the isotope uranium-235 in the form of reactor fuel;:up to 4 grams of contained uranium-235 of any enrichment in the form of fission chambers; up to 16.1 kilograms of contained uranium-235 enriched to less than 20 pecenR[[tin he isotope uranium-235 in the form of plates; and to possess, but not separate, such special nuclear material as may be produced by the operation of the facility.
3 (3) Pursuant to the Act and 1 0 CFR Part 30, "Rules of General Applicability to Domestic Licensing of Byproduct Material," to receive, possess, and use a 4-curie sealed americium-beryllium neutron source in connection with operation of the facility; a 55-millicurie sealed cesium-1 37 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 in reactor experiments, such byproduct material as may be produced by the 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, and orders of the Commission now or hereafter in effect and to the additional conditions specified below: (1) Maximum Power Level The licensee is authorized to operate the facility at steady-state power levels not in excess of 2300 kilowatts (thermal) and in the pulse mode with reactivity insertions not to exceed $ 1.75 (1.23 %Ak/k). (2) Technical Specifications The Technical Specifications contained in Appendix A are hereby incorporated in the license. The licensee shall operate the !facility in accordance with the Technical Specifications. (3) Physical Security Plan The 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 the authority of 10 CFR 50.90 and 1 0 CFR 50.54(p). The approved plan;i which is exempt from public disclosure pursuant to the provisions of 10 CFR 2.790, is entitled "Physical Security Plan for the McClellan Nuclear Radiation Center (MNRC) TRIGA Reacitor Facility," Revision 3, dated August 1996.
4
- 0. This license is effective as of the date of issuance and shall expire twenty (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
~NUCLEAR REGULATORY COMMISSION
*" ~UNITED STATES "
*. WASHINGTON 1 D.C. 20555-0001
* "'r*°*
'December 9, 1998 Brigadier General Michael P. Wiedemer, Commander Sacramento Air Logistics Center SM-ALC/TI- 1 5335 Price Avenue McClellan 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 License No. 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 to your 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 airborne radioactivity in the reactor room or the unrestricted area exceeding the applicable dose limits in 10 CFR Part 20. A copy of the safety evaluation supporting Amendment No. 1 is also enclosed. Sincerely, Warren J. Eresian, Project Manager Non-Power Reactors and Decommissioning
*Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No. 50-607
Enclosures:
- 1. Amendment No. 1
- 2. Safety Evaluation cc w/enclosures:
See next page
McClellan AFB TRIGA REACTORDcktN.067 cc: Dr. Wade J. Richards SM-ALCITI-1 5335 Price Avenue, Bldg. 258 McClellan AFB, California 95652-2504 Col. Robert Capell HQ AFMC/SGC 4225 Logistics Avenue, Suite 23 Wright-Patterson AFB, Ohio 45433-5762 Lt. Col. Catherine Zeringue HQ AFSC/SEW 9570 Avenue G, Building 24499 Kirtland AFB, New Mexico 87117-5670 Test, Research, and Training Reactor Newsletter 202 Nuclear Sciences Center University of Florida Gainesville, Florida 32611
~NUCLEAR REGULATORY COMMISSION
* .'
- O*,i.UNITED STATES WASHINGTON, D.C. 20888-0001 DEPARTMENT OF THE AIR FORCE AT McCLELLAN AIR FORCE BASE DOCKET NO. 50-607 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 1 License No. R-130
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that A. The application for an amendment to Facility Operating License No. R-1 30 filed by the Department of the Air Force at McClellan Air Force Base (the licensee) on November 18, 1998, conforms to the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the regulations of the Commission as stated in Chapter I of Title 10 of the Code of FederalRegulations (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 the Commission; D. The issuance of this amendment will not be inimical to the common defense and security or t*o 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; and F. 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.
2
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the enclosure to this license amendment, and paragraph 2.C.(ii) of Facility Operating 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 through Amendment No. 1, are hereby incorporated in the license. The licensee shall operate 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, Director Non-Power Reactors and Decommissioning Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation
Enclosure:
Appendix A, Technical Specifications Changes Date of Issuance:
ENCLOSURE TO LICENSE AMENDMENT NO. 1 FACILITY OPERATING LICENSE NO? R-1 30 DOCKET NO. 50-607 Replace the following pages of Appendix A, "Technical Specifications," with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. Remove Insert 24 24 25 25
.j° .;* ,
This specification restricting thec.quantity is intended to prevent damage to vital equipment by 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 TNT equivalent or less in any radiography bay external to the reactor tank will not result in damage to the reactor controls or the reactor tank. Safety Analyses have been performed (SAR Chapter 13, Section 13.2.6.2) which show that up to six (6) lbs of TNT equivalent can be safely irradiated in any radiography bay. Therefore, the three (3) lb limit gives a safety margin of two (2).
3.8.3 Failure and Malfunctions* Applicability. This specification applies to experiments installed in the reactor, 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 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 or aerosols into the reactor building or into the unrestricted area, the quantity and type of material in the experiment shall be limited such that the airborne radioactivity in the reactor room or the unrestricted area will not result in exceeding the applicable dose limits in 10 CFR Part 20, assuming 100% of the gases or aerosols escape.
- b. In calculations pursualtt 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 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 vapors formed by boiling this material can escape only through an undistributed column of water above the core, at least 10% of these vapors 'can escape. 24
- c. If a capsule fails and releases material which could damage the reactor fuel or structure by corrosion or other means, an evaluation shall be made to.
determine the need for corrective action. Insipection and any corrective action taken shall be reviewed by the Facility Director or his designated 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 reactor room or the unrestricted area will result in exceeding the applicable dose limits in 10 CFR Part 20.
- b. These assumptions are used to evaluate the potential airborne radioactivity release due to an experiment 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 must be brought to the attention of the Facility Director or his designated alternate for review to assure safe operation of the reactor (SAR Chapter 13, 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 end of the operating cycle. If the reactor is not operated for a reasonable .time, a r'eactor 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 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 level monitoring channels. Objective. The objective is to verify that the maximum power level of the reactor does not exceed the authorized limit. 25
~NUCLEAR REGULATORY COMMISSION
,* ' *o~i'-UNITED
- STATES WASHINGTON," O.C. 20865-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 1 TO FACILITY OPERATING LICENSE NO. R-130 DEPARTMENT OF THE AIR FORCE AT McCLELLAN AIR FORCE BASE DOCKET NO..POO-607
1.0 INTRODUCTION
By letter dated November 18, 1 998, the Department of the Air Force at McClellan Air Force 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 the unlikely 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 and its experimental facilities. Specifications.
- 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 or aerosols into the reactor building or into the unrestricted area, the quantity and type of material to be irradiated shall be limited such that the airborne concentration of radioactivity shall not exceed the applicable limits of 10 CFR Part 20 (at the operations boundary), assuming 100% of the gases or aerosols escape.
"h.
O *° °" 0 2 Bases.
- a. This specification is intended to reduce the likelihood that airborne radioactivity in excess of the limits of 10 CFR Part 20 shall be released into the reactor building or to the unrestricted area (SAR Section 13.2.6.2).
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, 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 or aerosols into the reactor building or into the unrestricted area, the quantity and type of material in the experiment shall be limited such that the airborne radioactivity in the reactor room or the unrestricted area will not result in exceeding the applicable dose limits in 10 CFR Part 20, assuming 100% of the gases or aerosols escape.
Bases.
- a. This specification is intended to reduce the likelihood that airborne radioactivity in the reactor room or the unrestricted area will result in exceeding the applicable dose limits on 10 CFR 20.
The licensee has proposed clarifying the TS by basing the TS on dose instead of concentrations of radioactive material. The purpose of this TS is to limit doses to members of 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 reactor confinement and subsequently to the environment. Doses to members of the reactor staff and members of the public from accidents at research reactors are limited to the doses given 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 the annual 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 50 reins 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 2000 working 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 if inhaled continually over the course of a year would produce a total effective dose equivalent of 0.05 rem. The licensee's proposed wording would be based on dose limits directly. The licensee is concerned that the TS as currently written could be interpreted to limit releases to the instantaneous concentration of airborne radioactive material in the reactor room and unrestricted areas. This would ignore the time integral aspects of the concentration 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 20 while 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 a TS based on dose would also include consideration of radiation shine from a cloud of radioactive material. This proposed change to the TSs is acceptable to the staff because the 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 located within 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 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 criteria for categorical exclusioni set forth in 10 CFR 51.22(c)(9). Pursuant to 10CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared 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 or consequences of accidents previously evaluated,* or create the possibility of a new or different kind of accident from 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 is reasonable assurance that the health and safety of the public will not be endangered by the proposed activities; and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or the health and safety of the public. Principal Contributor: Warren J. Eresian Date: December 9, 1998
*. *.--*,*UNITED STATES o NUCLEAR REGULATORY COMMISSION Z r~oWASHINGTON, D.C. 2055.5-0001 FACILI;TY OPERATING LICENSE DOCKET NO. 50-607 DEPARTMENT OF THE AIR FO.RCE.AT McCLELLAN AIR FORCE BASE License No. R-1 30
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A. The application for license, filed by the Department of the Air Force at McClellan Air Force Base, on October 23, 1996, as supplemented, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR 'Chapter I; B. Construction of the facility was completed in substantial conformity 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 be conducted without' endangering the health and safety of the 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 the activities authorized by this operating license in accordance with the regulations of the Commission;... F. The licensee is a Federal agency and will use the facility for defense programs and research. The licensee, in accordance with 10 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 the requirements of 10 CFR Part 140 of the Commission's regulations;
2 G. The issuance of this license will not be inimical to the common defense and 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 the Commission's regulations and all applicable requirements have been satisfied; and I.In*- c]p, 1 °os s e-ssio n,-a +n *ni !le*r
- 2. Facility Operating License No. R-130 is hereby issued to the Department of the 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 the Department of the Air Force at McClellan Air Force Base (the licensee). The facility is located on the licensee's site at McClellan Air Force Base and 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, the Commission hereby licenses the Department of the Air Force at McClellan Air Force Base: (1) Pursuant to Section 104c of the Act and 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," to possess, use, and operate the facility at the designated location at 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 to 21 .0 kilograms of contained uranium-235 enriched to less than 20 percent in the isotope uranium-235 in the form of reactor fuel; up to 4 grams of contained uranium-235 of any enrichment in the form of fission chambers; up 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 be produced by the operation of the facility.
3
**_*:*ea 4-curie sealed americium-beryllium neutron source in connection with operation of the facility; a 55-millicurie sealed cesium-i137 source for instrument calibrations; small instrument calibration and check 'sources of less than 0.1 millicurie eah.* *.
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, and orders of the Commission now or hereafter in effect and to.the additional conditions specified below: (1) Maximum Power Level The licensee is authorized to operate the facility at steady-state power levels not in excess of 2300 kilowatts (thermal) and in the pulse mode with reactivity insertions not to exceed $1.75. (1.23 %Ak/k). (2) Technical Specifications The Technical Specifications contained in Appendix A are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. (3) Physical Security Plan The licensee shall fully implement and maintain in effectel.- provisions of the Commission-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 approved plan, which is exempt from public disclosure pursuant to the provisions of 1 0 CFR 2.790, is entitled "Physical Security Plan for the McClellan Nuclear Radiation Center (MNRC) TRIGA Reactor Facility," Revision 3, dated August 1 996.
S .. * .:.
.4
- 0. This license is effective as of the date of issuance and shall expire twenty (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, 1999 Brigadier General Michael P. Wiedemer, Commander Sacramento Air Logistics Center SM-ALC/TI- 1 5335 Price Avenue McClellan AFB, California 95652-2504
SUBJECT:
ISSUANCE OF AMENDMENT NO. 2 TO AMENDED FACILITY OPERATING LICENSENO. R-130 - DEPARTMENT OF THE AIR FORCE AT McCLELLAN AIR FORCE BASE (TAC NO. MA3477)
Dear General Wiedemer:
The Commission has issued enclosed Amendment No. 2 to Facility Operating License No. 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 Safety Analysis Report (SAR) to support expanded experimental facilities in response to your submittal of January 11, 1999. The amendment provides for the installation of an Argon-41 Production Facility and a Central Irradiation Facility. The installation of the Argon-41 Production Facility does not require any change or expansion of the TSs since an experiment failure will not result in airborne radioactivity in the reactor room or the unrestricted area exceeding the applicable dose limits already prescribed. The installation of the Central Irradiation Facility requires 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 the Central 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 Manager Non-Power Reactors and Decommissioning Project Directorate Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No. 50-607
Enclosures:
- 1. Amendment No. 2
- 2. Safety Evaluation cc w/enclosures: See next page
McClellan AFB TRIGA REACTORDoktN.5-0 cc" Dr. Wade J. Richards SM-ALC/TI-1 5335 Price Avenue, Bldg. 258 McClellan AFB, California 95652-2504 Lt. Col. Marcia Thornton HQ AFSC/SEW" 9570 Avenue G., Bldg. 24499 Kirtland AFB, New Mexico 87117-5670 Col. Robert Capell HQ AFMC/SGC 4225 Logistics Avenue, Suite 23 Wright-Patterson AFB, Ohio 45433-5762
0* 0
*. UNITED STATES
.NucLEAR REGULATORY COMMISSIoN WHNToND.C. 208-o000 DEPARTMENT OF THE AIR FORCE AT Mc.CLELLAN AIR FORCE BASE DOCKET NO. 50-607 AMENDMENT TO FACILITY OPERATING LICENSE AmendmentNo. 2 License No. R-1 30 1.. The U.S. Nuclear Regulatory Commission (the Commission) has found that A. The application for an amendment to Facility Operating License No. R-1 30 filed by the Department of the Air Force at McClellan Air Force Base (the licensee) on January 11, 1999, conforms to the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the regulations of the Commission as stated in Chapter I of Title 10 of the Code of FederalRegulations (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 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 the Commission; D. The issuance of this amendment will not be inimical to the common defense and security 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; and F. 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.
- 2. Accordingly, the license is amended by changes to the Safety Analysis Report and Technical Specifications as indicated in the enclosure to this license amendment, and paragraph 2.C.(ii) of Facility 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 through Amendment No. 2, are hereby incorporated in the license. The licensee shall 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, Director Non-Power Reactors and Decommissioning Project Directorate Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation
Enclosure:
- Appendix A, Technical Specifications
*and Safety Analysis Report Changes Date of Issuance: March 1, 1999
.... 0.
ENCLOSURE TO LICENSE AMENDMENT NO. 2 FACILITY OPERATING LICENSE NO. R-130 DOCKET NO. 50-607 A. Replace the following page of Appendix A, "Technical Specifications," with the enclosed page. The revised page is identified by amendment number and contains vertical lines indicating the areas of change. Remove Insert 22 22 B. Insert the following sections into the Safety Analysis Report.
- 1. Add new Section 10.5.3
- 2. Add new Section 11.1.1.1.6
- 3. Append to Section 13.2.6.2
- 4. Add new Appendix A to Chapter 13
- 5. Add new Appendix .B to Chapter 13
- 6. Change Section 10.4.1
- 7. Add new Section 10.4.1.4
- 8. Append to Section 1 3.2.6.2
- 9. Add Reference 13.19 to ChaPter 13
- S unrestricted 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 Central Irradiation 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 be less than one (1) dollar (0.7% AK/K).
- b. The absolute reactivity worth of any single secured experiment shall be less than the maximum allowed pulse ($1.75) (1.23% AK/K).
- c. The absolute total reactivity worth of experiments installed in the reactor and in-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, or removal 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 the sample 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, Chapter 13.2.2.2.1 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 preclude pulsing if the experiment's fixturing should fail, since the resulting reactivity insertion would not cause prompt criticality if less than one dollar. Given that the reactor will not pulse inadvertently, the additional increase in transient power and temperature will be slow enough so that the fuel temperature scram will be effective.
- b. The absolute worst event which may be considered in conjunction with a single secured experiment is its sudden accidental or unplanned removal while the reactor is operating. This would result in a reactivity increase less than a pulse of $1.92, analyzed in SAR Chapter 13, Section 13.2.2.2.1.
- c. It is conservatively assumed that simultaneous removal of all experiments in the reactor and in-tank experiment facilities at any given time shall not exceed the maximum reactivity insertion limit. SAR, Chapter .13, Section 13.2.2.2.1 shows that an insertion 22 Amendmient No. 2
ADDITION TO MCCLELLAN NUCLEAR RADIATION CENTER SAFETY ANALYSIS REPORT - ARGON-41 PRODUCTION FACILITY NEW SECTION 10,5.3 10.5.3 Arcqon-41 Production Facility The Argon-41 Production Facility will produce 1-2 curies of 4 1Ar for research and commercial use. The 41Ar will be produced by introducing argon gas into a stainless steel container located in one of the silicon irradiation positions (adjacent to the graphite reflector and external to the reactor core - Figure 10.111A). 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 desired activity 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 valve is opened to pressurize the cooling coils above the liquid nitrogen bath. The Dewar is then raised 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 of 41Ar gas to the sample container. The liquid nitrogen Dewar is lowered, exposing the cooling coils to room temperature. When that portion of the system between the cooling coils and the sample container has reached equilibrium the sample container will be isolated and..removed from the room. The coil is surrounded with a lead shield to minimize the radiation exposure to personnel. A catch tank surrounds the Dewar to contain any liquid nitrogen escaping from'the Dewar or in the unlikely event of a total failure of the Dewar. Over pressure protection of the overall system is provided by several relief valves that vent to an over pressure tank. The over pressure ta~nk is protected by its own relief valve which vents to the reactor room. The tank is located as high as possible in the reactor room. 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 the 41 Ar. The Argon-41 Production Facility consists of several different components, with the major components listed below.
0 COMPONENT MATERIAl DESCRIPTION Irradiation Container 304 stainless The irradiation container is a 1000 ml sample cylinder with a working pressure steel of 1 800 psig and a burst pressure of 6000 psig. It conforms to the "Shipping Container Specifications" from the U.S. Code of Federal Regulations, Title 49 or Bureau of Explosives Tariff No.BOE 6000. Over Pressure 304 stainless The adjustable proportional pressure relief Relief Valves steel valves have a working pressure up to 6000 psig. When upstream pressure overcomes the force exerted by the spring, the poppet opens, allowing flow through the valve. As the upstream pressure increases, flow through the valve increases proportionately. Cracking pressure is only sensitive to inlet pressure and is not affected by outlet pressure. Over Pressure Carbon steel 30 gallon tank. Relief Tank Valves 304 stainless Bellows sealed valves. steel Tubing 304 stainless 1/4-inch and Y/=-inch. steel NEW SECTION 11.1.1.1.6 11.1.1.1.6 Araqon-41 from the Argon-41 Production Facility Ar-41 will be produced by the Ar-41 Production Facility (see Chapter 10) as needed. The Ar-41 that is produced by the Ar-41 Argon Production Facility will be contained in the system so there should be no increase in the Ar-41 levels in the reactor room or the Ar-41 that is released to the unrestricted area. Catastrophic failure of the system will not result in any 10 CFR 20 limit being exceeded and is further discussed in Chapter 13. APPEND TO SECTION 13.2.6.2 The Argon-41 Production Facility (see Chapter 10) can produce argon-41 in excess of the amounts analyzed in Appendix A of the MNRC Safety Analysis Report. However, if the system releases argon-41, the gas will be contained in the reactor room and the existing
0 reactor 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 Stack Continuous Air Monitor will also be used to verify any release outside the MNRC boundary. If the system had a catastrophic failure and 4 curies of argon-41 were released to the volume of the reactor room, the argon-41 concentration in the reactor room would be approximately 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 be restricted. The reactor room ventilation system (as described in Chapter 9) would, be operated in the recirculation mode for approximately one day before the dose rate from argon-41 decays to less than 1 mR/hr. Therefore, the argon-41 Discharge Limit defined in the MNRC Technical Specifications will not be exceeded due to the recirculation mode of the reactor room ventilation system. Other potential accidents include failure of the irradiation container due to overpressurization from the argon gas supply cylinder, since a new argon supply cylinder is typically delivered at 2200 psig and the container is rated for 1800 psig. However, this requires multiple failures and is considered non-credible: a) the operator would have to violate an operational procedure; b) the regulator would have to fail, and c) at the same time the pressure relief valve would have to fail. Also, liquid nitrogen could spill into the reactor tank, causing expansion of the water and expelling a portion of tank water. To prevent this, a catch basin surrounds the Cold Trap, and the liquid nitrogen is supplied through a pipe in the reactor room wall connecting the trap to a supply container in the equipment room. A third accident could result if the pressure relief valve became choked with supersonic flow; however, the flow rates are estimated to be less than sonic (see calculat~ion in Chapter 13, Appendix A). NEW APPENDIX A TO CHAPTER 13 ARGON-41 CONCENTRATION IN REACTOR ROOM GIVEN:
- 1. Reactor room volume =-7.39x10 3 ft 3 tReference 111
- 2. 4 curies Ar-41 in argon production system
- 3. D(y)=, 2 = O.25Evx [Reference 21 Dy== 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
0 X *= concentration of gamma emitting isotope in the cloud (Ci/m 3 ) CALCULATIO )N: X = (4Ci)/[7.39xl10 3 ft 3 )(1 m3/35.314 ft 3 ) = 1 .91!x 0.2 Gi/m 3 D(y)=,2 = 0. 25Eyx
= (0.25)( 1.2936 Mev/dis)(1 .91 xl 0.2 Cl/rn 3 )
= (0.0062 rads/sec)(3600 seclhr)
= 22.24 radslhr D = Doe~x t = -(1/A)In(0D/D)
= -(T112Iln2)ln(D/D0 )
For 0 = 1 mrad/hr t = -(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), edited by Shelein, p. 439
- 3. 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 PROJECT Assume: Perfect Gas Constants: Property Value Units R 208 N-rn/k g-degK k(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 C c =[1 .67(208 N-m/kg-degK)(-200 +273)K( 1 kg-rn/N-sec 2 )I*
= 159.2397 rn/s at -200 degrees C Next, 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 psi V = (1 ft 3 /min)(12 in/ft) 3 (2.54 cm/in) 3 (1 min/60 sec)
= 471.9474 cm 3 /sec Area = r*r 2 = 3.14(0.18in/2)2 = 0.025434 in2 based on 1/4 inch tubing
= 0.16409 cm 2 Velocity = V/Area = 28.7615 rn/sec Mach 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 and therefore 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 REPORT AND TECHNICAL SPECIFICATIONS - CENTRAL IRRADIATION FACILITY CHANGE SECTION 10.4.1 The Central Irradiation Facility, located in the center of the reactor core, may contain either a plug assembly (as described in sections 10.4.1 .1 through 10.4.1 .3 and Figure 10.7) or a moveable sample can system (as described in section 10.4.1.4). All parts are removable from the reactor using underwater tools. NEW SECTION 10.4.1.4 10.4.1 .4 Central Irradiation Facility The 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 the insertion and removal of samples from the central facility through the use of a drive mechanism similar to the control rods. The central thimble is approximately 52 inches in length and 4.22 inches outer diameter with 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. Aluminum shims 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 fuel element locations in the B hex ring 4.25-inch hole. Two captive bolts attach the central thimble to the upper grid plate. These bolts prevent the accidental removal of the facility when removing samples from the central thimble. An 1100 aluminum slug located inside the central thimble is normally positioned in the reactor core. The aluminum slug is 4 inches in diameter and 24.75 inches in length. This voids the water from the central thimble when the sample can is removed from the thimble. An orifice plate is located on the bottom of the central thimble. In the event the aluminum slug releases from the locating holes and falls to the, bottom of the central thimble, the rate 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 3.99 inches and an inside diameter of 3.75 inches. The can could be free flooding or dry, and is used to position samples for irradiation in the reactor core. The positioning of samples can 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 being rotated. The drive mechaauism has the same type of drive motor as the control rod drives except the 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 wiBl treat the insertion and removal of the samples as if they were control rods. The reactor room controls can only be enabled from the reactor console. The normal indicators are as follows:
"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 travel indications, stop driving limits and start/stop rotation of the carrier. APPEND TO SECTION 13.2.6.2 Another potential accident involves the Central Irradiation Facility (see Chapter 10) since it may be considered similar to a control rod. Therefore, consider three potential scenarios for an uncontrolled reactivity insertion analogous to the Uncontrolled Withdrawal Of a Control Rod (see Section 13.2.2.2.2). First, if the material in the sample can were of sufficiently different worth than the aluminum cylinder, the sample can would cause reactivity changes in the same fashion as a control rod, and either operator error or mechanical failure could cause an uncontrolled reactivity insertion. Second, if the aluminum cylinder failed to engage upon the sample can's insertion, a water void would be created in the central facility as the aluminum cylinder descended ahead of the sample can. Similarly, if the aluminum cylinder failed to replace the can upon removal from the central facility a water void would result. All three of the above scenarios can be bounded by the Uncontrolled Withdrawal of a Control Rod analysis (Section 13.2.2.2.2). Specifically, the Central Irradiation Facility must have less reactivity and must drive slower than the control rod analyzed ($3.50 and 42 inches/minute, respectively). To that end, the reactivity of any material in the sample can shall be measured at low power to verify it's worth is not only less than $3.50 but also less than $1.75, the reactivity limit for the Central Irradiation Facility (based on the Technical Specification limit of $1 .75 for the pulsed reactivity insertion). For example, the worth 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.). The worth of an aluminum cylinder vs. void and vs. water has been analyzed by computer simulation (Reference 13.19). The most positive reactivity effect in the computer simulation 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 Case 3 to Case 1 2, where in an accident the sample can not only floods but also the aluminum cylinder drops, resulting in a decrease of about $.1.76. Thus, the worth of the sample can or 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 rod speed analyzed in the MNRC SAR (Section 13.2.2.2.2). Furthermore, in the event of failure of the aluminum cylinder to engage upon installation of the sample can, the base of the Central Thimble is designed (by sizing the hole in the base) to allow the aluminum cylinder to descend at no more than the analyzed 42 inches/minute. Therefore, the accident analysis in Chapter 13 of the MNRC SAR for Uncontrolled Withdrawal of a Control Rod (Section 13.2.2.2.2) is sufficient to bound any accident associated with the Central Irradiation Facility since: a) the material in the sample can shall be measured and verified to be less than $1.75 (half of the analyzed $3.50); b) the drive speed cannot exceed the analyzed 42 inches/minute; and c) the aluminum cylinder cannot fall uncontrolled faster than the analyzed 42 inches/minute. Finally, physical impact on the fuel is considered non-credible since the sample can is always contained in a guide tube or attached to a drive mechanism such that it is unlikely to drop onto the core (see description in Section 10.4.1.4). ADD REFERENCE 13.19 TO CHAPTER 13 13.1 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 Central Irradiation 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 be less 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 Facility is based on the pulsing reactivity insertion limit (section 3.1.2) and on the design of the sample 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, Chapter 13.2.2.2.1 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 preclude pulsing if the experiment's fixturing. should fail, since the resulting reactivity insertion would not cause prompt criticality if less than one dollar. Given that the reactor will not pulse inadvertently, the additional increase in transient power and temperature will be slow enough so that the fuel temperature scram will be effective.
0 9
*'* ,o*.UNITED STATES"
- NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C..20588-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 2 TO AMENDED FACILITY OPERATING LICENSE NO. R-130 DEPARTMENT OF THE AIR FORCE AT McCLELLAN AIR FORCE BASE DOCKET NO. 50-607
1.0 INTRODUCTION
By letter dated January .11, 1999, the Department of the Air Force at McClellan Air Force 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), and changes to the Safety Analysis Report. The amendment provides for the installation of an Argon-41 Production Facility and a Central Irradiation Facility. The installation of the Argon-41 Production Facility does not require any change or expansion of the TSs since an experiment failure will not result in airborne radioactivity in the reactor room or the unrestricted area exceeding the applicable dose limits already prescribed. The installation of the Central Irradiation Facility requires 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 the Central Irradiation Facility to $1 .75, corresponding to the pulse limit specified in TS 3.1 .2.
2.0 EVALUATION Argon-41 Production Facility The licensee has requested amendment of the Safety Analysis Report to provide for the installation of an Argon-41 Production Facility. The purpose of the facility is to produce Argon-41 for research and commercial uses. Argon gas from a pressurized argon bottle is introduced into a stainless steel container located in a position external to the core, but in the 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 is removed cryogenically and admitted to sample containers. Overpressure protection is provided 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 of up to 6000 psig. The overpressure tank relief valve vents to the reactor room. All piping (1/4 and Y/=inch 304 stainless steel) is anchored to prevent pipe whip in the event of pipe failure. The irradiation container has a working pressure of 1 800 psig with a burst pressure of 6000 psig.
After the argon gas has been irradiated, the gas is transferred to the sample containers. A cooling coil which has been evacuated with a vacuum pump is immersed in a liquid nitrogen 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 valved off, 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 the reactor 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 day will result in a dose rate of approximately 1 mrad per hour. The argon-41 discharge limit as defined in the Technical Specifications will not be exceeded. The licensee has considered other potential accidents. These include overpressurization of the irradiation container, spilling liquid nitrogen into the reactor tank, and the choking of a relief valve due to supersonic flow. Overpressurization of the irradiation container requires multiple mechanical failures and operator violation of the procedure governing the use of the production facility. To prevent the spilling of liquid nitrogen into the reactor tank, a catch basin is installed around the liquid nitrogen bath. Finally, the licensee has analyzed the flow through the relief valves and has determined that the flow remains subsonic, thus preventing choking at the valve. Central. Irradiation Facility The licensee has requested amendment of the 'Technical Specifications and Safety Analysis Report to provide for the installation of a Central Irradiation Facility. The facility allows 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 normal control rods, and a reactor operator controls the insertion and removal of samples. Drive speeds are equal to those of the normal control rods. The central thimble is essentially a vertical guide tube which passes through the upper grid plate, the lower grid plate and the safety plate, resting on the tank floor. lA sample can and an aluminum slug move within the central thimble. An aluminum slug normally occupies a position in the reactor core. When the sample can is inserted, the aluminum slug moves downward out of the co)re, and its position in the core is replaced by the sample 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 rod malfunctions are applicable. In particular, the analysiz of an Uncontrolled Withdrawal of a Control 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 the maximum 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 rod worth of $3.50 was used to allow for reasonable variations. In order to bound accidents involving the Central Irradiation Facility, it is required to show that 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 computer simulation (SAR Reference 13.19) of the reactivity changes associated with various scenarios,- including normal operations and accidents. The most limiting case, the flooding of 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 rod withdrawal 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 located within 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 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 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 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 or consequences of accidents previously evaluated, or create the possibility of a new or different kind of accident from 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 is reasonable assurance that the health and safety of the public will not be endangered by the proposed activities; and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or the health and safety of the public. Principal Contributor: Warren J. Eresian Date: MArch 1, 1999
9 9 UNITt* STATES
- 9 9
*
*
,AUCLEAR REGULATORY COMMISSION
- 1~
WASHINGTON, D.C. 2O55*-0OO1 Brigadier Commander General Michael P. Wledemer Vice Chancellor Kevin Smith Office of the Chancellor Sacramento Air Logistics Center University of California, Davis SM-ALC/TI-1 One Shields Avenue 5335 Price Avenue Davis, California 95616-8558 McClellan AFB, California 95652-2504
SUBJECT:
ORDER APPROVING THE TRANSFER OF THE FACILITY OPERATING LICENSE FOR THE McCLELLAN NUCLEAR RADIATION CENTER FROM THE DEPARTMENT OF THE AIR FORCE TO THE REGENTS OF THE UNIVERSITY OF CALIFORNIA. AND APPROVING CONFORMING AMENDMENT (TAC NO. MA3477)
Dear General Wiedemer and Dr. Smith*:
The enclosed Order Is in response to the application dated April 13,.1999, as supplemented on July 19 and August 4-,1999, and January 18 and 27, 2000, requesting approval of the transfer of Operating License No. R-1 30 for the McClellan Nuclear Radiation Center from the Departm~ent of the AIr Force to the Regents of the University of California, and approval of a conforming amendment to reflect the transfer. The enclosed Order provides consent to the proposed 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 for the 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~sy Warreni~J. Ere fan, Project Manager Events Assessment, Generic C~ommunlcations and Non-Power Re~ctom Branch DivIsion of Reguiator*"1mp rovement Programs Office of Nuclear Reactor Regulation Docket No. 50-607
*
Enclosures:
.1. Order
- 2. Amendment No.3
*.3. Safety Evaluation 4, IndemnityAgreement. .
- Senextp ge
McClellan AFB TRIGA REACTOR Docket No, 50-607 cc; Dr. Wade J. Richards SM-ALC/TI- 1 6335 Price Avenue, Bldg. 258 McClellan AFB, CA 95652-2504 Col. Robert Capell HQ AFMC/SGC 4225 Logistics Avenue, Suite 23 Wright-Patterson AFB, OH 45433-5762 Lt, Cot. Catherine Zeringue HQ AFSC/SEW 9570 Avenue G, Building 24499 Kircland AFB, New Mexico 871 17-5670 Test. Research, and Training Reactor Newsletter 202 Nuclear Sciences Center University of Florida Gainesville, FL 32611
7590-01 -P UNITED STATES OF AME*RICA 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 LICENSE AND 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 .Codeof_Federal Regufation~s (10 CFR Part 50). The facility is located on McClellan Air Force Base In Sacramento, 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 proposed transfer of Operating License No, R-1 30 from the USAF to the University of California. The University of Calliornia at Davis (UCD), part of the University of California, was proposed to be the actual operator of the facility. The application was supplemented by submittals dated July 19 and August 4, 1999, and January 18 and 27, 2000, The initial application and the supplements 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 sole operator 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 the operating license and replace them with references to the UCD, as appropriate, as well as to make other miscellaneous administrative changes to the operating license to ref lect the transfer. Under 10 CFR 50.80, no license for a production or utilization facility, or any right thereunder, 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 the application and other information before the Commission, the NRC staff has determined that the University of California Is qualified to hold the license, and that the transfer of the license to the 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 for the proposed license amendment complies with the standards and requirements of the Atomic Energy Act of 1954, as amended, and the Commission's rules and regulations set forth in 10 CFR Chapter 1; the facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission: there Is.reasonable assurance that the activities authorized by the proposed license amendment can be conducted without endangering the health and safetyof the public and that such activities will be conducted in compliance with theCommission's regulations: the issuance of the proposed license amendment will not be inimical to the common defense and security or to the health and safety of the public; and the issuance of the proposed amendment will be in accordance with 10 CFR
T r-P ." =* 1af1 1 NU.SS5r0 r P.5/)
* *14 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 herein to 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 shall become null and void, provided, however, on written application arnd for good cause shown, 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, to conform 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, Director Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation 4.
'*"*' '* %*UN1TED STATES WASHINGT"ON, D.C. 20555-0001 DEPARTMENT O T.HE AIR FORCF AT MCCLELLAN. AIR FoRCE BASE DOCKET NO. 50-607 AMENDMENT TO AMENDED FACILITY OPERATING LICENSE Amendment No. 3 License No. R-130 1.The U.$. Nuclear Regulatory Commission (the Commission) has tound that A. The application for an amendment to Amended Facility Operating License No. R-130 filed by tile Department of the Air Force at McClellan Air Force Base and the Regents of the University of California on April 13, 1999, as supplemented on July 19 and August 4, 1999, and January 18 and 27, 2000, conmpiies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the regulations of the Commission as stated In Chapter I of Title 10 of the Code of Federal 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 the Commission; D. The issuance of this amendment will not be~inlrmicalto the common defense and security or to the health and safety of the public;and E. This issuance of this amendment is in accordance with the regulations of the Commission as stated in 10 CFR Part 51, and all applicable requirements have been satisfied.
- 2. Accordingly, the license is amendedas indicated in the attachment to thisilcense amendment, ENCLOSURE 2
N.955 FEB.*1.006
- 09M P.7/1.4
- 3. This license amendment is effective as of the date of issuance, FOR THE NUCLEAR REGULATORY COMMISSION Ledyard B. Marsh, Chief Events Assessment, Generic Communications and Non-Power Reactors Branch Divsion of Regulatory Improvement Programs Office of Nuclear Reactor Regulation
Enclosures:
- 1. Amended Facility License
- 2. Appendix A, Technical Specifications changes Date of Issuance: January 31, 2000 4
rI"l" I'"*L"]II * * '* O°* P.8/r*14
**
- NUCLEAR REGULATORY COMMISSION
* ~WASHINGTON, D.C, 20885,=0001
,* FACILITY OPERATING LICENSE
~DOCKET NO., 50-607
#t'z*,r _REGENTS oF THE UNIVERSITY OF C.ALIFORNI*I A License No. R-130 1.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 18 and 27, 2000 comnpiies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth 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 be conducted without endangering the health and safety of the public and (II)that such activities 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 the Commission; F. The licensee is a Nonprofit Educational institution and will use the facility for educational programs arnd research, and has satief led the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements of the Commission's regulations; G. The issuance of this license will not be inimical to the common defense and security or 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; and S.The receipt, possession, and use of the byproduct and special nuclear materials as authorized by this license will be in accordance with the Commission's regulations in 10 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 Regents of the University of California (the licensee), The facility is located on the McClellan Air Force Base, Sacramento, California, I 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 of Production and Utilization Facilities," to possess, use, and operate the facility at the designated location at McClellan Air Force Base in accordance with the procedures and limitations set forth in this license. (Ii) Pursuant to the Act and 10 CFR Part 70, "Domestic Licensing of Special Nuclear Material,= 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 the format 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 be produced by the operation of the facility. (iii) Pursuant to the Act and 10 CFR Part 30, "Rules of General Applicability to Domestic Licensing of Byproduct Material," to receive, possess, and use a 4-curie sealed americium-beryllium neutron source in connection with operation of the 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 In reactor experiments, such byproduct material as may be produced by the ape ration of the facility.
C. This license shall be deemed to contain and Is su~bj~ect 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, and orders of the Commission now or hereafter in effect and to the additional conditions specified, below: (i) Maximum Po~wer Level The licensee is authorized to operate the facility at steady-state power levels not in excess of 2300 kilowatts (thermal) arid in the pulse mode with reactivity insertions not to exceed $1.75 (1.23 %/0 k/k). Amndent N..h 3
3-3 (ii) Technical S~oecfficatlonis The Technical Specifications, as revised through Amendment No. 3, are hereby. incorporated in the license. The licensee shall operate the facility in accordance with f the Technical Specifications. (lii) Physical Securityv lan The licensee shall fully implement and maintain in effect all provisions of the Commission-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 approved plan, which is exempt from public disclosure pursuant to the provisions of 10 CFR 2.790, 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) years from its date of issuance. FOR THE NUCLEAR REGULATORY COMMISSION Previously signedI by Origina/signed by Samuel J, Collins, Director office of Nuclear Reactor Regulation Date of issuance: August 13, 1998 Amendment No. 3
Q E*NCLOSURE TO LICENSEAMENDMENT NO.3 AMENDED FACILITY OPERATING LI.CENSE NO. R-!30 DOCKET NO; 50-807 Replace the following pages of Appendix A, "T'echnlcal Specificationts,= with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
*Remove 1 *1 39 39 40 40 41 4.
TECHNICAL SPECIFICATIONS FOR THE U.C. DAVIS MCCLELLAN NUCLEAR RADIATION CENTER (MNRC) General The McClellan Nuclear Radiation Center (MNRC) reactor is operated by the University of California, Davis, CA. The MNRC research reactor Is a TRIGA type reactor. The MNRC provides state-of-the-art neutron radiography capabilities. In addition, the MNR~C provides a i
- wide range of irradiation servic~es far both research and industrial needs. The reactor operates at a nominal steady start power level up to and including 2 MW. The MNRC reactor is also capable 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 ,AsLow 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 the Nuclear Regulatory Commission (e.g., senior reactor operator or reactor operator) to carry out the duties and responsibilities associated with the position requiring the license. 1.2.1 Senior.ReactorQOerator. An individual who is licensed to direct the activities of reactor operators and to manipulate the controls of the facility. 1.,2.2 Reactor Onerator. An individual who is licensed to manipulate the controls of the facility and perform reactor-related maintenance. 1.3 Ch.* A channel is the combination of sensor, line amplifier, processor, and output 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 the channel for verification that it is operable..,.' 1.3.2 Channel Calibratlaon. A channel calibration is an adjustment of the channel such that its-output corresponds with acceptable accuracy to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm or trip, and shall be deemed to include a channel test. 1.3.3 Channel Ch~eck. A channel check is a qualit~ative verification of acceptable performance by observation of channel behavior. This verification, where possible, shall include comparison of the channel with other independent channels or systems measuring the same variable. 1Amendment No .3.
b ViCECHACELOR OR ESERCHVICE CHANCELLOR FOR ADMINISTRATION ai
. U .D SAFETY K CMMITTEE s I II suPERVISOR SUPERVwSOR ...
* -- - - - - - - - - --.
[OPERATIONS STAFFI HEALTh-PHYSICS STAFF]
~UNIVERSITY MANAGEMENT ORGANIZATION
~Figure 6, !
0: Normal Adminisirative Reporting Channel-Technical Review, Communications and Asisisance .
L'LV* *.,.J.,J *
- J.'-tr .L"*'
* ~ *I*~C*Lff**J~J .1. * .LC.[~I.j J.
7n - - - -L VICE CHA*.NCELLOR OFFzICE OF' RESEARCH I I. I 1. I.
- I
---- I
- TUCIEAR SAFETYL AND UCENSING NUCLEAR SAFETY AND LICENSING REVIEWS, APPROVALS AND RECOMMENDATIONS *-
COMMUNICATION OF* LICENSED ACTIVITIES UC. Davis McCleIlan Nuclear Radiatiog Center Lit'easing Organization Figure 6.2 54 An~Iendment Wo. .3
* ,%.UNITED
%*_*
,*"
- STATES
- S NUCLEAR REGULATORY COMMISSION
*o*_*' *WASHINGTON, 0, .0. S5-0001 Docket No. 50-607 This 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 the Atomic Energy Act of 1954, as amended (hereinafter referred to as the Act).
Article I As used in this agreement,
- 1. Nuclear reactor, byproduct material,, person, source material, specIal nuclear material, and precautionary 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 the radioactive, 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 use of property, arising out of or resulting from the radioactive, toxic,.explosive, or other hazardous properties of i, 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 the radioactive 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 as defined in any such other agreement; or ii.The radioactive material in the course of transportation and also arising out of such properties 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 wihich* is in the course 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, as amended.
- 4. In the course of transportation means Inthe course of transportation within the United States, or in the course of transportation outside the United States and any other nation, and moving from one person licensed by the Commission to another person licensed by the Commission, 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, such transportation is not by predetermination to be interrupted by the removal of the material from the 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 end when the radioactive material is removed from the transporting conveyance for any purpose other 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 of the r'adloactive material to the location if the material is also in the course of transportation from any 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 or precautionary evacuation (including all reasonable additional costs incurred by a State, or a political 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 of employees of persons indemnified who are employed (a) at the location or, if the nuclear Incident 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 loss of 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 In the 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 reactors subject to the license or licenses designated in the Attachment hereto, or (2) which is produced as 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 II
- 1. Any obligations of the licensee under subsection 53e(8.). of the Act to indemnify the United States and the Commission from public liability shall not in the aggregate exceed $250,000 with respe.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 their interests 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 person* or an act of God.
As used herein, conduct of the claimant includes conduct of persons through whom the claimant derives 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 ifsuit is instituted within 3 years from the date on which the claimant first knew, or reasonably could have known, of his injury or damage and the cause thereof.
*The waiver of any such issue or defense shall be effective regardless of whether such issue or defense may otherwise be deemed jurisdictional or relating to an element in the cause of action. The waivers shall be judicially enforceable In accordance with their te.r~ms by the claimant agaInst the person indemnified.
- 3. The waivers set forth in paragraph 2 of this article: (a) Shall not preclude a defense based upon 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 and wrongfully caused by the claimant; (c) Shall not apply to injury to a claimant who is employed at the site of and in connection with the activity where the extraordinary nuclear occurrence takes place if benefits therefor are either payable or required to be provided under any workmen~s compensationi or occupational disease law: Provided, however, That with respect to an extraordinary nuclear occurrence occurring at the facility, a claimant who is employed at the facility Inconnection with the construction of a nuclear reactor with respect to which no operating license has been issued by the Nuclear Regulatory Commission shall not be considered as employed in connection with the activity where the extraordinary nuclear occurrence takes place if:
(1) The claimant is employed exclusively in connection with the construction of a nuclear reactor, 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 transfer of nuclear material at the facility; (d) Shall not apply to anty claim for punitive or exemplary damages. provided, with respect to any claim for wrongful death under any State law which provides for damages only punitive in nature, this exclusion does not apply to the extent that the claimant has sustained actual damages, measured by the pecuniary injuries resulting from such death but not to exceed the maximum 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 claim which is not within the protection afforded under (1)the limit of liability provisions under subsection 170(e) of the Atomic Energy Act of 1954, as amended, and (b) the terms of this agreement. Article Ill
- 1. The Commission undertakes and agrees to Indemnify and hold harmless the licensee and other 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 legally liable for the nuclear incident, the Commission agrees to pay to such person those sums which such 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 with respect to: (a) Property which is located at the location and used in connection with the licensee'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, the transporting 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 to such 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 the aggregate 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 of persons legally liable for the nuclear Incident (other than such property described in the proviso to paragraph 2 of this Article) as in the aggregate exceed whichever of the following is lower: (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 as secondary financial protection, As used in this Article applicable agreements means each agreement entered into by the Commission pursuant to subsection 170(c) or (k)of the Act in which agreement the nuclear incident is defined as a common occurrence.
- 5. The obligations of the Commission under this agreement shall apply only with respect to nuclear incidents occurring during the term of this agreement.
- 6. The obligations of the Commission Uinder this and all other agreements and contracts to which 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 a common occurrence, $560,000,000 less the sum of the amounts of financial protection established under all applicable agreements.
- 7. Ifthe 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 the Commission 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 for damage to property of the licensee, shall not be affected by any failure on the part of the licensee to fulfill Its obligations under this agreement. Bankruptcy or insolvency of tihe licensee or any other person indemnified or of the estate of the licensee or any other person indemnified 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 make indemnity 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 defense of any claim Including such legal costs of the licensee as are approved by the Commission and shall have the right (a) to require the prior approval of the Commission for the settlement or payment of any claim or action asserted against: the Ilicensee or other person indemnified for public liability or damage to property of persons legally liable for the nuclear incident which claim or 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 or other 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 shall furnish 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 or transferred, without the approval of the Commission.
Article V The parties~agree that they will enter into appropriate amendments of this agreement to the extent that such amendments are required pursuant to the Atomic Energy. At of 1954, as amended, or licenses, regulations or orders of the Commission. Article VI The licensee agrees to pay to the Commission such fees as are established l~y the Commission pursuant to regulations or orders of the Commission,. Article Vii The term of this agreement shall commence as of the date and time specified in Item 4 of the Attachment and shall terminate at the time of expiration of that license specified in Item 2 of the Attachment, which is the last to expire; provided that, except as may otherwise be provided in applicable regulations or orders of the Commission, the term of this agreement shall not terminate until all the radioactive material has been removed from the location and transportation of the radioactive material from the location has ended as defined in subparagraph 4(b), Article I, Termination of the term of this agreement shall not affect any obligation of the licensee or any obligation of the Commission under this agreement with respect to any nuclear incident occurring during the term of this agreement. 4g.
