ML20196C082

Annual Progress Rept of Univ of Florida Training Reactor, Sept 1987 - Aug 1988

ML20196C082
Person / Time
Site:	05000083
Issue date:	08/31/1988
From:	Vernetson W, Whaley P FLORIDA, UNIV. OF, GAINESVILLE, FL
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
ORO-4014-18, NUDOCS 8812070170
Download: ML20196C082 (215)

Text

{{#Wiki_filter:_ - - - _ Contract # DE-AC05-76ER04014 Report # ORO-4014-18 ANNUAL PROGRESS REPORT OF THE UNIVERSITY OF FLORIDA TRAINING REACTOR September 1,1987 - August 31,1988 By Dr. William G. Vernetson, Director Mr. Paul M. Whaley, Acting Reactor Manager 7

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L NUCLEAR FACILITIES DIVISION DEPARTMENT OF NUCLEAR ENGINEERING SCIENCES College of Engineering University of Florida -f\ 1s Gainesville . 3p' g70 k O b !183 '

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s Contract #DE-AC05-76ER04014 Report #ORO.-4014-18 ANNUAL PROGRESS REPORT OF THE UNIVERSITY OF FLORIDA TRAINING REACTOR September 1,1987 - August 31,1988 Submitted By Dr. William G. Vernetson Director of Nuclear Facilities November,1988 Department of Nuclear Engineering Sciences University of Florida Gainesville, Florida 1 I __ _ ..

Contrcet #DE-AC05 76ER04014 Report #ORO -4014-18 f ANNUAL PROGRESS REPORT OF THE UNIVERSITY OF FLORIDA TRAINING REACTOR September 1,1987 - August 31,1988 i l Submitted By Dr. William G. Vernetson Director of Nuclear Facilities t l November,1988 4 l l Department of Nuclear Engineering Sciences University of Florida Gainesville, Florida

TABLE OF CONTENTS Eagg Number I. INTRODUCTION I-1 II. UNIVERSITY OF FLORIDA PERSONNEL ASSOCIATED WITH THE REACTOR II-1 III. FACILITY OPERATION III-1 IV. h10DIFICATIONS TO THE OPERATING CHARACTERIS-TICS OR CAPABILITIES OF THE UFTR FACILITY IV-1 V. SIGNIFICANT hiAINTENANCE, TESTS AND SURVEIL-LANCES OF UFTR REACTOR SYSTEhtS AND FACILITIES V-1 VI. CHANGES TO TECHNICAL SPECIFICATIONS, STANDARD OPERATING PROCEDURES AND OTHER DOCUh1ENTS VI-1 VII. RADIOACTIVE RELEASES AND ENVIRONhiENTAL SURVEILLANCE VII-1 VIII. EDUCATION, RESEARCH AND TRAINING UTILIZATION VIII 1 IX. TIIESES, PUBLICATIONS, REPORTS AND ORAL PRESENTATIONS OF WORK RELATED TO THE USE AND OPERATION OF TIIE UFTR IX-1 APPENDIX A: NOTICE OF VIOLATION FROh! NRC NRC INSPECTION REPORT NUh1BER 50 83/88 01 WITH UFTR FACILITY LICENSEE RESPONSE APPENDIX B: FINAL REPORT TO NRC ON INTER-hil'ITENT DOWNSCALE FAILURE OF SAFETY CHANNEL I INDICATION APPENDIX C: UFTR TECilNICAL SPECIFICATIONS APPROVED AhiENDhiENT 17 PAGES WIT) < NRC SAFETY EVALUATION REPORT APPENDIX D: UFTR SAFETY ANALYSIS REPORT REVISION 4 DOCUhtENTATION APPENDIX E: UETR SAFETY ANALYSIS REPORT REVISION 5 DOCUh!ENTATION i

TABLE OF CONTENTS (CONTINUED) APPENDIX F: UFTR STANDARD OPERATING PROCEDURES ORIGINAIE AND MAJOR REVISIONS FOR 1987-1988 REPORTING YEAR:

1. UFFR SOP-F.8, "UFTR SAFEGUARDS RE-PORTING REQUIPEMENTS"(REV 0)

APPENDIX G. DOCUMENTATION FOR QUALITY ASSURANCE PROGRAM APPROVAL FOR RADIOACTIVE MA-TERIAL PACKAGES NO. 0578, REVISION 1 APPENDIX II. CORRECTION PAGE FOR Tile 1985 1986 ANNUAL REPORT I I i

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e ,_ _ c--.-------------._ _ _ _ . .. i I. INTRODUCTION l L1 Overall Utilization i The University of Florida Training Reactor's overall utilization for the past reporting 1 year (September,1987 through August,1988) cor.tinued to be at high levels of quality usage characteristic of the last (19861987) reporting year when the 91.5% availability factor was 1 the highest in the last six years and probably in the 28-year history of the facility. Although several significant forced outages in the current year have limited utilization, the overall availability factor has still been maintained at nearly 80%. The UFTR continues to experience a high rate of utilization in a broad spectrum of areas with total utilization continuing near the highest levels recorded in the early 1970's. This broad based utilization has been supported by a variety of usages ranging from ! research and educational utilization by users within the University of Florida as well as by f other researchers and educators around the State of Florida througa the support of the DOE Reactor Sharing Program and several externally supported usages. Significant usage has also been devoted to facility enhancement where a key ingredient for this usage has been a stable management staff. Personnel associated with the UFTR are listed in Section II; facility operations for all usages are delineated in Section III. The yearly total energy generation of 26.68 hiegawatt hours for the 1987 1988 g reporting year represents a 10% decrease over the previous reporting year which is still the , fourth highest one year total energy generation over the last twelve years of UFITt i operation and represents the seventh highest one year value in the 29 year operational history of the UFTR. The decrease in en::rgy generation was primarily due to considerable low power usage for operator training and research on plasma kinetics parameters as well 11

as implementation of the neutron radiography facility. Additional large time and resource commitments were made I,r efforts related to decontamination, movement, inventory and other work with the LEU fuel stored and hoped to be used in the UFTR HEU to LEU conversion. Several extended outages (one to implement corrective and preventive maintenance on all control blade drive motor reduction gear assemblics to restore free motion and two others to evaluate and correct the intermittent loss of indication downscale on Safety Channel #1) also caused lost facility usage and hence negatively affected energy generation. The total run time for the facility was maintained somewhat above the previous year at 568.35 hours for this reporting year indicating considerable low power run time for neutron radiography and the plasma kinetics experiments as well as UFTR operator training. With the addition of one new Senior Reactor Operator (SRO) at the end of this year and another RO expected to be licensed early in the next reporting year, the availability of operating personnel should be improved. Overall, the indication is toward considerable low power usage and continued high utilization of the reactor subject to availability of the reactor and licensed operators. Analysis of facility utilization shows that the sustained usage and energy generation relative to the previous year are attributable to the same supportive conditions as in the last year. As noted for the last four years, the refurbishment of the Neutron Activation Analysis Laboratory has impacted favorably on all areas of utilization from research projects using

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Neutron Activation Analysis (NAA) to training and educational uses for students at all ) levels. With successful implementation of an improved remote sample handling "rabbit" facility, efforts to advertise availability and encourage usage of the UITR (especially for 3 research) have proceeded in a favorable light though always less quickly than hoped over the last four years. Implementation of the standard rabbit capsule size with larger carrying I2 L

i capacity during the 1986-1987 reporting year has further supported use of the facility. The additionalimplementation of two state-of-the art PC-based spectrum analysis systems with complete ORTEC software packages for spectrum analysis and data reduction has been a key factor supporting reactor utilization during the last two reporting years for education f and training uses as well as research projects, several of which constitute ongoing but promising seed projects to support proposals. Indeed, the 19871988 reporting year is the i first full year for availability of the PC based ORTEC analyzers with standardized rabbit system capsule size. The NA A Laboratory has also been outfitted with its own independent sample and standards drying facility during the 1937-1988 reporting year.The result of these various improvements has been an easier and faster turnaround of sainples submitted to be t irradiated for Neutron Activation Analysis with a resultant increase in interest. The implementation of these facilities has given the UFTR management the capability to promote it among University of Florida users and among researchers at other universi ties 1 and colleges around the State of Florida. As the availability of this high technology facility I becomes better advertised through its users, its usap continued to increase. 1 De primary catalyst for maintaining facility usage continues to be the Department of Energy's (DOE) Reactor Sharing Program. His reporting year was the fifth consecutive-year in which the UFTR was supported as part of DOE's Reactor Sharing Program. His program is designed to increase the availability of University reactor facilities such as the UFTR for non reactor owning educational (user) institutions ranging from high schools to } colleges and universities. Basically, this grant provides funds against which reactor operating - costs may be charged when the facilities are utilized by regionally affiliated user institutions 3 for student instruction / training or for student or faculty research that is not supported by outside funding. In all, sixteen (16) different academic institutions ranging from high schools 13

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to universities around the State of Florida made use of this program to utilize the UFTR for research (primarily via neutron activation analysis to determine trace element t compositions), for reactor facility demonstrations, experiments and course work related to various aspects of operation and for training of students in various community college i programs such as nuclear medicine technology and radiation protection technology and for research and training p'rograms for high school students for which several science fair projects are stillin progress. At years end, several unsupported research projects were still awaiting availability of the UFTR under the Reactor Sharing Program as UFTR usage attributable to this DOE sponsored program continues to grow. Despite considerab!c cost-sharing by the University of Florida, all of the reactor sharing funds allocated by the i Department of Energy for this supporting year were fully utilized. This program has been renewed at a 12% increased funding level for the upcoming year, so further expansion of this usage is possible and expected. Reactor use by University of Florida courses and laboratories continues at the substantial level establish.d in the last several years. Course and Department usages within the University range from the Environmental Engineering Sciences Department in its Health Physics courses to the Chemistry Department in a graduate level radiochemistry laboratory course. Of course, the biggest single user department remains the Nuclear Engineering Sciences Department which uses the reactor facility for both r;raduate and undergraduate laboratories, research projects and class demonstrations, la addition, plasma ), I kinetics research has expanded considerably as part of the nuclest space power research , program in the Nuclear Engineering Sciences Department. External users for courses ) include Central Florida Community College for its radiation protection technology courses as well as Santa Fe and liillsborough Community Colleges for their nuclear medicine 14 I - _

Clj technology courses. -? With many continuing usages already scheduled along with the state-of-the art analysis iristrumentation and support equipment in the NAA Laboratory, plus renewal of the Reactor Sharing Program support, facility utilization and energy generation for the upcoming year should be considerably augmented. The latter augmentation is particularly possible because the UFTR utilization under the DOE Reactor Shap.:g Program has spread publicity on the availability of the UFTR so that a number of investigators on the University of Florida campus and elsewhere around the state have again indicated an interest in using the reactor facility and its experimental systems during the upecming year. Several other state wide users are in the process of preparing proposals hopefully to provide funded usage of the UFTR within the next year. The large usages for the University of South Florida (both Tampa and St. Petersburg campuses), as well as for two groups at Florida State University and one at the University of Central Florida, are primarily to ' demonstrate capabilities to support proposals seeking external support as an outgrowth of the DOE Reactor Sharing Program support. Therefore, expectations of continued growth of reactor facility usage dependent on a continued upgrading of facility capabilities and staff 1 eapertise are quite realistic p I.2 Facility Improveme,tt M vy For facility enhano . n . neutron radiography facility (available during the entire a year for the first time but nearing optimization during the latter part of the year) provided 4 i a strong base for continued growth and diversification of usage during this year and should l continue to do so during the upcom!:,, year as the facility is further optimized to attract l more users, several of whom have expressed interest in demonstrations of radiography

l { rystem parameters. One possible user is interested in research on layered materials. Finally, plads are in progress to investigate the possibility of installing a prompt gamma analysis facility at the UFTR to complement the NAA 12b capabilities. This a a multiyear

 !' er.hancement project; if feasible it will require some designated suppurt.

Another area of enhancement receiving considerable attention this year was a series of measurements to characterize all experimental facility irradiation parano :s from b neutron flux and spectrum characteristics and gamma d,se levels and spectrum characteris-tics to ratios of neutron and gamma field parameters. It is hoped that a masters' level student will be able to bring this program to fruition during the upcoming year, though data to date is sufficient to support continued plasma kinetics research for the space power reactor program at the University of Florida and for research on radiation effects on dielectric materials for a researcher at Florida State University. For staff enhancement, the facility upper-level management is well set with a permanent full time ac ting reactor manager functioning effectively. Management staffing conditions are gene: ally supportive of the considerable broad based increases in facility usage for education and training of students as well as research by faculty at the University of Florida and other schools. Nevertheless, all other staff personnel are part time employees, two of whom previously wera full time employees and one of whom has been effectively lost as a part time SRO due to poor health. Although such employees provide a good experience base for operations, the lack of other licensed staff members during the , , [

                                                                                                         . 'l current reporting year has necessi;sted limitations in the growth of some usage programs, i

I It is expected that these limitations will be less restrictive during the upcoming reporting l year with one new part time SRO licensed at year's end and a fully certi'icJ RG aheduled early in the new year.The resultant ren. oval of the need for spemi Jccations of ' raining I-6 1

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1 l f i time in the classroom or the control room will .nhance facility education and research usages. 13 Administrative Commitment of Resources The level of administrative work dedicated to regulatory activities was considerable \ duiing the year and is expected to be at a similar level during this next reporting year as license related administrative activities continue to involve large commitments of personnel r resources. Although the facility was clted for no violations or deviations following the biennial NRC Operations Inspection conducted October 19 22, 1937, it was cited for two Level IV violations following the NRC Health Physics Inspectica conducted March 14-17, 1988. One violapon was for failure to conduct adequate surveys to evt.luato the extent of radiation hazards in liq id and gaseous effluer.ts released from the facility, and the other was for failure to have the Director cf Nuclear Facilities approve the Radiation Control Techniques used to conduct environmental surveillances and effluent measurements required by Technical Specifications. The items in these violatir;ns were primarily administrative and technical analysis problems; no actual safety problems were noted. Faci!!ty responses to the violations and full compliance were all cmapleted following the inspection 'y hy 31,1985 and occupied significant facility management and staff time during the reporting year. The notice of violation clong with the licensee response is

                                                                                                                . I wntained in Appendix A. Additional commitments to perform a complete documented                a J

evaluation of the Argon 41 measurement methodology and to evaluate all UFTR radiation , protection surveil'ances relative to instrumentation capabihties and needs will involve considerable additional time commitments. The net result is that administration efforts directed at compliance with NRC requirements will continue tc involve considerable ) l l7

( commitments of time and resources during the next year. It should be noted that considerable facility management effort will need to be devoted to preparing the license amendment package for the HEU to LEU conversion during the upcoming year so administrative efforts will not be reduced. In general, none of the NRC findings involved any actual safety problems. Sim!!arly, two inspections by representatives of the American Nuclear Insurers resulted in only minor recommendations. As indicated, the UITR continues to operate with an outstanding safety record. No uncontrolled releases of radioactivity have occurred from the facility and controlled releases remain well within established limits. The personnel radiation exposures for 19871988 have been maintained near the usual low yearly level since there was no extensive dose commitment for maintenance in the core or other high dose rate areas as for the control blade drive system project of the 1985-1986 reporting year. There was also no waste or special nuclear ma'erial shipped this year; however, waste is expected to be shipped in the upcoming reporting year to prepare the facility for the llEU-to LEU fuel conversion activities to commence within the next two years. With the correct!ve action implemented following the NRC Ilealth Physics inspection in February,1987, the upco ning waste ship.nent is assured to be properly controlled and documented. Environmental 4 radioactivity surveillances continue to show no detectable off site dose attributable to the UFTR facility as also noted in Section VII as the facility continues to operate within ALARA guidelines with minimal exposure of staff and visitors. *., J Other administrative activities have also involved large commitments of time and , resources during the year. First, the USNRC response to Amendment 17 to the UFITI Technical Specifications submitted originally in the previous reporting year was received on February 8,1988 requesting clarificatia for times when the reactor vent system could be I-8 1

secured with the stack count rate above 10 cps and the addition of provisions for controlled release of radioactive effluents to the environment during abnormal operat'ng conditions. The requested clarification and addition were included in the resubmission of Amendment 17 on March 7,1988. The approved Tech Spec Amendment 17 was finally received on May 3,1988. The required core vent sampling system was installed on May 4,1988 and was available for all subsequent operation, The revision permitting certain activities to be conducted when the cadiur is shutdown, the vent system secured and the stack monitor reading above 10 eps has not yet been incorporated into Standard Operating Procedures, but the work is in progress.The enti e amendment package including new Tech Spec pages (photostats of those submitted) and the NRC Safety Evaluation Report supporting the amendment is contained in Appendix C. Second, Revision 4 to the UFTR Safety Analysis Report (SAR) was submitted to update descriptions of the UITR Fire Protection System and Communications Systems. A 1 l c implete copy of the entire submittal for UITR SAR Revision 4 is contained in Appendix D. In addition, Revision 5 of the UITR SAR was submitted as part of ongoing reviews to assure the document remains updated. Revision 5 corrects a number of typographical errors and updates some reactor descriptive oata and parameters as well as the description of the instrumentation operation i.. the UFTR console. A complete copy of the entire subm.ittal for Revision 5 is contained in Appendix E. Both Revision 4 and 5 were evaluated not to involve any unreviewed safety questions and were incorporated into official copies of the

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UITR SAR. Review efforts to assure an updated and accurate UITR SAR continue with , a revision of Chapter 11 i preparation independent of any changes needed for the IIEU-

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to LEU conversion submittal. I9

Next, although temporary change notices (TCNs) were issued for thirteen (13)  ; different standard operating procedures, some for multiple TCNs, no revisions of standard operating procedures (SOPS) were issued and only one new procedure was generated during the year. UFTR SOP F.8,'UFTR Safeguards Reporting Requirements" was generated to delineate requirements for reporting safega. ds events with the full text included in Appendix F for reference purposes and t meet Tech Spec requirements for such submissions. Another adr.inistrative effort during the year involved submission of a revised Diennial Reactor Operator Requalification and Recertification Program Plan to reflect new l requirements in 10 CFR 55 for a comprehensive examination once every two years and an i operations tests every year. These changes will be reflected in future issues of the Program Plan as it is resubmitted for each successive two-year training cycle. l l Considerable administrative efforts were also devoted to llEU to LEU Conversion Documents. A new proposal updating the UFTR conversion schedule and work status per 10 CFR 50.64(b)(2) requirements was submitted in March,1988. With receipt of DOE funding to support conversion analysis in November,1987, considerable effort was des ed to clearing a new facility for the SPERT fuel held under the SNM-1050 license. In addition to efforts to decontaminate the new facility, upgrade the fire alarm system, revise the security system and security plan, move the fuel and coriduct a defen sd visual inspection I and inventory of fuel pin serial numbers, considerable effort was dcyoted to the required administrative controls and plans for making the decision on whether to implement UFTR

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llEU to LEU conversion with the SPERT fuel or with standard uranium-silicide plate type fuel. This decision is expected early in the next reporting year after X radiographic inspection of previously identified SPERT fuel pins is used as the basis for the choice of 140

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l l l l conversion options. A final administrative effort was devoted to generating a OA Program suitable to control shipment of SPERT F 1 LEU fuel pins. Complete documentation for NRC OA Program Approval for Radioactive Materials Packages No. 0578, Revision No.1 is contained in Appendix G. The program approval was obtained to be used to ship some of the SPERT fuel to an Oak Ridge National Laboratory reactor facility; however, it will also be useable to ship SPERT fuel to the UFTR if this conversion option is selected. The level of administrative work dedicated to regulatory and licensing activities is expected to remain at a similar level during the next reporting year. Commitments to perform a complete documented evaluation of the Argon-41 measurement methodology and to evaluate all UFTR radiation protection and control surveillance measurements relative l to instrumentation capabilities and requirements will involve considerable administrative effort. He same is true of the continuing effort to update the UFI'R SAR, especially Chapter 11. Of course, considerable facility management effort will be devoted to prepare the license amendment package for IIEU to LEU conversion during the upcoming year. The net result is that administrative efforts directed at compliance with NRC requirements i will not be reduced but will likely be significan:ly increased during the next reporting year. De considerable test, maintenance and surveillance activities required by the facility license Technical Specifications and other controls also contributed significantly to usage and personnel commitments. Details on these surveillance and maintenance usages are } presented in Section V of this report, while any associated mcdifications or evaluations of potential unreviewed safety questions are tabulated in Section IV. This contribution has 3 increased from last year because of several outstanding late maintenance projects for corrective action. The first significant outage was for a failure of Safety 2 control blade to 1 11

withdraw on demand; the second was initiated to address the brief downscale loss and gredual recovery of the Safety Channel #1 indication. The failure of the S 2 control blade to withdraw upon demand was due to hardened , grease deposits and worn bearings in the drive motor gear assembly. Since the other control I blade drive motor gear assemblies were noted to be developing similar problems, corrective and preventive maintenance was performed on all control blade drive motor reduction gear  ; assemblies to restore and assure free removal on demand. No further problems have occurred since September,1987.

          'nic second large maintenance project involved corrective maintenance, replacement of failed components, implementation of a detailed testing and evaluation program and final system checkout to restore t e UFITI to operational status following the brief downscale failure of Safety Channel #1 and the sabsequent recurrence of the event for which a root cause was not definitively identified.

The failure of the Safety Channel #1 monitoring function is considered to be the , most serious of the repo: table occurrences for the year.The complete final report on this 1 failure is contained in Appendix B of this report. Generation of this and several other I reports for UFTR promptly reportable (unuf aal) occurrences occupied cons!derable commitment of management as well as technical resources during the year. l 1.4 Facility Summary Overview ) The rea tor and associated facilities continue to maintain a high in state visibility and - strong industry relationships. With the DOE Reactor Sharing Program to support UFTR- ) t related research by faculty and students at other academic institutiam as well as training for various high school, community college ar.d university programs around the state, the 1-12

reactor facility is also maintaining high in state visibility with other educational institutions. His situation is particularly true among high school science departments where reactor sharing supported usage increased significantly last year with even larger increases in size and diversity of usages expected during the upcoming year. The interactions of several small externally supported research programs as a result of the Reactor Sharing Work is further proof of its effectiveness as is the continued generation of proposals to obtain external funding based on results of research obtained under Reactor Sharing support, ne description of various projects associated with the UlTR is given in Section Vill; the listing of projects has become extensive over the past few years of increased utilization. He same is true of the list of publications and reports associated with the UITR; the listing given in Section IX of this report is the i.ncst extensive list in the last ten (10) years and generally delineates the diversity and quantity of facility usage. With the sustained statewide interest, the facility is being included in several proposals to provide for funded usage of the UITR and the NAA Laboratory. Several such usages occurred during each of the past two reporting years (19861988). De Reactor Sharing Program began in late 1983 and is directly responsible for the generation of a number of these proposals. As more of these proposals are submitted and funded, further increases in UFIR usage can be expected in any case, on campus research and senice usage of the UFFR is also increasing because of the visibility generated sia the Reactor Sharing Program. In general the level of interest in the facility is high though expanded on.  :) campus usage for funded rescarch is a continuing objective. i s Finally, it is hoped that more direct industry training will be accomplished in the j upcoming year. One small usage is tentatively scheduled for early 1989; nevertheless, the lack of utility interest in training programs other than operations usage for SRO l 15

certification makes it ualikely significant growth will occur in this area. With the rabbit system and the associated NAA and neutron radiography facilities plus the increased DOE s Reactor Sharing Program and expectations for increased research funding from other I agencier, further increases in facility usage are realistic and should be significant, especially with a newly licensed part time SRO and a new part time RO expected to be licensed early in the next reporting year. The expectations for the 19881989 year are outstanding. Significant opportunities for expanded education and research usages are appaient. The significant possibilities for continued growth in existing and new program areas are a challenge that is being addressed vigorously.

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II. UNIVERSITY OF FLORIDA PER3ONNEL ASSOCIATED WITil Tit.E REACFOR A. Esrsonnel Emoloyed by the UFTR W.G. Vernetson - Associate Engineer and Director of Nuc'.:ar Facilities (September 1,1987 - August 31,1988) P.M. Whaley - Senior Reactor Operator and Acting Reactor Manager (September 1,1987 August 31,1988) H. Gogun >- - ' Senior Reactor Operator (part time) (September, 1987 August,1988) G.W. Fogle - Reactor Operator (1/4 time) (September,1987 - August,1988) R. Piciullo - Student Reactor Operator Trainee (1/2 time) (September,1987 - July,1988)

                                     -    Senior Reactor Operator (1/2 time) (July,1988 -

August,1988) M. Wachtel - Student Reactor Operator Trainee (1/3 time) (September,1987 August,1988) CJ. Stiehl - Student Reactor Operator Trainee / Technician (part time) (September,1987 - February,1988) J. Godfrey - Student Reactor Operator Trainee (1/2 time)

    -                                     (January,1988 June,1988)

P. Stevens - Secretary Specialist (3/4 time) (September,1987 - l August,1988) l l l B. Endiation control office D.L Munroe' - Radia'. ion Control Officer (September,1987 - August,1988) II.G. Norton - Radiation Control Technician (September,1987 - August,19S8)

   ' ' Die specified alternates for the Radiation Control Officer position are Mr. William Coughlin who works out of the Shands Teaching Ilospital on campus and II. Norton listed below Mr. Munroe as a Radiation Control Technician.

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LP. Nichols - Radiation Control Technician (September,1987 - August,1988) R.N. Hagen - Nuclear Technician (September,1987 - June,1988) R.K. Ilansen - Nuclear Technician (September,1987 - August, 1988) M.W. Wilkerson - Fuclear Technician (hiay,1988 - August,1988) W.G. Wabbersen - Nuclear Technician (August,1988) Basic routine health physics is performed by UFTR staff; however, assistance from the Radiation Control Office is required for operations where a significant dose (Imel I RWP) is expected or possible and where certain experiments are inserted or removed from the reactor ports. 3ese personnel are also required fo. certain operations where j high contamination levels may be expected. They also periodically review routine UFTR i radiation control records and operations and assist in performance of certain radiation l safety and control related surveillances. As a result, many radiation control office persennel are listed and though employed 1/3,1/2 or full time, only a small fraction of their work effort supports UFTR activities. C. Reactor Safety Review Subcommittee (RSRS) ht.J. Ohanian - RSRS Chairman, Associate Dean for Research, College of Engineering and Professor, Nuclear Engineering Sciences Department W.G. Vernetson - hier.iber - Reactor hianager and Director of Nuclear Facilities I J.S. Tulenko - hiember (NES Department Chairman) W.E. Bolch - hiember at large D.L hiunroe - hiem'oer (Radiation Control Officer) D. Line Responsibility for UFTR Administration hi.ht. Criser, Jr. - President, University of Florida W.ll. Chen - Dean, College of Engineering (September 1,1987 - July 31,1988) W.ht. Phillips - Dean, College of Engineering (August 1,1988 - August 31,1988) 11-2

J.S. Tulenko - Chairman, Department of Nuclear Engineering - , Sciences l W.G. Vernetson' - Director of Nuclear Facilities P.M. Whaley - Acti.ig Reactor hianager i E. Line Responsibility for the Radiation Control Office { M.M. Criser, Jr. - President, University of Florida W.E. Elmore - Vice President, Administrative Affairs c W.S. Properzio - Director, Emironmental IIcalth and Safety - - l D.L Munroe - Radiation Control Officer For line responsibility for the Radiation Control Office, all personnel were employed in permanent positions for the full year. 4

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l 2 Dr. W.G. Vernetson continues to serve as Director of Nuclear Facilities and Reactor Manager with Mr. P.M. Whaley sening as full-time Acting Reactor Manager, 11-3 1

l Ill. FACllJTY OPERATION The UFTR continues to experience a high rate of utilization especially when compared to the 1985-1986 reporting year when large outages limited reactor operation, . with total utilization continuing near the highest levels recorded in the early 1970's. This continuation of a high rate of UFTR facility usage has been supported by a variety of usages ranging from research and educational uititration by users within the University of Florida as well as research and educational utilization by researchers and educators around I the State of Florida through the support of the DOE Reactor Sharing Program. Again this year several externally supported usages have also continued to impact reactor utilization and support the continued diversification of facility activities and capabilities. a~ As noted the last four years, the refurbishment of the Neutron Activation Ana!y:!: . biboratory has impacted favorably on all areas of utilization from research projects using NAA to training and educational uses for students at all levels. With successful implementa- )

tion of an improved remote sample handling "rabbit" facility, efforts to advertise availability 1

and encourage usage of the UFTR (especially for research) have proceeded in a favorable light though always less quickly than hoped over the last four years. Implementation of the , 1. standard rabbit capsule size with larger carrying capacity during the 1986 1937 reporting year has further supported use of the facility. The additional implementation of two state- , of-the art PC based spectrum analysis systems with complete ORTEC software packages for spectrum analysis and data reduction has been a key support factor for reactor utilization during the last two reporting years for education and training uses as well as research projects, several of which constitute ongoing but promising seed projects to support . proposals. Indeed, the 19S7-1988 reporting year is the first full year for availability of the 111-1 r

PC-based ORTEC analyzers with standardized rabbit system capsule size. The NAA 12boratory has also been outfitted with its ewn independent sample and standards drying facility during the 19871988 reporting year.The result of these various improvements has been an easier and faster turnaround of samples submitted to be irradiated for Neutron l Activation Analysis. With the continued and increased support of the DOE Reactor Sharing Program, l there has been continued significant usage by a wide variety of users from a broad spectrum of schools for educational as well as research purposes; again, several proposals for separate research funding are in progress. There has also been continued slow growth in reactor usage for both educational and research programs sponsored by the University of Florida e but spurred by Reactor Sbaring users, with the research area showing several relatively large projects at the proposal stage. 'Ite plasma kinetics research has been a particularly active . area and should continue to expand. Finally, there were also several commercial research 9 irradiations and related projects this year; with the computational analysis capabilities for NAA, it is hoped more such irradiations will be forthcoming during this next year to further complement UFFR research and educational uti lization activities whether supported by the

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University of Florida, Reactor Sharing or externally funded sources. The level of administrative work dedicated to regulatory activities is expected to be at a similar level during th!s next reporting year. Although the facility was cited for no violations or deviations following the biennial NRC Operations Inspection conducted October 19 22, 1987; it was cited for two level IV violations following the NRC IIcalth Physics inspec: ion conducted March 1417,1988. One violation was for failure to conduct adequate surveys to evaluate the extent of radiation hazards in liquid and gaseous ef0uents released from the facility and the other was for failure to have the Director of Nuclear ,

                                                                      'IIc2

Facilities approve the Radiation Control Techniques used to conduct environmental { surveillances and effluent measurements required by Technical Specifications. The items in these violations were primarily administrative and technical analysis problems; no actual i safety problem was noted. Facility responses to the violations and full conipliance were all completed following the inspection by July 31, 1988 and occupied significant facility management and sta'f time during the reporting year. Additional commitments to perform a complete document 1d evaluation of the Argon-41 measurement methodology and to evaluate all UFTR radiation protection and control surveillances relative to instrumentation , capabilities and needs wMl involve considerable additional time commitments. It is also expected that considerable facility management effort will be devoted to preparing the license amendment package for the HEU to LEU conversion during the upcoming year. The net result is that administration efforts directed at compliance with NRC requirements will not be reduced but likely will be increased during the next year. Shown in Table 1111 is a summary breakdown of the reactor utilization for this reporting period. The list delineates UITR utilization divided into sixty (60) different educational, research, training, tests, surveillances and facility enhancement operations and general tour /demonntation activities.The total reactor run time was about 568 hours while various experiments and other projects used over 1828 hours of facility time, not counting a large block of time devoted to daily and weekly checkouts. In addition, there were many concurrent usages during the year to optimize utilization of available personnel. The run time represents a significant increase of over 39c from last year despite effectively reduced licensed operating staff and reduced availability due to several medium size outages reducing availability below S0rc. In contrast, the experiment time represents a very large 36% imease without accounting for over 500 hours of concurrent experiment time in a 111-3 l ____ ____________-____ ______-_______

variety of areas. The relatively large increase here is because of large time commitments I for training new operators and for setting up to run the plasma kinetics and neutron radiography experiments as well as better record keeping of project times using the facility or its staff but not the reactor such as nearly 300 hours for project work with the LEU  ; SPERT fuel as well as unloading and transferring of the two Co 60 irradiator sources for the University of Florida Departments of Radiochemistry and Microbiology. Maintenance time on two medium size projects (restoration of free movement to control blade drive motor gear box assemblies and corrective action to address intermittent downscale failure of Safety Channel 1 Indication) contributed considerable time also. The large increase in experiment time along with a small increase in run time are directly attributable to the relatively high reactor availability (79.2%) for the year and to continued high interest in usage of the UI-TR for education, training, research and service I activiti:s. With additional personnellicensed, run time might well have exceeded the highest level recorded in the 1983-1984 reporting year; with one more part time person licensed (SRO) at year's end and another RO expected early in the next reporting year, the outlook is good for increased run time in the next year.

            'In summary these figures in Table 1111 indicate continued high and diverse utilization of the UFTR facility over the last five (5) years with research and educational usage maintained this year despite the presence of several large outages contributing in contrast to the previous reporting s ear. The design and implementation of various new facilities has played a key role here to enhance and promote educational, training and research utilization at all levels, in addition, the newly implemented neutron radiography facility has been available for the entire year and is now nearing optimization to provide a strong base for continued growth and diversification of um;e during the upcoming year 111-4 1                              - - - - - - - - - _ _ _ _ _ - - . - - - - -

l as the facility is further optimized to attract mare users, several of whom have expressed interest in its use for research projects. Of course, the Reactor Sharing Program is planned to coatinue to play a key overall support role in encouraging facility usage in all categories as this support has again been renewed at a 12% increased level for the r, ext year. Table Ill 2 summarizes the different categories of reactor utilintion: (1) college and unier*% teaching, (2) research projects, (3) UITR operator training, requalification and recertification, (4) utility operator training, (5) experimental facilities enhancement plus l UFFR testing, maintenance, surveillance activities,(6)IIEU to LEU fuel conversion related l effortt, and (7) various tours and reactor operations demonstrations which is a final l category to account for all other planned usages.The absence of significant utility operator training is a noteworthy point; efforts continue to schedule some utility usages during the upcoming year but, othi than an occasional SRO requiring a few hours of training for a _ utility managc .aent positin. there is Nie interest by utilities in training programs so this i is not a likely area for 11rge scale increases in facility usage. 1 College course vcilization involved 19 different courses, some many times to account for over 197 hours of actual run time, an increase of nearly 16G over the previous year. The research utilization consisted of some 22 projects using nearly 264 hours of actual reactor run time exclusive of internal research into reactor characteristics. This number of usage hours was decreased by about 12% from the previous year, primarily because of staff commitments to other activities including the SPERT 1.EU fuel inventory and inspute 4 as well as efforts to license additional operators. Both these categories included con- , siderable concurrent usage to optimize personnel utilization. As noted, there are iacreases in several areas from the last reporting year, especially in the training and educational programs supported under the DOE Reactor Sharing Program. This program plus a large r 111-5 L -

amount of internally supported usage for education and research plus several service activities all contribute to maintain the total facility utili:.ation at high levels especially since growth in University of Florida course usage has slowed considerably. With many educational and several large research projects (including several sponsored by reactor sharing and several deriving from the University of Florida Nuclear Engineering Sciences Department) already scheduled for the upcoming year, this next year promises to produce facility utilization at a similar or even higher level than that experienced during the last two reporting years, primarily because of the availability of more licensed personnel. A single utility operator training program could also produce a substantial increase in usage time by itself. With several significant maintenance projects completed and performed during the reporting year, this expected high usage in the upcoming year is realistic especially in the areas of educational usage for college courses and for research, both on and off campus. I Table III 3 contains a breakdown delinc.ating the 16 schools and their 109 usages of the UFTR facilities which were sponsored under the Department of Energy Reactor Sharing Program grant. Dese Reactor Sharing usages account for nearly 31 hours of run time in Categories 1 and 5 in Table III 2 and nearly 220 hours of run time in Category 2

          ~

related to research and have resulted in maintaining and fostering improved visibility for the UFTR around the State of Florida and also among researchers and other users at the University of Florida many of whom are just beginning to recognize the unique capabilities of the UFTR facilities. Several new inquiries for involvement in the Reactor Sharing program have been received again this year; several new users have also been accom-modated, in all, the 109 usages represent a small decrease from last year although the total of 31 participating faculty represents a significant increase as does the diversity and length of individual usages. The 120 students involved also represent a decrease a!though the 1116

diversity of groups involved again balances this decrease as a positive factor. Much of the increased diversity is due to the effort to involve high school science students in research and education programs at the UITR which will receive renewed emphasis for the upcoming year. Obviously this DOE Program is a key driving force behind the continued utilization and growth of interest in the UFTR facility. This publicity is certainly n key factor is explaining the continued large number of visitors (569) who toured the facility again this year. Therefore, the UFTR facility continues to build and support a j base for long term permanent growth nnd support of facility utilization with the Reactor Sharing Program serving as the catalyst for this growth.The implementation of the various facility improvements such as the PC based analyzers and software in the NAA Laboratory l and the radiography facility are simply spinoffs from the various expressed needs of those visiting the facility in conjunction with staff interests in diversification of capabilities and can only serve to increase usage possibilities. Similarly, r.s the neutron radiography facility has i l become functional though optimization and final design efforts continue, plans are being formulated to investigate the feasibility cf implementing a prompt gamma analysis facility at the UITR. l Detailed in Table 1114 are the monthly and total energy generation figures, as well r as the hours at full power per month and totals for this past year, The UFTR generated 26.6S Mw brs during this twelve month reporting period, down somewhat (~10%) from last year but still the seventh largest value L UFTR operating history and the fourth highest in the last thirteen years despite several significant outages. Since there were several research usages such as the Plasma Kinetics and Neutron Radiography projects as well as extensive operations training to license new UFTR operators where tne usage was lengthy but at relatively low or fluctuating power levels, the power generation could have been 1117 l - - - . - - - - - - - - - - - - - - - -

considerably higher. Indeed, even with a 79.2% availability factor for the year, the real limitation on usage has been personnel availability and funded support for desired usages. Described in Table 1115 is the monthly breakdown of usage and availability data. As noted in Section 1 of this report, thcre were several significant forced outages for maintenance during the 1987 1988 reporting year in contrast to the previous year so the overall availability is down somewhat to 79.2% from 91.5% with 2 months at 100%. Nevertheless, a significant part of the 20.8% unavailability (nearly 3%) is attributed to personnel vacations and leave as well as the administrative shutdown required to allow fuel cooling prior to the biennial tuel inspection, not malfunctions. Similarly, Table 1116 contains a detailed breakdown of days unavailable each month with a brief description of the l primary contributors. The overall availability of 79.2% is approximately the average over l the last five years but improvement is to be expected in the upcoming year as several l outages were utilized to perform multiple maintenance projects and, as shown in the data l in Table 1116, key causes of failures have generally been isolated and corrected to limit recurrences of related failures. Such a maintenance philosophy is expected to assure l continued high availability, hopefully above 90% in the next year.

           ~

Described in Table ill 7A is an explanation and date for one unscheduled trip for ( the reporting period. As explained, the trip was on overpower due to student error with evaluation indicating it was not promptly reportable. Table 1117B also contains one entry for scheduled trips, in this case the trip was used to demonstrate the rapid decay and recovery of stack count rate with power reduction and increase for radiation protection technology students, ne lack of more scheduled trips is primarily due to the lack of utility training programs where such trips are part of the training exercises. All safety systems responded properly for both trips as described in Table lil 7A and Table lil-7B with neither 111-8 L _ _ _ _ _ _ _ _ _ _ _ .________ ___

considered to be promptly reportable. Several repartable incidents described as unusual occurrences (and per UITR Tech Specs sometimes potentially abnormal occurrences) occurred during ;his reporting year. Table 1118 contains a descriptive log of seven (7) unusual occurrences with relatively brief descriptive evaluations of each. Several of these occurrences as the more significant entries were promptly reportable to include those in entries 1,3,4 and 6 Entry 1 addresses the failure of the S 2 control blade to withdraw upon demand due to hardened grease deposits in the drive motor gear assembly with similar problems noted to be developing in the other drive motor gear assemb!!es. Entries 3 and 4 address two occurrences where the Safety Channel 1 Indication momentarily dropped to zero and gradually recovered over a few l second interval. Finally, Entry 6 addresses the burn out of the Control Blade Safety 2 clutch indicating lamp causing Safety 2 to drop to the fully inserted position while at full power. Although unusual occurrence Entries 1,3 and 4 are most significant and were promptly reported along with Entry 6, the rest are reportcd via this report. In some cases these may not need to be reported at all except as required by the UFTR Reactor Safety Review 1 Subcommittee and good practice to document and assure proper facility management control of operations. No uncontrolled releases of radioactivity have occurred from the facility and controlled releases remain well within established limits. The personnel radiation exposures for 1987-1988 have been maintained near the usual low yearly level since there was no extensive dose commitment for maintenance in the core or other high dose rate areas as for the control blade drive system project of the 1985-1936 reporting year. There was also no waste or special nuclear material shipped this year; however, waste is expected to be shipped in the upcoming reporting year to prepare the facility for the llEU to LEU fuel Ill-9 l- - _-- - - - - _ _ _ _ - - - - - - - - - - _ _ _ _ - - - _ - - _ - - - - _ - - - - - - - - - - - - - - - - - - - . _ - - - - - - - - - - - - -

conversion activities to commence within the next two years. With the corrective action implemented following the NRC Health Physics inspection in February,1987, the upcoming waste shipment is assured to be properly controlled and documented. Environmental radioactivity surveillances continue to show no detectable off site dose attributable to the UFIR facility as also noted in Section VII as the facility continues to operate within ALARA guidelines with minimal exposure of staff and visitors. 111-1 0

TABLE 111-1

SUMMARY

OF FACillrY UTIIIZATION (September,1987 - August,1988) NOTE: The projects marked with one asterisk (*) indicate irradiations or neutron activations. The projects marked with two asterisks (") indicate train-ing/ educational use. The projects marked with three asterisks ("*) indicate demonstrations of reactor operations."Experiment Time"is total time that the facility dedicates to a particular use; it includes *Run Time." *Run Time" is inclusive time commencing with reactor startup and ending with shutdown and securing of the reactor. RUN EMRIMENP TIME TIME PROJECT AND USER TYPE OF ACTIVITY (hours) (hours)

    "ENU 5176L         -   Dr.      Independent Reactor Operations                                                                                       152.87    298.70 W.G. Vernetson, P.M.            Laboratory Course for Under-                                                                                         (22.06)   (31.59) l Whaley and Reactor              graduate and Graduate Nuclear Staff                           Engineering Sciences Students

{

    "CFCC Radiation Pro-            Three Reactor Operations Based                                                                                       24.79     195.90 tection Technology Pro-         Radiological Control and Protec-                                                                                     (2.99)    (33.31) gram      -   Mrs. R.           tion Training Programs of Coop-Raw's/'.str. S. MacKen-         erative Work Exercises zie - Reactor Sharing

{ 'NAA Research on NAA Evaluation and De- 21.94 28.20 Volcanic Rock Samples velopment ofIrradiation Schemes (4.42) (4.82) Dr.' Mark DeFant - for Identification and Quantifica-University of South Flo- tion of Rare Earth and Other rida (Tampa) - Reactor Elemental Constituents in Stan-Sharing dards and Volcanic Rock Samples With Subsequent Companson With Other Laboratory Results

     'NAA Research - Dr.             Estimation of the 1 123/l 127                                                                                       2.38      3.75 C. Williams / Dr. M.             Ratio in Radiopharmaceuticals                                                                                       (0.35)    (0.63) nornor/       Gainesville        Using Instrumental Neutron Acti-V.A. Ilospital, Dr.         vation Analysis W.G. Vernetson and P.M. Whaley, UFTR l

11.11

TABLE III-1 (CONTINUED) RUN EXIGIMEN1' TIME TIME PROJECT AND USER TYPE OF ACTIVITY (hours) (hours)

  'NAA        Research    on  Evaluation of Effects of Oil Re-                                          78.21              85.05 Seagrass        Community   lated Drilling Fluids on Various                                          (4.21)             (5.92)

Samples - Dr. C. Seagrass Community hiodels D'Asaro - University of Containing Shellfish, Grasses and West Florida Dr. D. Other Organic and Inorganic Webber - EPA and Components Reactor Sharing Faeility Charac- Evaluation and Optimization of 0.20 032 terization - Dr. W.G. Thermal Column Beam for (0.20) (032) . Vernetson Planned Neutron Irradiation of Electronic Components { innovative Nuclear Pulsed Ionization Chamber Plas- 28.55 98.42 Space Power Institute - ma Kinetics Diagnostic System (11.83) Plasma Kinetics Pars- Operational Tests to Indude { meter Determinations - Temperature Dependent Plasma Partial Seed Project - Kinetics Analysis of UF Ile Plas-Dr. W.11. Ellis mas Within Small IIxternally IIcated Detectors hiaintenance Activities Corrective and Preventive Main- 0.45 2438 on Control Blade Drive tenance on All Control Blade (2.53) Motor Reduction Gears Drive Motor Reduction Gear As-

  -      W.G.      Vernetson,  semb!!cs to Restore and Assure Reactor Staff                Free Removal on Demand
     'hAA      Research on     Neutron Activation Analysis of                                             4.54             8.41 Biological Media Dr.        DNA to Determine Elemental                                                                               i R. Rill, Biology Dept.,   Sodium Content Florida State University "ENU-4905       -  NAA    Special     Project on Veri-                                                 6.46            9.02 Research on NBS Stan-     fication/ Benchmarking of Trace                                               (1.22)         (13S) dards Dr. W.G. Ver-       Elements in Various NBS Stan-                                                                            l netson/L Tryboski         dards for Use as Secondary Stan-                                                                         J dards and Crosschecks.

i 111-1 2

TABLE 1111 (CONTINUED) RUN EXPERIMbNT TIME TIME PROJECT AND USER TYPE OF ACTIVITY (hours) (hours) Research on Properties Use of Neutron Radiography, 18.55 62.33 of Materials - Dr. S. Transmission and Scattering Ex-Turner, Nusertech periments and Other Analytical Techniques to Examine and Cha-racterize Used and Unused Bora-Gex Absorber Liners From Utility Spent Fuel Pools

  '"Ilawthorne Middle       Lecture, Tour and Demonstration                                             0.42                         3.25 School Science Students   of Reactor Facility Operations
  - Ken Wilson - Reactor    and Use of Rabbit System for Sharing                   Neutron Activation Analysis
  "*FAS 6428        -   Dr. Lecture, Tour and Demonstration                                             0.50                         1.00 f  M.O. Balaban              of Reactor Operations and Fa.                                               (0.50)                       (1.00) cility Capabilities l  Neutron Radiography       Continued Neutron and Gamma                                                 26.92                        82.83 Facility Development      Flux Measurements in the Ther-                                               (1.33)                      (14.30) and        implementation mal Column Facility With Rear-Studies - Dr. W.G. Ver-   rangement of Thermal Column netson,       Dr. A.M. Graphite and Other Special Ma-Jacobs, P.M. Whaley,      terials Plus Beam Quality Analy.

