ML100550816

University of Utah - Copy of Proposed TS: Final Form of Chapter 6.0 Titled Administrative Controls

ML100550816
Person / Time
Site:	University of Utah
Issue date:	02/11/2010
From:	Univ of Utah
To:	Office of Nuclear Reactor Regulation
References
Download: ML100550816 (18)

Text

50 S Central Campus Dr. Room MEB 2298 Salt Lake City, UT 84112 Phone: 801.587.9696 Web site: http://www.nuclear.utah.edu/nep.html February 11, 2010 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Mail stop: 12D03 Washington, DC 20555-0001

REFERENCE:

Docket 50-407 The University of Utah -TRIGA Research Reactor (UUTR)

License No. R-126 SUBJECT 2): Copy of proposed TS: final form of Chapter 6.0 titled ADMINISTRATIVE CONTROLS The copy of the proposed TS includes the copy of TS Chapter 14. Only the Chapter 6.0 is provided. The rest of the TS document is not changed.

Chapter 14 Technical Specifications Note: changes from the TS submitted in 2005 are in CHAPTER 6 [Fig.6.1 is clarified; whole Chapter is rearranged to follow the ANSI/ANS 15.1-2007 and NUREG 1537 Part 1 & 2]

i I

FACILITY LICENSE R-126 TECHNICAL SPECIFICATIONS AND BASES FOR THE UNIVERSITY OF UTAH TRIGA REACTOR DOClK-T50-407 Amendment No. 7 ii

TECHNICAL SPECIFICATIONS AND BASES FOR THE UNIVERSITY OF UTAH TRIGA NUCLEAR REACTOR This document constitutes the Technical Specifications for the Facility License No. R-126 and supersedes all prior Technical Specifications. Included in these Technical Specifications are the "Basis" to support information the selection and significance of the specification. The bases are included for information purposes only. They are not part of the technical Specifications, and they do not constitute limitations or requirements to which the licensee must adhere.

Furthermore, the dimensions, measurements and other numeric values given in these specifications may differ slightly from actual values because of normal construction and manufacturing tolerances, or normal degree of accuracy of the instrumentation.

Amendment No. 5

TABLE OF CONTENTS 1.0 DEFINITIO NS ..................................................................................... 2 1.1 Reactor Operating Conditions ................................................................................... 2 1.2 Reactor Experiments and Irradiations ........................................................................ 3 1.3 Reactor Components ................................................................................................. 4 1.4 Reactor Instrumentation ............................................................................................. 4 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS ............ 6 2.1 Safety Limit - Fuel Element Temperature .................................................................... 6 2.2 Limiting Safety System Settings ................................................................................. 6 3.0 LIMITING CONDITIONS OF OPERATION ........................................... 9 3.1 Normal Operation ..................................................................................................... 9 3.2 Reactivity limitations ................................................................................................. 9 3.3 Control and Safety System ........................................................................................ 10 3.4 Argon-41 Discharge Limit ........................................................................................ 13 3.5 Engineered Safety Feature - Ventilation System ........................................................ 13 3.6 Limitations on Experiments ...................................................................................... 14 3.7 As Low As Reasonable Achievable (ALARA) Radioactive Effluent Releases ................... 14 3.8 Primary Coolant Conditions ...................................................................................... 15 4.0 SURVEILLANCE REQUIREMENTS ..................................................... 16 4.1 General ....................................................................................................................... 16 4.2 Safety Limit - Fuel Element Temperature .................................................................. 16 4.3 Limiting Conditions for Operation ............................................................................. 17 4.4 Reactor Fuel Elements ............................................................................................. 19 4.5 Primary Coolant Conditions ...................................................................................... 20 5.0 D ESIG N FEATURES ........................................................................... 21 5.1 Reactor Fuel ................................................................................................................ 21 5.2 Reactor Core ............................................................................................................... 21 5.3 Control Elements ...................................................................................................... 22 5.4 Radiation Monitoring System ................................................................................... 22 5.5 Fuel storage ................................................................................................................ 23 5.6 Reactor Building and Ventilation System .................................................................... 23 5.7 Reactor Pool Water Systems .................................................................................... 24 6.0 ADMINISTRATIVE CONTROL ........................................................... 26 6.1 Organization ................................................................................................................ 26 6.1.1 Structure ................................................................................................................. 26 6.1.2 Responsibility ....................................................................................................... 28 6.1.3 Staffing ................................................................................................................... 28 6.1.4 Selection and Training of Personnel ....................................................................... 29 6.2 Reactor Safety Committee ........................................................................................ 29 6.2.1 Function .................................................................................................................. 29 6.2.2 Composition and Qualifications ........................................................................... 29 6.2.3 Operation ................................................................................................................ 30 6.2.4 Review Function ................................................................................................... 30 6.2.5 Audits ...................................................................................................................... 31 6.2.6 Records ................................................................................................................... 31 6.3 Radiation Safety ...................................................................................................... 31 6.4 Operating Procedures ............................................................................................... 32 Amendment No. 5 i