FEB 12009NO.957 554p P.6/6 Attachment to Indemnity Agreement No. E-40 Item 1- Licensee University of California, Davis Address-- One Shields Avenue, Davis, California 9561648558 Item 2- License number or numbers R-130 Item 3- Location The reactor is located in the McClellan Nuclear Radiation Center Building on McClellan AFB, located approximately 8 miles northeast of Sacramento, California. Item 4-.. The indemnity agreement designated above, of which this Attachment Is a 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 Branch Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Dated at Rock'ville, MD, the ,* day of * ,2000. _________________By Kevin Smith Vice C~hancelior University of California, Davis
Fz~? 0 UNITED STATES
/
%" NUCLEARWASHINGTON, REGULATORY COMMISSION D.C. 20555-0001
*o*,'*August 9, 2001 Dr. Kevin Smith, Vice Chancellor Office of the Chancellor University of California, Davis One Shields Avenue Davis, 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 Facility 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 May 11, 2001. The amendment reflects the administrative changes to the TSs as a result of the transfer of the license 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 and which are discussed in the enclosed safety evaluation report. Sincerely, Warren J. Eresian, Project Manager Operational Experience and Non-Power Reactors Branch Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No. 50-607
Enclosures:
- 1. Amendment No. 4
- 2. Safety Evaluation cc w/enclosures:
Please see next page
University of California - Davis/McClellan MNRC Docket No. 50-607 co: Dr. Wade J. Richards 5335 Price Avenue, Bldg. 258 McClellan AFB, CA 95652-2504 Test, Research, and Training Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611
- "* w*'oUNITEDNUCLEAR REGULATORY STATES COMMISSION
.* WASHINGTON, D.C. 20555-0001 REGENTS OF THE UNIVERSITY OF CALIFORNIA AT McCLELLAN NUCLEAR RADIATION CENTER DOCKET NO. 50-607 AMENDMENT TO AMENDED FACILITY OPERATING LICENSE Amendment No. 4 License No. R-130
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that A. 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 Nuclear Radiation Center (the licensee) on May 11, 2001, conforms to the standlards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the regulations of the Commission as stated in Chapter I of Title 10 of the Code of FederalRegulations (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 the Commission; D. The issuance of this amendment will not be inimical to the common defense and security 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; and F. 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, license is to the enclosure amended by changes to the Technical Specifications as indicated in the this license amendment, and paragraph 2.C.(ii) of Amended Facility 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 through Amendment No. 4, are hereby incorporated in the license. The licensee shall 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 Warren J. Eresian, Project Manager Operational Experience and Non-Power Reactors Branch Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation
Enclosure:
Appendix A, Technical Specification Changes Date of Issuance: August 9, 2001
S 0 ENCLOSURE TO LICENSE AMENDMENT NO. 4 AMENDED FACILITY OPERATING LICENSE NO. R-130 DOCKET NO. 50-607 Replace the following pages of Appendix A, Technical Specifications, with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. Remove Insert ii ii iii iii iv iv V v vi vi 1 I 2 2 3 3 4 4 6 6 7 7 9 9 13 13 14 14 15 15 16 16 17 17 18 18 19 19 25 25 26 26 27 27 28 28 29 29 30 30 31 31 32 32 33 33 34 34 35 35 36 36 39 39 40 40
UNITED STATES 1"%" NUCLEAR REGULATORY COMMISSION
~WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 4 TO AMENDED FACILITY OPERATING LICENSE NO. R-130 REGENTS OF THE UNIVERSITY OF CALIFORNIA AT McCLELLAN NUCLEAR RADIATION CENTER DOCKET NO. 50-607
1.0 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, to Facility 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 request under oath. The resubmittal contained no new information.) The request provides for the following 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 the Department of the Air Force to the Regents of the University of California. As a result of this transfer, a nUmber of administrative changes simply involving name changes (e.g., changing references from "Responsible Commander" to "Vice Chancellor of the Office of Research" 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 channel calibrations 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 loading termed a 30B core.
- 6. Section 6.0. A request to revise the organization and duties of the Nuclear Safety Committee and to clarify the Committee's review and audit functions to reflect the new licensee.
- 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 of the Air Force to the Regents of the University of California, the TSs must be modified to take account of administrative changes. These changes will occur in a number of places, and consist of the substitution of Department of the Air Force organizational and position titles with corresponding 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 and Operation." 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 are acceptable. 2.2 Section 2.1, Basis b. The previous version of the Technical Specifications addressed the issue of the effect of pulsing on fuel clad integrity and concluded that TRIGA fuel of the type used in the McClellan reactor could 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 present analysis expands the discussion to include more recent measurements of hydrogen pressure resulting from pulses and concludes that the cladding will not rupture if fuel temperatures are never greater than 1200 to 1250 degrees Centigrade, providing the cladding temperature is less than 500 degrees Centigrade. Since the pulse reactivity limit remains at $1.75, the staff concludes 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 by Research Reactor Safety Analysis Services (RRSAS-99-6-1, December 1999) to determine if the conservative maximum core inlet temperature (heat exchanger outlet temperature) as set by the U.S. Air Force in the original design could be raised from 35 degrees Centigrade to 45 degrees Centigrade. The effect of the lower limit is that the reactor power is required to be reduced below the license limit of 2 MW whenever ambient local weather conditions prevent the system 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 state operations, when compared with previous calculations by Argonne National Laboratory, General Atomics published reports, and results from power upgrades at the Sandia Annular Core
0 Research Reactor facility shows that the maximum core inlet temperature can be raised to 45 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. The calculations also show that there is no increase in the maximum fuel temperature or the maximum fuel clad surface temperature, two of the most important parameters which measure fuel integrity. Accordingly, the staff concludes that safety limits will not be reduced and that there 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 the stack monitor system, the reactor room radiation monitor and the reactor room continuous air monitor. These systems are presently required to be calibrated semiannually. The licensee has requested that they be calibrated annually. The requirement for semiannual calibrations stems from the original Department of the Air Force licensing organization, but has no operational safety basis. Research reactors of similar power levels currently licensed by the NRC (National Institute of Standards and Technology, Rhode Island AEC) are permitted to calibrate similar instruments on an annual basis, since there are no operating experience data to suggest that this practice would compromise safety. In addition, the American National Standard ANSI/ANS 15.11 "Radiation Protection at Research Reactor 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 was operating with a mixed core of 8.5/20 and 20/20 fuel loading (referred to as the MixJ core in the original SAR). At that time it was understood that the reactor would eventually transition to a core consisting of 20/20 and 30/20 fuel, termed a 30B core. The 30B core was analyzed in the original SAR and found to be acceptable by the NRC staff in the SER. In addition, the NRC staff had previously approved the generic use of TRIGA fuels with uranium loadings of up to 30 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 the operation and maintenance of the reactor and associated equipment. There are a number of minor changes with respect to titles and some changes with .respect to the composition and duties of the Nuclear Safety Committee (NSC). The review and inspection functions of the NSC have been expanded to provide additional oversight. These expanded functions include review of the Emergency Plan and Physical Security Plan, review and update of the NSC Charter every two years, review of inspections conducted by other agencies, assessment of actions taken to correct deficiencies, inspection of currently active experiments, and inspection of future plans for facility modifications or facility utilization. Since these changes increase oversight of facility operations, the staff concludes that they are acceptable.
0 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 the installation of an Iodine-125 production loop. The purpose of the loop is to produce from ten to twenty 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 the reactor core.
- 2. The Xenon-I 24 is irradiated over an eight to sixteen hour time span and by neutron activation results in the production of Xenon-125. The activated Xenon-I124 gas contains 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 decays with a 17-hour half-life to Iodine-I125. After a few days, most of the Xenon-I125 has decayed 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 to as decay storage 2.
- 5. The Iodine-125 in decay storage I is recovered by washing the tank with a NaOH solution, resulting in a Nal solution which is packaged as a liquid and sent to an off-site user in an appropriate DOT container.
All equipment used in the production loop is located within a primary containment and a secondary containment. The primary containment houses the irradiation chamber, tubing, pneumatically operated valves, transfer vessel, decay storage I and decay storage 2. The secondary containment is placed around the primary containment to the irradiation chamber and allows 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 these containments are within the reactor room, which has a ventilation system with isolation/recirculation capability. There are two other structures within the reactor room which are confinement barriers designed for the safety of personnel working with the production loop. The first is a glove box which contains controls for operation of the Iodine-125 recovery system. The glove box has its own ventilation and filtration system which exhausts into the reactor room ventilation system. The second is a fume hood in which quality assurance of the Iodine-125 is performed. The fume hood also contains its own ventilation and filtration system which exhausts into the reactor room ventilation system. The licensee has analyzed the situation (worst-case) whereby all of the Xenon-I125 from the primary containment leaks into the secondary containment and subsequently leaks into the reactor room at the design leak rate of the secondary containment. Their analysis shows that exposures to personnel in the reactor room would result in a deep dose equivalent (DDE) of 17 millirem 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 at the 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 cladding rupture of one highly irradiated fuel element with no decay followed by instantaneous release of fission products into the air. At the closest distance to the site boundary (10 meters), the maximum 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 the margin of safety with respect to 10 CFR 20 limits and that the installation of the production loop is 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 of Iodine-125 present in the reactor room glove box and the reactor room fume hood. Limiting the amount of Iodine-125 in these areas will reduce the occupational dose and dose to personnel in the unrestricted areas to less than 10 CFR 20 limits if the inventories of Iodine-I125 are totally released 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 located within 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 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 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 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 or consequences of accidents previously evaluated, or create the possibility of a new or different kind of accident from 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 is reasonable 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 or the health and safety of the public. Principal Contributor: Warren J. Eresian Date: August 9, 2001
0I 0 TECHNICAL SPECIFICATIONS FOR THE UNIVERSITY OF CALIFORNIA - DAVIS MCCLELLAN NUCLEAR RADIATION CENTER (UCD/MNRC) DOCUMENT NUMBER: MNRC-0004-DOC-1 1 Rev 11, 12/10/99 Amendment No. 4 i
0 TECHNICAL SPECIFICATIONS APPROVAL These "Technical ReactorSpecifications" have undergone the University for the of California at Davis/McClellan Nuclear Radiation Center following coordination: (UCD/MNRC) Reviewed bY:H*alF* z 'ODte Reviewed by: (*-- k_,. Q- _* Reactor Operations S~pervisor (Date) Approved by: 9 ,U4 UCD/MNI Director (Date) Approved by:________________ Chairman, UCD/MNRC (Date) Nuclear Safety Committee Amendment No. 4 ii
0 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Page 1.0 Definitions...................................................................................................................1 2.0 Safety Limit and Limiting Safety System Setting (LSSS)............................................................... 5 2.1 Safety Limits*...................................................................................................... 5 2.2 Limiting Safety System Selling (LSSS)......................................................................... 6 2.2.1 Fuel Temperature .................................................................................... 6 3.0 Limiting Conditions for Operations (LCO) ............................................................................... 7 3.1 Reactor Core Parameters ....................................................................................... 7 3.1.1 Steady-State Operation.................................................. *............................ 7 3.1.2 Pulse or Square Wave Operation................................................................... 7 3.1.3 Reactivity Limitations................................................................................. 8 3.2 Reactor Control and Safety Systems........................................................................... 8 3.2.1 Control Rods........................................................................................... 8 3.2.2 Reactor Instrumentation ............................................................................. 9 3.2.3 Reactor, Scrams and Interlocks .................................................................... 10 3.2.4 Reactor Fuel Elements ............................................................................. 12 3.3 Reactor Coolant Systems...................................................................................... 13 3.4 Reactor Room Exhaust System ............................................................................... 14
.3.5 Intentionally Left Blank .................................... ..................................................... 14 3.6 Intentionally Left Blank ......................................................................................... 14 3.7 Reactor Radiation Monitoring Systems ....................................................................... 14 3.7.1 Monitoring Systems................................................................................. 14 3.7.2 Effluents - Argon-41 Discharge Limit .............................................................. 16 Amendment No. 4 iii
Pane 3.8 Experiments ..................................................................................................... 16 3.8.1 Reactivity Limits ..................................................................................... 16 3.8.2 Materials Limit ....................................................................................... 17 3.8.3 Failure and Malfunctions ........................................................................... 18 4.0 Surveillance Requirements .................................................................. ........................... 19 4.1 Reactor Core Parameters...................................................................................... 19 4.1.1 Steady State Operation ............................................................................ 19 4.1.2 Shutdown Margin and Excess Reactivity.......................................................... 20 4.2 Reactor Control and Safety Systems ......................................................................... 20 4.2.1 Control Rods......................................................................................... 20 4.2.2 Reactor Instrumentation ............................................................................ 21 4.2.3 Reactor Scrams and Interlocks .................................................................... 22 4.2.4 Reactor Fuel Elements ............................................................................. 23 4.3 Reactor Coolant Systems...................................................................................... 24 4.4 Reactor Room Exhaust Systerm........ ....................................................................... 25 4.5 Intentionally Left Blank ......................................................................................... 25 4.6 Intentionally Left Blank ......................................................................................... 25 4.7 Reactor Radiation Monitoring Systems ....................................................................... 25 4.8 Experiments ..................................................................................................... 26 5.0 Design Features.......................................................................................................... 27 5.1 Site and Facility Description.................................................................................... 27 5.1.1 Site.................................................................................................... 27 5.1.2 Facility Exhaust...................................................................................... 28 5.2 Reactor Coolant System........................................................................................ 28 Amendment No. 4 iv
0 0 Page 5.3 Reactor Core and Fuel ........................................................................................... 29 5.3.1 Reactor Core........................................................................................... 29 5.3.2 Reactor FueL........................................................................................... 30 5.3.3 Control Rods and Control Rod Drives .............................................................. 31 5.4 Fissionable Material Storage .................................................................................... 31 6.0 Administrative Controls..................................................................................................... 31 6.1.1 Structure................................................................................................ 32 6.1.2 Responsibilities........................................................................................ 32 6.1.3 Staffing.................................................................................................. 32 6.1.4 Selection and Training of Personnel ................................................................ 32 6.2 Review, Audit, Recommendation and Approval............................................................... 32 6.2.1 NSC Composition and Qualifications............................................................... 33 6.2.2 NSC Charter and Rules` ............................................................................. 33 6.2.3 Review Functiont...................................................................................... 33 6.2.4 Audit/Inspection Function ............................................................................ 34 6.3 Radiation Safety. ............................................. ..................................................... 34 6.4 Procedures ........................................................................................................ 34 6.4.1 Reactor Operations Procedures........ .................. ........................................... 34 6.4.2 Health Physics Procedures .......................................................................... 35 6.5 Experiment Review and Approval............................................................................... 35 6.6 Required Actions.................................................................................................. 35 6.6.1 Actions to be taken in case of a safety limit violation.............................................. 35 6.6.2 Actions to be taken for reportable occurrences`................................................... 36 Amendment No. 4 V
6.77Rep Ortstn R po t ..................................................... ........ 36 6.7.2 Special Reports........................................................................................ 38 6.8 Records ............................................................................................................. 39 Fig. 6.1 UCD/MNRC Organization for Licensing and Operation*........................................................... 40 Amendment No. 4 vi
0 TECHNICAL SPECIFICATIONS FOR THE UNIVERSITY OF CALIFORNIA - DAVIS/MCCLELLAN NUCLEAR RADIATION CENTER (UCD/MNRC) General The University of California - Davis/McClellan Nuclear Radiation Center (UCD/MNRC) reactor is operated by the University 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 provides a wide range of irradiation services for both research and industrial needs. The reactor operates at a nominal steady state power level up to and including 2 MW. The UCD/MNRC reactor is also capable of square wave and 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 Nuclear Regulatory Commission (e.g., senior reactor operator or reactor operator) to carry out the duties and responsibilities associated with the position requiring the license. 1.2.1 Senior Reactor Operator. An individual who is licensed to direct the activities of reactor operators 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 and perform reactor-related maintenance. 1.3 Channel. A channel is the combination of sensor, line amplifier, processor, and output 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 the channel for verification that it is operable. 1.3.2 Channel Calibration. A channel calibration is an adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm or trip, and shall be deemed to include a channel test. 1.3.3 Channel Check. A channel check is a qualitative verification of acceptable performance by observation of channel behavior. This verification, where possible, shall include comparison of the channel with other independent channels or systems measuring the same variable. 1.4 Confinement. Confinement means isolation of the reactor room air volume such that the movement of 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, fission chambers, foils, etc), which is designed to investigate specific reactor characteristics or which is intended for irradiation within an experiment facility and which is not rigidly secured to a core or shield structure so as to be a part of their design. 1.5.1 Experim~ent. Moveable. A moveable experiment is one where it is intended that the entire experiment may be moved in or near the reactor core or into and out of reactor experiment facilities while the reactor is operating. Amendment No. 4 1
- 0 1.5.2 Experiment. Secured. A secured experiment is any experiment, experiment facility, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means. The restraining force rmust be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the experiment, or by forces which can arise as a result of credible 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 their associated input circuits, which are designed to initiate a scram for the primary purpose of protecting an experiment or to provide information which requires manual protective action to be initiated. 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 can be either 8.5, 20, or 30 wt%, with an enrichment of less than 20% U-235. A standard fuel element may contain a burnable poison. 1.7 Fuel Element. Instrumented. An instrumented fuel element is a standard fuel element fabricate~d with thermocouples for temperature measurements. An instrumented fuel element shall have at least one operable 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 a channel. 1.9 Mode, Steady-State. Steady-state mode operation shall mean operation of the UCD/MNRC reactor with 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/MNRC reactor 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 the selector switch 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 the UCD/MNRC reactor safety system in response to a variable or condition of the UCD/MNRC reactor facility having 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 the generation 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. 4 2
1.15.3 Instrument generation System Level. and transmission At the protective of the command signal for instrument level, protective the safety shutdown equipmentaction is the to operate. 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 QOerational Core. A pulse operational core is a reactor operational core for which the maximum 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 reactor is at ambient temperature and the reactor is critical. (K*, = 1) 1.18 Reactivity Limits. The reactivity limits are those limits imposed on the reactivity conditions of the reactor core. 1.19 Reactivity Worth of an Experiment. The reactivity worth of an experiment is the maximum value of the reactivity 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 which directly 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 the parameters 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 in these Technical 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 input channels, 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 off position and the key.is removed from the lock and under the control of a licensed operator, and the conditions of a or b exist:
- a. (1) The minimum number of control rods are fully inserted to ensure the reactor is shutdown, as required by 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 being moved or serviced ifthe experiments have, on movement, a reactivity worth exceeding the maximum value allowed for a single experiment or $1.00, whichever is smaller, or
- b. The reactor contains insufficient fissile materials in the reactor core, adjacent experiments or control rods to attain criticality under optimum available conditions of moderation and reflection.
1.25 _Reactor Shutdown. The UCD/MNRC reactor is shutdown ifit 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. Amendment No. 4 3
- 0 1.26 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 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, and which 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 a fuel 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 allow it to perform its control function, and its safety function when the coupling is disengaged. This safety function is commonly termed a scram. 1.28.1 Regqulatingq Rod. A regulating rod is a Control rod'used to maintain an intended power level and 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 of radioactivity. 1.31 Scram Time. Scram time is the elapsed time between reaching a limiting safety system set point and the 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 denote a 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 to provide confidence that the reactor can be made subcritical by means of the control and safety system starting from any permissible operating condition with the most reactive rod assumed to be in the most reactive position, and once this action has been initiated, the reactor will remain subcritical without further operator 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 manual shutdown 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 checks and tests, and calibrations. Operability checks and tests are generally specified as daily, weekly or 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 surveillance interval is the interval between a check, test, or calibration, whichever is appropriate to the item Amendment No. 4 4
0 fuel element temperature. This parameter is well suited as it can be measured directly. A loss in the integrity of the fuel element cladding could arise ifthe cladding stress exceeds the ultimate strength of the cladding material. The fuel element cladding stress is a function of the element's internal pressure while the ultimate strength of the cladding material is a function of its temperature. The cladding stress is a result of the internal pressure due to the presence of air, fission product gasses and hydrogen from the disassociation 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 hydrogen to zirconium in the alloy. At a fuel temperature of 930°C for ZrH1 .7 fuel, the cladding stress due to the internal pressure is equal to the ultimate strength of the cladding material at the same temperature (SAR Fig 4.18). This is a conservative limit since the temperature of the cladding material is always 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 50000. Analysis (SAR Chapter 4, Section 4.5.4.1.1), shows that a maximum temperature for the clad during a pulse which gives a peak adiabatic fuel temperature of 100000 is estimated to be 470°C. Further analysis (SAR Section 4.5.4.1.2), shows that the internal pressure for both Zr. 65 (at 115000) and Zr1 .7 (at 1100°C) increases to a peak value at about 0.3 sec, at which time the pressure is about one-fifth of the equilibrium value or about 400 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 were approximately the same as for the 115000 case. Similar results were found for fuel with ZrH1 .7. Measurements of hydrogen pressure in TRIGA fuel elements during transient operations have been made and compared 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 100000, the pressure (ZrH1 .65 ) was only about 6% of the equilibrium value evaluated at the peak temperature. Calculations of the pressure gave values about three times greater than the measured values. The analysis gives strong indications that the cladding will not rupture if fuel temperatures are never greater than 120000 to 125000, providing the cladding temperature is less than 5000 C. For fuel with ZrH 1.7 ,a conservative safety limit is 110000. As a result, at this safety limit temperature, the class pressure 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 element temperature. 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 more conservatively) as measured in an instrumented fuel element. One instrumented element shall 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 a temperature which, ifexceeded, shall cause a reactor scram to be initiated preventing the safety limit from being exceeded. A setting of 75000 provides a safety margin at the point of the measurement of at least 13700 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 measured temperatures 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 the center and mid-plane of the fuel element. For pulse operation of the reactor, the same limiting safety system setting applies. However, the temperature channel will have no effect on limiting ;the peak power generated because of its Amendment No. 4 6
0 ! relatively long time constant (seconds) as compared with 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 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 the purpose 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/MNRC TRIGA 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 result 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% 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 phase changes at temperatures in excess of 1000 C. However, after extensive steady state operation at two (2) MW the hydrogen will redistribute due to migration from the central high temperature regions of 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 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. Ifthe pulse insertion is such that the temperature of the fuel exceeds about 8750 C, then the pressure may be sufficient to cause expansion of microscopic holes in the fuel that grow with each pulse. Analysis (SAR Chapter 13, Section 13.2.2.2.1), shows that the limiting pulse, for the worst case conditions, is 1.34% Ak/k ($1.92). Therefore, the 1.23% Ak/k ($1.75) limit is below the worse case reactivity insertion accident limit. The $0.90 square wave step insertion limit is also well below the worse case reactivity insertion accident limit. Amendment No. 4 7
0 Basis -
- a. The apparent condition of the control rod assemblies shall provide assurance that the rods shall continue to perform reliably as designed.
- b. This assures that the reactor shall shut down promptly when a scram signal is initiated (SAR Chapter 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 reactor operator during reactor operations. Objective - The objective is to require that sufficient information is available to the operator to assure safe operation of the reactor. Specification - The reactor shall not be operated unless the channels described in Table 3.2.2 are operable and the information is displayed on the reactor console. Table 3.2.2 Required Reactor Instrumentation (Minimum Number Operable) Measuring Steady Square Channel Surveillance Channel State Wave Function Requirements* Pulse
- a. Reactor Power 2 0 2 Scram at 2.2 D,M,A Level Safety MW or less Channel
- b. Linear Power Automatic D,M,A 1 0 Channel Power Control
- c. Log Power 1 0 Startup D,M,A Channel Control
- d. Fuel Temperature 2 2 Fuel DM,A 2
Channel Temperature
- e. Pulse Channel I Measures PA Pulse NV & NVT
(*)Where: D - Channel check during each day's operation M - Channel test monthly A - Channel calibration annually P - 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 is properly monitored and indicated in the reactor control room (SAR Chapter 7, Sections 7.1.2 &
7.1.2.2). Amendment No. 4 9
- 0 3.3 Reactor Coolant Systems Applicability - 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 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 channels described in Table 3.3 are operable, and the information is displayed locally or in the control room. Table 3.3 REQUIRED WATER SYSTEMS AND INSTRUMENTATION Minimum Measuring Number Surveillance Channel/System Operable Requirements* Function: Channel/System
- a. Primary Coolant For operation of the D,Q,A Core Inlet reactor at 1.5 MW or Temperature higher, alarms on high Monitor heat exchanger outlet temperature of 45°C (1 130 F)
- b. Reactor Tank Alarms ifwater level M Low Water. drops below a depth of Monitor 23 feet in the reactor tank
- c. Purification** Alarms ifthe primary D,M,S Inlet Conduc- coolant water conductivity tivity Monitor is greater than 5 micromhos/cm
- d. Emergency Core For operation of the reactor D,S Cooling System at 1.5MW or higher, provides water to cool fuel in the event of a Loss of Coolant Accident for a minimum of 3.7 hours at 20 gpm from an appropriate nozzle
(*)Where: D - channel check during each day's operation A - channel calibration annually Q - 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 shall be shutdown. Amendment No. 4 13
Basis -
.a. Table 3.3. The primary coolant core inlet temperature alarm assures that large power fluctuations will not occur (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 is 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 ifthe water level drops below a height of 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 a week 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 System Applicability - 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 reactor 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.
Specification -
- a. The reactor shall not be operated unless the reactor room exhaust system is in operation and the pressure 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 Coolant 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 room will be reduced due to mixing with exhaust system air (SAR Chapter 9, Section 9.5.1). Pressure in the reactor room will be negative relative to surrounding areas due to air flow patterns created, by the reactor room 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 fuel temperature 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 Systems 3.7.1 Monitorinq Systems Applicability - This specification applies to the information which shall be available to the reactor operator during reactor operation. Amendment No. 4 14
0 Objective - 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.
.Specification - The reactor shall not be operated unless the channels described in Table 3.7.1 are operable, the readings are below the alarm setpoints, and the information is displayed in the control room. The stack and reactor room CAMS shall not be shutdown at the same time during reactor operation.
Table 3.7.1 REQUIRED RADIATION MONITORING iNSTRUMENTATION Minimum Measuring Number Channel Surveillance Equipment Operable** Function Requirements*
- a. Facility I Monitors Argon-41 and D,W,A Stack Monitor radioactive particu-lates, and alarms
- b. Reactor Room 1 Monitors the radiation D,W,A Radiation level in the reactor Monitor room and alarms
- c. Purification 1 Monitors radiation D,W,A System Radia- level at the demineral-tion Monitor izer station and alarms
- d. Reactor Room 1 Monitors air from the D,W,A Continuous reactor room for parti-Air Monitor culate and gaseous radioactivity and alarms
(*)Where: AD -- channel channel calibration check during each day's operation annually W - 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 release of radioactive material to the environment (SAR Chapter 11, Section 11.1.1.1.4). The alarm setpoint 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 the operations area.
- b. Table 3.7.1. The reactor room radiation monitor provides information regarding radiation levels in the reactor 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 provides information regarding radioactivity in the primary system cooling water (SAR Chapter 11, Section 11.1.5.4.2)
Amendment No. 4 15
and allows assessment of radiation levels in the area to ensure that personnel radiation doses will be below 10 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 by the discharge of Argon-41 from the UCD/MNRC reactor facility. Specification - The annual average unrestricted area concentration of Argon-41 due to releases of this radionuclide from the UCD/MNRC, and the corresponding annual radiation dose from Argon-41 in 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 is specified in Appendix B, Table 2, Column 1 of 10 CER Part 20. 10 CER 20.1301 specifies dose limitations in the unrestricted area. 10 CFR 20.1101 specifies a constraint on air emissions of radioactive material to the environment. The SAR Chapter 11, Section 11.1.1.1.4 estimates that the routine Argon-41 releases and the corresponding doses in the unrestricted area will be below these limits. 3.8 Experiments 3.8.1 Reactivity Limits. Aoplicability - This specification applies to the reactivity limits on experiments installed in specific reactor experiment facilities. Obiective - The objective is to assure control of the reactor during the irradiation or handling of experiments 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-tank irradiation 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 or in-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 in any other reactor in-core and in-tank irradiation facilities at any given time shall be less than one dollar and ninety-two cents ($1.92) (1.34% Ak/k), including the potential reactivity which might result from malfunction, flooding, voiding, or removal and insertion of the experiments.
Amendment No. 4 16
Basis -
- a. A limitation of less than dollar ($1.00) onepneumatic (0.7% Ak/k) on the reactivity worth of a single movable experiment positioned in the 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 limit of $1.75 is not exceeded (SAR Chapter 13, Section 13.2.2.2.1). In addition, 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 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 secured experiment is its sudden accidental or unplanned removal while the reactor is operating. For such 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 Chapter 10, 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 control over the positioning of samples into and out of the central core region 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 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 the fuel temperature safety limit.
3.8.2 Materials Limit 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 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 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 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 200 millicuries. An additional 800 millicuries of 1-125 in sealed storage containers may also be present in the reactor room fume hood.
Amendment No. 4 17
S
- e. Explosive materials in quantities greater than 25 milligrams of TNT equivalent shall not be irradiated 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 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 any radiography bay. The irradiation of explosives in any bay is limited to those assemblies where a safety analysis has been performed that shows that there is no damage to the 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 of materials.
- b. The 1.5 curies limitation on iodine 131 through 135 assures that in the event of failure of a fueled experiment leading to total release of the iodine, occupational doses and doses to members 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 twenty (20.0) curies in the reactor room glove box and to one (1.0) curie in the reactor room fume hood assures that, ifthese inventories of 1-125 are totally released into their respective containments, occupational doses and doses to members of the general 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 of 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 any radiography bay external to the reactor tank will not result in damage to the reactor controls or the 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 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 and type of material in the experiment shall be limited such that the airborne radioactivity in the reactor room will not result in exceeding the applicable dose limits in 10 CFR Part 20 in the unrestricted area, assuming 100% of the gases or aerosols escapes. Amendment No. 4 18
0
- 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 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 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 by boiling this material can escape only through an undistributed column of water above the core, at least 10% of these vapors can escape.
- c. If a capsule fails and releases material which could damage the reactor fuel or structure by corrosion 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 his designated 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 reactor room 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 an experiment 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 be brought to the attention of the UCD/MNRC Director or his designated alternate for review to assure 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 reactor operating 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 time period. Prior to reactor system or measuring channel operation, the surveillance shall be performed for each reactor 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 level monitoring channels. Obiective - The objective is to verify that the maximum power level of the reactor does not exceed the authorized limit. Amendment No. 4 19
0 0 Discovery of noncompliance with Technical Specifications 4.3.a-c shall limit operations to that required to perform the 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 input signal. 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 input 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 for power levels above 1.5 MW. A channel calibration semiannually assures that the Emergency Core Cooling System performs as required for power levels above 1.5 MW.
4.4 Reactor Room Exhaust System Applicability - This specification applies to the surveillance requirements for the reactor room exhaust 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's operation. 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 verify that the exhaust system is maintaining a negative pressure in the reactor room relative to the surrounding facility areas. 4.5 This section intentionally left blank 4.6 This section intentionally left blank. 4.7 Reactor Radiation Monitorinq Systems Applicability - 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. 4 25
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 perform the surveillance. Basis -
- a. A channel check of the facility stack monitor system during each day's operation will assure the monitor is operable. A channel test weekly will assure that the system responds correctly to a 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 assure that the monitor is operable. A channel test weekly will ensure that the system responds to a known source. 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 assures that the monitor is operable. A channel test weekly will ensure that the system responds to a known source. 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 responds correctly to a known source. A channel calibration annually will assure that the CAM reads correctly.
4.8 Experiments Applicability - This specification applies to the surveillance requirements for experiments installed in any UCD/MNRC reactor experiment facility. Amendment No. 4 26
0 Objective - The objective is to prevent the conduct of experiments or irradiations which may damage the 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 a written safety analysis has been performed and reviewed by the UCD/MNRC Director, or his designee, 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 Facility Use Authorization. Facility Use Authorizations and experiments carried out under these authorizations 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/MNRC experiment facility, until it has been reviewed for compliance with the approved experiment and Facility Use Authorization 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 other UCD/MNRC reactor in-core or in-tank irradiation facility shall be estimated or measured, as 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 in any UCD/MNRC reactor experiment facility.
Basis - a & b. Experience at most TRIGA reactor facilities verifies the impo'rtance of reactor staff and safety committee reviews of proposed experiments.
- c. Measurement of the reactivity worth of an experiment, or estimation of the reactivity worth based 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 will facilitate maintaining surveillance over such experiments.
5.0 Desigqn Features 5.1 Site and Facility Description. 5.1.1 Sit..ee Applicability - This specification applies to the UCD/MNRC site location and specific facility design features. Objective - The objective is to specify those features related to the Safety Analysis evaluation. Specification - Amendment No. 4 27
0
- a. The site location is situated approximately 8 miles (13 kin) north-by-northeast of downtown Sacramento, 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. The unrestricted 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 has 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 of the site 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 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 to be beyond a credible accident scenario.
5.1.2 Facility Exhaust 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 of radioactivity released into the environment, or during a Loss of Coolant Accident, the system is to 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 and exhaust air from the UCD/MNRC facility. The system shall have an exhaust stack height of a minimum of 18.2m (60 feet) above ground level.
- b. Manually activated shutdown controls for the exhaust system shall be located in the reactor control room.
Basis - The UCD/MNRC facility exhaust system is designed such that the reactor room shall be maintained at a negative pressure with respect to the surrounding areas. The free air volume within the 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 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 to operating personnel. 5.2 Reactor Coolant System Applicability - This specification applies to the reactor coolant system. Amendment No. 4 28
0 Objective - The objective is to assure that adequate water is available for cooling and shielding during normal 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 ifthe water level in the reactor tank drops below a depth of 23 ft.
- c. For operations at 1.5 MW or higher during a Loss of Coolant Accident the reactor core shall be cooled for 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 be exceeded under natural convection flow conditions.
- b. A reactor tank water low level alarm sounds when the water level drops significantly. This alarm annunc~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 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 Reactor Core Applicability - This specification applies to the configuration of the fuel. Objective - The objective is to assure that provisions are made to restrict the arrangement of fuel elements 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 96 or more fuel elements to include fuel followed control rods. Below 0.5 MW there is no minimum 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 be loaded 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 20E core configurations. Amendment No. 4 29
0 S 308 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 done before 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 above loading restrictions will be allowed in an operational 0.5 MW or greater core. Specifications for the 20E core and 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 Fuel Applicability - 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 their physical and 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 to 2.3 megawatts steady state or to a 36 megawatt-sec pulse assures an ample margin of safety between 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 equal to 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 normal operation, 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 of purchased materials.
Amendment No. 4 30
0 0 5.3.3 Control Rods and Control Rod Drives Applicability - This specification applies to the control rods and control rod drives used in the reactor core. Obiective - The objective is to assure the control rods and control rod drives are of Such a design 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, 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 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, B4 C powder, or boron and its compounds. These materials shall be contained in a suitable clad material such as stainless steel or aluminum to assure mechanical stability during movement and to isolate the neutron poison from the tank water environment. 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 at the 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 Storaqe Applicability - This specification applies to the storage of reactor fuel at a time when it is not in the reactor core. Obiective - The objective is to assure that the fuel which is being stored will not become critical and will not 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 the keff 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 Controls 6.1 Orqanization. The Vice Chancellor for Research shall be the licensee for the UCD/MNRC. The UCD/MNRC facility shall be under the direct control of the UCD/MNRC Director or a licensed senior reactor operator (SRO) designated by the UCD/MNRC Director to be in direct control. The UCD/MNRC Director shall be accountable to the Vice Chancellor of the Office of Research for the safe operation and maintenance of the reactor and its associated equipment. Amendment No. 4 31
0 6.1.1 Structure. The management for operation of the UCD/MNRC facility shall consist of the organizational structure as shown in Figure 6.1. 6.1.2 Responsibilities. The UCD/MNRC Director shall be accountable to the Vice Chancellor of the 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 all experiments and experiment*{ procedures prior to their use in the reactor. Individuals in the management organization (e.g., Reactor Manager, Health Physics Manager, etc.) 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 the operating license and technical specifications. The Reactor Manager and Health Physics Manager report directly to the UCD/MNRC Director. 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 reactor operator should be within thirty (30) minutes of the facility and reachable by telephone, and;
- d. A senior reactor operator shall be present whenever a reactor startup is performed, 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 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 and Training of Personnel for Research Reactors (ANS 15.4). Qualification and requalification of licensed operators shall be subject to an approved Nuclear Regulatory Commission (NRC) program. 6.2 Review. Audit. Recommendation and Approval General Policy. Nuclear facilities shall be designed, constructed, operated, and maintained in such 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 with applicable regulatory requirements. The UCD Vice Chancellor of the Office of 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. Amendment No. 4 32
0 6.2.1 NSC Composition and Qualifications. The UCD/MNRC Director shall appoint the Chairperson 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 shall evaluate 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 proposed Class I modifications, prior to implementing (Class I) modifications, prior to taking action under the preceding 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 an unreviewed safety question or require a change to the license, to a Facility Use Authorization, or to the Technical Specifications. This determination may be in the form of verifying a decision already made by the UCD/MNRC Director.
- d. Review reactor operations and operational maintenance, Class I modification records, and 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 for UCDIMNRC 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 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 other inspections of these facilities conducted by other agencies.
Amendment No. 4 33
0 6.2.4 AuditlInspection Function. The NSC or a subcommittee thereof, shall audit/inspect reactor operations 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 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 and Technical 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 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 the UCD/MNRC Radiation Safety Program. The program should use the guidelines of the American National Standard for Radiation Protection at Research Reactor Facilities (ANSI/ANS 15.11). The Health Physics Manager 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 of procedures shall be performed and documented in a timely manner by the UCD/MNRC staff to assure that procedures are current. Procedures shall be adequate to assure the safe operation of the 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 or other routine maintenance that could have an effect on reactor safety.
- e. Testing and calibration of reactor instrumentation and controls, control rods and control rod drives.
Amendment No. 4 34
f.that Administrative controlssafety could affect reactor for operations, maintenance, and conduct of irradiations and experiments 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 alarms 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 solid and liquid waste and gaseous effluents released from the facility. The program shall include a management 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 items from 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 Approval. Experiments having similar characteristics are grouped together for review 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/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 approval as a new experiment.
- c. Minor changes to an experiment that do not significantly alter the experiment may be approved by a senior reactor operator.
6.6 Required Actions 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: Amendment No. 4 35
0.Q
- 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 the UCD/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 and contributing factors. (2) Effect of the violation upon reactor facility components, systems, or structures, and on 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 the NRC 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 is necessary to shut down the reactor to correct the occurrence, operations shall not be resumed unless 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 the NRC Operations Center within 24 hours of the occurrence. A written preliminary report shall be sent to the NRC, Attn: Document Control Desk, I White Flint North, 11555 Rockville Pike, Rockville MD 20852, within 14 days of the occurrence. A final written report shall be sent to the above address within 30 days of the occurrence.
- d. Reportable occurrences should be reviewed by the NSC prior to forwarding any written report to 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: Document Control 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 the previous calendar year shall be submitted within six months following the end of each calendar year. Each annual report shall include the following information:
- a. A brief summary of operating experiences including experiments performed, changes in facility design, 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 reactor was critical, and the cumulative total energy output since initial criticality.
Amendment No. 4 36
*
(2) The written report (and, to the extent possible, the preliminary telephone report or report by similar conveyance) shall describe, analyze, and evaluate safety implications, and outline 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 or previously 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 the Safety Analysis Report (SAR); (3) A personnel change involving the positions of UCD/MNRC Director or UCD Vice Chancellor 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 or development 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 required information may be contained in single or multiple records, or a combination thereof. Records and logs shall 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 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 shipped off 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. 4 39
0 72hZ I - UNIVERSITY OF CALIFORNIA - DAVIS CE CHANCELLOR FOR RESEARCH (Licensee) I -
,TOR TIONS NJCHI
....................... Formal Licensing Channel Administrative Reporting Channel
- - - - Communications Channel UCD/MNRC ORGANIZATION FOR LICENSING AND OPERATION FIGURE 6.1 40
Dr. Barry M. Klein Vice Chancellor for Research University of California, Davis One Shields Avenue Davis, 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 Facility Operating License No. 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 to your submittal of October 17, 2002, and is discussed in the enclosed Safety Evaluation Report. Sincerely, Warren J. Eresian, Project Manager Research and Test Reactors Section Operating Reactor Improvements Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No. 50-607
Enclosures:
- 1. Amendment No. 5
- 2. Safety Evaluation Report
S University of California - Davis/McClellan MNRC Docket No. 50-607 cc: Dr. Wade J. Richards 5335 Price Avenue, Bldg. 258 McClellan AFB, CA 95652-2504 Test, Research, and Training Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611
Dr. Barry M. Klein Vice Chancellor for Research University of California, Davis One Shields Avenue Davis, 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 Facility 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 October 17, 2002, and is discussed in the enclosed Safety Evaluation Report. Sincerely, Warren J. Eresian, Project Manager Research and Test Reactors Section Operating Reactor Improvements Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No. 50-607
Enclosures:
- 1. Amendment No. 5
- 2. Safety Evaluation Report DISTRIBUTION:
PUBLIC RORP\R&TR r/f SHolmes OGC MMendonca WEresian TDragoun PMadden AAdams PDoyle CBassett DMatthews EHylton Plsaac DHughes WBeckner GHiIl (2) (T5-C3) LBerg ADAMS ACCESSION NO: ML02 TEMPLATE #: NRR-058 NAME WEresian:rdr EHylton SUttal PMadden WBeckner IiDATE 111/ /2002 11/ /2002 11/ /2002 11/ /2002 11/ /2002J OFFICIAL RECORD COPY
- 0 REGENTS OF THE UNIVERSITY OF CALIFORNIA AT McCLELLAN NUCLEAR RADIATION CENTER DOCKET NO. 50-607 AMENDMENT TO AMENDED FACILITY OPERATING LICENSE Amendment No. 5 License No. R-130
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that A. 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 Nuclear Radiation 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 the regulations of the Commission as stated in Chapter I of Title 10 of the Code of FederalRegulations (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 the Commission; D. The issuance of this amendment will not be inimical to the common defense and security 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; and F. 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. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the enclosure to this license amendment, and paragraph 2.C.(ii) of Amended Facility 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 through Amendment No. 5, are hereby incorporated in the license. The licensee shall 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 Warren J. Eresian, Project Manager Research and Test Reactors Section
*Operating Reactor Improvements Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation
Enclosure:
Appendix A, Technical Specification Changes Date of Issuance:
ENCLOSURE TO LICENSE AMENDMENT NO. 5 AMENDED FACILITY OPERATING LICENSE NO. R-130 DOCKET NO. 50-607 Replace the following pages of Appendix A, Technical Specifications, with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. Remove Insert 17 17 18 18 40 40
Basi.__s-
- a. A limitation positioned of less in the than one pneumatic transfer ($1 .00)(0.7%Ak/k) dollar tube, on the reactivity the central irradiation of a single movable experiment worth Chapter facility (SAR, 10, Section 10.4.1 ), the central 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). In addition, 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 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 secured experiment is its sudden accidental or unplanned removal while the reactor is operating. For such 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 (AC1F)(SAR Chapter 10, Section 1.0.4.2) 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 control .over the positioning of samples into and out of the central core region 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 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 the fuel temperature safety limit.
3.8.2 Materials Limit 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 radioactivity by limiting material 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 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 I-125 in the 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 1-125 being 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 600 millicuries in sealed storage containers) may also be present in the 1-125 fume hood. Amendment No. 5 17
- 0
- e. Explosive materials in quantities greater than 25 milligrams of TNT equivalent shall not be irradiated 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 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 any radiography bay. The irradiation of explosives in any bay is limited to those assemblies where a safety analysis has been performed that shows that there is no damage to the reactor safety systems upon detonation (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 fueled experiment leading to total release of the iodine, occupational doses and doses to members 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 I-125 fume hood assures that, if either of these inventories of 1-125 is totally released into their respective containments, the occupational doses and doses to members of the general 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 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 bay external to the reactor tank will not result in damage to the reactor controls or the 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. 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 or aerosols into the reactor building or into the unrestricted area, the quantity and type of material in the experiment shall be limited such that the airborne radioactivity in the reactor room will not result in exceeding the applicable dose limits in 10 CFR Part 20 in the unrestricted area, assuming 100% of the gases or aerosols escapes. Amendment No. 5 18
S 0
- 0 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 5 TO AMENDED FACILITY OPERATING LICENSE NO. R-130 REGENTS OF THE UNIVERSITY .OF CALIFORNIA AT McCLELLAN NUCLEAR RADIATION CENTER DOCKET NO. 50-607
1.0 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 Facility Operating License No. R-i130 for the McClellan Nuclear Radiation Center (MNRC) TRIGA research reactor. The request provides for the following changes, which if implemented, will result in 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 a Health 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 to whom reports the Health Physics Manager and the Reactor Operations Manager. The proposed management 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. This amendment approved the installation of an Iodine-125 production loop. The production loop included 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, the reactor 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 exhaust system, eventually to make its way to the unrestricted area. The exposure (CEDE to the thyroid) to a person in the unrestricted area for the entire 30 second duration of this event is much less than 1 millirem. Ifthe exposure duration is increased to 10 minutes, the estimated CEDE to the thyroid would still be less than 1 millirem. For those exposed in the reactor room for the maximum assumed occupancy time of 5 minutes the CEDE to the thyroid would be about 67 millirem. The results of all of the assumptions and calculations in the accident sequence are directly proportional to the initial inventory of Iodine-125 in the production system. Increasing the initial assumed inventory from 20 curies to 61 curies will simply result in a tripling of the exposure. The analysis in the SAR that supports the increase in iodine inventory shows that the CEDE to the thyroid for a 10-minute exposure in the unrestricted area would be about 3.0 millirem, For those exposed in the reactor room for the maximum assumed occupancy time of 5 minutes the CEDE to the thyroid would be about 205 millirem. In order to assess the potential consequences of the worst-case assumption, the resulting doses are 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 been analyzed in the licensee's Safety Analysis Report (SAR), and is a complete cladding rupture of a highly-irradiated single fuel element, followed by the instantaneous release of fission products into the air.. The accident analysis calculates the radiological consequences of the MHA with regard to doses to the general public in the unrestricted area, and also calculates occupational doses within the site boundary. The MHA results in a CEDE of 53 millirem in the unrestricted area. Since the release of 61 curies of Iodine-125 through the glovebox exhaust system and eventually to the unrestricted 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 less than 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 the allowable activity of lodine-125 in the Iodine Production Facility from 20 curies to 61 curies does not significantly reduce the margin of safety with respect to the Maximum Hypothetical Accident and 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 within the 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 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 criteria for categorical exclusion set forth in 10 Amendment No. 5
0 0 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 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 or consequences of accidents previously evaluated, or create the possibility of a new or different kind of accident from 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 is reasonable 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 or the health and safety of the public. Principal Contributor: Warren J. Eresian Date: Amendment No. 5
- U
.*NUCLEAR REGUALTORY COMMISSION i __ _ _ _ _ _ _ _ _ _ _ _
.Ii UNIVERSITY OF l CALIFORNIA - DAVIS
- VICE CHANCELLOR FOR I ~RESEARCHi
- (Licensee)
I I I II SDIRECTOR NUCLEAR. H_____SAFETYCO L I l A-tC--SITE MANAGER[ 1
- COMMITITEE I
i i-***-* HEALTH PHYSICS REACTOR BRANCH OPERATIONS FormlChnne Liensig ___________ Aminstrtie RpotinBCANnel CormmunLicatinsin Channel UCD/MNRC ORGANIZATION FOR LICENSING AND OPERATION FIGURE 6.1
RE NIE SAE
** NUCLEAR REGULATORY COMMISSION
~WASHJNGTON, D.C. 20555-0001 N~ovemb~er 2_5, 2003 Dr. Barry M. Klein Vice Chancellor for Research University of California, Davis One Shields Avenue Davis, 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 to Facility 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 Safety Evaluation Report. Sincerely, 6~)4A,~.6 Warren J. Eresian, Project Manager Research and Test Reactors Section New, Research and Test Reactors Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No. 50-607
Enclosures:
- 1. Amendment No. 6
- 2. Safety Evaluation Report
- 0 University of California - Davis/McClellan MNRC Docket No. 50-607 cc:
Dr. Wade J. Richards 5335 Price Avenue, Bldg. 258 McClellan AFB, CA 95652-2504 Test, Research, and Training Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611
OUNITED W STATES
'-*-"NUCLEAR REGULATORY COMMISSION e* *r/*WASHINGTON, D.C. 20555-0001 e- ~ * *~
REGENTS OF THE UNIVERSITY OF CALIFORNIA AT McCLELLAN NUCLEAR RADIATION CENTER DOCKET NO. 50-607 AMENDMENT TO AMENDED FACILITY OPERATING LICENSE Amendment No. 6 License No. R- 130
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that A. 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 Nuclear Radiation Center (the licensee) on March 31, 2003, conforms to the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the regulations of the Commission as stated in Chapter I of Title 10 of the Code of FederalRegulations (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 the Commission; D. The issuance of this amendment will not be inimical to the common defense and security 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; and F. 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. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the enclosure to this license amendment, and paragraph 2.C.(ii) of Amended Facility 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 through Amendment No. 6, are hereby incorporated in the license. The licensee shall 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 Patrick M. Madder Seto Chief Research and Test Reactors Section New, Research and Test Reactors Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation
Enclosure:
Appendix A, Technical Specification Changes Date of Issuance: November 25, 2003
- 0 ENCLOSURE TO LICENSE AMENDMENT NO. 6 AMENDED FACILITY OPERATING LICENSE NO. R-130 DOCKET NO. 50-607 Replace the following pages Of Appendix A, Technical Specifications, with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Remove Insert. 31 31 32 32 33 33 Figure 6.1 Figure 6.1
- 0 5.4 Fissionable Material Storage AppDlicabilitv - This specification applies to the storage of reactor fuel at a time when itis not in the reactor core.