II. Ilicks and Reactor sis and Optimization to Evaluate Staff Neutron Radiography Pot ntial

           -                 Based on a Continuing Series of Test Neutron Radiographs for FacilityImplementationIncluding Darkroom Development "Utility Reactor Oper-    Performance of a Set of Meaning-                                             12.48                       19.15 ations Usage - W.G.       ful Reactor Operations Exercises Vernet:,on                Involving Significant Reactivity i                             Manipulations Plus a Minimum of 10 Startups and 10 Shutdowms for Georgia Power Company Plant Vogtle Operations Supervisor SRO Candidate 111-1 3                                                                                  l i                                                                  - - - - - - - - - _ - - - - - - - -          - - - - - - - - -            ----- o

TABLE 111-1 (CONTINUED) RUN EXERIMENT TIME 'I1ME PROJECT AND USER TYPE OF ACTIVITY (hours) (hours) ,

     "* Brownie Scouts of         Tour and Demonstration of Reac. 0.00       2.50 America      -   Mrs. K. tor and NAA Laboratory Facility McCarthy                     Features for Brownie Scout Troop
     'NAA        Research    on   NAA to Evaluate Rare Earth          18.35      23.17 Estuarine Samples Dr.        Elements in Tampa Bay Estuary       (0.98)     (1.33)

G. Smith / Dr. R. Byrne Sediments Using Special Sample

     - University of South        Iloider to Irradiate Five Samples Florida, St. Petersburg -    in UFTR CVP With Results Aug-Reactor Sharing           mented by Short Re irradiatir.ns of Samples Previously Irradiated in the Sample lioider                                     r l
      ' Materials /NAA Re-         Irradiation er Geologic Quartz      17.42      19.58 i     search -        Dr. A.L       Samples to D:termme Uranium,                   (0.59) l     Odom, Geology Dept.,         Thorium and Saminium Trace Florida State University     Element Content for input into Research on Effects of Natural
!                                  Radiation on Geologic Quartz and Geologic Dating Based on Radioactive Decay
       ' Materials Research on     Determination of Elemental          7.84       8.99 Silica-Crystal Genera-       Chlorine andTitanium Contentin      (3.97)     (3.97) tion Processes - Dr. C.      Silica Samples (SiO2 ) Generated Balaban and Mr. G.           Using a Special Process Under LaTorre     -  Advanced      Development Materials      Research Center "EMA 3050 - NAA             NAA Class Team Project to Cha-     8.82        10.53 Project for L Worth          ractuize the Major Constituent     (1.59)      (2.00) and     R. llanrahan     -   Metallic Elements in Coal Fly W.G. Vernetson               Ash including Tour for Project Participants
        ' Florida Foundation of     Continuation of Summer 1987        0.00        6.50 Future Scientists - Faci-   Student Research Program: Ex-o        lity Characterization -     perimental Characterization of D r. W.G. Ve r-             the Neutron Spectra in Various netson/Kurt Mondlak         UFTR Ports HI-14

f TABLE 111-1 (CONTINUED) { RUN EXPERL\ TEST TIhiE TIME PROJECT AND USER TYPE OF ACTIVITY (hours'l (hours) l 'NAA Research on NAA to Evaluate the Trace hie- 28.51 33.19 Coal Slurry Samples - tal Content of Sediments From (8.94) (10.05) Dr. Ralph Llewellyn - Coal Slurry Settling Ponds and Physics Dept., Umver- the hiagnitude of the Potential sity of Central Florida - Source of Such Trace hietals Per Reactor Sharing BTU of Energy Recovered Ver-sus the Use of Virgin Coal l

        "ENU 4194       -   NAA      ENU-4194 Senior Project to            2.82       3.83 Educational Research         Learn INAA and Apply NAA to

[ Project on Ash From Evaluate Rare Earth Trace Ele-hit. St. IIelen's Vol- ment Content of Ash Obtained

                                                        ~

canic Eruption - Dr. Following the htt. St. IIelen's l W.G. Eruption Vernetson/P. Kuta, University of Florida [ Facility Equipment hiaic Storage, llandling, Inven- 0.00 21.17 Usage and Special hia- tory, Preparation and Disposition (4.75) , terials llandling b/ Fa- of irradiated Steel Specimens for cility Staff - W.G. Ver- hiaterials Science Analysis,Trans-netson and UFTR Staff fer of Non Radioactive Chemicals From UFTR Building Plus Re-ceipt, Ilandling and Transfer of Two Co 60 Sources for Radiation Chemistry and Radiation Biology Programs NRC and Other Inspec- Regular Biennial NRC Opera- 2.50 61.00 tions - W.G. Vernetson tions Inspection, Increased Fte- (2.50) (32.00) quency llealth Physics Inspection, Plus ANI Property and Nuclear Liability inspections "Licensed Operator NRC Requalification Training 3.81 61.80 Requalification and Re- Requirements Including Lectures, (2.25) (7.92) certification Program Practical Training. Examinations, Training Including Staff Startups, Shutdowns and Reac-Planning / Review hicet- tivity hianipulations as Necessary ings - Dr. W.G. Vernet-son / Reactor Staff / Rad Con Staff III-15 1 .. ..

i TABLE III 1 (CONTINUED) RUN EXWRIMENT f TIME TIME PROJECT AND USER TYPE OF ACTIVITY (hours) (hours)

 "UFTR Reactor Oper-      Individual Reactor Operator Li-                           101.89    285.73 ator Candidate Training  cense Training for UFTR Reactor                           (87.18)   (160.29)
 - Dr. W.O. Vernetson/. Operator Candidates M. Wachtel, i Reactor Staff            R. Piciullo and J. Godfrey as Well as Rabbit System Operator                                                 l Candidates t
 "Union County High       Lecture, Tour and Damonstration                           1.07      5.08 SchoolScience Program    of Reactor and Rabbit System
 - Renae Allen - Reac-    Operation for Neutron Activation tor Sharing              Analysis
 "

• Florida Regional Four Ixetures, Tours and De- 0.67 4.58 Junior Science, En- monstrations of Facility Opera- (0.67) (133) gineering and Humani- tions and Capabilities for High ties Symposium - Dr. School Student and Teacher Par-

! W.G. Vernetson/- ticipants Reactor Staff l "ENV-4201/5206 - Dr. Lectures Tour and Demonstra- 0.85 233 C.E. Roessler tions of Reactor Operations and Radiation Protection Related l Features of the UFTR Facility "ENU 6935 - Nuclear Lecture, Tour and Demomtration 033 1.17 L Seminar - Prof. J.S. of Reactor Operations and Fa- (033) (0.50) Tulenko cility Capabilities "ENU-4505L - Dr. Senior Level Nuclear Engineering 10.91 2738 W.11. Ellis, Dr. G.R. 12boratory Exercises and Experi- (0.25) Dalton and Dr. W.G. ments Including Foil Irradiations, l Vernetson - University Flux Mapping. Hot Channel Fac-of Florida tors, Reactor Calorimetry, Blade , Reactivity Worth Calibration, Dif-l fusion Length in Graphite,1/M Approach to Critical and Neutron Activation Analysis ( III-16 L - - - - - - - - - - - - - - - - - -

TABLE 111-1 (CONTINUED) RUN EXWRINiENT f TIh1E TlhiE PROJECT AND USER TYPE OF ACTIVITY (hours) (hours) f I'

hiaintenance Project to Corrective hiaintenance, Replace. 8.89 101.92 Correct Interm.ittent ment of Failed Parts, implemen-(22.42)

Downscale Failure of tation of Testing and Evaluation UFTR S:=fety Channel Program and System Checkout to

    #1 Indication - Two       Restore UFTR to Operational Occurrences    -     W.G. Status Following Brief Downscale Vernetson,     Reactor    Failure of Safety Channel #1 and l    Staff                     Subsequent Recurrence of Event SPERT Iow Enriched        Decontamination Work, Radia-          0.00      199.25 Fuel Conversion Re-       tion / Contamination Surveys, Pro-             (46.25) lated Efforts             perty Surveys, Facility hiodifica-tions, Fire Alarm System hiain-tenance, LEU SFERT Fuel hiovement, Security System hiodification, NRC Radiation Safety inspection, and LEU Fuel Inventory and Visual Inspection

Efforts "ENU-4144 - Dr. Lecture, Tour and Discussion of 0.00 2.00 W.G. Vernetson Facility Operations for a Senior Level Systems Course Compar-ing/ Contrasting UFTR Systems With Corresponding Power Reac-tor Systems "ENV 6211L - Dr. Lecture, Tour and Demonstration 0.00 1.50 C.E. Roessler/Dr. W.G. of Facility Capabilities Emphasiz- )

Vernetson ing Radiation hionitoring Instru-mentation

     '"1988 Engineers' Fair    Lecture, Tour and Demonstration      0.00      0.75
     - Dr. W.G. Vernetson/-    of Facility Capabilities l     Reactor Staff "Santa Fe Community        Lecture, Tour and Demonstration      0.00     3.50 l

College Nuclear hiedi. of UFTR Operations with Radia-cine Radiologic Tech- tion Surveys and NAA Training nology Program - S. Exercises h!archionno - Reactor Sharing i L III-17 l [ . l

TABLE 1111 (CONTINUED) RUN EXERIMENT TIME TIME PROJECT AND USER TYPE OF ACTIVITY (hours) (hours)

  "Hillsborough      Com-   Lecture, Tour and Demonstration      0.00     3.00 munity College Nuclear    of Facility Operations with Radia-l  Medicine and Radiation    tion Surveys and NAA Lectures Therapy      Technology   and Training Exercises Program - Dr. M.

i l Lombardi - Reactor Sharing "S t. Augustine High Ixeture, Tour and Demonstration 1.48 5.75 School Science Class - of Reactor Facility Operations Ms. E. Doyle/Mr. S. and Use of Rabbit System for Buell- Reactor Sharing NAA Exercises and Thermal Column for Neutron Radiography "ENU 3002 - Dr. G.S. Lecture, Tour and Demonstration 0.50 4.17 Roessler/Dr.W.G.Ver- of Reactor Operations With Neu-netson tron Astivation Analysis "Florida Institute of Lecture, Tour and Demonstration 2.58 7.17 Technology Society of of UFTR Operations With Radia-Physics Students, Phy- tion Surveys, Use of Rabbit Sys- l sies Dept. - Dr. W.G. tem for NAA and Use of "Dier-Vernetson/E, Thomas - mal Column for Neutron Radio-Reactor Sharing graphy l "Boca Ciega High Lectures, Tours and Demonstra- 0.00 1.50 School Science Dept. - tions of Reactor and NAA Labo-Dr.11. Bevis - Univer- ratory Facility Capabilities sity of Florida - Reactor Sharing l

   "* Region IV Seminar      Lecture, Tour and Demonstration      0.00      1.25 on Advanced X Ray         of Reactor, NAA btboratory Fa-Procedures - W.S. Pro-    cility and Radiography Capabili-(   perzio                    ties

( 111 18

\                                        _____-__-- _

l TABLE 111-1 (CONTINUED) RUN EXIBIMENT TIME TIME PROJECT AND USER TYPE OF ACTIVITY (hours) (hours)

    '"National Junior         Lectures, Tours and Demonstra-                0.00     3.50 Science and Humanities    tions of Reactor and NAA 1. abo-                                i Symposium - Dr. B. Ab-    ratory Facility Capabilities for bott                      Participants in the National JSil Symposium
    'NAA      Research     on NAA to Evaluate the Rare Earth                7.78      12.17   I Phosphate Ore - Dr. P. Elemental Content of Phosphate                (1,05)    (2.75)

Glelisse, Mechanical Orcs Above the Trace Element Engineering (FAMU Level for Possible Mining Appli-and FSU) and Dr. cations Clark, Chemistry Dept., l Florida State University

    - Seed Project - Reac-tor Sharing                                                                               l l
    'NAA      Research     on NAA to Evaluate the Feasibility               1.52      3.00    l Florida lake Sediments    of Determining the Environmen-                          (0.25)

Dr. Claire Schelske, tal Level of Elemental Germani-Fis h e rie s and um in 12ke Sediments in Florida Aquaculture Laborato-ry, University of Florida

    - Seed Project
     ' Physics of Materials    Investigation of Fast Neutron and            3.40      9.17 Properties Research -      Gamma Ray Fluence Induced                              (0.67)

Dr. Ilans Plendl, Phy. Lattice Disturbances and Optical sics Dept., Florida State Properties in Dielectric (Topaz) University and Dr. Meterials; Work to Date involves Peter Gielisse - Me- Design of Cadmium Covered Ma-chanical Engineering terial IIolder and NAA to Eval-Dept., FAMU/FSU - unte and Quantify Trace Element Reactor Sharing Content of Holder Material "CilS-5110/5110L - 1ecture, Tour and Demonstration 0.52 3.33 Dr. Muga and Dr. K. of Reactor and NAA Lab Opera- (1.00) Williams tions for Radiochemistry Re-search Lecture and laboratory Courses 111-1 9 L .. - - - - - - - - . -

TABLE 1111 (CONTINUED) RUN EXERIMENT TlhtE 'I1ME PROJECT AND USER TYPE OF ACTIVITY (hours) (hours)

  "CHS 5110L      Dr. K. Radiochemistry Laboratory Pro-           5.45       7.83 l  Williams                ject for NAA of Powdered htilk/-

Exercises on llatf-Life Determin-ation,' on Neutron Activation 1 Analysis of Silver and Aluminum in hietal Samples and for one Student Project on NAA of Pow-l dered htilk

  'ENV-6936 - Health       Evaluation of the hiethodology          37.67      48.98 Physics    Rad!oactMty   used to hicasvre Argon 41 Stack         (12.60)    (16.17)

Release Research - Dr. Effluent Releases at the UFFR W.G. Vernetson, Nu- to Include Implementation of a clear Engineering Better Source to hiatch Sample Sciences Dept. and Dr. Geometry in Response to NRC W.E. Bolch, Emiron- Inspection Findings mental Engineering Sciences Dept., Univer-sity of Florida "ENU 6516L - Dr. Graduate Level Nuclear En- 10.49 25.57 W.H. Ellis, Dr. G.R. gineering 12boratory Exercises (2.08) (7.12) i Dalton and Dr. W.G. and Experiments Including Foil Vernetson - University Irradiations,1/ht Approach To- , of Florida Critical, Neutron / Gamma Flux l l hiapping, Neutron Activation l Anc. lysis, Inverse Reactor Kinetics hieasurements, Control Blade Reactivity Worth hicasurements and Demonstration of Neutron Radiography Implementation 1

   '"Florida Foundation     lecture, Tour and Demenstration         0.00       9.67 of Future Scientists -   of Reactor Facility Operations                     (3.00) l Dr. W.G. Vernetson -     and Experimental Capabilities Reactor Sharing          Plus Project Selection for Two FFFS liigh School Students (Jas-on hiusgrove of Escambia Ifigh

( School and Joe Nefflen of Glades Central Community liigh School) 111-2 0 l

t TABLE 1111 (CONTINUED) i RUN EXIYRIMENT TihtB TIhiE f l PROJECT AND USER TYPE OF ACTIVITY (hours) (hours) l

 ' Florida Foundation of Summer 1988 Student Research                            12.40     14.67 Future Scientists - NAA Program: Comparison Bench-                              (8.42)    (9.68)

Research on NBS Stan- marking of Non Cer'.ified Ele. dards - Dr. W.G. Ver- ments in NBS Standards Using , netson/J. Nefflen NAA

• Florida Foundation of Summer 1988 Student Research 12.40 14.67 Future Scientists NAA Program: NAA to Evaluate the (8.40) (9.70)

Research on Volcanic Rare Earth Trace Element Con-Ash - Dr. W.G. Vernet- tent in hit. St. IIelen's Volcanic son /J. hiusgrove Ash I

 "* Florida Foundation   Lecture, Tour and Demonstration                         0.62      2.50 of Future Scientists -  of Reactor Operations for FFFS Dr. W.G. Vernetson/-    Summer Research Program liigh hir D. Roberts - Reac-  School Students I tor Sharing
  'NAA Research on Soil  NAA Research for Blogeochemi-                            16.62     19.57 l and VegetationSamples   cal Assessment of Soil and Vege-                        (0.57)    (0.68)
 - Dr. Gary Cwick, Uni-  tation Samples From the Pollard, versity of West Florida Alabama Oil Field to Quantify
 - Reactor Sharing       Potentially Abnormal Levels of Various hietals

( *"htiscellaneousTours hiiscellaneous Tours invoMng 4.58 15.42

 - Dr. W.G. Vernetson    Facility Demonstrations for Vari-                       (4.58)     (8.92) ous Visitors Including Groups of

( Students Representing Various Special Interests, Alumni, Poten-tial New Staff hiembers, NES ( Seminar $peakers, ROTCinstruc-tors, UPD Officers, NRC Ucense Examiners, Visits by Potential or Actual Facility Users and Various Other Interested Individuals and Small Groups 111-2 1

h TABLE 1111 (CONTINUED) RUN EXIBllMENT TIh1E TIME PROJECT AND USER TYPE OF ACTIVITY (hours) (hours) l l Required Surveillances Scheduled UFTR Facility Com- 52.25 138.78

       -      W G. Vernetson/-                  ponent and System Testing, Sur-                           (13.61)    (23.83)

Reactor Staff veillance, Calibration and Related hicasurement and Verification Actisities Required by Technical Srcifications, Procedures or NRC Commitments hiaintenance Activities Preventive and Corrective hiain- 1,15 170.18 Reactor Staff tenance and/or Replacement of (15.58) UFTR Facility Components Ex-cluding hiinor Items and Those Listed Individually to include System Testing as Necessary TOTAL 763.39 2328.97 l (197.00) (500.63) TOTAL ACTUAL 56S.35 1823.34 1 1 1, Values in parentheses represent eiultiple or concurrent facility utilization (Run or Expetiment time); l that is, the resetor was already being utilized in a primary run or methity for a project so a reactor [ training or demonstration utilization could be conducted concurrently with a scheduled NAA trradiation, I course experiment, or other reactor run.Thus, the actual reactor run time for the 19S719S$ reporting l year is $68.35 hours, an increase of nearly 3% over the presious year (552.52 hours). In contrast, the l actual experiment time for the 198719SS reporting year is significantly increased at 1828.34 hours, an increase of about 36% Indicating inercased utilliation of staff time this year for reactor usage and other projects including better record keeping of project times using the facility but not the reactor. Indeed, l nearly 200 hours of experiment time was devoted to non reactor senices such as work with or related l to the LEU SPERT fuel, acceptance and transfer of Co-00 sources for other departments and transfer of non radioactive chemicals from the radiochemistry laboratory.The run t'me and experiment time before the reduction for concurrent usages shows many simultaneous multiple usages assured optimal effort of staff time despite the relathe unavailability of one long time SRO due to illness for most of the year until a replacement SRO was licensed in July,1983. Of course, the experiment time continues to include considerable reactor usage for coricethe maintenance and surveillance acthities; however, the numbers this year indicate high tescis of quality faellity usage directed to research, education, training and penice, especially as driven by the Reactor Sharing Program usages.

2. Exp. Time is run time (total Ley on time minus chedout time) plus set up time for esperiments or other reactor or facility usage including checkouts, tests and maintenance insching the reactor facility.

111-2 2 h- .

TABLE 1112 UFTR UTIUZATION

SUMMARY

Utilization Categories Run Time Experiment Time (hours) (hours)

1. College Courses and Laboratories (18) 227.89 (30.79) 606.93 (78.15)

2. Research Activities (22) 318.08 (53.91) 499.78 (79.04)

3. UFTR Operator Training and Re-qualificat;on (2) 105.70 (89.43) 347.53 (16821)

4. Utility Operator Training (1) 12.48 19.15

( 5. UFTR Maintenance, Testing and Sur-veillance ActMties, Experimental Facilities Enhancement and Facility Equipment Usage and Special Materials Ilandling (10) 92.36 (17.64) 600.58 (115.73)

6. l{EU to LEU Fuel Conversion Related l Efforts (1) 0.00 199.25 (46.25)

7. Reactor Tours and Demonstrations l Including liigh School Classes (11) 8.84(5.251 5125 (1325)

TOTAL 765.35 (197.00) 2IE97 (50063) NOTE 1: The same meaning is attached to values in parentheses in Table III 2 as in Table 111 1. Values in parentheses adjacent to topic areas indicate the number of entries from Table 1111 that were collapsed into this utilization category. NOTE 2: The first two categories of College Courses and 12boratories as well as ( Research Activities plus the last category for high school group demonstra-tions include significant usages sponsored under the Department of Energy ) UFTR Reactor Sharing Program which allowed sixteen (16) schools to have l 109 usages of the UFTR facilities as delineated in Table 111-3. His usage by 16 schools is the most diverse usage yet recorded under the University of Florida Reactor Sharing Program. NOTE 3: In some cases the assignment of items to one of the seven (7) categories is somewhat arbitrary especially for non college tour groups for whom lectures and other training is conducted or research performed to aid facihty l modification or development. I NOTE 4: Console checks are excluded from this Utilization Summary but are estimated to account for about 10 hours additional utilization per month, in addition, non specialized and usually non scheduled murs for one or a few persons are nc, srmally tracked in this Utilization Summ:'.ry. Dese types of tour actiuties typically involve about 510 hours of additional time per month and are offered on an as available basis depending on staff availability. 111-2 3

E f TABLE 1113 1987-1938 REACTOR SIIARING PROGRAM

SUMMARY

OF USAGE OF JFTR FACILITIFE Users School Usages' Faculty Students Boca Ciega fligh School 1 2 13 Central Florida Community College (CFCC) 43 2 23 Florida Institute of Technology (FIT) 1 1 13 Florida Foundation of Future Scientists 7 2 2 (Escambia High Schrol Ells and Glades Central Conar'anity fligh School - GCCilS) Florida State University (FSU) 9 4 2 Ilawthctne Middle School (HMS) 1 2 15 i Hillsborough Community College (llCC) 1 1 7 Santa Fe Community College (SFCC) 1 1 8 Spruce Creek High School (SCIIS) 1 1 1 St. Augustine fligh School (SAllS) 1 3 18  ; i Union County liigh School (UCllS) 1 1 9 - 1 University of Central Florida (UCF) 11 1 2  ! University of South Florida, 3 2 2 St. Petersburg (USF SP) l University of South Florida, Tampa (USF T) 9 2 2 l University of West 71orida (UWF) 19 6 3 i TOTAL 109 31 120  !

1. Usage is defined as utilization of the Unhersity of Florida Training Reactor facilities for all or any part l of a day. In many esses a school can have multiple usages but all related to the same research project l or training program such as the one project for the Unnersity of West Morida that involved long term l irradiations or the multiple usage training program for Central norida Community College. l l

111-2 4 f

TABLE III-4 i MONTHLY REACTOR ENERGY GENERATION' (September,1987 - August,1988) Hours at ' l Monthly Totals Kw-Hrs Full Power September,1987 4208.156 41.267 October,1987 2274.868 21.817 November,1987 1976.683 17.650 December,1987 2652.842 25.184 January,1988 1905.342 18.667  ! February,1988 1881.840 18.334 March,1988 - 644.085 5.250 t April,1988 1104.453 ?734 May,1988 1413.164 13.151  : June,1988 1221.938 10.667 July,1988 3181.SS9 28.216 i August,1988 4211352 41.439 YEARLY TOTAL 26,676.6122 250376

1. The yearly total energy generation cf 26.68 hiegawatt. hours for the 19S71938 reporting year represents a 10% decrease over the previous reporting 3 rat while the 25033 hours at full power represent a similar t 10% decrease owr the prenous year. This decrease in energy generation was primarily due to considerable low power usage of the UFTR for IRTR operator training and rescarch on plasma kinetics e parameters as well as implementation of the neutron radiography fccility plus large time commitments l for work efforts related to decontamination, mowment, inventory and other work with the LEU fuel stored and hoped to be used in the UFTR IIEU.to LEU conwrsion. Several oatages, one toimplement r corrective and prevenths maintenance on all control blade drive motor reduction gear assemblics to restore free motion and two others to evaluate and cortcet the intermittent loss of indication downseale on Safety Channel #1 also c.used lost facility usage and hence affected energy generation negathcly.  ;

The total run time for the facility was maintained somewhat above the previous year at 56835 hours for  ; this reporting year indicating considerable low power run time for neutron rad;ography and the UF  ; plasma Linctics experiments as well as UFTR operator training; owrall, the indication is toward l considerabic low power usage and continued high utilitation of the reactor when the reactor and the i necessary licensed operators are available. With the addition of one new SRO at the end of thh year f and another RO expected early in the next reporting year, the availability of operating personnel should be improved. With the additional continued high t.tilitation and with the good availability experienced owr the final two months of the reporting year coupled with more licensed personnel, larger yearly energy generation values can be expected for the neu reporting year.

2. Tbe 26,676.612 Kw-llrs of energy generation is still the fourth hi hest g one year tot:.1 energy generation i o er the last twche years of UFTR operation and represents the sewnth highest one year value in the 2Sycar operational history of the UFTR. l I

I t 111-2 5 l

p TABLE 1115 r MONTilLY REACTOR USAGE /AVAllA31LITY DATA (September,1987 - nugust,1988) Monthlv Totals Kev On Time Exo. Time' Run Time Availability September,1987 73.50 hrs. 163.28 hrs. 61.40 hrs. 79.2 % . October,1987 55.60 hrs. 146.00 hrs. 49.58 hrs. 93.5 % November,1987 42.60 hrs. 109.73 hrs. 38.42 hrs. 93.3 %  : December,1987 52.00 hrs. 121.55 hrs. 45.92 hrs. 87.9 % I January,1988 37.20 hrs. 153.72 hrs. 29.58 hrs. 71.8 % February,1988 60.80 hrs. 151.08 hrs. 54.82 hrs. 100.0 % March,1988 34.50 hrs. 175.00 hrs. 29.72 hrs. 43.6 %

• April,1988 32.60 hrs. 154.00 hrs. 29.75 hrs. 38.3 %

May,1988 42.60 hrs. 143.85 hrs. 35.35 hrs. 100.0 % . r June,1988 60.80 hrs. 170.08 hrs. 50.48 brs. 57.5 % l July,1988 75.60 hrs. 146.18 hrs. 70.48 hrs. 99.2 %  ! August,1988 77.20 hrs. 193.87 hrs. 72.85 hrs. 85.5 % TOTALS: 645.00 hrs. 1828.34 hrs. 56835 hrs. 79.2%' l Experiment T'me is Run Time (Total Key-On Time minus Checkout Time) plus set up time for experiments, tours, or other facility usage induding checkouts, tests and maintenance involong reactor running or facility usage. htonthly Average availabilityis 79.2%; on the basis of days of the ) ear, the availability is similarly 79.2% as indicated in Table 111-6. The yearly availability is down from the historical high of 91.5% recorded last year. Newrtheless, the 79.2% availability is accentable and with repairs made, the avail.bility in the upcoming year is en, :cted to again return to 90%. De large value of run time shows continued high utilization of the UFTR facility. III-26

I r C , TABLE III 6 UFTR AVAILABILITY

SUMMARY

(September,1987 - August,1988) 1 i Days Primary Cause of Month Availability Unavailable Lost Aval! ability September,1987 70.2 % 6.25 days Maintenance to replace a failed Safety-2 control blade drive motor , (3/4 day) and to replace the pri-mary coolant demineralizer resins and connections on the deminera-lizer (1/2 day). Maintenance to clean the drive mo- l tor gea assemblies, to free them of hardened grease and replace worn i bearings to restore free withdrawal on the Safety-2 control blade with similar preventive actions on the L other three control blade drive mo-tor gear assemblies (5 days). October,1987 93.5 % 2.00 days Maintenance and repairs related to restoring and assuring proper dilute  ; fan operation and RPM indication l (1 1/2 days) plus replacement of  : the flex coupling on the dilute fan i duct (1/2 day). [ November,1987 93.3 % 2.00 days Maintenance to repair the stack  ! dilute fan shaft (1 1/4 days) and to replace the temperature recorder l ink pads, to replace the shield tank [ ceramic filter and to replace the  ! control blade clutch current indicat- L ing lamps (1/2 day). December,1987 87.9 % 3.75 days Maintenance to restore the safety blade S 2 blade position indicator l plus vacation /hohday leave time (3  ! days). l i 111-2 7 l r

TABLE III 6 (CONTINUED) UFTR AVAllABILITY

SUMMARY

(September,1987 - August,1988) Days Primary Cause of hiODth Availtbility Unavailable Lost Availability , January,1988 71.8 % 8.75 days Maintenance to re. store proper re-sponse of Safety 2 Regulating Blade Position Indicators (1 1/2 days), maintenance to check out and re. [ place the chopper card of the two-pen recorder (6 days) plus vacation-holiday leave time (1 day). February,1988 100.0 % 0 days - - - - - - - - - - - - - - - - - - March,1988 43.6 % 17.50 days Maintenance to isolate and replace l a failed feedback capacitor in the noise filter circuit of Safety Channel 1 following a few second lots of channel indication (transient loss of indication and trip function) to al- ' low restart (171/2 days). Mainte-nance also to replaced failed con-sole analog clock with digital clock l and to clean the Safety 1/ Log cali-brate switch to remove noise in circuit response (concurrent). April,1988 38.3.s 18.50 days Maintenance to address recurrence of Safety Channel 1 transient down-scale circuit failure (161/2 days) and replacement of failed APD motor (2 days) plus relamping of the reactor cell (concurrent). May,1988 100.0 % 0 days - - - - - - - - - - - - - - 111-2 8

TABLE !!I-6 (CONTINUED) UFTR AVAILABILITY

SUMMARY

(September,1987 - August,1988) < Days Primary Cause of Month Availability Unavailable Lmt Availabilltv. , June,1988 57.5 % 12.75 days Maintenance to replace burned out ( S-2 control blade clutch current lamps and maintenance on the two-3en strip chart recorder used for alade drop time measurements (4 1/2 days). Maintenance to replace failed meter movement and GM Tube in the APD (2 days) and maintenance to clean and oil the temperature re-corder (1/4 day). Administrative shutdown required to allow fuel cooling prior to fuel inspection (5 days) as well as unstacking / restack-ing and preparation time (1 day). July,1988 99.2 % 0.25 days Maintenance to refill the primary coolant storage tank (1/4 day). August,19 A 85.5 % 4.50 days Maintenance to repair the stack dilute fan RPM indicator and then install and calibrate a new stack dilute fan RPM indicator (31/2 - days) arid maintenance to move the control room status board, install a dustless marker board and paint the control room (1 day). TOTAL ANNUAL UNAVAILABILITY: 76.25 days = 20.8% TOTAL ANNUAL AVAILABILITY: 289.75 days = 79.2% NOTE 1: This availability summary neglects all minor unavailabilities for periods smaller than one quarter day. In most cases these periods are for much less than an hour as some minor problem is corrected. This availability summary also neglects unavailability for scheduled tests and surveillances except where roed. NOTE 2: Of the 76.25 dap unavailability, ordy 66.25 dap were due to forced unavailability due to maintenance for repairs, delay awaiting parts arrival, trip evaluations, etc. The remaining 10 dap were for personnel vacations, leaves, decay of the fuel radionuclide inventory prior to fuel inspection, etc. where the reactor was or could hae been fully operationat. 111-2 9

TABLE lil 7A l UNSCllEDULED TRIPS' i During this reporting year, the UFTR experienced only one unscheduled trip which is ( der,cribed below; the trip is not considered to have affected reactor safety or the health and safety of UFTR personnel or the public. All safety systems responded properly and a full review was conducted prior to restart. Date Description of Occurrence 4 April 1988 At 1700 hours during a Reactor Operations 1.aboratory (ENU 5176L) training exercise, with SRO P.M Whaley directing operations, while W. Coughl!n was noting in the log that he was increasing to 100 kw using the automatic servo control mechanism, J. Riverota incorrectly adjusted the linear range selector switch to the next position (calibrate). At this point, the demanded power was 100%, the actual power was about 95%, but the signal to the servo control mechanism was 85%. The reacter autocontrol system responded by withdrawal of the regulating blade to a 30 see period until, at 125% power, Safety Channel I responded properly in less than ~10 seconds (<7 seconds) by initiating a reactor trip through the reactor protection system.The reactor operator in charge noted the incorrect switch setting upon occurrence of this event but within the few second time frame did not see the event developing in time to correct the switch setting. This event was noted to constitute a reactor trip from a known cause and was therefore considered not a promptly reportable occurrence. All student operators were counselled on verbatim compliance with procedures and the proper way to undertake control manipulations during such training excretses. For evaluation of the overpower condition, the trip evaluation file for the April 1,1987 overpower trip was referenced v,here operator error (and hence potential prompt reportability) was involved. 'Ihe UFTR is designed with safety analyses addressing up to 625 kw so the overpower event has no impact on system safety or the health and safety of the public and all safety systems responded properly to cause the tr p. i All safety systems responded to perform their intended safety function for the trip listed in this Table. 111 30

TABLE 111-7B SCIIEDULED TRIPS There was only one scheduled trip performed for training or ex; imental purposes during this reportin,g year. Part of the reason for this lack of schedu eu trips was the failure to schedule any utility operator training programs where such trips are a designed part of the training program. It is expected that some training trips will be included in the ENU 5176L Reactor Operations Laboratory course for the upcoming reporting year. Date Description of Occurrence 15 June 1988 At 1615 hours on 15 June 1988, with the UFTR operating at full power for 3 minutes, a manual training trip was conducted by SRO P.M. Whaley to demonstrate rapid decay and recovery of stack count rate with power reduction and increase as part of Argon 41 Stack Effluent Measurernent Exercise for two Cooperative Work Training Prograrn students from Central Florida Community College. All safety systems responded properly with the renetor restarted beginning at 1619 hours.  ! t i 5 l 3 l l [

                                                                                                                                                         }

l. p

)

a l I i i l f

t 111-3 1 l 1

l

TABLE 111-8 LOG OF UNUSUAL OCCURRENCES During this reporting year there were no events which are considered to have compromised reactor safety or the health and safety of the public. Several events, classified as unusual occurrences, are described below as they deviated from the norma functioning of the facility and are included here as the most important such deviations for the reporting year. Unscheduled shutdowns are included here as well. Trips are not addressed here since they are inclucid in Table III 7. Date Description of Occurrence

1. 25 Sep 87 On September 8,1987, when the Safety 2 control blade failed to withdraw upon demand. replacement of the failed S 2 drive motor with an identical spare restored the system to normal operation (un-der MLP #87 26). On September 25,1987, the Safety 2 control blade again failed to withdraw during a daily checkout. Under MLP #87 29, subsequent checks showed the motor to drive unit coupling to be bound due to hemy hardened deposits of waxy material binding the pinion gear in the reduction gear assembly. The remaining control blade gear assemblies were also inspected and found to be in similar though earlier stages of the same condition. All gear cases were re-moved, disassembled, soaked in a solvent, cleaned and inspected under standard work procedures outlined in MLP #87 29. The lower and motor side bearings of the Safety 2 and other control blade reduction gears (worm gear and main shaft) were also found to be coated with the same deposits inhibiting smooth operation so they were replaced with duplicates. Following reassembly the only modification involved was replacement of the bearing retainer brass C-clips with commercial-ly standard steel E clips to perform the same function through a small enlargement of the E clip 5 ot in the retainer end of the vettical worm gear shaft. nis change was reviewed under 50.59 Evaluation No. 87-15 not to involve an unreviewed safety question.

Detailed disassembly of this gear box arrangement had not previously been considered necessary in assuring integrity of the drive system. Ilowever, recurrence of this failure event will be prevented through

                                    ?criodic inspection of the reduction gear as part of the five year nspection of the control blade drive systeme (V-1 Surveillance). nis is the only portion of the control blade drive system operation not previously examined under the V 1 Surveillance. Such periodic checks will be facilitated with installation of the modified E-clip washers. Fol-lowing completion of all maintenance and surveillance checks, the reactor was returned to normal operat .,ns with no further problems noted.

Ill '.,2 L

TABLE III 8 (CONTINUED) LOG OF UNUSUAL OCCURRENCES __D,at e Description of Occurre_nce Reportability under Tech Specs Section 6.6.2 "Special Reports," Para-graph 3(c) was not considered to require a special report since the failed gear system is not part of the safety system as delineated in UFTR Tech Specs Section 5.5.2. In addition, the failure was in a fail- , safe mode discovered during shutdown and did not render the reactor safety system incapable of performing its intended function. For these reasons the event was not considered to be promptly reportable to  ; NRC, though the NRC was notified in a timely T. anner prior to restart for their consideration of the occurrence.

2. 26 Oct 87 At 1602 hours during the ENU 5176L Reactor Operations 12boratory class with the reactor critical at 100 watts, the diluting fan rpm reading dropped to 410 rpm versus a normal value of 480 520 rpm and a required value per SOP. A.1 of 425 rpm. Following a normal un-scheduled shutdown, the diluting fan belts were founo to be slipping and in need of replacement. Under h!LP #87 33 the old belts were removed and new ones from stocked spares were installed to restore the rpm reading to 510 rpm. The new belts were run overnight and verified to be cperating properly on 27 October 1987 prior to com-mencing the daily preoperational checkout. De evaluation of the event indicated no radiological impact and no prompt reportability since the UFTR was shut down immediately from ~100 watts upon discovery of the system failure with the stack count rate at only ~2 cps. Upon completion of all restart conditions, the reactor was restored to normal operation with no further dilute fan rpm problems noted.

3. 14 hfar 88 At 1437 hours with a Reactor Operations l2boratory class (ENU-5176L) in progress at 50% power, Safety Channel 1 failed to the bottom meter stop. P.ht. Whaley, operator at the controls, noted the indications on Safety Channel 2, the log pen recorder, and the wide range indicator were normal and directed a reactor shutdown. Before the shutdown could be started (a few seconds), Safety Channel I returned to normal indication. The subjective evaluation was that the return was not instantaneous, but the meter returned to normal indication relatively slowly (i.e., not as if switched on, but rather as if recovering from an electrical transient). ne shutdown was completed with all instruments responding normally at 1438 hours.