6.5 Experiments Review and Approval ............................................................................ 32 6.6 Required Actions ...................................................................................................... 32 6.6.1 Action To Be Taken in the Event a Safety Limit is Exceeded ................................... 32 6.7 Reports ....................................................................................................................... 33 6.8 Facility Operating Records ........................................................................................ 36 6.9 Quality Assurance ................................................................................................... 37 Amendment No. 5 ii

6.0 ADMINiSTRAT CONTROLS:*

6.1 Organization 6.1.1 Structure

1) The UUTR is an integral part of the Nuclear Engineering Facilities of the University of Utah. The organization of the facility management and operation shall be as shown in Figure 6.1. The responsibilities and authority of each member of the operating staff shall be defined in writing.

2) As indicated in Fig. 6.1, the Reactor Safety Committee shall report to Level 1.

Radiation safety personnel shall report to Level 2. Additional description of levels follows:

Level 1: Individual responsible for the reactor facility's licenses, i.e., the Associate Vice President for Research in the Office of Vice President for Research; The Vice President for Research will assign which of the Associate Vice Presidents for Research will be the responsible Level 1 individual.

a.

b. Level 2. Individual responsible for reactor facility operation, i.e., the facility director shall be the Director of the Utah Nuclear Engineering Program, who shall also be the Director of the Nuclear Engineering Facilities of the University of Utah; Level$. Individual responsible for day-to-day operation or shift shall be the reactor supervisor (RS). This person shall be a senior reactor operator (SRO).

c.

d. Level 4: Operating staff shall be senior reactor operators, reactor operators, and trainees.

3) Transition plan shall be defined as follows:

a. In the absence of the Level 2 individual [sabbatical leave, sudden leave]:

Level 1 shall have the authority to automatically appoint the RS as an interim reactor administrator (RA) with direct access to Level 1 and communication to U.S.N.R.C; Radiation safety personnel shall report to RA unless Level 1 decides differently;

b. In the absence of the Level 3 individual [sabbatical leave, sudden leave]:

,Level 2 shall have the authority to automatically appoint a new RS who shall be a most senior Level 4 individual (SRO).

Amendment No. 7 26

"LEVELI Thu University of Lhah President j

((YELl*

The University of,-Lkth _ _ _ _ _ _ _ _ _

Vice President for

-Reseaohruh NRC LEVEL 71

,-Iu-,turut -3o Ratr Safety Do, Npleear EnginmrIn1 Committee Radiation Safety - - LEVEL ,

Committee , ServIso r t 4- - > Line of communication LEVEL 4 - Line of responsibility Lp nd reporting erfigS.j Figure 6.1 University of Utah Administrative Organization for Nuclear Reactor Operations [Clarified - December 2009; in compliance with ANSI/ANS 15.1-2007]

Amendment No. 7 27

6.1.2 Responsibility Responsibility for the safe operation of the reactor facility shall be with the chain of command established in Fig. 6.1 following ANSI/ANS 15.1-2007:

1) Individuals at the various management levels, in addition to having responsibility for the policies and operation of the reactor facility, shall be responsible for safeguarding the public and facility personnel from undue radiation exposures and for adhering to all requirements of the operating license or charter and technical specifications.

2) In all instances, responsibilities of one level may be assumed by designated alternates or by higher levels, conditional upon appropriate qualifications.