Obiective - The objective is to assure that the fuel which is being stored will not become critical and will not reach 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 keff 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 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 Controls 6.1 Organization. The Vice Chancellor for Research shall be the licensee for the UCD1MNRC. The UCD/M*NRC facility shall be under the direct control of the UCD/MNRC Director. The UCD/MNRC Director shall be accountable to the Vice Chancellor for Research for the safe operation and maintenance of the facility. 6.1.1 Structure. The management for operation of the UCD/MNRC facility shall consist of the organizational structure as shown in Figure 6.1. 6.1.2 Responsibilities. The UCDIMNRC Director shall be accountable to the Vice Chancellor for Research for the safe operation and maintenance of the facility. The UCDIMNRC Director, or his designated alternate, shall review and approve all experiments and experiment procedures prior to their use in the reactor. Individuals in 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 the operating license and technical 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 final decision. The Reactor Supervisor and Health Physics Supervisor report directly to the Operations 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 who can perform prescribed instructions;
- c. A senior reactor operator readily available. The available senior reactor operator should be within thirty (30) minutes of the facility and reachable by telephone, and;
- d. A senior reactor operator shall be present whenever a reactor startup is performed, 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 be available to the reactor operator in the control room. The list shall include: Amendment No. 6 31
- 0
- 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 and Training of Personnel for Research Reactors (ANS 15.4). Qualification and requalification of licensed operators shall be subject to an approved Nuclear Regulatory Commission (NRC) program. 6.2 Review. Audit. Recommendation and Approval General Policy. Nuclear facilities shall be designed, constructed, operated, and maintained in such amanner 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 with applicable 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 NSC Composition and Qualifications. The UCD Vice Chancellor for Research shall appoint the 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. The NSC Shall evaluate 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 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 incl-*u~de "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 proposed Class I modifications, prior to implementing (Class I) modifications, prior to taking action under the preceding 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 an unreviewed safety question or require a change to the license, to a Facility Use Authorization, or to the Technical Specifications. This determination may be in the form of verifying a decision already made by the UCD/MNRC Director.
Amendment No. 6 32
- d. Review health reactor physics operations program and operational and associated recordsmaintenance, Class Inuclear for all UCDIMNRC modification records, and the facilities.
- e. Review the periodic updates of the Emergency Plan and Physical Security Plan for UCD/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 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 other inspections of these facilities conducted by other agencies.
6.2.4 Audit/Inspection Function. The NSC or a subcommittee thereof, shall audit/inspect reactor operations 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 UCDIMNRC 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 and Technical 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 the UCD/MNRC Radiation Safety Program. The program should use the guidelines of the American National Standard for Radiation Protection at Research Reactor Facilities (ANSI/ANS 15.11). The Health 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 nthssction. 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 to assure that procedures are current. Procedures shall be adequate to assure the safe operation of the Amendment No. 6 33
- 0 I-..... COMMISSION UNVRIYO CAIOM AI SAFETYECOITYTEE AIFRI 1 C-MDATEES I VIEMANAGELLRFO I, I SUPERISRECREANCTO AR.. SFT OPERSUPERVISOR MA ANG ER_______________________________________________ i FormlChnne Liensig UCD/NRC ORGAIZAIOSOR REACTOR OETINLICNIGADSFT FIGURE 6.1
R R E*O4"*O OUNITED STATES
-* ,NUCLEAR REGULATORY COMMISSION
*.* WASHINGTON, D.C. 20555-0001 "
1-
**
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 6 TO AMENDED FACILITY OPERATING LICENSE NO. R-130 REGENTS OF THE UNIVERSITY OF CALIFORNIA AT McCLELLAN NUCLEAR RADIATION CENTER DOCKET NO. 50-607
1.0 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 Facility Operating License No. R-1 30 for the McClellan Nuclear Radiation Center (MNRC) TRIGA research reactor. The request provides for the following changes, which if implemented, will result in 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, and change the Technical Specifications and UCD/MNRC Organization Chart to reflect this change.
Each of these requests is discussed below. 2.0 EVALUATION The current organization structure includes an UCD/MNRC Director to whom reports a Site Manager. 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/MNRC Director, 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 the Chairperson of the NSC. In the proposed organization structure, that responsibility is given to the UCD Vice Chancellor for Research, who is also the licensee for the UCD/MNRC. Since this proposed change increases the level of oversight from the licensee's staff to the licensee, the staff concludes that it is acceptable.
The staff has reviewed the proposed changes to the TSs and concluded that they are administrative in nature and do not impact the licensee's ability to continue to meet the relevant requirements of 10 CFR 50.36.
3.0 ENVIRONMENTAL CONSIDERATION
This amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51 .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 the amendment 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 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 is reasonable 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 or the health and safety of the public. Principal Contributor: Warren J. Eresian Date: November 25, 2003 Amendment No. 6
0 0 TECHNICAL SPECIFICATIONS FOR THE UNIVERSITY OF CALIFORNIA - DAVIS MCCLELLAN NUCLEAR RADIATION CENTER (UCD/MNRC) DOCUMENT NUMBER: MNRC-0004-DOC-1 2 Rev 12 09/02
1 Oct. 16 0? 11:OOa Ar t~~r 6. Johnson tS4lJ 753-9743 p. 1,JLaJ. rwdIL. .atDfJh.L.* ~ r. t.wc. TECHNICAL SPECtFICATIONS APPROVAL Revision 12 Radiation Cencer "Teclnical Gpo of me(UOI)/MNRG) Reactor havefor ctifoons* the Universit undergone of California the following at DavistlMcCleIlan, Nuclear coordination: 10 ~ 02-Reviewed by;*ell/ Dale Rcvicwcd by'." ' floa rMnae " " D~kcI R~eviewed by: Site Manager Date I Approved by:
/~zL7z~OZ-UCD/MNRc~bir4ctor Data Approvod by; Date
- 0 Technical Specifications Rev 12 09/2002 TtePageRe12 902 Titovle Page Rev 12 9/2002 32 Rev 12 9/2002 Figure 6.1 Rev 12 9/2002
S 0 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Paage 1.0 Definitions ................................... ............................................................................ 2.0 Safety Limit and Limiting Safety System Setting (LSSS.)............................................................ 5 2.1 Safety Limits................................................................................................... 5 2.2 Limiting Safety System Setting (LSSS)...................................................................... 6 2.2.1 Fuei Temperature.................................................................................. 6 3.0 Limiting Conditions for Operations (LC.O.0............................................................................. 7 3.1 Reactor Core Parameters.................................................................................... 7 3.1 .1 Steady-State Operation ........................................................................... 7 3.1.2 Pulse or Square Wave Operation ................................................................ 7 3.1.3 Reactivity Limitations .............................................................................. 8 3.2 Reactor Control and Safety Systems........................................................................ 8 3.2.1 Control Rods....................................................................................... 8 3.2.2 Reactor Instrumentation................................... ........................................ 9 3.2.3 Reactor Scrams and Interlocks.................................................................. 10 3.2.4 Reactor Fuel Elemenis........................................................................... 12 3.3 Reactor Coolant Systems................................................................................... 13 3.4 Reactor Room Exhaust System ............................................................................ 14 3.5 Intentionally Left Blank ...................................................................................... 14 3.6 Intentionally Left Blank ...................................................................................... 14 3.7 Reactor Radiation Monitoring Systems .................................................................... 14 3.7.1 Monitoring Systems .............................................................................. 14 3.7.2 Effluents - Argon-41 Discharge Limit..............,.............................................. 16
9 0 Page* 3.8 Experiments ................................................................................................. 16 3.8.1 Reactivity Limits................................................................................... 16 3.8.2 Materials Limit.................................................................................... 17 3.8.3 Failure and Malfunctions......................................................................... 18 4.0 Surveillance Requirements .......................................................................................... 19 4.1 Reactor Core Parameters................................................................................... 19 4.1.1 Steady State Operation.......................................................................... 19 4.1.2 Shutdown Margin and Excess Reactivity ....................................................... 20 4.2 Reactor Control and Safety Systems ...................................................................... 20 4.2.1 Control Rods ..................................................................................... 20 4.2.2 Reactor Instrumentation ......................................................................... 21 4.2.3 Reactor Scrams and Interlocks.................................................................. 22 4.2.4 Reactor Fuel Elemen*ts........................................................................... 23 4.3 Reactor Coolant Systems*................................................................................. .24 4.4 Reactor Room Exhaust System ............................................................................ 25 4.5 Intentionally Left Blank...................................................................................... 25 4.6 Intentionally Left Blank...................................................................................... 25 4.7 Reactor Radiation Monitoring Systems .................................................................... 25 4.8 Experiments ................................................................................................. 26 5.0 Design Features ...................................................................................................... 27 5.1 Site and Facility Description ................................................................................ 27 5.1.1 .Site................................................................................................ 27 5.1.2 Facility Exhaust .................................................................................. 28 5.2 Reactor Coolant System..................................................................................... 28
0 S 5.3 Reactor Core and F~ue]........................................................................................ 29 5.3.1 Reactor Core .......................................................................... :............. 29 5.3.2 Reactor .FuelJ........................................................................................ 30 5.3.3 Control Rods and Control Rod Drives ............................................................ 31 5.4 Fissionable Material Storage.................................................................................. 31 6.0 Administrative Controls.................................................................................................. 31 6.1 Organization.................................................................................................... 31 6.1.1 Structure............................................................................................. 32 6.1.2 Responsibilities..................................................................................... 32 6.1.3 Staffing .............................................................................................. 32 6.1.4 Selection and Training of Personnel.............................................................. 32 6.2 Review, Audit, Recommendation and Approval............................................................. 32 6.2.1 NSC Composition and Qualifications ................................................ i............ 33 6.2.2 NSC Charter and Rules ........................................................................... 33 6.2.3 Review Function.................................................................................... 33 6.2.4 Audit/Inspection Function.......................................................................... 34 6.3 Radiation Safety................................................................................................ 34 6.4 Procedures ..................................................................................................... 34 6.4.1 Reactor Operations Procedures................................................................... 34 6.4.2 Health Physics Procedures........................................................................ 35 6.5 Experiment Review and Appro~ial ............................................................................ 35 6.6 Required Actions............................................................................................... 35 6.6.1 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 Reports .......................................................................................................... 36 6.7.1 Operating Reports.................................................................................. 36 6.7.2 Special Reports..................................................................................... 38 6.8 Records ......................................................................................................... 39 I Fig. 6.1 UCD/MNRC Organization for Licensing and Operation......................................................... 40
0 0 TECHNICAL SPECIFICATIONS FOR THE UNIVERSITY OF CALIFORNIA - DAVIS/MCCLELLAN NUCLEAR RADIATION CENTER (UCD/MNRC) General The University of California - Davis/McClellan Nuclear Radiation Center (UCD/MNRC) reactor is operated by the University 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 provides a wide range of irradiation services for both research and industrial needs. The reactor operates at a nominal steady state power level up to and including 2 MW. The UCD/MNRC reactor is also capable of square wave and 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 Nuclear Regulatory Commission (e.g., senior reactor operator or reactor operator) to carry out the duties and responsibilities associated with the position requiring the license. 1.2.1 Senior Reactor Operator. An individual who is licensed to direct the activities of reactor operators 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 and perform reactor-related maintenance. 1.3 Channel. A channel is the combination of sensor, line amplifier, processor, and output 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 the channel for verification that it is operable. 1.3.2 Channel Calibration. A channel calibration is an adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm or trip, and shall be deemed to include a channel test. 1.3.3 Channel Check. A channel check is a qualitative verification of acceptable performance by observation of channel behavior. This verification, where possible, shall include comparison of the channel with other independent channels or systems measuring the same variable. 1.4 Confinement. Confinement means isolation of the reactor room air volume such that the movement of 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, fission chambers, foils, etc), which is designed to investigate specific reactor characteristics or which is intended for irradiation within an experiment facility and which is not rigidly secured to a core or shield structure 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 entire experiment may be moved in or near the reactor core or into and out of reactor experiment facilities while the reactor is operating. I
0 0 1.5.2 Experiment. Secured. A secured experiment is any experiment, experiment facility, or component 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 experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the experiment, or by forces which can arise as a result of credible 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 their associated input circuits, which are designed to initiate a scram for the primary purpose of protecting an experiment or to provide information which requires manual protective action to be initiated. 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 can be either 8.5, 20, or 30 wt%, with an enrichment of less than 20% U-235. A standard fuel element may contain a burnable poison. 1.7 _Fuel Element. Instrumented. An instrumented fuel element is a standard fuel element fabricated with thermocouples for temperature measurements. An instrumented fuel element shall have at least one operable 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 a channel. 1.9 Mode, Steady-State. Steady-state mode operation shall mean operation of the UCD/MNRC reactor with 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/MNRC reactor 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 the selector 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 within the UCD/MNRC reactor safety system in response to a variable or condition of the UCD/MNRC reactor facility having 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 the generation 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. 2
- 0 1.15.3 Instrument System Level. At the protective instrument level, protective action is the generation 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 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 the maximum 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 reactor is at ambient temperature and the reactor is critical. (Keff = 1) 1.18 Reactivity Limits. The reactivity limits are those limits imposed on the reactivity conditions of the reactor core. 1.19 Reactivity Worth of an Experiment. The reactivity worth of an experiment is the maximum value of the reactivity 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 which directly 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 the parameters 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 in these 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 for initiation of manual protective action. 1.24 Reactor Secured. The UCD/MNRC reactor is secured when the console key switch is in the off position and the key is removed from the lock and under the control of a licensed operator, and the conditions of a or b exist:
- a. (1) The minimum number of control rods are fully inserted to ensure the reactor is shutdown, as required by 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 being moved or serviced ifthe experiments have, on movement, a reactivity worth exceeding the maximum value allowed for a single experiment or $1.00, whichever is smaller, or
- b. The reactor contains insufficient fissile materials in the reactor core, adjacent experiments or control rods to attain criticality under optimum available conditions of moderation and reflection.
1.25 Reactor Shutdown. The UCDIMNRC reactor is shutdown ifit 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
0 1.26 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, and which 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 a fuel 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 allow it to perform its control function, and its safety function when the coupling is disengaged. 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 power level and 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 of radioactivity. 1.31 Scram Time. Scram time is the elapsed time between reaching a limiting safety system set point and the 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 denote a recommendation; the word "may" to denote permission, neither a requirement nor a recommendation. 1.34 Shutdown Margin. Shutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence that the reactor can be made subcritical by means of the control and safety system starting from any permissible operating condition with the most reactive rod assumed to be in the most reactive position, and once this action has been initiated, the reactor will remain subcritical without further operator 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 manual shutdown 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 checks and tests, and calibrations. Operability checks and tests are generally specified as djaily, weekly or 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 surveillance interval is the interval between a check, test, or calibration, whichever is appropriate to the item being 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 involve an unreviewed safety question:
- a. If the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or
- b. If a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or
- c. 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. 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 reactor computerized 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 fuel element. 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 during steady-state operation.
- b. The maximum temperature in a standard TRIGA fuel element shall not exceed 1100 0C during pulse operation.
Basis -
- a. This fuel safety limit applies for conditions in which the cladding temperature is above 500 °C (Safety Analysis Report (SAR), Chapter 4, Section 4.5.4.1.3). The important parameter for a TRIGA reactor is 5
0 0 the fuel element temperature. This parameter is well suited as it can be measured directly. A loss in the integrity of the fuel element cladding could arise if the cladding stress exceeds the ultimate strength of the cladding material. The fuel element cladding stress is a function of the element's internal pressure while the ultimate strength of the cladding material is a function of its temperature. The cladding stress is a result of the internal pressure due to the presence of air, fission product gasses and hydrogen from the disassociation 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 hydrogen to zirconium in the alloy. At a fuel temperature of 930 °C for ZrH1 7 fuel, the cladding stress due to the internal pressure is equal to the ultimate strength of the cladding material at the same temperature (SAR Fig 4.18). This is a conservative limit since the temperature of the cladding material is always 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 a pulse which gives a peak adiabatic fuel temperature of 1000 °C is estimated to be 470 °C. Further analysis (SAR Section 4.5.4.1.2), shows that the internal pressure for both Zr 1.65 (at 11 50°C) and Zr17z (at 11 00°C) increases to a peak value at about 0.3 sec, at which time the pressure is about one-fifth of the equilibrium value or about 400 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 ZrH 1.65 fuel temperature greater than 1150 °C, over a 200°C range, show that the time to reach the peak pressure and the fraction of equilibrium pressure value achieved were 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 and compared 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 6. 5) was only about 6% of the equilibrium value evaluated at the peak temperature. Calculations of the pressure gave values about three times greater than the measured values. The analysis gives strong indications that the cladding will not rupture iffuel 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 for cladding failure. 2.2 Limiting Safety System Setting. 2.2.1 Fuel Temperature. Applicability - This specification applies to the protective action for the reactor fuel element temperature. Objective - The objective is to prevent the fuel element temperature safety limit from being reached. Specification - The limiting safety system setting shall be 750 °C (operationally this may be set more conservatively) as measured in an instrumented fuel element. One instrumented element shall 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 a temperature which, ifexceeded, shall cause a reactor scram to be initiated preventing the safety limit from being exceeded. A setting of 750 °C provides a safety margin at the point of the measurement 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 measured temperatures 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 the center and mid-plane of the fuel element. For pulse operation of the reactor, the same limiting safety system setting applies. However, the temperature channel will have no effect on limiting 6
0 the peak power generated because of its relatively long time constant (seconds) as compared with the width of the pulse (milliseconds). In this mode, however, the temperature trip will act to limit the energy release after the pulse ifthe transient rod sho~uld 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 the purpose 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/MNRC TRIGA 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 result 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 to 1.7. 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 state operation at two (2) MW the hydrogen will redistribute due to migration from the central high temperature regions of 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 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. Ifthe pulse insertion is such that the temperature of the fuel exceeds about 875 °C, then the pressure may be sufficient to cause expansion of microscopic holes in the fuel that grow with each pulse. Analysis (SAR Chapter 13, Section 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% 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 the reactivity 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 at all 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 by the 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) shall not exceed 6.65% Ak/k ($9.50).
Basis -
- a. This specification assures that the reactor can be placed in a shutdown condition from any operating condition and remain shutdown, even if the maximum worth control rod should stick in the fully withdrawn position (SAR Chapter 4, Section 4.5.5).
- b. This specification sets an overall reactivity limit which provides adequate excess reactivity to override the xenon buildup, to overcome the temperature change in going from zero power to 2 MW, to permit pulsing at the $1.75 level, to permit irradiation of negative worth experiments and account for fuel burnup over time. An adequate shutdown margin exists with an excess of $9.50 for the two analyzed cores: (SAR Chapter 4, Section 4.5.5).
3.2 Reactor Control and Safety Systems 3.2.1 Control Rods Applicability - 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 drive assemblies.
- b. The scram time measured from the instant a signal reaches the value of a limiting safety system setting to the instant that the slowest control rod reaches its fully inserted position shall not exceed one (1) second.
8
Basis -
- a. Thecontinue apparent to condition of the control rod assemblies shall provide assurance that the rods shall perform reliably as designed.
- b. This assures that the reactor shall shut down promptly when a scram signal is initiated (SAR Chapter 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 reactor operator during reactor operations. Objective - The objective is to require that sufficient information is available to the operator to assure safe operation of the reactor. Specification - The reactor shall not be operated unless the channels described in Table 3.2.2 are operable and the information is displayed on the reactor console. Table 3.2.2 Required 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,A Level Safety MW or less Channel
- b. Linear Power 10 1Automatic D,M,A Channel Power Control
- c. Log Power 10 1Startup D,M,A Channel Control
- d. Fuel Temperature 2 2 2Fuel D,M,A Channel Temperature
- e. Pulse Channel 0 10Measures P,A Pulse NV & NVT
(*)Where: 0 - Channel check during each day's operation M- Channel test monthly A - Channel calibration annually P - 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 is 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 assure that the reactor power level and energy are adequately monitored (SAR Chapter 7, Sections 7.1.2 & 7.1.2.2).
9
the fuel temperature that Chapter assure(SAR is properly d. monitored and The Table 3.2.2. indicated in the reactorchannels fuel temperature control room 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 in Table 3.2.3 are operable: Table 3.2.3 Required Scrams and Interlocks Steady Square Channel Surveillance State Wave Function Requirements* Scram. Pulse
- a. Console 1 1 Manual Scram M Manual and Automatic Scram 1 Scram Alarm
- b. Reactor Room Manual Scram M Manual Scram and Automatic Scram Alarm
- c. Radiography 4 Manual Scrams M 4
Bay Manual and Automatic Scrams Scram Alarms
- d. Reactor Power 0 Automatic M 2
Level Safety Scram Alarms & Scrams Scrams 2 at 2.2 MW or less
- e. High Voltage Automatic M 2 Scram Alarms &
Power Supplies Scrams Scrams on Loss of High Voltage to 2 the Reactor Power Level Safety Channels
- f. Fuel Automatic Scram M 2 Alarms & Scrams Temperature Scrams on indicated fuel 2 2 temperature of 750°C or less
- g. Watchdog 2 Automatic Scram M Circuit Alarms & Scrams 10
- h. External 2 2 Automatic M Scrams Scrams and Alarms if an experiment or radiography scram interlock is activated
- i. One Kilowatt Prevents initiation M 0 I Pulse & of a step reactivity Square Wave insertion above a Interlock reactor power level of I KW
- j. Low Source I I Prevents withdrawal M Level Rod of any control rod Withdrawal ifthe log channel Prohibit reads less than 1.5 Interlock times the indicated log channel current level with the neutron source removed from the core
- k. Control Rod I I Prevents simul- M Withdrawal taneous withdrawal Interlock of two or more rods in manual mode I. Magnet I I De-energizes the M Power Key control rod Switch Scram magnets, scram &
alarm (*)Where: M - channel test monthly Basis -
- a. Table 3.2.3. The console manual scram allows rapid shutdown of the reactor from the control room (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 the reactor room.
- c. Table 3.2.3. The radiography bay manual scrams allow rapid shutdown of the reactor from any 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 if the 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 safety channels operate within their intended range as required for proper functioning of the power level 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 fuel temperature exceeds 7500° C, therefore ensuring the safety limit will not be exceeded (SAR Chapter 4, Sections 4.5.4.1 & 4.7.2).
11
- g. Table 3.2.3.
acquisition The watchdog computer circuits properly are functioning assure that (SARtheChapter control 7,system computer Section 7.2). and the data
- 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 into the bays (SAR Chapter 9, Section 9.6). External scrams from experiments, a subset of the external scrams, also assure the integrity of the reactor system, the experiment, the facility, and the 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 above one (1) kilowatt assures that the pulse magnitude will not allow the fuel element temperature to exceed 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 source of 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 two or more control rods thus limiting the reactivity-insertion rate from the control rods in manual mode (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 rod magnets and results in a scram (SAR Chapter 7, Section 7.1.2.5). 3.2.4 Reactor Fuel Elements Applicability - This specification applies to the physical dimensions of the fuel elements as measured 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 fuel elements 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 core if:
- a. In measuring the transverse bend, the bend exceeds 0.125 inch (3.175 mm) over the full length 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 the reactor, during which differential expansion between the fuel and the cladding occurs and the pressure of the gases within the elements increases sharply. The above limits on the allowable distortion of a fuel element correspond to strains that are considerably lower than the strain expected to cause rupturing of a fuel element. Limited operation in the steady state or pulsed mode may be necessary to identify a leaking fuel element especially if the leak is small. 12
3.3 Reactor Coolant Systems Applicability - 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 channels described in Table 3.3 are operable, and the information is displayed locally or in the control room. Table 3.3 REQUIRED WATER SYSTEMS AND INSTRUMENTATION Minimum Measuring Number Surveillance Channel/System Operable Function: Channel/System Requirements*
- a. Primary Coolant 1 For operation of the D,Q,A Core Inlet reactor at 1.5 MW or Temperature higher, alarms on high Monitor heat exchanger outlet temperature of 45 °C (113°F)
- b. Reactor Tank I Alarms ifwater level M Low Water drops below a depth of Monitor 23 feet in the reactor tank
- c. Purification** I Alarms ifthe primary D,M,S Inlet Conduc- coolant water conductivity tivity Monitor is greater than 5 micromhos/cm
- d. Emergency Core I For operation of the reactor D,S Cooling System at 1.5MW or higher, provides water to cool fuel in the event of a Loss of Coolant Accident for a minimum of 3.7 hours at 20 gpm from an appropriate nozzle
(*)Where: D -- channel A channel check during calibration each day's operation annually Q - 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 shall be shutdown. Basis -
- a. Table 3.3. The primary coolant core inlet temperature alarm assures that large power fluctuations will not 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 there is 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 a week 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 System Applicability - 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 the reactor 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 the pressure 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 Coolant 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 room will be reduced due to mixing with exhaust system air (SAR Chapter 9, Section 9.5.1). Pressure in the reactor room will be negative relative to surrounding areas due to air flow patterns created by the reactor room 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 fuel temperature 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 Systems 3.7.1 Monitoring Systems Applicability - This specification applies to the information which shall be available to the reactor operator 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. Specification - The reactor shall not be operated unless the channels described in Table 3.7.1 are operable, the readings are below the alarm setpoints, and the information is displayed in the 14
- 0 control room. The stack and reactor room CAMS shall not be shutdown at the same time during reactor operation.
Table 3.7.1 REQUIRED RADIATION MONITORING INSTRUMENTATION Minimum Measuring Number Channel Surveillance Equipment Operable** Function Requirements*
- a. Facility I Monitors Argon-41 and D,W,A Stack Monitor radioactive particu-lates, and alarms
- b. Reactor Room I Monitors the radiation D,W,A Radiation level in the reactor Monitor room and alarms
- c. Purification I Monitors radiation D,W,A System Radia- level at the demineral-tion Monitor izer station and alarms
- d. Reactor Room I Monitors air from the D,W,A Continuous reactor room for parti-Air Monitor culate and gaseous radioactivity and alarms
(*)Where: D - channel check during each day's operation A - channel calibration annually W - 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 release of radioactive material to the environment (SAR Chapter 11, Section 11.1.1.1.4). The alarm setpoint 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 the operations area.
- b. Table 3.7.1. The reactor room radiation monitor provides information regarding radiation levels in the reactor 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 provides information regarding radioactivity in the primary system cooling water (SAR Chapter 11, Section 11.1.5.4.2) 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 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.
15
3.7.2 Effluents - Arqon-41 Dischargle Limit Applicability - 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 releases of this radionuclide from the UCD/MNRC, and the corresponding annual radiation dose from Argon-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 is specified in Appendix B, Table 2, Column 1 of 10 CFR Part 20. 10 CFR 20.1301 specifies dose limitations in the unrestricted area. 10 CFR 20.1101 specifies a constraint on air emissions of radioactive material to the environment. The SAR Chapter 11, Section 11.1.1.1.4 estimates that the routine Argon-4l releases and the corresponding doses in the unrestricted area will be below these limits. 3.8 Exp~eriments 3.8.1 Reactivity Limits. Applicability - This specification applies to the reactivity limits on experiments installed in specific reactor experiment facilities. Objective - The objective is to assure control of the reactor during the irradiation or handling of experiments 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 (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 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 or 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, and in any other reactor in-core and in-tank irradiation facilities at any given time shall be less than one dollar and ninety-two cents ($1.92) (1.34% A k/k), including the potential reactivity which might 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 single movable 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 limit of $1.75 is not exceeded (SAR Chapter 13, Section 13.2.2.2.1). In addition, limiting the worth of each movable experiment to less than $1.00 will assure that the additional increase in transient 16
- 0 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 secured experiment is its sudden accidental or unplanned removal while the reactor is operating. For such 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 Chapter 10, 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 control over the positioning of samples into and out of the central core region 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 less thanthe 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 the fuel temperature safety limit.
3.8.2 Materials Limit 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 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 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 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 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 total inventory of 1-125 being handled in the 1-125 fume hood at any one time in preparation for shipping shall not exceed 20 curies. An Additonal 1.0 curie of 1-125 (up to 400 millicuries in the form of quality assurance samples and up to 600 millicuries in sealed storage containers) may also be present in the 1-125 fume hood.
- e. Explosive materials in quantities greater than 25 milligrams of TNT equivalent shall not be irradiated 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 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 any radiography bay. The irradiation of explosives in any bay is limited to those 1"7
assemblies where a safety analysis has been performed that shows that there is no damage to the 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 of materials.
- b. The 1.5 curies limitation on iodine 131 through 135 assures that in the event of failure of a fueled experiment leading to total release of the iodine, occupational doses and doses to members 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 is totally released into its respective containment, or if both inventories are simultaneously released into their respective containments, the occupational doses and doses to members of the general public in the unrestricted areas will 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 of 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 any radiography bay external to the reactor tank will not result in damage to the reactor controls or the 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 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 and type of material in the experiment shall be limited such that the airborne radioactivity in the reactor room will not result in exceeding the applicable dose limits in 10 CFR Part 20 in 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 on high radiation levels, at least 10% of the gaseous activity or aerosols produced will escape. 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 by boiling this material can escape only through an undistributed column of water above the core, at least 10% of these vapors can escape.
- c. If a capsule fails and releases material which could damage the reactor fuel or structure by corrosion 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 his designated 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 reactor room or the unrestricted area will result in exceeding the applicable dose limits in 10 CFR Part 20.
- b. These assumptions are used to evaluate the potential airborne radioactivity release due to an experiment 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 be brought to the attention of the UCD/MNRC Director or his designated alternate for review to assure 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 reactor operating 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 time period. Prior to reactor system or measuring channel operation, the surveillance shall be performed for each reactor 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 level monitoring channels. Obiective - The objective is to verify that the maximum power level of the reactor does not exceed 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 a significant change in core configuration occurs, a channel calibration shall be required. Discovery of noncompliance with this specification shall limit reactor operations to that required to perform the surveillance. Basis - The annual power level channel calibration will assure that the indicated reactor power level is correct. 4.1.2 Shutdown Margin and Excess Reactivity 19
- 0 Applicability - These specifications apply to the surveillance requirements for reactivity control of the reactor core.
Objective - The objective is to measure and verify the reactivity worth, performance, and operability 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 shutdown margin 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 required to perform the surveillance. Basis -
- a. The reactivity worth of the control rods is measured to assure that the required shutdown margin is available and to provide an accurate means for determining the excess reactivity of the core. Past experience with similar reactors gives assurance that measurements of the 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 been made.
- 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 change unexpectedly.
4.2 Reactor Control and Safety Systems 4.2.1 Control Rods Applicability - This specification applies to the surveillance of the control rods. Objective - The objective is to inspect the physical condition of the reactor control rods and establish the operable condition of the rods. Specification - Control rod worths shall be determined annually or after physical removal or any significant change in core or control rod configuration.
- a. Each control rod shall be inspected at annual intervals by visual observation of the fueled sections 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 required to perform the surveillance. Basis (Technical Specifications 4.2.1 .a-b) - Annual determination of control rod worths or measurements after any physical removal or significant change in core loading or control rod 20
configuration provides information worths. The frequency about of inspection changes for the controlinrods reactor shalltotal reactivity provide and verification periodic individual rod of the condition of the control rod assemblies. The specification intervals for scram time assure operable 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 fuel temperature 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 required to perform the surveillance. Basis - 21
- a. A daily channel check and monthly test, plus the annual calibration, will assure that the reactor power level safety channels operate properly.
- b. A channel test monthly of the reactor power level multi-range channel will assure that the channel is operable and responds correctly. The channel check will assure that the reactor power 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 is operable and responds correctly. A channel check of the reactor power level wide range log channel will assure that the channel is operable on a daily basis. A channel calibration will assure 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, will assure that the fuel temperature measuring channels operate properly.
- e. A channel test prior to pulsing plus the annual channel calibration will assure the pulse energy integrating 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 perform the surveillance. Basis - 22
0
- a. 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 is operable.
- c. A channel test monthly of the Radiography Bay Manual Scrams will assure that the scrams are operable.
- d. A channel test monthly of the Reactor Power Level Safety Scrams will assure that the scrams are operable.
- e. A channel test monthly of the Loss-of-High-Voltage Scram will assure that the high voltage power supplies are operable and respond correctly.
- f. A channel test monthly of the Fuel Temperature Scrams will assure that the scrams are operable.
- g. A channel test monthly of the Watchdog Circuits Scrams will assure that the scram circuits are operable.
- h. A channel test monthly of the External Scrams will assure that the scrams are operable and respond 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 assure that the interlock is operable.
- k. A channel test monthly of the Control Rod Withdrawal Interlock will assure that the interlock is operable.
I. A channel test monthly of the Magnet Current Key Switch will assure that the scram is operable. 4.2.4 Reactor Fuel Elements Applicability - 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 of instrumented fuel elements), and all elements adjacent to the central irradiation facility shall be inspected annually.
- b. Instrumented fuel elements shall be inspected ifany of the elements adjacent to it fail to pass the visual and/or physical measurement requirements of Section 3.2.4. Discovery of noncompliance with Technical Specification 4.2.4 shall limit operations to that required to perform the surveillance.
Basis (Technical Specifications 4.2.4.a-b) - The above specifications assure that the lead fuel elements shall be inspected regularly and the integrity of the lead fuel elements shall be maintained. These are the fuel elements with the highest power density as analyzed in the SAR Chapter 4, Section 4.5.5.6. The instrumented fuel element is excluded to reduce the risk of damage to the thermocouples. 23
.0*
4.3 Reactor Coolant Systems Applicability - 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, the tank water level alarm, the water conductivity cells and the emergency core cooling system are all operable. 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 to perform the 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 input signal. 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 input 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 for power levels above 1.5 MW. A channel calibration semiannually assures that the Emergency Core Cooling System performs as required for power levels above 1.5 MW.
24
0 4.4 Reactor Room Exhaust System Applicability - This specification applies to the surveillance requirements for the reactor room exhaust 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's operation. 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 verify that the exhaust system is maintaining a negative pressure in the reactor room relative to the surrounding facility areas. 4.5 This section intentionally left blank 4.6 This section intentionally left blank. 4.7 Reactor Radiation Monitoring Systems Applicability - 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 to perform the surveillance. Basis -
- a. A channel check of the facility stack monitor system during each day's operation will assure the monitor is operable. A channel test weekly will assure that the system responds correctly to a 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 assure that the monitor is operable. A channel test weekly will ensure that the system responds to a known source. 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 assures that the monitor is operable. A channel test weekly will ensure that the system responds to a known source. 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 responds correctly to a known source. A channel calibration annually will assure that the CAM reads 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 may damage the 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 a written safety analysis has been performed and reviewed by the UCD/MNRC Director, or his designee, 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 Facility Use Authorization. Facility Use Authorizations and experiments carried out under these authorizations 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/MNRC experiment facility until it has been reviewed for compliance with the approved experiment and Facility Use Authorization 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 other UCD/MNRC reactor in-core or in-tank irradiation facility shall be estimated or measured, as 26
0I! 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 in any UCD/MNRC reactor experiment facility.
Basis - a & b. Experience at most TRIGA reactor facilities verifies the importance of reactor staff and safety committee reviews of proposed experiments.
- c. Measurement of the reactivity worth of an experiment, or estimation of the reactivity worth based 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 will facilitate maintaining surveillance over such experiments.
5.0 Design Features 5.1 Site and Facility Description. 5.1.1 Sit__e Applicability - This specification applies to the UCD/MNRC site location and specific facility design 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 downtown Sacramento, 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. The unrestricted 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 has 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 of the 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 reactor building. This accident was analyzed in Chapter 13 of the Safety Analysis Report, and found to be beyond a credible accident scenario.
27
0 0 5.1.2 Facility Exhaust Applicability - This specification applies to the facility which houses the reactor. Objective - 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 proper removal of heat from the reactor room. Specification -
- a. The UCD/MNRC reactor facility shall be equipped with a system designed to filter and exhaust air from the UCD/MNRC facility. The system shall have an exhaust stack height of a minimum of 18.2m (60 feet) above ground level.
- b. Manually activated shutdown controls for the exhaust system shall be located in the reactor control room.
Basis - The UCD/MNRC facility exhaust system is designed such that the reactor room shall be maintained at a negative pressure with respect to the surrounding areas. The free air volume within the 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 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 to operating personnel. 5.2 Reactor Coolant System Applicability - This specification applies to the reactor coolant system. Obiective - The objective is to assure that adequate water is available for cooling and shielding during normal 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 ifthe water level in the reactor tank drops below a depth of 23 ft.
- c. For operations at 1.5 MW or higher during a Loss of Coolant Accident the reactor core shall be cooled for 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 be exceeded 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 and 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 28
- 0 5.3.1 Reactor Core Applicability - This specification applies to the configuration of the fuel.
Objective - The objective is to assure that provisions are made to restrict the arrangement of fuel elements 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 of 96 or more fuel elements to include fuel followed control rods. Below 0.5 MW there is no minimum 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 be loaded into corner positions ony and graphite dummy elements in the flat positions. The performance of fuel temperature measurements shall apply to variations to the as-analyzed 20E core 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 done before 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 above loading restrictions will be allowed in an operational 0.5 MW or greater core. Specifications for the 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 Argonne National Laboratory Report AN L/ED 97-54). 5.3.2 Reactor Fuel Applicability - 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 their physical and nuclear characteristics. 29
0 0 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 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 limiting operation to 2.3 megawatts steady state or to a 36 megawatt-sec pulse assures an ample margin of safety between the maximum temperature generated in the fuel and the safety limit for fuel 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, is equal to the clad yield strength when both fuel and cladding temperature are at the safety limit 9300 C. Since the fuel temperatures are not expected to exceed 630 0C during any condition of normal operation, 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 of purchased materials.
5.3.3 Control Rods and Control Rod Drives Applicability - This Specification applies to the control rods and control rod drives used in the reactor core. Objective - The objective is to assure the control rods and control rod drives are of such a design 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 and regulating rods shall have fuel followers sealed in stainless steel. The transient rod shall have an air 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 assure mechanical stability during movement and to isolate the neutron poison from the tank water environment. 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 at the 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.
30
- 0 5.4 Fissionable Material Storage Applicability - This specification applies to the storage of reactor fuel at a time when it is not in the reactor core.
Objective - The objective is to assure that the fuel which is being stored will not become critical and will not 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 the keff 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 Controls 6.1 Organization. The Vice Chancellor for Research shall be the licensee for the UCD/MNRC. The UCD/MNRC facility shall be under the direct control of the UCD/MNRC Director or a licensed senior reactor operator (SRO) designated by the UCD/MNRC Director to be in direct control. The UCD/MNRC Director shall be accountable to the Vice Chancellor of the Office of Research for the safe operation and maintenance of the reactor and its associated equipment. 6.1.1 Structure. The management for operation of the UCD/MNRC facility shall consist of the organizational structure as shown in Figure 6.1. 6.1.2 Responsibilities. The UCD/MNRC Director shall be accountable to the Vice Chancellor of the 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 all experiments and experiment procedures prior to their use in the reactor. Individuals in the management 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, and shall be responsible for safeguarding the public and facility personnel from undue radiation exposures and for adhering to the operating license and technical specifications. The Site Manager shall report directly to the UCD/MNRC Director. The Reactor Manager and Health Physics 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 reactor operator should be within thirty (30) minutes of the facility and reachable by telephone, and;
- d. A senior reactor operator shall be present whenever a reactor startup is performed, 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 be available 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 and Training of Personnel for Research Reactors (ANS 15.4). Qualification and requalification of licensed operators shall be subject to an approved Nuclear Regulatory Commission (NRC) program. 6.2 Review. Audit. Recommendation and Approval General Policy. Nuclear facilities shall be designed, constructed, operated, and maintained in such 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 with applicable regulatory requirements. The UCO Vice Chancellor of the Office of 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 NSC Composition and Qualifications. The UCD/MNRC Director shall appoint the Chairperson 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. The NSC shall evaluate and review nuclear safety associated with the operation and use of the UCD/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 and proposed Class 1 modifications, prior to implementing (Class I) modifications, prior to taking action under the preceding 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 an unreviewed safety question or require a change to the license, to a Facility Use Authorization, or 32
- 0 to the Technical Specifications. This determination may be in the form of verifying a decision already made by the UCD/MNRC Director.
- d. Review reactor operations and operational maintenance, Class I modification records, and 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 for UCD/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 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 other inspections of these facilities conducted by other agencies.
6.2.4 Audit/Inspection Function. The NSC or a subcommittee thereof, shall audit/inspect reactor operations 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 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 license and Technical 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 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 the UCD/MNRC Radiation Safety Program. The program should use the guidelines of the American National Standard for Radiation Protection at Research Reactor Facilities (ANSI/ANS 15.11). The Health Physics Manager shall report to the Site 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 to assure that procedures are current. Procedures shall be adequate to assure the safe operation of the 33
0 i 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 or other routine maintenance that could have an effect on reactor safety.
- e. Testing and calibration of reactor instrumentation and controls, control rods and control rod drives.
- 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 alarms 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 solid and liquid waste and gaseous effluents released from the facility. The program shall include a management 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 items from 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 Approval. Experiments having similar characteristics are grouped together for review and approval under specific Facility Use Authorizations. All specific experiments to be 34
- 0 performed under the provisions of an approved Facility Use Authorization shall be approved by the UCD/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 approval as a new experiment.
- c. Minor changes to an experiment that do not significantly alter the experiment may be approved by a senior reactor operator.
6.6 Required Actions 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 the UCD/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 and contributing factors. (2) Effect of the violation upon reactor facility components, systems, or structures, and on 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 the NRC 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 is necessary to shut down the reactor to correct the occurrence, operations shall not be resumed unless 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 the NRC Operations Center within 24 hours of the occurrence. A written preliminary report shall be sent 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 the above address within 30 days of the occurrence.
- d. Reportable occurrences should be reviewed by the NSC prior to forwarding any written report to the Vice Chancellor of the Office of Research or to the Nuclear Regulatory Commission.
35
- 0 6.7 Reports. All written reports shall be sent within the prescribed interval to the NRC, Attn: Document Control 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 during the previous calendar year shall be submitted within six months following the end of each calendar year. Each annual report shall include the following information:
- a. A brief summary of operating experiences including experiments performed, changes in facility design, 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 reactor was critical, and the cumulative total energy output since initial criticality.
- c. The number of emergency shutdowns and inadvertent scrams, including reasons for the shutdowns or scrams.
- d. Discussion of the major maintenance operations performed during the period, including the effect, 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 or in procedures, and of tests and experiments carried out pursuant to Section 50.59 of 10 CFR Part 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 as follows: 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 is greater than 1xl0 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 release occurs, 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) released during 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 0 I 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 during the reporting period based on data from an appropriate monitoring system. 3 The estimated maximum annual average concentration of Argon-41 in the unrestricted area (in microcuries/ml), the estimated corresponding annual radiation dose at this location (in millirem), and the fraction of the applicable 10 CFR 20 limits for these values. 4 The total estimated quantity of radioactivity in particulate form with half lives greater than eight days (in curies) released during the reporting 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, by facility users, and by visitors in terms of the average radiation exposure per individual and the greatest exposure per individual in each group.
- h. An annual summary of the radiation levels and levels of contamination observed during routine 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 planned administrative 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 in Section 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 could render the system incapable of performing its intended safety function unless the failure is 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 a condition 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 through 6.7.2.a.8. (2) The written report (and, to the extent possible, the preliminary telephone report or report by similar conveyance) shall describe, analyze, and evaluate safety implications, and outline 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.
(1) Any significant variation of measured values from a corresponding predicted or previously 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 the Safety Analysis Report (SAR); (3) A personnel change involving the positions of UCD/MNRC Director or UCD Vice Chancellor 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 or development 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 required information may be contained in single or multiple records, or a combination thereof. Records and logs shall 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 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 shipped off site.
38
0 0
- 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.
39
... NUCLEARcoMSSoREGUALTORY UNIVERSITY OF CALIFORNIA - DAVIS VICE CHANCELLOR FOR RESEARCH (Licensee) UCD/MNRC UCDIMNRC [ DIRECTOR NUCLEAR ._._ l COMMITJTEE [ 8At-SITE , ' MANAGERi HEALTH PHYFIGUREAC6.1
- EGu Att _; UNITED STATES "
. ** * * *NUCLEAR REGULATORY COMMISSION
*,.,**. *WASHINGTON, D*.C. 20555-0001 March 30, 2004
- Dr. Barry M. Klein Vice Chancellor for Research University of California, Davis One Shields Avenue Davis, CA 95616-8558
SUBJECT:
REVISION TO SAFETY EVALUATION 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 revision to the Safety Evaluation 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 issued on December 30, 2003 and is available on the Commission's ADAMS system, Accession Number ML033421339. Sincerely, _
- WreJ EeIn . rjc aae Research and Test Reactors Section New, Research and Test Reactors Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No. 50-607
Enclosure:
Revision to Amendment No. 7 Safety Evaluation Report cc w/enclosure: Please see next page
0.. University of California - Davis/McClellan MNRC Docket No. 50-607 cc: Mr. Jeff Ching 5335 Price Avei~ue, Bldg. 258 McClellan AFB, CA 95652-2504 Test, Research, and Training Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611
0 UNITEb STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 REVISION TO SAFETY EVALUATION REPORT SUPPORTING AMENDMENT NO. 7 TO AMENDED .FACILITY OPERATING LICENSE NO. R-130 REGENTS OF THE UNIVERSITY OF CALIFORNIA AT McCLELLAN NUCLEAR RADIATION CENTER DOCKET N.O. 50-607
1.0 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 the McClellan Nuclear Radiation Center (MNRC) TRIGA research reactor. The request provided for the 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 Facility Operating License be amended to include an additional section 2.B.(4) as follows: 2.B.(4) Inaddition to those items specified in2.B.(1), 2.B.(2) and 2.B.(3) the following radioactive materials may be received, possessed, and used at the facility. Radioactive Material Chemical and/or Maximum (element and mass number) Physical Form. Possess atQuantity Any OneLicensee Time May A. Any radioactive material A. Any A. 20 curies (1 curie each, except as between atomic number I through provided below) 83, Inclusive B. Any radioactive material with A. Any A. 4 Curies (100 milllcuries each, atomic numbers 84 and above except as provided below) or up to 20 micrograms c.. Iodine-125 c. Iodide/LIquid C. 40 Curies D. Source material (but only trace D. Any D. 4 grams per radionuclide, not to amounts of Th-234) exceed 10 grams total E. Special nuclear material E. Any E. 2 grams per radionuclide, not to exceed 5 grams total This amendment request was approved and issued on .December 30, 2003.