111-3 3

TABLE III 8 (CONTINUED) LOO OF UNUSUAL OCCURRENCES Date Descriotion of Occurrence De immediate indications were that an intermittent fault had developed in the circuitry for Safety Channel 1 (part of the wide range drawer) but not in any other secticn of the wide range drawer. Under MLP #88 9, failure of a feedback capacitor was determined to be one cause of such a failure so it was replaced with a substitute of different manufac urer (see 10 CFR 50.59 Evaluation #88-4) with identical specifications. During a restart run to verify problem correction, the event recurred after 5 minutes at 100 kw, At t als point per MLP #88 10, extensive checks were made of all components in Safety Channel I with the Channel responding properly to a wide range of tests, checks and surveillances including external signals. The cause of the problem was not specifically isolated though the fission chamber, the preamplifier and the connections and cables were variously suspected and recommended as the root cause of the downscale failure occur. rence with the final consensus being that the problem may have been a cable / connection problem that was fixed by the checks or there could be a problem only identifiable and isolatable with the system operating at power (current from the fission chamber) Tnd special voltage and current monitoring instrumentation temporanty installed. Derefore, a special test procedure was generated and approved to al-low reactor restart in stepst the reactor was declared operable with propr checkouts performed and compensating features implemented to melude a second reactor operator assigned to observe Safety Channel 1 indications during operation. His special test procedure was intended to verify proper operation of Safety Channel I by monitoring the voltage level in the preamplifier with respect to ground, the current drawn by fission chamber detector operation from the high voltage supply and the high voltage power supply output voltage while the UFTR was operated at power levels in steps up to full power for an exte ded tun to demonstrate correction of the Safety Channel 1 failure problem or, in the event of recurrence, to enable isolation of  ; the fault. De result was a successful restart with the reactor declared < ready to return to normal operations on 1 April 1988 per RSRS and  ! NRC communications though the root cause had not been specifically isolated. Except during the transient, all functions of indication and trips were not inhibited or changed; that is, there was only a temporary loss of indication and function in Safety Channel #1. During the test restart , the compensating measure of two operators was more than adequate to assure reactor safety and protection of the health and safety of the public as the UFTR was declared operable. De impact of this failure on system operation was evaluated to be negligible. i 111-3 4 l

r - 3 TABLE III 8 (CONTINUED) LOG 0F UNUSUAL OCCURRFMCES l i Date DesHption of Occurrence

4. 9 April 88 At 1209 hours, with Reactor Operations Laboratory class (ENU-5176L)in progress at ~75% power, Safety Channel 1 again failed to the bottom meter stop. The aperator agam noted the indications on {

Safety Channel 2, the ) og pen recorder, and th wide range meter were all normal and commenced a reactor shutdown. Again, before the shutdown could be completed, Safety Channel I returned to normal indication.ne subjective evaluation was again thH the return was not instantaneous, but the meter returned to normal Indication relatively slowly. The shutdown with the reactor secured was completed with all instruments responding normall" at 1214 hours. The immediate indications were that the intermittent fault had recurred in the circuitry for Safety Channel 1 (part of the wide range drawer) but not in any other section of the wide range drawer. Under MLP #8S-14, the noise suppression capacitors in both Safety Channel feedback loops were checked. ne failed SC-2 capacitor was replaced with a substitute of different manufacturer (see 10 CFR 50.59 Evaluation No. SS 9) with identical specifications. In addition under MLP #8S 14, 50.59 Evaluation No. SS 9 was used to control the change of unlabeled RG71 cable connectors to Amphenol 68175 connectors while 50.59 Evaluation No. SS-11 was used to replace the wide range drawer Safety Channel 1 signal cable compression type RG62 fitting with a crimp type which, due to corrosion and looseness, was thought to be a good candidate as the root cause of the intermit-tent failure. Since this corrective action was thought to be the most likely problem area but could not be verified, the two Safety Channel amplifiers and meters were switched to provide conclusive evidence that the problem was internal or external to the console should it recur. Per MLP #SS 14. extensive shecks et all components in Safety Channel I wera again made with the Safety Channel responding properly to a wide range of tests, checks and surveillances including external signals.The cause of the problern.was not specifically isolated ! though the fission chamber, the preamplifier and the connections and cables were variously suspected and recommended as the root cause of the events with the final consensus being that the problem could l have been a cable / connection problem that was likely corrected by the use of the crimp type connectors. Of course, there could be a problem only identifiable and isolatable with the system operating at poveer l (current from fission chamber) and special voltage and current i monitoring Iris?rumentation temporarily installed, 111-3 5 t _ _ ------------ I

1 i TABLE 111-8 (CONTINUED) , LOG OF UNUSUAL OCCURRENCES ' Date Description of Occurrence , i Another special test procedure was generated and approved to allow  ; reactor restart in steps; the reactor was declared operable with proper checkouts performed and compensating features implemented to

        'nclude a second staff member assigned to observe the Safety Channel                                      l 1 and 2 meters during operation. His special test procedure was inte.

nded to verify proper operation of Safety Channels 1 and 2 by  ! monitoring the voltage level in the preamplifier with respect to grou. , nd, the current drawn by detector operation from the high voltage  ! supply and the high voltage power supply output voltage while the  ! UFTR was operated at power levels in steps up to full power for an  !; extended run to demonstrate correction of the Safety Channel 1 failure problem or,in the event of recurrence, to enable isolation of the fault  ! external or internal to the console due to the amplifier switch, t i, As part Of the plans connected with the Special Test Procedure, reeurrenn ut the Safety Channel problem in Safety Channel 2 would i isolate the pblem to the am plifier/ meter circuit outside the console ' with the next nep planned to be replacement of the preamplifier and,  ! for another recurrence, replacement of the fission chamber. De result  ! was a successful restart with the reactor declared ready to return to [ normal operations except for the extra staff person monitoring safety , channels for all operations until 10 hours of operation above 50 kw  ! could be logged. This return to normal operations was authorized on l 28 April 19S8 per RSRS and NRC communications though the root  ! cause had again not been specifically isolated. I Again, except during the transient, all functions of indication and trips  ; were not inhibited or changed; that is, there was only a temporary loss l of indication and function (probably though not confirmed) in Safety  ; Channel #1. During the test restart the compensating measure of two , indhiduals was more than adequate to assure reactor safety and j protection of the health and safety of the public as the UFTR was r declared operable. l t I l [ 111-3 6  ! l i

TABLE 111-8 (CONTINUED) IDG OF UNUSUAL, OCCURRENCES l Date Descriotion of Occurrence 1 i After successful completion of the staged restart begun on April 25  ; and completed on April 27,1988, the UFTR was authorized to return < to normal operations with only the requirement for compensated , operations in that a second competent staff person had to be in the control room to monitor the Safety Channel meters for all operations until 10 hours operation above 50 kw had been completed with normal experimental and training usages of the UFTR to be approved and i conducted provided there was no recurrence of safety channel failure. ' After having met this compensated operations requirement, the UFTR was returned to uncompensated operations; that is, no extra person i monitoring the Safety Channels. At this point the corrective action was , considered successful and the reactor returned to normal cocrations l but with a caution to operations staff that no root cause had et been found. His return to uncompensated operations was compfeted on May 23,1988 with no recurrence of the Safety Channel failure to date. He corrective action taken is considered to have corrected the failure problems though admittedly no root cause has been found. At this  ! time the Safety Channel failure incident is considered closed and a final report to this effect was transmitted to NRC with a letter dated l June 9,1988 (see Appendix E of this report).  ;

5. 25 April 88 At 1230 hours during the restart following completion of work to i address the intermittent downscale failure of Safety Channel 1 '

indication, the UFTR was undergoing a hold at 50 kw during the phased return to full power operations with special monitoring  ; equipment installed in the Safety Channel I circuit when the Air i Particulate Detector (APD) motor was noted to be smoking. An unscheduled shutdown was undertaken by SRO P.M Whaley with the r UFTR secured at 1231 hours. Under MLP #8817, the APD was taken  ! out of service and a new motor assembly from stocked spares was , installed under 10 CFR 50.59 Evaluation No. 8814 which was required f since the existing motor had been a modification when installed. Going i back to the original motor mounting per the tech manual required some modifications per 50.59 Evaluation No. 8814 and restored air flow to a high but desirable level. Upon return to service no further >

                ?roblems were noted. Dough not promptly reportaole since the                                                  !

I JFTR was promptly shutdown to avoid violation of a Limiting  : l Condition for Operation, the NRC was informed of this occurrence on ( i 27 April 19SS in updating the status of UFTR preparations for return i to normal operation with only two individuals monitoring SC-1 and l SC 2. 111-3 7 r i

l l l 6 TABLE 1118 (CONTINUED) LOG OF UNUSUAL OCCURRENCES Date Descriotion of Occurrence

6. 10 June 88 At 1720 hours after approximately 51/2 hours of sample irrtdiation at full power (supporting neutron activation analysis), the Control Blade Safety.2 clutch indicating lamp burned out, dropping Control  !'

Blade Safety.2 from 56% withdrawn to the fully inserted pos. tion.De ' reactor operator at the controls responded promptly by conducting an unscheduled reactor shutdown per SOP A.4 (Reactor Shutdown) with the reactor secured at 1722 hours with all control and safety systems  ! responding properly. Subsequent restart was recommended following replacement and retest of all clutch indicating lamps including associated required survell-lances such as control blade drive and drop time checks, Discussion and review of Technical Specifications requirements indicated that a special report was required for such an uncontrolled or unanticipated change in reactivity greater than one dollar (where reactor trips from known causes only are excepted this event not technically meeting the definition of trip).  ; Immediate corrective action consisted of replacing the burned out clutch current indicating bulb and all o'hers to include the necessary { control blade drive and drop time checks. All checks were successful so the reactor was restored to operating status on June 15,198tl  ; Although staff and ksi.$ evaluations showed there were no radiologi-cal or safety consequences in this event, such failures are w x . avoided; also, since a clutch indicating lamp failure occurred during  ! the previous reporting year (at shutdown conditions), tha, frequency of  ; preventive clutch current indicating lamp replacement was increased t from annual to semiannual scheduled preventive maintenance.His l action is in agreement with previous management evaluations and is  ! expscted to reduce significar.tly the likelihood of recurrence of this failure event. To date there have been no recurrences. l l I h l l 111 38 i i

TABW 1118 (CONTINUED) l l LOG OF UNUSUAL OCCURRENCES [ i l Date Description of Occurrence l l 7, 10 Aug 88 At 0958 hours the dilution fan RPM indication dropped from 495 l RPM to 465 RPM. Although a shutdown was not required, the drop , in indicated RPM was a potential developing loss of a required [ indication so the reactor was shutdown and secured by the SRO at  ! 0959 hours with other systems responding normally. Under MLP #88-  : 39 the connectors on the RPM indicator were cleaned and the  ; coupling tightened to restore proper RPM indication. Subsequent  ! reactor operations on 10 August were made with no further problems ; noted; however, on 12 August 1988, the RPM indication was again l found to be low, As a result, under MLP #88 40, a replaccinent RPM i indicator was installed and calibrated on 15 August 1988 with no further problems noted, t l l c l i I l

l. k I

i I i i r r l { L l l I l 111-3 9 i l c

I IV. MODIFICATIONS TO Tile OPERA 11NG CilARACTERIS11CS OR CAPABILITIES OF Tile UFTR A number of modifications were made to the operating characteristics or capabilities of the , UFTR and directly related facilities during the 1987 1988 reporting period. These I modifications were all subjected to 10 CFR 50.59 cvaluations and then determinations (as necessary) to assure no unreviewed safety questions were involved. Carried over from the 19841985 Lporting Year: Modification 6: Replacement of Vent System Manometers Modification 7: Addition of City Water Flow Sensors (Rotameters) Carried over from the 19861987 Reporting Year: Modification 87 4: SAR Revision 3, Part I (Emironmental Monitoring)

1. SAR Revhion 3. Part I (Emiromntal Monitoring) (Permanent Closed item)

(Modification 87 4; Evaluation / Determination Completed 21 May 1987) Part I of this revis!on substantively changes specifications to permit the use of emironmental monitoring devices other than film badges; it also allows monitoring wints to be selected based on evaluation rcther than the SAR specification of

           ;ocations and changes the specification that the UFTR staff collect film badges for processing (a radiation control function as the Radiation Control staff acts on behalf of the UFTR staff).

Part I of this revision also corrects typographical errors (a misspelling) and c4erical errors (omissions from the table of contents, a misplaced paragraph in the description of radiation monitors, an insorrect specification of a Radirdion Control SOP). Part 11 of Revision 3 simply adds an Appendix 15G to Chapter 5 of the SAR to l address Wigner Energy Considerations for UFTR graphite fires. De entire SAR Revision 3 has been included in all official copies of the UFTR FSAR; however, the changeover to exclusive use of thermoluminescent dosimeters instead of film badges has not yet been implemented. Controlling Document: UFTR SAR Revision 3 Documentation Package Submitted to USNRC . See Appendix C of 1986 1987 Annual Report IV 1 _ _ _ - - - - - - - - - - - - - - - - - - - - - - - I

i

2. Temocrary Removal of hinterialt Science Annex Fire (Stuckc)_lktection capability j (Temporary Closed Item)  ;

t J i (hfodification 8710, Evf :. n Completed 27 August 1987) Evaluation was made to allow the removat of a smoke detector in the reactor building annex; fire detection capability for the reactor cell and the ares directly adjacent to the reactor cell rema ned unaffected. l Controlling Document: hialntenance log Page 87 22 (Closed on 3 November 1987) l Primary Coolant Return Une Trio Timing Check (01 Surveillance) (Temporary - Closed Item) ! (htodification 8711: Evaluation Completed 24 September 1987) I' The Primary Coolsnt Return Une Trip function is tested as part of the quarterly scram check Q 1 surveillance Part of this test involves noting the time required for , f the system to drain sufficiently to initiate a return line trip. During the 27 August 1987 checks, the time for trip initiation following securing primary coolant Dow was l noted to have decreased significantly a conse wative change.This change was noted i and then evaluated not to involve any unreviewed safety question. [ Controlling Document: 10 CFR 50.59 Evaluation No. 8711 i

4. Ruet of Primary Coolant Return Une Reed Switch (Permanent Closed item) l (blodification 8712: Evaluation Completed 24 September 19S7)  ;

i Investigation into the decreased time for the no return line Dow trip as noted on the  ! quarterly scram checks performed on 27 August 1987 indicated that the Dow rate  ! (water level) trip point in the Dow switch is particularly sensitive to the position of l the active magnetic reed switch. Since the reed switch was not secured in the switch j housing. It was evaluated to have shifted position. Therefore, the reed switch was i subsequently repositioned to make the timing check value approximately the same as historical values after an evaluation was performed to document agreement that i repositioning the reed switch to its previous position would not involve any l unreviewed safety questions. Controlling Doeament: hisintenance leg Page S7 25 (Closed on31 August 1987) l r 1 IV 2

5. UFTR SAR Revhlon 4 on Fire Protection and Communicptions Systems (Permanent Closed)

(Modification 8713: Evaluation / Determination Completed 24 September 1987) Revision 4 to the UFTR SAR was reviewed and submitted ta the USNRC. This revision affected a number of pages to include:

1) Updating the FSAR to reflect an upgraded (previously reviewed) four rone fire detection system for the entire reactor building with monitor box at the Emergency Support Facility cutside the building:

2) Correcting an error regarding the description of the UFTR cell fire extinguishersi and

3) Updating the FSAR description of the communications between the UFD1 and University of Florida Radiation Control office.

This revision was noted to be essentially administrative in nature to update existing previously reviewed UFTR status and conditions and was determired not to involved any unreviewed safety questions. Controlling Document: UFTR SAR Resision 4 Documentation Package . See Appendix D of this Report

6. Temperature Dependant Plasma Kinetics Experiments on Solid Urnnium FluiOD Qttg.bers Up to 10 atm Ratira (Experimental Closed item) l (Modification 8714: Evaluation / Determination Completed 24 September 1987) l A proposal was presented for the use of fission chambers containing solid urardum depos ts with a heater assembly and with specified fill gas compositions and pressures I in the UFTR thermal column to perform temperature dependent plasma kinetics j measurements. This experiment was evaluated and determined not to involve any unreviewed safety questions.

Controlling Document: Run Request 87 49 i IV-3

7. Washer Reolacement/ Shaft Groove Enlargement for Bearing Retention on all Control Bladp_ Drive Motor Vertical Shafts (Permanent - Closed Item)

(Modification 87-15: Evaluation Completed 23 October 1987) During inspection, overhaul and repair of the control blade drive motor gear assembly mechanisms relative to Maintenance Log Page 87-29, removal of shaft bearings was accomplished resulting in the deformation of bearing retainer clips. An exact replacement for the clips was not commercially available. Replacement of the brass C-clips with standard commercial stainless steel E-clips was proposed with supporting work to expand the retaining clip groove to fit the standard clip. This modification was implemented in the restoration of the control blade drive motor gear assemblies to proper operation after being evaluated not to involve any unreviewed safety questions. Controlling Document: Maintenance Log Page 87-29 (Closed on 30 September 1987) H

8. Low Level Radioactive Material Storage Enclosure (Permanent - C'-

(Modification 87-16: Evaluation Completed 22 October 1987) The installation of a woven mesh wire cage to segregate and control the cell area used for storage of low level radioactive materials (experiments, port plugs, reactor waste, etc.) from the remainder of the reactor cell (with the capability for securing access) was evaluated not to involve any unreviewed nfety questions. Controlling Document: Maintenance Log Page 87-35 (Closed on 13 November 1987)

9. Correction of Control Blade S-3 Reactivity Worth Curve (Permanent - Closed Item)

(Modification 87-17: Evaluation Completed 17 December 1987) The reactivity axis of the Control Blade Safety 3 reactivity worth curve was noted to have an inadvertent scale change from 0.0002 units per division to 0.0001 units per division in the range of 0.01 to 0.012 ak/k on the integral worth curve. The data was reviewed and the curve reconstructed with consistent scale. This ano.naly was evaluated not to involve any unreviewed safety questions. IV-4

10. Temoorary Reolacement of Teletector in Room 108 NSC (Temporary - Closed Item)

(Modification 87-18: Evaluation Completed 27 January 1988) One GM tube failed in the  : range (accident) beta-gamma radiation detector maintained in the Emergency .ipport Facility. During the period when the part was on order and the unit undergoing repair and calibration, the substitution of an alternate high range beta-gamma survey instrument was evaluated to be acceptable and not to involve any unreviewed safety questions. Controlling Document: Maintenance Iag Page 87-46 (MLP 87 64 Closed on 6 January 1988)

11. Technical Specifications Aniendment #17 Core Vent Syggm Ooeration and Post Accident Samoling (Permanent - Closed Item)

(Modification 88-1: Evaluation Completed 22 March 1988) In support of UFTR Technical Specification Amendment 17, a valved penetration was proposed for installation on the rabbit system exhaust line (an auxiliary connection to the core vent system). This penetration, along with a rabbit exhaust line isolation valve as shown in Figure IV-1 allows sampling the core vent system prior to filtering and discharge and therefore the reactor cell atmosphere in the event that controlled venting in an accident scenario should be required. This modification was evaluated not to involve any unreviewed safety questions. Controlling Document: Maintenance Log Page 88-19 (Closed on 4 May 1988)

12. Use of Alternate Clock in Cell / Control Room (Temporary - Closed Item)

(Modification 88-2: Evaluation Completed 22 March 1988) Following a failure of the console analog clock, an evaluation was made that the use of an alternate clock permanently mounted on the UFTR cell north wall (in clear view for operating personnel) was acceptable for time-keeping functions and did not involve any unreviewed safety questions. Controlling Document: 10 CFR 50.59 Evaluation No. 88 2 IV-5

 .     -                                                              .         ~-              .-                         ~.  - . . .

13. Insertion of PuBe sources (1 Ci or 10 Ci) into UFTR Thermal Column (Experi-mental - Closed Item)

(Modification 88 3: Evaluation Completed 22 March 1988) t An evaluation was made that insertion of Pu Be sources in the thermal column (primarily to test for the capability of generating a high level signal for the UFTR power monitoring channels to allow testing of response without operating the reactor) does not involve any unreviewed safety questions. Controlling Document: Run Request 88-15

14. Reolacement of Failed Feedback Caoacitor in UFTR Safety Channel #1 Circuit (Permanent - Closed Item)

(Modification 88-4: Evaluation Completed 15 March 1988) An evaluation was made that replacement of a failed noise filter feedback capacitor in the UFTR Safety Channel 1 linear amplifier with an identically (electrically) rated capacitor in a larger frame made by a different manufacturer does not involve any unreviewed safety questions. Controlling Document: Maintenance Log Page 88 9 (Closed on 15 March 1988) i

15. Replacement of Console Clock (Permanent - Closed Item)

(Modification 88-5: Evaluation Completed 22 March 1988) Following the failure addressed in 10 CFR 50.59 Evaluation No. 88-2, an alternate (digital) clock was proposed for installation in the UFTR console to replace the previous installed but failed analog clock. This modification was evaluated not to involve any unreviewed safety questions. Controlling Document: Maintenance Log Page 88-8 (Closed on 23 March 1988)

16. Monitoring Safety Channel 1 Signah (Temporary - Closed Item)

(Modification 88 6: Evaluation / Determination Completed 28 March 1988) As the result of an intermittent power level monitoring failure in Safety Channel 1 (downscale signal failure with slow transient recovery of monitoring signal), a test procedure involving monitoring of Safety Channel 1 signals at various circuit locations (indicated in Figure VI 2) during operations was generated. This test j procedure was determined not to involve any unreviewed safety questions. Controlling Documents: Maintenance Log Page 88-10 (Closed on 31 March 1988) SpecialTest Procedure (Verification of Proper Operation of Safety Channel 1 Preamp and Detector) l IV-6

p . f

17. IrJLPfogam in "estore Safety Channel 1 to Unrestricted Oper.ption: 1) Reterminat-ing.XCl Cab uLinterchanging SCI and SC2 Amolifiers. 2) Reolacing Preamp an.d.3) Rep 1 sion Chamber (Temporary - Closed Item)

(Modific- a 88- Evaluation Completed 11 April 1988) i FolN mg the successful completion of the test program referenced in 10 CFR 50.59 Evaluation No. 88-6 and a subsequent recurrence of the channel failure, a program of systematic replacement (as required) of all components with the potential for ( consing the problem followed by reactor operations with a second individual monitoring Safety Channel 1 for a specified test interval was devised and evaluated r.ct to involve any unreviewed safety questions. This evaluation was principally concerned with the technical aspects of the test program. Controlling Documents: Maintenance Log Page 8814 (Closed on 28 April 1988) Special Test Procedure (Test Program for Restoration of Safety Channel 1 to Unrestricted Operation)

18. Test Program for Restoration of Safety Channel 1 to Unrestricted Operation (Integrated Program Evaluation) (Temporary - Closed Item)

(Modification 88 8: Evaluation / Determination Completed 11 April 1988) The test program referenced in 10 CFR 50.59 Evaluation No. 88 7 was reviewed as an integrated test program and determined not to involve any unreviewed safety questions. Controlling Documents: 10 CFR 50.59 Evaluation / Determinations 88-6,88-7 Maintenance Log Page 8814 (Closed on 28 April 1988) Special Test Procedure (Verification of Proper Operation of Safety Channel 1 Preamp and Detector)

19. Change of Unlabeled RG-71 Cable Connector to Amohenol 68175 (RG59

_ Equivalent) (Permanent - Closed Item) (Modification 88 9: Evaluation Completed 13 April 1988) Subsequent actions for the test program referenced in 10 Ci R 50.59 Evaluation No. 88 7 and 10 CFR 50.59 Evaluation / Determination No. 88-8 included replacement of wide range drawer and pre-amplifier cable connectors. During the accomplishment of the initial step of the test program, the replacement of a non standard cable connector with an equivalent common usage, better contacting connector was evaluated not to involve any unreviewed safety questions. Controlling Document: Maintenance Log Page 8814 (Closed on 28 April 1988) IV-7

i

20. Reolacement of SC2 Failed Feedback Capacitor (Permanent - Closed Item)

(Modification 88-10: Evaluation Completed 13 April 1988) Subsequent actions for the test program referenced in 10 CFR 50.59 Evaluation No. 88 7 and 10 CFR 50.59 Evaluation / Determination No. 88 8 included interchanging the SC2 amplifier circuit with the SC1 amplifier circuit. Based on the results of troubleshooting procedures associated with 10 CFR 50.59 Evaluation No. 88-4 and MLP 88 9, the filter capacitor in SC2 was checked prior to accomplishment of the substitution and found failed; an evaluation similar to 10 CFR 50.59 Evaluation No. 88-4 was performed for the noise filter capacitor in the SC2 amplifier to allow installation of a replacement capacitor. " Controlling Document: Maintenance Log Page 88-14 (Closed on 28 April 1988)

21. Reolacement of Wide Range Drawer Safety Channel 1 Signal Cable Comoression Tyne RG62 Connector with Crimo Tvoc (Permanent - Closed Item)

(Modification 8811: Evaluation Completed 13 April 1988) Subsequent actions for the test program referenced in 10 CFR 50.59 Evaluation No. ' i 88 7 and 10 CFR 50.59 Evaluation / Determination No. 88-8 included replacement I of Safety Channel I cable connectors. The replacement of the wide range drawer Safety Channel 1 signal cable compression type connector with a crimp type i connector was implemented after being evaluated not to involve any unreviewed l safety questions. < l Controlling Document: Maintenance Log Page 88-14 (Closed on 28 April 1988) l

22. Operation of Safety Channel 1 with High DC Offset (Temporary - Closed Item)

(Modification 88-12: Evaluation / Determination Completed 18 April 1988) - Subsequent actions for the test program referenced in 10 CFR 50.59 Evaluation No. i 88-7 and 10 CFR 50.59 Evaluation / Determination No. 88 8 included interchanging the SC2 amplifier circuit with the SC1 amplifier circuit. Differences in the operating characteristics of the identical operational amplifiers in SC1 and SC2 caused the DC offset to be higher for Safety Channel 1 than the value specified in the technical manual. Following technical evaluations including consultation with the UFTR console vendor, operation of the Safety Channel 1 linear amplifier with a high DC offset was evaluated not to involve any unreviewed safety questions. Controlling Document: Maintenance Log Page 8814 (Closed on 28 April 1988) IV 8 _ _ _ _ _ _ _ .. . - - - _ _ - _ -. . _ -_ 2

I

23. Alternate Method of Testing Secondarv Water Low Flow Trio Function (Temporary

      - Closed Item)

(Modification 88-13: Evaluation / Determination Completed 21 April 1988) Subsequent actions for the test program referenced in 10 CFR 50.59 Evaluation No. 88-7 and 10 CFR 50.59 Evaluation / Determination No. 88-8 included interchanging the SC2 amplifier circuit with the SC1 amplifier circuit. The minor differences in operating characteristics between SC1 and SC2 linear amplifiers caused the SC1 response to a ganged (log scale and linear Safety Channel 1) calibration switch position (used in checking the secondary water trip, simulating log scale power level above 1 kW) to be greater than the 125 kW trip point. An alternate method for checking the secondary water trip was proposed and implemented after being evaluated not to involve any unreviewed safety questions. Controlling Document: Maintenance Log Page 8814 (Closed on 28 April 1988) Special Test Procedure (Alternate Method for Testing Secondary Water Low Level Trip Function)

24. Reolacement of APD Motor /Comoressor With On Hand Spare (Permanent-Closed Item) l (Modification 8814: Evaluation Completed 26 April 19S8) l l

Following a failure of the Air Particulate Detector (APD) vacuum pump, an on-hand spare vacuum pump previously acquired for use in the APD was found to have l a different physical configuration than the installed pump. A modification to adapt the pump mounting bracket to fit the UFTR APD was proposed and implemented l after being evaluated not to involve any unreviewed safety questions. Controlling Document: Maintenance Log Page 8317 (Closed on 27 April 1988) l 25. APD Flow Rate Increase with Motor / Blower Reolacement (Permanent - Closed Item) (Modification 88-15: Evaluation Completed 8 June 1988) l Following the APD pump replacement referenced in 10 CFR 50.59 Evaluation No. 88-14, the APD air flow rate was noted to be significantly higher than previous operational characteristics. Since the higher air flow was well within the operating flow range recommended in the technical manual, this change in flow was proposed to be acceptable and implemented after being evaluated not to involve any unreviewed safety questions. Controlling Document: Maintenance Log Page 8817 (Closed on 27 April 1988) IV-9

26. Cobalt-60 Source Storage and Handling (Permanent - Closed Item)

(Modifi ation 8816: Evaluation Completed 8 June 1938) Two 600 Curie Co-60 irradiation facilities are housed in the Nuclear Science Center, adja ,ent to the Reactor Building. The use of the UFTR freight door and overhead cr.v 4e for transferring and handling replacement sources (and the spent sources) was pr , posed and implemented after being evaluated not to involve any unreviewed srlety questions. Controlling Document: Radiation Work Permit 88 31 (Closed on 6 May 1988) 27 Alternate Recorder Channel Pen hiotor (Permanent - Closed Item) (Modification 8817: Evaluation Completed 30 June 1983) The UFTR blade drop timing checks are performed with a high speed two channel recorder; when one channel pen motor failed, a replacement was obtained from an on hand spare recorder unit. The replacement motor unit was an older model with a slightly different physical configuration but the same characteristics. The replacement of the failed motor with the older model motor implemented after being evaluated not to involve any unreviewed safety questions. Controlling Document: Maintenance Log Page 88 28 (Closed on 14 June 1988)

28. UFTR FSAR Revision 5 Submittal to USNRC (Permanent - Closed Item)

(Modification 8818: Evaluation / Determination Completed 30 June 1988) Revision 5 to the UFTR SAR indicates updated values for UFTR operating paremeters, corrects typographical errors and provides a better description of UFTR experimental facilities and console instrumentation. 'Ihis change to the FSAR was determined not to represent any unreviewed safety question. Controlling Document: UFTR SAR Revision 4 Documentation Package See Appendix E of this Report

29. Substitute of Meter Movement for Air Particulate Detector (Temporary / Permanent -

Closed Item) (Modification 8819: Evaluation Completed 18 August 1988) The substitution of a larger and higher ranging meter movement for the APD was evaluated not to involve any unreviewed safety questions. Controlling Document: Maintenance Log Page 88 30 (Closed on 28 June 1988) IV-10

l l 30. Installation of Reactor Cell hianual Shutoff Valves for Rabbit System Samole and l Gas Return Lines (Permanent - Closed Item) (Modification 88 20: Evaluation Completed 18 August 1988) The installation of manual shutoff valves in the UFTR cell for the rabbit system sample transit line and gas supply / return line (indicated in Figure IV 3) was considered as a backup means of assuring control over sample insertions using the rabbit system. De manual shutoff valves were implemented after being evaluated not to involve any unreviewed safety questions. Controlling Document: Maintenance Img Page 88-35 (Closed on 25 July 1988)

31. Vertical Port Plug Material Modification (Permanent - Open Item)

(Modification 88 21: Evaluation Completed 18 August 1988) The construction and use of more effective vertical port shield plugs for better shielding around the ports was checked out after being evaluated not to involve any ., unreviewed safety questions.

32. Repositioning of Control Room Status floard and installation of Erasable Marker Board (Permanent - Closed Item)

(Modification 88-22: Evaluation Completed 18 August 1988) To improve control room presence for operating staffinvolved in reactor operations instruction and to reduce chalkdust in the control room, the Surveillance Status Board from the east control room wall to the south control room wall and an erasable marker board was installed on the east control room wall after these changes were evaluated not to involve any unreviewed safety questions. Controlling Document: Maintenance Log Page 88 38 (Closed on 31 August 1988)

33. Changes in Indicated Stack Dilute Fan RPM from Recalibration with Stroboscone (Permanent Closed Item)

(Modification 88 23: Evaluation Completed 18 August 1988) Following the replacement of a failed stack dilute fan tachometer with an identical replacement, a calibration was performed using a strobe tachorneter, ne indicated RPM values were noted to be about 530 RPM as opposed to values from the previous unit of about 490 RPM. This change was evaluated to be acceptable and not to involve any unreviewed safety questions. Controlling Document: Maintenance Log Page 88-40(Closed on 15 August 1988) l l IV 11

34. Installation of Optically Couoted Tachometer for Redundant Stack DilutiortEan RPM Indication (Permanent - Open Item)

(Modification 88-24: Evaluation Completed 18 August 1988) The installation of an independent channel for stack dilution fan RPM indication in the control room using an optically coupled tachometer was proposed and accepted as a reasonable potential redundant and diverse backup for the occasional failures of the existing mechanical coupling. Installation as a backup channel was approved after being evaluated not to involve any unreviewed safety questions. This modification remains open at the end of the reporting year. Controlling Document: 10 CFR 50.59 Evaluation No. 88 24 i i l i i l IV 12 . l s, ___ - _ _ _ _ _ _ _ __ _ . . _ _ -. _ _ _ _ _ _ . - _ _ . _

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                                                                                                                           '. 8 l                                l BIOLOGICAL                        ll l                                l SHIELDING ll h                        l, UFTR CELL                                                             l 3

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V. SIGNIFICANT MAINTENANCE, TESTS AND SURVEILLANCES OF UFFR REACTOR SYSTEMS AND FACILITIES i A review of records for the 1984-1985 reporting year shows extensive corrective and preventive maintenance was performed on all four control blade drive systems external to the biological shield. Similarly maintenance work during the 1985-1986 reporting year was even more extensive as the problem of a sticking safety blade (S-3) recurred on September 3,1985. The recurrence necessarily demanded a detailed and complete check of all control blade drive systems to determine finally ad correct the cause of the sticking blade internal to the biological shield with the 1986 1987 reporting year invohing relatively little maintenance and no large maintenance projects. For the current 1987-1988 reporting year, there were two dominant though . man.geable mairanance projects. The first large scale maintenance project during the ' 1987-1988 reporting year involved an extensive effort to clean the control blade drive motor gear assemblies to free them of hardened grease and replace worn bearings. Though only Safety-2 had failed to withdraw on demand, all gear assemblies had grease in various stages of hardening which was cleaned out and then replaced with fresh grease and new bearings, to restore free withdrawal of S 2 and assure free motion of all control blades. The second large scale project was involved with the evaluation, corrective action, testing and monitoring of the two safety channels due to two occurrecces of the downscale failure of the Safety Channel 1 meter indication (and probably the function). The extensive checks, maintenance efforts to clean connections, change connections and replace parts and special test development and implementation as well as the monitoring involved for the two occurrence.i easily make this the largest maintenance effort since the control blade drive system maintenance performed internal to the biological shield in the 1985-1986 reporting year. Other significant maintenance efforts in 1987-1988 were devoted to the diluting fan motor and RPM indicating system, the two pen recorder response and the blade position indicators for all control blades. Although corrective maintenance in the current reporting year is considerably increased over the previous reporting year, it is expected that much of the corrective and preventive maintenance performed this year will assure a retum to over 90% availability in the 1988-1989 reporting year. Indeed, the 79.2Fo availability for the year indicates more or less routine maintenance and surveillance checkt and tests throughout the year except for the two projects cited to demonstrate again the worth of the maintenance performed in the 1984-1986 reporting years. In the tables that follow, all significant maintenance, tests and surveillances of UFTR reactor systems and facilities are tabulated and briefly described in chronological order; these tabulations also include administrctive checks. Table V a contains all regularly scheduled surveillances, tests or other checks and maintenance required by the Technical Specifications, NRC commitments, UFfR Standard Operating Procedures, or other administrative controls; these items are normally delineated with a prefix letter and a number for tracking purposes. Table V-2 contains a listing of all the maintenance projects required to repair a failed system or component or to prevent a failure of a degraded , system or component. These are frequen;1y no: scheduled though they can be when a problem is noted to be developing and preventive actions are implemented. In addition, V-1

they frequently are associated with reactor unavailability. Finally, these maintenance items can be associated with surv'illances, checks or test items listed in Table V-1 since some of these scheduled surveillances are also required to be performed on a system after the system undergoes maintenance. For example, 'when the area monitor check sources or detectors are the subject of preventive or corrective maintenance as listed in Table V-2, the Q-2 calibration check of the area monitors must be completed as listed in Table V-1 before 7 the reactor is considered operable. In Table V 2 the first date for each entry is the date when the Maintenance Log , Page (MLP) was opened; the date for work completion and the MLP number are included at the end of the maintenance description. As a result, the first two items are listed in Table V-2 on starting dates prior to the beginning of the current reporting year. Dey are entered here because the maintenance was completed in the current year. i 1 4 1 j d V-2

TABLE V-1 CIIRONOLOGICAL TABUIATION AND DESCRil'I' ION OF SCIIEDULED UFTR SURVEILLANCES, CIIECKS AND TESTS Date Surveillance / Check / Test Descriotion 15 September 1987 S8- Semiannual Leak Check of SbBe Neutron Source 22 September 1987 S Semiannual Leak Check of PuBe Neutron Source 25-30 September 1987 V-I- Five Year Surveillance Inspection of hiechanical Integrity of Control Blade and Drive Systems (Completed Inspection begun in 1984 with hiaintenance to Correct Safety 3 Failure to Drop on Demand) 30 September 1987 S1- hicasurement of Control Blade Drop Times 30 September 1987 S 5 - hicasurement of Control Blade Controlled Insertion Times 2 October 1987 S Semiannual Inventory of Special Nuclear Material 6 October 1987 O Quarterly Radiological Survey of Restricted Areas 6/9 October 1987 S Semiannual Inventory of Security-Related Keys for UFTR and UFSA 13 October 1987 Q 2- Quarterly Check of Area and Stack Radiation hfonitors 20 October 1987 O Quarterly Radiological Survey of Unrestricted Areas 22 October 1987 A 3- AnnualMeasurementof UFTRTemperature Coefficient of Reactivity (Partial) 29 October 1987 Q Quarterly Radiological Emergency Evacuation Drill 5 November 1987 A 3- AnnualMeasurementof UFTRTemperature Coefficient of Reactivity (Completed)

 *i 6 November 1987      S1- Measurement of Control Blade Drop Times t

16 November 1987 S5- Measurement of Control Blade Controlled Insertion Times 16 November 1987 A Annual Replacement of Control Blade Clutch Current Light Bulbs V-3 i

TABLE V-1 (CONTINUED) CHRONOLOGICAL TABULATION AND DESCRIP'ITON OF SCIIEDULED UFTR SURVEILLANCES, CIIECKS AND TES'I3 Date Surveillance / Check / Test Descriotion 8 December 1987 S Semiannual Check / Replacement of Security System Satteries 9 December 1987 Q 1- Quarterly Check of Scram Functions 9 December 1987 Q Quarterly Check of Posting Requirements 21 December 1987 Q Quarterly Radiological Emergency Evacuation Drill (Large Annual DrillInvolving All Outside Agencies) 21 December 1987 A Annual Check of Emergency Call Lists 22 December 1987 S hicasurement of Argon 41 Stack Concentration 22 Decembe r 1987 S hicasurement of Stack Dilution Air Flow Rate (For-merly A-1) l 22-31 December 1987 B Biennial Evaluation of UFTR SOP hianuals for Com-pleteness 22 January 1988 Q4- Quarterly Radiological Survey of Unrestricted Areas l 28 January 1988 Q Calibratin" Check of Area and Stack Radiation Mon-i itors l 10 February 1988 Q 5- Quarterly Radiological Survey of Restricted Areas 17 24 February 1988 A 2- UFTR Nuclear Instrumentation Calibration Check and Calorimetric Heat Balance 29 February 1988 S9- Semiannual Replacement of Deep Well Pump Fuses

, 29 February 1988        S2- Annual Reactivity hicasurem:nts (Worth of Control Blades, Total Excess Reactivity, Reactivity Insertion Rate and Shutdown hfargin) (Partiai; 12 hiarch 1988         S2- Annual Reactivity hicasurements (Worth of Control Blades, Total Execss Reactivity, Reactivity Insertion Rate and Shutdown hfargin) (Partial)

V-4

TABLE V-1 (CONTINUED) CIIRONOLOGICAL TABULATION AND DESCRIIrFION OF SCIIEDULED UFTR SURVEILLANCES, CIIECKS AND TESTS Date Surveillance / Check / Test Descriotion 15 March 1988 S Semiannual Leak Check of PuBe Neutron Source 30 March 1988 S Semiannual Leak Check of SbBe Neutron Source 30 March 1988 Q 1- Quarterly Check of Scram Functions 30 March 1988 Q 6- Quarterly Check of Posting Requirements j 6 April 1988 S 3 - Semiannual Inventory of Special Nuclear Materials 615 April 1988 S 6 - Semiannual Inventory of Security Related Keys. for UFTR and UFSA 12 April 1988 Q 6- Quarterly Check of Posting Requirements l 21 April 1988 Q 2- Calibration Check of Area and Stack Radiation Mon- l itors 21 April 1988 Q 3- Quarterly Radiological Emergency Evacuation Drill 28 April 1988 Q 4- Quarterly Radiological Survey of Unrestricted Areas 28 April 1988 Q 5- Quarterly Radiological Survey of Restricted Areas 28 April 1988 S2- Annual Reactivity Measurements (Worth of Control Blades, Total Excess Reactivity, Reactivity insertion Rate and Shutdown Margin) (Completed) 29 April 1988 S7- Semiannual Check / Replacement of Security System Batteries 6 May 1988 Q 6- Quarterly Check of Posting Requirements 24 May 1988 S6- Semiannual Inventory of Security Related Keys for UFSA 13-15 June 1988 A 4- Annual Replacement of Control Blade Clutch Current Light Bulbs V-5

TABLE V-1 (CONTINUED) CHRONOLOGICAL TABULATION AND DESCRIPTION OF SCIIEDULED UFTR SURVEILLANCES, CIIECKS AND TES'13 Date Surveillance / Check / Test Descriotion 13/14 June 1988 S hicasurement of Control Blade Drop Times 15 June 1988 S5- hicasurement of Control Blade Controlled Insertion Times 17 June 1988 Q 1- Quarterly Check of Scram Functions 22 29 June 1988 B Biennial Inspection of Incore Reactor Fuel Elements 18 June 1988 A 5- Annual Check of Emergency Call Lists 26 June 1988 Q Quarterly Radiological Emergency Evacuation Drill 10/11 August 1988 Q Quarterly Radiological Survey of Unrestricted Areas 11 August 1988 Q Quarterly Radiological Survey of Restricted Areas 15 August 1988 A Annual Check of Emergency Call Lists 18 August 1988 Q 2- Calibration Check of Area and Stack Radiation hionitors 30 August 1988 S hicasurement of Argon 41 Stack Concentration 30 August 1968 S hicasurement of Stack Dilution Air Flow Rate (For-l merly A 1) l l l t V-6

TABLE V-2 CHRONOLOGICAL TABULATION OF UFTR PREVENTIVE /CORRECrlVE MAINTENANCE Date Maintenance Descriotion 8 September 1987 Replaced the failed S 2 control blade drive motor with an identical stocked spare with all S 2 control blade system checks and surveillances including withdrawal and controlled insertion times as well as drop time confirmed to be acceptable to restore the system to proper operation (8 Sep 87, htLP #87-26). l 8 September 1987 Replaced a failed capacitor (temporarily mounted) in the brush ! recorder amplifier used to perform the Control Blade Drop Time Surveillance to restore brush recorder operation for timing blade drop times via referencing to 60 hertz line voltage. l The capacitor will be permanently mounted at a later dato 6 though the temporary mounting will serve indefinitely (8 Sep 88, h!LP #87-27). 1 t 22 September 1987 Removed a broken quick disconnect connection on the primary coolant demineralizer and replaced it with an identical spare l fron' supplies on hand. Subsequent leak checks verified restaration of the demineralizer system to proper operation with no leakage detected (22 Sep 87, htLP #87 28). 25 September 1987 Performed preventive and corrective maintenance on all four (4) control blade drive motor gear assembly systems to include cleaning out congealed and hardened grease and oil from all assemblies and replacing worn bearings to restore the as-semblies to proper operation. Followi'1g completion of main-tenance operations, the necessary tests, checks and surveillances l to include bla.fe withdrawal times, controlled insertion times and blade 6 rop times as well as listening for rough operation f were completed successfully to restore proper operation to all drive motor gear assemblies. This maintenance activity is to be included in the V 1 five year surveillance of the control blade l drive system to assure prevention of future problems of im,neded operation due to hardened grease and oil (30 Sep 87, hiLP #87 29). 8 October 1987 Retermir.ated the termination connection on the diluti.1g fan tachometer to return the unit to service (8 Oct 87, hiLP #87 l 30). 1 V-7 l 1

T TABLE V-2 (CONTINUED) CIIRONOLOGICAL TABULATION OF UITR PREVENTIVE / CORRECTIVE MAINTENANCE Date Maintenance Descriotion 12 October 1987 Moved the motor mount on the dilute fan drive motor to tighten the belts for further service with no further problems noted (12 Oct 87, MLP #87 31), 12 October 1987 Added demineralized water (37.5. gallons) to the primary coolant storage tank to raise the level from 20.5 inches to the 26,0 inch level to fill the tank (12 Oct 87, hiLP #87 32). 26 October 1987 Replaced the belts on the diluting fan with stocked spares to restore the diluting fan control room indication to the normal l 510 rpm indication from the 410 rpm indication which caused l an unscheduled shutdown. Next day reverification of proper l operation showed no further problems (27 Oct 87, htLP #87-33). , 29 October 1987 Replaced the flex coupling on the diluting fan duct with work accomplished by UF technical support staff personnel under ! htWO-452726 with no furiher problems noted with the flex coupling (29 Oct 87, htLP #87-34). , 5 November 1987 Assembled and installed a screen enclosure to provide better l control of the contents of the cell low level storage area used for storage of experiments, equipment and other activated

products. Implementation of the screen enclosure was evaluated l negatively under 50.59 Evaluation No. 8716. The enclosurc had been planned since the previous reporting year and had been so indicated to NRC Inspector B. Revsin during the IIcalth l Physics Inspection in February, 1987 (15 Nov 87, htLP #87-35).