3) The reactor facility shall be under the direct control of a licensed Senior Reactor Operator (SRO) designated by the Reactor Supervisor (RS) who is also a licensed Senior Reactor Operator. The SRO shall be responsible to the RS for the overall facility operation including the safe operation and maintenance of the facility and associated equipment. The SRO shall also be responsible for ensuring that all operations are conducted in a safe manner and within the limits prescribed by the facility license, federal and state regulations, and requirements of the Reactor Safety Committee.

6.1.3 Staffing

1) When the reactor is not secured, the minimum staff shall consist of:

a. A licensed reactor operator in the control room, e.g., Reactor Operator (RO)

(may be the SRO or RS).

b. A designated Senior Reactor Operator (SRO) on call but not necessarily on site.

c. Another person present at the facility complex who is able to carry out prescribed written instructions.

2) A list of reactor facility personnel by name and telephone number shall be readily available in the control room for use by the operator. The list shall include:

a. management personnel,

b. radiation safety personnel,

c. other operations personnel;

3) Events requiring the presence at the facility of the senior reactor operator are:

a. initial startup and approach to power, Amendment No. 7 28

b. all fuel or control-rod relocations within the reactor core region,

c. relocation of any experiment with reactivity worth greater than one dollar;

d. recovery from unplanned or unscheduled shutdown or significant power reduction.

6.1.4 Selection and Training of Personnel The selection, training, and requalification of operations personnel shall meet or exceed the requirements of American National Standard "Selection and Training of Personnel for Research Reactors," ANSIOANS-15.4-1988 ,-R1999, Sections 4 through 7.

The Reactor Supervisor shall be responsible for the facility's Requalification Training Program and Operator Training Program.

6.2 Reactor Safety Committee 6.2.1 Function The Reactor Safety Committee (RSC) shall function to provide an independent review and audit of the facility's activities including:

1) reactor operations

2) radiological safety

3) general safety

4) testing and experiments

5) licensing and reports

6) quality assurance 6.2.2 Composition and Qualifications The RSC shall be composed of at least five members knowledgeable in fields that relate to nuclear reactor safety. The members shall collectively represent a broad spectrum of expertise in the appropriate reactor technology.

Amendment No. 7 29

The members of the committee shall include the Reactor Supervisor and faculty and staff members designated to serve on the committee. The University's Radiation Safety Officer shall be an ex officio member of the RSC.

Members and alternates shall be appointed by and report to Level 1 management.

Individuals may be either from within or outside the operating organization. Qualified and approved alternates may serve in the absence of regular members.

6.2.3 Operation The Reactor Safety Committee shall operate in accordance with a written charter, including provisions for:

1) meeting frequency: not less than once per calendar year and more frequently as circumstances warrant, consistent with effective monitoring of facility activities

2) voting rules and quorums: chairman or his designate and two members, e.g., not less than one-half of the voting membership where the operating staff (i.e., the Level 2 facility director and anyone who reports to that person) does not constitute a majority

3) use of subcommittees

4) review, approval, and dissemination of minutes in timely manner 6.2.4 Review Function The responsibilities of the RSC or designed subcommittee(s) thereof shall include, but is not limited to the following:

1) review and approval of all new experiments utilizing the reactor facility

2) review and approval of all proposed changes to the facility license by amendment, and to the Technical Specifications

3) review of the operation and operational records of the facility

4) review of significant operating abnormalities or deviations from normal and expected performance of facility equipment that affect nuclear safety

5) review of approval of all determinations of whether a proposed change, test, or experiment would constitute a change in the Technical Specifications or on unreviewed safety questions as defined by 10 CFR 50.59

6) review of reportable occurrences and the reports filed with the Commissions for said occurrences Amendment No. 7 30

7) review and approval of all standard operating procedures and changes thereto

8) biennial review of all standard procedures, the facility emergency plan, and the facility security plan.

A written report or minutes of the findings and recommendations of the review group shall be submitted to Level 1 and the review and audit group members in a timely manner after the review has been completed.

6.2.5 Audits The RSC or a subcommittee thereof shall audit reactor operations semiannually, but at intervals not to exceed 8 months. The semiannual audit shall include at least the following:

1) review of the reactor operating records

2) inspection of the reactor operating areas

3) review of unusual or abnormal events

4) radiation exposures at the facility and adjacent environs 6.2.6 Records The activities of the RSC shall be documented by the committee and the RSC shall maintain a file of the minutes of all meetings.