0 " 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 the reactor room glove box. The significance of this assumption is related to the ability of the reactor 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 the above table is 64 curies. The maximum activity In category D is about 0.1 curie, while the maximum 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 the bounding MHA, and the amendment request was approved. Instead of locating all of the radioactive materials shown in above table in the reactor room glove box, some of the materials will be located in the restricted area of the McClellan Nuclear Radiation Center. Non-volatile material will be handled in accordance with approved procedures. 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 the consequences of complete volatilization of the unsealed material (e.g., the reactor room glove box and reactor room fume hood), as previously analyzed. The staff has reviewed the proposed change to the Facility Operating License and concluded that itdoes not impact the licensee's ability to continue to meet the relevant requirements of 10 CFR 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 and surveillance requirements. The staff has determined that this amendment involves no significant 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 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 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 or consequences of accidents previously evaluated, or create the possibility of a new or different kind of accident from 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 is reasonable 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 or the health and safety of the public. Principal Contributor: Warren J. Ereslan Date: March 30, 2004
'5%. i ./ uNITED STATES
*/*"* NLCLEAR REGULATORY COMMISSION 0 ASIGTNDC.205-00 Deceeiber: 30, 2003 Dr. Barry M.Klein Vice Chancellor for Research University of California, Davis One Shields Avenue Davis, 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 to Facility 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 is discussed in the enclosed Safety Evaluation Report. Sincerely,*,*, 69 ~4~tey/ Warren J. Eresian, Project Manager Research and Test Reactors Section New, Research and Test Reactors Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No. 50-607
Enclosures:
- 1. Amendment No. 7
- 2. Safety Evaluation Report
University of California - Davis/McClellan MNRC Docket No. 50-607 cc: Dr. Wade J. Richards 5335 Price Avenue, Bldg. 258 McClellan AFB, CA 95652-2504 Test, Research, and Training Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611
UNITED STATES
"*%- NUCLEAR REGULATORY COMMISSION
- WASHINGTON, D.C. 20555-0001 REGENTS OF THE UNIVERSITY OF CALIFORNIA AT McCLELLAN NUCLEAR RADIATION CENTER DOCKET NO. 50-607 AMENDMENT TO AMENDED FACILITY OPERATING LICENSE Amendment No. 7 License No. R-1 30
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that A. 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 Nuclear Radiation Center (the licensee) on October 21, 2003 and November 6, 2003, conforms to the standards and requirements of the Atomic Energy Act of 1954:, as amended (the Act), and the regulations of the Commission as stated in Chapter I of Title 10 of the Code of FederalRegulations (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 the Commission; D. The issuance of this amendment will not be Inimical to the common defense and security 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; and F. 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. Accordingly, the license is amended by changes to the Facility Operating License as indicated below, and paragraph 2.B of Amended Facility Operating License No. R-130 is hereby amended to read as follows:
2.B.(4) Inaddition 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 Chemical and/or Maximum Quantity (element and mass Physical Form May Possess at AnyLicensee One Time number) A. Any radioactive A. Any A. A. 20 Curies (I Curie each, material between except as provided below) atomic number 1 through 83, inclusive B. Any radioactive A. Any A. 4 Curies (100 millicuries material with atomic each, except as provided numbers 84 and below) or up to 20 above micrograms C. Iodine-125 C. 40OCuries C. Iodide/Liquid
- 0. Any D. 4 grams per radionuclide, D. Source material (but only trace amounts not to exceed 10 grams of Th-234) total E. Special nuclear E. Any E. 2 grams per radionuclide, material not to exceed 5 grams total
- 3. This license amendment is effective as of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Research and Test Reactors Section New, Research and Test Reactors Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Date of Issuance: December 30, 2003
~UNITED O STATES 0
o* NUCLEAR REGULATORY COMMISSION o~WASHINGTON, D.C. 20555-0001 o# SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 7 TO AMENDED FACILITY OPERATING LICENSE NO. R-130 REGENTS OF THE UNIVERSITY OF CALIFORNIA AT McCLELLAN NUCLEAR RADIATION CENTER DOCKET NO. 50-607
1.0 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 the allowance of radioactive materials not produced by the reactor to be received, possessed and used 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 Chemical and/or Maximum Quantity Licensee May (element and mass Physical Form Possess at Any One Time number) A. Any radioactive A. Any A. A. 20 Curies (1 Curie each, material between except as provided below) atomic number 1 through 83, inclusive B. Any radioactive A. 4 Curies (100 mlllicuries A. Any material with atomic each, except as provided numbers 84 and below) or up to 20 above micrograms C. Iodine-I125 C. Iodide/Liquid C. 40 Curies D. Any D. 4 grams per radionuclide, D. Source material (but only trace amounts of not to exceed 10 grams total Th-234) E. Special nuclear E. Any E. 2 grams per radionuclide, material not to exceed 5 grams total
- 0
-2,-
This request is discussed below. 2.0 EVALUATION All of the radioactive materials to be received, possessed and handled In accordance with this amendment request will be located in the reactor room glove box. In November of 2002, the NRC approved 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 reactor 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 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 reactor room for the maximum assumed occupancy time of 5 minutes the CEDE to the thyroid would be about 205 millirem. These doses were compared to the expected doses (CEDE) resulting from the Maximum Hypothetical Accident (MHA), which serves as the bounding accident for radiological 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 the complete 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 box from 61 curies to 64.4 curies. (The total activity in categories A, B, and C In the above table Is 64 curles. The maximum activity in category D corresponds to 10 grams of Uranium-233, or about 0.1 curie. The maximum activity in category E corresponds to 5 grams of Plutonium-239, or about 0.3 curie.) For the complete volatilization of 64.4 curies, doses in the unrestricted area and in the reactor room will scale up proportionally from 61 curies, resulting in a dose in the unrestricted area of 3.2 mlllirem, and a dose in the reactor room of 216 millirem (i.e., the doses have increased by 5.6 percent.) The staff concludes that the consequences of the complete volatilization of 64.4 curies are much less than the consequences of the bounding MHA, and that increasing the allowable activity in the reactor room glove box from 61 curies to 64.4 curies does not significantly reduce the margin of safety 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 that it does not impact the licensee's ability to continue to meet the relevant requirements of 10 CFR Part 50.36.
3.0 ENVIRONMENTAL CONSIDERATION
This amendment does not involve changes in the installation or use of a facility component located within 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 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 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 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 or consequences of accidents previously evaluated, or create the possibility of a new or different kind of accident from 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 is reasonable 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 or the health and safety of the public. Principal Contributor: Warred J. Eresian Date: December 30, 2003
*NUCLEAR REGULATORY COMMISSION ,,.. * ,* UNITED D.C.
WASHINGTON, STATES 20555-0001
,* February 17, 2000 *i7/tlJ*
Brigadier General Michael P. Wiedemer Vice Chancellor Kevin Smith Commander Office of the Chancellor .. ~Sacramento Air Logistics Center University of California, Davis SM-ALCITI-1 One Shields Avenue 5335 Price Avenue Davis, California 95616-8558 McClellan AFB, California 95652-2504
SUBJECT:
RE-ISSUANCE OF NOTICE OF CONSIDERATION OF APPROVAL OF TRANSFER OF FACILITY OPERATING LICENSE NO. R-130 FOR THE McCLELLAN NUCLEAR RADIATION CENTER FROM THE DEPARTMENT OF THE 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 some administrative errors. We. apologize for any inconvenience this may have caused. Sincerely, Ledyard B. Marsh, Chief Events Assessments, Generic Communications and Non-Power Reactors Branch Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No. 50-607
Enclosure:
As stated cc: wlenclosures
McClellan AFB TRIGA REACTORDcktN.0-7 CC: Dr. Wade J. Richards SM-ALC/TI-1 5335 Price Avenue, Bldg. 258 McClellan AFB, CA 95652-2504 Cot. Robert Capell HQ AFMC/SGC 4225 Logistics Avenue, Suite 23 Wright-Patterson AFB, OH 45433-5762 Lt. Col. Catherine Ze~ringue HQ AFSCISEW 9570 Avenue G, Building 24499 Kirtland AFB, New Mexico 871 17-5670 Test, Research, and Training Reactor Newsletter 202 Nuclear Sciences Center University of Florida Gainesville, FL 3261 1
- L0 TECHNICAL SPECIFICATIONS FOR THE UNIVERSITY OF CALIFORNIA - DAVIS MCCLELLAN NUCLEAR RADIATION CENTER (UCDIMNRC)
DOCUMENT NUMBER: MNRC-0004-DOC-13 Rev 13 4/03 p
.~.
0 ! Revision ,13 of the "Technical Specifications" for the University of California at Davis/McClellan Nuclear Radiation ~.1 *>~ Center (UCD/MNRC) Reactor have undergone the following coordination: Reviewed by: Hel-'* !.0 eltPyiSpesoDate Reviewed by: __________* 0. R ator Su~dpervisrDt Approved by: *~Ij.e/o3
~i'I~toY Chairman, UO*/M NRC ~l1.
Date Nuclear Safety Committee I. (K ~/
-- I
0 Technical Specifications Rev 13 412003 Title Page Rev 13 4/2003 Approval Page Rev 13 4/2003 31 Rev 13 4/2003 32 Rev 13 4/2003 33 Rev 13 4/2003 Figure 6.1 Rev 13 4/2003
......................----.--...
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- 0
* " TECHNICAL SPECIFICATIONS TABLE OF CONTENTS 1.0 Definitions .............................................................................................................. 1 2.0 Safety Limit and Limiting Safety System Setting (LSSS.*........................................................6...
2.1 Safety Limits.................................................................................................. 2.2 Limiting Safety System Setting (LSSS)..................................................................... 6 2.2.1 Fuel Temperature ............................................................ i.................... 6 3.0 Limiting Conditions for Operations (LC.O.) ........................................................................... 7 3.1 Reactor Core Parameters................................................................................... 7 3.1.1 Steady-State Operation ....................................... :................................... 7 3.1.2 Pulse or Square Wave Operation ............................................................... 7 3.1.3 Reactivity Limitation~s............................................................................. 8 3.2 Reactor Control and Safety Systems .... .................................................................. 8 3.2.1 Control Rods...................................................................................... 8 3.2.2 Reactor Instrumentation.......................................................................... 9 3.2.3 Reactor Scrams and Interlocks................................................................. 10 3.2.4 Reactor Fuel Elementts.......................................................................... 12 3.3 Reactor Coolant Systems.................................................................................. 13 3.4 Reactor Room Exhaust System ........................................................................... 14 3.5 Intentionally Left Blank ..................................................................................... 14 3.6 Intentionally Left Blank..................................................................................... 14 3.7 Reactor Radiation Monitoring Systems.................................................................... 14 3.7.1 Monitoring Systems ... ......................................................................... 14 3.7.2 Effluent~s - Argon-41 Discharge Limit. .......................................................... 16 )
0 0 3.8 Experiments ................................................................................................ 16 3.8.1 Reactivity Limits ........................................ *........................................ 16 3.8.2 Materials Limit................................................................................... 17 3.8.3 Failure and Malfunctions ...................... ................................................. 18 4.0 Surveillance Requirements.......................................................................................... 19 4.1 Reactor Core Parameters ................................................................................. 19 4.1.1 Steady State Operation......................................................................... 19 4.1.2 Shutdown Margin and Excess Reactivity ....................................................... 20 4.2 Reactor Control and Safety Systems...................................................................... 20 4.2.1 Control Rods ................................................... ................................. 20 4.2.2 Reactor Instrumentation............................................................... *......... 21 4.2.3 Reactor Scrams and interlocks.................... ............................................. 22 4.2.4 Reactor Fuel Elements................................ .................................... 23 4.3 Reactor Coolant Systems ................................................................................. 24 4.4 Reactor Room Exhaust System ........................................................................... 25 4.5 Intentionally Left Blank..................................................................................... 25 4.6 Intentionally Left Blank..................................................................................... 25 4.7 Reactor Radiation Monitoring Systems.................................................................... 25 4.8 Experiments ................................................................................................ 26 5.0 Design Features ........................................................... ;......................................... 27 5.1 Site and Facility Description ............................................................................... 27 5.1.1 .Site............................................................................................... 27 5.1.2 Facility Exhaust ......................... ....................................................... 28 5.2 Reactor Coolant system ................................................................................... 28
5.3 Reactor Core and F.uel .................................................................................... 29 5.3.1 Reactor Care .................................................................................... 29 5.3.2 Reactor F..u~l..................................................................................... 30 5.3.3 Control Rods and Control Rod Drives.......................................................... 31 5.4 Fissionable Material Storage............................................................................... 31 6.0 Administrative Controls ............................................................................................ .31 6.1 Organization................................................................................................ 31 6.1.1 Structure......................................................................................... 32 6.1.2 Responsibilities ................................................................................. 32 6.1.3 Staffing .......................................................................................... 32 6.1.4 Selection and Training of Personnel ........................................................... 32 6.2 Review, Audit, Recommendation and Approvial........................................................... 32 6.2.1 NSC Composition and Qualifications........................................................... 33 6.2.2 NSC Charter and Rules .......................... i.............................................. 33 I 6.2.3 Review Function................................................................................. 33 6.2.4 Audit/Inspection Function ....................................................................... 34 6.3 Radiation Safety............................................................................................ 34 1 6.4 Procedures ................................................................................................. 34 6.4.1 Reactor Operations Procedur~es................................................................. 34 I 6.4.2 Health Physics Procedures ..................................................................... 35 6.5 Experiment Review and Approlval ......................................................................... 35 6.6 Required Actions ........................................................................................... 35 6.6.1 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................................................................................. 36 6.7.2 Special Reports ................................................................................... 38 6.8 Records........................................................................................................ 39 Fig. 6.1 UCD/MNRC Organization for Licensing and Operation........................................................ 40
- /
- 0 TECHNICAL SPECIFICATIONS FOR THE UNIVERSITY OF CALIFORNIA - DAVIS/MCCLELLAN NUCLEAR RADIATION CENTER (UCDIMNRC)
The University of California - Davis/McClellan Nuclear Radiation Center (UCD/MNRC) reactor is operated by the University 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 provides a wide range of irradiation services for both research and industrial needs. The reactor operates at a nominal steady state power level up to and including 2 MW. The UCD/MNRC reactor is also capable of square wave and 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 Nuclear Regulatory Commission (e.g., senior reactor operator or reactor operator) to carry out the duties and responsibilities associated with the position requiring the license. 1.2.1 Sernior Reactor Operator. An individual who is licensed to direct the activities of reactor operators 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 facility and perform reactor-related maintenance. 1.3 Channel. A channel is the combination of sensor, line amplifier, processor, and output devices which are connected for the purpose of measuring the value of a parameter. 1.3.1 Channel Test. Achannel 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 is*n adjustment of the channel such that its output corresponds with acceptable accuracy to known-'-values of the parameter which the channel measures. Calibration shall encompass the entire channel,.including equipment actuation, alarm or trip, and shall be deemed to include a channel test. " 1.3.3 Channel. Check. A channel check is a qualitative verification of acceptable performance by observation of channel behavior. This verification, where possible, shall include comparison of the channel with other independent channels or systems measuring the same variable. 1.4 Confinement. Confinement means isolation of the reactor room air volume such that the movement of 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, fission chambers, foils, etc), which is designed to investigate specific reactor characteristics or which is intended for irradiation within an experiment facility and which is not rigidly secured to a core or shield structure 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 entire experiment 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, or component 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 are normal to the operating environment of the experiment, or by forces which can arise as a result of 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 their associated input circuits, which are designed to initiate a scram for the primary purpose of protecting an experiment or to provide information which requires manual protective action to be initiated. ' 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 can be either 8.5, 20, or 30 wt%, with an enrichment of less than 20% U-235. A standard fuel element may contain a burnable poison. 1.7 Fuel Element., Instrumented. An instrumented fuel element is a standard fuel element fabricated with thermocouples for temperature measurements. An instrumented fuel element shell have at least one operable 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 a channel. 1.9 Mode. Steady-State. Steady-state mode operation shall mean operation of the UCDIMNRC reactor with 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/MNRC reactor 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 the selector 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 within the UCDIMNRC reactor safety system in response to a variable or condition of the UCDIMNRC reactor facility having 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 the generation 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 i: equipment. 2
1.15.3 Instrument generation System Level. At the protective instrument level, protective action is the 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 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 the maximum allowable pulse reactivity insertion has been determined. 1; 17 Reactivity. Exce~ss, Excess reactivity is that amount of reactivity that would exist ifall control rods (control, regulating, etc.) were moved to the maximum reactive position from the point where the reactor is 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 the reactor core. 1.19 R~eactivity Worth of an Exoeriment. The reactivity worth of an experiment is the maximum value of the reactivity 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 which directly 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 the parameters 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 in these 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 for initiation of manual protective action. 1.24 R~eactor Secured. The UCO/MNRC reactor is secured when the console key switch is in the off position and the key is removed from the lock and under the control of a licensed operator, and the conditions of a or b exist:
- a. (1) The minimum number of control rods are fully inserted to ensure the reactor is shutdown, as required by 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 being moved or serviced ifthe experiments have, on movement, a reactivity worth exceeding the maximum value allowed for a single experiment or $1.00, whichever is smaller, or
- b. The reactor contains insufficient fissile materials in the reactor core, a~djacent experiments or control rods to attain criticality under optimum available conditions of moderation and reflection.
1.25 Reactor Shut~down. The UCD/MNRC reactor is shutdown ifit 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
- 0 1.26 Reference Cpre Condition. The condition of the core when it is at ambient temperature (cold T<280 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 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, and which 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 a fuel 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 allow it to perform its control function, and its safety function when the coupling is disengaged. 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 power level and 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 of radioactivity. 1.31 Scram Time. Scram time is the elapsed time between reaching a limiting safety system set point and the 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 denote a recommendation; the word "may" to denote, permission, neither a requirement nor a recommendation. 1.34 Shut~down Margin. Shutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence that the reactor can be made subcritical by means of the control and safety system starting from any permissible operating condition with the most reactive rod assumed to be in the most reactive position, and once this action has been initiated, the reactor will remain subcritical without further operator action. 1.35 Shutdown. Unsched.u.led. An unscheduled shutdown is any unplanned shutdown of the UCD/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 safe operation, 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, weekly or quarterly. Calibration times are generally specified as quarterly, semi-annually, annually, or biennially. ...../) 1.37 to Surveillance reduce Intervals. frequency. Maximum Established intervalsshall frequencies are established to provide be maintained over theoperational long term. flexibility and not The allowable 4
0 0*** surveillance interval is the interval between a check, test, or calibration, whichever is appropriate to the item being 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 involve an unreviewed safety question:
- a. If the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or
- b. If a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or
- c. 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 reactor computerized 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 fuel element. 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 during steady-state operation.
- b. The maximum ten'perature in a standard TRIGA fuel element shall not exceed 1100 0C during pulse operation.
....... 'a. This fuel safety limit applies for coniditions in which the cladding temperature is above 500 °C (Safety Analysis Report (SAR), Chapter 4, Section 4.5.4.1.3). The important parameter for a TRIGA reactor is 5
0 =° the fuel element temperature. This parameter is well suited as it can be measured directly. A loss in the integrity of the fuel element cladding could arise ifthe cladding stress exceeds the ultimate strength of the cladding material. The fuel element cladding stress is a function of the element's internal pressure while the ultimate strength of the cladding material is a function of its temperature. The cladding stress is a result of the internal pressure due to the presence of air, fission product gasses and hydrogen from the disassociation 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 hydrogen to zirconium in the alloy. At a fuel temperature of 930 0C for ZrH 1.7 fuel, the cladding stress due to the internal pressure is equal to the ultimate strength of the cladding material at the same temperature (SAR Fig 4.18). This is a conservative limit since the temperature of the cladding material is always 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 a pulse which gives a peak adiabatic fuel temperature of 1000 °C is estimated to be 470 °C. Further analysis (SAR Section 4.5.4.1.2), shows that the internal pressure for both Zr *.s (at 1150 0 C) and Zr 17z (at 11 00°C) increases to a peak value at about 0.3 sec, at which time the pressure is about one-fifth of the equilibrium value or about 400 psi (a stress of 14,700 psi). The yield strength of the cladding at 500 0C is about 59,000 psi. Calculations for step increases in power to peak ZrH 1.85 fuel temperature greater than I115 0 C, over a 200°C range, show that the time to reach the peak pressure and the fraction of equilibrium pressure value achieved were 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 and compared 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 (ZrH 1.=) was only about 6% of the equilibrium value evaluated at the peak temperature. Calculations of the pressure gave values about three times greater than the measured values. The analysis gives strong indications that the cladding will not rupture iffuel 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 for cladding failure. 2.2 Limiting Safety System Setting. 2.2.1 Fuel Temperature. Applicability - This specification applies to the protective action for the reactor fuel element temperature. Obiective - The objective is to prevent the fuel element temperature safety limit from being reached. Specification - The limiting safety system setting shall be 750 0C (operationally this may be set more conservatively) as measured in an instrumented fuel element. One instrumented element shall 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 a temperature which, ifexceeded, shall cause a reactor scram to be initiated preventing the safety limit from being exceeded. A setting of 750 °C provides a safety margin at the point of the measuremenrA 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 measured temperatures 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 the center and mid-plane of the fuel element. For pulse operation of the reactor, the same limiting safety system setting applies. However, the temperature channel will have no effect on limiting 6
O ...... 0 the peak power generated because of its relatively long time constant (seconds) as compared with the width of the pulse (milliseconds). in this mode, however, the temperature trip will act to
"\ limit the energy release after the pulse ifthe transient rod should not reinsert and the fuel ! temperature continues to increase.
3.0 Limiting Conditions For Operat~ion* 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 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. 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. For the purpose 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/MNRC TRIGA 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 Aoolicabilitv - This specification applies to the peak temperature generated in the fuel as the result of a step insertion of reactivity.
Obiective - The objective isto 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 to 1.7. 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 state operation at two (2) MW the hydrogen will redistribute due to migration from the central high temperature regions of 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 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. Ifthe pulse insertion is such that the temperature of the fuel exceeds about 875 0C, then the pressure may be sufficient to cause expansion of microscopic 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 the reactivity 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 at all 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 by the 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) shall not exceed 6.65% Ak/k ($9.50).
Basis -
- a. This specification assures that the reactor can be placed in a shutdown condition from any operating condition and remain shutdown, even if the maximum worth control rod should stick in the fully withdrawn position (SAR Chapter 4, Section 4.5.5).
- b. This specification sets an overall reactivity limit which provides adequate excess reactivity to override the xenon buildup, to overcome the temperature change in going from zero power to 2 MW, to permit pulsing at the $1.75 level, to permit irradiation of negative worth experiments and account for fuel burnup over time. An adequate shutdown margin exists with an excess of $9.50 for the two analyzed cores: (SAR Chapter 4, Section 4.5.5).
3.2 Reactor Control and Safety Systems 3.2.1 Control Rods Aooljcjili*- 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 ifdamage is apparent to the rod or drive assemblies.
- b. The scram' time measured from the instant a signal reaches the value of a limiting safety system setting to the instant that the slowest control rod reaches its fully inserted position shall
. not exceed one (1) second.
8
0 S
- a. The apparent condition of the control rod assemblies shall provide assurance that the rods shall continue to perform reliably as designed.
- b. This assures that the reactor shall shut down promptly when a scram signal is initiated (SAR Chapter 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 reactor operator during reactor operations. Obiective - The objective is to require that sufficient information is available to the operator to assure safe operation of the reactor. Specification - The reactor shall not be operated unless the channels described in Table 3.2.2 are operable and the information is displayed on the reactor console. Table 3.2.2 Required 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,A Level Safety MW or less Channel b5.Linear Power 101Automatic D,M,A Channel -Power Control
- c. Log Power 101Startup D,M,A Channel Control
- d. Fuel Temperature 2 2 2 Fuel D,M,A Channel Temperature
- e. Pulse Channel 0 10Measures P,A Pulse NV & NVTr
(*) Where: D - Channel check during each day's operation M - Channel test monthly A - Channel calibration annually P - Channel test prior to pulsing operation
- a. Table 3,2.2. The two reactor power level safety channels assure that the reactor power level is 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 assure that the reactor power level and energy are adequately monitored (SAR Chapter 7, Sections
.... 7.1.2 & 7.1.2.2). 9
Ia1,.1.2 monitored and The fuel temperature indicated in the reactorchannels assure(SAR control room that Chapter the fuel temperature is properly 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 in Table 3.2.3 are operable: Table 3.2.3 Required Scrams and I!nterlocks Steady Square Channel Surveillance Scram State Pulse Wave Reouirements*
- a. Console I I .1 Manual Scram M Manual and Automatic Scram Scram Alarm
- b. Reactor Room 1 I Manual Scram M Manual Scram and Automatic Scram Alarm 4 4 Manual Scrams M
- c. Radiography Bay Manual 4 and Automatic Scrams Scram Alarms
- d. Reactor Power 2 0 2 Automatic M Level Safety Scram Alarms & Scrams Scrams at 2.2 MW or less
- e. High Voltage 2 1 2 Automatic M Power Supplies Scram Alarms &
Scrams *i Scrams onLosf High Voltage to the Reactor Power Level Safety Channels
- f. Fuel 2 2 2 Automatic Scram M Temperature Alarms & Scrams Scrams on indicated fuel temperature of 750°C or less
- g. Watchdog 2 2 2 Automatic Scram M
.Circuit Alarms & Scrams 10 N.,
0
- h. External 2 2 Automatic M Scrams Scrams and Alarms
\ ifan experiment or radiography scram interlock
*is activated i.One Kilowatt 0 1 Prevents initiation M Pulse & of a step reactivity Square Wave insertion above a Interlock reactor power level of 1 KW
- j. Low Source Prevents withdrawal M 1 1 Level Rod of any control rod Withdrawal if the log channel Prohibit reads less than 1.5 Interlock times the indicated log channel current level with the neutron source removed from the core
- k. Control Rod Prevents simul- M Withdrawal I 1 taneous withdrawal Interlock of two or more rods in manual mode I. Magnet De-energizes the 1 1 1 M Power Key control rod Switch Scram magnets, scram &
alarm (*)Where: M- channel test monthly Basis -
- a. Table 3.2.3. The console manual scram allows rapid shutdown of the reactor from the control room (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 the reactor room.
p,.Table 3.2.3. The radiography bay manual scrams allow rapid shutdown of the reactor from any 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 if the power level exceeds 2.2 MW, therefore not exceeding the safety limit (SAR Chapter 4, Section 4.7.2).
- e. Table 3,.2,*. The loss-of-high-voltage scram assures that the reactor power level safety channels, operate within their intended range as required for proper functioning of the power level 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 ifthe fuel temperature exceeds 7500o C, therefore ensuring the safety limit will not be exceeded (SAR Chapter 4, Sections 4.5.4.1 &4.7.2).
11
- a. Table 3.2.3, acquisition The watchdog computer circuitsproperly are functioning assure that (SARtheChapter control 7, system computer Section 7.2). and the data
- h. Table.3.2.3, The external scrams assure that the reactor will be shut down ifthe radiography bay doors and reactor concrete shutters are not in the proper position for personnel entry into the bays (SAR Chapter 9, Section 9.6). External scrams from experiments, a subset of the external scrams, also assure the integrity of the reactor system, the experiment, the facility, and the 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 above one (1) kilowatt assures that the pulse magnitude will not allow the fuel element temperature to exceed 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 source of 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 two or more control rods thus limiting the reactivity-insertion rate from the control rods in manual mode (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 rod magnets and results in a scram (SAR Chapter 7, Section 7.1.2.5). 3.2.4 Rea~ctor Fuel Elements Aoolicabilitv - This specification applies to the physical dimensions of the fuel elements as measured 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 fuel elements 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 core if:
- a. In measuring the transverse bend, the bend exceeds 0.125 inch (3.175 mam) over the full length 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 the reactor, during which differential expansion between the fuel and the cladding occurs and the pressure of the gases within the elements increases sharply. The above limits on the allowable distortion of a fuel element correspond to strains that are considerably lower than the strain expected to ".ause rupturing of a fuel element. Limited operation in the steady state or pulsed mode may be necessary to identify a leaking fuel element especially if the leak is small. 12
3.3 Reactor Coolant Systems Aoolicability - 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 channels described in Table 3.3 are operable, and the information is displayed locally or in the control room. Table 3.3 REQUIRED II WATER SYSTEMS I n I AND INSTRUMENTATION
* . ...
Minimum Measuring Number Surveillance Chann~el/System Ooe~rable. Requirements* Function: Channel/System
- a. Primary Coolant I For operation of the D,Q,A Core Inlet reactor at 1.5 MW or Temperature higher, alarms on high Monitor heat exchanger outlet temperature of 45 °C (113°F)
- b. Reactor Tank 1 Alarms ifwater level M Low Water drops below a depth of
*Monitor 23 feet in the reactor tank
(.
- c. Purification** 1 Alarms ifthe primary DM,S Inlet Conduc- coolant water conductivity tivity Monitor is greater than 5 micromhos/cm
- d. Emergency Core I For operation of the reactor D,S Cooling System at 1.5MW or higher, provides water to cool fuel in the event of a Lois of Coolant Accident for a minimum of 3.7 hours at 20 gpm from an appropriate nozzle
(*)Where: AD-- channel check during channel calibration each day's operation annually Q - 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 shall be shutdown. Basis - a.Table 3.3, The primary coolant core inlet temperature alarm assures that large power fluctuations will not 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 there is 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 ifthe water level drops
) below aheight 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 a week 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; System Aoplicability - 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 reactor 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 the pressure 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 Coolant 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 room will be reduced due to mixing with exhaust system air (SAR Chapter 9, Section 9.5.1). Pressure in the reactor room will be negative relative to surrounding areas due to air flow patterns created by the reactor room 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 fuel temperature 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 Systems AnoDlicability - This specification applies to the information which shall be available to the reactor operator 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 3.7.1 are operable, the readings are below the alarm setpoints, and the information is displayed in the 14
control reactor room. The stack and reactor room CAMS shall not be shutdown at the same time during operation. Table 3.7.1 REQUIRED RADIATION MONITORING INSTRUMENTATION Minimum Measuring Number Channel Surveillance Eduioment O, erable** Function Requirements*
- a. Facility I Monitors Argon-41 and D,W,A Stack Monitor radioactive particu-lates, and alarms
- b. Reactor Room 1 Monitors the radiation D),W,A Radiation level in the reactor Monitor room and alarms
- c. Purification 1 Monitors radiation D:,W,A System Radia- level at the demineral-tion Monitor izer station and alarms
- d. Reactor Room Monitors air from the D,W.A 1
Continuous reactor room for parti-Air Monitor culate and gaseous radioactivity and alarms (*)Where: D -- channel A check during channel calibration each day'ls operation annually W - 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 the release of radioactive material to the environment (SAR Chapter 11, Section 11.1.1.1.4). The alarm setpoint 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 the operations area.
- b. Table 3.7.1. The reactor room radiation monitor provides information regarding radiation levels in the reactor 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 provides information regarding radioactivity in the primary system cooling water (SAR Chapter 11, Section 11.1.5.4.2) 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 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. 15
3.7.2 .Effluents -.Argon-41 Discharge Limit AppJicaiity~- This specification from the UCD/MNRC applies to the concentration of Argon-41 that may be discharged 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 releases of this radionuclide from the UCD/MNRG, and the corresponding annual radiation dose from Argon-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 is specified in Appendix B, Table 2, Column 1 of 10 CFR Part 20.10 CFR 20.1301 specifies dose limitations in the unrestricted area. 10 CFR 20.1101 specifies a constraint on air emissions of radioactive material to the environment. The SAR Chapter 11, Section 11.1.1.1.4 estimates that the routine Argon-41 releases and the corresponding doses in the unrestricted area will be below these limits. 3.8 Experiments 3.8.1 React~ivity Limnits. Applicability - This specification applies to the reactivity limits on experiments installed in specific reactor experiment facilities. Obiective - The objective is to assure control of the reactor during the irradiation or handling of experiments 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 (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 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 or 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, and in any other reactor in-core and in-tank irradiation facilities at any given time shall be less than one dollar and ninety-two cents ($1.92) (1.34% A k/k), including the potential reactivity which might 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 single movable 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 limit of $1 .75 is not exceeded (SAR Chapter 13, Section 13.2.2.2.1 ). In addition, limiting the worth of each movable experiment to less than $1.00 will assure that the additional increase in transient 16
power Chapter (SAR and temperature will13.2.2.2.1). 13, Section be slow enough so that the fuel temperature scram will be effective
- b. The absolute worst event which may be considered in conjunction with a single secured experiment is its sudden accidental or unplanned removal while the reactor is operating. For such 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 Chapter 10, 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 control over the positioning of samples into and out of the central core region 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 less thanthe 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 the fuel temperature safety limit.
3.8.2 .Materials Limit Aoplicability - 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 limiting material quantity and the radioactive material inventory of the experiment.
.Specification - The reactor shall nct 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 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 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 total inventory of 1-125 being handled in the 1-125 fume hood at any one time in preparation for shipping shall not exceed 20 curies. An Additonal 1.0 curie of 1-125 (up to 400 millicuries in the form of quality assurance samples and up to 600 millicuries in sealed storage containers) may also be present in the 1-125 fume hood.
- e. Explosive .materialsin quantities greater than 25 milligrams of TNT eqluivalent shall not be irradiated in th~e reactor tank. Explosive materials in quantities of 25 milligrams of TNT equivalent or less may be irradiatedl 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 any radiography bay. The irradiation of explosives in any bay is limited to those
.17
...... " 0 assemblies where systems the reactor safety a safety analysis has been(SAR upon detonation performed Chapterthat 13,shows Sectionthat there is no damage to 13.2.6.2). ") Basis -
- a. Appropriate 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 fueled experiment leading to total release of the iodine, occupational doses and doses to members 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 is totally released into its respective containment, or if both inventories are simultaneously released into their respective containments, the occupational doses and doses to members of the general public in the unrestricted areas will 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 of 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 any radiography bay external to the reactor tank will not result in damage to the reactor controls or the 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 of radioactive 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 and type of material in the experiment shall be limited such that the airborne radioactivity in the reactor room will not result in exceeding the applicable dose limits in 10 CFR Part 20 in 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) Ifthe 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) Ifthe 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 by boiling this material can escape only through an undistributed column of water above the core, at least 10% of these vapors can escape.
- c. If a capsule fails and releases material which could damage the reactor fuel or structure by corrosion 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 his designated 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 reactor room or the unrestricted area will result in exceeding the applicable dose limits in 10 CFR Part 20.
- b. These assumptions are used to evaluate the potential airborne radioactivity release due to an experiment 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 be brought to the attention of the UCD/MNRC Director or his designated alternate for review to assure 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 reactor operating 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 time period. Prior to reactor system or measuring channel operation, the surveillance shall be performed for each reactor 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 level monitoring channels. O biecltive - The objective is to verify that the maximum power level of the reactor does not exceed 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 required to 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
................................................................ ~
....... /the Aplcbiiy reactor core.These specifications apply to the surveillance requirements for reactivity control of betve- The objective is to measure and verify the reactivity worth, performance, and operability 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 shutdown margin 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 required to perform the surveillance. Basis -
- a. The reactivity worth of the control rods is measured to assure that the required shutdown margin is available and to provide an accurate means for determining the excess reactivity of the 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 been made.
- 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 change unexpectedly.
4.2 Reactor Control and Sa~fet,! Systems 4.2.1 Control Rods Applicability - This specification applies to the surveillance of the control rods. Objective - The objective is to inspect the physical condition of the reactor control rods and establish the operable condition of the rods. Spoecification - Control rod worths shall be determined annually or after physical removal or any significant change in core or control rod configuration.
- a. Each control rod shall be inspected at annual intervals by visual observation of the fueled sections and absorber sections plus examination of the linkages and drives.
- b. The scram time .ofeach control rod shall be measured semiannually.
I. Discovery of noncompliance with Technical Specifications 4.2.1 .a-b shall limit reactor operations to that required to 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.
- 0 configuration provides information about changes in reactor total reactivity and individual rod worths. 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 assure operable 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 fuel temperature 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 the reactor power level safety channels operate properly.
\J b. A channel test monthly of the reactor power level multi-range channel will assure that the channel is operable and responds correctly. The channel check will assure that the reactor power 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 is operable and responds correctly. A channel check of the reactor power level wide range log channel will assure that the channel is operable on a daily basis. A channel calibration will assure 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, will assure that the fuel temperature measuring channels operate properly.
- e. A channel test prior to pulsing plus the annual channel calibration will assure the pulse energy integrating 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. 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 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. 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 is operable.
- c. A channel test monthly of the Radiography Bay Manual Scrams will assure that the scrams are operable.
- d. A channel test monthly of the Reactor Power Level Safety Scrams will assure that the scrams are operable.
- e. A channel test monthly of the Loss-of-High-Voltage Scram will assure that the high voltage power supplies are operable and respond correctly.
- f. A channel test monthly of the Fuel Temperature Scrams will assure that the scrams are operable.
- g. A channel test monthly of the Watchdog Circuits Scrams will assure that the scram circuits are operable.
- h. A channel test monthly of the External Scrams will assure that the scrams are operable and respond 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 is operable.
I. A channel test monthly of the Magnet Current Key Switch will assure that thescram is operable. 4.2.4 Reactor Fuel Element~s A**iailitv- 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 of instrumented fuel elements), and all elements adjacent to the central irradiation facility shall be inspected annually.
- b. Instrumented fuel elements shall be inspected ifany 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 to perform the surveillance.
Basis (Technical Specifications 4,2,4.a-b) - The above specifications assure that the lead fuel elements shall be inspected regularly adteintegrity.o h edfe lmnssalb maintained. These are the fuel elements with the highest power density as analyzed in the SAR Chapter 4, Section 4.5.5.6. The instrumented fuel element is excluded to reduce the risk of damage to the thermocouples.
4.3 Reactor Coolant Systems Aoolicability - 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, the tank water level alarm, the water conductivity cells and the emergency core cooling system are all operable.
.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) Achannel 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 to perform the 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 input signal. 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 input 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 for power levels above 1.5 MW. A channel calibration semiannually assures that the Emergency Core Cooling System performs as required for power levels above 1.5 MW.
*"L'1
4.4 Reactlor Room Exha~ust System
\ Applicability - This specification applies to the surveillance requirements for the reactor room exhaust 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's operation. 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 verify that the exhaust system is maintaining a negative pressure in the reactor room relative to the surrounding facility areas. 4.5 This section intentionally left blank 4.6 This section intentionally left blank. . 4.7 ,Rea~ctor Radiation Monitoring Systems Applicability - 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) Achannel 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 to perform the surveillance. Basis -
- a. A channel check of the facility stack monitor system during each day's operation will assure the monitor is operable. A channel test weekly will assure that the system responds correctly to a 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 assure that the monitor is operable. A channel test weekly will ensure that the system responds to a known source. A channel calibration of the monitor annually will assure that the monitor reads correctly.
- c. Achannel 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 system responds to a known source. 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 responds correctly to a known source. A channel calibration annually will assure that the CAM reads 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 may damage the reactor or release excessive amounts of radioactive materials as a result of experimenta* failure. Soecification -
- a. A new experiment shall not be installed in any UCD/MNRC reactor experiment facility until a written safety analysis has been performed and reviewed by the UCD/MNRC Director, or his designee, 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 approved Facility Use Authorization. Facility Use Authorizations and experiments carried out under these authorizations 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 UCDIMNRC experiment facility until it has been reviewed for compliance with the approved experiment and Facility Use Authorization 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 other UCD/MNRC reactor in-core or in-tank irradiation facility shall be estimated or measured, as
.................
~...............
---. '
Iff_/,/ shall be donebefore appropriate, at ambient reactorconditions. operation with said experiment. Whenever a measurement is done it
- d. Experiments shall be identified and a log or other record maintained while experiments are in any UCD/MNRC reactor experiment facility.
Basis - a & b. Experience at most TRIGA reactor facilities verifies the importance of reactor staff and safety committee reviews of proposed experiments.
- c. Measurement of the reactivity worth of an experiment, or estimation of the reactivity worth based 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 will facilitate maintaining surveillance over such experiments.
5.0 Design Feat~ures 5.1 Site and Facility Description!.- 5.1.1 Site Applicability - This specification applies to the UCD/MNRC site location and specific facility design 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 downtown Sacramento, 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. The unrestricted 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 has 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 of the 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 bnclosi 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 to be 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 of radioactivity released into the environment, or during a Loss of Coolant Accident, the system is to 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 and exhaust air from the UCD/MNRC facility. The system shall have an exhaust stack height of a minimum of 18.2rn (60 feet) above ground level.
- b. Manually activated shutdown controls for the exhaust system shall be located in the reactor control room.
Basis - The UCD/MNRC facility exhaust system is designed such that the reactor room shall be maintained at a negative pressure with respect to the surrounding areas. The free air volume within the 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 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 to operating personnel. 5.2 Reactor Coolant System Applicability - This specification applies to the reactor coolant system.
.Obiective - The objective is to assure that adequate water is available for cooling and shielding during normal 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 ifthe water level in the reactor tank drops below a depth of 23 ft.
- c. For operations at 1.5 MW or higher during a Loss of Coolant Accident the reactor core shall be cooled for 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 be exceeded 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 and i 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 RatrCr Aorolicalbility - This specification applies to the configuration of the fuel. Obiective - The objective is to assure that provisions are made to restrict the arrangement of fuel elements 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 of 96 or more fuel elements to include fuel followed control rods. Below 0.5 MW there is no minimum 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 be loaded 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 20E core 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 done before 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 above loading restrictions will be allowed in an operational 0.5 MW or greater core. Specifications for the 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 Argonne National Laboratory Report ANLIED 97-54). 5.3.2 Reactor Fuel I Applicability - 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 their physical and nuclear characteristics.
- Sp~ecification.-
characteristics: The individual unirradiated TRIGA fuel elements shall have the following 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 (inthe 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 limiting operation to 2.3 megawatts steady state or to a 36 megawatt-sec pulse assures an ample margin of safety between 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 equal to 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 0C during any condition of normal operation, 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 of purchased materials.
5.3.3 Contr~ol .Rodsand Control Rod Drives
.. / A oolicabilitv - This specification applies to the control rods and control rod drives used in the reactor core.
Obiective - The objective is to assure the control rods and control rod drives are of such a design 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 and regulating rods shall have fuel followers sealed in stainless steel. The transient rod shall have an air 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 assure mechanical st~bility during movement and to isolate the neutron poison from the tank water environment. 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"rapid insertion capability. These drives have been tested and proven in many TRIGA reactors.
- A 5.4 Fissionable Materiall Storaae
.... " ADp1icabilitraco coe. This specification applies to the storage of reactor fuel at a time when itis nat in the reacto core
- Objective - The objective is to assure that the fuel which is being stored will not become critical and will not reach an unsafe temperature.
* -a. All fuel elements not inthe reactor core shall be stored (wet or dry) in a geometrical array where the kef 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 Controls 6.1 Organization. The Vice Chancellor for Research shall be the licensee for the UCDIMNRC. The UCD/MNRC facility shall be under the direct control of the UCD/MNRC Director.: The UCD/MNRC Director shall be accountable to the Vice Chancellor for Research for the safe operation and maintenance of the fac~ility.
*6.1.1 'Struciture. The management for operation of the UCD/MNRC facility shall consist of the organizational 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 the operating license and technical specifications. The Operations Manager shall report directly to the UCDIMNRC Director, and shall immediately report all items involving safety and licensing to the Director for a final decision. The Reactor Supervisor and Health Physics Supervisor report directly to the Operations Manager. 6.1.3 Stffn 6.1.3.1 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 reactor operator should be within thirty (30) minutes of the facility and reachable by telephone, and;
- d. A senior reactor operator shall be present whenever a reactor startup is performed, 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 be available 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 licensed operators shall be subject to an approved Nuclear Regulatory Commission (NRC) program.
6.2 Review. Audit. Recommendation anld Approval Genleral Policy. Nuclear facilities shall be designed, constructed, operated, and maintained in such 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 with applicable 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 shall appoint the Chairperson 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. The NSC shall evaluate and review nuclear safety associated with the operation and use i' 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 and proposed Class I modifications, prior to implementing (Class I) modifications, prior to taking action under the preceding documents or prior to forwarding any of these documents to the i' Nuclear Regulatory Commission for approval.
i*......"Jc. Review and determine whether a proposed change, test, or experiment would constitute an unreviewed safety question or req.uire a change to the license, to a Facility Use Authorization, or 32
to the Technical Specifications. This determination may be in the form of verifying a decision already 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 for UCD/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 .ofsuch 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 other inspectionsof these facilitieg conducted by other agencies.
-*-6.2.4 Audit/Inspection Function. The NSC or a subcommittee thereof, shall audit/inspect reactor operations 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 license and Technical 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 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). The Health 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 to assure that procedures are current, Procedures shall be adequate to assure the safe operation of the 33
reactor, but shall ..... *Procedures shall not preclude be in effect forthetheusefollowing of independent items: judgment and action should the situation require. 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 other routine maintenance that could have an effect on reactor safety.
- e. Testing and calibration of reactor instrumentation and controls, control rods and control rod drives.
- 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 alarms 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 solid and liquid waste and gaseous effluents released from the facility. The program shall include a management 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 items from 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 together for 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 the UCD/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 approval as a new experiment.
- c. Minor changes to an experiment that do not significantly alter the experiment may be approved by a senior 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 the UCD/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 and contributing factors. (2) Effect of the violation upon reactor facility components, systems, or structures, and on 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 the NRC 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 is necessary to shut down the reactor to correct the occurrence, operations shall not be resumed unless 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 the NRC Operations Center within 24 hours of the occurrence. A written preliminary report shall be sent to the NI*.C, 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 the above address within 30 days of the occurrence.
- d. Reportable occurrences should be reviewed by the NSC prior to forwarding any written report 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: Document Control 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 during the previous calendar year shall be submitted within six months following the end of each calendar year. Each annual report shall include the following~information:
- a. A brief summary of operating experiences including experiments performed, changes in facility design, 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 reactor was critical, and the cumulative total energy output since initial criticality.
- c. The number of emergency shutdowns and inadvertent scrams, including reasons for the shutdowns or scrams.
- d. Discussion of the major maintenance operations performed during the period, including the effect, ifany, 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 or in procedures, and of tests and experiments carried out pursuant to Section 50.59 of 10 CFR Part 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 as follows: 1 The total estimated quantity of radioactivity released (in curies). 2 An estimation of the specific activity for each detectable radionuclide present ifthe specific activity of the released material after dilution is greater 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 release occurs, 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) released during 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:
1determined The total estimated quantity sampling by an appropriate of radioactivity releasedmethod. and counting (in curies) 2 The total estimated quantity (in curies) of Argon-41 released during the reporting period based on data from an appropriate monitoring system. 3 The estimated maximum annual average concentrationof Argon-41 in the unrestricted area (in microcuries/mi), the estimated corresponding annual radiation dose at this location (in millirem), and the fraction of the applicable 10 CFR 20 limits for these values. 4 The total estimated quantity of radioactivity in particulate form with half lives greater than eight days (in curies) released during the reporting 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, by facility users. and-by visitors in terms of the average radiation exposure per individual and the greatest exposure per individual in each group.
- h. An annual summary of the radiation levels and levels of contamination observed during routine 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 planned administrative 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 in Section 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 could render the system incapable of performing its intended safety function unless the failure is 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 a condition 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 through 6.7.2.a.8. (2) The written report (and. to the extent possible, the preliminary telephone report or report by similar conveyance) shall describe, analyze, and evaluate safety implications. and outline the corrective measures taken or planned to prevent re~occurrence of the event.
- 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 or previously 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 the Safety Analysis Report (SAR); (3) A personnel change involving the positions of UCD/MNRC Director or UCO Vice Chancellor 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 or development 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 required information may be contained in single or multiple records, or a combination thereof. Records and logs shall 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 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 shipped 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.
.1 J" 39
a ~
. . . - .
a . a.' I. Formal Licensing Channel Administrative Reporting Channel
............. Communications Channel UCD/MNRC ORGANIZATION FOR LICENSING AND OPERATION FIGURE 6.1
UCDAVIS 5335 PRICE AVENUE MNRC McCLELLAN NUCLEAR RESEARCH CENTER BUILDING 258 McCLELLAN, CA 95652 PHONE: (916) 614-6200 FAX: (916) 614-6250 WEB: http://mn rc.ucdavis.edu U.S. Nuclear Regulatory Commission October 29, 2015 Attn: Linh N. Tran, Senior Project Manager, NRR Mail Stop: 012 D20 One White Flint North 11555 Rockville Pike Rockville, MD 20852 RE: 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.
Dear Ms. Tran,
In response to your letter dated June 3, 2015, we are submitting the requested documentation per said letter 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 October29, 2015. Assoca'e Director of Operations Reactor Supervisor McClellan Nuclear Research Center University of California-Davis Facility Operating License No. R-130. C: B. Klein, UCD/MNRC
P
- - *.*oNUCLEAR REGULATORY COMMISSION
- WASHINGTON, 0.0. zt&o5S5O1Q FACILITY OPERATING LICENSE DOCKET NO. 50-607 DEPARTMENT OF THE AIR FORCE AT McCLELLAN AIR FORCE BASE
- , License No. R-130
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A." The application for license, filed by the Department of the Air Force at McClellan Air Force Base, on October 23, 1 996, as supplemented, complies with the standards and requirements of the Atomic Energy Act of 1!954, as =amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. Construction. of the facility was completed in substantial conformity with the provision's 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;
- 0. There is reasonable assurance (i) that the activities authorized by this license can be conducted ,without endangering the health and safety of the 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 the activities authorized by this operating license in accordance with the regulations of t/he Commission; ..