9 November 1987 Retapped and replaced the set screws in the stack dilute fan motor bearing assembly to restore proper bearing holddown to restrict axial motor shaft movement with no further problems noted (11 Nov 87, htLP 437 36). 16 November 1)S7 During the weekly checkout, the chart markings for the temperature recorder were dis, overed to be faint. Replaced the temperature recorder stamp pads to restore proper functioning of the ink marking on the recorder chart paper with no further nroblems noted (16 Nov 87, htLP #87 37). V-8 t-

TABLE V-2 (CONTINUED) CIIRONOLOGICAL TABULATION OF UFTR PREVENTIVE / CORRECTIVE hfAINTENANCE Date Maintenance Descriotion 16 November 1987 Replaced the ceramic filter and demineralizer cartridge on the shield tank demineralizer system to restore normal flow through the system (16 Nov 87, MLP #87-38). 16 November 1987 Replaced (A-4 Surveillance) all four clutch current indicating lamps after the Safety 3 bulb was discovered burned out during a weekly checkout (S 3 bulb previously burned out in May, ' 1987); per the Tech Specs, all blade drop times were measured (S 2 Surveillance), all blade controlled insertion times were measured (S 5 Surveillance) and all withdrawal times were measured to assure restoration of the reactor control system to proper operation with no further problems noted (16 Nov 87, MLP #87-39). 23 November 1987 Conducted a detailed inspection as well as a checkout and test including inspection of cylinder threads on all MSA bottles by a qualified inspector. All bottles were found to be functioning properly vith no potential regulator problems or cracks noted (24 Nov 87, MLP #87-40).  ! 30 November 1987 Performed routine preventive maintenance to replace the filters, grease the bearings and generally check out the cell and building air conditioning / air handling systems to assure proper i operation (30 Nov 87, MLP #87 41). 30 November 1987 Replaced the resins in the two demineralizers connected to the l cell city water line with fresh resins to restore availability of i i high resistivity primary coolant makeup water with no further problems noted (30 Nov 87, MLP #87-42). l i 14 December 1987 Removed the caps from the safe housing on the diluting fan l motor bearings to adjust the locknuts and washers on the hearings as part of a general service checkout as followup to  ; I

                           ?revious dilute fan motor bearing problems to assure continu-
                           'ng satisfactory operation of the dilution fan (14 Dec 87, MLP
                           #87-43).                                                         ;

i [ V9 L

l l TABLE V-2 (CONTINL ED) CHRONOLOGICAL TABULATION OF UFTR PREVENTIVE / CORRECTIVE h!AINTENANCE Date Maintenance Descriotion 14 December 1987 hionitored the motor of Safety Blade S-1 for excessive noise level during updnve and compared it with other blades. The inside mechamsm was also checked out with the cover removed with no defects noted. Though this motor is somewhat louder than the others during updrhe, no further action was con-sidered to be warranted though the updrive will be monitored for sound and changes more frequently in the future (14 Dec 87, h!LP #87-44). 17 December 1987 Cleaned the contacts on the Safety-2 blade position indicator (BPI) as well as its panel connection and tightened the DC input board from a loose position to restore proper functioning of the Safety 2 blade position indicator (17 Dec 87, hiLP #87 45). 21 December 1987 Installed a replacement high radiation level Ghi-Tube in the , teletector and completed calibration of the teletector on the l high range to assure proper response and operation of the unit ' with no further problems noted (6 Jan 88, hiLP #87-46). 4 January 1988 Removed the non. functioning Safety 2 blade position indicator (BPI) for inspection and cleaning and replaced one contact pad missing off the PC board to restore proper functioning over the full range of blade operation (4 Jan 88, h!LP #881). 8 January 1988 Replaced the card containing a failed oscillator circuit driving the stuck chopper for the linear (red) pen of the two pen recorder to restore proper response of the repaired recorder with alignment and all operating characteristics verified to be correct and unchanged from normal with no further problems noted (13 Jan 88, htLP #88 2). i 14 January 1988 Conducted heat testing on the non functioning Safety 2 blade position indicator after it began functioning upon cooldown from normal operation. Next the Regulating and Safety 2 BPI's were exchanged in place with both found to work properly so the units were returned to service after verifying proper response on all drive times and position indications on all blades for full removal and insertion (14 Jan 88, h1LP #88 3). V 10 i .-

TABLE V-2 (CONTINUED) CIIRONOLOGICAL TABULATION OF UFTR PREVENTIVE /CORRECrlVE MAINTENANCE _ Date Maintenance Descriotion 18 January 1988 Replaced the clock card with a bad connection for the Regulat-ing Blade Position Indicator from stocked spares to restore proper Reg Blade BPI response over the full range of blade motion (18 Jan 88, htLP #88-4). 18 January 1988 Added 68 gallons of demineralized water to raise the PC tank water level from 20.75" to 30" (18 Jan 88, h1LP #88-5). 2 February 1988 Spliced a new piece of wire onto the temperature recorder light j switch and reterminated it to restore proper operation of the light for the 12 point temperat.ae recorder (2 Feb 88, MLP

                        #88-6).

l 17 February 1988 Performed adjustments mads as part of the A 2 Annual UFTR l i Nuclear Instrumentation Calibration Check and Calorimetric IIcat Balance. Checks include flow meter, temperature sensors, voltage checks and adjustments made to Safs y Channels 1 and 2 based on the calorimetric heat balance so Safety Channels #1 and #2 read 100 kw at full power (24 Feb 88, MLP #88 7). 8 March 1988 Replaced the failed console analog clock with a new digital clock mounted in the console and evaluated to involve no i unreviewed safety questions per 10 CFR 50.59 Evaluation No. 3 l 88 5 (23 Feb 88, MLP #88 8). 14 March 1988 Relamped the reactor cell under MWO #53 6088 to restore full cell lighting level (19 Apr 88, MLP #8816). 14 March 1988 Replaced a failed feedback capacitor used for noise control in response to Safety Channel #1 failure where indication was noted to peg downscale and then recover slowly to the proper level after a few seconds (14 Mar 88, MLP #88 9). l f 1 L V-11 ( - - - - - - - - -

TABLE V-2 (CONTINUED) CIIRONOLOGICAL TABULATION OF UFTR PREVENTIVE / CORRECTIVE MAINTENANCE Date Maintenance Descriotion 16 March 1988 Checked all components in Safety Channel 1 in response to a full range of tests, checks and surveillances including external signals to isolate the probable cause of the transient downscale pegging and slow recovery of the Safety Channelindication to the fission chamber, the preamplifier as well as the connections and cables of Safety Channel 1. Evaluation indicated the failure may have been caused by a cable / connection problem that was fixed by the checks or there could be a problem only identifi-able and isolatable with the system operating at power (current from fission chamber) and special voltage and current monitor-ing instrumentation temporarily installed. Implemented a special test procedure to allow reactor restart in steps with the reactor declared operable following performance of checkouts and implementation of compensating features to include a second reactor operator assigned to observe the Safety Channel 1 meter during operation. The special test procedure was intended to verify proper operation of Safety Channel 1 by monitoring the voltage level in the preamplifier with response to ground, the current drawn by detector operation from the high voltage supply and the high voltage power supply output voltage while. "1e UFTR was operated at power levels in steps up to full power for an extended run to demonstrate correction of the Safety Channel 1 failure or, in the event of recurrence, to enable isolation of the fault. The result was a successful return to normal cperations on 1 April 1988 though the root cause was not isolated (1 Apr 88, MLP #8810). 24 March 1988 Cleaned the contacts on the Safety 1/ Log Calibrate Switch to eliminate previously r.oisy operation and restore stable sw:tch operation (24 Mar 80, MLP #S811). 31 March 1988 Tied and stabilized the cables for Safety Channel 1 up off the preamplifier to reduce noise and spurious signals and eliminate spurious period trip indications obtained during preoperational checks (31 Mar 88, MLP #8812). 7 April 1988 Checked out and released the stuck stack monitor needle to  ! restore proper free motion of the stack monitor needle (7 Apr i 88, MLP #8813). 1 V-12 I

TABLE V-2 (CONTINUED) CHRONOLOGICAL TABULATION OF UFTR  ; PREVENTIVE / CORRECTIVE MAINTENANCE Date hiaintenance Descriotion i 13 April 1988 Performed maintenance checks, te.,ts and monitoring in  : response to recurrence of the intermittent downscale failure of the Safety Channel 1 Indication that first occurred on 14 March 1988. Checked feedback noise suppression capacitors in both  ; Safety Channels and replaced the one in SC-2. Replaced SC-1 ' cable connectors with crimp type connectors for better contact (thought to be the source of problem) and also replaced unlabeled RG71 cable connections with Arnphenot 68175 s connectors to assure good contact; interchanged SC-1 ar.d SC-  ; 2 circuit boards to provide a means of isolating the problem l should it recur later. Finally, the special test procedure was , l im alemented to control return to power while monitoring key l po nts in the SC-1 circuit. The UFTR was returned to normal operations on 28 April 1988 with only an extra staff member assigned to monitor Safety Channel responses for the next 10 hours of operation above 50 kw as a compensating measure with no recurrence of i the problem for the remainder of the year (28 Apr 88, MLP  !

                       # 88-14).                                                        ;

15 April 1988 Manufactured and installed a new cover on the electrical junction box in the primary equipment pit to provide proper protective covering in response to an American Nuclear  ; Insurers inspection report (15 Apr 88, MLP #8815). 25 April 1988 Replaced the smoking and failed APD motor with the tech manual recommended replaecment approved under 50.59 l Evaluation No. 5814 due to the necessity to modify the motor mount to install the motor unit to restore the APD to proper operation with no further problems noted (27 Apr 88, MLP ,

                       #88 17).

29 April 1988 Reterminated a broken wire connection on the UFR security system panel to restore the security system to proper operation  : (29 Apr 88, MLP #8818). V 13 I L

f TABLE V-2 (CON TINUED) CHRONOLOGICAL TABUIATION OF UFTR ,' PREVENTIVE / CORRECTIVE hfAINTENANCE l Date hiaintenance Description 4 hiay 1988 Installed a backup core vent sampling system to allow drawing a sample via the rabbit tube for monitoring and quantification I of cell radionuclides p-for to release in abnormal and emergen-cy conditions following NRC approval of Tech Spee Amend-ment No.17 requiring such a backup means for quantifying l such release as previoT,1y approved under 50.59 Evaluation No. 881 (4 hiay 88, htLP #8819). ( 10 hiay 1988 Repaired a split rabbit system hose discovered prior to system use to return the experimental facility to normal operation with no further problems noted as the reactor was restarted and samples irradiated (10 hiay 88, h!LP #88 20). 25 hiay 1988 Checked and reset all (drifted) portal monitor setpoints to eliminate spurious alarms and return the monitor to senice with no further problems noted (25 hfay 88, hiLP #88 21). 31 hlay 1988 Replaced the ceramic filter on the shield tank recirculation loop to restore normal full flow rate with no further problems noted (31 hiay 88, htLP #88 22). 1 June 1988 Replaced the takeup reel drive belts of the 2 pen recorder takeup reel to restore free movement to the takeup reel with no further problems noted (1 Jun 88, htLP #88 23). 7 June 1988 Replaced the ink pads in the pad cartridge of the 12 point temperature recorder to restare clear indications on the chart , printout (7 Jun 88, h!LP #88 24). l 8 June 1988 Replaced the belts and filters on the reactor cell air condition-ing unit as preventive maintenance to assure continued proper operation (8 Jun 88, htLP #88 25). l 9 June 1988 Replaced the slide wire and switching rotor contacts, cleaned the switch contacts and slide bar and oiled the print head ! bushings and wire positioning floating bearings to eliminate I premature striking and restore proper temocrature printout without smearing on the 12 point temperature recorder (9 Jun r 88, htLP #88 26). l { V-14 i

1 TABLE V-2 (CC" TINUED) CHRONOLOGICAL TABULATION OF UFTR PREVENTIVE / CORRECTIVE MAINTENANCE Date Maintenance Description 9 June 1988 Pumped the equipment pit dry, rinsed with 5 gallons of demineralized water and then repumped dry after the rupture disk was broken by operator error; added approximately 18 } gallons of demineralized water to the coolant storage tank to ) restore level to 22 3/4"; replaced the rupture disk with a stocked spare and returned the system to normal operations (9 Jun 88, h!LP #88 27). l 13 June 1988 Replaced all four clutch current bulbs following the reportable failure of the SM clutch current light bulb at power and, during the S 1 surveillance measurement of control blade drop times, j replaced the pen motor in one channel of the two channel strip chart recorder with an on hand spare from an earlier model recorder per 10 CFR 50.59 Evaluation No. 8817 to restore proper functioning of the test unit to allow successful comple-tion of the drop time measurements and return to normal operations (14 Jun 88, h!LP #SS 28). i 24 June 1988 Replaced the hose clamps with on hand spares to restore proper funedoning of the remote fuel element lifting tools i during fuel bundle surveillance inspection with no further problems noted (24 Jun 88, MLP #88 29). ( 27 June 1988 Replaced the failed 0 50 microamp meter movement on the air i particulate detector with a 0-100 microamp meter movement I per 10 CFR 50.59 Evaluation Fo. 8819; replaced the failed ' Gh! tube with an on hand spare; and installed a new plug and meter mounting. Following source calibration to assure sen-sithiry to radioactivity (beta particles), the unit was determined to be function!ng properly with no further pisolems noted (28 Jun 88, hlLP #88 30). l 5 July 1988 Performed voltage at i other checks on the secondary deep well ' l pump to establish operating characteristics and assure proper operation with no problems noted (6 Jul 88, htLP #88 31). 5 July 1988 Cleaned debris and realigned the microswitch on the portal monitor treadle to restore uninhibited operational response to individuals standing on the treadle with no further problems noted (5 Jul 88, htLP #88 32). V 15

TABLE V-2 (CONTINUED) CilRONOIMGICAL TABULATION OF UFTR PREVENTIVE / CORRECTIVE h!AINTENANCE Date hiaintenance Descriotion f 18 July 1988 Added 90 gallons of demineralized water to the primary coolant storage tank to refill the tank to the 34 inch level for normal operation of the reactor (18 Jul 88, htLP #88 33). 21 July 1988 Rolled the west v belt for the stack dilution fan back on to restore proper system operation and proper RPh! indication with no further problems noted (21 Jul 88, htLP #88 34). 22 July 1988 Installed quick disconnect valves f.i the rabbit system insertion lines per the modification package with 10 CFR 50.59 Un- ) reviewed Safety Question Evaluation No. 88-20 to provide I redundant means of assuring the system cannot ha used to insert samples when not energized (25 Jul 88, htLP #88 35). 26 July 1988 Replaced two flanges at the strainer on the cell city water line ! with new fittings and then sealed and leak tested them under normal city water pressure with no further leaks detected (26 Jul 88, htLP #88 36). 27 July 1988 Cleaned off the tops of the liquid waste holdup tanks and washed them clean to restore ea:e of accessing the tanks (27 l Jul 88, htLP #88 37). 1 August 1988 hioved the surveillance status board from the east control room wall to the south wall, replaced the chalkdust producing blackboard behind the console with an erasable marker board mounted or the east wall and painted the walls of the control I too.n. The holes drilled were evaluated under 50.59 F, valuation No. 88 22 not to involve any unreviewed safety question ' Die result is a much enhanced training emironment for the opera-l t tor (31 Aug 88, h!LI' #88 38), I 8 August 1988 Opened a University of Florida hiaintenance Work Order $6-3942 to address cleaning out limuock from the storm sewer in the west reactor lot (htWO #56 3942 remains open). l 10 August 1988 Cleaned connections and tightened the coupling on the stack dilute fan RPh! Indicator to restore proper RPhi indication at the control console following an unscheduled shutdown due to I low RPh! indication (10 Aug 88, htLP #68 39). V-16 L - - - - - - _ - - - - - - - - - -

TABLE V-2 (CONTINUED) CIIRONOLOGICAL TABULATION OF UFTR PREVENTIVE /CORRECrlVE MAINTENANCE Date Maintenance Description 12 August 1988 Attempted repair and subsequently installed and calibrated a replacement diluting fan RPM indicator with no further problems noted (15 Aug 88, MLP #88-40). 29 August 1988 Investigated several small seepage leaks of water into the reactor cell along the east cell wall where it meets the floor and then had the UF Architectural Engineer check the leakage location and discuss several possible corrective actions including reworking the building exterior and using epoxy on the interior with Maintenance Work Order No. 56 6246 assigned by Work Management to address correction of this leakage. No actual work has been performed and may be unnecessary (MLP #88-41 remains open). J MWO No. 56-3942 Remains Open From August 8,1988. MLP #88-41 (MWO No. 56-6246) Remains Open From August 29,1988. i V-17

k f VI. ClIANGES TO TECIINICAL SPECIFICATIONS, SAFIRY ANALYSIS REPORT, STANDARD OPERATINO PROCEDURES AND OTHER KEY DOCUMENTS This Chapter contains a narrative deceription and status report on the .arious changes to key UFTR license-related documents that occurred during the 1987-1988 reporting year. As l such, this Chapter provides a ready reference for the status of various license related dccuments to include Technical Specifications, Safety Analysis Report, Standard Operating ! Procedures, Emergency Plan, Security Response Plan, Reactor Operator Training Requalification and Recertification Program, HEU to LEU Conversion Documents as well as Quality Assurance Program Approval for Radioactive Material Shipments and other key documents as they are generated or changed. A. Changes to Technical Soecifications The new Technical Specifications for the UFTR were issued on August 30,1982 and officially established on September 30, 1982. Two sets of requested corrections /- changes to the Technical Specifications were submitted to the NRC during the 1982-1983 reporting period. As noted in the 19831984 Annual Report, the UFTR facility received approval for Amendment No.14 and No.15 to the UFTR Technical Specifications during that reporting year. As noted in the 19851986 Annual Report, the UFTR facility requested and received approval for Amendment No.16 to correct an error in numbering Section 3.5 which had been incorrectly numbered Section 3.4 On 11 December 1986, the stack dilute fan and the core vent fan were secured by actuation of the evacuation alarm and the evacuation alarm / core vent system ) interlocks while the stack count rate was approximately 300 cps due to a normal Argon 41 vent and stack inventory buildup established by a prior run. The automatic evacuation occurred as part of the Q 3 Quarterly Evacuation Drill scenario. Establishment of two area monitors at the high level trip setpoint initiated the core vent / diluting fan interlock with the evacuation alarm actuated as required by Technical Specifications. However, UFTR Technical Specifications in Section 3.4.3 as a limiting condition for operation states that 'the vent system shall be operated until the stack monitor indicates less than 10 counts per second;" as a result, the l actuation for the drill contstituted a potential violation of Technical Specifications on Limiting Conditions for Operation (es en though the reactor was not running) and was reported as such. At its December 19,1986 meeting the Reactor Safety Review Subcommittet required (also subsequently committed to NRC in the report letter dated December 19,1986) l that a Tech Spec change be developed on the requirement for the core vent system j operation with stack monitor count rate above 10 cps; after re-evaluation and with support on a technical basis, Section 3.4.3 was committed to be modified so that the requirement for not securing the reactor vent system above 10 cps could be cased, perhaps with only a recommendation that it not be secured above 10 cps. This change was to be based upon the lack of safety and/or radiological effects from securing the reactor vent system for short periods of time or even with a higher stack count rate. This tech spec change was intended to eliminate the conflict involved in l 1 VI-1 l 1 l

in securing the vent fan system for an actual emergency following a reactor run should such occur; this work was committed to be completed by May 30,1987. The proposed Tech Spec change (Amendment 17) as submitted finally lnvolved a [ complete reorganization of Sections 3.3 and 3.4 of the Tech Specs into a format to I match the remainder of the UFTR Tech Specs so that currently mixed and/or missing elements would be contained in the proper Section (either 3.3 or 3.4) plus incorporation of several minor changes along with the easing of the requirement that the vent system shall be operated anytime the stack count rate is not less than 10 cps. l A brief summary will clarify the proposed Amendment 17. First, Amendme't '7 provided logical reorganization of Sections 3.3 and 3.4 of the UFTR Tech ' corJorm with the remainder of the existing Tech Specs where cach Sectio., < introduction, a listing of specifications and finally a set of bases to support the :is specifications. Second, these changes were to provide better defined, consistent b: . for the Technical Specifications on the Reactor Vent Systcm (Section 3.3.2 l augmented) and addition of previously lacking bases for the Technical Specifications on the Radiation Monitoring Systems and Radioactive Effluents (Section 3.4.7).The purpose of the substantive che.nge in Section 3.4.3 is to allow securing the core vent { fan when necessary without necessarily violating the Tech Specs. With this amendment, if the Reactor Vent System is secured, as it must and should { be for a valid emergency condition or a system failure, the event is not necessariy a violation of the Technical Specifications simply because the vem system was secured at > 10 eps. Otherwise, the content and intentions of the Tech Specs were i l not considered to be changed by this Amendment. l The proposed Tech Spec change (License Amendment No.17) with supporting { information and calculations was submitted to NRC with a letter dated June 2,1987. I A response dated February 5,1988 was finally received on February 8,1988  ; indicating two areas remaining to be clarified. The UFTR licensee was requested to  ! l revise th: Tech Specs to include the following two areas: l

a. Listing items for exception in TS 3.3.2 Paragraph I when the reactor vent system can be secured with the stack count above 10 cps,

b. Addressing provisions for controlled release of radioactive effluents to the environment during abnormal operating conditions.

The UFTR response to the NRC request that the additional arcas be addressed was submitted with a letter dated March 7,1988. As requested, this submittal presented and explained reasons for each of the exceptions allowing securing of the reactor vent system with stack count rate above 10 eps. The submittal also incorporated provisions into Tech Spec See: ion 3.3.1 allowing the controlled release of radioactivity effluents to the emironment during abnormal and emergency operating conditions and into Section 3.4.3 requiring that radioactivity in the effluent be quantified prior to initiating controlled venting whenever such venting is to be used to reduce cell radionuclide concentrations in addressing unlikely though possible VI-2

i { abnormal or emergency conditions involving high concentrations of airborne j radioactivity within ALARA guidelines. The approved license (Tech Spec) Amendment 17 was finally received on May 3, i 1988 per a letter from NRC dated April 27,1988. The approved Amendment 17 i corresponds exactly to the second license submitted on March 7,1988. The amendment consists of a revision to the Tech Specs to permit conducting certain activities when the reactor is shutdown, the reacto*: vent system is secured and the stack monitor is reading greater than 10 eps. These permitted activities have not yet been incorporated into UFTR Standard Operating Procedures but the work is in progress. As requested by NRC and submitted by the licensee, the Tech Specs were also revised to include a backup means for quantifying the radioactivity in the effluent during abnormal or emergency operating conditions in addition to administrative changes. Under Maintenance Log Page #8819, the backup core vent sampling system was installed on May 4,1988 into the rabbit system line per 10 CFR 50.59 Evaluation I Number 881 with availability for all subsequent reactor operations. The complete package sent by NRC also contains Amendment 17 pages (photstat copies of the submitted text) as well as the Safety Evaluation Report supporting Amendment No. l 17.The entire package received from NRC is contained in Appendix C of this report. No further requests for changes in the approved Tech Specs are anticipated for the { operation of the UFTR with its present h gh enriched fuel at a rated power level of 100 kWth. It is expected, however, that another substantive amendment to the Technical Specifications will be required before the UFTR can be converted from utilizing high enriched MTR plate-type fuel to utilizing low enriched SPERT pin-type or silicide plate type fuel. A decision will be made early in the upcoming i reporting year as DOE support to analyze conversion options became available in  ! ( November,1987. B. Revisions to UFTR Saferv Analysis Rcoort Revision 4 to the UFTR Safety Analysis Report, after review and approval by the RSRS under 10 CFR 50.59 Evaluation and Determination #87-13, was submitted to NRC in Washington with a copy to Region II with a letter dated September 25,1987. Revision 4 is not considered to involve any unreviewed safety questions and has been inserted into the official copies of the UFTR Safety Analysis Report. A complete copy of the entire submittal for UFTR SAR Revision 4 is contained in Appendix D. The Revision 4 alters the Safety Analysis Report Section 9.5.1 Fire Protection System on Page 910 to correct the number of CO, fire extinguishers indicated to be available in the cell and on Page 913 to describe the new improved four zone automatic fire alarm system. The change addresses the minimum claimed installed equipment for the new fire alarm system which replaced a two zone system and was installed per recommendations resulting from inspections by the American Nuclear Insurers. ne new monitoring station is located outside the reactor building (Licensee site) adjacent to the Emergency Response Center used for addressing radiological, VI 3

_1 fire and other building emergencies to provide optimal resp' onse to emergency conditions and their proper evaluation. Revision 4 to the UFTR Safety Analysis Report also alters Section 9.5.2 Communica-tions Systems on Page 913 to describe UFTR personnel positions by proper titles and to indicate what direct communications exist to the control room from other building locations including the llealth Physics office which is no longer in the UFTR building. This change actually predates the new SAR submittal (1981 and i license renewal (1982). Several outdated claims of intercom connectio deleted. This change is also not considered to affect Health Physics capabilities or response and is not considered to involved any unreviewed safety questions. Revision 5 to the UFTR Safety Analysis Report was submitted to NRC in Washington with a copy to NRC Region II with a letter dated June 30, 1988. Revision 5 was reviewed and not considered to involve any unreviewed safety questions or to impact the UFTR Safety Analysis; it has also been inserted into the official copies of the UFTR Safety Analysis Report. Changes here were the result of ongoing reviews of the UFTR Safety Analysis Report to assure updated contents. This revision was in progress when the NRC Operator Licensing Examiner J. Arildsen noted several minor typographical errors and the outdated control blade integral reactivity worth curves in his exam preparations using the FSAR Changed pages include Page 14 updating descriptions of experimental facilities and control blade integral reactivity worths, Page 15 correcting typographical errors and providing better descriptions of the reactor vent system, Page 3 6 correcting typographical errors and indicating the correct unchanged location of the emergency , personnel exit in the cell freight door, Page 4 9 (Table 41) updating UFTR operating characteristics and correcting several typographical errors, Page 71 reflecting instrumentation operation in the UFTR console and updating the list of control and indicating instrumentation to reflect changes previously reviewed and implemented, Page 91 correcting a typographical error for the crane capacity and Page 15 2 correcting several typographical errors including a sentence repeated twice. Most of these changes correct obvious typographical errors, text inconsistencies, or minor changes in current operating characteristics. Ilowever, for the changes on Page 71, Section 7.2.1 is changed to reflect instrumentation operation in the UFIR console as it has been in place prior to relicense submission in 1981 as well as additions to provide contral of rabbit system energization and communications with the rabbit system operator added since relicensing. Also on Page 71, Section 7.2.1, three items are added to the list of console instrumentation to include a digital clock replacing a previously installed analog clock per 10 CFR 50.59 Evaluation No. 88 5, a PuBe source alarm indicator present for over 10 years in response to a commitment to NRC, and the energization switch and communication line for the pneumatic operated rapid sample insertion (rabbit) system. A copy of the complete FSAR Revision 5 package submitted to NRC is contained in Appendix E. Further updating changes to FSAR Chapter 11 are in progress as the l review and updating of the UFTR Safety Analysis Report is a continuing effort, vs i .

C. Generation of New Standard Operating Procedures Only one new Standard Operating Procedure was generated during the current reporting year. UFTR SOP F.8, "UFTR Safeguards Reporting Requirements" was generated to delineate the requirements for reporting of safeguards events to the NRC fer the UFTR in response to new regulations in 10 CFR 70 and 10 CFR 73. Items addressed include safeguards events that must be reported to the NRC, designation of how communications are to be made to NRC concerning safeguards events and specification of time intervals for telephone communication and submittal of licensee written reports for applicable safeguards events. Key features of SOP-F.8 include UFTR Form SOP F.8A to document safeguards related telephone communications and UFTR Form SOP F.8B as the log of UFTR safeguards events required to be maintained at the facility and submitted quarterly to NRC when it contains new entries. I Since this procedure is newly generated during the latest reporting year, the full text ) of UFTR SOP F.8 (UFTR Safeguards Reporting Requirements) is contained as currently implemented in Appendix F for reference purposes and to meet Tech Spec requirements for such submissions. It is expected that procedure review and upgrading in response to the NRC Inspection of the Radiation Protection Program conducted on March 14 17, 1988, will result in at least two new procedures to control UFTR radioactive material transfers and to control utilization of the rabbit system as well as a major revision of UFTR SOP A.5 (Experiments) used to control and document review of experiments run in the UFTR. These procedures are nearing final form after several internal reviews during the current reporting year. D. Revisions to Standard Operating Procedures All existing UFTR Standard Operating Procedures were reviewed and rewritten into a standard format during the 198219S3 reporting period as required by a commitment to NRC following an inspection during that year. As committed to NRC, the final approved version of each SOP (except security response procedures which are handled separately)is permanently stored in a word processor to facilitate revisions and updates which are incorporated on a continuing basis in the standard format. l Table VI 1 contains a complete list of the approved UlTR Standard Operating Procedures as they existed at the end of the previous (19S6-1987) reportmg year l exclusive of applicable temporary change notices (TCNs) since these do not change procedure intent. Table VI 2 contains a similar complete up to date list of the approved Standard Operating Procedures as they exist at the end of the current l reporting year. The latest revision number and date for each non security (not withheld from public disclosure) related procedure is listed in Table VI 2. The latest revision number and date is in parentheses for each SOP; temporary change notices [ (TCNs) refer to minor changes made to an SOP in lieu of a full revision and are not noted on the two tables to simplify the presentation. A comparison of Tables VI 1 VI-5 1 _ _ _ _ _ _ _ _ _

{ and VI 2 indicates that there were no revisions to SOPS generated during this reporting year and only one new procedure (UFTR SOP F.8) as discussed in Section [ C of this Chapter was generated. I During the 19871988 reporting year, a number of minor changes were incorporated I into the UFTR Standard Operating Procedures as needs and/or errors were identified. ' Temporary Change Notices" were issued to correct minor discrepancies or better express the unchanged intent of thirteen (13) different procedures, some several times to include SOP 0.2, SOP O.5, SOP O.7, SOP A.1, SOP-A.2, SOP A 5, SOP A.6, SOP C.3, SOP D.1, SOP.D.4, SOP E.2, SOP E.6 and SOP E.7. It should be noted that the temporary change notices for SOP O.5 implemented, among other things; increased frequencies for several surveillances to assure the emergency call lists are checked semiannually (S 10) as required by the Emergency Plan (an evaluation showed the check of the emergency calllists was already being conducted semiannually) versus annually (A 5) and that the control blade clutch current light bulbs are replaced semiannually (S 11) versus annually (A4) following several bulb failures. Following an NRC Inspection of the Radintion Protection Program conducted on March 1417,1988 and as part of NRC Inspection Report No. 50-83/88 01 dated April 7,1988, the facility licensee was cited for a Seve;ity Level IV violation relative ' to use of SOP E.6, Argon 41 Concentration Measurements." As a result a Temporary Change Notice was implemented for SOP E.6 to clarify allowable I calibrated source usage and to make several other minor corrections not changing the intent of the SOP. The remaining Temporary Change Notices all involved relatively minor changes affecting one or a few sections of the respective 50Ps. All were fully reviewed by UFTR facility management and approved by the RSRS. Because of the quantity of paper involved and the relatively minor nature of Temporary Change Notices, copies of these SOP changes or the SOPS as currently revised and implemented are not included in this report. A copy of each may, however, be obtained directly from the UFTR facility if desired. E. Revisions to UFTR Emergency Plan With a letter dated March 3,1937, two revisions to the approved UFTR Emergency l Plan were submitted to the NRC during the previous reporting year. Both have been implemented since they had been reviewed by UFTR management and the Reactor l Safety Review Subcommittee to assure they did not decrease the effectiveness of the i Plan. l i Revision one consisted of individual page changes in the body of the Plan (pages l 5-2,81,8 3 and 84) to correct a typographical error on Page 5 2, to correct a location description on Page 8-1, to correct the function description of the decontamination facilities on Page 8-3 and finally to correct the description i.f the communications equipment at the Emergency Response Center on Page 84. VI 6 L, .

ne second resision consisted of changes in the Emergency Re,ponse Procedures contained in Appendix 111 of the Plan. First, Page 9 of SOP B.1, "Radiological Emergency"is changed to reficct a pager number for the Radiation Control Officer. l Second, SOP B.2, "Emergency Procedure - Fire" was rewritten to nssure better classification and response to fire events though its intent was not changed. At the end of the reporting year, no response has been received from NRC on this submittal which is not considered to involve any unreviewed safety question. No new revisions were submitted this year, though several are expected to be submitted l during the upcoming year, l F. Biennial Reactor Onerator Reaualifbation and Recett;fication Prneram i De existing approved biennial reactor operator requalifiestion and recertification

      . program expired at the end of June,1987. Therefore, a new program was submitted to NRC with a letter dated May 26,198'i, to cover the July,1987 through June,1989        i period. The new program had only minor changes (additions) from the previous f        program so it has served as an upgraded program. However, the new 10 CFR Part 55 (Operator's Licenses) became effective on May 26,1987 so the requirement that alllicensed personnel exercise the RO/SRO license for a minimum of 4 hours of I        !! censed activities during each calendar quarter has involved additional administrative I

time as this requirement is now being tracked on training forms. During the current reporting year, an upgraded resision of the UFTR Operator Requalification and Recertification Program Plan to be good through June,19S9 was submitted to NRC with a letter dated August 19, 1988. The revised Program plan reflects the new requirements (and NRC's interpretation of these) in 10 CFR 55 for a comprehensive examination once every two years and operations test every year. These two changes were reviewed by the Reactor Safety and Review Subcommittee and are not considered to rec} aire NRC approval, especially since they clearly upgraded the Program. Otherwise the Program remains essentially the same as that previously submitted in May,1987, in the Revised Program Plan, the annual operations test is , scheduled for December,1988 while the Biennial Comprehensive Examination it scheduled for June,1989 - the last month of the current Requalification Program Plan whereupon the existing UFTR Operator Requalification and Recertification Program Plan will be resubmitted for the next two year cycle. At the end of the reporting year no response has been received to either the May, 1987 or the August,1988 submittals as the facility continues to follow the Psogram Plan as upgraded until notified to do otherwise. Consideration is also given to the r new 10 CFR 55 (Operator's Ucenses) for any specific requirements as they relate ( to the UFTR Requalification and Recertification Program. l G. IIEU to LEU Fuel Comersion Documents The orig!nal proposal submitted to NRC to meet 10 CFR 50.64 requirements for l scheduling 12TR conversion from llEU to LEU fuel was accepted as meeting the legal requirements for submission in March,1987 of the previous reporting year. VI 7 ( _ _ _ _ _ _ _ _ _ _ _ _ . .-

( llowever,in a letter dated April 17,1987 and recclued on April 22,1987, the NRC claimed the scheduled span of time from receipt of funding to submittal of our ( application to convert was too long. The updated (reduced) schedule (Revision 1) showing a reduction of 8 months as presented in Table VI 3 was then submitted to NRC licensing in Washington with a cover letter dated hiay 14,1987. No further ( response was received to this submittal which was considered acceptable. During this reporting year, a new proposal updating the UFTR conversion schedule and work status per 10 CFR 50.64(b)(2) requirements was submitted to NRC with a letter { dated hiarch 22, 1988 to meet the annual hiarch 27, 1988 deadline for such submission with no subsequent response from NRC during the remainder of the year. This new schedule (Revision 2) is presented as Table VI 4 of this report and shows the schedule lengthened aaproximately two (2) months compared with Revision 1 which assumed receipt of funding on September,1987. The proposal for financial support of UFTR conversion from IIEU to LEU fuel was submitted to the Department of Energy with a letter dated August 7,1987. Official notice of funding for the first two years to support submisalon to NRC of the license amendment documentation for conversion was received on November 24 and effective Noveraber 15,1987; however, the de.cription of work was incorrect. A new grant description of work was finally received on December 29,1987 when the grant j document was signed for record purposes per conversations with Keith Brown at EG&O Idaho and hfartha Lyle, DOE Oak Ridge. Since recching funding, work has been proceeding as cuickly as possible though a shortage of graduate students to perform the neutronic and other analyses have caused this work to lag. In addition, because of extensive efforts to decontaminate and remodel a room in which to store the SPERT LEU fuel, to change the license description of the SPERT storage facility, to move the fuel to the new facility, to release the previous storage room to unrestricted usage, to revise the facility security plan (SNht 1050) and then to aerform a detailed pin by pin sisual inspection and verification of serial numbers, tae conversion analysis is lagging behind the schedule submitted to NRC in h! arch,1988. I At the end of the reporting year, the sisualinspection of pins is nearly complete with I X radiography scheduled to be performed early in the next reporting year so a 1 decision can be made on whether to proceed with the lieu to LEU conversion I analyses for the UFTR using SPERT 4.8% enriched UO, fuel pins or 19.8% enriched alumimtm silicide plates. It is expected that the delays in implementing all aspects of the conversion work funded by DOE will impact the Revision 2 schedule presented in Table VI-4 so tF t the schedule submittal required in htarch,1989 per 10 CFR 50.64(b)(2) as Revision 3 will likely show a further schedule slippage from Resision 2.

11. Quality Assurance Program Approval For Radioactive hf aterial Packagg l During the middle of the reporting year plans were being made by the University of Florida to ship ~1200 SPERT fuel pins held under the SNht 1050 license to Oak Ridge National Laboratory (ORNL). Since ORNL wanted the University of Florida VI S I

( to be the shipper of record, an approved Quality Assurance Program was needed r with the University to be responsible to see that the shipment would meet all 10 L CFR 71 requirements. ORNL was planning to have these pins shipped in 6M Type drums on which they will have performed the necessary enticality calculations.The initial request for OA Program approval to ship SPERT F-1 LEU fuel pins was ( submitted to NRC with a letter dated September 2,1987; a resubmittal deleting the requirement that it be withheld from public disclosure was transmitted with a letter dated September 17, 1987. NRC Ouality Assurance Program Approval for Radioactive Materials Packages No. 0578, Revision No. I with an expiration date of October 31,1992 and dated November 5,1987 was received on November 9,1987 and is contained in Appendix G along with the QA Program submittal. Because of a forced shutdown of the Oak Ridge reactor in which the SPERT pins were to be used for an experiment, plans to ship this fuel are in abeyance at the years end. Nevertheless, there is a likely possibility that Oak Ridge National I.aboratory will want to reclaim these pins upon restart of their reactor facility expected in late 1988. Even if some of the pins are not wanted by ORNI, the OA Program approval will also allow transfer shipment of the SPERT fuel to the UFTR facility onto the amended R 56 license from the SNM 1050 license if the decision is made to use the SPERT fuel for the llEU to LEU conversion which is not yet clear at the end of the reporting year. i l l VI 9 l l ..

TABLE VI 1 LISTING OF APPROVED UITR STANDARD OPERATING PROCEDURES (August 31,1987). O. ADMINISTRATIVE CONTROL PROCEDURES O.1 Operating Document Controls (REV 1,5/87) 0.2 Control of Maintenance (REV 4,5/87) } O3 Control and Documentation of UFTR Modifications (REV 0,10/85) l O.4 10 CFR 50.59 Evaluation and Determination (REV 1,5/86) O.5 UITR Quality Assurance Program (REV 1,2/86) O.6 Reactor Trip and Unscheduled Shutdown Review and Evaluation (REV 0, S/87) 0.7 Control of NRC 10 CFR 50 Written Communications Requirements (REV 0, 7/87) 0.8 Operator Licensing Requalification Examination Controls (REV 0,8/87) A. ROUTINE OPERATING PROCEDURES A.1 Pre-Operational Checks (REV 13,6/85) l A.2 Reactor Startup (REV 12,5/87) l A3 Reactor Operation at Power (REV 11, S/87) A.4 Reactor Shutdown (REV 9,6/85) A.5 Experiments (REV 3,4/83) A.6 Operation of Secondary Cooling Water (REV 1,10/S3) A.7 Determination of Control Blade Integral or Differential Reactivity Worth (REV 1,6/85) B. EMERGENCY PROCEDURES B.1 Radiological Emergency (REV 3,5/83) B.2 Fire (REV 5,5/85) ) B.3 'Ihreat to the Reactor Facility (Superseded by F Series Procedures) } B.4 Flood (REV 1,4/83) , C FUEL IIANDLING PROCEDURES l C1 Irradiated Fuel Handling (REV 4,2/85) , C2 Fuel Loading (REV 4,4/83) } C3 Fuel Inventory Procedure (REV 3,2/85) C4 Assembly and Disassembly of Irradiated Fuel Elements (REV 0,9/84) f ) l l VI 10 L _ _ - - - - - - - - - - -

TABLE VI-1 (CONTINUED) LISTING OF APPROVED UFTR STANDARD OPERATING PROCEDURES (August 31,1987) D. RADIATION CONTROL PROCEDURES D.1 UFTR Radiatfor Protection and Control (REV 3,1/83) D.2 Radiation Work Permit (REV 10,3/87) D3 Primary Equipment Pit Entry (REV 2,5/85) D.4 Removmg Irradiated Samples From UFTR Experimental Ports (REV 3,5/85) D.5 UFTR Reactor Waste Shipments: Preparations and Transfer (REV 0,5/87) E. MAINTENANCE PROCEDURES E.1 Changing Primary Purification Demineralizer Resins (REV 3,6/83) E.2 Alterations to Reactor Shielding and Graphite Configuration (REV 3,5/87) E3 Shield Tank and Shield Tank Recirculation System hiaintenance (REV 2, 4/83) E.4 Superseded E.5 Superseded E.6 Argon-41 Concentration hicasurement (REV 0,1/84) E.7 hicasurement of Temperature Coefficient of Reactivity (REV 0,5/85) I E.8 Verification of UFTR Negat ive Void Ccefficient of Reactivity (REV 0,12/85) F. SECURITY PIAN RESPONSE PROCEDURES (Reactor Safeguards hiaterial, Disposition Restricted) F.1 Physical Security Controls (Confidential, except for UFTR Form SOP F.1A) F.2 Bomb nreat (Confidential, except for UFTR Form SOP F.2A) F3 Theft of (or Dreat of the Theft of) Special Nuclear hiaterial (Confidential, except for UFTR Form SOP F3A) F.4 Civil Disorder (Confidential) F.5 Fire or Explosion (Confidential) f F.6 Industrial Sabotage (Confidential) F7 Security Procedure Controls (REV 1,9/84) l l l VI-11 t - - - - - - - - - - - - - - - - -

s TABLE VI 2 LISTING OF APPROVED UFFR STANDARD OPERATING PROCEDURES (August 31,1988) O. ADMINISTRATIVE CONTROL PROCEDURES 1 O.1 Operating Document Controls (REV 1,5/87) 0.2 Control of Maintenance (REV 4,5/87) O.3 Control and Documentation of UFTR Modifications (REV 0,10/85) O.4 10 CFR 50.59 Evaluation and Determination (REV 1, S/86) O.5 UFTR Ouality Assurance Program (REV 1,2/86) O.6 Reactor Trip and Unscheduled Shutdown Review and Evaluation (REV 0, 5/87) O.7 Control of NRC 10 CFR 50 Written Communications Requirements (REV 0, 7/87) O.8 Operator Licensing Requalification Examination Controls (REV 0,8/87) A. ROUTINE OPERATING PROCEDURES A.1 Pre Operational Checks (REV 13,6/85) l A.2 Reactor Startup (REV 12,5/87) ) A.3 Reactor Operation at Power (REV 11,5/87) A.4 Reactor Shutdown (REV 9,6/85) A.5 Experiments (REV 3,4/83) A.6 Operation of Secondary Cooling Water (REV 1,10/83) A.7 Determination of Control Blade Integral or Dif.erential Reactivity Worth (REV 1,6/85) B. EMERGENCY PROCEDURES B.1 Radiological Emergency (REV 3,5/83) l B.2 Fire (REV 5,5/85) l B.3 Threat to the Reactor Facility (Superseded by F Series Procedures) l B.4 Flood (REV 1,4/83) C. FUEL IIANDLING PROCEDURES l C.1 Irradiated Fuel Handling (REV 4,2/85) l C.2 Fuel Loading (REV 4,4/83) C.3 Fuel Inventory Procedure (REV 3,2/85) { C.4 Assembly and Disassembly of irradiated Fuel Elements (REV 0,9/84) VI 12 i __ - - - - - - - - - - - - - - -

TABLE VI 2 (CONTINUED) LISTING OF APPROVED UFTR STANDARD OPERATING PROCEDURES (August 31,1988) l \ D. RADIATION COKIROL PROCEDURES D.1 UFTR Radiation Protection and Control (REV 3,1/83) D.2 Radiation Work Permit (REV 10,3/87) D.3 Primary Equipment Pit Entry (REV 2,5/85) D.4 Removing Irradiated Samples From UFTR Experimental Ports (REV 3,5/85) D.5 UFTR Reactor Waste Shipments: Preparations and Transfer (REV 0,5/87) E. MAINTENANCE PROCEDURES E.1 Changing Primary Purification Demineralizer Resins (REV 3,6/85) F.2 Alterations to Reactor Shielding and Graphite Configuration (REV 3,5/87) E.3 Shield Tank and Shield Tank Recirculation System hiaintenance (REV 2, 4/83) E.4 Superseded E.5 Superseded E.6 Argen 41 Concentration hicasurement (REV 0,1/84) E.7 hicasurement of Temperature Coefficien: of Reactivity (REV 0,5/85) , E.8 Verification of UFTR Negative Void Coefficient of Reactivity (REV 0,12/85) ! l F. SECURITY PLAN RESPONSE PROCEDURES (Reactor Safeguards hiaterial, Disposition Restricted) l F.1 Physical Security Controls (Confidential, except for UFTR Form SOP F.1A) P.2 Bomb nreat (Confidential, except for UFTR Form SOP F.2A) F.3 Reft of (or Dreat of the Deft of) Special Nuclear hiaterial (Confidential, except for UFTR Form SOP F.3A) F.4 Civil Disorder (Confidential) F.5 Fire or Explosion (Confidential) F.6 Industrial Sabotage (Confidential) F.7 Security Procedure Controls (REV 1,9/84) l ( F.8 UPTR Safeguards heporting Requirements (REV 0,9/87) l VI-13

k TABLE VI 3 TABLE II (Revision 1) UNIVERSITY OF FLORIDA TRAINING REACTOR TENTATIVE MILESTONE SCIIEDULE FOR IIEU TO LEU FUEL CONVERSION

1. Date of Receipt of Funding (expected) September 30.1987 II. Date of Full Submittal to NRC of Application to Convert (including all necessary documents) October,1989 III. Date of NRC Order to Convert February,1990 A. Date of Completion of All Plans to Convert September,1990 i B. Date of Receipt of LEU Fuel November,1990 C. Date of Completion of Any Final Tests With IIEU Fuel January,1991 D. Date of Removal of IIEU Fuel March,1991 E. Date of Shipment of HEU Fuel June,1991 F. Date of Loading of LEU Fuel August,1991 i

e G. Date of Completion of Determination of Initial l Operational Parameters With LEU (Startup and Power Operations Testing) October,1991 II. Date of Submittal of Report to NRC/ DOE Summarizing l New Operational Characteristics and Comparing With Predictions of Safety Analysis January,1992 VI 14

TABLE VI 4 TABIE II (Revision 2) UNIVERSITY OF FLORIDA TRAINING REACTOR TENTA*11VE MILESTONE SCIIEDULE FOR IIEU TO I EU FUF.L CONVERSION I. Effective Date of Receipt of Funding Novamber,1987 } II. Date of Full Submittal to NRC of Application to Convert (including all necessary documents) December,1989 111. Date of NRC Order to Convert April,1990 A. Date of Completion of All Plans to Convert November,1990 1 B. Date of Receipt of LEU Fuel January,1991 C. Date of Completion of Any Final Tests With IIEU Fuel March,1991 D. Date of Removal of lieu Fuel May,1991 ) } E. Date of Shipment of HEU Fuel August,1991 i F. Date of Loading of LEU Fuel October,1991

G. Date of Completion of Determination of Initial Operational Parameters With LEU (Startup and Power Operations Testing) December,1991 l

        } {. Date of Submittal of Report to NRC/ DOE Summarizing New Operational Characteristics and Comparing l               With Predictions of Safety Analysis                    M-   't,1992 l

VI-15

I VII. RADIOACTIVE RELEASES AND ENVIRONMENTAL SURVEILIANCE Dis chapter summarizes the gaseous, liquid and solid radioactive releases from the UFTR facility for this reporting year. Argon 41 is the primary gaseous relcase while there were ! several low level liquid releases :.nd no solid release at all. Finally, this chapter includes a J summary of personnel exposures at the UFTR facility. A. Gaseous (Argon-41) The gaseous releases from the UFTR Facility for this reporting year are summarized in Table VII 1. The basis for the gaseous activity relcue values is indicated in Table VII.