6.3 Radiation Safety The Radiation Safety Office of the University of Utah shall be assigned responsibility for implementing the radiation protection program at the reactor using the guidelines of American National Standard "Radiation Protection at Research Reactor Facilities," ANSI/ANS-15.11-1993 (R2004). The Radiation Safety Officer shall report to Level 2 [Fig. 6.1].

Amendment No. 7 31

6.4 Operating Procedures Written operating procedures shall be adequate to ensure the safety of operation of the reactor, but shall not preclude the use of independent judgment and action should the situation require such. Operating procedures shall be in effect for the following items:

1) performing irradiations and experiments 1

2) startup, operation, and shutdown of the reactor

3) emergency situations including provisions for building evacuation, earthquake, radiation emergencies, fire or explosion, personal injury, civil disorder, and bomb threat

4) core changes and fuel movement

5) control element removal and replacement

6) performing preventive maintenance and calibration tests on the reactor and associated equipment

7) power equipment Substantive changes to the above procedures shall be made only with the approval of the Level 2. Minor modifications to the original procedures that do not change their original intent may be made by Level 3 or higher, but the modifications must be approved by Level 2.

Temporary changes to the procedures that do not change their original intent may be made by a licensed SRO. All such changes shall be documented and subsequently reviewed by the Reactor Safety Committee. Such deviations shall be documented and reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the next working day to the Level 2.

6.5 Experiments Review and Approval Approved experiments shall be carried out in accordance with established and approved procedures as described in 6.4.

6.6 Required Actions 6.6.1 Action To Be Taken in the Event a Safety Limit is Exceeded In the event a safety limit is exceeded:

Amendment No. 7 32

1) The reactor shall be shut down and reactor operation shall not be resumed until authorized by the U.S. Nuclear Regulatory Commission (U.S.N.R.C.).

2) An immediate report of the occurrence shall be made to Level 2 and to the Chairman of the Reactor Safety Committee, and reports shall be made to the U.S.N.R.C in accordance with Section 6.7 of these specifications.

3) A report shall be prepared that shall include an analysis of the causes and extent of possible resultant damage, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence. This report shall be submitted to the Reactor Safety Committee for review and then submitted to the NRC when authorization is sought to resume operation of the reactor.

4) A report shall be made~to the U.S.N.R.C. in accordance with Section 6.7 of these specifications.

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

1) A report within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone to the Project Manager U.S.N.R.C., of

a. any accidental release of radioactivity above permissible limits in unrestricted areas whether or not the release resulted in property damage, personal injury, or exposure;

b. any violation of the safety limit;

c. any reportable occurrence as defined in Section 1.1, "Reportable Occurrence," of these specifications.

2) A report within 10 days in writing to the Document Control Center, U.S.N.R.C.,

Washington, D.C., with a copy to the U.S.N.R.C. Operations, of

a. any accidental release or radioactivity above permissible limits in unrestricted areas whether or not the release resulted in property damage, personal injury, or exposure. The written report (and, to the extent possible, the pre-liminary telephone or telegraph report) shall describe, analyze, and evaluate safety implications, and outline the corrective measures taken or planned to prevent recurrence of the event.

b. any violation of safety limit,

c. any reportable occurrence as defined in Section 1.1, "Reportable Occurrence," of these specifications.

Amendment No. 7 33

3) A report within 30 days in writing to the Document Control Center, U.S.N.R.C.,

Washington, D.C., with a copy to the U.S.N.R.C. Operations, of

a. 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,

b. any significant change in the transient or accident analysis is described in the Safety Analysis Report,

c. any significant changes in facility organization,

d. any observed inadequacies in the implementation of administrative or procedural controls.

4) A report within 60 days after completion of startup testing of the reactor (in writing to the Director, Office of Nuclear Reactor Regulation, USNRC. Washington, D.C.

20555) upon receipt of a new facility license or an amendment to the license authorizing an increase in reactor power level describing the measured values of the operating conditions including:

a. an evaluation of facility performance to date in comparison with design predictions and specifications,

b. a reassessment of the safety analysis submitted with the license application in light of measured operating characteristics when such measurements indicate that there may be substantial variance from prior analysis.