F. The licensee is a Federal agency and will use the facility for defense programs and research. The licensee, in accordance with 10 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 the requirements o~f 10 CFR Part 140 of the Commission's regulations;
G. The issuance of this license will not be inimical to the common defense and 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 the Commissioh's regulations and all applicable requirements have been satisfied; and I.The receipt:, possession, and use of the byproduct and special nuclear materials as authorized by this license will be in accordance with the Commissioa 's regulations in 10 CFR Parts 30 and 70, including Sections 30.33, 70.23, and 70.31.
- 2. Facility Operating License No. R-130 is hereby issued to the Department of the 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 the Department 'of the Air Force at McClellan Air Force Base (the licensee). The facility is located on the licensee's site at McClellan Air Force Ease and 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, the Commission ~hereby licenses the Department of the Air Force at McClellan Air Force Base: (1) Pursuant to Section 104c of the Act and 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," to possess, use, and operate the facility at the designated location at 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 to 21 .0: kilograms of contained uranium-235 enriched to less than 20 percent in the isotope uranium-235 in the form of reactor fuel;:up to 4 grams of contained uranium-235 of any enrichment in the form of fission chambers; up to 16.1 kilograms of contained uranium-235 enriched to less than 20 pecenR[[tin he isotope uranium-235 in the form of plates; and to possess, but not separate, such special nuclear material as may be produced by the operation of the facility.
3 (3) Pursuant to the Act and 1 0 CFR Part 30, "Rules of General Applicability to Domestic Licensing of Byproduct Material," to receive, possess, and use a 4-curie sealed americium-beryllium neutron source in connection with operation of the facility; a 55-millicurie sealed cesium-1 37 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 in reactor experiments, such byproduct material as may be produced by the 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, and orders of the Commission now or hereafter in effect and to the additional conditions specified below: (1) Maximum Power Level The licensee is authorized to operate the facility at steady-state power levels not in excess of 2300 kilowatts (thermal) and in the pulse mode with reactivity insertions not to exceed $ 1.75 (1.23 %Ak/k). (2) Technical Specifications The Technical Specifications contained in Appendix A are hereby incorporated in the license. The licensee shall operate the !facility in accordance with the Technical Specifications. (3) Physical Security Plan The 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 the authority of 10 CFR 50.90 and 1 0 CFR 50.54(p). The approved plan;i which is exempt from public disclosure pursuant to the provisions of 10 CFR 2.790, is entitled "Physical Security Plan for the McClellan Nuclear Radiation Center (MNRC) TRIGA Reacitor Facility," Revision 3, dated August 1996.
4
- 0. This license is effective as of the date of issuance and shall expire twenty (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
~NUCLEAR REGULATORY COMMISSION
*" ~UNITED STATES "
*. WASHINGTON 1 D.C. 20555-0001
* "'r*°*
'December 9, 1998 Brigadier General Michael P. Wiedemer, Commander Sacramento Air Logistics Center SM-ALC/TI- 1 5335 Price Avenue McClellan 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 License No. 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 to your 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 airborne radioactivity in the reactor room or the unrestricted area exceeding the applicable dose limits in 10 CFR Part 20. A copy of the safety evaluation supporting Amendment No. 1 is also enclosed. Sincerely, Warren J. Eresian, Project Manager Non-Power Reactors and Decommissioning
*Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No. 50-607
Enclosures:
- 1. Amendment No. 1
- 2. Safety Evaluation cc w/enclosures:
See next page
McClellan AFB TRIGA REACTORDcktN.067 cc: Dr. Wade J. Richards SM-ALCITI-1 5335 Price Avenue, Bldg. 258 McClellan AFB, California 95652-2504 Col. Robert Capell HQ AFMC/SGC 4225 Logistics Avenue, Suite 23 Wright-Patterson AFB, Ohio 45433-5762 Lt. Col. Catherine Zeringue HQ AFSC/SEW 9570 Avenue G, Building 24499 Kirtland AFB, New Mexico 87117-5670 Test, Research, and Training Reactor Newsletter 202 Nuclear Sciences Center University of Florida Gainesville, Florida 32611
~NUCLEAR REGULATORY COMMISSION
* .'
- O*,i.UNITED STATES WASHINGTON, D.C. 20888-0001 DEPARTMENT OF THE AIR FORCE AT McCLELLAN AIR FORCE BASE DOCKET NO. 50-607 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 1 License No. R-130
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that A. The application for an amendment to Facility Operating License No. R-1 30 filed by the Department of the Air Force at McClellan Air Force Base (the licensee) on November 18, 1998, conforms to the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the regulations of the Commission as stated in Chapter I of Title 10 of the Code of FederalRegulations (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 the Commission; D. The issuance of this amendment will not be inimical to the common defense and security or t*o 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; and F. 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.
2
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the enclosure to this license amendment, and paragraph 2.C.(ii) of Facility Operating 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 through Amendment No. 1, are hereby incorporated in the license. The licensee shall operate 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, Director Non-Power Reactors and Decommissioning Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation
Enclosure:
Appendix A, Technical Specifications Changes Date of Issuance:
ENCLOSURE TO LICENSE AMENDMENT NO. 1 FACILITY OPERATING LICENSE NO? R-1 30 DOCKET NO. 50-607 Replace the following pages of Appendix A, "Technical Specifications," with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. Remove Insert 24 24 25 25
.j° .;* ,
This specification restricting thec.quantity is intended to prevent damage to vital equipment by 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 TNT equivalent or less in any radiography bay external to the reactor tank will not result in damage to the reactor controls or the reactor tank. Safety Analyses have been performed (SAR Chapter 13, Section 13.2.6.2) which show that up to six (6) lbs of TNT equivalent can be safely irradiated in any radiography bay. Therefore, the three (3) lb limit gives a safety margin of two (2).
3.8.3 Failure and Malfunctions* Applicability. This specification applies to experiments installed in the reactor, 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 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 or aerosols into the reactor building or into the unrestricted area, the quantity and type of material in the experiment shall be limited such that the airborne radioactivity in the reactor room or the unrestricted area will not result in exceeding the applicable dose limits in 10 CFR Part 20, assuming 100% of the gases or aerosols escape.
- b. In calculations pursualtt 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 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 vapors formed by boiling this material can escape only through an undistributed column of water above the core, at least 10% of these vapors 'can escape. 24
- c. If a capsule fails and releases material which could damage the reactor fuel or structure by corrosion or other means, an evaluation shall be made to.
determine the need for corrective action. Insipection and any corrective action taken shall be reviewed by the Facility Director or his designated 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 reactor room or the unrestricted area will result in exceeding the applicable dose limits in 10 CFR Part 20.
- b. These assumptions are used to evaluate the potential airborne radioactivity release due to an experiment 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 must be brought to the attention of the Facility Director or his designated alternate for review to assure safe operation of the reactor (SAR Chapter 13, 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 end of the operating cycle. If the reactor is not operated for a reasonable .time, a r'eactor 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 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 level monitoring channels. Objective. The objective is to verify that the maximum power level of the reactor does not exceed the authorized limit. 25
~NUCLEAR REGULATORY COMMISSION
,* ' *o~i'-UNITED
- STATES WASHINGTON," O.C. 20865-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 1 TO FACILITY OPERATING LICENSE NO. R-130 DEPARTMENT OF THE AIR FORCE AT McCLELLAN AIR FORCE BASE DOCKET NO..POO-607
1.0 INTRODUCTION
By letter dated November 18, 1 998, the Department of the Air Force at McClellan Air Force 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 the unlikely 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 and its experimental facilities. Specifications.
- 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 or aerosols into the reactor building or into the unrestricted area, the quantity and type of material to be irradiated shall be limited such that the airborne concentration of radioactivity shall not exceed the applicable limits of 10 CFR Part 20 (at the operations boundary), assuming 100% of the gases or aerosols escape.
"h.
O *° °" 0 2 Bases.
- a. This specification is intended to reduce the likelihood that airborne radioactivity in excess of the limits of 10 CFR Part 20 shall be released into the reactor building or to the unrestricted area (SAR Section 13.2.6.2).
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, 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 or aerosols into the reactor building or into the unrestricted area, the quantity and type of material in the experiment shall be limited such that the airborne radioactivity in the reactor room or the unrestricted area will not result in exceeding the applicable dose limits in 10 CFR Part 20, assuming 100% of the gases or aerosols escape.
Bases.
- a. This specification is intended to reduce the likelihood that airborne radioactivity in the reactor room or the unrestricted area will result in exceeding the applicable dose limits on 10 CFR 20.
The licensee has proposed clarifying the TS by basing the TS on dose instead of concentrations of radioactive material. The purpose of this TS is to limit doses to members of 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 reactor confinement and subsequently to the environment. Doses to members of the reactor staff and members of the public from accidents at research reactors are limited to the doses given 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 the annual 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 50 reins 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 2000 working 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 if inhaled continually over the course of a year would produce a total effective dose equivalent of 0.05 rem. The licensee's proposed wording would be based on dose limits directly. The licensee is concerned that the TS as currently written could be interpreted to limit releases to the instantaneous concentration of airborne radioactive material in the reactor room and unrestricted areas. This would ignore the time integral aspects of the concentration 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 20 while 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 a TS based on dose would also include consideration of radiation shine from a cloud of radioactive material. This proposed change to the TSs is acceptable to the staff because the 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 located within 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 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 criteria for categorical exclusioni set forth in 10 CFR 51.22(c)(9). Pursuant to 10CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared 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 or consequences of accidents previously evaluated,* or create the possibility of a new or different kind of accident from 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 is reasonable assurance that the health and safety of the public will not be endangered by the proposed activities; and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or the health and safety of the public. Principal Contributor: Warren J. Eresian Date: December 9, 1998
*. *.--*,*UNITED STATES o NUCLEAR REGULATORY COMMISSION Z r~oWASHINGTON, D.C. 2055.5-0001 FACILI;TY OPERATING LICENSE DOCKET NO. 50-607 DEPARTMENT OF THE AIR FO.RCE.AT McCLELLAN AIR FORCE BASE License No. R-1 30
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A. The application for license, filed by the Department of the Air Force at McClellan Air Force Base, on October 23, 1996, as supplemented, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR 'Chapter I; B. Construction of the facility was completed in substantial conformity 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 be conducted without' endangering the health and safety of the 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 the activities authorized by this operating license in accordance with the regulations of the Commission;... F. The licensee is a Federal agency and will use the facility for defense programs and research. The licensee, in accordance with 10 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 the requirements of 10 CFR Part 140 of the Commission's regulations;
2 G. The issuance of this license will not be inimical to the common defense and 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 the Commission's regulations and all applicable requirements have been satisfied; and I.In*- c]p, 1 °os s e-ssio n,-a +n *ni !le*r
- 2. Facility Operating License No. R-130 is hereby issued to the Department of the 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 the Department of the Air Force at McClellan Air Force Base (the licensee). The facility is located on the licensee's site at McClellan Air Force Base and 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, the Commission hereby licenses the Department of the Air Force at McClellan Air Force Base: (1) Pursuant to Section 104c of the Act and 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," to possess, use, and operate the facility at the designated location at 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 to 21 .0 kilograms of contained uranium-235 enriched to less than 20 percent in the isotope uranium-235 in the form of reactor fuel; up to 4 grams of contained uranium-235 of any enrichment in the form of fission chambers; up 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 be produced by the operation of the facility.
3
**_*:*ea 4-curie sealed americium-beryllium neutron source in connection with operation of the facility; a 55-millicurie sealed cesium-i137 source for instrument calibrations; small instrument calibration and check 'sources of less than 0.1 millicurie eah.* *.
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, and orders of the Commission now or hereafter in effect and to.the additional conditions specified below: (1) Maximum Power Level The licensee is authorized to operate the facility at steady-state power levels not in excess of 2300 kilowatts (thermal) and in the pulse mode with reactivity insertions not to exceed $1.75. (1.23 %Ak/k). (2) Technical Specifications The Technical Specifications contained in Appendix A are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. (3) Physical Security Plan The licensee shall fully implement and maintain in effectel.- provisions of the Commission-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 approved plan, which is exempt from public disclosure pursuant to the provisions of 1 0 CFR 2.790, is entitled "Physical Security Plan for the McClellan Nuclear Radiation Center (MNRC) TRIGA Reactor Facility," Revision 3, dated August 1 996.
S .. * .:.
.4
- 0. This license is effective as of the date of issuance and shall expire twenty (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, 1999 Brigadier General Michael P. Wiedemer, Commander Sacramento Air Logistics Center SM-ALC/TI- 1 5335 Price Avenue McClellan AFB, California 95652-2504
SUBJECT:
ISSUANCE OF AMENDMENT NO. 2 TO AMENDED FACILITY OPERATING LICENSENO. R-130 - DEPARTMENT OF THE AIR FORCE AT McCLELLAN AIR FORCE BASE (TAC NO. MA3477)
Dear General Wiedemer:
The Commission has issued enclosed Amendment No. 2 to Facility Operating License No. 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 Safety Analysis Report (SAR) to support expanded experimental facilities in response to your submittal of January 11, 1999. The amendment provides for the installation of an Argon-41 Production Facility and a Central Irradiation Facility. The installation of the Argon-41 Production Facility does not require any change or expansion of the TSs since an experiment failure will not result in airborne radioactivity in the reactor room or the unrestricted area exceeding the applicable dose limits already prescribed. The installation of the Central Irradiation Facility requires 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 the Central 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 Manager Non-Power Reactors and Decommissioning Project Directorate Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No. 50-607
Enclosures:
- 1. Amendment No. 2
- 2. Safety Evaluation cc w/enclosures: See next page
McClellan AFB TRIGA REACTORDoktN.5-0 cc" Dr. Wade J. Richards SM-ALC/TI-1 5335 Price Avenue, Bldg. 258 McClellan AFB, California 95652-2504 Lt. Col. Marcia Thornton HQ AFSC/SEW" 9570 Avenue G., Bldg. 24499 Kirtland AFB, New Mexico 87117-5670 Col. Robert Capell HQ AFMC/SGC 4225 Logistics Avenue, Suite 23 Wright-Patterson AFB, Ohio 45433-5762
0* 0
*. UNITED STATES
.NucLEAR REGULATORY COMMISSIoN WHNToND.C. 208-o000 DEPARTMENT OF THE AIR FORCE AT Mc.CLELLAN AIR FORCE BASE DOCKET NO. 50-607 AMENDMENT TO FACILITY OPERATING LICENSE AmendmentNo. 2 License No. R-1 30 1.. The U.S. Nuclear Regulatory Commission (the Commission) has found that A. The application for an amendment to Facility Operating License No. R-1 30 filed by the Department of the Air Force at McClellan Air Force Base (the licensee) on January 11, 1999, conforms to the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the regulations of the Commission as stated in Chapter I of Title 10 of the Code of FederalRegulations (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 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 the Commission; D. The issuance of this amendment will not be inimical to the common defense and security 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; and F. 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.
- 2. Accordingly, the license is amended by changes to the Safety Analysis Report and Technical Specifications as indicated in the enclosure to this license amendment, and paragraph 2.C.(ii) of Facility 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 through Amendment No. 2, are hereby incorporated in the license. The licensee shall 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, Director Non-Power Reactors and Decommissioning Project Directorate Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation
Enclosure:
- Appendix A, Technical Specifications
*and Safety Analysis Report Changes Date of Issuance: March 1, 1999
.... 0.
ENCLOSURE TO LICENSE AMENDMENT NO. 2 FACILITY OPERATING LICENSE NO. R-130 DOCKET NO. 50-607 A. Replace the following page of Appendix A, "Technical Specifications," with the enclosed page. The revised page is identified by amendment number and contains vertical lines indicating the areas of change. Remove Insert 22 22 B. Insert the following sections into the Safety Analysis Report.
- 1. Add new Section 10.5.3
- 2. Add new Section 11.1.1.1.6
- 3. Append to Section 13.2.6.2
- 4. Add new Appendix A to Chapter 13
- 5. Add new Appendix .B to Chapter 13
- 6. Change Section 10.4.1
- 7. Add new Section 10.4.1.4
- 8. Append to Section 1 3.2.6.2
- 9. Add Reference 13.19 to ChaPter 13
- S unrestricted 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 Central Irradiation 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 be less than one (1) dollar (0.7% AK/K).
- b. The absolute reactivity worth of any single secured experiment shall be less than the maximum allowed pulse ($1.75) (1.23% AK/K).
- c. The absolute total reactivity worth of experiments installed in the reactor and in-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, or removal 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 the sample 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, Chapter 13.2.2.2.1 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 preclude pulsing if the experiment's fixturing should fail, since the resulting reactivity insertion would not cause prompt criticality if less than one dollar. Given that the reactor will not pulse inadvertently, the additional increase in transient power and temperature will be slow enough so that the fuel temperature scram will be effective.
- b. The absolute worst event which may be considered in conjunction with a single secured experiment is its sudden accidental or unplanned removal while the reactor is operating. This would result in a reactivity increase less than a pulse of $1.92, analyzed in SAR Chapter 13, Section 13.2.2.2.1.
- c. It is conservatively assumed that simultaneous removal of all experiments in the reactor and in-tank experiment facilities at any given time shall not exceed the maximum reactivity insertion limit. SAR, Chapter .13, Section 13.2.2.2.1 shows that an insertion 22 Amendmient No. 2
ADDITION TO MCCLELLAN NUCLEAR RADIATION CENTER SAFETY ANALYSIS REPORT - ARGON-41 PRODUCTION FACILITY NEW SECTION 10,5.3 10.5.3 Arcqon-41 Production Facility The Argon-41 Production Facility will produce 1-2 curies of 4 1Ar for research and commercial use. The 41Ar will be produced by introducing argon gas into a stainless steel container located in one of the silicon irradiation positions (adjacent to the graphite reflector and external to the reactor core - Figure 10.111A). 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 desired activity 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 valve is opened to pressurize the cooling coils above the liquid nitrogen bath. The Dewar is then raised 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 of 41Ar gas to the sample container. The liquid nitrogen Dewar is lowered, exposing the cooling coils to room temperature. When that portion of the system between the cooling coils and the sample container has reached equilibrium the sample container will be isolated and..removed from the room. The coil is surrounded with a lead shield to minimize the radiation exposure to personnel. A catch tank surrounds the Dewar to contain any liquid nitrogen escaping from'the Dewar or in the unlikely event of a total failure of the Dewar. Over pressure protection of the overall system is provided by several relief valves that vent to an over pressure tank. The over pressure ta~nk is protected by its own relief valve which vents to the reactor room. The tank is located as high as possible in the reactor room. 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 the 41 Ar. The Argon-41 Production Facility consists of several different components, with the major components listed below.
0 COMPONENT MATERIAl DESCRIPTION Irradiation Container 304 stainless The irradiation container is a 1000 ml sample cylinder with a working pressure steel of 1 800 psig and a burst pressure of 6000 psig. It conforms to the "Shipping Container Specifications" from the U.S. Code of Federal Regulations, Title 49 or Bureau of Explosives Tariff No.BOE 6000. Over Pressure 304 stainless The adjustable proportional pressure relief Relief Valves steel valves have a working pressure up to 6000 psig. When upstream pressure overcomes the force exerted by the spring, the poppet opens, allowing flow through the valve. As the upstream pressure increases, flow through the valve increases proportionately. Cracking pressure is only sensitive to inlet pressure and is not affected by outlet pressure. Over Pressure Carbon steel 30 gallon tank. Relief Tank Valves 304 stainless Bellows sealed valves. steel Tubing 304 stainless 1/4-inch and Y/=-inch. steel NEW SECTION 11.1.1.1.6 11.1.1.1.6 Araqon-41 from the Argon-41 Production Facility Ar-41 will be produced by the Ar-41 Production Facility (see Chapter 10) as needed. The Ar-41 that is produced by the Ar-41 Argon Production Facility will be contained in the system so there should be no increase in the Ar-41 levels in the reactor room or the Ar-41 that is released to the unrestricted area. Catastrophic failure of the system will not result in any 10 CFR 20 limit being exceeded and is further discussed in Chapter 13. APPEND TO SECTION 13.2.6.2 The Argon-41 Production Facility (see Chapter 10) can produce argon-41 in excess of the amounts analyzed in Appendix A of the MNRC Safety Analysis Report. However, if the system releases argon-41, the gas will be contained in the reactor room and the existing
0 reactor 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 Stack Continuous Air Monitor will also be used to verify any release outside the MNRC boundary. If the system had a catastrophic failure and 4 curies of argon-41 were released to the volume of the reactor room, the argon-41 concentration in the reactor room would be approximately 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 be restricted. The reactor room ventilation system (as described in Chapter 9) would, be operated in the recirculation mode for approximately one day before the dose rate from argon-41 decays to less than 1 mR/hr. Therefore, the argon-41 Discharge Limit defined in the MNRC Technical Specifications will not be exceeded due to the recirculation mode of the reactor room ventilation system. Other potential accidents include failure of the irradiation container due to overpressurization from the argon gas supply cylinder, since a new argon supply cylinder is typically delivered at 2200 psig and the container is rated for 1800 psig. However, this requires multiple failures and is considered non-credible: a) the operator would have to violate an operational procedure; b) the regulator would have to fail, and c) at the same time the pressure relief valve would have to fail. Also, liquid nitrogen could spill into the reactor tank, causing expansion of the water and expelling a portion of tank water. To prevent this, a catch basin surrounds the Cold Trap, and the liquid nitrogen is supplied through a pipe in the reactor room wall connecting the trap to a supply container in the equipment room. A third accident could result if the pressure relief valve became choked with supersonic flow; however, the flow rates are estimated to be less than sonic (see calculat~ion in Chapter 13, Appendix A). NEW APPENDIX A TO CHAPTER 13 ARGON-41 CONCENTRATION IN REACTOR ROOM GIVEN:
- 1. Reactor room volume =-7.39x10 3 ft 3 tReference 111
- 2. 4 curies Ar-41 in argon production system
- 3. D(y)=, 2 = O.25Evx [Reference 21 Dy== 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
0 X *= concentration of gamma emitting isotope in the cloud (Ci/m 3 ) CALCULATIO )N: X = (4Ci)/[7.39xl10 3 ft 3 )(1 m3/35.314 ft 3 ) = 1 .91!x 0.2 Gi/m 3 D(y)=,2 = 0. 25Eyx
= (0.25)( 1.2936 Mev/dis)(1 .91 xl 0.2 Cl/rn 3 )
= (0.0062 rads/sec)(3600 seclhr)
= 22.24 radslhr D = Doe~x t = -(1/A)In(0D/D)
= -(T112Iln2)ln(D/D0 )
For 0 = 1 mrad/hr t = -(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), edited by Shelein, p. 439
- 3. 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 PROJECT Assume: Perfect Gas Constants: Property Value Units R 208 N-rn/k g-degK k(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 C c =[1 .67(208 N-m/kg-degK)(-200 +273)K( 1 kg-rn/N-sec 2 )I*
= 159.2397 rn/s at -200 degrees C Next, 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 psi V = (1 ft 3 /min)(12 in/ft) 3 (2.54 cm/in) 3 (1 min/60 sec)
= 471.9474 cm 3 /sec Area = r*r 2 = 3.14(0.18in/2)2 = 0.025434 in2 based on 1/4 inch tubing
= 0.16409 cm 2 Velocity = V/Area = 28.7615 rn/sec Mach 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 and therefore 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 REPORT AND TECHNICAL SPECIFICATIONS - CENTRAL IRRADIATION FACILITY CHANGE SECTION 10.4.1 The Central Irradiation Facility, located in the center of the reactor core, may contain either a plug assembly (as described in sections 10.4.1 .1 through 10.4.1 .3 and Figure 10.7) or a moveable sample can system (as described in section 10.4.1.4). All parts are removable from the reactor using underwater tools. NEW SECTION 10.4.1.4 10.4.1 .4 Central Irradiation Facility The 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 the insertion and removal of samples from the central facility through the use of a drive mechanism similar to the control rods. The central thimble is approximately 52 inches in length and 4.22 inches outer diameter with 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. Aluminum shims 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 fuel element locations in the B hex ring 4.25-inch hole. Two captive bolts attach the central thimble to the upper grid plate. These bolts prevent the accidental removal of the facility when removing samples from the central thimble. An 1100 aluminum slug located inside the central thimble is normally positioned in the reactor core. The aluminum slug is 4 inches in diameter and 24.75 inches in length. This voids the water from the central thimble when the sample can is removed from the thimble. An orifice plate is located on the bottom of the central thimble. In the event the aluminum slug releases from the locating holes and falls to the, bottom of the central thimble, the rate 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 3.99 inches and an inside diameter of 3.75 inches. The can could be free flooding or dry, and is used to position samples for irradiation in the reactor core. The positioning of samples can 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 being rotated. The drive mechaauism has the same type of drive motor as the control rod drives except the 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 wiBl treat the insertion and removal of the samples as if they were control rods. The reactor room controls can only be enabled from the reactor console. The normal indicators are as follows:
"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 travel indications, stop driving limits and start/stop rotation of the carrier. APPEND TO SECTION 13.2.6.2 Another potential accident involves the Central Irradiation Facility (see Chapter 10) since it may be considered similar to a control rod. Therefore, consider three potential scenarios for an uncontrolled reactivity insertion analogous to the Uncontrolled Withdrawal Of a Control Rod (see Section 13.2.2.2.2). First, if the material in the sample can were of sufficiently different worth than the aluminum cylinder, the sample can would cause reactivity changes in the same fashion as a control rod, and either operator error or mechanical failure could cause an uncontrolled reactivity insertion. Second, if the aluminum cylinder failed to engage upon the sample can's insertion, a water void would be created in the central facility as the aluminum cylinder descended ahead of the sample can. Similarly, if the aluminum cylinder failed to replace the can upon removal from the central facility a water void would result. All three of the above scenarios can be bounded by the Uncontrolled Withdrawal of a Control Rod analysis (Section 13.2.2.2.2). Specifically, the Central Irradiation Facility must have less reactivity and must drive slower than the control rod analyzed ($3.50 and 42 inches/minute, respectively). To that end, the reactivity of any material in the sample can shall be measured at low power to verify it's worth is not only less than $3.50 but also less than $1.75, the reactivity limit for the Central Irradiation Facility (based on the Technical Specification limit of $1 .75 for the pulsed reactivity insertion). For example, the worth 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.). The worth of an aluminum cylinder vs. void and vs. water has been analyzed by computer simulation (Reference 13.19). The most positive reactivity effect in the computer simulation 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 Case 3 to Case 1 2, where in an accident the sample can not only floods but also the aluminum cylinder drops, resulting in a decrease of about $.1.76. Thus, the worth of the sample can or 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 rod speed analyzed in the MNRC SAR (Section 13.2.2.2.2). Furthermore, in the event of failure of the aluminum cylinder to engage upon installation of the sample can, the base of the Central Thimble is designed (by sizing the hole in the base) to allow the aluminum cylinder to descend at no more than the analyzed 42 inches/minute. Therefore, the accident analysis in Chapter 13 of the MNRC SAR for Uncontrolled Withdrawal of a Control Rod (Section 13.2.2.2.2) is sufficient to bound any accident associated with the Central Irradiation Facility since: a) the material in the sample can shall be measured and verified to be less than $1.75 (half of the analyzed $3.50); b) the drive speed cannot exceed the analyzed 42 inches/minute; and c) the aluminum cylinder cannot fall uncontrolled faster than the analyzed 42 inches/minute. Finally, physical impact on the fuel is considered non-credible since the sample can is always contained in a guide tube or attached to a drive mechanism such that it is unlikely to drop onto the core (see description in Section 10.4.1.4). ADD REFERENCE 13.19 TO CHAPTER 13 13.1 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 Central Irradiation 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 be less 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 Facility is based on the pulsing reactivity insertion limit (section 3.1.2) and on the design of the sample 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, Chapter 13.2.2.2.1 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 preclude pulsing if the experiment's fixturing. should fail, since the resulting reactivity insertion would not cause prompt criticality if less than one dollar. Given that the reactor will not pulse inadvertently, the additional increase in transient power and temperature will be slow enough so that the fuel temperature scram will be effective.
0 9
*'* ,o*.UNITED STATES"
- NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C..20588-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 2 TO AMENDED FACILITY OPERATING LICENSE NO. R-130 DEPARTMENT OF THE AIR FORCE AT McCLELLAN AIR FORCE BASE DOCKET NO. 50-607
1.0 INTRODUCTION
By letter dated January .11, 1999, the Department of the Air Force at McClellan Air Force 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), and changes to the Safety Analysis Report. The amendment provides for the installation of an Argon-41 Production Facility and a Central Irradiation Facility. The installation of the Argon-41 Production Facility does not require any change or expansion of the TSs since an experiment failure will not result in airborne radioactivity in the reactor room or the unrestricted area exceeding the applicable dose limits already prescribed. The installation of the Central Irradiation Facility requires 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 the Central Irradiation Facility to $1 .75, corresponding to the pulse limit specified in TS 3.1 .2.
2.0 EVALUATION Argon-41 Production Facility The licensee has requested amendment of the Safety Analysis Report to provide for the installation of an Argon-41 Production Facility. The purpose of the facility is to produce Argon-41 for research and commercial uses. Argon gas from a pressurized argon bottle is introduced into a stainless steel container located in a position external to the core, but in the 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 is removed cryogenically and admitted to sample containers. Overpressure protection is provided 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 of up to 6000 psig. The overpressure tank relief valve vents to the reactor room. All piping (1/4 and Y/=inch 304 stainless steel) is anchored to prevent pipe whip in the event of pipe failure. The irradiation container has a working pressure of 1 800 psig with a burst pressure of 6000 psig.
After the argon gas has been irradiated, the gas is transferred to the sample containers. A cooling coil which has been evacuated with a vacuum pump is immersed in a liquid nitrogen 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 valved off, 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 the reactor 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 day will result in a dose rate of approximately 1 mrad per hour. The argon-41 discharge limit as defined in the Technical Specifications will not be exceeded. The licensee has considered other potential accidents. These include overpressurization of the irradiation container, spilling liquid nitrogen into the reactor tank, and the choking of a relief valve due to supersonic flow. Overpressurization of the irradiation container requires multiple mechanical failures and operator violation of the procedure governing the use of the production facility. To prevent the spilling of liquid nitrogen into the reactor tank, a catch basin is installed around the liquid nitrogen bath. Finally, the licensee has analyzed the flow through the relief valves and has determined that the flow remains subsonic, thus preventing choking at the valve. Central. Irradiation Facility The licensee has requested amendment of the 'Technical Specifications and Safety Analysis Report to provide for the installation of a Central Irradiation Facility. The facility allows 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 normal control rods, and a reactor operator controls the insertion and removal of samples. Drive speeds are equal to those of the normal control rods. The central thimble is essentially a vertical guide tube which passes through the upper grid plate, the lower grid plate and the safety plate, resting on the tank floor. lA sample can and an aluminum slug move within the central thimble. An aluminum slug normally occupies a position in the reactor core. When the sample can is inserted, the aluminum slug moves downward out of the co)re, and its position in the core is replaced by the sample 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 rod malfunctions are applicable. In particular, the analysiz of an Uncontrolled Withdrawal of a Control 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 the maximum 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 rod worth of $3.50 was used to allow for reasonable variations. In order to bound accidents involving the Central Irradiation Facility, it is required to show that 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 computer simulation (SAR Reference 13.19) of the reactivity changes associated with various scenarios,- including normal operations and accidents. The most limiting case, the flooding of 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 rod withdrawal 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 located within 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 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 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 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 or consequences of accidents previously evaluated, or create the possibility of a new or different kind of accident from 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 is reasonable assurance that the health and safety of the public will not be endangered by the proposed activities; and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or the health and safety of the public. Principal Contributor: Warren J. Eresian Date: MArch 1, 1999
9 9 UNITt* STATES
- 9 9
*
*
,AUCLEAR REGULATORY COMMISSION
- 1~
WASHINGTON, D.C. 2O55*-0OO1 Brigadier Commander General Michael P. Wledemer Vice Chancellor Kevin Smith Office of the Chancellor Sacramento Air Logistics Center University of California, Davis SM-ALC/TI-1 One Shields Avenue 5335 Price Avenue Davis, California 95616-8558 McClellan AFB, California 95652-2504
SUBJECT:
ORDER APPROVING THE TRANSFER OF THE FACILITY OPERATING LICENSE FOR THE McCLELLAN NUCLEAR RADIATION CENTER FROM THE DEPARTMENT OF THE AIR FORCE TO THE REGENTS OF THE UNIVERSITY OF CALIFORNIA. AND APPROVING CONFORMING AMENDMENT (TAC NO. MA3477)
Dear General Wiedemer and Dr. Smith*:
The enclosed Order Is in response to the application dated April 13,.1999, as supplemented on July 19 and August 4-,1999, and January 18 and 27, 2000, requesting approval of the transfer of Operating License No. R-1 30 for the McClellan Nuclear Radiation Center from the Departm~ent of the AIr Force to the Regents of the University of California, and approval of a conforming amendment to reflect the transfer. The enclosed Order provides consent to the proposed 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 for the 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~sy Warreni~J. Ere fan, Project Manager Events Assessment, Generic C~ommunlcations and Non-Power Re~ctom Branch DivIsion of Reguiator*"1mp rovement Programs Office of Nuclear Reactor Regulation Docket No. 50-607
*
Enclosures:
.1. Order
- 2. Amendment No.3
*.3. Safety Evaluation 4, IndemnityAgreement. .
- Senextp ge
McClellan AFB TRIGA REACTOR Docket No, 50-607 cc; Dr. Wade J. Richards SM-ALC/TI- 1 6335 Price Avenue, Bldg. 258 McClellan AFB, CA 95652-2504 Col. Robert Capell HQ AFMC/SGC 4225 Logistics Avenue, Suite 23 Wright-Patterson AFB, OH 45433-5762 Lt, Cot. Catherine Zeringue HQ AFSC/SEW 9570 Avenue G, Building 24499 Kircland AFB, New Mexico 871 17-5670 Test. Research, and Training Reactor Newsletter 202 Nuclear Sciences Center University of Florida Gainesville, FL 32611
7590-01 -P UNITED STATES OF AME*RICA 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 LICENSE AND 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 .Codeof_Federal Regufation~s (10 CFR Part 50). The facility is located on McClellan Air Force Base In Sacramento, 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 proposed transfer of Operating License No, R-1 30 from the USAF to the University of California. The University of Calliornia at Davis (UCD), part of the University of California, was proposed to be the actual operator of the facility. The application was supplemented by submittals dated July 19 and August 4, 1999, and January 18 and 27, 2000, The initial application and the supplements 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 sole operator 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 the operating license and replace them with references to the UCD, as appropriate, as well as to make other miscellaneous administrative changes to the operating license to ref lect the transfer. Under 10 CFR 50.80, no license for a production or utilization facility, or any right thereunder, 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 the application and other information before the Commission, the NRC staff has determined that the University of California Is qualified to hold the license, and that the transfer of the license to the 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 for the proposed license amendment complies with the standards and requirements of the Atomic Energy Act of 1954, as amended, and the Commission's rules and regulations set forth in 10 CFR Chapter 1; the facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission: there Is.reasonable assurance that the activities authorized by the proposed license amendment can be conducted without endangering the health and safetyof the public and that such activities will be conducted in compliance with theCommission's regulations: the issuance of the proposed license amendment will not be inimical to the common defense and security or to the health and safety of the public; and the issuance of the proposed amendment will be in accordance with 10 CFR
T r-P ." =* 1af1 1 NU.SS5r0 r P.5/)
* *14 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 herein to 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 shall become null and void, provided, however, on written application arnd for good cause shown, 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, to conform 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, Director Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation 4.
'*"*' '* %*UN1TED STATES WASHINGT"ON, D.C. 20555-0001 DEPARTMENT O T.HE AIR FORCF AT MCCLELLAN. AIR FoRCE BASE DOCKET NO. 50-607 AMENDMENT TO AMENDED FACILITY OPERATING LICENSE Amendment No. 3 License No. R-130 1.The U.$. Nuclear Regulatory Commission (the Commission) has tound that A. The application for an amendment to Amended Facility Operating License No. R-130 filed by tile Department of the Air Force at McClellan Air Force Base and the Regents of the University of California on April 13, 1999, as supplemented on July 19 and August 4, 1999, and January 18 and 27, 2000, conmpiies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the regulations of the Commission as stated In Chapter I of Title 10 of the Code of Federal 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 the Commission; D. The issuance of this amendment will not be~inlrmicalto the common defense and security or to the health and safety of the public;and E. This issuance of this amendment is in accordance with the regulations of the Commission as stated in 10 CFR Part 51, and all applicable requirements have been satisfied.
- 2. Accordingly, the license is amendedas indicated in the attachment to thisilcense amendment, ENCLOSURE 2
N.955 FEB.*1.006
- 09M P.7/1.4
- 3. This license amendment is effective as of the date of issuance, FOR THE NUCLEAR REGULATORY COMMISSION Ledyard B. Marsh, Chief Events Assessment, Generic Communications and Non-Power Reactors Branch Divsion of Regulatory Improvement Programs Office of Nuclear Reactor Regulation
Enclosures:
- 1. Amended Facility License
- 2. Appendix A, Technical Specifications changes Date of Issuance: January 31, 2000 4
rI"l" I'"*L"]II * * '* O°* P.8/r*14
**
- NUCLEAR REGULATORY COMMISSION
* ~WASHINGTON, D.C, 20885,=0001
,* FACILITY OPERATING LICENSE
~DOCKET NO., 50-607
#t'z*,r _REGENTS oF THE UNIVERSITY OF C.ALIFORNI*I A License No. R-130 1.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 18 and 27, 2000 comnpiies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth 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 be conducted without endangering the health and safety of the public and (II)that such activities 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 the Commission; F. The licensee is a Nonprofit Educational institution and will use the facility for educational programs arnd research, and has satief led the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements of the Commission's regulations; G. The issuance of this license will not be inimical to the common defense and security or 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; and S.The receipt, possession, and use of the byproduct and special nuclear materials as authorized by this license will be in accordance with the Commission's regulations in 10 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 Regents of the University of California (the licensee), The facility is located on the McClellan Air Force Base, Sacramento, California, I 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 of Production and Utilization Facilities," to possess, use, and operate the facility at the designated location at McClellan Air Force Base in accordance with the procedures and limitations set forth in this license. (Ii) Pursuant to the Act and 10 CFR Part 70, "Domestic Licensing of Special Nuclear Material,= 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 the format 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 be produced by the operation of the facility. (iii) Pursuant to the Act and 10 CFR Part 30, "Rules of General Applicability to Domestic Licensing of Byproduct Material," to receive, possess, and use a 4-curie sealed americium-beryllium neutron source in connection with operation of the 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 In reactor experiments, such byproduct material as may be produced by the ape ration of the facility.
C. This license shall be deemed to contain and Is su~bj~ect 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, and orders of the Commission now or hereafter in effect and to the additional conditions specified, below: (i) Maximum Po~wer Level The licensee is authorized to operate the facility at steady-state power levels not in excess of 2300 kilowatts (thermal) arid in the pulse mode with reactivity insertions not to exceed $1.75 (1.23 %/0 k/k). Amndent N..h 3
3-3 (ii) Technical S~oecfficatlonis The Technical Specifications, as revised through Amendment No. 3, are hereby. incorporated in the license. The licensee shall operate the facility in accordance with f the Technical Specifications. (lii) Physical Securityv lan The licensee shall fully implement and maintain in effect all provisions of the Commission-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 approved plan, which is exempt from public disclosure pursuant to the provisions of 10 CFR 2.790, 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) years from its date of issuance. FOR THE NUCLEAR REGULATORY COMMISSION Previously signedI by Origina/signed by Samuel J, Collins, Director office of Nuclear Reactor Regulation Date of issuance: August 13, 1998 Amendment No. 3
Q E*NCLOSURE TO LICENSEAMENDMENT NO.3 AMENDED FACILITY OPERATING LI.CENSE NO. R-!30 DOCKET NO; 50-807 Replace the following pages of Appendix A, "T'echnlcal Specificationts,= with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
*Remove 1 *1 39 39 40 40 41 4.
TECHNICAL SPECIFICATIONS FOR THE U.C. DAVIS MCCLELLAN NUCLEAR RADIATION CENTER (MNRC) General The McClellan Nuclear Radiation Center (MNRC) reactor is operated by the University of California, Davis, CA. The MNRC research reactor Is a TRIGA type reactor. The MNRC provides state-of-the-art neutron radiography capabilities. In addition, the MNR~C provides a i
- wide range of irradiation servic~es far both research and industrial needs. The reactor operates at a nominal steady start power level up to and including 2 MW. The MNRC reactor is also capable 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 ,AsLow 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 the Nuclear Regulatory Commission (e.g., senior reactor operator or reactor operator) to carry out the duties and responsibilities associated with the position requiring the license. 1.2.1 Senior.ReactorQOerator. An individual who is licensed to direct the activities of reactor operators and to manipulate the controls of the facility. 1.,2.2 Reactor Onerator. An individual who is licensed to manipulate the controls of the facility and perform reactor-related maintenance. 1.3 Ch.* A channel is the combination of sensor, line amplifier, processor, and output 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 the channel for verification that it is operable..,.' 1.3.2 Channel Calibratlaon. A channel calibration is an adjustment of the channel such that its-output corresponds with acceptable accuracy to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm or trip, and shall be deemed to include a channel test. 1.3.3 Channel Ch~eck. A channel check is a qualit~ative verification of acceptable performance by observation of channel behavior. This verification, where possible, shall include comparison of the channel with other independent channels or systems measuring the same variable. 1Amendment No .3.
b ViCECHACELOR OR ESERCHVICE CHANCELLOR FOR ADMINISTRATION ai
. U .D SAFETY K CMMITTEE s I II suPERVISOR SUPERVwSOR ...
* -- - - - - - - - - --.
[OPERATIONS STAFFI HEALTh-PHYSICS STAFF]
~UNIVERSITY MANAGEMENT ORGANIZATION
~Figure 6, !
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- TUCIEAR SAFETYL AND UCENSING NUCLEAR SAFETY AND LICENSING REVIEWS, APPROVALS AND RECOMMENDATIONS *-
COMMUNICATION OF* LICENSED ACTIVITIES UC. Davis McCleIlan Nuclear Radiatiog Center Lit'easing Organization Figure 6.2 54 An~Iendment Wo. .3
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- STATES
- S NUCLEAR REGULATORY COMMISSION
*o*_*' *WASHINGTON, 0, .0. S5-0001 Docket No. 50-607 This 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 the Atomic Energy Act of 1954, as amended (hereinafter referred to as the Act).
Article I As used in this agreement,
- 1. Nuclear reactor, byproduct material,, person, source material, specIal nuclear material, and precautionary 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 the radioactive, 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 use of property, arising out of or resulting from the radioactive, toxic,.explosive, or other hazardous properties of i, 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 the radioactive 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 as defined in any such other agreement; or ii.The radioactive material in the course of transportation and also arising out of such properties 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 wihich* is in the course 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, as amended.
- 4. In the course of transportation means Inthe course of transportation within the United States, or in the course of transportation outside the United States and any other nation, and moving from one person licensed by the Commission to another person licensed by the Commission, 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, such transportation is not by predetermination to be interrupted by the removal of the material from the 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 end when the radioactive material is removed from the transporting conveyance for any purpose other 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 of the r'adloactive material to the location if the material is also in the course of transportation from any 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 or precautionary evacuation (including all reasonable additional costs incurred by a State, or a political 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 of employees of persons indemnified who are employed (a) at the location or, if the nuclear Incident 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 loss of 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 In the 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 reactors subject to the license or licenses designated in the Attachment hereto, or (2) which is produced as 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 II
- 1. Any obligations of the licensee under subsection 53e(8.). of the Act to indemnify the United States and the Commission from public liability shall not in the aggregate exceed $250,000 with respe.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 their interests 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 person* or an act of God.
As used herein, conduct of the claimant includes conduct of persons through whom the claimant derives 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 ifsuit is instituted within 3 years from the date on which the claimant first knew, or reasonably could have known, of his injury or damage and the cause thereof.
*The waiver of any such issue or defense shall be effective regardless of whether such issue or defense may otherwise be deemed jurisdictional or relating to an element in the cause of action. The waivers shall be judicially enforceable In accordance with their te.r~ms by the claimant agaInst the person indemnified.
- 3. The waivers set forth in paragraph 2 of this article: (a) Shall not preclude a defense based upon 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 and wrongfully caused by the claimant; (c) Shall not apply to injury to a claimant who is employed at the site of and in connection with the activity where the extraordinary nuclear occurrence takes place if benefits therefor are either payable or required to be provided under any workmen~s compensationi or occupational disease law: Provided, however, That with respect to an extraordinary nuclear occurrence occurring at the facility, a claimant who is employed at the facility Inconnection with the construction of a nuclear reactor with respect to which no operating license has been issued by the Nuclear Regulatory Commission shall not be considered as employed in connection with the activity where the extraordinary nuclear occurrence takes place if:
(1) The claimant is employed exclusively in connection with the construction of a nuclear reactor, 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 transfer of nuclear material at the facility; (d) Shall not apply to anty claim for punitive or exemplary damages. provided, with respect to any claim for wrongful death under any State law which provides for damages only punitive in nature, this exclusion does not apply to the extent that the claimant has sustained actual damages, measured by the pecuniary injuries resulting from such death but not to exceed the maximum 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 claim which is not within the protection afforded under (1)the limit of liability provisions under subsection 170(e) of the Atomic Energy Act of 1954, as amended, and (b) the terms of this agreement. Article Ill
- 1. The Commission undertakes and agrees to Indemnify and hold harmless the licensee and other 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 legally liable for the nuclear incident, the Commission agrees to pay to such person those sums which such 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 with respect to: (a) Property which is located at the location and used in connection with the licensee'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, the transporting 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 to such 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 the aggregate 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 of persons legally liable for the nuclear Incident (other than such property described in the proviso to paragraph 2 of this Article) as in the aggregate exceed whichever of the following is lower: (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 as secondary financial protection, As used in this Article applicable agreements means each agreement entered into by the Commission pursuant to subsection 170(c) or (k)of the Act in which agreement the nuclear incident is defined as a common occurrence.
- 5. The obligations of the Commission under this agreement shall apply only with respect to nuclear incidents occurring during the term of this agreement.
- 6. The obligations of the Commission Uinder this and all other agreements and contracts to which 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 a common occurrence, $560,000,000 less the sum of the amounts of financial protection established under all applicable agreements.
- 7. Ifthe 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 the Commission 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 for damage to property of the licensee, shall not be affected by any failure on the part of the licensee to fulfill Its obligations under this agreement. Bankruptcy or insolvency of tihe licensee or any other person indemnified or of the estate of the licensee or any other person indemnified 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 make indemnity 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 defense of any claim Including such legal costs of the licensee as are approved by the Commission and shall have the right (a) to require the prior approval of the Commission for the settlement or payment of any claim or action asserted against: the Ilicensee or other person indemnified for public liability or damage to property of persons legally liable for the nuclear incident which claim or 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 or other 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 shall furnish 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 or transferred, without the approval of the Commission.
Article V The parties~agree that they will enter into appropriate amendments of this agreement to the extent that such amendments are required pursuant to the Atomic Energy. At of 1954, as amended, or licenses, regulations or orders of the Commission. Article VI The licensee agrees to pay to the Commission such fees as are established l~y the Commission pursuant to regulations or orders of the Commission,. Article Vii The term of this agreement shall commence as of the date and time specified in Item 4 of the Attachment and shall terminate at the time of expiration of that license specified in Item 2 of the Attachment, which is the last to expire; provided that, except as may otherwise be provided in applicable regulations or orders of the Commission, the term of this agreement shall not terminate until all the radioactive material has been removed from the location and transportation of the radioactive material from the location has ended as defined in subparagraph 4(b), Article I, Termination of the term of this agreement shall not affect any obligation of the licensee or any obligation of the Commission under this agreement with respect to any nuclear incident occurring during the term of this agreement. 4g.