2. These values are obtained by periodic measurements of stack concentrations as required by Technical Specifications following UFTR SOP E.6, "Argon 41 Concentration Measure-ment."

TABLE VII 1 UFTR GASEOUS RELEASE

SUMMARY

Month Release Month 1v Average Concentration September,1987 22.6S x 105 pCi/ Month 7.137 x 10 pCi/ml October,1987 12.26 x 10' pCi/ Month 3.858 x 10 pCi/ml 5 November,1987 10.65 x 10 .nci/ Month 3.352 x 10 pCi/ml December,1987 13.42 x 10' pCi/ Month 4.160 x 10 pCi/ml January,1988 9.639 x 10' pCi/ Month 2.988 x 10 pCi/ml February,1988 9.520 x 10' pCi/ Month 2.951 x 10 pCi/ml March,1988 3.258 x 10' pCi/ Month 1.010 x 10 pCi/ml l April,1988 5.5S7 x 10' pCi/ Month 1.732 x 10 pCi/ml May,1988 7.149 x 105 pCi/ Month 2.216 x 10 pCi/ml June,1988 6.182 x 105 pCi/ Month 1.916 x 10 pCi/ml { July,1988 16.10 x 10' pCi/ Month 4.989 x 10 pCi/ml August,1988 21.08 x 10' pCi/ Month 6.490 x 10 pCi/ml TOTAL ARGON-41 Releases for the Reporting Year: 137.80 Ci YEARLY AVERAGE ARGON-41 Release Concentration: 3.57 x 10 pCi/ml Vil-1

UFTR Technical Specifications require average Argon 41 release concentration averaged over a month to be less than 4.0 x 10'8 pCi/ml. All such monthly values are well below this limiting release concentration and the average monthly release concentration of 3.57 x 104 pCi/ml is more than an order of magnitude below the limiting value. f Total releases and average monthly concentrations are based upon periodic Argon-41 release concentration measurements made at equilibrium full power (100 kw) conditions. The results for these experimental measurements used in calculating the gaseous Ar-41 release data are summarized in Table VII.2. Entries in Table VII 2 represent the average results of analyses of a minimum of three (3) samples per UFTR SOP.E.6. TABLE Vil 2 UFTR GASEOUS RELEASE DATA BASE Releases Per Unit Instantaneous Argon 41 Month Ener_ev Generation Coneentration at Full Power' , Sept.1987 Nov.1987 53S7.2 pCi/kw br 12.2 x 10 pCi/ml Dec.1987 - June 1988 5060.45 pCi/kw.hr 11.3 x 10.ap Ci/ml July 1988 - Aug.1988 5005.21 pCi/kw hr 11.1 x 10-a pCi/ml

t. Values used to assure average release concentration meets t0 CFR 20 limits.

l l B. 1.iquid Waste From the UFTR/ Nuclear Sciences Comolex l Dere were approximately 617,280 liters discharged from the liquid waste holdop tanks to the campus sanitary sewage system during this reporting period. For this period i there were batch discharges as summarized in Table VII 3. 1 The effluent discharged into the holding tanks com:s from tv :nty laboratories within l the Nuclear Sciences Center, the University Radiation Cc,ntrol Office as well as the UITR l complex.The UFTR normally releases approximately 1 liter of primary coolant per week to the holdup tanks as waste from primary coolant sampling. A total of 52 weekly samples the average activity for these coolant samples was l were taken 1.3 x 10 pCi/mlduring (#1) andthis 1.1 yreporting 10' yeaj:pCi/ml (a) for this 19S7-1938 reporting per l ne only other primary coolant sample released to the holdup tanks during the I reporting year was approximately 0.2 liters as a result of the broken ru occurred on 9 June 1988, ne total activity of this sample was 4.0#1 pCi/ml x 10'pture and disk Iwhich i 2.15 x 10 pCi/ml (a). The remaining 68 liters (18 gal ons) of primary coolant was then l l held for decay in the cell until the radioactive materials concentration was at background prior to being discharged to the waste holdup tanks. j I There were no other primary coolant samples removed for analysis or as a result of failures or maintenance during the 1987-1988 reporting period. Vil-2

i { TABLE Vil 3 l LIQUID WASTE RELEASES FROM llOLDUP TANKS Volume Concentration' Total Release Date (liters) (uCl/ml) Activity (uCl) September 14, 1987 78,585 5.1 x 10.s 4.0 September 21,1987 95,395 <LLD(3.72 x 10'8) 3.5' j February 2,1988 93,300 7.74 x 10 O.7 June 10,1988 84,500 1.1 x 10.s 0.9 8 July 11,1988 88,500 < LLD(2.69 x 10) 0.24 July 29,1988 88,500 2.59 x 10 O.23 l August 16,1988 88,500 2.82 x 10 O.25 l

t. The reported actisity concentrations are based on gross beta actisity determinations. Aetisity levets for tritium and carbon 14 are not included in the gross beta values; however, these concentrations were determined separately to be less than 0.2% of the allowable htPC for release to the sanitary sewer l system for all releases.

i 2. The actisity was determined for these entries using the LLD. Actual activity released in these cases is l less than this value.

      ~

( C. Solid Waste Shipped Offsite De UFTR facility made no shipments of solid wuste during this reporting year.De last shipment was made on December 10,1985 through ADCO Senices, Inc. and consisted on one 55 gallon drum containing radioactive scrap metal parts as well as paper, plastic and other reactor related waste materials associated primarily with the work to restore proper functioning of the UFTR control blade drive systems. De actisity of the shipment was approximately 3.125 curies with the activity primarily attributed to Cobalt 60. Dough a similar shipment of two drums was planned for the last reporting year and again this reporting year to remove all the products resulting from the control blade restoration and maintenance project of 19851986, this shipment has not occurred to date. No date has been set for this next shipment though it is expected to occur sometime during the next reporting year as waste from several other small maintenance projects is consolidated for shipment to clear space for waste expected to be generated during the UFTR conversion from IIEU. to LEU fuel expected with 3-4 years.The new Standard Operating Procedure UFTR SOP-D.5,"UFTR Reactor Waste Shipments: Preparations and Transfer" generated in the 19S6-1987 reporting year will be used to assure proper control of the waste for shipment. Vll-3

D. Environmental Monit IIing The UFTR maintains continuoi s film badge as well as thermoluminescent dosimeter monitoring (new for the 19821983 reporting period) in areas adjacent to and in the vicinity of the UFTR complex. The badge and TLD cumulative totals for this reporting period from September,1987 through August,1988 are summarized in Table Vil 4. As can be noted, the values for the 12 months of the reporting period are either minimal or very low j in all cases. Overall, the values in Table VII 4 show minimal environmental radiation dose from UFTR operations. All yearly exposures recorded via TLD's are zero while those recorded via film badges are also essentially background to within the accuracy of the monitoring instruments. TABLE Vll-4 CUh1ULATIVE RESULTS OF ENVIRONhtENTAL htONITORING FOR Tile 1987 - 1988 REPORTING YEAR Film Badge Total Yearly Total Yearly Designation Exposure (mremV TLD'st Exposure (mrem)3 A1 40 1 hi A2 30 2 h1 A3 10 3 hi A4 20 4 h1 A5 40 5 h! A6 30 6 hi A7 30 7 hi 8 h1 9 hl 10 h1 11 hi 12 h1 film badge yearly exposures include significant contributions from May,1988 cuposure wbich was biased in comparison with other monthly esposure records occurred beuuse no control badge was included with the May film badges.

2. The first Seven TLD's are attached adjacent to the corresponding numbered film badge monitors.

3. M denotes minimal (<10 mrem) exposure.

Film badge yearly exposures include significant contributions from hiay,19SS  ! exposure which was biased in comparison with other monthly exposure records; the bias occurred because no control badge was included with the hiay,19S8 film badges. Film badges normally receive about 30 mrem during film badge handling and processing which  ! accounts for most or even all of the htay exposures. De exposure for June,1988 might be i attributabic to the Biennial Fuel Inspection (B 2) Surveillance though it is not really  ; significant, especially when the TLD exposures are considered. De accumulation of exposure recorded by month of exposure is presented in Table Vll 5. Vll-4 ( _ - _ _ _ _ _

l l t Based on Revision 3 of the UFTR Safety Analysis Report submitted to the NRC on l May 29,1987, plans are to eliminate some of the film badges currently used since the  ! l thermoluminescent dosimeters are preferred and were intended to replace the film badges  ! previously referenced in the Safety Analysis Report. No action has been taken on this l change to date, though plans are to implement this change in the next reporting year. l , TABLE Vil 5  ! ENVIRONMENTAL BADGE EXPOSURE RECORD BY MONTil OF EXPOSURE I Film Badge Total hiay,1988 June,1988 August,1988 Designation Excosure _ Exposure Exoosure Exposurt A1 40 30 hl 10 i A2 30 20 hi 10 i A3 10 10 M M i t A4 20 20 M M  !' I A5 40 20 10 10 A6 30 20 M 10 A7 30 20 10 M  ! l  ; I E. Personal Radiation Exposure  ! Maintenance and experimental work requiring significant exposure commitment was j minimized during this 19871988 reporting year as in the 19861987 reporting year following ) f the two years when major maintenance in the core area involved relatively large dose L commitments. UFTR associated personnel exposures significantly greater than minimum  ! detectable during the reporting period are summarized in Table Vil 6. l { l Table Vll 6 lists the permanent badge exposures recorded above background for the { reporting year for personnel employed directly at the UFTR. These exposures are  ! ( summarized for all badged UFTR personnel on an annual basis because all exposures with i one exception are well below 100 mrem. In addition, the largest exposures are generally l spread over several months primarily for support of experimental, research and educational  ; as well as maintenance sunelllances projects. l i Exposures for University of Florida personnel employed by the Radiation Control 9 l Office where the exposure is attributed to radiation control work associated with UITR [ activities was minimal with no individunt receiving a recorded exposure above background j in excess of 11 mrem whole body dose. Several individuals from the Radiation Control j Office periodically assigned to support UFTR related activities and special projects received l a non minimal dose for the year as listed in Table Vil 7 for the biennial fuel inspection (B- l

2) surveillance and for the semiannual antimony beryllium neutron source leak check (S- l

8) surveillance. The fuel inspection surveillance is typical of the type of project requiring additional radiation control support personnel usually at widely spaced intervals. During the 198719SS year only the fuel inspection surveillance required the utilization of radiation control personnel not normally assigned to support special UFTR activities requiring the presence of personnel from the Radiation Control Office. f Yli 5 L _ ___

b TABLE VII 6 l ANNUAL UFTR PERSONNEL EXPOSURE' Permanent sure Name Position (mrem 8'8 W.G. Vernetson Director of Nuclear Facilities M P.M. Whaley Senior Reactor Operator / Acting Reactor Manager 250 ( II Gogun Senior Reactor Operator M O.W. Fogle Reactor Operator M f R. Piciullo Student Reactor Operator Trainee /SRO 70 ( CJ. Stiehl Student Reactor Operator Trainee / Technician M M. Wachtel Student Reactor Operator Trainee 10 ( J. Godfrey Student Reactor Operator Trainee M ( 1. Severalindhiduals f, a the Radiation ControlOmce personnel periodically assigned to support UfTR. related aethities and recching a non minimal dose for the year are listed in Table VII 7.

2. M denotes minimal (<10 mrem) meaning background only.

3. All exposures reported here are for film badge readings for deep /whole body crposure.

( As delineated in Table Vll.7, shielding removal for the biennial fuel inspection was accomplished as a separate work item for this project with one staff person (Piciutto) ( receiving a measurable Aq whole body dose. Two fuel bundles were removed from the UFTR for the routine biennial fuel ( inspection surveillance (B 2). All personnel involved in the project were monitored by t'ilm badge dosimetry, with personnel ditectly involved also monitored by local use TLD dosimetry. In this activity which incleded replacement of shielding after the fuel l surveillance, sh (6) personnel received measurable exposures, three (3) form the UFTR operations staff and three (3) from the Radiation Control Office. All exposures listed in Table Vll 7 are for self reading pocket dosimeters used as whole body monitors unless otherwise noted, it should be noted that the exposure for Mr. Piciullo at ~60 mR whole body accounted for most of his yearly exposure recorded at 70 mR via a permanent film badge. Finaliy, the two other small projects included in Table Vil 7 account for small additional exposures. One individual from the Radiation Control Office receiving ~5 mR whole body dose during the semiannualleak check (S-8) surveillance of the Sb Be neutron source and one UFTR operations staff member receiving -8 mR during the replacement of the primary coolant purification system resins and ceramic filter in late September,1987. Vil-6 L

s / TABLE VII 7 RADIATION EXPOSURE ACQUIRED DURING SPECIAL UFIR PROJECl3' ( UFFR BIENNIAL FUEL INSPECTION (B 2 Surveillance) (June 22,1988) Shielding Removal For B 2 Fuel inspection UFFR Personnel { R. Piciullo 8mR Biennial B 2 Fuel Insocetion andJ,hielding Replacement k Radiation Control Personnel T. Ballard' 11/10 mR [ ht. Wilkerson 5mR K. Barker8 10 mR UFTR Personnd [ ht. Wachtel 7mR P. Whaley 17 mR ( R. Piciullo 55/60 mR (whole body) 280 mR (right I ankle) 80 mR (right wrist) 90 mR (left wTist) 61 mR (head) NEITIRON SOURCE (St> Be) LEAK CIIECK (S4 Surveillance) (htarch 30,1988) Radiation Control Personnel R. Ilansen 5 mR (whole body) 17 mR (right wTist) PRIMARY COOLANT DEhtlNERAll7.ER RESINS /CERAhtlC F111rER REPIACEMFNP (September 22,1987) UFFR Personnel R. Piciullo S/7 mR whole body

1. All ciposures listed are for se'J reading pocket dosimeters useJ as whole tWy monitors unless c4herwise neded.

2. Radiation Control Personnel not normally assigned to assist in UFTR operations except for targe projects-VII7 '

t

I A final special serdee project involved two 600 Ci C0 60 sources which were r transferred at the UFTR to the University of Florida, Department of Radiochemistry and I Department of Microbiology using the cell crane for off loading the new sources and on-loading the spent sources. All personnel exposures during this transfer were to personnel

,  from those departments with UFTR personnel assuring proper controls and handling as well l  as performing crane operations. All personnel participating in the transfer were monitored by film badge dosimetry. All exposures were to individuals from the two departments receiving the r,ources; the exposures are listed in Table Vil 8 and are for self reading pocket

{ dosimeters used as whole body monitors unless otherwise noted. ( TABLE Vll 8 RADIATION EXPOSURE ACOUIRED DURING SPECIAL PROJECTS NOT RELATED TO UFTR WORK ACTIVITIES COHALT-60 SOURCE TRANSFERS FOR 1RRADIATORS (May 6,1988)na l Dr. R. llanrahan 62 mR extremity,12 mR whole body Charles Crawford 7 mR Ravindra Bhave 10 mR

t. All exposures listed are for self reading podet dosimeters used as whole body monitors unless otherwise noted.

2. Att personnet listed are associated with the University of norida. Department of Radiochemistry or the

           . Department of Microbiology.

{ For visitors, students, or other non permanent UFTR personnel, a few individuals had a non zero dosimeter exposure measurement above 1Fo of the allowable quarterly limit for the entire reporting period as indicated on Table Vll 9. In most cases, the values of one up to ten mrem exposures recorded for self reading pocket dosimeters are probablv due to uncertainty in reading the devices or having dropped the dosimeter as noted in Table Vil-

 } 9. In all cases in Table Vil 9 except for llouck (on 2/15/88) dosimeters nonitoring other e   students participating in the same exercise or project indicated no exposure. Additionally, in all cases except for llouck, the projects did not involve any activities that would generate radiation exposure, llouck was performing work associated with an external project not related to the UFTR.

It should be noted that tours of reactor facilities are strictly controlled and limited during periods when the reactor is running or ports are open or other opportunities for significant radiation fields are present. Therefore, the lack of significant visitor exposure is expected and in agreement with ALARA guidelines. Vil-8 l _ - - - - - . -

1 TABLE Vll 9 EXPOSURES RECORDED FOR NON PERMANENT UITR PERSONNEL Personnel Date Exoosure Comments Patricia Kuta 1/19/88 10 mR Dropped Dosimeter (Evaluated 0 mrem) John Houck 2/15/88 10 mR (Left Hand) 8 mR Chest Film Badge indicates O mR David Browder 2/22/88 5mR Dropped Dosimeter (LabPartner had no Exposure) { James Monroe 3/4/88 10 mR Dropped Dosimeter (Film Badge, O mR) t s Gall Martin 4/14/88 Off Scale Dropped Dosimeter (Lab Partner, 0 mR) { V,G. Todd 5/18/88 6mR Dropped Dosimeter (bb Partner, 0 mR) f lieatherlyllicks 5/21/88 Off Scale Dropped Dosimeter (bbPartner had no Exponre) { lleatherlyllicks 6/16/88 5mR Bumped Dosimeter (bbPartner , had no Exposure) ( Patty Yawn 9/15/88 10 mR bb Student (8 Others),0 mR, 3 mR on Backup Dosimeter { Ed Styre 9/22/8S Off Scale bb Student (8 Others) had no Recorded Exposure (0 mR) l Vll 9 i _ _ _ _ _ _ _ _ _ . - . - _ _ . _ _ _ - - - . - - - - _ - - - - - - - J

l i i VIII. EDUCATION, RESEARCil AND TRAINING UTILIZATION l i NOTE: The participating students are indicated with an *, Other participants are faculty l or staff members of the University of Florida, unless speci'ically designated otherwise. A " indicates those studerts working on theses, projects or dissertations. , Badiation Protection Training Reactor Operations Based Radiation Protection licalth l Physics Cooperative Work Tralrdng Program, Dr. W.G. Vernetson, R. Rawls (CFCC), S. j l MacKenzie (CFCC), A. Mackovjak",11. Ilicks', R. Ilanrahan', Reactor Staff.  ; I A set of reactor operations based radiation protection health physics cooperative work  ! training exercises have been developed to meet the cooperative work needs of Radiation  ! Protection Technology students at Central Florida Community College (CFCC). Three (3) L l of these courses were conducted during this reporting year for a total of 23 students with , j great success. Students who take these courses are well suited to work as radiation control

  !                                 technicians and health ?hysics assistants at nuclear power plants. The exercises are also l                                    extremely adaptable anc. some of them have been upgraded and used in the undergraduate and graduate health physics laboratory and other courses at the University of F'orida. De       ;

development of this course and its subsequent presentation to CFCC students has been l partially supported under the UFTR DOE Reactor Sharing Program and 1as been a

valuable resource in the continuing effort to sustain and even increase reactor utilization. ,

 ;                                  During this reporting year a senior project was used to produce improved visual aids for        r some segments of the program.                                                                   j l

l f j UFTR Reactor Operations With NAA and Neutron Radiographic Imb Exercises Dr.W.G. t

Vernetson, Dr. II. Abbott, P.M. Whaley, R. Rawls/S. MacKenzie (CFCC), Dr. M. Lombardi l 1 (IICC). Dr. S. Marchionno (SFCC), E nomas' (FIT), S. Buell (SAllS), K. Wilson (IIMS), (

[ R. Allen (UCilS), V. Venkataktishnan'. D. Roberts', R. Ilanrahan',11. !!icks', Reactor i Staff, i ( i Mini courses (including lectures, tours, demonstrations, reactor operations, NAA of f l unknown and standard samples, demonstrations of neutron radiography, etc.) have been  ! , developed and presented as part of the UFTR DOE Reactor Sharing Program to provide l t aractical reactor operations, radiation protection and health physics training as well as NAA [ l aboratory experience and neutron radiography for groups of students from Central Florida j 1 Community College Radiation Protection Technology Program, Santa Fe Community  ; College Nuclear Medicine Technology / Radiologic Programs, the liillsborough Community l i College Nuclear Medicine / Allied IIcalth Technology programs and the Florida Institute of  ! Technology's Society of Physics Students. Other participants in all or part of such mini-l courses this year include a Boca Ciega liigh School physics class, a Union County liigh j

School Science class, a St. Augustine liigh School Physics class, a liswthorne Middle School t j Science class as well as individual students from Escambia liigh School and Glades Central Community liigh School.

[ Vill 1 }

 \                                                                                                                                   i
/

(

Reactor Operatlons 12horaton'(ENU 5176L) . Dr. W.G. Vernetson, P.ht. Whaley, Reactor Staff Students in the reactor operations course spend about two end a half hours weekly at the controls of the UFTR performing reactor operations exercises under supervision oflicensed reactor operators, ne lab encompasses training in reacthity manipulations, reactor checkouts, operating procedures, standard and abnormal operations and applicable regulations. Specific exercises directed toward development of understanding of light water l power reactor behavior are included as this laboratory course serves as basic preparation for students entering the utility industry in the test and startup area as well as plant operations. When this course is not interrupted by outages, students perform a series of j exercises designed to assure them of conducting 10 meaningful startups and 10 shutdowns along with a broad usage of reactivity manipulations. A special effort is made to correlate UFTR exercises with the classroom Icetures on various aspects of LWR operations. This ! stand alone lab course was offered three (3) times during the current reporting year as the l laboratory is now approved as a separate stand.alone course. Has!c Physles _Research Development of Pulsed lonization Chamber Plasma Kinetics Diagnostics Capabilities Dr. W.ll. Ellis, Dr. E.T. Dugan, W Y. Chol *, J.S. Parks', h1J. Baumgartner", J. hionroe' Exper! mental measurements have been made with several pulsed ionization chamber designs to determine plasma kinetic properties including first and second order recombina-tion coefficients as well as lon number densities in a fissioning plasma. Earlier work was confined to helium plasmas. During the current y ear work was extended to heated chambers containing higher pressures of UFglie mixtures. His work is ongoing as part of the Innovative Nuclear Space Power Institute research efforts in the Stretegic Defense initiative for supporting the development of space nuclear power 5;cneration sources. Service to Florida Foundation _ofluture Scientists - Lectures, Tours and Demonstrations of Reactor Operations - Dr. B. Abbott, Dr. W.G. Vernetson, R. Ilanrahan' 11. Ilicks', D. Roberts', UFTR Staff A series of lectures, tours and demonstrations of reactor operations and nuclear facility capabilities are conducted for a large number of student and facuhy participants in the annual Junior Science, Engineering and liumanities Symposium jointly sponsored each winter by the Florida Foundation of Future Scientists and the University of Ilorida for promising high schooljuniors and their teachers, his year the same sersice was extended { for participants in the National Jun!ar Science and llumanities Symposium held at the University of Florida in Spring,19SS with the Florida Foundation of Future Scientists serving as the host chtpter and also for groups of high school students la the Summer Research Program. Reacter Operations Demomtratiom Reactor Operations Instruction and Demomtrations for Various Courses Within the University of Ilorida Dr. W.G. Vernetson Reactor Staff. The following courses are identified where one or in some cases as many as four or five l classes or labs in a course would be conducted using the UFTR facility. All would begin with the lecture, tour and reactor operations and facility capabilities demonstration with ( Vill.2 t __ - - - - --

I L r L later classes, where needed, devoted to more detailed lab instruction in one or mora areas r of UFTR facility operations such as instrumentation demonstrations, rndiation surveys, L neutron activation analpsis using the rabbit system for short irradiations or the vertical ports for longer irradiations. Courses include: f pcurse Instructor ENU 3002 Dr. G.S. Roessler [ EMA 3050 Dr, D. Clark ENU 4144 Dr. W.G. Vernetson ENU-4194 Dr. W.G. Vernetson ( ENV-4201/5206 Dr. C.E. Roessler ENU 4905 Dr. W.G. Vernetson ENU 5005 Dr. G.R. Dalton { CilS 5110 Dr. M.L htuga CilS 5110L Dt'. K. Williams ENV 6211 Dr. C.E. Roessler ( ENV 6211L Dr. CE. Roessler/Dr. W.E. Bolch FAS 6428 Dr. M.O. Halaban ENV 6932 Dr. W.S. Properzio ( ENU 6935 Prof. J.S. Tulenko Radiation Protection and Control Health Physics Practigg - (ENV-4932/6932) Dr. W.E, ( Bolch, Dr. W.S. Properrio, Dr, W.G. Vernetson, D.L Munroe, II. Norton, R. IIagen, Reactor Staff. nis course provides students in various disciplines with practical crpe ience in radiation [ protection and control such as performing radiation surveys in and around the UITR cell and erwirons, calibrating area radiation monitors, determining effluent levels, setting up emergency exercises, etc. Rese exercises also serve as training for potential radiation { control technidans, most of whom are students in Nuclear or Emironmental Engineering Selences. Most of the actisity occurred in this category during this reporting period, f b'uckar Engineering laboratory 1 - (ENU-4505L) - Dr. W.l! Ellis, Dr. G.R. Dalton, Dr. WG. Vernetson, P.M. Whaley, J. Monroe', Reactor Staff. k ENU 45ML is the nuclear engineering laboratory course for undergraduate senior level students in Nuclear Engineering Sciences. De UITR is used for a variety of exercises and experiments, including NAA exercises, radiation dose measurements, measurement of { induced radioactivity, foil irradiations, flax mapping, evaluation of hot channel factors, calorimetry, blade worth reactivity calibration, deterraination of diffusion length in graphite j and 1/M approach to critical as well as a variety of other reactor physics parameter determinations and operational measurements. Nuclear Engineering I aboratory II . (ENU-6516L) Dr. W,il. Ellis, Dr. G.R. Dalton, Dr. W.C. Vernetson, P.M. Whaley, J. Monroc*, R. llagen', Reactor Staff. ENU 6516L is the nuclear engineering laboratory course for graduate students in Nuclear Engineering Sciences. De UFTR is used for a variety of exercises and experiments Vill-3

l including foil irradiations for coincidence counting,1/M approach-to-critical, neutron / gam-ma flux and energy mapping, neutron activation analysis, inverse reactor kinetics I m:asurements, control blade reactivity worth measurements and demonstration of the neutron radiography raethodology. NAA Research - Neutron Activation Analysis of Seagrass Community Components - Dr. l G. Chiu (UWF), Dr. Ranga Rao (UWF), Dr. W.G. Vernetson, D. Morton* (UWF), V. Venkatakrishnan*, R. Hanrahan*, Reactor Staff.

           - Various seagrass communities have been exposed to used drilling fluids off the gulf coast of northwest Florida. The components of one of these communities consisting of sediments, water samples, grasses, shells and shellfish meats have been subjected to long and short irradiations to monitor the uptake of certain hensy metals, principally barium and chromium, suitable for detection using neutron activation analysis. Reactor time for this                  ,

work was supported under the DOE Reactor Sharing Program. Results to date are I encouraging with work concluded except for a possibility that some samples will need to be reirradiated to support a paper and/or proposal submittal. 1 NAA Research Neutron' Activation Analysis of a Seagrass Ded Exposed to Drilling Fluids l

            - Dr. C.N. D'Asaro (UWF), Dr. T. Duke (UWF), Dr. D. Weber (EPA), R. Montgomery"                             '

(UWF), S. Macauley" (UWF), D. Morton' (UWF), V. Venkatakrishnan', R. Hanrahan*, H. Hicks *, Reactor Staff. [ This project involves moving cores from a seagrass bed to the laboratory where they are exposed to various drilling fluids to determine possible effects on seagrass community structure and biomass. Barium, chromium and scandium are present in the drilling fluids l and are known to impact negatively on animals and plants. However, knowing the correct concentrations of these metals is critical in order to correlate observed effects with metal concentrations to explain the phenomena involved. Use of the UFTR facility for the I irradiation and subsequent NAA provides an effective means of performing the chemical analyses. Reactor time for this work was partially supported by the University of West Florida through a grant from the Environmental Protection Agency with the remainder suppoited under the DOE Reactor Sharing Program and the UFTR facility. The external support was provided as an outgrowth of research during the 19851986 year supported as a seed project under the DOE Reactor Sharing Program. N6&B.esearch - Neutron Activation Analysis of Archeological Seashells Dr. T. Stocker (UWF), Dr. W.G. Vernetson, R. Hanrahan*, UFTR Staff. l Under the Reactor Sharing Program, neutron activation analysis is being applied to various archeological seashell specimens ranging up to nearly 1800 years old. Since shells were used as trade items by the American Indians in the Eastern half of the United States, the research is directed toward identifying enough trace element constituents in these seashells to develop a method for determining Indian trade routes in the Eastern United States.This research is in its early stages on a time available basis with no work performed during the current reporting year. Some information on this type of work may be available from a European reactor facility which has been requested to supply reprints of their work. Vill-4 i l .-- _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ __

i f NAA Research - Trace Element Evaluation of Seashells - Dr. Guy Prentice, Dr. G.S. I Roessler, R. Hanrahan*, UFTR Staff. Neutron activation analysis is being applied to identify the trace element composition of environmental seashells from various locations in Florida. The purpose of this research is to determine whether a set of key trace elements (nuclides) can be identified as signatures

     ~ for shells from various locations. The work continues as its purpose is being reevaluated and the work progresses on a time available basis with no irradiations performed during the current reporting year.

NAA Research - Neutron Activation Analysis of Estuary Sediments - Dr. R. Byrne (USF-St. Petersburg), Dr. G. Smith (USF-St. Petersburg), L Parlatore*, M. Lectzow', S. Knapp*, R. Hanrahan*, V. Venkatakrishnan', UFFR Staff. Under the DOE Reactor Sharing Grant, Instrumental Neutron Activation Analysis (INAA) is being applied to estuary sediments from the Tampa Bay region of Florida to determine ( and quantify the spatial distributing of various rare earth metals. Work to date has included preparatory work to map the spatial variation of the flux in the UFFR vertical ports and another exercise to determine accurate values for the cadmium ratios for ports to be used l in the activations for this research in a special graphite sample holder manufactured for this project. These are key parameters because of the resonance absorption characteristics of t many rare earth metals. Virgin teflon tube sample holders were demonstrated to withstand l extended reactor rur,s and have been analyzed for impurity content using NAA. During the current year one extended irradiation and analysis was performed with several relatively short irradiations performed to confirm previous results. The remaining samples in this project are expected to be irradiated and analyzed during the next reporting year with a proposal to obtain external support to follow. I l Investication of Properties of Fuel Storace Pit Liners Dr. S. Turner, Dr. W.G. Vernetson,

                                                                      ~

P.M. \9haley, R. Robinson *, J. Houck', UFFR Staff. Power reactor high density spent fuel racks typically are separated by sheet metal-enclosed I boron'silicide material. This project is intended to define parameters that may be used to i > gauge radiation damage and incipient failure (including significant absorber loss as well as l } mechanical failure) in boraflex. Specific procedures applied to date involve relative derisity i measurements, modulus of rupture tests, neutron transmission coefficient measurements and i neutron radiography of used as well as unused liner samples from utility spent fuel pools j with consistent results obtained to date. NAA Research Neutron Activation Analysis of Volcanic Rock Samples - Dr. M. DeFant (USF Tampa), Dr. W.G. Vernetson, V. Venkatakrishnan', H. Hicks", R. Hanrahan', UFFR r Staff. Under the DOE Reactor Sharing Program Neutron Activation Analysis is being applied to various volcanic rock samples from widely dispersed geographic locations ranging from Central America to both North and South America. The research is directed toward identifying the proper standards as well as effective irradiation am' decay schemes to facilitate trace element identification of sufficient numbers of different rare carth nuclides including uranium and thorium in the volcanic rock samples. During the current reporting i Vill 5 1 ____ ________

year this project involved expanded investigations of irradiation and decay schemes to c provide a larger data base of identifiable rare earth nuclides to support a proposal for I future funding. Eventually,information on geologic origins and rau earth mineral deposits is expected as NAA on such samples continues periodically. l Optical Physics Research - Analysis of Radiation Induced Lattice Disturbances in Dielectric Materials - Dr. H. Plendl (FSU), Dr. P. Gielisse (FSU/FAMU), J. Rink' (FSU), R. Hanrahan*, J. Monroe *, C.J. Stiehl, Reactor Staff. Various types and cuts of dielectric materials, primarily topaz, have been subjected to various thermal and fast neutron flueuces in the UFTR. Similar irradiations with 3 MeV electrons are being performed at Florida State University. The objective of this work is to I analyze the response of the material lattice to the disturbances caused by the various { components of the radiation field to include thermal neutrons, fast neutrons and gamma rays. Comparisons are being made with previous results of irradiations with X-rays and electrons and with thermal neutrons, all in isoiation. The purpose of the work is to gain a comprehensive understanding of how certain dielectrics such as Al2 (SO4 )(OH) and similar f lattices response to different types of radiation in the generation and destruction of color sites. The next phase will involve primarily high energy gamma and neutron irradiation in a UFTR experimental facility which has been under development and characterization for insertion in the UFTR shield tank. Cerenkov Noise Detector Development - Development of a Detector of Reactor Core Perturbations - Dr. E.E. Carroll, Prof. GJ. Schoessow, Reactor Staff. A new design Cererkov detector is being developed and tested using the prompt gamma , radiation deriving from the reactor core. The detector is being Scated in the thermal ' column entrance port with shielding plugs removed and substi*ated by lithiated paraffin plugs made for the purpose of reducing the neutron flux to rcceptable values when the reactor is running at power. Samples of the lithiated paraffin plugs were irradiated to assure that no unexpected activation products would be formed ".ere the plugs to see a large flux. l Other, work has involved spectroscopic analysis of the gamma energies emitted from the thermal column where the detector will be placed. The Cerenkov detector has been moved at various angles for various power levels with the ultimate objective to develop a detector system that is able to detect reactor perturbations at various power levels through large thicknesses of material by means of high-energy, penetrating, fission-produced gamma rays. The work to date has produced a doctoral dissertation and results are encouraging. This project has been in abeyance during the current year out is expected to be restarted in the upcoming year, possibly as part of the design element in the graduate level nuclear engineering laboratory course. UFTR Core Redesign (LEU Program) - Thermal hydraulic Analysis for Core Redesign - Dr. W.G. Vernetson, P.M. Whaley. As part of the DOE LEU Conversion Program, thermal hydraulic analysis related to redesign of the UFTR core using SPERT fuel rods has been performed. Computer analysis has been undertaken to evaluate the UFIR/SPERT design for steady state conditions as well as transients arising in response to a step insertion of reactivity, a loss of coolant flow, and a loss of coolant accident. Results to date indicate required safety margins and transient Vill-6 L

E i f h response conditions can be maintained with the UFTR/SPERT core design. Since support L has been providcd to analyze conversion alterations, the decision on whether to go with SPERT or plate fuel will be made in the near future with thermal hydraulic rela .1 conversion analysis expected to begin during the upcoming year to provide input to support the license amendment for the HEU to LEU conversion. , 1 l NAA Research - Determination of Sodium Concentrations in DNA Samples - Dr. Randolph Rill (FSU - Chemistry Dept.), T. Strecleclia" (FSU), R. Hanrahan*, UFTR Staff. ( I NAA is being used to characterize and quantify the uptake of sodium by DNA to investigate phase transitions in concentrated solutions that use sodium as a counter lon. Since the concentration of sodium is the major determinant of phase transition behavior, the determination of sodium concentration in DNA samples is being used to describe the liquid crystalline phases of DNA and the anomalous behavior of DNA phase transitions at low ionic strengths. The high purity of the sample as well as the element of interest (sodium) makes the determination of sodium concentration in these samples ideal candidates for NAA using short term irradiations via the rabbit system. This work has been supported by a federal research grant and is nearing completion. I L ) UFTR Risk Assessment - Dr. W.G. Vernetson. A preliminary probabilistic risk assessment of the University of Florida Training Reactor l has been conducted. This project has determined an estimate of the probability of l occurrence of a set of postulated maximum credible UFTR accidents. The results will be used to show that the UFTR poses no significant risk to the general population and environment around the UFTR and has demonstrated proficiency in PRA analyses as additional PRA projects are undertaken. Specifically, research is continuing to obtain better i data for the maximum credible accidents and extend the methodology to examine risk associated with less serious but higher probability UFTR related accidents or failures of key systems such as safety channels. This project is relatively inactive at present awaiting further student interest; it should be noted that NRC has shown some interest in this area which may lead to its reactivation. NAA Research - Trace Elements in Coal Slurry Samples - Dr. R.A. Llewellyn (UCF, Dept. of Physics), R. Vargas* (UCF), R. Hanrahan*, Reactor Staff. This project involves determining the concentrations or trace metals and uranium decay products taken from coal slurry settling ponds. The specific clements of interest are routinely mined from coal deposits; the potential for increased yields per energy used in l recovery 'is being tested, with NAA providing an assessment of the trace element concentration for specified settling pond sites. The first stage of this project is nearing completion with the potential for future commercial studies well established. Reactor time for this work was supported under the DOE Reactor Sharing Program. NAA Research Determination of Chlorine (and Titanium) Concentrations in Quartz - G.P. LaTorre (GelTech), Dr. C. Balaban (Advanced Materials Research Company), R. Hanrahan', Reactor Staff. Different manufacturing techniques and parameters are used to reduce the concentration Vill-7

L of chlorine in quartz glass (silica) produced for optical uses. Compositional characterization of the glass is based on the titanium / silicon ratio. The high purity of the sample matrix and [ the elements ofinterest (Cl, Ti) for this project make NAA ideally suited to determine the concentrations of chlorine and titanium remaining after various processing stages. Funding for this service work is supplied through the Advanced Materials Research Center. k NAA Research -TrialIrradiation of Phosphate for Rare Earth Element and Other Element Characterization - Dr. P. Gielisse (FAMU/FSU, Dept. of Mechanical Engineering), Dr. R. {. Clark (FSU, Chemistry Dept.). Various phosphate ore samples are being assessed using NAA to identify significant concentrations of rare earth elements for potential mining applications. Interest in this project is spurred by the large mined phosphate deposits in Florida as well as the recent advances in superconductors involving various composite materials containing rare earth elements. Analysis is in progress for short and long duration irradiations. Reactor time for i this work is being initially supported under the DOE Reactor Sharing Program as data is being generated to support a proposal for external funding. l I NAA Research - Germanium Trace Element Concentrations in Lake Sediments in Florida - Dr. C.L Schelske (UF, Fisheries and Aquaculture Dept.), R. Hanrahan*, Reactor Staff. I A feasibility study is being conducted to determine the suitability for using NAA to determine trace concentrations of germanium in sediments taken from several north Florida j lakes. Efforts to date have been inconclusive as sample spectrum analysis is hampered by sample and standard matrix elemental composition as well as the expected low germanium concentration values. Work is expected to continue as revised irradiation schemes will be implemented to attempt qualitative identification of germanium in the lake sediments. NAA Research - Blogeochemical Assessment of the Pollard, Alabama Oil Field - Dr. G. Cwick (UWF), D. Boudreau* (UWF), R. Hanrahan*, H. Hicks *, Reactor Staff. The biogeochemical analysis of soil and vegetation samples is the first phase of a three-phase' study to determine if hypothesized biogeochemical anomalies occur in the Pollard, Alabama oil field and can be correlated to tonal anomalies in satellite imaging that corresponds to hydrocarbon deposits. Potentially abnormal concentrations of selected elements characteristic of hydrocarbon seepage from underground deposits could produce identifiable stress type conditions or growth reactions in the vegetation. These environmen-tal characteristics may be correlated to satellite mapping of hydrocarbon production potential. Environmental vegetative anomalies detected by neutron activation analysis will be correlated to image anomalies. This work is initially supported under the DOE Reactor Sharing Program as data is being generated to support a proposal for external funding. Elasma Physics Studies - High Temperature Pulsed lon Chamber Plasma Diagnostic Reactor Shield Tank Irradiation Facility Design - Dr. W.II. Ellis, Dr. I. Maya, P.M. Whaley, J. Monroe * *, W.Y. Choi*, A. Ferrari*, Reactor Staff. In support of the design of a high temperature irradiation facility for pulsed ion chamber diagnostic experiments to be performed in the shield tank of the UFFR, flux mapping is b:ing carried out. The purpose of this flux mapping is to determine the general radiation Vill-8 1

a

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flux profile in the shield tank, both gamma and neutron, and locate the highest usable flux field therein, a determining factor for placement of the irradiation facility. Gold foils and thermoluminescent dosimeters have been used for neutron and gamma field flux mapping with additional measurements in progress to better define the flux distribution. When ( completed, the shield tank facility will provide a more flexible pulsed ion chamber plasma diagnostic experimental arrangement to facilitate loading and unloading of experimental 5 chambers to allow non disruptive temporary storage without complete removal between experiments.This arrangement will promote the multiple simultaneous usages of the UFTR and reduce personnel exposure. The design and operation of the facility is in support of plasma diagnostic studies associated with establishing the engineering design parameters for r gaseous core reactor /hiHD converter space power systems currently under study by the . L Innovative Nuclear Space Power Institute (INSPI). r Plasma Physics Studies - hiultiprobe PIC Diagnostic Studies of Nuclear Enhanced MHD il L Plasmas - Dr. W.H. Ellis, Dr.1. hiaya, Dr. W.G. Vernetson, W.Y. Choi", J. hionroe*, A. Ferrari*. I The objective of this research is to investigate those characteristics of nuclear generated plasmas that are related to critical engineering design parameters for gas-core reac- , tw.dHD converter systems. The work will be directed toward the development of an [ rimental system to measure the various design parameters as functions of temperature  ; e . pressure for nuclear generated plasmas to include the nuclear ionization source rate, l [Jasma loss coefficients, and electrical conductivity. Ionizatiou chambers filled with L candidate reactor fuel gas /h!HD working fluids will be placed into the UFTR equipped with a high temperature heater syrtem, with gas purge, plasma diagnostics, power, control i , and environmental monitoring systems, hieasurements will be performed over a range of l [ temperature and pressure conditions and for a range of reactor power levels (and nuclear ionization source intensities) and gas compositions in support of the University of Florida INSPI space power research program. Gaseous Release Determinations Evaluation of UFTR Gaseous Release Determination hiethodology - Dr. W.G. Vernetson, Dr. W.E. Bolch, P.ht. Whaley, B. hiurray", R. I ( Ilanrahan', Reactor Staff. I In response to USNRC Region 11 Inspection Report No. 50-83/88-01, the methodology for ( determining Argon 41 concentrations in UFTR stack effluent is being evaluated and modified as necessary to assure accuracy of the measurement. The evaluation and modifications are being addressed as part of a graduate student masters project. The { principle modification involves the use of a low density (simulated gas geometry), multiple nuclide source to provide calibration data for the concentration determination procedure. Work completed during the reporting year included obtaining a properly sized source and ( usage in the semiannual gaseous release surveillance measurement of the Argon 41 stack concentration. Evaluations of the concentration determination using the new source as compared to the concentration determination using the previously utilized resin. cast Cobalt-l 60 source are being made; in addition, extensive measurements are being made on the effects of variation of sample geometry and votume in the concentration determination.This work will evtend into the next reporting year and should provide a better indication of the accuracy of Argon 41 stack release measurements. ( Vill 9 i

{ UFTR Core Redesign (LEU Program)- Neutronics Analysis for UFTR Core Redesign - 7 Dr. W.G. Vernetson, Dr. E.T. Dugan, P.ht. Whaley, hl. Salih*. l As part of the DOE Low Enriched Uranium Conversion Program, investigations have been performed on the UFTR to determine the feasibility and desirability of replacing the 93% ( enriched hiTR plate type fuel with 4.8% enriched, cylindrical SPERT fuel pins. For this redesign, the only permanent structural modification had been hoped to be the insertion , of new grid assemblies into existing fuel boxes. Acceptable neutronic criteria (possible k,,, ( range, maximum flux and degree of undermoderation) have been determined using industry-accepted, 4-group cross sections in one, two and three-dimensional d!ffusion theory r c iculations of k,,,, flux profiles, power peaking factors and coefficients of reactivity. First j order perturbation calculations have been used to determine key kinetic parameters. Neutronic results to date indicate that the UFTR/SPERT core redesign can be j accommodated to meet requisite neutronic criteria with an actual increase in peak therrnal l ( flux levels which will be very useful for NAA and other research projects requiring high thermal flux levels. The UFTR finally received a grant to support during the current j i reporting year to begin the conversion process beginning with a decision on whether to go I with SPERT or plate-type fuel. Neutronics analysis to date on this project has involved obtaining and setting up the code methodology to be utilized in producing the licensing package for submission to USNRC. UFTR Operator Training and Reaualification - Dr. W.G. Vernetson, Reactor Staff. ( Lectures and hands on operations on the reactor are recessary to license operators for the UFTR. the requalification program establishes a required number of startups, weekly checks, daily checks, drills, practical exercises and lectures for each operator. Operator participation is mandatory in order to maintain assurance of proficiency levels and to be [ able to demonstrate the requisite operator skills. Operational proficiency is assured by  ; written and oral tests as well as observed practical exercises. The same program in an l accelerated mode is used to train UFTR reactor operator license candidates. Current 10 1 CFR Part 55 (Operator's Licenses) requirements have been considered in continuing the UFTR Operator Requalification and Recertification Training Program.Three trainees were involv'ed in the initial training this year; one dropped out to take a position elsewhere, one sat for and passed the SRO exam and the third set for the RO license but failed one i section of the exam which will be retaken early in the next reporting year. i f Reactor Operations - Utility SRO Certification Operations Dr. W.G. Vernetson, P.hi. Whaley, D. Scukanec (GPC), C. Narmi (GPC), hf. Rowe (GPC). l Periodically, utilities with nuclear power plants require certification operations to be performed by management personnel for SRO certification. This operations usage involves the performance of a set of meaningful reactor operations exercises involving significant reactivity manipulations plus a minimum of 10 startups and 10 shutdowns. This usage was provided for only one Georgia Power Company (GPC) Plant Vogtle Operations supervisor SRO candidate during this reporting year. l Vill 10

Q;neous Release Determinations - Argon-41 Stack Measurements - Dr. W.G. Vernetson, Dr. W.E. Bolch, P.M. Whaley, D.L Munroe, B. Murray*, R. Hanrahan*, Reactor Staff. A Cobalt-60 Standard Sample has been applied in standardized controlled measurements of radioactivity (Ar-41) in stack effluent. A direct detailed standard operating procedure (UFTR SOP-E.6: Argon-41 Concentration Measurement) has been developed and approved as the best practicable evaluation of Ar 41 releases from the UlTR facility as required by UFTR Technical Specifications on Effluents Surveillance in Section 4.2.4, Paragraph (2). l Application of this SOP continues to obtain h statistically significam number of data points l and eventually to investigate the effect of variable core vent flow on total Ar-41 releases. j Other comrnitments during this reporting year have limited progress on this project; with the expectation of eventually raising power levels plus the dec. eased Ar-41 release limit in the proposed 10 CFR 20 revision, this work will be moved to a higher priority in the next reporting year if a student can be found to work on it especially if other work to characterize the Argon 41 measurement methodology is en aluded successfully, l l NAA Research - Neutron Activation Analysis for Characterization of Various NBS and , USGS Standards - Dr. W.G. Vernetson, Dr. W.H. Ellis, P.M. Whaley, I Tryboski", H. Hicks', J. Nefflen", R. Hanrahan*, Reactor Staff. Various NBS standard reference source samples in various dilutions are being irradiated l for neutron activation analysis to determine the NAA lower limit of detection for the l various standards and to identify and benchmark secondary standards based on NBS noncertilled concentration values and USGS (US Geological Survey) standards obtained from USGS. This work formed the basis for training a high school student in research methods under the 1986 and again under the 198S Florida Foundation of Future Scientists Summer High School Student Research Program as well as for a students senior project l during the current year. Limited results have been obtained to date, although good reports in limited areas have been prepared by the students in each case, the work has continued to progress slowly as various reliable secondary standards are to be developed to facilitate i NAA on samples where multiple trace element concentrations are to be determined. This l ongoing project provides data on which to base generating irrediation and decay schemes  ! targeted to measure concentrations of specific elements in NBS Standards to assure certified l comparisons with unknown samples are available. Work to dele is progressing well;

considerable additional effort is required to benchmark uncertified contents of standards.