5) An annual report within 60 days following the 3 0 th of June of each year (in writing) to the Document Control Center, U.S.N.R.C., Washington, D.C. with a copy to the U.S.N.R.C. Operations, providing the following information:

a. a brief narrative summary of (i) operating experience (including experiments performed), (ii) changes in facility design, performance characteristics, and operating procedures related to reactor safety and occurring during the reporting period, and (iii) results of surveillance tests and inspections:

b. tabulation of the energy output (in Megawatt-days) of the reactor, hours reactor was critical and the cumulative total energy output since initial criticality;

c. the number of emergency shutdowns and inadvertent scrams, including reasons for them;

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 10 CFR 50.59; Amendment No. 7 314

f. a summary of the nature, and amount of radioactive effluents released or discharged to the environs beyond the effective control of the licensee as measured at or before the point of such release or discharge:

Liquid Waste (summarized on a monthly basis)

L radioactivity discharged during the reporting period

" total estimated quantity of radioactivity released (in curies) an estimation of the specific activity for each detectable radionuclide present if the specific activity of the released material after dilution is greater than 1 x 10-7 gCi/cc.

" summary of the total release (in curies) of each nuclide determined just above for the reporting period based on representative isotopic analysis,

" estimated average concentration of the released radioactive material at the point of release for the reporting period in terms of gCi/cc and fraction of the applicable DAC or Effluent concentration ii. total volume (in gallons) of effluent water (including dilutant) released during each period of release.

Gaseous Waste (summarized on a monthly basis)

L. radioactivity discharged during the reporting period (in curies)

" total estimated quantity of radioactivity released (in curies) determined by an appropriate sampling and counting method,

" total estimated quantity of argon-41 released (in curies) during the reporting period based on data from an appropriate monitoring system.

" estimated average atmospheric diluted concentration of argon-41 released during the reporting period in terms of jiCi/cc and fraction of the applicable DAC value.

" total estimated quantity of radioactivity in particulate form with half-lives greater than 8 days (in curies) released during the reporting period as determined by an appropriate particulate monitoring system.

" average concentration of radioactive particulates with half-lives greater than 8 days released in pCi/cc during the reporting period, and

" an estimate of the average concentration of other significant radionuclides present in the gaseous waste discharge in terms of pCi/cc and fraction of the applicable DAC-value for the reporting period if the estimated release is greater than 20% of the applicable DAC.

Amendment No. 7 35

Solid Waste (summarized on an annual basis)

i. total amount of solid waste packaged (in cubic feet),

ii. total activity in solid waste (in curies),

iii. the dates of shipment and disposition (if shipped off site).

g. An annual summary of the radiation exposure received by facility personnel and visitors in terms of the average radiation exposure per individual and greater exposure per individual and greater exposure per individual in the two groups. Each significant exposure in excess of the limits of 10 CFR 20 should be reported, including the time and data of the exposure as well as the circumstances that led up to the exposure.

h. An annual summary of the radiation levels of contamination observed during routine surveys performed at the facility in terms of the average and highest levels.

j. An annual summary of any environmental surveys performed outside the facility.

6.8 Facility Operating Records In addition to the requirements of applicable regulations, and in no way substituting for those requirements, records and logs shall be prepared for at least the following items and retained for a period of at least 5 years for items 1) through 6) and indefinitely for items 7) through 11).

1) normal reactor operation

2) principal maintenance activities

3) abnormal occurrences

4) equipment and component surveillance activities required by the Technical Specifications

5) experiments performed with the reactor

6) gaseous and liquid radioactive effluents released to the environs

7) offsite inventories and transfers

8) fuel inventories and transfers

9) facility radiation and contamination surveys Amendment No. 7 36

10) radiation exposures for all personnel

11) updated, corrected, and as-built drawings of the facility 6.9 Quality Assurance In accordance with Regulatory Guide 2.5 and ANSI 402, "Quality Assurance Program Requirements for Research Reactors," Section 2.17, the facility shall not be required to prepare quality assurance documentation for the as-built facility. Quality assurance (QA) requirements will still be limited to those specified in Section 2.17 as follows:

"All replacements, modifications, and changes to systems having a safety related function shall be subjected to a QA review. Insofar as possible, the replacement, modification, or change shall be documented as meeting the requirements of the original system or component and have equal or better performance or reliability."

"The required audit function shall be performed as specified in Section 6.2.5."

Amendment No. 7 37