FEB 12009NO.957 554p P.6/6 Attachment to Indemnity Agreement No. E-40 Item 1- Licensee University of California, Davis Address-- One Shields Avenue, Davis, California 9561648558 Item 2- License number or numbers R-130 Item 3- Location The reactor is located in the McClellan Nuclear Radiation Center Building on McClellan AFB, located approximately 8 miles northeast of Sacramento, California. Item 4-.. The indemnity agreement designated above, of which this Attachment Is a 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 Branch Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Dated at Rock'ville, MD, the ,* day of * ,2000. _________________By Kevin Smith Vice C~hancelior University of California, Davis
Fz~? 0 UNITED STATES
/
%" NUCLEARWASHINGTON, REGULATORY COMMISSION D.C. 20555-0001
*o*,'*August 9, 2001 Dr. Kevin Smith, Vice Chancellor Office of the Chancellor University of California, Davis One Shields Avenue Davis, 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 Facility 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 May 11, 2001. The amendment reflects the administrative changes to the TSs as a result of the transfer of the license 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 and which are discussed in the enclosed safety evaluation report. Sincerely, Warren J. Eresian, Project Manager Operational Experience and Non-Power Reactors Branch Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No. 50-607
Enclosures:
- 1. Amendment No. 4
- 2. Safety Evaluation cc w/enclosures:
Please see next page
University of California - Davis/McClellan MNRC Docket No. 50-607 co: Dr. Wade J. Richards 5335 Price Avenue, Bldg. 258 McClellan AFB, CA 95652-2504 Test, Research, and Training Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611
- "* w*'oUNITEDNUCLEAR REGULATORY STATES COMMISSION
.* WASHINGTON, D.C. 20555-0001 REGENTS OF THE UNIVERSITY OF CALIFORNIA AT McCLELLAN NUCLEAR RADIATION CENTER DOCKET NO. 50-607 AMENDMENT TO AMENDED FACILITY OPERATING LICENSE Amendment No. 4 License No. R-130
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that A. 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 Nuclear Radiation Center (the licensee) on May 11, 2001, conforms to the standlards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the regulations of the Commission as stated in Chapter I of Title 10 of the Code of FederalRegulations (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 the Commission; D. The issuance of this amendment will not be inimical to the common defense and security 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; and F. 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, license is to the enclosure amended by changes to the Technical Specifications as indicated in the this license amendment, and paragraph 2.C.(ii) of Amended Facility 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 through Amendment No. 4, are hereby incorporated in the license. The licensee shall 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 Warren J. Eresian, Project Manager Operational Experience and Non-Power Reactors Branch Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation
Enclosure:
Appendix A, Technical Specification Changes Date of Issuance: August 9, 2001
S 0 ENCLOSURE TO LICENSE AMENDMENT NO. 4 AMENDED FACILITY OPERATING LICENSE NO. R-130 DOCKET NO. 50-607 Replace the following pages of Appendix A, Technical Specifications, with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. Remove Insert ii ii iii iii iv iv V v vi vi 1 I 2 2 3 3 4 4 6 6 7 7 9 9 13 13 14 14 15 15 16 16 17 17 18 18 19 19 25 25 26 26 27 27 28 28 29 29 30 30 31 31 32 32 33 33 34 34 35 35 36 36 39 39 40 40
UNITED STATES 1"%" NUCLEAR REGULATORY COMMISSION
~WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 4 TO AMENDED FACILITY OPERATING LICENSE NO. R-130 REGENTS OF THE UNIVERSITY OF CALIFORNIA AT McCLELLAN NUCLEAR RADIATION CENTER DOCKET NO. 50-607
1.0 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, to Facility 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 request under oath. The resubmittal contained no new information.) The request provides for the following 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 the Department of the Air Force to the Regents of the University of California. As a result of this transfer, a nUmber of administrative changes simply involving name changes (e.g., changing references from "Responsible Commander" to "Vice Chancellor of the Office of Research" 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 channel calibrations 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 loading termed a 30B core.
- 6. Section 6.0. A request to revise the organization and duties of the Nuclear Safety Committee and to clarify the Committee's review and audit functions to reflect the new licensee.
- 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 of the Air Force to the Regents of the University of California, the TSs must be modified to take account of administrative changes. These changes will occur in a number of places, and consist of the substitution of Department of the Air Force organizational and position titles with corresponding 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 and Operation." 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 are acceptable. 2.2 Section 2.1, Basis b. The previous version of the Technical Specifications addressed the issue of the effect of pulsing on fuel clad integrity and concluded that TRIGA fuel of the type used in the McClellan reactor could 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 present analysis expands the discussion to include more recent measurements of hydrogen pressure resulting from pulses and concludes that the cladding will not rupture if fuel temperatures are never greater than 1200 to 1250 degrees Centigrade, providing the cladding temperature is less than 500 degrees Centigrade. Since the pulse reactivity limit remains at $1.75, the staff concludes 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 by Research Reactor Safety Analysis Services (RRSAS-99-6-1, December 1999) to determine if the conservative maximum core inlet temperature (heat exchanger outlet temperature) as set by the U.S. Air Force in the original design could be raised from 35 degrees Centigrade to 45 degrees Centigrade. The effect of the lower limit is that the reactor power is required to be reduced below the license limit of 2 MW whenever ambient local weather conditions prevent the system 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 state operations, when compared with previous calculations by Argonne National Laboratory, General Atomics published reports, and results from power upgrades at the Sandia Annular Core
0 Research Reactor facility shows that the maximum core inlet temperature can be raised to 45 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. The calculations also show that there is no increase in the maximum fuel temperature or the maximum fuel clad surface temperature, two of the most important parameters which measure fuel integrity. Accordingly, the staff concludes that safety limits will not be reduced and that there 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 the stack monitor system, the reactor room radiation monitor and the reactor room continuous air monitor. These systems are presently required to be calibrated semiannually. The licensee has requested that they be calibrated annually. The requirement for semiannual calibrations stems from the original Department of the Air Force licensing organization, but has no operational safety basis. Research reactors of similar power levels currently licensed by the NRC (National Institute of Standards and Technology, Rhode Island AEC) are permitted to calibrate similar instruments on an annual basis, since there are no operating experience data to suggest that this practice would compromise safety. In addition, the American National Standard ANSI/ANS 15.11 "Radiation Protection at Research Reactor 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 was operating with a mixed core of 8.5/20 and 20/20 fuel loading (referred to as the MixJ core in the original SAR). At that time it was understood that the reactor would eventually transition to a core consisting of 20/20 and 30/20 fuel, termed a 30B core. The 30B core was analyzed in the original SAR and found to be acceptable by the NRC staff in the SER. In addition, the NRC staff had previously approved the generic use of TRIGA fuels with uranium loadings of up to 30 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 the operation and maintenance of the reactor and associated equipment. There are a number of minor changes with respect to titles and some changes with .respect to the composition and duties of the Nuclear Safety Committee (NSC). The review and inspection functions of the NSC have been expanded to provide additional oversight. These expanded functions include review of the Emergency Plan and Physical Security Plan, review and update of the NSC Charter every two years, review of inspections conducted by other agencies, assessment of actions taken to correct deficiencies, inspection of currently active experiments, and inspection of future plans for facility modifications or facility utilization. Since these changes increase oversight of facility operations, the staff concludes that they are acceptable.
0 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 the installation of an Iodine-125 production loop. The purpose of the loop is to produce from ten to twenty 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 the reactor core.
- 2. The Xenon-I 24 is irradiated over an eight to sixteen hour time span and by neutron activation results in the production of Xenon-125. The activated Xenon-I124 gas contains 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 decays with a 17-hour half-life to Iodine-I125. After a few days, most of the Xenon-I125 has decayed 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 to as decay storage 2.
- 5. The Iodine-125 in decay storage I is recovered by washing the tank with a NaOH solution, resulting in a Nal solution which is packaged as a liquid and sent to an off-site user in an appropriate DOT container.
All equipment used in the production loop is located within a primary containment and a secondary containment. The primary containment houses the irradiation chamber, tubing, pneumatically operated valves, transfer vessel, decay storage I and decay storage 2. The secondary containment is placed around the primary containment to the irradiation chamber and allows 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 these containments are within the reactor room, which has a ventilation system with isolation/recirculation capability. There are two other structures within the reactor room which are confinement barriers designed for the safety of personnel working with the production loop. The first is a glove box which contains controls for operation of the Iodine-125 recovery system. The glove box has its own ventilation and filtration system which exhausts into the reactor room ventilation system. The second is a fume hood in which quality assurance of the Iodine-125 is performed. The fume hood also contains its own ventilation and filtration system which exhausts into the reactor room ventilation system. The licensee has analyzed the situation (worst-case) whereby all of the Xenon-I125 from the primary containment leaks into the secondary containment and subsequently leaks into the reactor room at the design leak rate of the secondary containment. Their analysis shows that exposures to personnel in the reactor room would result in a deep dose equivalent (DDE) of 17 millirem 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 at the 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 cladding rupture of one highly irradiated fuel element with no decay followed by instantaneous release of fission products into the air. At the closest distance to the site boundary (10 meters), the maximum 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 the margin of safety with respect to 10 CFR 20 limits and that the installation of the production loop is 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 of Iodine-125 present in the reactor room glove box and the reactor room fume hood. Limiting the amount of Iodine-125 in these areas will reduce the occupational dose and dose to personnel in the unrestricted areas to less than 10 CFR 20 limits if the inventories of Iodine-I125 are totally released 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 located within 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 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 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 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 or consequences of accidents previously evaluated, or create the possibility of a new or different kind of accident from 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 is reasonable 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 or the health and safety of the public. Principal Contributor: Warren J. Eresian Date: August 9, 2001
0I 0 TECHNICAL SPECIFICATIONS FOR THE UNIVERSITY OF CALIFORNIA - DAVIS MCCLELLAN NUCLEAR RADIATION CENTER (UCD/MNRC) DOCUMENT NUMBER: MNRC-0004-DOC-1 1 Rev 11, 12/10/99 Amendment No. 4 i
0 TECHNICAL SPECIFICATIONS APPROVAL These "Technical ReactorSpecifications" have undergone the University for the of California at Davis/McClellan Nuclear Radiation Center following coordination: (UCD/MNRC) Reviewed bY:H*alF* z 'ODte Reviewed by: (*-- k_,. Q- _* Reactor Operations S~pervisor (Date) Approved by: 9 ,U4 UCD/MNI Director (Date) Approved by:________________ Chairman, UCD/MNRC (Date) Nuclear Safety Committee Amendment No. 4 ii
0 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Page 1.0 Definitions...................................................................................................................1 2.0 Safety Limit and Limiting Safety System Setting (LSSS)............................................................... 5 2.1 Safety Limits*...................................................................................................... 5 2.2 Limiting Safety System Selling (LSSS)......................................................................... 6 2.2.1 Fuel Temperature .................................................................................... 6 3.0 Limiting Conditions for Operations (LCO) ............................................................................... 7 3.1 Reactor Core Parameters ....................................................................................... 7 3.1.1 Steady-State Operation.................................................. *............................ 7 3.1.2 Pulse or Square Wave Operation................................................................... 7 3.1.3 Reactivity Limitations................................................................................. 8 3.2 Reactor Control and Safety Systems........................................................................... 8 3.2.1 Control Rods........................................................................................... 8 3.2.2 Reactor Instrumentation ............................................................................. 9 3.2.3 Reactor, Scrams and Interlocks .................................................................... 10 3.2.4 Reactor Fuel Elements ............................................................................. 12 3.3 Reactor Coolant Systems...................................................................................... 13 3.4 Reactor Room Exhaust System ............................................................................... 14
.3.5 Intentionally Left Blank .................................... ..................................................... 14 3.6 Intentionally Left Blank ......................................................................................... 14 3.7 Reactor Radiation Monitoring Systems ....................................................................... 14 3.7.1 Monitoring Systems................................................................................. 14 3.7.2 Effluents - Argon-41 Discharge Limit .............................................................. 16 Amendment No. 4 iii
Pane 3.8 Experiments ..................................................................................................... 16 3.8.1 Reactivity Limits ..................................................................................... 16 3.8.2 Materials Limit ....................................................................................... 17 3.8.3 Failure and Malfunctions ........................................................................... 18 4.0 Surveillance Requirements .................................................................. ........................... 19 4.1 Reactor Core Parameters...................................................................................... 19 4.1.1 Steady State Operation ............................................................................ 19 4.1.2 Shutdown Margin and Excess Reactivity.......................................................... 20 4.2 Reactor Control and Safety Systems ......................................................................... 20 4.2.1 Control Rods......................................................................................... 20 4.2.2 Reactor Instrumentation ............................................................................ 21 4.2.3 Reactor Scrams and Interlocks .................................................................... 22 4.2.4 Reactor Fuel Elements ............................................................................. 23 4.3 Reactor Coolant Systems...................................................................................... 24 4.4 Reactor Room Exhaust Systerm........ ....................................................................... 25 4.5 Intentionally Left Blank ......................................................................................... 25 4.6 Intentionally Left Blank ......................................................................................... 25 4.7 Reactor Radiation Monitoring Systems ....................................................................... 25 4.8 Experiments ..................................................................................................... 26 5.0 Design Features.......................................................................................................... 27 5.1 Site and Facility Description.................................................................................... 27 5.1.1 Site.................................................................................................... 27 5.1.2 Facility Exhaust...................................................................................... 28 5.2 Reactor Coolant System........................................................................................ 28 Amendment No. 4 iv
0 0 Page 5.3 Reactor Core and Fuel ........................................................................................... 29 5.3.1 Reactor Core........................................................................................... 29 5.3.2 Reactor FueL........................................................................................... 30 5.3.3 Control Rods and Control Rod Drives .............................................................. 31 5.4 Fissionable Material Storage .................................................................................... 31 6.0 Administrative Controls..................................................................................................... 31 6.1.1 Structure................................................................................................ 32 6.1.2 Responsibilities........................................................................................ 32 6.1.3 Staffing.................................................................................................. 32 6.1.4 Selection and Training of Personnel ................................................................ 32 6.2 Review, Audit, Recommendation and Approval............................................................... 32 6.2.1 NSC Composition and Qualifications............................................................... 33 6.2.2 NSC Charter and Rules` ............................................................................. 33 6.2.3 Review Functiont...................................................................................... 33 6.2.4 Audit/Inspection Function ............................................................................ 34 6.3 Radiation Safety. ............................................. ..................................................... 34 6.4 Procedures ........................................................................................................ 34 6.4.1 Reactor Operations Procedures........ .................. ........................................... 34 6.4.2 Health Physics Procedures .......................................................................... 35 6.5 Experiment Review and Approval............................................................................... 35 6.6 Required Actions.................................................................................................. 35 6.6.1 Actions to be taken in case of a safety limit violation.............................................. 35 6.6.2 Actions to be taken for reportable occurrences`................................................... 36 Amendment No. 4 V
6.77Rep Ortstn R po t ..................................................... ........ 36 6.7.2 Special Reports........................................................................................ 38 6.8 Records ............................................................................................................. 39 Fig. 6.1 UCD/MNRC Organization for Licensing and Operation*........................................................... 40 Amendment No. 4 vi
0 TECHNICAL SPECIFICATIONS FOR THE UNIVERSITY OF CALIFORNIA - DAVIS/MCCLELLAN NUCLEAR RADIATION CENTER (UCD/MNRC) General The University of California - Davis/McClellan Nuclear Radiation Center (UCD/MNRC) reactor is operated by the University 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 provides a wide range of irradiation services for both research and industrial needs. The reactor operates at a nominal steady state power level up to and including 2 MW. The UCD/MNRC reactor is also capable of square wave and 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 Nuclear Regulatory Commission (e.g., senior reactor operator or reactor operator) to carry out the duties and responsibilities associated with the position requiring the license. 1.2.1 Senior Reactor Operator. An individual who is licensed to direct the activities of reactor operators 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 and perform reactor-related maintenance. 1.3 Channel. A channel is the combination of sensor, line amplifier, processor, and output 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 the channel for verification that it is operable. 1.3.2 Channel Calibration. A channel calibration is an adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm or trip, and shall be deemed to include a channel test. 1.3.3 Channel Check. A channel check is a qualitative verification of acceptable performance by observation of channel behavior. This verification, where possible, shall include comparison of the channel with other independent channels or systems measuring the same variable. 1.4 Confinement. Confinement means isolation of the reactor room air volume such that the movement of 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, fission chambers, foils, etc), which is designed to investigate specific reactor characteristics or which is intended for irradiation within an experiment facility and which is not rigidly secured to a core or shield structure so as to be a part of their design. 1.5.1 Experim~ent. Moveable. A moveable experiment is one where it is intended that the entire experiment may be moved in or near the reactor core or into and out of reactor experiment facilities while the reactor is operating. Amendment No. 4 1
- 0 1.5.2 Experiment. Secured. A secured experiment is any experiment, experiment facility, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means. The restraining force rmust be substantially greater than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the experiment, or by forces which can arise as a result of credible 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 their associated input circuits, which are designed to initiate a scram for the primary purpose of protecting an experiment or to provide information which requires manual protective action to be initiated. 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 can be either 8.5, 20, or 30 wt%, with an enrichment of less than 20% U-235. A standard fuel element may contain a burnable poison. 1.7 Fuel Element. Instrumented. An instrumented fuel element is a standard fuel element fabricate~d with thermocouples for temperature measurements. An instrumented fuel element shall have at least one operable 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 a channel. 1.9 Mode, Steady-State. Steady-state mode operation shall mean operation of the UCD/MNRC reactor with 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/MNRC reactor 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 the selector switch 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 the UCD/MNRC reactor safety system in response to a variable or condition of the UCD/MNRC reactor facility having 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 the generation 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. 4 2
1.15.3 Instrument generation System Level. and transmission At the protective of the command signal for instrument level, protective the safety shutdown equipmentaction is the to operate. 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 QOerational Core. A pulse operational core is a reactor operational core for which the maximum 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 reactor is at ambient temperature and the reactor is critical. (K*, = 1) 1.18 Reactivity Limits. The reactivity limits are those limits imposed on the reactivity conditions of the reactor core. 1.19 Reactivity Worth of an Experiment. The reactivity worth of an experiment is the maximum value of the reactivity 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 which directly 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 the parameters 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 in these Technical 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 input channels, 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 off position and the key.is removed from the lock and under the control of a licensed operator, and the conditions of a or b exist:
- a. (1) The minimum number of control rods are fully inserted to ensure the reactor is shutdown, as required by 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 being moved or serviced ifthe experiments have, on movement, a reactivity worth exceeding the maximum value allowed for a single experiment or $1.00, whichever is smaller, or
- b. The reactor contains insufficient fissile materials in the reactor core, adjacent experiments or control rods to attain criticality under optimum available conditions of moderation and reflection.
1.25 _Reactor Shutdown. The UCD/MNRC reactor is shutdown ifit 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. Amendment No. 4 3
- 0 1.26 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 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, and which 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 a fuel 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 allow it to perform its control function, and its safety function when the coupling is disengaged. This safety function is commonly termed a scram. 1.28.1 Regqulatingq Rod. A regulating rod is a Control rod'used to maintain an intended power level and 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 of radioactivity. 1.31 Scram Time. Scram time is the elapsed time between reaching a limiting safety system set point and the 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 denote a 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 to provide confidence that the reactor can be made subcritical by means of the control and safety system starting from any permissible operating condition with the most reactive rod assumed to be in the most reactive position, and once this action has been initiated, the reactor will remain subcritical without further operator 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 manual shutdown 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 checks and tests, and calibrations. Operability checks and tests are generally specified as daily, weekly or 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 surveillance interval is the interval between a check, test, or calibration, whichever is appropriate to the item Amendment No. 4 4
0 fuel element temperature. This parameter is well suited as it can be measured directly. A loss in the integrity of the fuel element cladding could arise ifthe cladding stress exceeds the ultimate strength of the cladding material. The fuel element cladding stress is a function of the element's internal pressure while the ultimate strength of the cladding material is a function of its temperature. The cladding stress is a result of the internal pressure due to the presence of air, fission product gasses and hydrogen from the disassociation 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 hydrogen to zirconium in the alloy. At a fuel temperature of 930°C for ZrH1 .7 fuel, the cladding stress due to the internal pressure is equal to the ultimate strength of the cladding material at the same temperature (SAR Fig 4.18). This is a conservative limit since the temperature of the cladding material is always 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 50000. Analysis (SAR Chapter 4, Section 4.5.4.1.1), shows that a maximum temperature for the clad during a pulse which gives a peak adiabatic fuel temperature of 100000 is estimated to be 470°C. Further analysis (SAR Section 4.5.4.1.2), shows that the internal pressure for both Zr. 65 (at 115000) and Zr1 .7 (at 1100°C) increases to a peak value at about 0.3 sec, at which time the pressure is about one-fifth of the equilibrium value or about 400 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 were approximately the same as for the 115000 case. Similar results were found for fuel with ZrH1 .7. Measurements of hydrogen pressure in TRIGA fuel elements during transient operations have been made and compared 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 100000, the pressure (ZrH1 .65 ) was only about 6% of the equilibrium value evaluated at the peak temperature. Calculations of the pressure gave values about three times greater than the measured values. The analysis gives strong indications that the cladding will not rupture if fuel temperatures are never greater than 120000 to 125000, providing the cladding temperature is less than 5000 C. For fuel with ZrH 1.7 ,a conservative safety limit is 110000. As a result, at this safety limit temperature, the class pressure 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 element temperature. 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 more conservatively) as measured in an instrumented fuel element. One instrumented element shall 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 a temperature which, ifexceeded, shall cause a reactor scram to be initiated preventing the safety limit from being exceeded. A setting of 75000 provides a safety margin at the point of the measurement of at least 13700 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 measured temperatures 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 the center and mid-plane of the fuel element. For pulse operation of the reactor, the same limiting safety system setting applies. However, the temperature channel will have no effect on limiting ;the peak power generated because of its Amendment No. 4 6
0 ! relatively long time constant (seconds) as compared with 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 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 the purpose 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/MNRC TRIGA 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 result 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% 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 phase changes at temperatures in excess of 1000 C. However, after extensive steady state operation at two (2) MW the hydrogen will redistribute due to migration from the central high temperature regions of 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 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. Ifthe pulse insertion is such that the temperature of the fuel exceeds about 8750 C, then the pressure may be sufficient to cause expansion of microscopic holes in the fuel that grow with each pulse. Analysis (SAR Chapter 13, Section 13.2.2.2.1), shows that the limiting pulse, for the worst case conditions, is 1.34% Ak/k ($1.92). Therefore, the 1.23% Ak/k ($1.75) limit is below the worse case reactivity insertion accident limit. The $0.90 square wave step insertion limit is also well below the worse case reactivity insertion accident limit. Amendment No. 4 7
0 Basis -
- a. The apparent condition of the control rod assemblies shall provide assurance that the rods shall continue to perform reliably as designed.
- b. This assures that the reactor shall shut down promptly when a scram signal is initiated (SAR Chapter 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 reactor operator during reactor operations. Objective - The objective is to require that sufficient information is available to the operator to assure safe operation of the reactor. Specification - The reactor shall not be operated unless the channels described in Table 3.2.2 are operable and the information is displayed on the reactor console. Table 3.2.2 Required Reactor Instrumentation (Minimum Number Operable) Measuring Steady Square Channel Surveillance Channel State Wave Function Requirements* Pulse
- a. Reactor Power 2 0 2 Scram at 2.2 D,M,A Level Safety MW or less Channel
- b. Linear Power Automatic D,M,A 1 0 Channel Power Control
- c. Log Power 1 0 Startup D,M,A Channel Control
- d. Fuel Temperature 2 2 Fuel DM,A 2
Channel Temperature
- e. Pulse Channel I Measures PA Pulse NV & NVT
(*)Where: D - Channel check during each day's operation M - Channel test monthly A - Channel calibration annually P - 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 is properly monitored and indicated in the reactor control room (SAR Chapter 7, Sections 7.1.2 &
7.1.2.2). Amendment No. 4 9
- 0 3.3 Reactor Coolant Systems Applicability - 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 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 channels described in Table 3.3 are operable, and the information is displayed locally or in the control room. Table 3.3 REQUIRED WATER SYSTEMS AND INSTRUMENTATION Minimum Measuring Number Surveillance Channel/System Operable Requirements* Function: Channel/System
- a. Primary Coolant For operation of the D,Q,A Core Inlet reactor at 1.5 MW or Temperature higher, alarms on high Monitor heat exchanger outlet temperature of 45°C (1 130 F)
- b. Reactor Tank Alarms ifwater level M Low Water. drops below a depth of Monitor 23 feet in the reactor tank
- c. Purification** Alarms ifthe primary D,M,S Inlet Conduc- coolant water conductivity tivity Monitor is greater than 5 micromhos/cm
- d. Emergency Core For operation of the reactor D,S Cooling System at 1.5MW or higher, provides water to cool fuel in the event of a Loss of Coolant Accident for a minimum of 3.7 hours at 20 gpm from an appropriate nozzle
(*)Where: D - channel check during each day's operation A - channel calibration annually Q - 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 shall be shutdown. Amendment No. 4 13
Basis -
.a. Table 3.3. The primary coolant core inlet temperature alarm assures that large power fluctuations will not occur (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 is 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 ifthe water level drops below a height of 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 a week 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 System Applicability - 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 reactor 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.
Specification -
- a. The reactor shall not be operated unless the reactor room exhaust system is in operation and the pressure 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 Coolant 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 room will be reduced due to mixing with exhaust system air (SAR Chapter 9, Section 9.5.1). Pressure in the reactor room will be negative relative to surrounding areas due to air flow patterns created, by the reactor room 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 fuel temperature 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 Systems 3.7.1 Monitorinq Systems Applicability - This specification applies to the information which shall be available to the reactor operator during reactor operation. Amendment No. 4 14
0 Objective - 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.
.Specification - The reactor shall not be operated unless the channels described in Table 3.7.1 are operable, the readings are below the alarm setpoints, and the information is displayed in the control room. The stack and reactor room CAMS shall not be shutdown at the same time during reactor operation.
Table 3.7.1 REQUIRED RADIATION MONITORING iNSTRUMENTATION Minimum Measuring Number Channel Surveillance Equipment Operable** Function Requirements*
- a. Facility I Monitors Argon-41 and D,W,A Stack Monitor radioactive particu-lates, and alarms
- b. Reactor Room 1 Monitors the radiation D,W,A Radiation level in the reactor Monitor room and alarms
- c. Purification 1 Monitors radiation D,W,A System Radia- level at the demineral-tion Monitor izer station and alarms
- d. Reactor Room 1 Monitors air from the D,W,A Continuous reactor room for parti-Air Monitor culate and gaseous radioactivity and alarms
(*)Where: AD -- channel channel calibration check during each day's operation annually W - 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 release of radioactive material to the environment (SAR Chapter 11, Section 11.1.1.1.4). The alarm setpoint 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 the operations area.
- b. Table 3.7.1. The reactor room radiation monitor provides information regarding radiation levels in the reactor 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 provides information regarding radioactivity in the primary system cooling water (SAR Chapter 11, Section 11.1.5.4.2)
Amendment No. 4 15
and allows assessment of radiation levels in the area to ensure that personnel radiation doses will be below 10 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 by the discharge of Argon-41 from the UCD/MNRC reactor facility. Specification - The annual average unrestricted area concentration of Argon-41 due to releases of this radionuclide from the UCD/MNRC, and the corresponding annual radiation dose from Argon-41 in 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 is specified in Appendix B, Table 2, Column 1 of 10 CER Part 20. 10 CER 20.1301 specifies dose limitations in the unrestricted area. 10 CFR 20.1101 specifies a constraint on air emissions of radioactive material to the environment. The SAR Chapter 11, Section 11.1.1.1.4 estimates that the routine Argon-41 releases and the corresponding doses in the unrestricted area will be below these limits. 3.8 Experiments 3.8.1 Reactivity Limits. Aoplicability - This specification applies to the reactivity limits on experiments installed in specific reactor experiment facilities. Obiective - The objective is to assure control of the reactor during the irradiation or handling of experiments 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-tank irradiation 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 or in-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 in any other reactor in-core and in-tank irradiation facilities at any given time shall be less than one dollar and ninety-two cents ($1.92) (1.34% Ak/k), including the potential reactivity which might result from malfunction, flooding, voiding, or removal and insertion of the experiments.
Amendment No. 4 16
Basis -
- a. A limitation of less than dollar ($1.00) onepneumatic (0.7% Ak/k) on the reactivity worth of a single movable experiment positioned in the 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 limit of $1.75 is not exceeded (SAR Chapter 13, Section 13.2.2.2.1). In addition, 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 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 secured experiment is its sudden accidental or unplanned removal while the reactor is operating. For such 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 Chapter 10, 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 control over the positioning of samples into and out of the central core region 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 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 the fuel temperature safety limit.
3.8.2 Materials Limit 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 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 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 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 200 millicuries. An additional 800 millicuries of 1-125 in sealed storage containers may also be present in the reactor room fume hood.
Amendment No. 4 17
S
- e. Explosive materials in quantities greater than 25 milligrams of TNT equivalent shall not be irradiated 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 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 any radiography bay. The irradiation of explosives in any bay is limited to those assemblies where a safety analysis has been performed that shows that there is no damage to the 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 of materials.
- b. The 1.5 curies limitation on iodine 131 through 135 assures that in the event of failure of a fueled experiment leading to total release of the iodine, occupational doses and doses to members 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 twenty (20.0) curies in the reactor room glove box and to one (1.0) curie in the reactor room fume hood assures that, ifthese inventories of 1-125 are totally released into their respective containments, occupational doses and doses to members of the general 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 of 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 any radiography bay external to the reactor tank will not result in damage to the reactor controls or the 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 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 and type of material in the experiment shall be limited such that the airborne radioactivity in the reactor room will not result in exceeding the applicable dose limits in 10 CFR Part 20 in the unrestricted area, assuming 100% of the gases or aerosols escapes. Amendment No. 4 18
0
- 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 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 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 by boiling this material can escape only through an undistributed column of water above the core, at least 10% of these vapors can escape.
- c. If a capsule fails and releases material which could damage the reactor fuel or structure by corrosion 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 his designated 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 reactor room 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 an experiment 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 be brought to the attention of the UCD/MNRC Director or his designated alternate for review to assure 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 reactor operating 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 time period. Prior to reactor system or measuring channel operation, the surveillance shall be performed for each reactor 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 level monitoring channels. Obiective - The objective is to verify that the maximum power level of the reactor does not exceed the authorized limit. Amendment No. 4 19
0 0 Discovery of noncompliance with Technical Specifications 4.3.a-c shall limit operations to that required to perform the 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 input signal. 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 input 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 for power levels above 1.5 MW. A channel calibration semiannually assures that the Emergency Core Cooling System performs as required for power levels above 1.5 MW.
4.4 Reactor Room Exhaust System Applicability - This specification applies to the surveillance requirements for the reactor room exhaust 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's operation. 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 verify that the exhaust system is maintaining a negative pressure in the reactor room relative to the surrounding facility areas. 4.5 This section intentionally left blank 4.6 This section intentionally left blank. 4.7 Reactor Radiation Monitorinq Systems Applicability - 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. 4 25
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 perform the surveillance. Basis -
- a. A channel check of the facility stack monitor system during each day's operation will assure the monitor is operable. A channel test weekly will assure that the system responds correctly to a 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 assure that the monitor is operable. A channel test weekly will ensure that the system responds to a known source. 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 assures that the monitor is operable. A channel test weekly will ensure that the system responds to a known source. 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 responds correctly to a known source. A channel calibration annually will assure that the CAM reads correctly.
4.8 Experiments Applicability - This specification applies to the surveillance requirements for experiments installed in any UCD/MNRC reactor experiment facility. Amendment No. 4 26
0 Objective - The objective is to prevent the conduct of experiments or irradiations which may damage the 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 a written safety analysis has been performed and reviewed by the UCD/MNRC Director, or his designee, 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 Facility Use Authorization. Facility Use Authorizations and experiments carried out under these authorizations 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/MNRC experiment facility, until it has been reviewed for compliance with the approved experiment and Facility Use Authorization 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 other UCD/MNRC reactor in-core or in-tank irradiation facility shall be estimated or measured, as 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 in any UCD/MNRC reactor experiment facility.
Basis - a & b. Experience at most TRIGA reactor facilities verifies the impo'rtance of reactor staff and safety committee reviews of proposed experiments.
- c. Measurement of the reactivity worth of an experiment, or estimation of the reactivity worth based 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 will facilitate maintaining surveillance over such experiments.
5.0 Desigqn Features 5.1 Site and Facility Description. 5.1.1 Sit..ee Applicability - This specification applies to the UCD/MNRC site location and specific facility design features. Objective - The objective is to specify those features related to the Safety Analysis evaluation. Specification - Amendment No. 4 27
0
- a. The site location is situated approximately 8 miles (13 kin) north-by-northeast of downtown Sacramento, 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. The unrestricted 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 has 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 of the site 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 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 to be beyond a credible accident scenario.
5.1.2 Facility Exhaust 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 of radioactivity released into the environment, or during a Loss of Coolant Accident, the system is to 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 and exhaust air from the UCD/MNRC facility. The system shall have an exhaust stack height of a minimum of 18.2m (60 feet) above ground level.
- b. Manually activated shutdown controls for the exhaust system shall be located in the reactor control room.
Basis - The UCD/MNRC facility exhaust system is designed such that the reactor room shall be maintained at a negative pressure with respect to the surrounding areas. The free air volume within the 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 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 to operating personnel. 5.2 Reactor Coolant System Applicability - This specification applies to the reactor coolant system. Amendment No. 4 28
0 Objective - The objective is to assure that adequate water is available for cooling and shielding during normal 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 ifthe water level in the reactor tank drops below a depth of 23 ft.
- c. For operations at 1.5 MW or higher during a Loss of Coolant Accident the reactor core shall be cooled for 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 be exceeded under natural convection flow conditions.
- b. A reactor tank water low level alarm sounds when the water level drops significantly. This alarm annunc~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 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 Reactor Core Applicability - This specification applies to the configuration of the fuel. Objective - The objective is to assure that provisions are made to restrict the arrangement of fuel elements 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 96 or more fuel elements to include fuel followed control rods. Below 0.5 MW there is no minimum 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 be loaded 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 20E core configurations. Amendment No. 4 29
0 S 308 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 done before 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 above loading restrictions will be allowed in an operational 0.5 MW or greater core. Specifications for the 20E core and 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 Fuel Applicability - 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 their physical and 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 to 2.3 megawatts steady state or to a 36 megawatt-sec pulse assures an ample margin of safety between 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 equal to 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 normal operation, 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 of purchased materials.
Amendment No. 4 30
0 0 5.3.3 Control Rods and Control Rod Drives Applicability - This specification applies to the control rods and control rod drives used in the reactor core. Obiective - The objective is to assure the control rods and control rod drives are of Such a design 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, 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 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, B4 C powder, or boron and its compounds. These materials shall be contained in a suitable clad material such as stainless steel or aluminum to assure mechanical stability during movement and to isolate the neutron poison from the tank water environment. 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 at the 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 Storaqe Applicability - This specification applies to the storage of reactor fuel at a time when it is not in the reactor core. Obiective - The objective is to assure that the fuel which is being stored will not become critical and will not 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 the keff 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 Controls 6.1 Orqanization. The Vice Chancellor for Research shall be the licensee for the UCD/MNRC. The UCD/MNRC facility shall be under the direct control of the UCD/MNRC Director or a licensed senior reactor operator (SRO) designated by the UCD/MNRC Director to be in direct control. The UCD/MNRC Director shall be accountable to the Vice Chancellor of the Office of Research for the safe operation and maintenance of the reactor and its associated equipment. Amendment No. 4 31
0 6.1.1 Structure. The management for operation of the UCD/MNRC facility shall consist of the organizational structure as shown in Figure 6.1. 6.1.2 Responsibilities. The UCD/MNRC Director shall be accountable to the Vice Chancellor of the 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 all experiments and experiment*{ procedures prior to their use in the reactor. Individuals in the management organization (e.g., Reactor Manager, Health Physics Manager, etc.) 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 the operating license and technical specifications. The Reactor Manager and Health Physics Manager report directly to the UCD/MNRC Director. 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 reactor operator should be within thirty (30) minutes of the facility and reachable by telephone, and;
- d. A senior reactor operator shall be present whenever a reactor startup is performed, 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 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 and Training of Personnel for Research Reactors (ANS 15.4). Qualification and requalification of licensed operators shall be subject to an approved Nuclear Regulatory Commission (NRC) program. 6.2 Review. Audit. Recommendation and Approval General Policy. Nuclear facilities shall be designed, constructed, operated, and maintained in such 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 with applicable regulatory requirements. The UCD Vice Chancellor of the Office of 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. Amendment No. 4 32
0 6.2.1 NSC Composition and Qualifications. The UCD/MNRC Director shall appoint the Chairperson 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 shall evaluate 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 proposed Class I modifications, prior to implementing (Class I) modifications, prior to taking action under the preceding 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 an unreviewed safety question or require a change to the license, to a Facility Use Authorization, or to the Technical Specifications. This determination may be in the form of verifying a decision already made by the UCD/MNRC Director.
- d. Review reactor operations and operational maintenance, Class I modification records, and 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 for UCDIMNRC 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 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 other inspections of these facilities conducted by other agencies.
Amendment No. 4 33
0 6.2.4 AuditlInspection Function. The NSC or a subcommittee thereof, shall audit/inspect reactor operations 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 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 and Technical 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 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 the UCD/MNRC Radiation Safety Program. The program should use the guidelines of the American National Standard for Radiation Protection at Research Reactor Facilities (ANSI/ANS 15.11). The Health Physics Manager 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 of procedures shall be performed and documented in a timely manner by the UCD/MNRC staff to assure that procedures are current. Procedures shall be adequate to assure the safe operation of the 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 or other routine maintenance that could have an effect on reactor safety.
- e. Testing and calibration of reactor instrumentation and controls, control rods and control rod drives.
Amendment No. 4 34
f.that Administrative controlssafety could affect reactor for operations, maintenance, and conduct of irradiations and experiments 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 alarms 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 solid and liquid waste and gaseous effluents released from the facility. The program shall include a management 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 items from 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 Approval. Experiments having similar characteristics are grouped together for review 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/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 approval as a new experiment.
- c. Minor changes to an experiment that do not significantly alter the experiment may be approved by a senior reactor operator.
6.6 Required Actions 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: Amendment No. 4 35
0.Q
- 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 the UCD/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 and contributing factors. (2) Effect of the violation upon reactor facility components, systems, or structures, and on 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 the NRC 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 is necessary to shut down the reactor to correct the occurrence, operations shall not be resumed unless 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 the NRC Operations Center within 24 hours of the occurrence. A written preliminary report shall be sent to the NRC, Attn: Document Control Desk, I White Flint North, 11555 Rockville Pike, Rockville MD 20852, within 14 days of the occurrence. A final written report shall be sent to the above address within 30 days of the occurrence.
- d. Reportable occurrences should be reviewed by the NSC prior to forwarding any written report to 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: Document Control 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 the previous calendar year shall be submitted within six months following the end of each calendar year. Each annual report shall include the following information:
- a. A brief summary of operating experiences including experiments performed, changes in facility design, 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 reactor was critical, and the cumulative total energy output since initial criticality.
Amendment No. 4 36
*
(2) The written report (and, to the extent possible, the preliminary telephone report or report by similar conveyance) shall describe, analyze, and evaluate safety implications, and outline 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 or previously 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 the Safety Analysis Report (SAR); (3) A personnel change involving the positions of UCD/MNRC Director or UCD Vice Chancellor 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 or development 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 required information may be contained in single or multiple records, or a combination thereof. Records and logs shall 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 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 shipped off 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. 4 39
0 72hZ I - UNIVERSITY OF CALIFORNIA - DAVIS CE CHANCELLOR FOR RESEARCH (Licensee) I -
,TOR TIONS NJCHI
....................... Formal Licensing Channel Administrative Reporting Channel
- - - - Communications Channel UCD/MNRC ORGANIZATION FOR LICENSING AND OPERATION FIGURE 6.1 40
Dr. Barry M. Klein Vice Chancellor for Research University of California, Davis One Shields Avenue Davis, 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 Facility Operating License No. 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 to your submittal of October 17, 2002, and is discussed in the enclosed Safety Evaluation Report. Sincerely, Warren J. Eresian, Project Manager Research and Test Reactors Section Operating Reactor Improvements Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No. 50-607
Enclosures:
- 1. Amendment No. 5
- 2. Safety Evaluation Report
S University of California - Davis/McClellan MNRC Docket No. 50-607 cc: Dr. Wade J. Richards 5335 Price Avenue, Bldg. 258 McClellan AFB, CA 95652-2504 Test, Research, and Training Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611
Dr. Barry M. Klein Vice Chancellor for Research University of California, Davis One Shields Avenue Davis, 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 Facility 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 October 17, 2002, and is discussed in the enclosed Safety Evaluation Report. Sincerely, Warren J. Eresian, Project Manager Research and Test Reactors Section Operating Reactor Improvements Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No. 50-607
Enclosures:
- 1. Amendment No. 5
- 2. Safety Evaluation Report DISTRIBUTION:
PUBLIC RORP\R&TR r/f SHolmes OGC MMendonca WEresian TDragoun PMadden AAdams PDoyle CBassett DMatthews EHylton Plsaac DHughes WBeckner GHiIl (2) (T5-C3) LBerg ADAMS ACCESSION NO: ML02 TEMPLATE #: NRR-058 NAME WEresian:rdr EHylton SUttal PMadden WBeckner IiDATE 111/ /2002 11/ /2002 11/ /2002 11/ /2002 11/ /2002J OFFICIAL RECORD COPY
- 0 REGENTS OF THE UNIVERSITY OF CALIFORNIA AT McCLELLAN NUCLEAR RADIATION CENTER DOCKET NO. 50-607 AMENDMENT TO AMENDED FACILITY OPERATING LICENSE Amendment No. 5 License No. R-130
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that A. 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 Nuclear Radiation 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 the regulations of the Commission as stated in Chapter I of Title 10 of the Code of FederalRegulations (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 the Commission; D. The issuance of this amendment will not be inimical to the common defense and security 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; and F. 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. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the enclosure to this license amendment, and paragraph 2.C.(ii) of Amended Facility 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 through Amendment No. 5, are hereby incorporated in the license. The licensee shall 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 Warren J. Eresian, Project Manager Research and Test Reactors Section
*Operating Reactor Improvements Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation
Enclosure:
Appendix A, Technical Specification Changes Date of Issuance:
ENCLOSURE TO LICENSE AMENDMENT NO. 5 AMENDED FACILITY OPERATING LICENSE NO. R-130 DOCKET NO. 50-607 Replace the following pages of Appendix A, Technical Specifications, with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. Remove Insert 17 17 18 18 40 40
Basi.__s-
- a. A limitation positioned of less in the than one pneumatic transfer ($1 .00)(0.7%Ak/k) dollar tube, on the reactivity the central irradiation of a single movable experiment worth Chapter facility (SAR, 10, Section 10.4.1 ), the central 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). In addition, 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 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 secured experiment is its sudden accidental or unplanned removal while the reactor is operating. For such 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 (AC1F)(SAR Chapter 10, Section 1.0.4.2) 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 control .over the positioning of samples into and out of the central core region 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 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 the fuel temperature safety limit.
3.8.2 Materials Limit 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 radioactivity by limiting material 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 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 I-125 in the 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 1-125 being 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 600 millicuries in sealed storage containers) may also be present in the 1-125 fume hood. Amendment No. 5 17
- 0
- e. Explosive materials in quantities greater than 25 milligrams of TNT equivalent shall not be irradiated 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 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 any radiography bay. The irradiation of explosives in any bay is limited to those assemblies where a safety analysis has been performed that shows that there is no damage to the reactor safety systems upon detonation (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 fueled experiment leading to total release of the iodine, occupational doses and doses to members 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 I-125 fume hood assures that, if either of these inventories of 1-125 is totally released into their respective containments, the occupational doses and doses to members of the general 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 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 bay external to the reactor tank will not result in damage to the reactor controls or the 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. 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 or aerosols into the reactor building or into the unrestricted area, the quantity and type of material in the experiment shall be limited such that the airborne radioactivity in the reactor room will not result in exceeding the applicable dose limits in 10 CFR Part 20 in the unrestricted area, assuming 100% of the gases or aerosols escapes. Amendment No. 5 18
S 0
- 0 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 5 TO AMENDED FACILITY OPERATING LICENSE NO. R-130 REGENTS OF THE UNIVERSITY .OF CALIFORNIA AT McCLELLAN NUCLEAR RADIATION CENTER DOCKET NO. 50-607
1.0 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 Facility Operating License No. R-i130 for the McClellan Nuclear Radiation Center (MNRC) TRIGA research reactor. The request provides for the following changes, which if implemented, will result in 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 a Health 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 to whom reports the Health Physics Manager and the Reactor Operations Manager. The proposed management 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. This amendment approved the installation of an Iodine-125 production loop. The production loop included 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, the reactor 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 exhaust system, eventually to make its way to the unrestricted area. The exposure (CEDE to the thyroid) to a person in the unrestricted area for the entire 30 second duration of this event is much less than 1 millirem. Ifthe exposure duration is increased to 10 minutes, the estimated CEDE to the thyroid would still be less than 1 millirem. For those exposed in the reactor room for the maximum assumed occupancy time of 5 minutes the CEDE to the thyroid would be about 67 millirem. The results of all of the assumptions and calculations in the accident sequence are directly proportional to the initial inventory of Iodine-125 in the production system. Increasing the initial assumed inventory from 20 curies to 61 curies will simply result in a tripling of the exposure. The analysis in the SAR that supports the increase in iodine inventory shows that the CEDE to the thyroid for a 10-minute exposure in the unrestricted area would be about 3.0 millirem, For those exposed in the reactor room for the maximum assumed occupancy time of 5 minutes the CEDE to the thyroid would be about 205 millirem. In order to assess the potential consequences of the worst-case assumption, the resulting doses are 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 been analyzed in the licensee's Safety Analysis Report (SAR), and is a complete cladding rupture of a highly-irradiated single fuel element, followed by the instantaneous release of fission products into the air.. The accident analysis calculates the radiological consequences of the MHA with regard to doses to the general public in the unrestricted area, and also calculates occupational doses within the site boundary. The MHA results in a CEDE of 53 millirem in the unrestricted area. Since the release of 61 curies of Iodine-125 through the glovebox exhaust system and eventually to the unrestricted 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 less than 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 the allowable activity of lodine-125 in the Iodine Production Facility from 20 curies to 61 curies does not significantly reduce the margin of safety with respect to the Maximum Hypothetical Accident and 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 within the 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 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 criteria for categorical exclusion set forth in 10 Amendment No. 5
0 0 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 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 or consequences of accidents previously evaluated, or create the possibility of a new or different kind of accident from 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 is reasonable 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 or the health and safety of the public. Principal Contributor: Warren J. Eresian Date: Amendment No. 5
- U
.*NUCLEAR REGUALTORY COMMISSION i __ _ _ _ _ _ _ _ _ _ _ _
.Ii UNIVERSITY OF l CALIFORNIA - DAVIS
- VICE CHANCELLOR FOR I ~RESEARCHi
- (Licensee)
I I I II SDIRECTOR NUCLEAR. H_____SAFETYCO L I l A-tC--SITE MANAGER[ 1
- COMMITITEE I
i i-***-* HEALTH PHYSICS REACTOR BRANCH OPERATIONS FormlChnne Liensig ___________ Aminstrtie RpotinBCANnel CormmunLicatinsin Channel UCD/MNRC ORGANIZATION FOR LICENSING AND OPERATION FIGURE 6.1
RE NIE SAE
** NUCLEAR REGULATORY COMMISSION
~WASHJNGTON, D.C. 20555-0001 N~ovemb~er 2_5, 2003 Dr. Barry M. Klein Vice Chancellor for Research University of California, Davis One Shields Avenue Davis, 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 to Facility 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 Safety Evaluation Report. Sincerely, 6~)4A,~.6 Warren J. Eresian, Project Manager Research and Test Reactors Section New, Research and Test Reactors Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No. 50-607
Enclosures:
- 1. Amendment No. 6
- 2. Safety Evaluation Report
- 0 University of California - Davis/McClellan MNRC Docket No. 50-607 cc:
Dr. Wade J. Richards 5335 Price Avenue, Bldg. 258 McClellan AFB, CA 95652-2504 Test, Research, and Training Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611
OUNITED W STATES
'-*-"NUCLEAR REGULATORY COMMISSION e* *r/*WASHINGTON, D.C. 20555-0001 e- ~ * *~
REGENTS OF THE UNIVERSITY OF CALIFORNIA AT McCLELLAN NUCLEAR RADIATION CENTER DOCKET NO. 50-607 AMENDMENT TO AMENDED FACILITY OPERATING LICENSE Amendment No. 6 License No. R- 130
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that A. 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 Nuclear Radiation Center (the licensee) on March 31, 2003, conforms to the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the regulations of the Commission as stated in Chapter I of Title 10 of the Code of FederalRegulations (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 the Commission; D. The issuance of this amendment will not be inimical to the common defense and security 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; and F. 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. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the enclosure to this license amendment, and paragraph 2.C.(ii) of Amended Facility 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 through Amendment No. 6, are hereby incorporated in the license. The licensee shall 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 Patrick M. Madder Seto Chief Research and Test Reactors Section New, Research and Test Reactors Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation
Enclosure:
Appendix A, Technical Specification Changes Date of Issuance: November 25, 2003
- 0 ENCLOSURE TO LICENSE AMENDMENT NO. 6 AMENDED FACILITY OPERATING LICENSE NO. R-130 DOCKET NO. 50-607 Replace the following pages Of Appendix A, Technical Specifications, with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Remove Insert. 31 31 32 32 33 33 Figure 6.1 Figure 6.1
- 0 5.4 Fissionable Material Storage AppDlicabilitv - This specification applies to the storage of reactor fuel at a time when itis not in the reactor core.