Work on this project was partially supported via the DOE Reactor Sharing Program for a high school student research and science fair project and provided a valuable research experience. NAA Rexarch - Implementation of Upgradtd NAA Laboratory Facilities Dr. W.G. l Vernetson, Dr. W.H. Ellis, Dr. G.J. Schoe. .,0w, R. Ilanrahan', P.M. Whaley. The implementation of the two PC based ORTEC analyzers with spectrum analysis software l in the 19861987 reporting year caused the decision to be made not to upgrade an ND66 MCA since the NAA Lab now has state of the art analytical capabilitics for performing spectrum analysis and subsequent neutron activation analysis. The new larger standardized size sample holder is for the rabbit system has also worked well to facilitate case and speed of handling samples for NAA. During this year manual cell isolation valves were installed to provide a backup m ans to assure samples could not be inserted until allowed by the l Vill-11 l

reactor operator. Earlier in the year a post accident core vent sampling connection was also installed in the rabbit system lines to provide for sampling of cell air radioactivity levels prior to venting during abnormal or emergency operating conditions per UFIR Tech Spec Amendment No.17. Neutron Radiocraohv Facility Develooment - Determination of Beam Characteristics and Optimization of Facility - Dr. W.G. Vernetson, Dr. A.ht. Jacobs, Dr. S. Nagler, Dr. H. Van Rinsvelt, P.M. Whaley", H. Hicks *, L hforales, UFTR Staff.

     'P n al column and East West throughport facilities were evaluated for radiation beam
      ; hts: eristics with the thermal column being optimize as a neutron radiography facility. A
        . c61imator/collimator and drift tube assembly have been completed, a film cassette and de. '.oping facility have been implemented. The beam ccnfiguration modifications are nearing completion to attempt certifiable Class I (ANSI Standard E545) neutron radiographs. Following final beam configuration development, a shield and shutter assembly will be developed. One funded and several other repeated applications have been performed in this reporting year with interest expressed by several other potential users for the upcoming year. Checks to determine possibility of producing real time radiographs in several configurations were unsuccessful in the 1986-1987 reporting year. liowever, this developmental project is ongoing and a major enterprise for utilizing staff time and design efforts in the next reporting year as we hope to obtain a real time system. During the i     current year extensive work to optimize and characteri7e the facility parameters has been l     accomplished along with completion of complete darkroom facilities for radiograph development including the loan of an autoprocessor. During the upcoming year plans are to finalize characterization of facility parameters, install permanent shield facility and again try to accomplish real time radiography.

t- Basic Physics Studv Neutron Irradiation of Geologic Quartz Dr. A. Odom (FSU), Dr. l W.G. Vernetson, J. Rink", UFTR Staff, i The UFTR has been used to provide a source for fission of uranium traces in geologic l quartz to produce Frankel defects in the quartz crystal structure. This irradiation simulates the effects of exposure to cosmic radiation. The defects are then being analyzed to provide a calibration for dating techniques. During the current year NAA research has been emphasized to quantify U, Th and other rare earth constituents of the geologic quartz sarnples with emphasis on U, Th and Sm because of their long term radioactive effects. l l hinlical/ Physics Research Estimate of I 123/I-127 Ratio in Radiopharmaceuticals Using j INAA - Dr. C. Williams (VA Hospital), Dr. hi. Thor'nor (VA Hospital), Dr. W.G. ! Vernetson, P.M. Whaley, Reactor Staff. l Medical imaging with radio iodine (1 123/125) is performed via introduction of radioactive ) iodine into a biological system; the production of the imaging compound is improved f through the addition of stable iodine. This INA project is hoped to provide information regarding the amount of stable iodine used to pmvide maximum benefit to the imaging compound. Vill-12 \ _ _ _ _ _

l L r l NAA Research - Characterization of the Trace Element Content in Mt. St. Helen's Ash - Dr. W.G. Vernetson, P.M. Whaley, R. Hanrahan, P. Kuta* *, J. Musgrove (EHS). ( Neutron Activation Analysis is being applied to quantify the elemental constituents in ash obtained from the 1980 eruption of Mt. St. Helen's volcano. The objective of this work is ( to identify potentially hazardous elemental constituents and to determine if useful quantities of any elements such as rare earths were emitted. Various irradiation and decay schemes have been implemented to obtain partial, but as yet inconclusive and incomplete, data on ] ( the elemental constituents of the ash. Work on this project vy, partially supparted via the DOE Reactor Sharing Program for a high school student u ,earch and science fair project and provided a valuable research experience for the stu/ .nt involved { LEU Conversion - Special SNM 1050 SPERT Low Er. .ched Fuel Conversion Efforts - Dr. W.G. Vernetson k Extensive efforts are undenvay to qualify the SPERT ,.el for use in the UFTR. Work to date on the SPERT fuellicensed under SNM-1050 hac L:cluded extensive decontamination ( work, radiation and contamination surveys, property ..e ?ys, SNM 1050 facility modifica. tions, fire alarm system maintenance / upgrade, LEU SPERT fuel movement to a newly decontaminated room, security system modification, NRC Radiation Safety Inspection and ( complete pin by pin identification number verification for fuel inventory and visual inspection. Efforts in this area should conclude early in the next year with X-ray non-destructive examination of pins selected as candidates for the conversion. ( Facility Characterization - Determination of UFTR Beam Ports / Thermal Column Neutron Spectra Dr. W.G. Vernetson, Dr. W.H. Ellis, P.M. Whaley, K. Mondlak', J. Monroe *, UFTR Staff. [ The neutron spectra at the thermal cob i m, South beam port and South West beam port I are being determined to provide information for irradiation services. When t6e irradiation I ( and analysis protocol is established, variation in beam parameters will be attempted to , determine the viabihty of beam variations. This project was initiated by a participant in the 1 [ 1987 Summer Student Research Program and has been continued into this reporting year i to provide the basis for a science fcir entry. The work to date is progressing well as several ! laboratory exercises have contributed to the data base for this project. Electronic Material Irradiation Research - Dr. W.G. Vernetson, Prof. J.S. Tulenko, P.M. Whaley, R. Hanrahan', S. Knapp*, UFTR Staff. A series of measurements in the previous year provided evaluation of experimental ports for specific irradiations such as irradiation of electronic components and neutron j transmission measurements as well as data for future irradiations. Current plans are to l subject various hardened electronic components to various neutron and gamma fields to characterize their resistance to damage. This is applied materials research would support the University of Florida DOE Robotics Program. f { Vill 13 i

l L r L CHS 5510/5510L - Dr. K. Williams, Dr. h!.L hiuga, Dr. W.G. Vernetson, P.M. Whaley, R. Hanrahan*, C. Crawford*. ( Radiochemistry laboratory project exercises of half life determination, neutron activation analysis of silver and aluminum in metal samples and on identification of chlorine in ( chemical samples have been performed using both an Nal scaler system and a HPGe spectrum analysis system. Data from this set of class exercises has been used to develop a standardized UFTR exercise. Extensive work has also been performed as a project in the ( CHS-5510L Laboratory to identify the trace element concentrations in powdered milk to provide the basis for a yearly repeatable laboratory experiment. [ hiaintenance Activity - Activities to Correct Failures and Restore the UFTR to Operable Status - Dr. W.G. Vernetson, P.M. Whaley, UFIR Staff. ( Routine corrective maintenance on UFTR systems nnd facilities occupied a considerable amount of time during the reporting period, with two major maintenance projects requiring significant effort during the reporting year. [ Control Blade Drive Unit Maintenance - Maintenance was conducted in response to a series of events involving a control blade drive motor failure and a control blade unit drive train failure. During the maintenance activity, one control blade drive motor was replaced [ and bearings in the drive train from the motor to an intermediate shaft unaffected in I previous control blade major maintenance programs were replaced. Solidified oil and grease had been impeding the operation of control blade Safety 1; following definition of the [ problem, all control blade drive units were dismantled, cleaned, reassembled and lubricated with one modification installed during the operation to allow the use of commercially available bearing retainer spring clips. Following control blade drive retests, the UFTR was [ restored to normal operations. Power level Safety Channel 1 Monitoring Failure - An intensive maintenance program was [ conducted for Safety Channel 1 when the instrument failed downscale Initial indications were that a filter capacitor had failed and altered the circuit to cause the downseale j movernent. Following replacement of the capacitor and a week of normal reactor operations, the iailure recurred. Series of tests and checks revealed th: need for the performance of several maintenance items, but did not definitely identify the root cause. [ Since the failure appeared to be an intermittent fault and therefore, not isolatable by I ( systematic troubleshooting methods, a program of sequential replacement of all potential faulted components was devised with a retest for each step. Following the replacement of I cable connectors and a step intended to aid fault identification in the programmed t replacement series, Safety Channel 1 passed the retest and the UFTR was restored to normal operation. { 1 i I { Vill-14 t

L ( IX.11IESES, PUBLICATIONS, REPOR'13 AND ORAL PRESPNI'ATIONS OF WORK RELATED "ID THE USE AND OPERATION OF TIIB UFIR L r 1. "Distribution of Rare Earth and Other Elements in Some Egyptian Phosphorites," L M.A. El Haddad and E.A. Ahmed, paper presented at the Fourth Symposium on Phanerozoic and Development in Egypt held in Cairo, April 22,1987 (omitted from 19861987 report).

2. "Die Geochemistry of Rangwa and Homa Bay Carbonatites (West Kenya)," M.A. El Haddad, paper presented at the 14th Colloquium of African Geology held at the

( Berlin Technical University, Berlin, August 18-22,1"7 (omitted from 1986-1987 report). ( 3. "Dark Blue Green Beryl Produced by Electron Irradiation," WJ. Rink, PJ. Gielisse, T. Erch and H.S. Plendl, mill. Am. Phys. Soc.. 31,1293 (1986) (omitted from 1986-1987 report). (

4. "Government Support to Cover Cost of UFTR Conversion From IIEU to LEU Fuel,"

W.G. Vernetson, July,1987, Proposal Submitted to Department of Energy, Nuclear ( Engineering Sciences Dept., University of Florida, Gainesville, August 14, 1987 (Funded effective November 15,~ 1988). l ( 5. "Fall Semester Reactor Operations 1.aboratory hianual for ENU 51761," W.G. Vernetson and P.M. Whaley, Nuclear Engineering Sciences Dept., University of Florida, September,1987.

6. "Quality Assurance Program for Shipment of SPERT F 1 Fuel Pins Per 10 CFR Part 71," W.G. Vernetson, OA Program submhtal to USNRC Office of Nuclear Material

( Safety and Safeguards to obtain OA Program approval to ship SPERT F-1 fuel in DOT specification 6M shipping containers. University of Rorida, Gainesville, FI,  ! September 2,1987. j ( 7. "Reactor Usage Operations Programs for Georgia Power Company Degreed Personnel," W.G. Vernetson, UFTR SRO Certification Operations Manual for { Reactor Usage on September 9-11,1987. 1

8. "Reactor Usage Operations Programs for Georgia Power Company Degreed Personnel," W.G. Vernetson, Final Report on SRO Certification Operations, Nuclear

( Engineering Sciences Dept., University of Florida, Gainesville, F1, Seperaber 16, 1987.  ; { 9. "UFTR Final Safety Analysis Report Revision 4," W.G. Vernetson, Official Submittal to USNRC, Nuclear Engineering Sciences Dept., University of Rorida, Gainesville, FI, September 27,1987. l

10. "Initial Characterization of Boraficx Surveillance Coupons for Vogtle Plant," S.

Turner, Nusertech, Inc., Palm liarbor, FI, October,1987. { IX-1 l - - - - . - - - - -

[

11. "NAA - Trace Element Analysis for Rare Earths in Volcanic Rock," W.G. Vernetson, R. Hanrahan, Final Report to Dr. M. DeFant, Nuclear Engineering Sciences Dept.,

Gainesville, FL October 27,1987. (

12. "Major Maintenance Outages in a Nonpower Reactor Environment," W.G. Vernetson and P.M. Whaley, Trans. Amer. Nucl. Soc.. Z, p.192, November,1987.

(

13. "Nuclear Seeded MHD Plasma Diagnostic Experiment With the PIC System," W.H.

Ellis, Oral Presentation for the INSPI Gas Core Reactor Working Group Meeting, ( Los Angeles, CA, November,1987. ,

14. ' Trace Element Distribution in Chromites from Six Occurrences in the Eastern

[ Desert, Egypt," M.A. El Haddad and A.A. Khodeir, paper presented at the Twenty-Fifth Annual Meeting of the Geological Society of Egypt held in Cairo, November 14 17, 1987. [

15. "Annual Progress Report of the University of Florida Training Reactor for September 1,1986 - August 31,1987 Reporting Year," W.G. Vernetson, November,1987.

(

16. "Effects of Drilling Fluids on Seagrass Communities," W.G. Vemetson and R.

Hanrahan, Interim Report to Dr. D. Weber and Dr. C. D'Asaro, Nuclear Engineering ( Sciences Dept., Gainesville, FL, December 8,1987.

17. "Fine Florida Fibers, Inc. - A Proposal to Produce Fiberglass Woo! Insulation From Waste Product Coal Fly Ash," R. Hanrahan and L Worth, project utilizing results of NAA for EMA-3050 Introduction to Ceramics Course, Materials Science and Engineering Dept., University of Florida, Gainewille, FL December 9,1987.

[

18. "Verification of Non-Certified Elemental Con;entration In NBS Standards," IJ.

r Tryboski, ENU-4905 Special Project in Nuclear Engineering Sciences Dept., L University of Florida, Gainesville, FL December 14, 1987. r 19. "Analysis of Major Constituent Elements in Coal Fly Ash," RJ. Hanrahan, ENU-L 4905 Independent Study Project in Nuclear Engineering Sciences Dept., University of Florida, Gainesville, FL December 15, 1987.

20. "Development of Instructional Visual Aid Materials for Radiation Protection Technology Training at the University of Florida Training Reactor," K.A. Mackovjak, r ENU-4905 Special Project in Nuclear Engineering Sciences Dept., University of i Florida, Gainesville, FL December 15, 1987.

( 21. "Final Report on the Fall Semester Reactor Operations Based liealth Physics l Cooperative Work Training Program," conducted for Radiation Protection Technology Prograru Students at Central Florida Community College, W.G. Vemetson, Nuclear Engineering Sciences Dept., University of Florida, Gainesville, ( FL December,1987. . { ( IX 2 i

L r L

22. "Nuclear Seeded Plasma Diagnostic Experiments With the PIC System," W.H. Ellis, r W.Y. Choi and M.J Baumgartner, in INSPI-QR UF-008, lanovative Nuclear Space L Power Institute (INSPI) Quarterly Progress Report for Period Ending September 30, 1987, University of Florida, Gainesville, FL, January,1988.

23. "Spring Semester Reactor Operations I2boratory Manual for ENU-5176L," W.G.

Vernetson and P.M. Whaley, Nuclear Engineering Sciences Dept., University of Florida, Gainesville, FL, January,1988. [

24. "Effects of Drilling Fluids on Secgrass Communities," W.G. Vernetson and R.

Hanrahan, Interim Report to Dr. D. Weber and Dr. C. D'Asaro, Nuclear Engineering l Sciences Dept., University of Florida, Gainesville, FL, January 19, 1988. [

25. "NAA of Local Water Samples for Rare Earth Elements - Tampa Bay Estuarine l Samples," W.O. Vernetson, R. Hanrahan, Interim Report to Drs. R. Byrne/G. Smith, Nuclear Engineering Sciences Dept., University of Florida, Gainesville, FI, January 20,1988.

26. "An Analysis of the Neutron Energy Spectrum Within the Experimental Ports of the University of Florida Training Reactor," K.A. Mondlak, Abstract and Oral

{ Presentation on FFFS Summer Research Project presented at the 25th Annual Junior Science, Engineering & Humanities Symposium held at the University of Floric'-> Gainesville, FI, January 31 - February 2,19S8.

27. "Results of Activation Analysis of Geologic Quartz Samples for Uranium and  ;

Thorium," W.G. Vemetson, R. Hanrahan, Report to Dr. A.L Odom and J. Rink, ( Nuclear Engineering Sciences Dept., University of Florida, Gainesville, FI, February 3, 1988. ( 28. "Univarsity of Florida Reactor Sharing Program," W.G. Vernetson, proposal subadtted to Depar* ment of Energy, Nuclear Engineering Sciences Dept., University of Flarida, Gainesville, FI, February,1988. (

29. "Summan Report on Boraflex Samples Analysis for Vogtle Plant," S. Turner, Nusertech, Inc., Palm Harbor, FI, Februay,1988.

(

30. "Radiation Tests of Boraflex Coupons for Bisco Company," S. Turner, Nasertech,  !

Inc., Palm Harbor, FI, March,1988. , f

31. Examination of Boraflex Coupons From Grand Gulf Nuclear Station," S. Turner, l Nusertech, Inc., Palm liarbor, F1, March 1,1988. l

{ '

32. "Analysis of Trace Elements in Coal Slurg Samples," W.G. Vernetson, R. Ilanrahan, Interim Report to Dr. R.L Llewellyn, Nuclear Eng!ncering Sciences Dept., University

( of Florida, Gainesville, FL, March 10,1988.

33. "Die Neutron Energy Spectrum Within a 100-Kilowatt Light Water Reactor," K.

( Mondlak, Science Fair Presentation on FFFS Summer Research Project, University of Flerida, Gainesville, FI, March,1988. IX-3

r. .

t r L

34. "Proposal Submitted to the Nuclear Regulatory Commission to hicet 10 CFR 50.64 r Requirements for Scheduling UFTR Conversion From lieu to LEU Fuel," W.G.

L Vernetson, updated scheduling proposal submitted to USNRC, Nuclear Engineering Sciences Dept., University of Morida, Gainesville, FI, hfarch 22,1988.

35. "Effects of Drilling Fluids on Seagrass Communities," W.G. Vernetson and R.

Hanrahan, Interim Report to Dr. D. Weber and Dr. C. D'Asaro, Nuclear Engineering Sciences Dept., University of Florida, Gainesville, FI, hfarch 25,1988.

36. "Microcosm Studies on the Effects of Drilling Fluids on Seagrass Communities," D.

[ Morton, G. Chiu, et.al, Paper presented at the International Conference on Drilling i Wastes held in Calgary, Alberta, Canada, April 5-8,1988. i [

37. "Neutron Activation Analysis to Identify Elemental Constituents of Ash Samples t From hit. St. Helens Volcanic Eruption," P. Kuta, ENU 4944 Practical Work Project Report in Nuclear' Engineering Sciences Dept., University of Rorida, Gainesville, FI, April 29,1988.

38. "Summer Semester Reactor Operations Laboratory hianual for ENU-51761.," W.G.

Vernetson, Nuclear Engineering Sciences Dept., University of Florida, Gainesville, ( FL, hiay,1988.

39. "Final Report on the Spring Semester Reactor Operations Based IIealth Physics

( Cooperative Work Training Program," conducted for Radiation Protection Technology Program Students at Central Florida Community College, W.G. Vernetson, Nuclear Engineering Sciences Dept., University of Moride, Gainesville, { FL, hiay,1988.

40. "Liquid Crystalline Phases of DNA," T. Strecleclia, Oral Doctoral Dissertation

( Defense Picsentation, Biology Dept., Florida State University, Tallahassee, FL, hiay, , 1988. ( 41. "Liquid Crystalline Phases of DNA," T. Streeleclia, Doctoral Dissertation, Biology Dept., Florida State University, Tallahassee, FL, hiay,198S. ( 42. "Proposal for Reactor Usage Operations Training Program for Rorida Power ' Corporation Degreed Engineer," W.G. Vernetson, Nuclear Engineering Seier.ces Dept., University of Florida, Gainesville, FL, hiay 17,1988. f

43. "Reply to Notice of Violation, USNRC Inspection Report No. 50 83/88-01," W.G.

Vernetson, Nuclear Engineering Sciences Dept., University of Morida, Gainesville, ( FI, hiay 6,1988. 44

• Pulsed Ion Chamber Diagnostic Studies of Nuclear Seeded hillD Phtsmas," W.ll.

( Ellis and h1J. Baumgartner, Trans. Amer. Nucl. Soc.16, p. 494, June,1988. l l

45. "Effects of Drilling Fluids on Seagrass Communities," W.G. Vernetson and R.

llanrahan, Final Report to Dr. D. Weber and Dr. C. D'Asaro, Nuclear Engineering Sciences Dept., University of Florida, Gainesville, FI, June 8,1988. k IX 4 l l I , . _ _ _ _ _ _ _ _ _ _ _. l

46. "Final Report - Safety Channel 1 Circuit Failure," W.G. Vemetson, Official Submittal to USNRC, Nuclear Engineering Sciences Dept., University of Florida, Gainessille, l FI, June 9,1988.

47. "Final Report - Clutch Current Indicating Lamp Failure," W.G. Vernetson, Official Submittal to USNRC, Nuclear Engineering Sciences Dept., University of Florida, Gainesville, FI, June 23,1988.

48. "lon Loss Characteristics of UF6Gas Mixtures Exposed to Ionizing Radiation," M.

Baumgartner, Oral Presentation Defense of Desis Project, Nuclear Engineering Sciences Dept., University of Florida, Gainesville, FL, June 29,1988.

49. "Nuclear Seeded Plasma Diagnostic Experiments With the PIC System," W.H. Ellis, MJ. Baumgartner, tlatt, in INSPI-QR UF-011, Innovative Nuclear Spam Power l Institute Quarterly Progress Report for Period Ending June 30,1988, University of Florida, Gainesville, FI, July,1988.

[ 50. "UFTR Final Safety Analysis Report Revision 5," W.G. Vernetson, Official Submittal to USNRC, Nuclear Engineering Sciences Dept., University of Florida, Gainesville, FL, June 30,1988.

51. "Analysis of Trace Elements in Coal Slurry Samples," W.G. Vernetson, R. Hanrahan, p Interim Report to Dr. R. Uewellyn, Nuclear Engineering Sciences Dept., University

( of Florida, Gainesville, FI, July 12, 1988.

52. "Results of Followup Activation Analysis of Geologic Quartz Samples for Metals to I include Al, Eu, Fe, Rb, Na and Ti." W.G. Vernetsoa, Report to Dr. A.L Odom and J. Rink, Nuclear Engineering Sciences Dept., University of Florida, Gainesville, FL July 14,19S8.

t

53. "Neutron Activation Analysis for Trace Elements in Volcanic Ash From Mt. St.

r Helens Volcanic Eruption," JJ. Musgrove, summer research project submitted as a participant from Escambia High School in Florida Foundation of Futurc Scientists 1988 Summer Rese.trch Prograra (prepared also for use as a High School Science Fair Project), Nuclear Engineering Sciences Dept., University of Florida, Gainesville, ! FI, August 3,1988. 7

54. "Quantitative Study of Elemental Concentrations in Powdered Milk Using Neutron l Activation Analysis," Charles Crawford, Special Project for Radiochemistry 1Aboratory Course CllS 51101, University of Florida, Gainesville, FI, August 4, 1988.

{

55. "Comparison of NBS Standards Using Neutron Activation Analysis," J.C. Nefflen, summer research project report submitted as a participant from Glades Central

(' Community liigh School in Florida Foundation of Future Scientists 1988 Summer Research Program (prepared also for use as a High School Science Fair Project), Nuclear Engineering Sciences Dept., University of Florida, Gainesville, FI, August { 5,1988. IX 5 l . . . .

i

56. "Neutron Activation Analysis for Trace Elemems in Volcanic Ash From Mt. St.

Helens Volcanic Eruption," JJ. Musgrove, Oral Presentation on FFFS Summer Research Project, University of Florida, Gainesville, FL, August 5,1988. I 57. "Comparison of NBS Standards Using Neutron Activation Analysis," J.C. Nefflen, Oral Presentation on FFFS Summer Research Project, University of Florida, l Gainesville, FI, August 5,1988.

58. "Ion less Characteristics of UF6Gas Mixtures Exposed to Ionizing Radiation," M.

l Baumgartner, Masters' 'Diesis Project in Nuclear Engineering Sciences Dept., I University of Florida, Gainesville, F1, August,1988. l 59. "Final Report on the Summer Semester Reactor Operations-Based Health Physics i I Cooperative Work ' Training Program," conducted for Radiation Protection I Technology Program students at Central Morida Community College, W.G. ) Vernetson, Nuclear Engineering Sciences Dept., University of Florida, Gainesville, l FI, August,1988. l

60. "Master 'Ihesis and Other Research Opportunities Involving the University of Florida Training Reactor," W.G. Vernetson, Oral Presentation to Nuclear Seminar Course (ENU 6935) graduate students entering the Nuclear Engineering Sciences Dept.,

University of Florida, Gainesville, F1, August 25,1988.

61. "Status of Neutron Radiography Development at the University of Florida Training l Reactor," P.M. Whaley, Internal Report, Nuclear Engineering Sciences Dept.. I University of Florida, Gainesville, FI, August 30,1988.

6... "Results of Neutron Activation Analysis of Synthetic Quartz Glass for Chlorine and/or Titanium Concentrations," W.G. Vernetson and R. Hanrahan, Periodic Reports on Sample Analysis to C. Balaban and G. LaTorre, Nuclear Engineering Sciences Dept., University of Florida, Gainesville, FL February 16 and 29,1988; March 7,1988; April 5 and 29,1988 and June 23,1988.

63. "Pulsed lonization Chamber Measurement for Fissile Plasma Characterization," W.H.

f Ellis, Oral Presentation to be presented to the Directortte, innovative Nuclear Space Power Institute (INSPI), Gainesville, FI, September,1988. t l 64. "Development of Neutron Radiography Capabilities at the University of Florida Training Reactor," W.G. Vernetson and P.M. Whaley, paper accepted for presentation at the 1988 TRTR Annual Meeting to be held in Newport, Oregon, October 4-6,1988.

65. "Status of IIEU/ LEU Conversion.for the University of Florida Training Reactor,"

W.G. Vernetson, paper accepted for presentation at the 1988 TRTR Annual Meeting to be held in Newport, Oregon, October 4 6,1988. IX 6

y I (

66. "Pulsed Ion Chamber Diagnostic Studies of Nuclear Seeded Mild Plasmas," W.H.

Ellis, MJ. Baumgartner, W.Y. Choi, J.I. Monroe and J.S. Park, INSPI FR UF 012, [ University of Florida Innovative Nuclear Space Power Institute (INSPI) Final Report for Period September 12, 1985 to September 30,19S8, in preparation to meet reporting requirements of Contract #DNA00185-C-0329 with the Dept. of the Air ( Force, Wright Patterson Air Force Base through the Innovative Science and Technology Office, Strategic Defense Initiative Organization, University of Florida, Gainesville, FL

67. "Computer-Dased On Line Pulsed Ionization Chamber Diagnostic Systems," W.Y.

Choi, J.S. Park, MJ. Baumgartner and W.H. Ellis, summary accepted for publication in the transactions for the 1988 American Nuclear Society Winter Meeting International Conference to be held in Washington, D.C. November 4,1988. k 68. "Evidence for a New Quartz Geochronometer," A.L Odom and WJ. Rink, Chemistry Dept., Eorida State University, Tallahassee, FI, submitted to Malute for publication in late 1988 or 1989. (

69. "Pulsed Ion Chamber Diagnostic Studies of Nuclear Seeded Mild Plasmas," W.H.

Ellis, INSPI University of Florida Paper accepted for presentation to the First { Meeting on Ultrahigh Temperature Reactor and Energy Conversion Research Program, Washington, D.C., November,1988. ( 70. "Multiprobe PIC Diagnostic Studies of Nuclear Enhanced MIID Plasmas," W.Y. Choi, Doctoral Dissertation Research Proposal under preparation for presentation to advisory committee in Fall,1988.

71. "Optimization of the UFTR Neotron Radiography Facility," H. Hicks. ENU-4905 Special Senior Project Report, Nuclear Engineering Sciences Department, University

{ of Florida, Gainesville, FL, (completion expected in December,1988).

72. "Comparison of Argon 41 Effluent Concentration Determinations Relative to

{ Variations in Sample Volumes," B. Murray, Masters 'Diesis Project in Emironmental Engineering Sciences Dept., University of Florida, Gainesti'le, FL, degree expected May,1989. f

73. "A Comparison of Laboratory and Field Conditions on Seagrass Communities Exposed to Drilling Fluids," D. Weber (EPA), C. D'Asaro (UWF), cLal, in preparation for submittal for publication in 1989.

74. "Phase Transitions in Concentrated DNA Solutions at 1.ow bnic Strengths," R. Rill and T. Streeleclia, paper in preparation for submittal to Biopolymers in 1989.

75. "Sodium NMR Study of Sodium DNA Interactions in Concentrated DNA Solutions

( at Low lonic Strengths," R. Rill and T. Streeleclia, paper in preparations for submittal for publication in 1989. l NOTE: This list of reports and publications does not indude the sarious presentations with sisuat aids made for the several doien groups who sisit the UITR cach ) car for tours and demonstrations. l IX 7 I .

l [ ( l l f { l APPENDIX A J NOTICE OF VIOIATION FROM f t NRO !NSPECllON REPORT { NUMBER 50-83/88-01

  . WITII UITR FACILITY LICENSEE RESPONSE l

l l l I l

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( l ENCLOSURE 1 f NOTICE OF VIOLATION ( University of Florida Docket No. 50-83 University of Florida Training Reactor License No. R-56 During the Nuclear Regulatory Comission (NRC) inspection conducted on March 14-17, 1988, violations of NRC requirements were identified. In

accordance with the "General Statement of Policy and Procedure for NRC l Enforcement Actions," 10 CFR Part 2, Appendix C (1987), the violations are listed below

A. 10 CFR 20.201(b) requires the licensee to make or cause to be made such surveys as (1) may be .necessary for the licensee to comply with regulations in this part, and (2) are reasonable under the circumstances to evaluate the extent of radiation hazards that may be present. Technical Specification (TS) 4.2.4(2) requires the Argon-41 (Ar-41) [ concentration in stack effluents to be measured semiannually at intervals i not to exceed 8 months. TS 6.6.1(5) requires a routine annual repor' mvering the activities of { the reactor facility during the previous c .odar year. Each annual operating report shall include a sumary o- the nature and amount of radioactive effluents released or discharged to the environs. The sumary ( shall include an estime.te of individual radionuclides present. if the  ; estimated average relet.se after dilution is less than 25% of the concentration allowed, a statement to that effect is sufficient. Contrary to the above, for the period from September 1,1986, to August 31, 1987, the licensee failed to conduct adequate surveys to evaluate the extent of radiation hazards present in liquid and gaseous effluents released from the facility in that: r 1. For measurements of Ar-41 in gaseous effluents, the gama l spectroscopy detection system was calibrated using a 1,000 cubic cencimeter (cc) matrix calibration standard and sample concentration results were calculated for a 1,000 ce sample volume. The actual ( volume of the sample container utilized to measure concentrations in < Ar-41 gaseous effluents was 1,250 cc. l

2. The lower limit of detection for liquid waste tank effluent analyses,

[ 1.08 E-7 microcuries per milliliter (uci/ml), was greater than 25% of the concentration (4.0 E-7 uti/ml) allowed for release to the sanitary sewer and the individual isotopes present in the effluent were not identified as required by TS. ThisisaSeverityLevelIVviolation(SupplementIV). l . .

                                                          - - - - - - - -                            1

( ( University of Florida Docket No. 50-83 University of Florida Training Reactor 2 License No. R-56 f B. Technical Specification 6.3 requires th:.t the facility shall be operated ( and maintained in accordance with approved written procedures. All procedures and major revisions thereto shall be reviewed and approved by ' the Director of Nuclear Facilities before going into effect. Contrary to the above, for the reporting p triod from September 1,1986, to August 31, 1987, the licensee failed to have the Director of the Nuclear Facilities approve the Radiation Control Technique procedures used to conduct environmental surveillances and effluent release measurements required by TS. ( This is a Severity Level IV violation (Supplement IV). Pursuant to the provisions of 10 CFR 2.201, University of Florida is hereby ( required to submit a written statement or explanation to the Nuclear Regulatory Comission, ATTN: Document Control Desk, Washington, DC 20555, with a copy to the Regional Administrator, Region II, within 30 days of the date of the letter r transmitting this Notice. This reply should be clearly marked as a "Reply to a i Notice of Violation" and should include for each violation: (1)admissionor denial of the violation, (2) the reason for the violation if admitted, (3) the corrective steps which have been taken and the results achieved, (4) the { corrective steps which will be taken to avoid further violations, and (5) the date when full compliance will be achieved. !4here good cause is shown, consideration will be given to extending the response time. If an adequate ( reply is not received within the time specified in this Notice, an order may be issued to show cause why the license shculd not be modified, suspended, or revoked or why such other action as may be proper shculd not be taken. FOR THE NUCLEAR REGULATORY COMMISSION J hfu Nl. f44 4-. Douglas M. Collins, Chief ( Emergency Preparedness and Radiological Protection Branch Division of Radiation Safety ( and Safeguards Dated at Atlanta, Georgia this *1% day of March 1988 [ 1 l i

{ . NUCLEAR ENGINEERIND SCIENCES DEPARTMENT Nuclear Reactor Facility , [ University of Florida - mv s

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o ,- mn - - Pn.no (904) N244N *1st.s $4330 May 6, 1988 United States Nuclear Regulatory Commission Attn: Document Control Desk ( Washington, D.C. 20555 [ Ret Reply to Notice of Violation ( Inspection Report No. 50-83/88-01 Dear Sirt ( This report is divided into two parts to address the two violations cited in Inspection Repurt No. 50-83/88-01. A. Inspection Report No. 50-83/88-01 cites the UTTR facility with a Severity Level IV violation for f ailure to conduct adequate surveys to evaluate r the extent of radiation hazards present in liquid and gaseous effluents l released from the facility in two cases as quoted here: I

1. For measurer.ents of Ar-41 in gaseous effluents, the gamma, spectro-( scopy detection system was calibrated using a 1,000 cubic centimeter (cc) matrix calibration standard and sample concentratton results ,

were calculated for a 1,000 cc sample volume. The actual volume of I the sample container utilized to messare cor.centrations in Ar-41 gaseous ef fluents was 1,250 cc.

                       - 2. The lower limit of detection for liquid weste tank effluent onely-f                                ses, 1.08 E-7 microcuries per mi]If)1ter (pCi/ml), was greater than 25% of the concentration (4.0 E-7 IAi/mi) allowed for release to the sanitary sever and the ladividual isotopes present in the uffluent vere not identified as required by Technical Specific 2rionw.

{ la. Admission or Denial of the Violation l

1) For the Ar-41 measurements, the statem6ht of violation repeated above is admitted; however, this methodology has been con- l sidered conservative. '

2) For the failure to identify individual isotopes present in the liquid effluent, the statement of violation is also admitted,  ;

though no credit for dilution is taken for the radioactivity l { 1evel in the liquid effluent. I hamss.*sb/NWsb, Ace en new

{ U.S. Nuclear Regulatory Commission May 6, 1988 Page ho ( , Ib. Reasons for the Violation "

1) The reason for the violation on Ar-41 measurements is the be-lief that the methodology in use was conservative, though ad-( mittedly not as accurate as possible. The methodology had been reviewed independently by an Environmental Engineering Profes-sor; though he had documented his walk-through review of the

[ Argon-41 measurement considering the procedure adequate, he did I not document any consideration of whether the measurement is conservative.

2) The reason for the violation involving failure to identify in-dividual isotopes in the liquid effluent is that releases are usually less than 5-10 per cent or less of allowed release con-centrations. Such a concentration was probably applicable on the one 1986-1987 release for which we are cited. The 1.08E-7 pCi/n1 value is the lower limit of detection, not an actual quantifled release concentration. The only reason for the

{ quoted 1.08E-7 LCi/ml level in the effluent is that the count time for the sample was shortened representing an increase in the lower limit of detection to the point where the 1.1.D was j { greater than 25% of the allowable release concentration. As a result there was an oversight for the monitoring requirement for specifying individual isotopes for inclusion in the Annual ( Report per Paragraph 6.6.1(5) of the UFTR Technical Specifica-tions. Ic. Correctice Steps Takan/Results Achieved [

1) The next scheduled Argon-41 measurement will not be performed until we have documented analysis to assure the conservatista of l

[ the present methodology or obtained 2 calibration source thac ' more closely models the IMO cc sample containers and performed an appropriate evaluation for its use.

2) No further liquid releases have been mado since the NRC inspec- l tion on Marcl. 14-17, 1988. When releases are next made samples  !

will be counted sufficiently to assure the activity level is i below 25% of that allowed or the contributing individual iso- i topes will be identified. It is worth noting that two liquid wastereleasesin,geptember, 1987 and another in January, 1988 average 2.96 x 10 pCi/ml which is only about 7.4% of the al- i lowable concentration. All three have been well below the 25%  ! cutoff for requiriag identification of individual nuclidea. ( l l 1

L U.S. Nuclear Regulatory Commission May 6, 1988 g Page Rree - i

14. Corrective Steps to be Taken to Avoid Further Violations "

1) The Argon-41 methodology is being reviewed as part of a student project under the direction of the Director of Nuclear Facili-

[ ties. In addition, a new calibration source at 1250 cc is being i ordered and should be available for the next Argon-41 measure-ment due in June, 1988 and required by August, 1988. f 2) The Radiation Control Technique 9rocedure #21 used to control sampling and release of liquii . tuents will be reviewed and approvcd by the Director of Nuclo.. Facilities and it will be { revised to assure all changes to the technique are adequately reviewed prior to implementation before any further releases are made from the holdup ranks. A revised version of Radiation Control Technique #21 is currently under review with approval ( expected by May 31,198J. le. Date When Full Compliance Will Be Achieved (

1) Full compliance has effectively been achieved as of the NRC In-spection, in that certain evaluations of the current methodo-( logy and/or acquisition of a new calibration source will be accomplished before the next Argon-41 measurement due in June, 1988 with a 2 month window allowed. The corrective steps to be f taken to avoid further violations in the monitoring of gaseous l 1 effluents per Section 1d.1) above will be completed by August 31, 1988.

( { 2) The corrective steps to be tak.a to avoid further violations in I the monitoring of liquid effluent releases will be implemented fully by June 15, 1988. B. Inspertion Report No. 50-83/88-01 cites the UTTR f acility for a frverity Level IV violation for f ailure to follow Technical Specification b.3 re-r quiring that the facility be operated in accordance with approved written l procedures. All procedures and major revisions thereto shall be reviewed and approved by the Director of Nuclear Facilities before going into ef-feet. Contrary to the above, for the reporting period from September 1, { 1986, to August 31, 1987, the licensee is cited for failure to have the Director of Nuclear Facilities approve the Radiation Control Technique procedures used to conduct environmental surveillances and effluent re-lease measurements required by Technical Specifications.

a. _ Admission or Denial of the Violation he violation is admitted.

l

s - / U.S. Nuclear Regulatory Commission May 6, 1988 Page Four ( , [

b. Reason for the Violation -

( The manual of Radiation Contr31 Technique Procedures has been de-veloped by the Radiation Control Office to serve the entire Univer-f sity of Florida campus. Some of the Techniques applicable to the UFIR have been in the manual for many years prior to the relicensing of the UFTR in 1962 when the procedures were required to be reviewed by the Director of Nuclear Facilities. As a result, they were grand-( fathered in for the facility and occasinnally updated (improved) by the Radiation Cuntrol Of fice, sometimes based on input f rom t.he UFTR r staff and management. However, because of the historical development ( of these Radiation Control Techniques, they were not formally docu-mented as reviewed by the Director of Nuclear Facilities. The fail-ure to do so is an oversight.

c. Corrective Steps Taken to Date/Results Achieved All applicable Radiation Control Techniques used on a f requent basis

( have been reviewed by the Director of Nuclear Facilities as of May 5,1988 to assure no unreviewed Radiation Control Techniques pro-r cedures are used to support operation of the UFTR facilities. This ( step is assuring that this, violation will not recur.

d. Corrective Steps (e be Taken to Avoid Further Violations

               !ss a grot.p all the Radiation Contre. TecSnhues used to support op-eration of the UFIR f acility are being eviewed by the Director of

[ Nuclear Facilicios and current copiao will then be maintained in a l separate notebook at the UFTR facility with a cover page documenting a dated review by the Director of N>tclear Facilities. In addition, a memorandum of understanding is being generated between the Radiation I f Control Of fice and the Directo of Nucl ear Facilities to assure l changes to these Techniques are ?/eviewed by the Director of Nuciatur Facf.11 ties prior to implementatton to support UFFR operations. ( e. Date of Full Compliance 1 ( Compliance has been achieved via the interim measure noted in Para-l graph (c) as of May 5, 1988. Full compliance with documented review l of all applicable Radiation Control Techniques maintained in a sepa-rate notebook will be achieved by July 31, 1988. I

( U.S. Nuclear Regulatory Commission May 6, 1988 Page Five We trust this response satisfies the requirements delineated in Inspection Re- - port No. 50-83/88-01. If there are further questions, please advise. Sincerely, William C. Vernetson f Director of Nuclear Facilities WGV/ps cc NRC Region II Regional Mr.inistrator P.M. Whaley J.S. Tulenko Reactor Safety Review Subco:nmittee (RSRS) a~- " NotarhPublic { m- k V Datii e er ._ ( n*,m.sm..tno. W "JmWioon fr

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l ( ( k - { { { ( { APPENDIX B { FINAL REPORT TO NRC ON f INTERMITTENT DOWNSCALE FAILURE OF UFTR SAFETY CilANNEL 1 INDICATION I i ( ( ( l /

5 NUCl. EAR ENGINEERING SCIENCES DEPARTMENT Nuclear Reactor Facility , University of Florida . . a

l n v=-esse,0 , ewe .

t -s mu.ueo.. om mu me .cioo m.urs.t.i. sm June 9, 1988 Final Report safety channel 1 circuit Failure Nuclear Regulatory Corai .sion ( Suite 2900

• l 101 Marietta Street, N.W.

Atlanta, Georgia 30323 . f Attention: J. Nelson Grace. Regional Administrator, Region II Re: University of Florzia Training Reactor f Facility bicenset R-56 Docket No. 50-83 Gentlemen: Pursuant to the reporting requirerents of paragraph 6.6.2(3)(c) of the UPTR Technical Specifications, a doacription of a potential abnorcal occurreneo as defined in the UFTR Technical Specifications, Chapter 1 was previously de-scribed in an. interim 14-day report dato3 April 25,198? to includo NRC noti-fication, occurrenco scenario, corrective action and evalustion as well as curront status of the system. This transmittal is inter.ded to constitute a finsi report on the occurrence. Tho potential prompely reportable occurrence involved the recurrence of failure of tb- '.afwty Channel 81 circuit to provide , propor power indication for several secends on April 9,1989 af ter the retarn to normal operacions on April 1, 1983 folic, wing the previous failures on March 15 and 16,1988 (the latter daring a test prior to return to normal opera-tions) per previous re? ort dated March 29, 1988.

                  !!RC Notification Tho Executive Coesittee of tJ.9 Rea,ctor S.tinty Review Subcorsittee reviewed f                   this latest occurrence on April 11, 1938 and concluded that it is a potential abnormal occurrones as defined in UTTR Technical Specifications, Chapter 1 following NRC notification as per Section 6.6.2 of the UTTR Tech Specs earlier on the same day. This notifiestien was carried out by both telephono to Mr.

Paul Durnett and a following telecepy on April it , 1988. In addition to sev-eral discussions to updato Mr. Burnett on 11 April 1989, later conversations with Mr. Robert Carroll and Mr. Paul Frederickson of Projects bavo kept Region

                   !! appricco of reactor status including staged restart with extra monitors in-utalled s sich occurred on April 23-27 and subsequent UrTR return to normal op-f                  crations with an extra staf f mer.ber nonitoring .%foty Channols for all opera-tions until registering 10 hours co rensated operation above 50 kW and timlly a return to norral conitoring conditions on May 20 (irplemented on May 23 1980) with a caution nercrand'.::s issved to operators to rake them avaro that no root cause has been found f or the Saf ety Channel failure ( Attachment 1).

[ Ws twowrAwn. Acw usw

i }. (' Nuclear Regulatory Commission June 9,1988 Page Two { Initial Event Scensrio At 1209 on April 9,1988, with a Reactor Operations Laboratory class (ENU-

           $176L) in progress with power increasing at ~75% power, Safety Channel .1 failed to the bottom meter stop. G.W. Fogle, reactor operator at the controls,

( noted that the indications on Safety Channel 2, the log pen recorder, the wide range iriicator and other indicators were all normal and comenced a reactor shutdown while notif ying the SRO on call who concurred. As power reduction be- { gan, safety Channel 1 returned to norral indication as with the previous fail-ures on March 15 and March 16, 1988. Again the subjective evaluation was that l the return was not instantaneous, but the meter returned to normal indication l relatively slowly over several seconds (i.e., not as if switched on, but rather as if recovering from an electrical transient). Th'e shutdown was com-plated with all instru::ents responding normally at 1210 with the reactor f secured at 1214. Corrective Action Plan ( ror the first. occurrence the rear had been put on administrative shutdown and the full RSRS had est on Mar: 2, 1988 with this event as one item on its agenda. All agreed.the situation being addressed properly although the [ ( exact cause of the event had not p t been identified. Via a series of trouble-chooting and corrective r.sintenance activities, the problem was isolated to involve the fission chaeber, prearp or connections shown in Attachment II which is rigure 1-8 of the LTIR Safety Analysis Report. There was a strong possibility that cleaning connectors on these ecTponents had corrected the problem per conversations with one vencor and concurred with by two UrTR per-sonnel familiar with such instrumentation behavior. As a result, the UPTR was f returned to norrm1 operation on 1 April 1988 following completion of an ap-proved special test procedure. It should be noted that failed noise suppres-sion feedback capacitors have been replaced in both Safety Channels (original-ly thought to be the cause of SC-1 failure) but these were not at fault for I the current failure and, in failed state, have negligible irrpact on circuit ' [ operations because this is a tc amplifier where the feedback coefficient is ( set by a precision resister. Such a failure cou' d have occurred anytime since console ins talla tion. 1 The immediate indications this time were the same as for the previous occur-f' rences - mmely, that an intermittent f ault had developed in the circuitry for Safety Channel 1 (part of the wide range drawer) but not in any other section of 'the wide range drawer. With the reactor secured, Maintenance Log Page #38-f 14 was initiated to investigate and control correction of this failure recur-rence. Although another series of checks was performed, again no root cause could be identified.

i Nuclear Regulatory Commission June 9, 1988 page Three ( [ The recurrence of the Safety Channot 1 failure on April 9,1988, following l about a week of normal operation including 9.65 hours of operation above 50 kw indicates that the Safety Channel 1 fault is intomittent and not isolatable r by the usual test methods of investigation. Therefore, a new program was de- { veloped to isolate and correct the cause of the failures each potential prob-lem is to be dealt with in a systematic mnner followed by a rotest and spe-cial monitoring period prior to restoring the reactor to normal unrestricted f operation. Corrective actions as well as actions to expedite fault isolation are to be taken during each of three possible rajor steps in the mainterunce program. Therefore, the following program was implemented (per isolation of the fault to the connections, preamplifier or fission chamber shown in Attach- { rient II, Figure 1-8 of UTTR SAR) to isolate and correct the fault in Safety Channel 1 with the reactor to be restored to normal operations whenover the test program is successful for each of the following three (3) oteps:

1. Attempt to isolate the internittent failure as external to the console by ,

interchanging S0-1 and 50-2 linear arplifier circuits and change out con-nectors on the wide range drawer and on the prearplifer cables to the f wide range drawer. A criep type connector will be used to replace one clamp type connectors this modification is considered a pocsible fix for the failure while the interchango of amplifiers is only considered an aid { to fault isolation should the failure recur. , 1 [ 2. Replace the prearplifier with one equivalent to that presently in use at t the UFTR according to the vendor except that the replacement item uses one cable connection for the pulsed and the current instrurents while the l currently installed prearplifier uses two. This will a cquire a 10 CFR f 50.59 evaluation to bring both signal lines to a single conne: tor, but is not expected to present any significant difficul'ies technically or ad-ministra tively.

3. Replace the fission charter and its cables / cable connections. The fission chamber (previously, model RSW 314-02552) is a standard item, but not stocked by the current vender General Electric which requires 30 to 60

              & ys lead time. Ef forts are currently underway to obtain a detector from another source within the Department of Energy.

Eva lua tion Except during the transient, the functions of indicatica end trip were not in-hibited or changed; that is, enere was only a temporary loss of indication and f trip function in Safety Channel *1 The impact of this lailure on system op-eration is minimized beeruse it occurs for only a few seconds.

s Nuclear Regulatory Commission June 9,1988 f Page Four This safety Channel #1 Circuit failure is potentiially a promptly reportable ( occurrence per UPTR Technical Specifications, Section G.6.2 delineating re-quirements for Special Reports where Paragraph (3)(c) states certain safety system failures are promptly reportable. Specifically, a special report is ( needed for a "reactor safety system rulfunction that renders the reactor safety system incapable of performing its intended safety function, unless the [ malfunction or condition is discovered during raintenance tests or periods of l reactor shutdowns" or involves components or systems in addition to these re- i quired by Tech Specs. f Similarly one definition of Abnorral Occurrences for the UPTR in Toch Specs section 1.0 is "a ralfunction of a safety system corponent or other component or system rm1 function that could, or threatens to, render the (safety) system incapable of performing its intended safety function." Since Reactor Safety ( System is also defined in Tech specs Section 1.0 to be "a combination.of mea-suring channels and associated circuitry that forms the automatic protective [ actien. to be initiated, or provides inforcation which requires the initiation I of r%nual protective action " the initial and later occurrences of this event rmy not be strictly required to be promptly reported. Basical:y. this event was considered to have no direct impset on saJety and not to impact the health and safety of the public. However, the event was re-ported promptly on April 11, 1988 and later supported by the RSRS reco.T.Tenda-tion on the same day since there was at least a partial failure of the safety system. Nevertheless, safety irplications are negligible shee Safety Channel ,

        #2 was always operable and safety channel #1 was only lost for a few seconds.

Corrective Action - Current Status [ The special test procedure contained in the April 25 Interim Report was used l to control ret: art in March following the first occurrence. Except for an oc-casion when a r-onitoring connector slipped of f necessitating a shutdown to re-connect the device, the original monitored restart on 31 March 1988 was un-eventful with all systems responding prop'erly with no recurrence of the Safety Channel circuit failure. Af ter removing the monitoring instruentation nnd performing a daily checkout during which a spurious noise-induced period trip signal due to wires laying on the prearp was corrected by securing the wires, a final run at full power with no special monitoring instrumentation was con-ducted as the final requirement prior to the first return to norrul opera- [ tions. All systems functioned normally for this run also so with concurrence < by the RSRS (previously granted per the test procedure but reverified) and with NRC Region II verbal notification via telephone conversation with Paul Burnett, the UPTR was returned to norral operations with the problem con-sidered corrceted by the various raintenance activities to check and clean all connections. The recurrence on April 9 negated this declaration as the UPTR was returned to administrative shutdown to correct the cause of the Saf ety Channel failure recurrence.

Nuclear Regulatory Commission June 9, 1988 Page rivo For this recurrence, a modified form of the previous special test procedure was used to support again a staged restart to normal operation begun on 25 April 1988 with dela'yed completion on 27 April 1988 af ter replacement of a failed motor on an Air Particulate Detector. To date only the first of the three program steps listed above under Corrective Action Plan has been found necessary. As indicated in the April 25 interim report, this Special Test Pro-cedure was prepared for RSRS review and approval to allow declaring the UPTR operable'pending successful completion of all norral checks and again per-mitted restart in steps following correctivo and diagnostic maintenance ac-tivities as a test to verif y proper operation of Safety Channel #1 by provid-  ! f ing for continuous. visual monitoring of voltage levels in the linear channel l section of the preamplifier with respect to ground, the current drawn by de-tector operation from high voltage supply and the high voltago power supply output voltage. This procedure again provided compensation for possible recur-  ; { rence of the Safety Channel failure by having a second competent staff member l present in the control room to monitor both safety channels continuously dur- l ing tho entire restart program which included holds at 1 kw for 10 ninutes,10 [ kw for 10 minutes, 50 kw for 1 hour, 75 kw for 10 minutes and 100 be for 1 hour with monitoring devices'in place. This time the return to normal opera- / tions usage of the UTTR was accorpanied by the requirement that the second ( competent individusl be raintained for all operations until 10 hours operation above 50 kw was logged. Af ter successful compiscion of the st ,ed restart begun on April 25 and com-f plated on April 27, 1933, a memorandum ( Attachment III) authorizing UTTR Re-turn to Norral Operations Except for the Oxtra Staff Person Monitoring Safety Channels for all operations was then issued on April 28, 1983 at the UTTR was { declared ready to return to normal operations with only the requirement that a second competent staff person be in the control room to monitor the Safety Channel meters for all operations until 10 hours operation abovo 50 kw had [ I been completed. During the cperatiens to get 10 hours above 50 kw with an ex-tra monitoring individual, normal experimental and training usages of the UPTR were approved and conducted with no recurrence of ssf ety channel failure. This ten hours of operation above 50 kw was completed as of May 19, 1988 as indi-cated in a memorandum (see Attachment IV) dated Msy 20, 1988 f rom the racility Director to Acting Reactor Manager PJ!. Whaley documenting having eet the power requirement and approving the return to uncompensated operations; that is, no extra persen monitoring the Safety Channels. The record of operations above 50 kw af ter May 20 through June 8 is contained in Table 1 as Attachment J

      '/. At this point the corrective action was considered successful and the reac-t       tor declared ready for return to normal operations with normal personnel re-quirements suf ficient for f urther operations but with a caution to operations staf f that no root cause had >wt been found. This return to uncompensated op-f       crations was completed on May 23, 1938 and documented on that date for all operators via a memorandum ( Attschment I) f rom P.M. Whaley acknowledging the return to uncompensated operations but with a caution to operations staf f that no root cause has been identified for the Safety Channel failure.

f i Nt. clear Regulator Commission

  . June 9, t988 Page Six Since May 23 the UPTR has :)een conducting normal operations, with no recur-rence of the Sa fety Channel failure. Since May 20, 1988 the UPTR has operated l   above 50 kw for nearly (9) additional hours (see Attachment V). Based on the successful results of the ' staged test restart with special rnonitoring instru-mentation installed, the coerations with an extra individual monitoring until l'  completing 10 hours operation above 50 kw and the subsequent operations with                                                  j no additional conitoring, the corrective action taken f a ecm idered to have corrected the failure prot lems though admittedly no *' A causO has been found.

At this time the Safety Clannel failure incident f- icnwn.dtvol closed. Further information will be supplied and Region II will y co&Ae should this event recur whereupon anJther step in the Specin .At W: Aedut will be con-sidered necessary. ) [ If further inforeatio*. is needed, please advise. I Sincerely, j ,,/JN y / William G. Vernetson l Director of Nuclear racilities,, WGV/ps Attachments j cc: P.M. W .aley l Reactor Safety Review Subcor.T.ittee b, - h) -l0-Notary Date [ ' Netwy M5i,lted et flodde

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p ATTACIDtD4T I s NUCLEAR ENGINEERING SCIENCES DEPARTMENT Nuclear Reactor Facility University of Florida - l m, :l W.0,Vermeesea,0<. doe , IIWCUA4 M ACfo4 WSOiseo * - s ,mu. mm w.poena up.t.w.suas [ Hay 23, 1988 HEHORANDUM { TO: All UFTR Operators and Staf f { FROM: P.H. Whaley Y" (

SUBJECT:

Saf ety Channel 1 Tes t Program Status [ As of hay 20, 1938 the firs t s tep of the preposed test procedure has been completed with the accu-"lation of 10 hours 23 minutes of run time above 50 kw with no f.ailure of Saf ety Channel 1. Since this 10 hour interval was based on the longest time above ( - 50 kw between f ailures f or Saf ety Channel 1, the successful com-pletion of greater than 10 hours is evidence that the cable re-termination has repaired the Saf ety Channel 1 f ault; neverthe- [ less, the root cause has not been definitely determined except by the absence of a f ailure. Theref ore, all UTTR reactor operators [ and reactor operator trainees are cautioned to be particularly i vigilant of the perf ormance of Saf ety Channel 1 and Saf ety Chan-nel 2 during reactor operations in the power range. { raw /ps , cc: Required Reading . Director of Nuclear Facilities

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NI CHANNEE 1 .

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(LOG N, SAFETY I and . PERIOD) , LOG N

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) NI CFE NEL 1: UFTR Nuclear Instrumentation Channel 1 Diagram . o

                       . Figure 1-8.                                                                                                                                                                             ' -

g 5 i (Log N, Safety fl and Period Channels) . i c - _ _ _ _ _ _

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ATTACllMENT III NUCLEAR ENGINEERING SCIENCES DEPARTMENT a Nucloor Reacter Facility , ,, University of Florida  ; p:i

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April 28, 1988 HEHORANDUM TO: P.M. 'haley FROM:

                                     .A0 W.C. Vernetson'

SUBJECT:

,UTTR Return to Nordal Operations Based upon the successful completion of the special test pro-cedure with the monitoring equipment in place at 1617 hours on April 27 and subsequent removal of the equipment on April 28, 1988 to address the UPTR Safety Channel #1 Circuit railure and prior concurrences by the RSRs Executive and Full Committees as well as NRC Region II (Paul Frederickson), the tTTR is hereby                         ,

authorized to commence normal experimental operations as of 9:00 a.m. today, April 28, 1983. Remember that the second individual as a me:7.ber of our staf f cust be monitoring the safety channel indications for all operations until at least 10 more hours of norral operation above 50 kw have s been cotepleted and until I authorize otherwise. Only upon such successful completion will the tTTR be cleared for return to nor-nul operation with no extra monitoring. . WGV/ps l l l 1 l l 1 b - -- _ -___ _____

NUCL EAR ENGINEERING SCIENCES DEPARTMENT Nucl orRcactorFacility University of Florida - e c.v.. m m. t m ieuctoa m m May 20, 1988 , = ss , . ms rm mn

% pon m a m .t... n ue HEMORANDIAt TO:         P.M. Whaley          g FROM:

yfUP-Y W. C. Ve rne ts on

SUBJECT:

Approval for Return to Normal Control Roon Operations Staffing Requirements - Since the successful conclusion of the UFTR restart on April 27,1988, with special monitors ins talled per the Special Tes t Procedure approved. on April 11, 1988, the ITFTR has conducted power operations above 50 kw for the time in-tervals and on the dates s hown as f ollows : Date Ti=e Total Time (hr-min) 28 April 1009 - 1014 0-5 - 1102 - 1241 1 - 39 1331 - 1431 1 - 00 29 April 1645 - 1700 0 - 15 , 3 May 1214 - 1233 0 - 19 5 May 1220 - 1302 0 - 42 10 May 1538 - 1545 0-7 1628 - 1659 0 - 31 , 1 11 May 1603 - 1626 0 - 18 l l 12 May 1500 - 1703 2 - 03 ) 1 13 May 1613 - 1622 0 - 09 l 1655 - 1755 1 - 00 l 16 May 1402 - 1602 2 - 00 18 May 1612 - 1627 0 - 15 Total Tice Above 50 kw: 10 hr. 23 min. ' An extra monitoring s taf f member has been on duty to conitor the two saf ety channels during all of these operattor Since the saf ety channel f ailure loss of indication and signal has not recur.<d and since this 10.38 hours of opera-tion above 50 kw neets the condition set in the Special Tes t Procedure per NRC and RSRS commitments of greater than 10 tours above 50 kw with no f ailure, the UTTR is now approved to continue normal operations with the usual control room and other staff requirements. The requirement f or the separate individual to monitor the saf ety channels is hereby ended. t u o m ,v e n w .Ac w %

7 ATTACHMENT V TABLE 1 UFTR POWER (~> 50 KW) OPERATIONS SINCE 20 MAY 1988 THROUCR 8 JUNE 1988 Date Time Total Time (hr-min) 21 May 1734 - 1812 0 - 38 23 May 1044 - 1120 0 - 36 1632 - 1700 0 - 28 25 May - 1133 - 1322 1 - 49 26 May 0915 - 0930 0 - 15 ( 1228 - 1428 2 - 00 27 May 1507 - 1522 0 - 15 31 May 1214 - 1224 0 - 10 2 June 1737 - 1752 0 - 15 3 June 1439 - 1443 0-4 1638 - 1642 0-4 6 June 1244 - 1256 0 - 12 1316 - 1331 0 - 15 f 1858 - 1930 0 - 32 j 8 June 1433 - 1438 0-5 1717 - 1838 1 - 21 . 1 70TAL............................. 8 - 59 l blbzz)lk bu &/W

                                              ~

Facility Director (/ Ifa te l . _ - _ _ _ _ _ _ _ _ _ _ - . _ _ _ _ _ . .-

( (. l l I {' I l ( APPENDIX C l UFTR TECIINICAL SPECIFICATIONS {  ! APPROVED AMENDMENT 17 PAGES WITII ( NRC SAFETY EVALUATION REPORT I {

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     /          o                             UNITED STATES
, 8 ' ) - (f[',,p,                  NUCLEAR REGULATORY COMMISSION                  RECEIVED W O 3 g

g. f;E W ASHING TO N, D. C, 20555 t,pril 27, 1988 Docket No. 50-83 Dr. William G. Vernetson Director of fluclear Facilities 102 Nuclear Reactor Building Department of Nuclear Engineering Sciences University of Florida Gainesville, Florida 32611 (

Dear Dr. Vernetson:

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SUBJECT:

ISSUANCE OF AMENDNENT NO. 17 TO FACILITY OPERATING LICENSE NO. R TECHNICAL SPECIFICATION REVISIONS The Ccm ission has issued the enclosed Amendment No. 17 to facility Operating License No. R-56 for the University of Florida Training Reactor. The anendment consists of changes to the Technical Specificaticos (TS) in response to your application dated June 2, 1987 and as supplemented on itarch 7, 1968. The amendment consists of a revisien to your TS to permit you to conduct certain activities when the reactor is shut down, the reacter vent systen is { secured and the stack nonita' is reading greater than 10 counts per second. Also, the TS have been revised to include a hackup reans fer quantifying the radioactivity in the effluent during abnormal or emergency operating f conditions ir eddition to administrative changes. A copy of the related Safety Evaluation supporting Amencrent No. 17 is enclosed. f Sincerely,

                                                     /       v In<. $. ))      a
                                                   ' Theodore S. Michaels Project llanager Standardization and Non-Power                  i Reactor Project Directorate                  l Divisien of Reactor Frojects !!!, IV,          i V and Special Projects Office of Nuclear Reactor Regulation

Enclosures:

l 1. Amendment No. 17

2. Safety Evaluation l

l / ( - .

University of Flcrida Decket No. 50-83 cc: Mr. Jares S. Tulenko, Chairman Department of Nuclear Engineering Sciences University of Florida College of Engineering 202 Nuclear Sciences Center Gainesville, Flcrida 32611 Bureau cf Intergovernmental Relations 660 Apalachee Parkway Tallahassee, Florida 32304 l ( ( l l l l

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g UNITED STATES NUCLEAR REGULATORY COMMISSION WASHING T ON. D. C. 20555 a,.....j l I l ? UNIVERSITY OF FLORIDA DOCKERT NO. 50-83 j AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 17 License No. R-56

1. The Nuclear Regulatory Comission (the Comission) has found that:

A. The application for amendment to Facility Operating License No. R-56, filed by the University of Florida (the licensee), dated June 2, 1987 as supplemented on March 7, 1988, complies with the stardards and requirenents of the Atomic Energy Act of 1954, as amended (the Act), and the Ccenission's regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Comission; J C. There is reasonable assurance: (1) that the activities authorized by this amendeent can be conducted without endangering the health l and safety of the public, and (ii) that such activities will be  ; conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter It D. The issuance of this anendment will net be inimical to the comon defense and seca ;ty or to the health and safety of the public; E. The issuance of this amendeent is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirenents have been satisfied; and F. Publication of notice of this amendtrent is not required sirce it does not involve a significant hazards consideration nor amendment of a license of the type described in 10 CFR Section 2.106(a)(2). / l L _ - - - - - - - - - - - - -

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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.(2) of Facility OperatinD License No. R-56 is hereby arended to read as follows:

(2) Technical Specifications t The Technical Specifications contained in Appendix A, as revised I throughAmendmentNo.17,areherebyincorporatedinthelicense. The licensee shall operate the facility in accordance with the Technical Specifications, f 3. This license amendment is effective as of the date of issuance. FOR THE NUCLEAR REGULATORY CO WISSION j - ,- cm g.fifi lwd & O Lester'S, Rubenstein, Acting Director Standardization and Non-Power Reactor Project Directorate Division of Reactor Projects III, IV, Y and Special Projects Office of Nuclear Reactor Regulation

Enclosure:

Appendix A Technical Specifications Changes ( Date of Issuance: April 27, 19S8 ( l ( l l l I . - - - - - - - -

ENCLOSIRE To LICEf;SE AMENDitENT NO.17 FACILITY OPEATING LICENSE N0. R-56 DOCKET NO. 50-83 r Replace the following sages of the Appendix A Technical Specifications with the enclosed pages. Tie revised pages are identified by Amendment number and contain vertical lines indicating the area of changes. l Remove Pages Insert Pages 10 10 11 11 12 12 l l l l ( ( p I l I l t -

{. range drawer. Tha wid3 rango decwor provid:s protcetion during ctortup thrcugh tho sourca count rato intericek (2 cp3), 10-s:c period inhibit and tho 3-s:c peritd trip. The primary cnd cec:ndary coolont ficw rato, temp;roturo and lovel s:naing instrumonta-tiCn provides information and protection over the entire range of reactor operations and 10 proven' to be conservative f rom a saf ety viewpoint. The key switch prevents unsuthor-ized cperation of the reactor and is an additional full trip (manusi scram) control a-v;ilible to the operator. The core level trip provides redundant protection to the pri-mary flow trip. The core level trip acts as an inhibit during startup until the minimum coro water level is reached. . 3.3 Reactor Vent System Theso specifications apply to the equipment required for controlled release of gaseous ra'dioactive effluent to the environment via the stack or its confinement within the reactor cell. 3.3.1 Specifications (1) The reactor vent system shall be operated at all times during reactor operation. In cddition, the vent system shall be operated until the stack monitor indicates less  ; than 10 counts per second (eps) unless otherwise indicated by facility conditions j ta include loss of building electrical power, equipment failure, cycling console f power to dump primary coolant or to conduct tests and surveillances and initiating l the evacuation alarm for tests and surveill.nces including emergency drills. The i reactor vent system shr.11 be ice.ediately secured upon detection of: a failure in I ( the monitoring system, a failure of the absolute filter, or an unanticipated high  ! stack count rate. The reactor vent system shall be capable of maintaining an air flow rate between 1 I {(2) cnd 400 cfm from the reactor cavity whenever the rasctor is operating and as speci- I fied in these Technical Specifications. } (3) The diluting fan shall be operated whenever the reactor it in operation and as ctherwise specified in these Technical Specifiestions, c.t an exhaust flow rate 1crger than 10,000 cfm.

                                       .c (4)   The air conditioning / ventilation system and reactor vent systems are automatically chut off whenever the reactor building evacuatiun alarm is automatically or ranual-f            ly actuated.

(5) All doors to the reactor cell shall normally be closed while the reactor is operat-( ing. Transit is not prohibited through air lock and control room doors. (6) The reactor vent system shall have a backup e.eans for quantifying the radioactivity in the ef fluent during abnoreal or e .orgency operating conditions where venting [ could be used to reduce cell radionuclide concentrations for ALARA considerations. 3.3.2 Bases Und:r nornsi et 2tions, to ef fect controlled release of gaseous activity through the f reactor vent system, a negative cell pressure is required so that any building leakage ' l will be inward. Under nor:ml shutdown conditions with significant Argon-41 inventory in i the reactor cavity, operation of the core vent system prevents unnecessary exposure from ' gss leakage back into the cell. Under emergency conditions , the reactor vent system will j f be chut down and the daeper closed, thus minimizing Icekage of radioactivity from tho 6 reacpr cell unless venting is required. Amend:sent 17 I  % - - - - - - - - - -

3#4 Radi9tirn Monittring S/ stems and Radioactivo Effluents 3.4.1 Area Radicticn Mtnitors The reactor cell shal) be monitored by at least three area radiation monitors, two of which shall be capable of audibly warning personnel ci high radiation levels. The output ( cf et least two of the monitors shall be indicated and recorded in the control room. The l setpoints for the radiation monitors shall be in accordance with Table 3.3. 3.4.2 Argon-41 Discharge Tha following operational limits are specified for the discharge of Argon-41 to the en-(vironments (1) The concentration

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of Argon-41 in the gaseous effluent discharge of the UrTR is de-termined by averaging it over a consecutive 30-day period. f(2) The dilution resulting from the operation of the stack dilution fan (flow rate of 10,000 cfm or more) and atmospheric dilution of the stack plume (a factor of 2001 may be taken into account when calculating this concentration. (3) When calcugated as above, discharge concentration of Argon-41 shall not exceed MPC (4.0 x 10' pe/ml). operation of the UrTR shall be such that this maximum permis-sible concentration (averaged over a month) is not exceeded. Table 3.3 Radiation Monitoring Systee Settings No. of Required Type operable Functions Alam(s) Setting Purpose f Area Radiation 3 detecting 5 mr/hr low level Detect /ala m/ record Monitors 2 audio alarming 25 mr/hr high level low and high level 2 recording external radiation ( Air Particulate 1 detecting Range adjusted ac- Detec t/ alarm / record Monitors 1 audio alaming cording to APD* type airborne radioactivity 1 recording (according to moni- in the reactor cell ( toring requirements) [ Stack Radiation 1 detecting (1) Fixed alar i at Detect /alam/ record i Monitor 1 audio alaming 4000 eps release of gaseous 1 recording (2) Adjustable alam radioactive effluents as per power in the reactor vent ( I level duct to the environs

• Air particulate detector

( NOTES: For maintenance or repair, the required radiation monitors may be replaced by suitable portable instaments provided the intended function is being accomplished. Service, calibration, and testing interruptions for brief { periods are pemissible when the reactor is not in operation. 3.4.3 Reactor Vent / Stack Monitoring System (1) Whenever the resetor vent system is operating, air drawn through the reactor vent system shall' be continuously monitored for gross concentration of radioactive gases. The output of the monitor shall be indicated and recorded in the control l room. f (2) Whenever venting is to be used to reduce cell radionuclide concentrations during ' abnomal or emergency conditions, then the radioactivity in the ef fluent must be l quantified prior to initiating controlled venting. I JUL- - - - - - - - - - - - - - - . _ - - - - - - - - - - - - - - - - - - - - - - - - - - - -

U* *

 '(3)   Tho react:r cir c;vity flew chall bo p;riodically cnalyzed to minicizo Argon-41 r0-leasco to tho cnvironmrt whilo raintaining o nscativo pressure within tha reactor cavity to cinicizo radh ctivo h zords to reactor porconnol.

3.4.4 Air Particulate Monitor { Tho reactor cell environment shall be monitored by at lesst one air particulato monitor, c pable of audibly warning personnel of radioactive particulate airborne contamination in the cel' stmosphere. 3.4.5 Liquid Effluents Discharge j (1) The liquid waste f rom the radioactive liquid waste holding tanks shall be sampled I and the activity measured before release to the sanitary sewage system. feleases of radioactive liquid waste from the holding tanks / campus sanitary sewage f(2) mystem shall be in compliance with the limits specified in 10 CFR 20, Appendix B. Table 1, Column 2, as specified in 10 CPR 20.303. ( 3.4.6 solid Radioactive Waste Disposal solid radioactive waste disposal shall be accomplished in compliance with applicable . { riorids. regulations and under the control of the Radiation Control Of fice of the University of 3.4.7 Bases I Tho crea radiation monitoring system, stack monitoring system and air particulate detec t:r provide informtion to the operator indicating radiation and airborne contamination ( levels under the full range of operating conditions. Audible indicators and alarn lights j indicate (via monitored parameters) when corrective operator action is required, and (in a warning light indicates situations recomend- [ the case of the area radiation ronitors) inq cr requiring special operator attention and evaluation. Argon-41 discharges are lim-ited to a monthly average which is less than the unrestricted area limit, and liquid and Colid radioactive wastes are regulated and controlled to assure ecmpliance with legal requirements. ( , 3.5 Limitations on Experiments Applicability: These specifications apply to all experi ents or experimental dsvices installed an the reactor core or its experimental f acilities. Objectives: The objectives are to raintain operatio .al safety and prevent damage to tho' reactor facility, reactor fuel, reacter core, and associated equipments to prevent ex-ceeding the reactor safety limits; and to minimize potential hazards f rom experimental f devices. , specifications: ( (1) General The reactor ranager and the radiation control officer (or their duly appointed re- { presenta tive) shall review and approve in writing all proposed experie.ents prior to their perfortunce. The reactor canager shall refer to the Reactor Safety Review Subcommittee (RSRS) the evaluation of the safety aspects of new experiments and all changes to the facility that e.sy be necessitated by the requirements of the experi-ments and that may have safety significance. When experiments contain substances that irradiation in the reactor can convert into a caterial with eiignificant , l l [ Amendment 17 _ _ _ _ _ _ _ _ g l

Sa ma g UNITED STATES ( [#' ( 'k vygf E NUCLEAR REGULATORY COMMISSION WASHING TON. D. C, 20555 (ig .. .... SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ( SUFFORTING AMENDiiENT NO. 17 TO FACILITY OPERATING LICENSE NO. R-56 UN!YERSITY OF FLORIDA DOCKET NO. 50-83

1.0 INTRODUCTION

( By letter dated June 2,1987, the University of Florida (licensee) requested an arendment to their Technical Specifications (TS) for the University of Florida Training Reactor. The atendrent would permit securing the reactor vent system when stack counts are above 10 counts { per second (eps) under certain non-erergency conditions. r The need for the TS chance was discovered by the licensee during a l quarterly evacuation drill (Decerber 11, 19E6 and dccumented Decenber 19,19t>6) when two area radiatio'n nonitors were set at a high level trip set point, which secured the reacter vent systen and sounded the evac-( uation alarm as required by TS. Securing the reacter vent system, however, above 10 cps under nor-e?.ergency conditices is not permissible by the TS. The stack ecunt rate at the time was 300 cps. Therefore, the licensee proposed to clarify the TS to permit securing cf the reactor vent ( system under certain ren-erergency conditions. In addition the licensee proposed certain administrative changes to the TS. { The licensee's prepesed changes were reviewed by Region II in a menorandum dated January 22, 1988 (D. Verre111 to T. Michaels) and the licensee respended to further suggested changes on March 7, 1988 , [ l 2.0 EVALUATION The licensee has cutlined, in the March 7, 1938 letter, the conditions ( under which the reactor vent can be secured above 10 rps. These J conditions are (1) loss ef building electrical pewer (2) equiprent [ failure (3) cycling console pewer to dump prinary coolant or to ( conduct tests and surveillances, ard (4) initiating the evacuation alarm for tests and surveillances including emergency drills. Each of these conditions would be applicable when the reacter is shut down. Also, for ( cenditions 1, 2 and 4 there is no technical basis for requiring eperation of the Reactnr Vent System at stack count rates greater than 10 cps. When the core vent systen is secured, any effluent that would be released ( is contained within the cere/ reactor vent systen eith the only potential release path being backflew (diffusion driven) into the cell. The I licensee's calculations (see June 2, 1907 letter, rege 2) show that the [ Argon-41 cor. centration in the cell air space is less than 10 CFR 20 l restricted area concentration limits. These calculations assure all full I power, equilibriun Argon-41 in core voids w s instantaneously released into the cell air. Additicrally, existing constraints to maintain l l

                                                  .g.

Argon-41 discharge within effluent limits will autenatically prevent exceeding both restricted and unrestricted area concentration litrits, if such excesses were possible. The licensee observed on December 11, 1966, after the core vent fan was secured ar.d with a high stack count rate of 200-300 cps, that no increases in Air Particulate Detector level or Area Monitor indications resulted. For condition 3, the interruption of power to the console and the securing of the Reactor Vent Systen is usually only momentary and in such a time frame, there is no cause for concern about back leakage of stack effluents into the cell er control reem. The staff finds that the revisions to the TS (Section 3.3.2(1)), which pern\- securing the vent fan above 10 cps for the conditiens previously outlired, are acceptable. The licensee plans to install a backu) means to cuantify radioactive effluents to the environment during a> normal operating conditions such as when the vent moniter is inoperable or the absolute filter fails. Sections 3.3.1(6) and 3.4.3(2) have been revised to reflect this change, l which increases the safety of the f..ility, and is acceptable. Other changes to TS 3.3,3.4.3(3) and the addition of 3.4.7 are administrative; they improve the TS ard are acceptable.

3.0 ENVIRONMENTAL CONSIDERATION

( This amendrent involves changes in the installation or use of a facility components lecated within the restricted area as defined in 10 CFR Part 20 and changes in ins:ection and surveillanct requirerents. The staff ( has deternined that tie atendrent involves no significant incretse in the l amounts, and no significant change in the types, of any effluents that nay be released offsite, and there is no significant increase in t individual or curulative cccupational radiation exposure. Accordingly, l this amendment reets the eligibility criteria for categorical exclusion set fcrth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), ro envirennental impact stateeent er envircr. mental assesstent need be ( prepared in correction with the issuance of this atendment.

4.0 CONCLUSION

The staff has concluded, based on the considerations discussed above, that: (1) there is reaserrble assurance that the realtn and safety of the r public will not be endangered by the operatier in the proposed nanner, and ( (2) such activities will be corducted in corpliance with the Comission's regulations ard the issuance of this enerdrent will not be ininfeal to the cerron defense and security or the health and safety of the public, f Principal Contributor: Theodcre S. Plichaels Dated: April 27, 1988 ( (

I I ( APPENDIX D UFTR SAFETY ANALYSIS REPORT ( REVISION 4 DOCUMENTATION l ( ( {

F .. NUCLEAR ENGINEERIND SCIENCES DEPARTMENT Nuclear Reactor Facility , . , University of Florida - T As

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• e.wa.,n.,tes ami pm punesano.wours UPTR SAR Revision 4 Septeraber 25, 1987 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 ATTNt Document Control Desk Ret University of riorida Training Reactor racility License R-56, Docket No. 50-83 centlement The enclosed package contains Revision 4 prges for the UrTR Safety Analy-sis Report dated January, 1981 submitted as part of our relicensing ef fort.

Revision 4 consists of two pages (9-10 and 9-13) in Section 9.5 (other Auxi-liary Systems) pertaining to Fire Protection and Co..munications Systems. These changes are not considered to affect the UPTR Safety Analysis as outlired be-low. On Page 9-10 under Section 9.5.1 rire Protectien system the claimed num-ber of Coy extinguishers available in the reactor room is two, not five as previously indicated. The number of coy extinguishera in the reactor room has always been two (2). The claim of five (5) on Page 9-10 is attributed to a typing or other error as there are normally rnore than five CO2 extinguirhers available when one considers the entire first flocr of the reactor building which is where the error probably derived. With two (2) CO2 extinguishers in the reactor room, one in the control room, one ir.7ediately outside the control room and several others placed throughout the ground floor, there is no reduc- { tion of effective fire suppression capability frcm Coy extinguishers, j Second, on Page 9-13 in Section 9.5.1 the description of the new four-( zone fire alaru system and its minimum clained installed equipment is in-cluded: this system replaces the old two-zone system and was installed per recommendations resulting from inspections by American Nuclear Insurers. In-( stead of a two-zone system (reactor cell / control room and rensirider of build-ing with rnonitoring station outside the control room), the new four-zone sys-tem (reacter cell / control room, downstairs laboratories, upstairs labora-tories /of fices and annex laboratories /of fices) better delineates where a fire ( is located upon alarm at a new renitoring station that is now located just outside the downstairs exit of the building at the Ecergency Response Center l used for addressing radiological, fire and other building emergencies. In i i general, the new systen is a better system with more zones, rure datectors and l more pull stations making it a much more ef fective automatic fire alarm system. 1

U.S. Nuclear Requiatory Commission Septe cer 25, 1987 f Pane Two Finally, on Page 9-13 in Section C 5.2 Communications Systems, the first paragraph is changed to reflect ade.inistrative titles used elsewhere in the j SAR and to include Health Physics Of fico as one of the entities to be reached ( by ti.e full service telephone at the reactor console. The second paragraph is changed to reflect administrative titles and to delete the claim the Health Physics Office is connected to the control room by intercom. The Health Phy-sics Of fice is in the Nuclear Sciences Center (attached adjacent building) and can be reached by telephone with rapid response as necessary, Qualified health physics technicians (staff operations personnel) are available through the in-tercom system connection in the operations staff room. The claim that the in-tercom connects to the Health Physics Of fice at this locatim da' as to over ten years ago and was inadvertently included in the 1931 revised SAR. This j change is not considered to affect Health Physics capabilities or response and l does not impset safety analysis for the UTTR. All text changes to the current revised SAR are clearly indicated by ver-tical lines in the rargin. All of these changes for Revision 4 have been fully f reviewed by UrTR Management and the Reactor Safety Review Subcom.mittee and are not considered to relax the requirements for assuring protection of the health and safety of the public nor are they considered to impact the UTTR Safety { Ana lysis. In addition the changes are not considered to involve any unreviewed safety question per 10 CFR 50.59. The entire enclosure consists of one (1) signed original letter of trans-mittal with enclosure plus ten (10) copies of the entire package. f If furthur inforration is required, please let us know. Sincerely. I [ G#d4 Uilliam G. Vernetson l Associate Engineer and Director of Nuclear Facilities ( WGV/ps Enclosures cc: U.S. NRC Region II f P.H. Whaley Reactor Safety Review Subcom.mittee sa ti $ ~ ~

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/ k { ( . ATTACilMr.NT I FINAL SAFCTY ANALYSIS RI: PORT f UNIVCRSITY OF FLORIDA TRAINING RP. ACTOR F ACILITY LICENSC R-56; DOCKCT NLHut:R 50-83 Revision 4 The attachext Revisien 4 pages revise the t'niversity of Florida Training Reac-f tor Final Safety Analysis Report as of Septeber, 1987. Revision 4 pages should be substituted to replaec existing pages as follows:

                                                  *  . .s ion 4 F:r ola cement Pa ges Previous Paces

( 9-10 (RCV 1 5/82) 9-10 (Rr/ 4, 9/67) 9-13 (on!CINAL) 9-13 ( rd"/ 4, 9/87 ) NOTE: All Revision 4 changes to existir.q pages are clearly delineated by ver-tical lines in the c.argin adjacent es the char.ge. Both pages are labeled as Revisien 4 at the botto . l I l t 1 I j

9.4.2 Core Vent System As indicated in Section 9.4.1, in order to present radioactive gases and particulate matter formed in the reactor from escaping into the reactor room, the air surrounding the reactor core structure is withdrawn by the core vent system and then through a rough and an absolute filter. The air is then discharged through the stack where it is diluted with about 12,000 c.f.m. of outside air before it is released to the atmosphere. Vacuum breaker vent lines (l" dia:neter) connect the tops of the fuel boxes to the coolant storage tank to provide a'n air-return path allowing rapid dumping of the water from the boxes. The coolant storage tank vent connection to the reactor ventilation system is shown in the diagram of Figure 9-4 giving a vertical section view of the physical arrangement of the UFTR Core Vent System. The vent lines are positioned between the gr.1phite blocks that surround the fuel boxes and tha concrete shield tang. A schematic flow diagram of the core cooling and vent system is presented in Figure 9-4 On-line measurement of the vent flow rate is acco'nplished by a pitot tube in the outlet line of the core vent. A differential pressure, pro-l portional to the squara of the flow rate. is displayed on inclined nano-I maters on the north wall of the reactor. The differential pressure across the rough filtt' is indicated by another inclined minometer, anc the dif fer-ential pressure across the absolute filter is indicited by a "Magnehelic" f gauge. These three instruments display differential pressure in inches of water head. ( Gamma activity of the gaseous of fluent release is monitored by a GM detector located on the dcanstream side of the absolute filter af ter the pitot tube (see Figure 9-5) at the base of the stsc< before dilution occurs. An audible alarm will be actuated in the control ro'n, in the event the vent { flow activity reaches a preset level. The data fron this manitor is continV-ously recorded. I., the exhaust auct there is a mato" opened, spring-closed damper valve which autonatically closes whenever the fan is not operating. The Reactor Vent System prevents diffusion of radioactive gases or particulate matter irto the reactor room during reactor operation. Loss ( nf electrical power to eitner the r! actor vent daner or the dilution faa motor will result in a reacter trip without du, ping primary water. The vent damper is electrically interlocked with the dilution f an motor control circuit 50 that the camper control cannot be opened unless the dilution fan { is energized. This inter'3ck prevents the dischaage of undiluted air effluent via the stad , f 9.5 Other Auxiliary Systms 9.5.1 Fire Protection System Since none of the m3terials of construction of the reactor are inflam-mable, and since the reactor building is fireproaf construction and will ( not be used for storage of quantities of inflarnable materials, a fire of \ any consequence 1*, considered very unlikely. t Conventional fire equipment is located in the reactor cell and through-( out the reactor building. Two C0 extinguishers are available in the reactor room 1,self, and one more is loca$ed in the contral room at the control ( 9-10 Rev. 1, 5/82 Rev. 4, 9/37 ( _ _ - _ _ - _ - - _ _ - _ _ _ __

consoleo A fire hose ant fire extinguisher are also loca%ed outside the control room in the ground floor foyer area referred to as the Limited Access Area in Chapter 3 of this report. An automatic fire alarm systen monitors the reactor cell and the remainder of the reactor building continuously. The system used is a four-zone systen with local monitor'ng and control station. The system is completely supervised with einergency battery back-up. Minimum equipment installed includes:

1. Three (3) lonization Detectors

2. Two (2) Thennal (Heat) Detectors

3. Seven (7) Pull Stations 4 Six (6) Horns Remote supervision is performed by University Personnel. Operation of this system will turn on the emergency light in the reactor rocin (for illumination).