Obiective - The objective is to assure that the fuel which is being stored will not become critical and will not reach 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 keff 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 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 Controls 6.1 Organization. The Vice Chancellor for Research shall be the licensee for the UCD1MNRC. The UCD/M*NRC facility shall be under the direct control of the UCD/MNRC Director. The UCD/MNRC Director shall be accountable to the Vice Chancellor for Research for the safe operation and maintenance of the facility. 6.1.1 Structure. The management for operation of the UCD/MNRC facility shall consist of the organizational structure as shown in Figure 6.1. 6.1.2 Responsibilities. The UCDIMNRC Director shall be accountable to the Vice Chancellor for Research for the safe operation and maintenance of the facility. The UCDIMNRC Director, or his designated alternate, shall review and approve all experiments and experiment procedures prior to their use in the reactor. Individuals in 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 the operating license and technical 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 final decision. The Reactor Supervisor and Health Physics Supervisor report directly to the Operations 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 who can perform prescribed instructions;
- c. A senior reactor operator readily available. The available senior reactor operator should be within thirty (30) minutes of the facility and reachable by telephone, and;
- d. A senior reactor operator shall be present whenever a reactor startup is performed, 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 be available to the reactor operator in the control room. The list shall include: Amendment No. 6 31
- 0
- 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 and Training of Personnel for Research Reactors (ANS 15.4). Qualification and requalification of licensed operators shall be subject to an approved Nuclear Regulatory Commission (NRC) program. 6.2 Review. Audit. Recommendation and Approval General Policy. Nuclear facilities shall be designed, constructed, operated, and maintained in such amanner 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 with applicable 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 NSC Composition and Qualifications. The UCD Vice Chancellor for Research shall appoint the 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. The NSC Shall evaluate 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 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 incl-*u~de "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 proposed Class I modifications, prior to implementing (Class I) modifications, prior to taking action under the preceding 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 an unreviewed safety question or require a change to the license, to a Facility Use Authorization, or to the Technical Specifications. This determination may be in the form of verifying a decision already made by the UCD/MNRC Director.
Amendment No. 6 32
- d. Review health reactor physics operations program and operational and associated recordsmaintenance, Class Inuclear for all UCDIMNRC modification records, and the facilities.
- e. Review the periodic updates of the Emergency Plan and Physical Security Plan for UCD/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 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 other inspections of these facilities conducted by other agencies.
6.2.4 Audit/Inspection Function. The NSC or a subcommittee thereof, shall audit/inspect reactor operations 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 UCDIMNRC 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 and Technical 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 the UCD/MNRC Radiation Safety Program. The program should use the guidelines of the American National Standard for Radiation Protection at Research Reactor Facilities (ANSI/ANS 15.11). The Health 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 nthssction. 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 to assure that procedures are current. Procedures shall be adequate to assure the safe operation of the Amendment No. 6 33
- 0 I-..... COMMISSION UNVRIYO CAIOM AI SAFETYECOITYTEE AIFRI 1 C-MDATEES I VIEMANAGELLRFO I, I SUPERISRECREANCTO AR.. SFT OPERSUPERVISOR MA ANG ER_______________________________________________ i FormlChnne Liensig UCD/NRC ORGAIZAIOSOR REACTOR OETINLICNIGADSFT FIGURE 6.1
R R E*O4"*O OUNITED STATES
-* ,NUCLEAR REGULATORY COMMISSION
*.* WASHINGTON, D.C. 20555-0001 "
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 6 TO AMENDED FACILITY OPERATING LICENSE NO. R-130 REGENTS OF THE UNIVERSITY OF CALIFORNIA AT McCLELLAN NUCLEAR RADIATION CENTER DOCKET NO. 50-607
1.0 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 Facility Operating License No. R-1 30 for the McClellan Nuclear Radiation Center (MNRC) TRIGA research reactor. The request provides for the following changes, which if implemented, will result in 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, and change the Technical Specifications and UCD/MNRC Organization Chart to reflect this change.
Each of these requests is discussed below. 2.0 EVALUATION The current organization structure includes an UCD/MNRC Director to whom reports a Site Manager. 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/MNRC Director, 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 the Chairperson of the NSC. In the proposed organization structure, that responsibility is given to the UCD Vice Chancellor for Research, who is also the licensee for the UCD/MNRC. Since this proposed change increases the level of oversight from the licensee's staff to the licensee, the staff concludes that it is acceptable.
The staff has reviewed the proposed changes to the TSs and concluded that they are administrative in nature and do not impact the licensee's ability to continue to meet the relevant requirements of 10 CFR 50.36.
3.0 ENVIRONMENTAL CONSIDERATION
This amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51 .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 the amendment 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 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 is reasonable 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 or the health and safety of the public. Principal Contributor: Warren J. Eresian Date: November 25, 2003 Amendment No. 6
0 0 TECHNICAL SPECIFICATIONS FOR THE UNIVERSITY OF CALIFORNIA - DAVIS MCCLELLAN NUCLEAR RADIATION CENTER (UCD/MNRC) DOCUMENT NUMBER: MNRC-0004-DOC-1 2 Rev 12 09/02
1 Oct. 16 0? 11:OOa Ar t~~r 6. Johnson tS4lJ 753-9743 p. 1,JLaJ. rwdIL. .atDfJh.L.* ~ r. t.wc. TECHNICAL SPECtFICATIONS APPROVAL Revision 12 Radiation Cencer "Teclnical Gpo of me(UOI)/MNRG) Reactor havefor ctifoons* the Universit undergone of California the following at DavistlMcCleIlan, Nuclear coordination: 10 ~ 02-Reviewed by;*ell/ Dale Rcvicwcd by'." ' floa rMnae " " D~kcI R~eviewed by: Site Manager Date I Approved by:
/~zL7z~OZ-UCD/MNRc~bir4ctor Data Approvod by; Date
- 0 Technical Specifications Rev 12 09/2002 TtePageRe12 902 Titovle Page Rev 12 9/2002 32 Rev 12 9/2002 Figure 6.1 Rev 12 9/2002
S 0 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Paage 1.0 Definitions ................................... ............................................................................ 2.0 Safety Limit and Limiting Safety System Setting (LSSS.)............................................................ 5 2.1 Safety Limits................................................................................................... 5 2.2 Limiting Safety System Setting (LSSS)...................................................................... 6 2.2.1 Fuei Temperature.................................................................................. 6 3.0 Limiting Conditions for Operations (LC.O.0............................................................................. 7 3.1 Reactor Core Parameters.................................................................................... 7 3.1 .1 Steady-State Operation ........................................................................... 7 3.1.2 Pulse or Square Wave Operation ................................................................ 7 3.1.3 Reactivity Limitations .............................................................................. 8 3.2 Reactor Control and Safety Systems........................................................................ 8 3.2.1 Control Rods....................................................................................... 8 3.2.2 Reactor Instrumentation................................... ........................................ 9 3.2.3 Reactor Scrams and Interlocks.................................................................. 10 3.2.4 Reactor Fuel Elemenis........................................................................... 12 3.3 Reactor Coolant Systems................................................................................... 13 3.4 Reactor Room Exhaust System ............................................................................ 14 3.5 Intentionally Left Blank ...................................................................................... 14 3.6 Intentionally Left Blank ...................................................................................... 14 3.7 Reactor Radiation Monitoring Systems .................................................................... 14 3.7.1 Monitoring Systems .............................................................................. 14 3.7.2 Effluents - Argon-41 Discharge Limit..............,.............................................. 16
9 0 Page* 3.8 Experiments ................................................................................................. 16 3.8.1 Reactivity Limits................................................................................... 16 3.8.2 Materials Limit.................................................................................... 17 3.8.3 Failure and Malfunctions......................................................................... 18 4.0 Surveillance Requirements .......................................................................................... 19 4.1 Reactor Core Parameters................................................................................... 19 4.1.1 Steady State Operation.......................................................................... 19 4.1.2 Shutdown Margin and Excess Reactivity ....................................................... 20 4.2 Reactor Control and Safety Systems ...................................................................... 20 4.2.1 Control Rods ..................................................................................... 20 4.2.2 Reactor Instrumentation ......................................................................... 21 4.2.3 Reactor Scrams and Interlocks.................................................................. 22 4.2.4 Reactor Fuel Elemen*ts........................................................................... 23 4.3 Reactor Coolant Systems*................................................................................. .24 4.4 Reactor Room Exhaust System ............................................................................ 25 4.5 Intentionally Left Blank...................................................................................... 25 4.6 Intentionally Left Blank...................................................................................... 25 4.7 Reactor Radiation Monitoring Systems .................................................................... 25 4.8 Experiments ................................................................................................. 26 5.0 Design Features ...................................................................................................... 27 5.1 Site and Facility Description ................................................................................ 27 5.1.1 .Site................................................................................................ 27 5.1.2 Facility Exhaust .................................................................................. 28 5.2 Reactor Coolant System..................................................................................... 28
0 S 5.3 Reactor Core and F~ue]........................................................................................ 29 5.3.1 Reactor Core .......................................................................... :............. 29 5.3.2 Reactor .FuelJ........................................................................................ 30 5.3.3 Control Rods and Control Rod Drives ............................................................ 31 5.4 Fissionable Material Storage.................................................................................. 31 6.0 Administrative Controls.................................................................................................. 31 6.1 Organization.................................................................................................... 31 6.1.1 Structure............................................................................................. 32 6.1.2 Responsibilities..................................................................................... 32 6.1.3 Staffing .............................................................................................. 32 6.1.4 Selection and Training of Personnel.............................................................. 32 6.2 Review, Audit, Recommendation and Approval............................................................. 32 6.2.1 NSC Composition and Qualifications ................................................ i............ 33 6.2.2 NSC Charter and Rules ........................................................................... 33 6.2.3 Review Function.................................................................................... 33 6.2.4 Audit/Inspection Function.......................................................................... 34 6.3 Radiation Safety................................................................................................ 34 6.4 Procedures ..................................................................................................... 34 6.4.1 Reactor Operations Procedures................................................................... 34 6.4.2 Health Physics Procedures........................................................................ 35 6.5 Experiment Review and Appro~ial ............................................................................ 35 6.6 Required Actions............................................................................................... 35 6.6.1 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 Reports .......................................................................................................... 36 6.7.1 Operating Reports.................................................................................. 36 6.7.2 Special Reports..................................................................................... 38 6.8 Records ......................................................................................................... 39 I Fig. 6.1 UCD/MNRC Organization for Licensing and Operation......................................................... 40
0 0 TECHNICAL SPECIFICATIONS FOR THE UNIVERSITY OF CALIFORNIA - DAVIS/MCCLELLAN NUCLEAR RADIATION CENTER (UCD/MNRC) General The University of California - Davis/McClellan Nuclear Radiation Center (UCD/MNRC) reactor is operated by the University 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 provides a wide range of irradiation services for both research and industrial needs. The reactor operates at a nominal steady state power level up to and including 2 MW. The UCD/MNRC reactor is also capable of square wave and 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 Nuclear Regulatory Commission (e.g., senior reactor operator or reactor operator) to carry out the duties and responsibilities associated with the position requiring the license. 1.2.1 Senior Reactor Operator. An individual who is licensed to direct the activities of reactor operators 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 and perform reactor-related maintenance. 1.3 Channel. A channel is the combination of sensor, line amplifier, processor, and output 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 the channel for verification that it is operable. 1.3.2 Channel Calibration. A channel calibration is an adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm or trip, and shall be deemed to include a channel test. 1.3.3 Channel Check. A channel check is a qualitative verification of acceptable performance by observation of channel behavior. This verification, where possible, shall include comparison of the channel with other independent channels or systems measuring the same variable. 1.4 Confinement. Confinement means isolation of the reactor room air volume such that the movement of 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, fission chambers, foils, etc), which is designed to investigate specific reactor characteristics or which is intended for irradiation within an experiment facility and which is not rigidly secured to a core or shield structure 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 entire experiment may be moved in or near the reactor core or into and out of reactor experiment facilities while the reactor is operating. I
0 0 1.5.2 Experiment. Secured. A secured experiment is any experiment, experiment facility, or component 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 experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the experiment, or by forces which can arise as a result of credible 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 their associated input circuits, which are designed to initiate a scram for the primary purpose of protecting an experiment or to provide information which requires manual protective action to be initiated. 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 can be either 8.5, 20, or 30 wt%, with an enrichment of less than 20% U-235. A standard fuel element may contain a burnable poison. 1.7 _Fuel Element. Instrumented. An instrumented fuel element is a standard fuel element fabricated with thermocouples for temperature measurements. An instrumented fuel element shall have at least one operable 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 a channel. 1.9 Mode, Steady-State. Steady-state mode operation shall mean operation of the UCD/MNRC reactor with 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/MNRC reactor 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 the selector 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 within the UCD/MNRC reactor safety system in response to a variable or condition of the UCD/MNRC reactor facility having 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 the generation 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. 2
- 0 1.15.3 Instrument System Level. At the protective instrument level, protective action is the generation 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 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 the maximum 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 reactor is at ambient temperature and the reactor is critical. (Keff = 1) 1.18 Reactivity Limits. The reactivity limits are those limits imposed on the reactivity conditions of the reactor core. 1.19 Reactivity Worth of an Experiment. The reactivity worth of an experiment is the maximum value of the reactivity 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 which directly 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 the parameters 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 in these 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 for initiation of manual protective action. 1.24 Reactor Secured. The UCD/MNRC reactor is secured when the console key switch is in the off position and the key is removed from the lock and under the control of a licensed operator, and the conditions of a or b exist:
- a. (1) The minimum number of control rods are fully inserted to ensure the reactor is shutdown, as required by 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 being moved or serviced ifthe experiments have, on movement, a reactivity worth exceeding the maximum value allowed for a single experiment or $1.00, whichever is smaller, or
- b. The reactor contains insufficient fissile materials in the reactor core, adjacent experiments or control rods to attain criticality under optimum available conditions of moderation and reflection.
1.25 Reactor Shutdown. The UCDIMNRC reactor is shutdown ifit 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
0 1.26 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, and which 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 a fuel 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 allow it to perform its control function, and its safety function when the coupling is disengaged. 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 power level and 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 of radioactivity. 1.31 Scram Time. Scram time is the elapsed time between reaching a limiting safety system set point and the 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 denote a recommendation; the word "may" to denote permission, neither a requirement nor a recommendation. 1.34 Shutdown Margin. Shutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence that the reactor can be made subcritical by means of the control and safety system starting from any permissible operating condition with the most reactive rod assumed to be in the most reactive position, and once this action has been initiated, the reactor will remain subcritical without further operator 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 manual shutdown 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 checks and tests, and calibrations. Operability checks and tests are generally specified as djaily, weekly or 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 surveillance interval is the interval between a check, test, or calibration, whichever is appropriate to the item being 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 involve an unreviewed safety question:
- a. If the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or
- b. If a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or
- c. 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. 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 reactor computerized 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 fuel element. 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 during steady-state operation.
- b. The maximum temperature in a standard TRIGA fuel element shall not exceed 1100 0C during pulse operation.
Basis -
- a. This fuel safety limit applies for conditions in which the cladding temperature is above 500 °C (Safety Analysis Report (SAR), Chapter 4, Section 4.5.4.1.3). The important parameter for a TRIGA reactor is 5
0 0 the fuel element temperature. This parameter is well suited as it can be measured directly. A loss in the integrity of the fuel element cladding could arise if the cladding stress exceeds the ultimate strength of the cladding material. The fuel element cladding stress is a function of the element's internal pressure while the ultimate strength of the cladding material is a function of its temperature. The cladding stress is a result of the internal pressure due to the presence of air, fission product gasses and hydrogen from the disassociation 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 hydrogen to zirconium in the alloy. At a fuel temperature of 930 °C for ZrH1 7 fuel, the cladding stress due to the internal pressure is equal to the ultimate strength of the cladding material at the same temperature (SAR Fig 4.18). This is a conservative limit since the temperature of the cladding material is always 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 a pulse which gives a peak adiabatic fuel temperature of 1000 °C is estimated to be 470 °C. Further analysis (SAR Section 4.5.4.1.2), shows that the internal pressure for both Zr 1.65 (at 11 50°C) and Zr17z (at 11 00°C) increases to a peak value at about 0.3 sec, at which time the pressure is about one-fifth of the equilibrium value or about 400 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 ZrH 1.65 fuel temperature greater than 1150 °C, over a 200°C range, show that the time to reach the peak pressure and the fraction of equilibrium pressure value achieved were 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 and compared 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 6. 5) was only about 6% of the equilibrium value evaluated at the peak temperature. Calculations of the pressure gave values about three times greater than the measured values. The analysis gives strong indications that the cladding will not rupture iffuel 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 for cladding failure. 2.2 Limiting Safety System Setting. 2.2.1 Fuel Temperature. Applicability - This specification applies to the protective action for the reactor fuel element temperature. Objective - The objective is to prevent the fuel element temperature safety limit from being reached. Specification - The limiting safety system setting shall be 750 °C (operationally this may be set more conservatively) as measured in an instrumented fuel element. One instrumented element shall 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 a temperature which, ifexceeded, shall cause a reactor scram to be initiated preventing the safety limit from being exceeded. A setting of 750 °C provides a safety margin at the point of the measurement 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 measured temperatures 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 the center and mid-plane of the fuel element. For pulse operation of the reactor, the same limiting safety system setting applies. However, the temperature channel will have no effect on limiting 6
0 the peak power generated because of its relatively long time constant (seconds) as compared with the width of the pulse (milliseconds). In this mode, however, the temperature trip will act to limit the energy release after the pulse ifthe transient rod sho~uld 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 the purpose 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/MNRC TRIGA 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 result 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 to 1.7. 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 state operation at two (2) MW the hydrogen will redistribute due to migration from the central high temperature regions of 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 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. Ifthe pulse insertion is such that the temperature of the fuel exceeds about 875 °C, then the pressure may be sufficient to cause expansion of microscopic holes in the fuel that grow with each pulse. Analysis (SAR Chapter 13, Section 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% 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 the reactivity 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 at all 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 by the 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) shall not exceed 6.65% Ak/k ($9.50).
Basis -
- a. This specification assures that the reactor can be placed in a shutdown condition from any operating condition and remain shutdown, even if the maximum worth control rod should stick in the fully withdrawn position (SAR Chapter 4, Section 4.5.5).
- b. This specification sets an overall reactivity limit which provides adequate excess reactivity to override the xenon buildup, to overcome the temperature change in going from zero power to 2 MW, to permit pulsing at the $1.75 level, to permit irradiation of negative worth experiments and account for fuel burnup over time. An adequate shutdown margin exists with an excess of $9.50 for the two analyzed cores: (SAR Chapter 4, Section 4.5.5).
3.2 Reactor Control and Safety Systems 3.2.1 Control Rods Applicability - 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 drive assemblies.
- b. The scram time measured from the instant a signal reaches the value of a limiting safety system setting to the instant that the slowest control rod reaches its fully inserted position shall not exceed one (1) second.
8
Basis -
- a. Thecontinue apparent to condition of the control rod assemblies shall provide assurance that the rods shall perform reliably as designed.
- b. This assures that the reactor shall shut down promptly when a scram signal is initiated (SAR Chapter 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 reactor operator during reactor operations. Objective - The objective is to require that sufficient information is available to the operator to assure safe operation of the reactor. Specification - The reactor shall not be operated unless the channels described in Table 3.2.2 are operable and the information is displayed on the reactor console. Table 3.2.2 Required 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,A Level Safety MW or less Channel
- b. Linear Power 10 1Automatic D,M,A Channel Power Control
- c. Log Power 10 1Startup D,M,A Channel Control
- d. Fuel Temperature 2 2 2Fuel D,M,A Channel Temperature
- e. Pulse Channel 0 10Measures P,A Pulse NV & NVT
(*)Where: 0 - Channel check during each day's operation M- Channel test monthly A - Channel calibration annually P - 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 is 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 assure that the reactor power level and energy are adequately monitored (SAR Chapter 7, Sections 7.1.2 & 7.1.2.2).
9
the fuel temperature that Chapter assure(SAR is properly d. monitored and The Table 3.2.2. indicated in the reactorchannels fuel temperature control room 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 in Table 3.2.3 are operable: Table 3.2.3 Required Scrams and Interlocks Steady Square Channel Surveillance State Wave Function Requirements* Scram. Pulse
- a. Console 1 1 Manual Scram M Manual and Automatic Scram 1 Scram Alarm
- b. Reactor Room Manual Scram M Manual Scram and Automatic Scram Alarm
- c. Radiography 4 Manual Scrams M 4
Bay Manual and Automatic Scrams Scram Alarms
- d. Reactor Power 0 Automatic M 2
Level Safety Scram Alarms & Scrams Scrams 2 at 2.2 MW or less
- e. High Voltage Automatic M 2 Scram Alarms &
Power Supplies Scrams Scrams on Loss of High Voltage to 2 the Reactor Power Level Safety Channels
- f. Fuel Automatic Scram M 2 Alarms & Scrams Temperature Scrams on indicated fuel 2 2 temperature of 750°C or less
- g. Watchdog 2 Automatic Scram M Circuit Alarms & Scrams 10
- h. External 2 2 Automatic M Scrams Scrams and Alarms if an experiment or radiography scram interlock is activated
- i. One Kilowatt Prevents initiation M 0 I Pulse & of a step reactivity Square Wave insertion above a Interlock reactor power level of I KW
- j. Low Source I I Prevents withdrawal M Level Rod of any control rod Withdrawal ifthe log channel Prohibit reads less than 1.5 Interlock times the indicated log channel current level with the neutron source removed from the core
- k. Control Rod I I Prevents simul- M Withdrawal taneous withdrawal Interlock of two or more rods in manual mode I. Magnet I I De-energizes the M Power Key control rod Switch Scram magnets, scram &
alarm (*)Where: M - channel test monthly Basis -
- a. Table 3.2.3. The console manual scram allows rapid shutdown of the reactor from the control room (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 the reactor room.
- c. Table 3.2.3. The radiography bay manual scrams allow rapid shutdown of the reactor from any 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 if the 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 safety channels operate within their intended range as required for proper functioning of the power level 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 fuel temperature exceeds 7500° C, therefore ensuring the safety limit will not be exceeded (SAR Chapter 4, Sections 4.5.4.1 & 4.7.2).
11
- g. Table 3.2.3.
acquisition The watchdog computer circuits properly are functioning assure that (SARtheChapter control 7,system computer Section 7.2). and the data
- 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 into the bays (SAR Chapter 9, Section 9.6). External scrams from experiments, a subset of the external scrams, also assure the integrity of the reactor system, the experiment, the facility, and the 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 above one (1) kilowatt assures that the pulse magnitude will not allow the fuel element temperature to exceed 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 source of 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 two or more control rods thus limiting the reactivity-insertion rate from the control rods in manual mode (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 rod magnets and results in a scram (SAR Chapter 7, Section 7.1.2.5). 3.2.4 Reactor Fuel Elements Applicability - This specification applies to the physical dimensions of the fuel elements as measured 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 fuel elements 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 core if:
- a. In measuring the transverse bend, the bend exceeds 0.125 inch (3.175 mm) over the full length 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 the reactor, during which differential expansion between the fuel and the cladding occurs and the pressure of the gases within the elements increases sharply. The above limits on the allowable distortion of a fuel element correspond to strains that are considerably lower than the strain expected to cause rupturing of a fuel element. Limited operation in the steady state or pulsed mode may be necessary to identify a leaking fuel element especially if the leak is small. 12
3.3 Reactor Coolant Systems Applicability - 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 channels described in Table 3.3 are operable, and the information is displayed locally or in the control room. Table 3.3 REQUIRED WATER SYSTEMS AND INSTRUMENTATION Minimum Measuring Number Surveillance Channel/System Operable Function: Channel/System Requirements*
- a. Primary Coolant 1 For operation of the D,Q,A Core Inlet reactor at 1.5 MW or Temperature higher, alarms on high Monitor heat exchanger outlet temperature of 45 °C (113°F)
- b. Reactor Tank I Alarms ifwater level M Low Water drops below a depth of Monitor 23 feet in the reactor tank
- c. Purification** I Alarms ifthe primary D,M,S Inlet Conduc- coolant water conductivity tivity Monitor is greater than 5 micromhos/cm
- d. Emergency Core I For operation of the reactor D,S Cooling System at 1.5MW or higher, provides water to cool fuel in the event of a Loss of Coolant Accident for a minimum of 3.7 hours at 20 gpm from an appropriate nozzle
(*)Where: D -- channel A channel check during calibration each day's operation annually Q - 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 shall be shutdown. Basis -
- a. Table 3.3. The primary coolant core inlet temperature alarm assures that large power fluctuations will not 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 there is 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 a week 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 System Applicability - 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 the reactor 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 the pressure 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 Coolant 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 room will be reduced due to mixing with exhaust system air (SAR Chapter 9, Section 9.5.1). Pressure in the reactor room will be negative relative to surrounding areas due to air flow patterns created by the reactor room 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 fuel temperature 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 Systems 3.7.1 Monitoring Systems Applicability - This specification applies to the information which shall be available to the reactor operator 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. Specification - The reactor shall not be operated unless the channels described in Table 3.7.1 are operable, the readings are below the alarm setpoints, and the information is displayed in the 14
- 0 control room. The stack and reactor room CAMS shall not be shutdown at the same time during reactor operation.
Table 3.7.1 REQUIRED RADIATION MONITORING INSTRUMENTATION Minimum Measuring Number Channel Surveillance Equipment Operable** Function Requirements*
- a. Facility I Monitors Argon-41 and D,W,A Stack Monitor radioactive particu-lates, and alarms
- b. Reactor Room I Monitors the radiation D,W,A Radiation level in the reactor Monitor room and alarms
- c. Purification I Monitors radiation D,W,A System Radia- level at the demineral-tion Monitor izer station and alarms
- d. Reactor Room I Monitors air from the D,W,A Continuous reactor room for parti-Air Monitor culate and gaseous radioactivity and alarms
(*)Where: D - channel check during each day's operation A - channel calibration annually W - 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 release of radioactive material to the environment (SAR Chapter 11, Section 11.1.1.1.4). The alarm setpoint 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 the operations area.
- b. Table 3.7.1. The reactor room radiation monitor provides information regarding radiation levels in the reactor 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 provides information regarding radioactivity in the primary system cooling water (SAR Chapter 11, Section 11.1.5.4.2) 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 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.
15
3.7.2 Effluents - Arqon-41 Dischargle Limit Applicability - 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 releases of this radionuclide from the UCD/MNRC, and the corresponding annual radiation dose from Argon-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 is specified in Appendix B, Table 2, Column 1 of 10 CFR Part 20. 10 CFR 20.1301 specifies dose limitations in the unrestricted area. 10 CFR 20.1101 specifies a constraint on air emissions of radioactive material to the environment. The SAR Chapter 11, Section 11.1.1.1.4 estimates that the routine Argon-4l releases and the corresponding doses in the unrestricted area will be below these limits. 3.8 Exp~eriments 3.8.1 Reactivity Limits. Applicability - This specification applies to the reactivity limits on experiments installed in specific reactor experiment facilities. Objective - The objective is to assure control of the reactor during the irradiation or handling of experiments 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 (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 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 or 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, and in any other reactor in-core and in-tank irradiation facilities at any given time shall be less than one dollar and ninety-two cents ($1.92) (1.34% A k/k), including the potential reactivity which might 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 single movable 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 limit of $1.75 is not exceeded (SAR Chapter 13, Section 13.2.2.2.1). In addition, limiting the worth of each movable experiment to less than $1.00 will assure that the additional increase in transient 16
- 0 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 secured experiment is its sudden accidental or unplanned removal while the reactor is operating. For such 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 Chapter 10, 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 control over the positioning of samples into and out of the central core region 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 less thanthe 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 the fuel temperature safety limit.
3.8.2 Materials Limit 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 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 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 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 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 total inventory of 1-125 being handled in the 1-125 fume hood at any one time in preparation for shipping shall not exceed 20 curies. An Additonal 1.0 curie of 1-125 (up to 400 millicuries in the form of quality assurance samples and up to 600 millicuries in sealed storage containers) may also be present in the 1-125 fume hood.
- e. Explosive materials in quantities greater than 25 milligrams of TNT equivalent shall not be irradiated 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 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 any radiography bay. The irradiation of explosives in any bay is limited to those 1"7
assemblies where a safety analysis has been performed that shows that there is no damage to the 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 of materials.
- b. The 1.5 curies limitation on iodine 131 through 135 assures that in the event of failure of a fueled experiment leading to total release of the iodine, occupational doses and doses to members 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 is totally released into its respective containment, or if both inventories are simultaneously released into their respective containments, the occupational doses and doses to members of the general public in the unrestricted areas will 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 of 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 any radiography bay external to the reactor tank will not result in damage to the reactor controls or the 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 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 and type of material in the experiment shall be limited such that the airborne radioactivity in the reactor room will not result in exceeding the applicable dose limits in 10 CFR Part 20 in 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 on high radiation levels, at least 10% of the gaseous activity or aerosols produced will escape. 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 by boiling this material can escape only through an undistributed column of water above the core, at least 10% of these vapors can escape.
- c. If a capsule fails and releases material which could damage the reactor fuel or structure by corrosion 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 his designated 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 reactor room or the unrestricted area will result in exceeding the applicable dose limits in 10 CFR Part 20.
- b. These assumptions are used to evaluate the potential airborne radioactivity release due to an experiment 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 be brought to the attention of the UCD/MNRC Director or his designated alternate for review to assure 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 reactor operating 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 time period. Prior to reactor system or measuring channel operation, the surveillance shall be performed for each reactor 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 level monitoring channels. Obiective - The objective is to verify that the maximum power level of the reactor does not exceed 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 a significant change in core configuration occurs, a channel calibration shall be required. Discovery of noncompliance with this specification shall limit reactor operations to that required to perform the surveillance. Basis - The annual power level channel calibration will assure that the indicated reactor power level is correct. 4.1.2 Shutdown Margin and Excess Reactivity 19
- 0 Applicability - These specifications apply to the surveillance requirements for reactivity control of the reactor core.
Objective - The objective is to measure and verify the reactivity worth, performance, and operability 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 shutdown margin 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 required to perform the surveillance. Basis -
- a. The reactivity worth of the control rods is measured to assure that the required shutdown margin is available and to provide an accurate means for determining the excess reactivity of the core. Past experience with similar reactors gives assurance that measurements of the 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 been made.
- 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 change unexpectedly.
4.2 Reactor Control and Safety Systems 4.2.1 Control Rods Applicability - This specification applies to the surveillance of the control rods. Objective - The objective is to inspect the physical condition of the reactor control rods and establish the operable condition of the rods. Specification - Control rod worths shall be determined annually or after physical removal or any significant change in core or control rod configuration.
- a. Each control rod shall be inspected at annual intervals by visual observation of the fueled sections 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 required to perform the surveillance. Basis (Technical Specifications 4.2.1 .a-b) - Annual determination of control rod worths or measurements after any physical removal or significant change in core loading or control rod 20
configuration provides information worths. The frequency about of inspection changes for the controlinrods reactor shalltotal reactivity provide and verification periodic individual rod of the condition of the control rod assemblies. The specification intervals for scram time assure operable 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 fuel temperature 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 required to perform the surveillance. Basis - 21
- a. A daily channel check and monthly test, plus the annual calibration, will assure that the reactor power level safety channels operate properly.
- b. A channel test monthly of the reactor power level multi-range channel will assure that the channel is operable and responds correctly. The channel check will assure that the reactor power 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 is operable and responds correctly. A channel check of the reactor power level wide range log channel will assure that the channel is operable on a daily basis. A channel calibration will assure 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, will assure that the fuel temperature measuring channels operate properly.
- e. A channel test prior to pulsing plus the annual channel calibration will assure the pulse energy integrating 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 perform the surveillance. Basis - 22
0
- a. 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 is operable.
- c. A channel test monthly of the Radiography Bay Manual Scrams will assure that the scrams are operable.
- d. A channel test monthly of the Reactor Power Level Safety Scrams will assure that the scrams are operable.
- e. A channel test monthly of the Loss-of-High-Voltage Scram will assure that the high voltage power supplies are operable and respond correctly.
- f. A channel test monthly of the Fuel Temperature Scrams will assure that the scrams are operable.
- g. A channel test monthly of the Watchdog Circuits Scrams will assure that the scram circuits are operable.
- h. A channel test monthly of the External Scrams will assure that the scrams are operable and respond 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 assure that the interlock is operable.
- k. A channel test monthly of the Control Rod Withdrawal Interlock will assure that the interlock is operable.
I. A channel test monthly of the Magnet Current Key Switch will assure that the scram is operable. 4.2.4 Reactor Fuel Elements Applicability - 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 of instrumented fuel elements), and all elements adjacent to the central irradiation facility shall be inspected annually.
- b. Instrumented fuel elements shall be inspected ifany of the elements adjacent to it fail to pass the visual and/or physical measurement requirements of Section 3.2.4. Discovery of noncompliance with Technical Specification 4.2.4 shall limit operations to that required to perform the surveillance.
Basis (Technical Specifications 4.2.4.a-b) - The above specifications assure that the lead fuel elements shall be inspected regularly and the integrity of the lead fuel elements shall be maintained. These are the fuel elements with the highest power density as analyzed in the SAR Chapter 4, Section 4.5.5.6. The instrumented fuel element is excluded to reduce the risk of damage to the thermocouples. 23
.0*
4.3 Reactor Coolant Systems Applicability - 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, the tank water level alarm, the water conductivity cells and the emergency core cooling system are all operable. 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 to perform the 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 input signal. 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 input 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 for power levels above 1.5 MW. A channel calibration semiannually assures that the Emergency Core Cooling System performs as required for power levels above 1.5 MW.
24
0 4.4 Reactor Room Exhaust System Applicability - This specification applies to the surveillance requirements for the reactor room exhaust 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's operation. 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 verify that the exhaust system is maintaining a negative pressure in the reactor room relative to the surrounding facility areas. 4.5 This section intentionally left blank 4.6 This section intentionally left blank. 4.7 Reactor Radiation Monitoring Systems Applicability - 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 to perform the surveillance. Basis -
- a. A channel check of the facility stack monitor system during each day's operation will assure the monitor is operable. A channel test weekly will assure that the system responds correctly to a 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 assure that the monitor is operable. A channel test weekly will ensure that the system responds to a known source. 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 assures that the monitor is operable. A channel test weekly will ensure that the system responds to a known source. 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 responds correctly to a known source. A channel calibration annually will assure that the CAM reads 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 may damage the 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 a written safety analysis has been performed and reviewed by the UCD/MNRC Director, or his designee, 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 Facility Use Authorization. Facility Use Authorizations and experiments carried out under these authorizations 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/MNRC experiment facility until it has been reviewed for compliance with the approved experiment and Facility Use Authorization 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 other UCD/MNRC reactor in-core or in-tank irradiation facility shall be estimated or measured, as 26
0I! 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 in any UCD/MNRC reactor experiment facility.
Basis - a & b. Experience at most TRIGA reactor facilities verifies the importance of reactor staff and safety committee reviews of proposed experiments.
- c. Measurement of the reactivity worth of an experiment, or estimation of the reactivity worth based 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 will facilitate maintaining surveillance over such experiments.
5.0 Design Features 5.1 Site and Facility Description. 5.1.1 Sit__e Applicability - This specification applies to the UCD/MNRC site location and specific facility design 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 downtown Sacramento, 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. The unrestricted 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 has 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 of the 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 reactor building. This accident was analyzed in Chapter 13 of the Safety Analysis Report, and found to be beyond a credible accident scenario.
27
0 0 5.1.2 Facility Exhaust Applicability - This specification applies to the facility which houses the reactor. Objective - 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 proper removal of heat from the reactor room. Specification -
- a. The UCD/MNRC reactor facility shall be equipped with a system designed to filter and exhaust air from the UCD/MNRC facility. The system shall have an exhaust stack height of a minimum of 18.2m (60 feet) above ground level.
- b. Manually activated shutdown controls for the exhaust system shall be located in the reactor control room.
Basis - The UCD/MNRC facility exhaust system is designed such that the reactor room shall be maintained at a negative pressure with respect to the surrounding areas. The free air volume within the 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 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 to operating personnel. 5.2 Reactor Coolant System Applicability - This specification applies to the reactor coolant system. Obiective - The objective is to assure that adequate water is available for cooling and shielding during normal 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 ifthe water level in the reactor tank drops below a depth of 23 ft.
- c. For operations at 1.5 MW or higher during a Loss of Coolant Accident the reactor core shall be cooled for 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 be exceeded 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 and 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 28
- 0 5.3.1 Reactor Core Applicability - This specification applies to the configuration of the fuel.
Objective - The objective is to assure that provisions are made to restrict the arrangement of fuel elements 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 of 96 or more fuel elements to include fuel followed control rods. Below 0.5 MW there is no minimum 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 be loaded into corner positions ony and graphite dummy elements in the flat positions. The performance of fuel temperature measurements shall apply to variations to the as-analyzed 20E core 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 done before 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 above loading restrictions will be allowed in an operational 0.5 MW or greater core. Specifications for the 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 Argonne National Laboratory Report AN L/ED 97-54). 5.3.2 Reactor Fuel Applicability - 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 their physical and nuclear characteristics. 29
0 0 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 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 limiting operation to 2.3 megawatts steady state or to a 36 megawatt-sec pulse assures an ample margin of safety between the maximum temperature generated in the fuel and the safety limit for fuel 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, is equal to the clad yield strength when both fuel and cladding temperature are at the safety limit 9300 C. Since the fuel temperatures are not expected to exceed 630 0C during any condition of normal operation, 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 of purchased materials.
5.3.3 Control Rods and Control Rod Drives Applicability - This Specification applies to the control rods and control rod drives used in the reactor core. Objective - The objective is to assure the control rods and control rod drives are of such a design 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 and regulating rods shall have fuel followers sealed in stainless steel. The transient rod shall have an air 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 assure mechanical stability during movement and to isolate the neutron poison from the tank water environment. 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 at the 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.
30
- 0 5.4 Fissionable Material Storage Applicability - This specification applies to the storage of reactor fuel at a time when it is not in the reactor core.
Objective - The objective is to assure that the fuel which is being stored will not become critical and will not 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 the keff 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 Controls 6.1 Organization. The Vice Chancellor for Research shall be the licensee for the UCD/MNRC. The UCD/MNRC facility shall be under the direct control of the UCD/MNRC Director or a licensed senior reactor operator (SRO) designated by the UCD/MNRC Director to be in direct control. The UCD/MNRC Director shall be accountable to the Vice Chancellor of the Office of Research for the safe operation and maintenance of the reactor and its associated equipment. 6.1.1 Structure. The management for operation of the UCD/MNRC facility shall consist of the organizational structure as shown in Figure 6.1. 6.1.2 Responsibilities. The UCD/MNRC Director shall be accountable to the Vice Chancellor of the 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 all experiments and experiment procedures prior to their use in the reactor. Individuals in the management 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, and shall be responsible for safeguarding the public and facility personnel from undue radiation exposures and for adhering to the operating license and technical specifications. The Site Manager shall report directly to the UCD/MNRC Director. The Reactor Manager and Health Physics 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 reactor operator should be within thirty (30) minutes of the facility and reachable by telephone, and;
- d. A senior reactor operator shall be present whenever a reactor startup is performed, 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 be available 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 and Training of Personnel for Research Reactors (ANS 15.4). Qualification and requalification of licensed operators shall be subject to an approved Nuclear Regulatory Commission (NRC) program. 6.2 Review. Audit. Recommendation and Approval General Policy. Nuclear facilities shall be designed, constructed, operated, and maintained in such 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 with applicable regulatory requirements. The UCO Vice Chancellor of the Office of 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 NSC Composition and Qualifications. The UCD/MNRC Director shall appoint the Chairperson 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. The NSC shall evaluate and review nuclear safety associated with the operation and use of the UCD/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 and proposed Class 1 modifications, prior to implementing (Class I) modifications, prior to taking action under the preceding 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 an unreviewed safety question or require a change to the license, to a Facility Use Authorization, or 32
- 0 to the Technical Specifications. This determination may be in the form of verifying a decision already made by the UCD/MNRC Director.
- d. Review reactor operations and operational maintenance, Class I modification records, and 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 for UCD/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 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 other inspections of these facilities conducted by other agencies.
6.2.4 Audit/Inspection Function. The NSC or a subcommittee thereof, shall audit/inspect reactor operations 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 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 license and Technical 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 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 the UCD/MNRC Radiation Safety Program. The program should use the guidelines of the American National Standard for Radiation Protection at Research Reactor Facilities (ANSI/ANS 15.11). The Health Physics Manager shall report to the Site 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 to assure that procedures are current. Procedures shall be adequate to assure the safe operation of the 33
0 i 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 or other routine maintenance that could have an effect on reactor safety.
- e. Testing and calibration of reactor instrumentation and controls, control rods and control rod drives.
- 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 alarms 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 solid and liquid waste and gaseous effluents released from the facility. The program shall include a management 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 items from 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 Approval. Experiments having similar characteristics are grouped together for review and approval under specific Facility Use Authorizations. All specific experiments to be 34
- 0 performed under the provisions of an approved Facility Use Authorization shall be approved by the UCD/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 approval as a new experiment.
- c. Minor changes to an experiment that do not significantly alter the experiment may be approved by a senior reactor operator.
6.6 Required Actions 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 the UCD/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 and contributing factors. (2) Effect of the violation upon reactor facility components, systems, or structures, and on 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 the NRC 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 is necessary to shut down the reactor to correct the occurrence, operations shall not be resumed unless 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 the NRC Operations Center within 24 hours of the occurrence. A written preliminary report shall be sent 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 the above address within 30 days of the occurrence.
- d. Reportable occurrences should be reviewed by the NSC prior to forwarding any written report to the Vice Chancellor of the Office of Research or to the Nuclear Regulatory Commission.
35
- 0 6.7 Reports. All written reports shall be sent within the prescribed interval to the NRC, Attn: Document Control 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 during the previous calendar year shall be submitted within six months following the end of each calendar year. Each annual report shall include the following information:
- a. A brief summary of operating experiences including experiments performed, changes in facility design, 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 reactor was critical, and the cumulative total energy output since initial criticality.
- c. The number of emergency shutdowns and inadvertent scrams, including reasons for the shutdowns or scrams.
- d. Discussion of the major maintenance operations performed during the period, including the effect, 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 or in procedures, and of tests and experiments carried out pursuant to Section 50.59 of 10 CFR Part 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 as follows: 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 is greater than 1xl0 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 release occurs, 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) released during 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 0 I 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 during the reporting period based on data from an appropriate monitoring system. 3 The estimated maximum annual average concentration of Argon-41 in the unrestricted area (in microcuries/ml), the estimated corresponding annual radiation dose at this location (in millirem), and the fraction of the applicable 10 CFR 20 limits for these values. 4 The total estimated quantity of radioactivity in particulate form with half lives greater than eight days (in curies) released during the reporting 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, by facility users, and by visitors in terms of the average radiation exposure per individual and the greatest exposure per individual in each group.
- h. An annual summary of the radiation levels and levels of contamination observed during routine 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 planned administrative 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 in Section 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 could render the system incapable of performing its intended safety function unless the failure is 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 a condition 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 through 6.7.2.a.8. (2) The written report (and, to the extent possible, the preliminary telephone report or report by similar conveyance) shall describe, analyze, and evaluate safety implications, and outline 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.
(1) Any significant variation of measured values from a corresponding predicted or previously 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 the Safety Analysis Report (SAR); (3) A personnel change involving the positions of UCD/MNRC Director or UCD Vice Chancellor 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 or development 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 required information may be contained in single or multiple records, or a combination thereof. Records and logs shall 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 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 shipped off site.
38
0 0
- 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.
39
... NUCLEARcoMSSoREGUALTORY UNIVERSITY OF CALIFORNIA - DAVIS VICE CHANCELLOR FOR RESEARCH (Licensee) UCD/MNRC UCDIMNRC [ DIRECTOR NUCLEAR ._._ l COMMITJTEE [ 8At-SITE , ' MANAGERi HEALTH PHYFIGUREAC6.1
- EGu Att _; UNITED STATES "
. ** * * *NUCLEAR REGULATORY COMMISSION
*,.,**. *WASHINGTON, D*.C. 20555-0001 March 30, 2004
- Dr. Barry M. Klein Vice Chancellor for Research University of California, Davis One Shields Avenue Davis, CA 95616-8558
SUBJECT:
REVISION TO SAFETY EVALUATION 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 revision to the Safety Evaluation 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 issued on December 30, 2003 and is available on the Commission's ADAMS system, Accession Number ML033421339. Sincerely, _
- WreJ EeIn . rjc aae Research and Test Reactors Section New, Research and Test Reactors Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No. 50-607
Enclosure:
Revision to Amendment No. 7 Safety Evaluation Report cc w/enclosure: Please see next page
0.. University of California - Davis/McClellan MNRC Docket No. 50-607 cc: Mr. Jeff Ching 5335 Price Avei~ue, Bldg. 258 McClellan AFB, CA 95652-2504 Test, Research, and Training Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611
0 UNITEb STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 REVISION TO SAFETY EVALUATION REPORT SUPPORTING AMENDMENT NO. 7 TO AMENDED .FACILITY OPERATING LICENSE NO. R-130 REGENTS OF THE UNIVERSITY OF CALIFORNIA AT McCLELLAN NUCLEAR RADIATION CENTER DOCKET N.O. 50-607
1.0 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 the McClellan Nuclear Radiation Center (MNRC) TRIGA research reactor. The request provided for the 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 Facility Operating License be amended to include an additional section 2.B.(4) as follows: 2.B.(4) Inaddition to those items specified in2.B.(1), 2.B.(2) and 2.B.(3) the following radioactive materials may be received, possessed, and used at the facility. Radioactive Material Chemical and/or Maximum (element and mass number) Physical Form. Possess atQuantity Any OneLicensee Time May A. Any radioactive material A. Any A. 20 curies (1 curie each, except as between atomic number I through provided below) 83, Inclusive B. Any radioactive material with A. Any A. 4 Curies (100 milllcuries each, atomic numbers 84 and above except as provided below) or up to 20 micrograms c.. Iodine-125 c. Iodide/LIquid C. 40 Curies D. Source material (but only trace D. Any D. 4 grams per radionuclide, not to amounts of Th-234) exceed 10 grams total E. Special nuclear material E. Any E. 2 grams per radionuclide, not to exceed 5 grams total This amendment request was approved and issued on .December 30, 2003.
0 " 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 the reactor room glove box. The significance of this assumption is related to the ability of the reactor 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 the above table is 64 curies. The maximum activity In category D is about 0.1 curie, while the maximum 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 the bounding MHA, and the amendment request was approved. Instead of locating all of the radioactive materials shown in above table in the reactor room glove box, some of the materials will be located in the restricted area of the McClellan Nuclear Radiation Center. Non-volatile material will be handled in accordance with approved procedures. 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 the consequences of complete volatilization of the unsealed material (e.g., the reactor room glove box and reactor room fume hood), as previously analyzed. The staff has reviewed the proposed change to the Facility Operating License and concluded that itdoes not impact the licensee's ability to continue to meet the relevant requirements of 10 CFR 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 and surveillance requirements. The staff has determined that this amendment involves no significant 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 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 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 or consequences of accidents previously evaluated, or create the possibility of a new or different kind of accident from 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 is reasonable 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 or the health and safety of the public. Principal Contributor: Warren J. Ereslan Date: March 30, 2004
'5%. i ./ uNITED STATES
*/*"* NLCLEAR REGULATORY COMMISSION 0 ASIGTNDC.205-00 Deceeiber: 30, 2003 Dr. Barry M.Klein Vice Chancellor for Research University of California, Davis One Shields Avenue Davis, 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 to Facility 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 is discussed in the enclosed Safety Evaluation Report. Sincerely,*,*, 69 ~4~tey/ Warren J. Eresian, Project Manager Research and Test Reactors Section New, Research and Test Reactors Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No. 50-607
Enclosures:
- 1. Amendment No. 7
- 2. Safety Evaluation Report
University of California - Davis/McClellan MNRC Docket No. 50-607 cc: Dr. Wade J. Richards 5335 Price Avenue, Bldg. 258 McClellan AFB, CA 95652-2504 Test, Research, and Training Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611
UNITED STATES
"*%- NUCLEAR REGULATORY COMMISSION
- WASHINGTON, D.C. 20555-0001 REGENTS OF THE UNIVERSITY OF CALIFORNIA AT McCLELLAN NUCLEAR RADIATION CENTER DOCKET NO. 50-607 AMENDMENT TO AMENDED FACILITY OPERATING LICENSE Amendment No. 7 License No. R-1 30
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that A. 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 Nuclear Radiation Center (the licensee) on October 21, 2003 and November 6, 2003, conforms to the standards and requirements of the Atomic Energy Act of 1954:, as amended (the Act), and the regulations of the Commission as stated in Chapter I of Title 10 of the Code of FederalRegulations (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 the Commission; D. The issuance of this amendment will not be Inimical to the common defense and security 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; and F. 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. Accordingly, the license is amended by changes to the Facility Operating License as indicated below, and paragraph 2.B of Amended Facility Operating License No. R-130 is hereby amended to read as follows:
2.B.(4) Inaddition 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 Chemical and/or Maximum Quantity (element and mass Physical Form May Possess at AnyLicensee One Time number) A. Any radioactive A. Any A. A. 20 Curies (I Curie each, material between except as provided below) atomic number 1 through 83, inclusive B. Any radioactive A. Any A. 4 Curies (100 millicuries material with atomic each, except as provided numbers 84 and below) or up to 20 above micrograms C. Iodine-125 C. 40OCuries C. Iodide/Liquid
- 0. Any D. 4 grams per radionuclide, D. Source material (but only trace amounts not to exceed 10 grams of Th-234) total E. Special nuclear E. Any E. 2 grams per radionuclide, material not to exceed 5 grams total
- 3. This license amendment is effective as of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Research and Test Reactors Section New, Research and Test Reactors Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Date of Issuance: December 30, 2003
~UNITED O STATES 0
o* NUCLEAR REGULATORY COMMISSION o~WASHINGTON, D.C. 20555-0001 o# SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 7 TO AMENDED FACILITY OPERATING LICENSE NO. R-130 REGENTS OF THE UNIVERSITY OF CALIFORNIA AT McCLELLAN NUCLEAR RADIATION CENTER DOCKET NO. 50-607
1.0 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 the allowance of radioactive materials not produced by the reactor to be received, possessed and used 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 Chemical and/or Maximum Quantity Licensee May (element and mass Physical Form Possess at Any One Time number) A. Any radioactive A. Any A. A. 20 Curies (1 Curie each, material between except as provided below) atomic number 1 through 83, inclusive B. Any radioactive A. 4 Curies (100 mlllicuries A. Any material with atomic each, except as provided numbers 84 and below) or up to 20 above micrograms C. Iodine-I125 C. Iodide/Liquid C. 40 Curies D. Any D. 4 grams per radionuclide, D. Source material (but only trace amounts of not to exceed 10 grams total Th-234) E. Special nuclear E. Any E. 2 grams per radionuclide, material not to exceed 5 grams total
- 0
-2,-
This request is discussed below. 2.0 EVALUATION All of the radioactive materials to be received, possessed and handled In accordance with this amendment request will be located in the reactor room glove box. In November of 2002, the NRC approved 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 reactor 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 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 reactor room for the maximum assumed occupancy time of 5 minutes the CEDE to the thyroid would be about 205 millirem. These doses were compared to the expected doses (CEDE) resulting from the Maximum Hypothetical Accident (MHA), which serves as the bounding accident for radiological 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 the complete 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 box from 61 curies to 64.4 curies. (The total activity in categories A, B, and C In the above table Is 64 curles. The maximum activity in category D corresponds to 10 grams of Uranium-233, or about 0.1 curie. The maximum activity in category E corresponds to 5 grams of Plutonium-239, or about 0.3 curie.) For the complete volatilization of 64.4 curies, doses in the unrestricted area and in the reactor room will scale up proportionally from 61 curies, resulting in a dose in the unrestricted area of 3.2 mlllirem, and a dose in the reactor room of 216 millirem (i.e., the doses have increased by 5.6 percent.) The staff concludes that the consequences of the complete volatilization of 64.4 curies are much less than the consequences of the bounding MHA, and that increasing the allowable activity in the reactor room glove box from 61 curies to 64.4 curies does not significantly reduce the margin of safety 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 that it does not impact the licensee's ability to continue to meet the relevant requirements of 10 CFR Part 50.36.
3.0 ENVIRONMENTAL CONSIDERATION
This amendment does not involve changes in the installation or use of a facility component located within 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 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 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 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 or consequences of accidents previously evaluated, or create the possibility of a new or different kind of accident from 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 is reasonable 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 or the health and safety of the public. Principal Contributor: Warred J. Eresian Date: December 30, 2003
*NUCLEAR REGULATORY COMMISSION ,,.. * ,* UNITED D.C.
WASHINGTON, STATES 20555-0001
,* February 17, 2000 *i7/tlJ*
Brigadier General Michael P. Wiedemer Vice Chancellor Kevin Smith Commander Office of the Chancellor .. ~Sacramento Air Logistics Center University of California, Davis SM-ALCITI-1 One Shields Avenue 5335 Price Avenue Davis, California 95616-8558 McClellan AFB, California 95652-2504
SUBJECT:
RE-ISSUANCE OF NOTICE OF CONSIDERATION OF APPROVAL OF TRANSFER OF FACILITY OPERATING LICENSE NO. R-130 FOR THE McCLELLAN NUCLEAR RADIATION CENTER FROM THE DEPARTMENT OF THE 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 some administrative errors. We. apologize for any inconvenience this may have caused. Sincerely, Ledyard B. Marsh, Chief Events Assessments, Generic Communications and Non-Power Reactors Branch Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No. 50-607
Enclosure:
As stated cc: wlenclosures
McClellan AFB TRIGA REACTORDcktN.0-7 CC: Dr. Wade J. Richards SM-ALC/TI-1 5335 Price Avenue, Bldg. 258 McClellan AFB, CA 95652-2504 Cot. Robert Capell HQ AFMC/SGC 4225 Logistics Avenue, Suite 23 Wright-Patterson AFB, OH 45433-5762 Lt. Col. Catherine Ze~ringue HQ AFSCISEW 9570 Avenue G, Building 24499 Kirtland AFB, New Mexico 871 17-5670 Test, Research, and Training Reactor Newsletter 202 Nuclear Sciences Center University of Florida Gainesville, FL 3261 1
- L0 TECHNICAL SPECIFICATIONS FOR THE UNIVERSITY OF CALIFORNIA - DAVIS MCCLELLAN NUCLEAR RADIATION CENTER (UCDIMNRC)
DOCUMENT NUMBER: MNRC-0004-DOC-13 Rev 13 4/03 p
.~.
0 ! Revision ,13 of the "Technical Specifications" for the University of California at Davis/McClellan Nuclear Radiation ~.1 *>~ Center (UCD/MNRC) Reactor have undergone the following coordination: Reviewed by: Hel-'* !.0 eltPyiSpesoDate Reviewed by: __________* 0. R ator Su~dpervisrDt Approved by: *~Ij.e/o3
~i'I~toY Chairman, UO*/M NRC ~l1.
Date Nuclear Safety Committee I. (K ~/
-- I
0 Technical Specifications Rev 13 412003 Title Page Rev 13 4/2003 Approval Page Rev 13 4/2003 31 Rev 13 4/2003 32 Rev 13 4/2003 33 Rev 13 4/2003 Figure 6.1 Rev 13 4/2003
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* " TECHNICAL SPECIFICATIONS TABLE OF CONTENTS 1.0 Definitions .............................................................................................................. 1 2.0 Safety Limit and Limiting Safety System Setting (LSSS.*........................................................6...
2.1 Safety Limits.................................................................................................. 2.2 Limiting Safety System Setting (LSSS)..................................................................... 6 2.2.1 Fuel Temperature ............................................................ i.................... 6 3.0 Limiting Conditions for Operations (LC.O.) ........................................................................... 7 3.1 Reactor Core Parameters................................................................................... 7 3.1.1 Steady-State Operation ....................................... :................................... 7 3.1.2 Pulse or Square Wave Operation ............................................................... 7 3.1.3 Reactivity Limitation~s............................................................................. 8 3.2 Reactor Control and Safety Systems .... .................................................................. 8 3.2.1 Control Rods...................................................................................... 8 3.2.2 Reactor Instrumentation.......................................................................... 9 3.2.3 Reactor Scrams and Interlocks................................................................. 10 3.2.4 Reactor Fuel Elementts.......................................................................... 12 3.3 Reactor Coolant Systems.................................................................................. 13 3.4 Reactor Room Exhaust System ........................................................................... 14 3.5 Intentionally Left Blank ..................................................................................... 14 3.6 Intentionally Left Blank..................................................................................... 14 3.7 Reactor Radiation Monitoring Systems.................................................................... 14 3.7.1 Monitoring Systems ... ......................................................................... 14 3.7.2 Effluent~s - Argon-41 Discharge Limit. .......................................................... 16 )
0 0 3.8 Experiments ................................................................................................ 16 3.8.1 Reactivity Limits ........................................ *........................................ 16 3.8.2 Materials Limit................................................................................... 17 3.8.3 Failure and Malfunctions ...................... ................................................. 18 4.0 Surveillance Requirements.......................................................................................... 19 4.1 Reactor Core Parameters ................................................................................. 19 4.1.1 Steady State Operation......................................................................... 19 4.1.2 Shutdown Margin and Excess Reactivity ....................................................... 20 4.2 Reactor Control and Safety Systems...................................................................... 20 4.2.1 Control Rods ................................................... ................................. 20 4.2.2 Reactor Instrumentation............................................................... *......... 21 4.2.3 Reactor Scrams and interlocks.................... ............................................. 22 4.2.4 Reactor Fuel Elements................................ .................................... 23 4.3 Reactor Coolant Systems ................................................................................. 24 4.4 Reactor Room Exhaust System ........................................................................... 25 4.5 Intentionally Left Blank..................................................................................... 25 4.6 Intentionally Left Blank..................................................................................... 25 4.7 Reactor Radiation Monitoring Systems.................................................................... 25 4.8 Experiments ................................................................................................ 26 5.0 Design Features ........................................................... ;......................................... 27 5.1 Site and Facility Description ............................................................................... 27 5.1.1 .Site............................................................................................... 27 5.1.2 Facility Exhaust ......................... ....................................................... 28 5.2 Reactor Coolant system ................................................................................... 28
5.3 Reactor Core and F.uel .................................................................................... 29 5.3.1 Reactor Care .................................................................................... 29 5.3.2 Reactor F..u~l..................................................................................... 30 5.3.3 Control Rods and Control Rod Drives.......................................................... 31 5.4 Fissionable Material Storage............................................................................... 31 6.0 Administrative Controls ............................................................................................ .31 6.1 Organization................................................................................................ 31 6.1.1 Structure......................................................................................... 32 6.1.2 Responsibilities ................................................................................. 32 6.1.3 Staffing .......................................................................................... 32 6.1.4 Selection and Training of Personnel ........................................................... 32 6.2 Review, Audit, Recommendation and Approvial........................................................... 32 6.2.1 NSC Composition and Qualifications........................................................... 33 6.2.2 NSC Charter and Rules .......................... i.............................................. 33 I 6.2.3 Review Function................................................................................. 33 6.2.4 Audit/Inspection Function ....................................................................... 34 6.3 Radiation Safety............................................................................................ 34 1 6.4 Procedures ................................................................................................. 34 6.4.1 Reactor Operations Procedur~es................................................................. 34 I 6.4.2 Health Physics Procedures ..................................................................... 35 6.5 Experiment Review and Approlval ......................................................................... 35 6.6 Required Actions ........................................................................................... 35 6.6.1 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................................................................................. 36 6.7.2 Special Reports ................................................................................... 38 6.8 Records........................................................................................................ 39 Fig. 6.1 UCD/MNRC Organization for Licensing and Operation........................................................ 40
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- 0 TECHNICAL SPECIFICATIONS FOR THE UNIVERSITY OF CALIFORNIA - DAVIS/MCCLELLAN NUCLEAR RADIATION CENTER (UCDIMNRC)
The University of California - Davis/McClellan Nuclear Radiation Center (UCD/MNRC) reactor is operated by the University 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 provides a wide range of irradiation services for both research and industrial needs. The reactor operates at a nominal steady state power level up to and including 2 MW. The UCD/MNRC reactor is also capable of square wave and 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 Nuclear Regulatory Commission (e.g., senior reactor operator or reactor operator) to carry out the duties and responsibilities associated with the position requiring the license. 1.2.1 Sernior Reactor Operator. An individual who is licensed to direct the activities of reactor operators 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 facility and perform reactor-related maintenance. 1.3 Channel. A channel is the combination of sensor, line amplifier, processor, and output devices which are connected for the purpose of measuring the value of a parameter. 1.3.1 Channel Test. Achannel 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 is*n adjustment of the channel such that its output corresponds with acceptable accuracy to known-'-values of the parameter which the channel measures. Calibration shall encompass the entire channel,.including equipment actuation, alarm or trip, and shall be deemed to include a channel test. " 1.3.3 Channel. Check. A channel check is a qualitative verification of acceptable performance by observation of channel behavior. This verification, where possible, shall include comparison of the channel with other independent channels or systems measuring the same variable. 1.4 Confinement. Confinement means isolation of the reactor room air volume such that the movement of 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, fission chambers, foils, etc), which is designed to investigate specific reactor characteristics or which is intended for irradiation within an experiment facility and which is not rigidly secured to a core or shield structure 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 entire experiment 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, or component 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 are normal to the operating environment of the experiment, or by forces which can arise as a result of 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 their associated input circuits, which are designed to initiate a scram for the primary purpose of protecting an experiment or to provide information which requires manual protective action to be initiated. ' 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 can be either 8.5, 20, or 30 wt%, with an enrichment of less than 20% U-235. A standard fuel element may contain a burnable poison. 1.7 Fuel Element., Instrumented. An instrumented fuel element is a standard fuel element fabricated with thermocouples for temperature measurements. An instrumented fuel element shell have at least one operable 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 a channel. 1.9 Mode. Steady-State. Steady-state mode operation shall mean operation of the UCDIMNRC reactor with 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/MNRC reactor 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 the selector 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 within the UCDIMNRC reactor safety system in response to a variable or condition of the UCDIMNRC reactor facility having 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 the generation 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 i: equipment. 2
1.15.3 Instrument generation System Level. At the protective instrument level, protective action is the 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 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 the maximum allowable pulse reactivity insertion has been determined. 1; 17 Reactivity. Exce~ss, Excess reactivity is that amount of reactivity that would exist ifall control rods (control, regulating, etc.) were moved to the maximum reactive position from the point where the reactor is 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 the reactor core. 1.19 R~eactivity Worth of an Exoeriment. The reactivity worth of an experiment is the maximum value of the reactivity 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 which directly 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 the parameters 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 in these 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 for initiation of manual protective action. 1.24 R~eactor Secured. The UCO/MNRC reactor is secured when the console key switch is in the off position and the key is removed from the lock and under the control of a licensed operator, and the conditions of a or b exist:
- a. (1) The minimum number of control rods are fully inserted to ensure the reactor is shutdown, as required by 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 being moved or serviced ifthe experiments have, on movement, a reactivity worth exceeding the maximum value allowed for a single experiment or $1.00, whichever is smaller, or
- b. The reactor contains insufficient fissile materials in the reactor core, a~djacent experiments or control rods to attain criticality under optimum available conditions of moderation and reflection.
1.25 Reactor Shut~down. The UCD/MNRC reactor is shutdown ifit 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
- 0 1.26 Reference Cpre Condition. The condition of the core when it is at ambient temperature (cold T<280 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 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, and which 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 a fuel 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 allow it to perform its control function, and its safety function when the coupling is disengaged. 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 power level and 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 of radioactivity. 1.31 Scram Time. Scram time is the elapsed time between reaching a limiting safety system set point and the 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 denote a recommendation; the word "may" to denote, permission, neither a requirement nor a recommendation. 1.34 Shut~down Margin. Shutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence that the reactor can be made subcritical by means of the control and safety system starting from any permissible operating condition with the most reactive rod assumed to be in the most reactive position, and once this action has been initiated, the reactor will remain subcritical without further operator action. 1.35 Shutdown. Unsched.u.led. An unscheduled shutdown is any unplanned shutdown of the UCD/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 safe operation, 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, weekly or quarterly. Calibration times are generally specified as quarterly, semi-annually, annually, or biennially. ...../) 1.37 to Surveillance reduce Intervals. frequency. Maximum Established intervalsshall frequencies are established to provide be maintained over theoperational long term. flexibility and not The allowable 4
0 0*** surveillance interval is the interval between a check, test, or calibration, whichever is appropriate to the item being 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 involve an unreviewed safety question:
- a. If the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or
- b. If a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or
- c. 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 reactor computerized 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 fuel element. 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 during steady-state operation.
- b. The maximum ten'perature in a standard TRIGA fuel element shall not exceed 1100 0C during pulse operation.
....... 'a. This fuel safety limit applies for coniditions in which the cladding temperature is above 500 °C (Safety Analysis Report (SAR), Chapter 4, Section 4.5.4.1.3). The important parameter for a TRIGA reactor is 5
0 =° the fuel element temperature. This parameter is well suited as it can be measured directly. A loss in the integrity of the fuel element cladding could arise ifthe cladding stress exceeds the ultimate strength of the cladding material. The fuel element cladding stress is a function of the element's internal pressure while the ultimate strength of the cladding material is a function of its temperature. The cladding stress is a result of the internal pressure due to the presence of air, fission product gasses and hydrogen from the disassociation 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 hydrogen to zirconium in the alloy. At a fuel temperature of 930 0C for ZrH 1.7 fuel, the cladding stress due to the internal pressure is equal to the ultimate strength of the cladding material at the same temperature (SAR Fig 4.18). This is a conservative limit since the temperature of the cladding material is always 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 a pulse which gives a peak adiabatic fuel temperature of 1000 °C is estimated to be 470 °C. Further analysis (SAR Section 4.5.4.1.2), shows that the internal pressure for both Zr *.s (at 1150 0 C) and Zr 17z (at 11 00°C) increases to a peak value at about 0.3 sec, at which time the pressure is about one-fifth of the equilibrium value or about 400 psi (a stress of 14,700 psi). The yield strength of the cladding at 500 0C is about 59,000 psi. Calculations for step increases in power to peak ZrH 1.85 fuel temperature greater than I115 0 C, over a 200°C range, show that the time to reach the peak pressure and the fraction of equilibrium pressure value achieved were 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 and compared 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 (ZrH 1.=) was only about 6% of the equilibrium value evaluated at the peak temperature. Calculations of the pressure gave values about three times greater than the measured values. The analysis gives strong indications that the cladding will not rupture iffuel 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 for cladding failure. 2.2 Limiting Safety System Setting. 2.2.1 Fuel Temperature. Applicability - This specification applies to the protective action for the reactor fuel element temperature. Obiective - The objective is to prevent the fuel element temperature safety limit from being reached. Specification - The limiting safety system setting shall be 750 0C (operationally this may be set more conservatively) as measured in an instrumented fuel element. One instrumented element shall 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 a temperature which, ifexceeded, shall cause a reactor scram to be initiated preventing the safety limit from being exceeded. A setting of 750 °C provides a safety margin at the point of the measuremenrA 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 measured temperatures 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 the center and mid-plane of the fuel element. For pulse operation of the reactor, the same limiting safety system setting applies. However, the temperature channel will have no effect on limiting 6
O ...... 0 the peak power generated because of its relatively long time constant (seconds) as compared with the width of the pulse (milliseconds). in this mode, however, the temperature trip will act to
"\ limit the energy release after the pulse ifthe transient rod should not reinsert and the fuel ! temperature continues to increase.
3.0 Limiting Conditions For Operat~ion* 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 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. 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. For the purpose 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/MNRC TRIGA 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 Aoolicabilitv - This specification applies to the peak temperature generated in the fuel as the result of a step insertion of reactivity.
Obiective - The objective isto 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 to 1.7. 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 state operation at two (2) MW the hydrogen will redistribute due to migration from the central high temperature regions of 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 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. Ifthe pulse insertion is such that the temperature of the fuel exceeds about 875 0C, then the pressure may be sufficient to cause expansion of microscopic 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 the reactivity 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 at all 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 by the 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) shall not exceed 6.65% Ak/k ($9.50).
Basis -
- a. This specification assures that the reactor can be placed in a shutdown condition from any operating condition and remain shutdown, even if the maximum worth control rod should stick in the fully withdrawn position (SAR Chapter 4, Section 4.5.5).
- b. This specification sets an overall reactivity limit which provides adequate excess reactivity to override the xenon buildup, to overcome the temperature change in going from zero power to 2 MW, to permit pulsing at the $1.75 level, to permit irradiation of negative worth experiments and account for fuel burnup over time. An adequate shutdown margin exists with an excess of $9.50 for the two analyzed cores: (SAR Chapter 4, Section 4.5.5).
3.2 Reactor Control and Safety Systems 3.2.1 Control Rods Aooljcjili*- 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 ifdamage is apparent to the rod or drive assemblies.
- b. The scram' time measured from the instant a signal reaches the value of a limiting safety system setting to the instant that the slowest control rod reaches its fully inserted position shall
. not exceed one (1) second.
8
0 S
- a. The apparent condition of the control rod assemblies shall provide assurance that the rods shall continue to perform reliably as designed.
- b. This assures that the reactor shall shut down promptly when a scram signal is initiated (SAR Chapter 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 reactor operator during reactor operations. Obiective - The objective is to require that sufficient information is available to the operator to assure safe operation of the reactor. Specification - The reactor shall not be operated unless the channels described in Table 3.2.2 are operable and the information is displayed on the reactor console. Table 3.2.2 Required 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,A Level Safety MW or less Channel b5.Linear Power 101Automatic D,M,A Channel -Power Control
- c. Log Power 101Startup D,M,A Channel Control
- d. Fuel Temperature 2 2 2 Fuel D,M,A Channel Temperature
- e. Pulse Channel 0 10Measures P,A Pulse NV & NVTr
(*) Where: D - Channel check during each day's operation M - Channel test monthly A - Channel calibration annually P - Channel test prior to pulsing operation
- a. Table 3,2.2. The two reactor power level safety channels assure that the reactor power level is 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 assure that the reactor power level and energy are adequately monitored (SAR Chapter 7, Sections
.... 7.1.2 & 7.1.2.2). 9
Ia1,.1.2 monitored and The fuel temperature indicated in the reactorchannels assure(SAR control room that Chapter the fuel temperature is properly 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 in Table 3.2.3 are operable: Table 3.2.3 Required Scrams and I!nterlocks Steady Square Channel Surveillance Scram State Pulse Wave Reouirements*
- a. Console I I .1 Manual Scram M Manual and Automatic Scram Scram Alarm
- b. Reactor Room 1 I Manual Scram M Manual Scram and Automatic Scram Alarm 4 4 Manual Scrams M
- c. Radiography Bay Manual 4 and Automatic Scrams Scram Alarms
- d. Reactor Power 2 0 2 Automatic M Level Safety Scram Alarms & Scrams Scrams at 2.2 MW or less
- e. High Voltage 2 1 2 Automatic M Power Supplies Scram Alarms &
Scrams *i Scrams onLosf High Voltage to the Reactor Power Level Safety Channels
- f. Fuel 2 2 2 Automatic Scram M Temperature Alarms & Scrams Scrams on indicated fuel temperature of 750°C or less
- g. Watchdog 2 2 2 Automatic Scram M
.Circuit Alarms & Scrams 10 N.,
0
- h. External 2 2 Automatic M Scrams Scrams and Alarms
\ ifan experiment or radiography scram interlock
*is activated i.One Kilowatt 0 1 Prevents initiation M Pulse & of a step reactivity Square Wave insertion above a Interlock reactor power level of 1 KW
- j. Low Source Prevents withdrawal M 1 1 Level Rod of any control rod Withdrawal if the log channel Prohibit reads less than 1.5 Interlock times the indicated log channel current level with the neutron source removed from the core
- k. Control Rod Prevents simul- M Withdrawal I 1 taneous withdrawal Interlock of two or more rods in manual mode I. Magnet De-energizes the 1 1 1 M Power Key control rod Switch Scram magnets, scram &
alarm (*)Where: M- channel test monthly Basis -
- a. Table 3.2.3. The console manual scram allows rapid shutdown of the reactor from the control room (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 the reactor room.
p,.Table 3.2.3. The radiography bay manual scrams allow rapid shutdown of the reactor from any 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 if the power level exceeds 2.2 MW, therefore not exceeding the safety limit (SAR Chapter 4, Section 4.7.2).
- e. Table 3,.2,*. The loss-of-high-voltage scram assures that the reactor power level safety channels, operate within their intended range as required for proper functioning of the power level 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 ifthe fuel temperature exceeds 7500o C, therefore ensuring the safety limit will not be exceeded (SAR Chapter 4, Sections 4.5.4.1 &4.7.2).
11
- a. Table 3.2.3, acquisition The watchdog computer circuitsproperly are functioning assure that (SARtheChapter control 7, system computer Section 7.2). and the data
- h. Table.3.2.3, The external scrams assure that the reactor will be shut down ifthe radiography bay doors and reactor concrete shutters are not in the proper position for personnel entry into the bays (SAR Chapter 9, Section 9.6). External scrams from experiments, a subset of the external scrams, also assure the integrity of the reactor system, the experiment, the facility, and the 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 above one (1) kilowatt assures that the pulse magnitude will not allow the fuel element temperature to exceed 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 source of 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 two or more control rods thus limiting the reactivity-insertion rate from the control rods in manual mode (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 rod magnets and results in a scram (SAR Chapter 7, Section 7.1.2.5). 3.2.4 Rea~ctor Fuel Elements Aoolicabilitv - This specification applies to the physical dimensions of the fuel elements as measured 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 fuel elements 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 core if:
- a. In measuring the transverse bend, the bend exceeds 0.125 inch (3.175 mam) over the full length 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 the reactor, during which differential expansion between the fuel and the cladding occurs and the pressure of the gases within the elements increases sharply. The above limits on the allowable distortion of a fuel element correspond to strains that are considerably lower than the strain expected to ".ause rupturing of a fuel element. Limited operation in the steady state or pulsed mode may be necessary to identify a leaking fuel element especially if the leak is small. 12
3.3 Reactor Coolant Systems Aoolicability - 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 channels described in Table 3.3 are operable, and the information is displayed locally or in the control room. Table 3.3 REQUIRED II WATER SYSTEMS I n I AND INSTRUMENTATION
* . ...
Minimum Measuring Number Surveillance Chann~el/System Ooe~rable. Requirements* Function: Channel/System
- a. Primary Coolant I For operation of the D,Q,A Core Inlet reactor at 1.5 MW or Temperature higher, alarms on high Monitor heat exchanger outlet temperature of 45 °C (113°F)
- b. Reactor Tank 1 Alarms ifwater level M Low Water drops below a depth of
*Monitor 23 feet in the reactor tank
(.
- c. Purification** 1 Alarms ifthe primary DM,S Inlet Conduc- coolant water conductivity tivity Monitor is greater than 5 micromhos/cm
- d. Emergency Core I For operation of the reactor D,S Cooling System at 1.5MW or higher, provides water to cool fuel in the event of a Lois of Coolant Accident for a minimum of 3.7 hours at 20 gpm from an appropriate nozzle
(*)Where: AD-- channel check during channel calibration each day's operation annually Q - 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 shall be shutdown. Basis - a.Table 3.3, The primary coolant core inlet temperature alarm assures that large power fluctuations will not 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 there is 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 ifthe water level drops
) below aheight 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 a week 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; System Aoplicability - 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 reactor 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 the pressure 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 Coolant 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 room will be reduced due to mixing with exhaust system air (SAR Chapter 9, Section 9.5.1). Pressure in the reactor room will be negative relative to surrounding areas due to air flow patterns created by the reactor room 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 fuel temperature 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 Systems AnoDlicability - This specification applies to the information which shall be available to the reactor operator 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 3.7.1 are operable, the readings are below the alarm setpoints, and the information is displayed in the 14
control reactor room. The stack and reactor room CAMS shall not be shutdown at the same time during operation. Table 3.7.1 REQUIRED RADIATION MONITORING INSTRUMENTATION Minimum Measuring Number Channel Surveillance Eduioment O, erable** Function Requirements*
- a. Facility I Monitors Argon-41 and D,W,A Stack Monitor radioactive particu-lates, and alarms
- b. Reactor Room 1 Monitors the radiation D),W,A Radiation level in the reactor Monitor room and alarms
- c. Purification 1 Monitors radiation D:,W,A System Radia- level at the demineral-tion Monitor izer station and alarms
- d. Reactor Room Monitors air from the D,W.A 1
Continuous reactor room for parti-Air Monitor culate and gaseous radioactivity and alarms (*)Where: D -- channel A check during channel calibration each day'ls operation annually W - 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 the release of radioactive material to the environment (SAR Chapter 11, Section 11.1.1.1.4). The alarm setpoint 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 the operations area.
- b. Table 3.7.1. The reactor room radiation monitor provides information regarding radiation levels in the reactor 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 provides information regarding radioactivity in the primary system cooling water (SAR Chapter 11, Section 11.1.5.4.2) 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 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. 15
3.7.2 .Effluents -.Argon-41 Discharge Limit AppJicaiity~- This specification from the UCD/MNRC applies to the concentration of Argon-41 that may be discharged 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 releases of this radionuclide from the UCD/MNRG, and the corresponding annual radiation dose from Argon-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 is specified in Appendix B, Table 2, Column 1 of 10 CFR Part 20.10 CFR 20.1301 specifies dose limitations in the unrestricted area. 10 CFR 20.1101 specifies a constraint on air emissions of radioactive material to the environment. The SAR Chapter 11, Section 11.1.1.1.4 estimates that the routine Argon-41 releases and the corresponding doses in the unrestricted area will be below these limits. 3.8 Experiments 3.8.1 React~ivity Limnits. Applicability - This specification applies to the reactivity limits on experiments installed in specific reactor experiment facilities. Obiective - The objective is to assure control of the reactor during the irradiation or handling of experiments 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 (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 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 or 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, and in any other reactor in-core and in-tank irradiation facilities at any given time shall be less than one dollar and ninety-two cents ($1.92) (1.34% A k/k), including the potential reactivity which might 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 single movable 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 limit of $1 .75 is not exceeded (SAR Chapter 13, Section 13.2.2.2.1 ). In addition, limiting the worth of each movable experiment to less than $1.00 will assure that the additional increase in transient 16
power Chapter (SAR and temperature will13.2.2.2.1). 13, Section be slow enough so that the fuel temperature scram will be effective
- b. The absolute worst event which may be considered in conjunction with a single secured experiment is its sudden accidental or unplanned removal while the reactor is operating. For such 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 Chapter 10, 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 control over the positioning of samples into and out of the central core region 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 less thanthe 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 the fuel temperature safety limit.
3.8.2 .Materials Limit Aoplicability - 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 limiting material quantity and the radioactive material inventory of the experiment.
.Specification - The reactor shall nct 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 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 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 total inventory of 1-125 being handled in the 1-125 fume hood at any one time in preparation for shipping shall not exceed 20 curies. An Additonal 1.0 curie of 1-125 (up to 400 millicuries in the form of quality assurance samples and up to 600 millicuries in sealed storage containers) may also be present in the 1-125 fume hood.
- e. Explosive .materialsin quantities greater than 25 milligrams of TNT eqluivalent shall not be irradiated in th~e reactor tank. Explosive materials in quantities of 25 milligrams of TNT equivalent or less may be irradiatedl 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 any radiography bay. The irradiation of explosives in any bay is limited to those
.17
...... " 0 assemblies where systems the reactor safety a safety analysis has been(SAR upon detonation performed Chapterthat 13,shows Sectionthat there is no damage to 13.2.6.2). ") Basis -
- a. Appropriate 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 fueled experiment leading to total release of the iodine, occupational doses and doses to members 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 is totally released into its respective containment, or if both inventories are simultaneously released into their respective containments, the occupational doses and doses to members of the general public in the unrestricted areas will 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 of 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 any radiography bay external to the reactor tank will not result in damage to the reactor controls or the 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 of radioactive 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 and type of material in the experiment shall be limited such that the airborne radioactivity in the reactor room will not result in exceeding the applicable dose limits in 10 CFR Part 20 in 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) Ifthe 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) Ifthe 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 by boiling this material can escape only through an undistributed column of water above the core, at least 10% of these vapors can escape.
- c. If a capsule fails and releases material which could damage the reactor fuel or structure by corrosion 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 his designated 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 reactor room or the unrestricted area will result in exceeding the applicable dose limits in 10 CFR Part 20.
- b. These assumptions are used to evaluate the potential airborne radioactivity release due to an experiment 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 be brought to the attention of the UCD/MNRC Director or his designated alternate for review to assure 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 reactor operating 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 time period. Prior to reactor system or measuring channel operation, the surveillance shall be performed for each reactor 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 level monitoring channels. O biecltive - The objective is to verify that the maximum power level of the reactor does not exceed 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 required to 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
................................................................ ~
....... /the Aplcbiiy reactor core.These specifications apply to the surveillance requirements for reactivity control of betve- The objective is to measure and verify the reactivity worth, performance, and operability 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 shutdown margin 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 required to perform the surveillance. Basis -
- a. The reactivity worth of the control rods is measured to assure that the required shutdown margin is available and to provide an accurate means for determining the excess reactivity of the 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 been made.
- 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 change unexpectedly.
4.2 Reactor Control and Sa~fet,! Systems 4.2.1 Control Rods Applicability - This specification applies to the surveillance of the control rods. Objective - The objective is to inspect the physical condition of the reactor control rods and establish the operable condition of the rods. Spoecification - Control rod worths shall be determined annually or after physical removal or any significant change in core or control rod configuration.
- a. Each control rod shall be inspected at annual intervals by visual observation of the fueled sections and absorber sections plus examination of the linkages and drives.
- b. The scram time .ofeach control rod shall be measured semiannually.
I. Discovery of noncompliance with Technical Specifications 4.2.1 .a-b shall limit reactor operations to that required to 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.
- 0 configuration provides information about changes in reactor total reactivity and individual rod worths. 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 assure operable 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 fuel temperature 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 the reactor power level safety channels operate properly.
\J b. A channel test monthly of the reactor power level multi-range channel will assure that the channel is operable and responds correctly. The channel check will assure that the reactor power 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 is operable and responds correctly. A channel check of the reactor power level wide range log channel will assure that the channel is operable on a daily basis. A channel calibration will assure 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, will assure that the fuel temperature measuring channels operate properly.
- e. A channel test prior to pulsing plus the annual channel calibration will assure the pulse energy integrating 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. 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 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. 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 is operable.
- c. A channel test monthly of the Radiography Bay Manual Scrams will assure that the scrams are operable.
- d. A channel test monthly of the Reactor Power Level Safety Scrams will assure that the scrams are operable.
- e. A channel test monthly of the Loss-of-High-Voltage Scram will assure that the high voltage power supplies are operable and respond correctly.
- f. A channel test monthly of the Fuel Temperature Scrams will assure that the scrams are operable.
- g. A channel test monthly of the Watchdog Circuits Scrams will assure that the scram circuits are operable.
- h. A channel test monthly of the External Scrams will assure that the scrams are operable and respond 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 is operable.
I. A channel test monthly of the Magnet Current Key Switch will assure that thescram is operable. 4.2.4 Reactor Fuel Element~s A**iailitv- 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 of instrumented fuel elements), and all elements adjacent to the central irradiation facility shall be inspected annually.
- b. Instrumented fuel elements shall be inspected ifany 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 to perform the surveillance.
Basis (Technical Specifications 4,2,4.a-b) - The above specifications assure that the lead fuel elements shall be inspected regularly adteintegrity.o h edfe lmnssalb maintained. These are the fuel elements with the highest power density as analyzed in the SAR Chapter 4, Section 4.5.5.6. The instrumented fuel element is excluded to reduce the risk of damage to the thermocouples.
4.3 Reactor Coolant Systems Aoolicability - 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, the tank water level alarm, the water conductivity cells and the emergency core cooling system are all operable.
.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) Achannel 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 to perform the 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 input signal. 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 input 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 for power levels above 1.5 MW. A channel calibration semiannually assures that the Emergency Core Cooling System performs as required for power levels above 1.5 MW.
*"L'1
4.4 Reactlor Room Exha~ust System
\ Applicability - This specification applies to the surveillance requirements for the reactor room exhaust 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's operation. 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 verify that the exhaust system is maintaining a negative pressure in the reactor room relative to the surrounding facility areas. 4.5 This section intentionally left blank 4.6 This section intentionally left blank. . 4.7 ,Rea~ctor Radiation Monitoring Systems Applicability - 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) Achannel 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 to perform the surveillance. Basis -
- a. A channel check of the facility stack monitor system during each day's operation will assure the monitor is operable. A channel test weekly will assure that the system responds correctly to a 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 assure that the monitor is operable. A channel test weekly will ensure that the system responds to a known source. A channel calibration of the monitor annually will assure that the monitor reads correctly.
- c. Achannel 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 system responds to a known source. 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 responds correctly to a known source. A channel calibration annually will assure that the CAM reads 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 may damage the reactor or release excessive amounts of radioactive materials as a result of experimenta* failure. Soecification -
- a. A new experiment shall not be installed in any UCD/MNRC reactor experiment facility until a written safety analysis has been performed and reviewed by the UCD/MNRC Director, or his designee, 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 approved Facility Use Authorization. Facility Use Authorizations and experiments carried out under these authorizations 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 UCDIMNRC experiment facility until it has been reviewed for compliance with the approved experiment and Facility Use Authorization 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 other UCD/MNRC reactor in-core or in-tank irradiation facility shall be estimated or measured, as
.................
~...............
---. '
Iff_/,/ shall be donebefore appropriate, at ambient reactorconditions. operation with said experiment. Whenever a measurement is done it
- d. Experiments shall be identified and a log or other record maintained while experiments are in any UCD/MNRC reactor experiment facility.
Basis - a & b. Experience at most TRIGA reactor facilities verifies the importance of reactor staff and safety committee reviews of proposed experiments.
- c. Measurement of the reactivity worth of an experiment, or estimation of the reactivity worth based 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 will facilitate maintaining surveillance over such experiments.
5.0 Design Feat~ures 5.1 Site and Facility Description!.- 5.1.1 Site Applicability - This specification applies to the UCD/MNRC site location and specific facility design 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 downtown Sacramento, 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. The unrestricted 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 has 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 of the 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 bnclosi 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 to be 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 of radioactivity released into the environment, or during a Loss of Coolant Accident, the system is to 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 and exhaust air from the UCD/MNRC facility. The system shall have an exhaust stack height of a minimum of 18.2rn (60 feet) above ground level.
- b. Manually activated shutdown controls for the exhaust system shall be located in the reactor control room.
Basis - The UCD/MNRC facility exhaust system is designed such that the reactor room shall be maintained at a negative pressure with respect to the surrounding areas. The free air volume within the 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 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 to operating personnel. 5.2 Reactor Coolant System Applicability - This specification applies to the reactor coolant system.
.Obiective - The objective is to assure that adequate water is available for cooling and shielding during normal 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 ifthe water level in the reactor tank drops below a depth of 23 ft.
- c. For operations at 1.5 MW or higher during a Loss of Coolant Accident the reactor core shall be cooled for 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 be exceeded 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 and i 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 RatrCr Aorolicalbility - This specification applies to the configuration of the fuel. Obiective - The objective is to assure that provisions are made to restrict the arrangement of fuel elements 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 of 96 or more fuel elements to include fuel followed control rods. Below 0.5 MW there is no minimum 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 be loaded 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 20E core 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 done before 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 above loading restrictions will be allowed in an operational 0.5 MW or greater core. Specifications for the 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 Argonne National Laboratory Report ANLIED 97-54). 5.3.2 Reactor Fuel I Applicability - 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 their physical and nuclear characteristics.
- Sp~ecification.-
characteristics: The individual unirradiated TRIGA fuel elements shall have the following 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 (inthe 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 limiting operation to 2.3 megawatts steady state or to a 36 megawatt-sec pulse assures an ample margin of safety between 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 equal to 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 0C during any condition of normal operation, 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 of purchased materials.
5.3.3 Contr~ol .Rodsand Control Rod Drives
.. / A oolicabilitv - This specification applies to the control rods and control rod drives used in the reactor core.
Obiective - The objective is to assure the control rods and control rod drives are of such a design 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 and regulating rods shall have fuel followers sealed in stainless steel. The transient rod shall have an air 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 assure mechanical st~bility during movement and to isolate the neutron poison from the tank water environment. 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"rapid insertion capability. These drives have been tested and proven in many TRIGA reactors.
- A 5.4 Fissionable Materiall Storaae
.... " ADp1icabilitraco coe. This specification applies to the storage of reactor fuel at a time when itis nat in the reacto core
- Objective - The objective is to assure that the fuel which is being stored will not become critical and will not reach an unsafe temperature.
* -a. All fuel elements not inthe reactor core shall be stored (wet or dry) in a geometrical array where the kef 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 Controls 6.1 Organization. The Vice Chancellor for Research shall be the licensee for the UCDIMNRC. The UCD/MNRC facility shall be under the direct control of the UCD/MNRC Director.: The UCD/MNRC Director shall be accountable to the Vice Chancellor for Research for the safe operation and maintenance of the fac~ility.
*6.1.1 'Struciture. The management for operation of the UCD/MNRC facility shall consist of the organizational 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 the operating license and technical specifications. The Operations Manager shall report directly to the UCDIMNRC Director, and shall immediately report all items involving safety and licensing to the Director for a final decision. The Reactor Supervisor and Health Physics Supervisor report directly to the Operations Manager. 6.1.3 Stffn 6.1.3.1 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 reactor operator should be within thirty (30) minutes of the facility and reachable by telephone, and;
- d. A senior reactor operator shall be present whenever a reactor startup is performed, 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 be available 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 licensed operators shall be subject to an approved Nuclear Regulatory Commission (NRC) program.
6.2 Review. Audit. Recommendation anld Approval Genleral Policy. Nuclear facilities shall be designed, constructed, operated, and maintained in such 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 with applicable 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 shall appoint the Chairperson 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. The NSC shall evaluate and review nuclear safety associated with the operation and use i' 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 and proposed Class I modifications, prior to implementing (Class I) modifications, prior to taking action under the preceding documents or prior to forwarding any of these documents to the i' Nuclear Regulatory Commission for approval.
i*......"Jc. Review and determine whether a proposed change, test, or experiment would constitute an unreviewed safety question or req.uire a change to the license, to a Facility Use Authorization, or 32
to the Technical Specifications. This determination may be in the form of verifying a decision already 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 for UCD/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 .ofsuch 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 other inspectionsof these facilitieg conducted by other agencies.
-*-6.2.4 Audit/Inspection Function. The NSC or a subcommittee thereof, shall audit/inspect reactor operations 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 license and Technical 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 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). The Health 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 to assure that procedures are current, Procedures shall be adequate to assure the safe operation of the 33
reactor, but shall ..... *Procedures shall not preclude be in effect forthetheusefollowing of independent items: judgment and action should the situation require. 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 other routine maintenance that could have an effect on reactor safety.
- e. Testing and calibration of reactor instrumentation and controls, control rods and control rod drives.
- 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 alarms 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 solid and liquid waste and gaseous effluents released from the facility. The program shall include a management 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 items from 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 together for 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 the UCD/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 approval as a new experiment.
- c. Minor changes to an experiment that do not significantly alter the experiment may be approved by a senior 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 the UCD/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 and contributing factors. (2) Effect of the violation upon reactor facility components, systems, or structures, and on 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 the NRC 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 is necessary to shut down the reactor to correct the occurrence, operations shall not be resumed unless 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 the NRC Operations Center within 24 hours of the occurrence. A written preliminary report shall be sent to the NI*.C, 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 the above address within 30 days of the occurrence.
- d. Reportable occurrences should be reviewed by the NSC prior to forwarding any written report 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: Document Control 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 during the previous calendar year shall be submitted within six months following the end of each calendar year. Each annual report shall include the following~information:
- a. A brief summary of operating experiences including experiments performed, changes in facility design, 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 reactor was critical, and the cumulative total energy output since initial criticality.
- c. The number of emergency shutdowns and inadvertent scrams, including reasons for the shutdowns or scrams.
- d. Discussion of the major maintenance operations performed during the period, including the effect, ifany, 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 or in procedures, and of tests and experiments carried out pursuant to Section 50.59 of 10 CFR Part 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 as follows: 1 The total estimated quantity of radioactivity released (in curies). 2 An estimation of the specific activity for each detectable radionuclide present ifthe specific activity of the released material after dilution is greater 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 release occurs, 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) released during 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:
1determined The total estimated quantity sampling by an appropriate of radioactivity releasedmethod. and counting (in curies) 2 The total estimated quantity (in curies) of Argon-41 released during the reporting period based on data from an appropriate monitoring system. 3 The estimated maximum annual average concentrationof Argon-41 in the unrestricted area (in microcuries/mi), the estimated corresponding annual radiation dose at this location (in millirem), and the fraction of the applicable 10 CFR 20 limits for these values. 4 The total estimated quantity of radioactivity in particulate form with half lives greater than eight days (in curies) released during the reporting 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, by facility users. and-by visitors in terms of the average radiation exposure per individual and the greatest exposure per individual in each group.
- h. An annual summary of the radiation levels and levels of contamination observed during routine 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 planned administrative 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 in Section 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 could render the system incapable of performing its intended safety function unless the failure is 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 a condition 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 through 6.7.2.a.8. (2) The written report (and. to the extent possible, the preliminary telephone report or report by similar conveyance) shall describe, analyze, and evaluate safety implications. and outline the corrective measures taken or planned to prevent re~occurrence of the event.
- 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 or previously 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 the Safety Analysis Report (SAR); (3) A personnel change involving the positions of UCD/MNRC Director or UCO Vice Chancellor 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 or development 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 required information may be contained in single or multiple records, or a combination thereof. Records and logs shall 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 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 shipped 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.
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. . . - .
a . a.' I. Formal Licensing Channel Administrative Reporting Channel
............. Communications Channel UCD/MNRC ORGANIZATION FOR LICENSING AND OPERATION FIGURE 6.1}}