9.5.2 communications Systems A full service telephone is installed within easy reach of the reactor I operator at the console. This provides direct comunication within the building on and cff-campus including: Facility Director Reactor Manager, Radiation Control Of fice. Health Physics Uf fue, University of Florida Police l Department. Gainesville Fire Depart.nent anu Senior Reactor Operator. An into

• System is set up providing direct comunication f rom reactor ,

console to the Reactor Manager and 5 nior Reactor Opeastor (not present in effect). l In case of a po.er f ailure, the telepnene will ce available for comuni-cation within the building a> well as on and off-campus. 9.b.3 Lionting System The reactor building is provided with everhead fluorescent lighting. Additional supplementary lignting is possible via 115v wall outlets. In case of a power f ailure. cree gency lighting is provided automati. cally throughout the building by the e w rgency diesel generator located i outside the reactor building. l 9.5.4 Diesel Generator Fuel Oil Storage and Transfer System ( The diesel generator is a Turbo-Charge 0-6 Cr,terpillar type tsc erator and is available for emergency conditions in case of a power f ailure. The system is designed to com on line automatically within 10 seconds af ter the power f ailure, operating 10 to 11 minutes af ter power recovery, as a back-up [ power supply in case of repeated f ailure within this short period of tire. The automatic starting system provides for three start-up events within a 90 l second period, af ter which it goes into a manual stand-by condition with the l option of a manual start-up or a reset mechanism for start-up. Fu?) oil storage provistens consist of an unterground tank with a capa-city of approximately 2000 gallons. Fuel oil transfer is accorplished by 9-13 i _ __ _ _ _ _ _ _ _ _ R@n en @/87 )

{ \ i APPENDIX E UFTR SAFETY ANALYSIS REPORT REVISION 5 DOCUMENTADON l I i ,

n . NUCl. EAR ENGINEERING SCIENCES DEPARTMENT Nuclear Reactor Facility University of Florida . x n ec.v . o., . hucLLA4 kl ACf04 8J ADWG ' o m r m usu June 30, 1988 "8 5

• n.a. c,co m.un .r.w som U.S. Nuclear Regulatory Commission Was hing ton , D.C. 20555 Attn: Document Control Desk Re: University of Florida Training Reactor Facility License: R-56, Docket No. 50-83 Revision 5, Saf ety Analysis Report l

                         -       '.ne n :                                                                              l e enclosed package contains Revision 5 pages for the UF!R Saf ety Analy-ort dated January,1981 submitted as part of our r*1icensing effort.

a 5 consis ts of seven pages to iaciude pages 14, 1-5, 3-6, 4-9, 7-1, e t.d 15-2. These changes have been reviewed bv 0 FIR management and the UFTR N sy Revieu Subcommittee and are not consideied to involve any unreviewed ,

                              .ty question or to impact the UFTR Saf en Analysis as outlined below; all                i sext changes are denoted by vertical lie.es in the right hand margin of the attached af f ected replacement pages. *.easons for all changes are explained in the f ollowing paragraphs.

On Page 1-4, several experimental facilities are better delineated and the reactivity worths and s hutdown cargin with the mos t reactive control blade out have been updated based on the lates t measurements made in February,1988. On Page 1-5, a typogr:phical error is corrected in the fits e paragrap;i. In the second paragraph, the purpose and f unctioning of reactor vent system is better described than in the original description. In the next to las t para-graph, the f act that the UC1.A and VPI reactors are being decommissioned is l noted. On Page 3-6, the third paragraph is changed to indicate the proper loca-tion of the emergency personnel exit in the Icf t-hand panel of the f reight door, not the right-hand panel as viewed f rom inside the cell. Several typo-graphical and grammatical errors are also corrected in this paragraph. Table 4-1 on Page 4-9 is updated to reflect present UFTR characteristics and to correct several typographical errors such as use of a 1.0 curie PuBe source, not the previously lis ted 10 curie PuBe source per UFTR Tech Spees. Other changes include approximate values on maximum thermal flux and excess core reactivity, approximate current f uel loading, current flow rates and equilibrium core inlet / outlet teeperatures and the control binde reactivity worthi noted previously as changes on Page 1-4. On Page 7-1, Section 7.2.1 is changed to reflect ins trumentation op-eration in the UFTR console. This section previously indicated all "ins trumen-tation contained in the console accepts or sends signals f rom or to the con-trol rod drives, the reactor interlock system, and various detectors and transducers loested around the reactor core and the reactor coolant sys tem."

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U.S. Nuclear Regulatory Commission June 30, 1988 Page Two Since the panel contains several other indicators such as a clock and door controls in place prior to relicense submission in 1981 and the energization switch and communication wire for the pneumatic-operated rapid aample inser- ) tion sys tem added since relicensing, this change simply. provides a correct, up-to-date console ins trumentation description. Also on Page 7-1, Section 7.2.1, three items are added to the lis t of control and indicating ins trumentation to include a digital clock replacing a previous mechanical analog clock, a PuBe source alarm indicator and the ener-gization switch and communication line for the pneumatic-operated rapid sample insertion sys tem. For the firs t two items all is essentially unchanged since the relicensing except for replacenent of the analog clock with a digital clock. Both the analog clock and the PuBe source alarm indicator were in place during relicensing in 1981 but were inadvertantly omitted f rom the list. As noted above the energitation switch and communication line f or the pneumatic-operated rapid sample insertion system represent a later addition which was f ully reviewed prior to implementation. 1 On Page 9-1, the firs t line of Sec tion 9.2.1 is changed to correct a l typographical error and specif ying a 3.0 ton crane, not a 30.0 ton crane I availabic f or use in the reector cell. l Finally on Page 15-2, the firs t paragraph in Section 15.1.1.1 is changed l to correct several typographical errors including the unnecessary repetition i of the next to las t independent clause in the last sentence. I As indicated, all Revision 5 changes have been f ully reviewed by UFTR Management and the Reactor Saf ety Review Subcommittee to involve no unreviewed safety ques tion per 10 CFR 50.59 and so are not considered to relax the re-quirements for assuring protection of the health and saf ety of the public and of the reactor f acility. The entire enclosure consists of one (1) signed original letter of trans-mittal with enclosure plu.s ten (10) copies of the entire package. If further information is required, please advise. Sincerely, YY William G. Veruetson Associate Engineer and Director of Nuclear Facilities N_" .u. hbdk-w No tarfPublic L. l% ](Y

                                                                                                       ' page r

( WGV/ps Enclosures W* M M' d ' cc U.S. NRC Region 11 ih '(mMn '""'"'** 5.h~ M

                                                                                               ' "R E      '

P.H. Wha 1ey l Reactor Saf ety Review Subcommittee I

a, s - ,

                                                                                                      .i j.

ATTACHMENT I FINAL SAFETY ANALYSIS REPORT UNIVERSITY OF FLCRIDA TRAINING REACTOR l FACILITY LICENSE R-56; DOCKET NUMBER 50-83 l- Revision 5 l The attached Revision 5 pages revise the University of Florida Training Reac-tor Final Saf ety Analysis Report as of June, 1988. Revision 5 pages should be j subs ti';uted to replace exis ting pages as f ollows : l

                       . Previous Pages                                                                    Revision 5 Replacement Pages 1-4 (ORIGINAL)                                                                           1-4 (REV 5, 6/88)                                     1 1-5 (ORIGINAL)                                                                           1-5 (REV 5, 6/88) 3-6 (ORIGINAL)                                                                           3-6 (REV 5, 6/88) l

[' 4-9 (REV 1, 3/82) 4-9 (REV 5, 6/88) i 7-1 (ORIGINAL) 7-1 (PSV 5, 6/88) 9-1 (ORIGINAL) 9-1 (REV 5, 6/88) 15-2 (ORIGINAL) 15-2 (REV 5, 6/88) l NOTE: All Revision 5 changes to exis ting pages are clearly delineated by I vertical lines in the c argin adjacent to the change. All pages are labeled as Revision 5 at the bottoci.

Although the Radiation Control Office provides solid radioactive waste disposal service, labeling and bagging of waste is the respon-sibility of the UFTR personnel. All pertinent information must be pro-vided to this office by the UFTR personnel. These and any other matters concerning radiation and safety procedures are covered in detail in the "Standard Operating Procedures" manual of the UFTR. (3) The major experimental facilities in the UFTR are illustrated in the vertical view line drawing of the UFTR shown in Figure 1-2 and include:

1. Sixteen (16) vertical foil slots placed at intervals in the graphite between the fuel compartments, each are 3/8 in. x 1 in. - infrequently used.

2. Three (3) vertical experimental holes located centrally with respect to the six (6) fuel compartments (boxes):

i) Center Vertical Port (CVP) with 2 inch diameter

11) West Vertical Port with 11/4 inch diameter 111) East Vertical Port with 11/2 inch diameter

3. Five (,' vertical square holes filled with 4 inch x 4 inch romvable graphite stringers;

4. A horizontal themal column having'six (6) 4 inch x 4 inch removable stringers flanked on each side by 2 add-itional themal column positions with removable stringers which are infrequently used;

5. A shield tank placed against the west face of the reactor opposite the fuel boxes and themal column;

6. Six (6) horizontal openings, 4 inches in diameter, located symmetrically on the center plane of the reactor and nor-mally filled with shield plugs, only one of which (south) goes all the way to the core region;

7. A removable horizontal threughport consisting of a 2.05 ,

inch 10 aluminum tube with 20 ft. length running east-west '

          -           across the reactor. Shield plugs or other shielding appro-priate to experiments in progress are nomally inserted into these ports which are clearly identified in Figure 1.2. A pneumatic-operated rapid sample insertion device is normally inserted in the west throughport access.

As quoted in Section 1.3.1, the safety rods have the following experi-mentally verified reactivity worths measured in February,1988: Safety 1 with a 1.49% 6k /k Safety 2 with a 1.45%'6k/k Safety 3 with 2.10% a k/k with the regulating blade having a total worth of 4 0.757. Ak/k. The maxi-mum allowable worth of any single unconstrained experiment is 0.6% reac-tivity. The measured shutdown margin with the most reactive blade out was N 2.7% 6 k/k in February,1988. 1-4 "REV 5, 6/88

The UFTR is a reactor used for instructional and university re-search activities, therefore it is desigr.ed so that safety is maxi-mized without excessite restraints on the different activities planned. As quoted in Reference 3, the inherent safety of the UFTR is based on four design features. First, the amount of excess reactivity in the reactor is limited to less.than 2.37. A k/k. Sctond, the reactor has negative temperature and void coefficients. In addition, the reactor is provided with sufficient interlocks and safety trips to make a hazardous incident extremely improbable. l Third, the amount of contained fission products is relatively small. And fourth, there is an extremely low probability that these fission products can escape. Nevertheless, because of the high popu-lation density of the campus, the reactor is housed in a structure with a minimum number of penetrations sealed against gas leakage. A negative pressure is maintained in the reactor building such that air and airborne contaminants within the cell are withdrawn by means of the reactor vent system through a filter system which is continu-ously monitored for radiation activity. Possible f ailures or accident situations have been analyzed and dis-cussed in Chapter (15), including the effects of a rapid reactivity inse'rtion,, radioactive fission product release and loss of coolant flow in the case of 100 Kw (thermal) operation of the UFTR.

         ~

1.3 Comparison Tables

1. 3.1 Comparison with Similar Facility Designs The UFTR which has been operational since May, 1959, is currently licensed for operation at 100 Kw (thermal).

Similar functional, licensed r' actors were located at the University of California, Los Angeles - (UCLA), at the University of Vnhington in Seattle, Washington, at the Virginia Polytechnic Institute at Blacksburg, Virginia and in the United Kingdom. A comparison of the nuclear charac-teristics of the UFTR to those of the UCLA Nuclear Reactor is shown in Table 1-1. The UCLA Neelear Reactor was chosen because of the great similarity between the UCLA R-1 reactor and the UFTR as briefly desuibed in the following paragraphs. Both the UCLA and the Virginia Polytect.aic reactors are being decor.missioned as of June,1988. The 100 Kw UCLA Argonaut Reactor (UCLA R-1) consists of a core of six aluminum boxes arranged in two parallel rows of three boxes each, the rows being separated by and surrounded with graphite. Four fuel bundles are placed within each box, each bundle consisting of 11 uran-ium-aluminum alloy fuel plates clad with aluminum. The graphite on one side of the reactor is extended to provide a thermal column, and on the 1-5 REV 5, 6/88

o The UFTR reactor building has five entrances (exits), but only two--one upstairs and one downstairs--leading into the reactor building from the Huclear Sciences Center, will be in normal use during regular work hours. The other three entrances (exits) are kept locked at all times. The vehicle / freight doors on the West side of the reactor cell (area 101) are used only in special situations such as refueling the UFTR and now have a personnel door installed for an emergency exit. This door is monitored on the reactor control console by the reactor operator on duty. The door on the West side of the radiochemistry laboratory (area 104) is also only used in special situatlets

• such as equipment delivery but is also available for emergency exit from th building. The final exit is on the second floor on the East side outsl65 the offices (area 201) and is also kept locked. This entrace (exit) is in general use for authorized keyed personnel to enter the building at all times.

All doors are steel fire-rated doors. The main reactor room entrance opens close to and in view of the reactor  ; operator in the control room (area 102). The entrance door from the control room to the hall can be eaily opened from the inside for use as an emergency exit. This door is weather-stripped with neoprene and is equipped with a door closer. The main reactor room exit (and occasional equipment entrance) (area 103) is equipped with radiation detection / monitoring devices for personnel. This exit has an air lock set-up and is 8 f t., 4 in, long, 7 f t. wide and 8 ft. high. The air lock also opens in view of the reactor opcrator in the control room and both of its doors are metal fire doors. Both of the doors to the air lock (area 103) are weather stripped with neoprene and have door-closers. These air lock doors are also monitored on the UFTR reactor control console by the reactor operator on duty. The freight doors will be closed at all times during operation of the reactor and will be opened only during the actual transfer of material or special maintenance activities. The door is 10 f t wide by 12 ft. high, four-paneled, steel-skinned, honeycombed construction, and hinged door. The sill, jambs, astragals and head have sponge-rubber seals and caulking to minimize leakage. The bottom, left-hand panel of the freight door also contains an emergency personnel exit which can be cpened by a panic release. This emergency exit is , also supplied with a door closer. The reactor is an elongated octagon located in the center of the 30 f t. dimension of the room,12 f t. f rom the West end. It has an East-West axis of 20 ft., 4 in, and a North-South axis of 15 ft., 6 in. The clear floor dimen-sions around the reactor are summarized in Table 3-1. An observation window was originally provided between the second floor hall and the reactor room and was made up of stationary 1/4 in, thick LEXAN , plate which was a shatter and bullet proof plate, sealed in aluminum frames. An additional observation window was provided between the second floor hall and the hot cave area in the radiochemistry laboratory area in Room 104. For security reasons, these windows were sealed with solla concrete blocks and painted over on the outside with sealer and latex paint. Subsequently, the offices referred to earlier have been added in area 201 on the second floor. These offices are not considered to have any effect on the structural integrity of the reactor building. u

       .                                                                                  TABLE. 4-1 PRESENT UFTR CilARACTERISTICS o

General Features Reac to r Type . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . He te roge neous , T he rmal Licens ed Ra ted Power Level . . . . . . . . . . . . . . . . 100 Kw Thermal Maximum tMemal flux level in center 12 2 vertical port at 100 Kw................. N 1.5 x 10 n/cm see Excess reac tivity (at 72'F) . . . . . . . . . . . . . . . N 1.0% 6 k/k Clean , , cold critical masn . . . . . . . . . . . . . . . . . 3.07 kg g235 Ef f ective prompt neutron lif etime. . . . . . . . . 2.8 x 10 sec Uniform water void coefficient............ /% voids Tempera ture coef f icient. . . . . . . . . . . . . . . . . . . -0.2% a kg% A k/k per *F j

                                                                                                             -0.3 x 10 U-235 mas s coef f icient. . . . . . . . . . . . . . . . . . . . 0.4% 6 k/% U-235 Startup source............................                                                     Sb-Be < 25 curies or PuBe <~
                                                                                                                        ~

1.0 curies t Re f l e c to r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . g ra p hi te ( 1. 6 g m/ co )  ! Mode ra to r. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . H2 O and g rap hi te ,

                                                                                                                                                                       )

( Fuel Plates Fu e 1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 3 % e n r i c te d , U-Al Fuel loading.............................. 3408.95 gm U-235 i I { Pla te thicknes s . . . . . . . . . . . . . . . . . . . . . . . . . . 0. 07 0 in . Pl a te wid t h. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2. 8 4 5 in . Pl a te le ng th. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 5. 6 2 5 i n . ( 0.13 7 in. t Wa te r c hannel wid t h. . . . . . . . . . . . . . . . . . . . . . Aluminum to water ra tio (volume) . . . . . . . . . 0. 49 "He a t" comp os i t ion . . . . . . . . . . . . . . . . . . . . . . . . 14.05 w/o U ( Coolant Type...................................... H,0 f Flow ( a t 10 0 Kw) . . . . . . . . . . . . . . . . . . . . . . . . . 41. 0 gpm I Equilibrium Inlet Temperaturo (100 Kw).... 115'F - j Equilibriun Outlet Temperature (100 Kw). . 130'F l' f-Control Blades Typ e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Cd , s w ing ing vane , grav i ty f all Numb e r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 s af e ty ; 1 regulating [ Ins e r t io n t i m e . . . . . . . . . . . . . . . . . . . . . . . . . . . . < 1 see Removal time . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 100 s e c (minimum) Blade wor th, s af e ties . . . . . . . . . . . . . . . . . . . . . Saf e ty #1 N 1. 49% a k/k

                                                           . . . . . . . . . . . . . . . . . . . . . Saf e ty # 2 N 1. 4 5% 6 k/k

[ ..................... Safety #3 N 2.1% 6k/k Blade worth, regulating................... Reg. Rod N O.75% 6 k/k Reactivity addition rate, caximum allowed. 0.06% a k/k/sec { Shield (concreto) Sid es , cen te r . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 f t. , cas t , bary tes ends............................... 6 f t. 9 in., cast, barytes [ l Sides, Midd1c . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Ba ry tes conc re te blocks 5 ft. 10 in, Top...................................... End...................................... 3 ft. 4 in. f Experimental Facilities Thernal column , h rizon ta1. . . . . . . . . . . . . . . 60 in. x 60 in. x 56 in. high { Thermal column, ver tical . . . . . . . . . . . . . . . . . 2 f t. diam. x 5 f t. ; H O or D20 Shield tes t tank. . . . . . . . . . . . . . . . . . . . . . . . . 5 f t . x 5 f t . x 14 f t . 2hig h Ex p e r i m e n t a l ho 1 c s . . . . . . . . . . . . . . . . . . . . . . . . 5 vertical, 4 in. x 4 in.

                                                  ........................ 3 vertical, 1 1-1/2 in.

Foil n1ots................................ 16 vertical, 3/8 in. x 1.0 in. l ._ _ - _ _ _ _ _ _ _ _ . _ _ _ _

7. INSTRUMENTATION AND CONTROL S 7.1 Introduction The reactor instrumentation monitors several reactor parameters and trans-mits the appropriate signals to the regulating system during normal operation, b and during abnormal and accident conditions to the reactor trip and safety systems. Since the UFTR is a low power, self-lin.iting reactor, the instru,en-tation and associated controls are considerably simplified when compared to ,

instrumentation and control systems of large power reactors. / l 7.2 Identification of Safety-Related Systems The safety-related instrumentation end controls for the UFTR include the control console, the control and safety channels, the facility interlock sys-tem, control drive switches, and the reactor scram circuitry. Table 7-1 con-tains a list of abbreviations used in the UFTR instrumentation and control diagrams; it is repeated from Chapter 1 for completeness and ease of reference I in this chapter. Figure 7-1 shows a block diagram of the nuclear instrumenta- l tion and scram logic of the UFTR. I I

7. 2.1 Console

 ,             All the functions essential to the operation of the UFTR are controlled by                                                                l the operator from a desk-type control cunsole. The reactor console is conven-iently located near the reactor to allow the reactor operator to monitor ac-tivities in the reactor cell during operation. All of the instrumentation con-tained in the console that is essential to the operation of the reactor ac-cepts or sends signals from or to the control rod drives, the reactor inter-(          lock system, and various detectors and transducers located around the reactor core, the reactor coolant system, and auxiliary systems such as the reactor vent system and the secondary coulant system.

f The reactor control panel contains the following control and indicating instrumentation:

1. A console pcwer (POWER ON) switch.

2. A three-position 0FF/0PERATE/ RESET key switch, r 3. A set of four control-blade switches for the three safety blades (1, l - 2, and 3) and the regulating blade. One set of switches for control-ling the secondary system city water valve.

4. Four control blade position digital indicators.

l 5. A MODE SELECTOR switch (mode switch) f,or automatic or manual operation.

6. A REACTOR POWER range switch (range switch).

7. A dual-pen strip-chart recorder.

( 8. A %-DEMAND control potentiometer.

9. A manual SCRAM bar.

I 10. A REACTOR PERIOD meter and calibrate / test controls, t 11. A set of scram (14) and blade interlock (3) annunciator lights, left panel.

12. Safety Channel Meter #1 and test controls.

I 13. Safety Channel Meter #2 and test controls.

14. Log Power Meter and calibrate controls.

15. Reactor cell entrance / exit door monitors.

16. Reactor equipment control switches and annunciator lights, right panel.

17. Digital clock.

18. PuBe source alarm indicator.

l 19. Energization switch and co.nmunication line for the pneumatic-operated rapid sample insertion system. I - A - L __ _____________ ________________________________

a

9. nUXILIARY SYSTEMS 9.1 Fuel Storage and Handling 9.1.1 New Fuel Storage 1

Unirradiated reactor fuel is normally stored in a 5-drawer, fire-resistant Diebold Safe equipped with a combination lock. Supports are ~ provided to space the plates in such a manner that no more than 56 plates can be placed in a drawer. The bottom of each drawer is lined with cad-mium. The fuel storage safe, which is locked at all times except during [ transfer of feel or inventory is located in the rea . tor cell. An authorized person is present at all times when the reactor cell (which comprises the reactor room and the control room) is unlocked. The reactor cell is pro-tected by a security system *,,hich alanns at the University of Florida cam-pus police headquarters. [ Loading and unloading of the fuel into and out of the reactor will only be performed by qualified reactor operators and staff, and under the supervision of the reactor supervisor as specified in the UFTR S0P C.1 and C.2. [ 9.1.2 Spent Fuel Storage

!                              Irradiated fuel is removed from the reactor in a lead transfer cask           l using the crane and special handling tools (Section 9.1.2.1); a contir.uous radiation survey is made while the fuel is being transferred. Irradiated

{ fuel assemblies or plates are stored in the spent fuel storage area located in the concrete floor at the northwest corner of the reactor cell as shown in Figure 3-2. This storage area is readily accessible to the crane and [ contains 27 steel-lined storage pits, each of which is 4" in diameter x l 4 f t, deep. These storage pits are arranged so that k ' O.8 under optimum conditions of reflection and moderatT$$.will be less Padlocked than shield

    .                    plugs are provided for these storage pits and are keyed to the University of Florida Proprietary Keyway, Sargent Grand Master Series. The key is kept in e safe, available to the Reactor Administration and under established
,                        conditions can be used by qualified reactor operators. Therefore, all re-l                        actor fuel which is not in the reactor will be locked either in the fuel safe or in the fuel storage pits, or in active transfer between these places.

{ Fuel plates a e replaced when necessary. The irradiated fuel can be l shipped to a fuel reprocessing plant after sufficient cooling. l 9.1.3 Bridge Crane [ A 3-ton bridge crane is provided for handling shield blocks, lead casks, and other heavy equipment. The crane travel allows coverage of the entire area of the reactor cell as shown in Figure 3-2. Maximum clear-ance of 11 ft., 9 in, can be obtained between the top of the reactor, which extends 11 f t.,10-1/2 in, above the floor, and the crane honk. l The clearance is reduced to 8 ft., 9 in. over the water tank which ex- l tends 3 f t above the top of the reactor. This clearance is adequate for use of the lead transfer cask to remove irradiated fuel elements from the reactor 9-1 6 - -. - .

15.11 Nuclear Excursions 15.1.1.1 Nuclear Excursions During Operation. It is difficult to visualize any circumstances which would result in a reactivity increase of a magnitude sufficient to cause serious degradation of the UFTR core. The design of the cooling system insures that the temperature of the reactor cannot be changed suddenly by the introduction of cold water. The maximum excursion which could occur with the rormal fuel loading would result from the sudden insertion of l all the available excess reactivity ; 21.0% ak/k available. A maximum of 2.3% excess o k/k can be loaded. Only two (2) methods are considered possible for loadi.g such an excess reactivity. First, the maximum excess reactivity could be reached by having the reactor temperature lowered to the freezing point of l water; second, the maximum excess reactivity could be reached.by violation of the standard operating procedures. The first method for insertion of maximum excess reactivity by reduction of reactor temperatures to.the freezing point is not considered feasible or plausible, not only because of the building and climate involved but also be-cause of the time element that would be required during which some abnormali-ties would be noted. As explained in the original UFTR Hazards Summary Report, the second method for insertion of maximum excess reactivity violation of the ) standard operating procedures if a possibility.(2) The Hazards Summary addresses two possible violations of S0ps by which the maximum excess reactivity in the UFTR could be achieved. The first violation involves loading a sample into the reactor with sufficient absorption proper-ties to prevent startup or reaching criticality regardless of the amount of control blade withdrawal If the control blades were fully withdrawn in this situation and criticality were not achieved, the maximum reactivity could be l

      .added if the sample were then removed without reinserting the control blades.

l The other possible, although extremely difficult, manner by which the maximum excess reactivity can be inserted would be by purposely and wantonly l l bypassing the Reactor Control and Safety System interlocks and trips and sub-sequently withdrawing the blades, in violation of the Technical SpeciTRations  ; l and the Standard Operating Procedures. I If all the circuits of the Reactor Protection System were to fail or be incapacitated, the power level would continue to rise until the available ex-i cess reactivity were overcome by the temperature and void coefficients charac- l teristic of the present reactor configuration. l As a result of studies made for the original Hazards Summary Report I (2) concerning the effects of e large reactivity addition in the UFTR during 100 KWth operation, it was also determined that the required power excursion in order to raise the temperature of the fuel plates to the melting point of aluminum (1220*F) involves an energy generation of 32 MW-sec, as explained in Appendix 15A (22). The corresponding exponen-tial period for this excursion is 8.3 milliseconds; therefore, the UFTR will tolerate a power excursion with a period at least as short as 8.3 15-2 REV 5, 6/88

u I ( f L r L

                                                                        'A

( l [ [ L APPENDIX F f L. UFTR STANDARD OPERATING PROCEDURES ORIGINAlE AND MAJOR REVISlQNS FOR 1987-1988 REPORTING YEAR

1. UFTR SOP-F.8, "UFIR SAFEGUARDS r REPORTING REQUIREMEN'IF (REV 0)

L 1 i

                                                                                                /

SOP-F.8 PAGE 1 of 9 I UFTR OPERATING PROCEDURE F.8 l 1.0 UFTR Safegtnrds Reporting Requirements

 . 2.0  Approval                                          !,         /
                                                            .[        /         IJ 77 [

Reactor Safety Review Subcommittee . . . . . . / L -{ "

                                                                    ,[o- -              /

Da te Director, Nuclear Facilities . . . . . . . . . f> J. - /dJJ//'/

                                                                                  / Da t'e i

l p [ REV 0. 9/87 i.

SOP-P.8 PAGE 2 of 9 3.0 The purpose of this procedure is to delinea te the reporting of safeguards events to the NRC for the UPTR R-56 license. Items l addressed include: I 3.1 Safeguards events that must be reported to the NRC. 3.2 Designation of how communications are to be made to NRC concern-ing safeguards events. 3.3 Specifica tion of time intervals for telephone communica tion and s ubmi t ta l of licensee written reports f or applicable sa f egua rds events. 4.0 Precautions and Limits 4.1 The reporting of safeguards events to the NRC is necessary 4.1.1 To assure safety during sa f egua rds-rela ted emergencies 4.1.2 To allow the Commission to identif y a nd c ha ra c t eri z e generic and f a cili ty-specific precursors to certain safeguards events. 4.2 Both telephone and written communications of safeguards events are required. 4.2.1 The 24-hour telephone number for the NRC Ope ra tion s Center is (301) 951-0550 4.2.2 Addresses for submission of written reports are as follows: 4.2.2.1 Original report to: U.S. Nuclear Regulatory Commission Washington, D.C. 20555 ATTN: Document Control Desk 4.2.2.2 One (1) copy of report to: US Nuclear Regula tory Commission Region II P. O. Dox 2203 A tla n ta , GA 30301 , 4.3 Safeguards ovonts are defined as viola tions of the UPTR Security l Plan as follows: 4.3.1 Actual, attempted or (credibly) threatened theft of special , nuclear material (SNM). REV 0, 9/87

k I SOP-F.8 PAGE 3 of 9 L 4.3.2 Actual, attempted or (credibly) threatened acts or events which interrupt normal ope ra ti on s at the UPTR due to un-authorized use of or tampering with machinery, components or ( controls. 4.3.3 Any loss, theft or unlawful diversion of special nuclea r ma-terial (SNM) under the R-56 license or any incident in which  ; an a ttempt has been made or is believed to have been made to commit a theft or unlawful diversion of such material. 4.3.4 Attempts to bring contraband into the reactor cell. 4.3.5 Any threatened, attempted, or committed act with the poten-tial for reducing the effectiveness of the safeguards system below commitments of the Security Plan. 5.0 References 5.1 10 CFR 70, "Domestic Licensing of Special Nuclear Me terial" 5.2 10 CPR 73, "Physical Protection of Plants and Materials" 5.3 UPTR Security Plan { 6.0 Records Required 6.1 UFTR Form SOP-F.8A, "Record of NRC Safeguards-Related Telephone Communications" 6.2 UFTR Form SOP-F.8B, "Loa of UFTR Safeguards Events" 6.3 Written Reports of Safeguards Events. 7.0 Instructions 7.1 The NRC Opera tions Center shall be notified by telephone (301-951-0550) within one hour after discovery of: 7.1.1 Any case of accidental criticality or any loss, other than normal ope ra ting loss, of special nuclear material. 7.1.2 Any event in which there is reason to believe that a person has committed or caused, or attempted to commit, or has made a credible threa t to commit or cauce: 7.1.2.1 A theft or unlawful diversion of special nuclear ma teria l ; 7.1.2.2 Significant physical damage to the UPTR facility or its fuel;

s SOP-F 8 PAGE 4 of 9 L 7.1.2.3 Intorruption of normal ope ra tion through the unauthorized use of or tampering with UFTR machinery, components, or controls including the security system. ( 7.1.3 An actual entry of an unauthorized person into the reactor cell. i 7.1.4 Any failure, degradation, or discovered vulnerability in the j safeguards system that could have allowed unauthorized or un-t detieted access into the reactor cell for'which compensatory { measures have not been employed. J 7.1.5 The actual or a ttempted introduction of c o n t ra ba n d into the f reactor cell. 7.2 Telephone communica tions should be recorded on UFTR Form SOP-F.8A, "Record of NRC Safeguards-Related relophone Communica-tions" as' contained in Appendix I or equivalent form. f 7.3 Followen Written Notifica tion 7.3.1 Initial telephone no ti f ica tion shall be followed within a period of 30 days by a written report which includes suffi- {' cient in f o rma ti on for NRC analysis and evaluation. 7.3.2 Written reports shall be submitted to two (2) NRC addresses. 7.3.2.1 Original report to: U.S. Nucicar Regulatory Commission Washington, D.C. 20555 ATTN: Document Control Desk 7.3.2.2 One (1) copy of report to: US Nuclear Regulatory Commission { Region II P. O. Box 2203 A tla nta , GA 30301 7.3.3 Errors discovered in a written report must be corrocted in a revised report with revisions indicated. 7.3.3.1 The revised report must replace the previous reports 7.3.3.2 The updato must be a complete revised report and not con-tain only supplomontary or revised informations j 7.4 A copy of each written report of an event shall be kept aa a

t. record for a period of three years from the date of the report.

I _ _ _ . _ _ _ _ _ _ _ _ _ - RP.V 0, 9/87

m- , k

• SOP-P.8 PAGE 5 of 9 l

7.5 Significant supplemental information which becomes available after the initial telephonic notification to the NRC Opera tions Center or after the submission of the written report must be reported by telephone to the NRC Opera tions Center and also sub-mitted in a revised written report (with the revisions indi-cated). 7.6 A current log record of safeguards events shall be ma i n ta ined using UPTR Form SOP-F.83, "Log of UPTR Safeguards Events" or l equivalent form. 7.6.1 Events should be recorded by the year and number; for exam-ple, the second event in 1987 would be entered as 87-2. 7.6.2 Safeguards events to be recorded within 24 hours of discovery are as follows 7.6.2.1 Any failure, degrada tion , or discovered vulnerability in the safeguards system that could have allowed unauthorized or undetected access to the reactor cell had compensa tory [ measures not been established; l 7.6.2.2 Any other threatened, attempted, or committed act with the I potential for reducing the effectiveness of the sa f egua rds j system below that committed to in the Physical Security Pla n or the actual condition of such reduction in effec-tiveness. NOTE: Yhe 24-hour time limit includes time for discovery by UPTR staff pe- onnel or by University Police De- { partment personnal acting on behalf of the staff.

            ~

Therefore, prompt response, followup and acknow-ledgement of UPD telephone reports is essential. 7.6.3 The log of events shall be retained as a record for three years after the last entry is made in the log. 7.6.3.1 Log entries shall not include details that communicate de-scriptive inf orma tion about the physical security system, or about security response procedures; 7.6.3.2 Log entries shall indicate the occurrence of the event to include: 7.6.3.2.1 Date/ Time of Discovery; l 7.6.3.2.2 Name anG Position of Discoveror; 7.6.3.2.3 Estimated Time. of Occurrence; i l t _ --- --

SOP-F.8 PAGE 6 of 9 7.6.3.2.4 Limited Description of Event; and i 7.6.3.2.5 Reactor Manager or Facility Director Acknowledgement. 7.6.3.3 Supplemental information detailing the physical security system cr security response procedures should be recorded on UFTR Form SOP-F.7 A , "Security Information Form: Physi-cal Security Evaluation" and controlled a s safeguards in-formation if applicable per 10 CPR 73.21. 7.6.4 Copies of all safeguards event log entries not previously submitted after October 8, 1987 shall be submitted quarterly to the Nuclear Regulatory commission, Document control Desk, Washington, D.C. 20555. 7.6.4.1 Quarterly reports will be made on UPTR Form SOP-F.0B. a s a Q-7 report. 7.6.4.2 Quarterly reports will be made by calendar qua rters due at the end of March, June, September and December each year. l l 1 l l l l REV 0, 9/07

SOP-P.8 PAGE 7 of 9 l l l l l l APPENDIX I UPTR Porm SOP-F.8A 1

   "Record of NRC Safeguards-Related Telephone communications" UPTR Form SOP-F.8B "Log of UPTR Safoguards Events" A,.

t l l r P lti;V 0, 9/07

SOP-F.8 PAGE 8 of 9 UFTR Form SOP-F.8A RECORD OF NRC SAFEGUARDS-RELATED TELEPHONE COMMUNICATIONS Incoming / Outgoing (circle) DATE: NAME OF CONTACT: ORGANIZATION: LOCATION: PHONE # IS TilIS A SAFEGUARDS EVENT:

• SUBJECT AREA:

ACTION REQUIRED: $

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DY: (Dato) SIGNATURE' FACILITY DIRECTOR ACLNOWLEDGFNENT: _ REV 0, 9/87 E

E SOP-P.8 PACE 9 of 9 Uf TR Form 50p.7.83 LOO CF UFTR SAFIGOARDS EVENTS Events are to be recorded mathin 24 hours and sutaitted to the NAC in a Quarterly Log per 10 CFR 73. A;9endia G. Pa*43raph !!(a) and ll(b). E<ent Oste / Time of N4 e/Pcsttien Est. Tira cf Description er twent As Mgr/Fac Olr M ,* Der OllcCytry Cf Ollcoverer Occurrence (Reference Attactr'4nts) Ackevledgement

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l l l APPENDIX G l DOCUMENTATION FOR QUALITY ASSURANCE PROGRAM APPROVAL FOR RADIOACTIVE MATERIALS PACKAGES NO. 0578, REVISION 1 l l l l l l

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{ l SGTB:0578 71-0578 l University of Florida ATTN: Mr. W. G. Vernetson Nuclear Reactor Facility Nuclear Reactor Bldg. Gainesville, FL 32611 Gentlemen: 4 Enclosed is Quality Assurance Program Approval for Radioactive Material Packages No. 0578, Revision No. 1. { Quality Assurance Program Approval No. 0578, Revision No. O has been revised to reflect the appropriate conditions of your approval. ( Sincerely, I Charles E. MacDonald, Chief i Transportation Branch I Division of Safeguards and Transportation, NMSS ( Enclosu're: As stated l l l e

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QUALITY ASSURANCE PROGRAM APPROVAL g ,3,Onnuusta FOR RADIOACTIVE MATERIAL PACKAGES 1 g Pursuant to the Atomic Energy Act of 1954.as amended.the Energy Reorganization Act of 1974.as amended, and Title 10. Code of Federal ) Regulations, Chapter 1. Part 71. and in rehance on statements and representations beretof ore made in item 5 by the person named in item p 2 the Quahty Assurance Program identified in item 5 is hereby approved This approvalis issued to satisfy the requirements of Section g 71.101 of 10 CFR Part 71. TNs approvalis subject to all appHcable rules, regulations and oroers of the Nuclear Regulatory Commission g now or nereafter in ettect and to any conditions spectried beio . p h 3 EXPlaAteON oATE k~

7. NAMbniversity of Florida, Nuclear Reactor facility g stasETaooaEss October 31, 1992 h Nuclear P.eactor Bldg. 4 oocxetsuvees k -

CIT y STATE ZIP COoE h Gainesville FL 32611 71-0578 p

5. Quatiiv Assua=NcE enOca4u apetiC4 tion o=Teisi > -

September 2, 1987

6. CONoiTIONS f

h k Activities authorized by this approval are use and maintenance applicable to f-shipment of SPERT F-1 fuel pins in 00T Specification 6M Shipping Containers. g It shall remain the responsibility of the licensee-user that all transportation p I activities meet the requirements of 10 CFR 71 Subpart H. > i > l > c hr{ <; V

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• DIVISION OF S AFIGUARDS AND TH ANSPOaT All0N L' Of flCE OF NUCLEAR MAlf RIAL SAF E e Y AND SAF LGUARDS x=xxramuur:raxxxrmurmuurxraxura:rwr:rwraxxt:r: ras rwr:r:grxrxua

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September 17, 1987 Director Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Corr. mission Washington, D.C. 20555 Re: Snot-1050 License Gentlemen: 1 Enclosed is a second submittal of a Quality Assuranco Pro-gram for Shipment of SPERT F-1 Fuel Pins por 10 CPR Part 71. The previous document was stamped as containing 2.790 inforration to l be withheld from public disclosure which is not correct. This program has undergone all necessary reviews as indicated by the signed cover page and will be fully implemented to control this shipment. This shipment is for a rosearch project at Oak Ridge Na-tional Laboratory and involves less than 25% of tho fuel pins held in storago currently under the Stoi-1050 license. I would like to got NRC approval of this plan at your earliest con-venionco. We would like to have approval of this program for the standard fivo (5) year period. If there are any questions, pleaso lot us know. Thank you for your assistance. l f Sincerely, l h 6C $4 h Wil'iam G. Vernetson l Assoc.lato Engineer and Director of Nuclear racilities WCV/ps Enclosure 1 ( cc: J.S. Tulenko M.J. Ohanian t us cwrarM%sw AcNntasww

L QUALITY ASSURANCE PROGRAM FOR SHIPMENT OF SPERT F-1 FUEL PINS PER 10 CFR PART 71 APPROVAL SPERT Facility SNM-1050 Safety Review Subcommittee . _I .A 'i (7 / Date Director, Nuclear Facilities . . . . . . . . . . . .df f e

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( QUALITY ASSURANCE PROGRAM FOR SHIPMENT OF SPERT F-1 FUEL PINS 10 CPR PART 71

1. ORGANIZATION The final responsibility for this Quality Assurance (QA) Program for Part 71 Requirements rests with the University of Florida. Design, f abrica t! on, re-pair and maintenance shall not be condacted under this QA program. This QA program is established to ensure compliance with 10 CFR 71 for the transporta-tion of up to twelve hundred (1200) stainless steel clad, 4.81% enriched ura-nium-dioxide fuel pins with unique serial numbers from the University of Flo-rids SPERT storage facility (SNM-1050 license) to a secure facility at the Oak Ridge National Latoratory. A sketch of a SPERT F-1 fuel pin is included for f reference purposes as Figure 1. The pins have been used in various neutron pulsing inverse multiplication and other experiments in suberitical but highly multiplyi g, n water-moderated lattices from 1968 - 1933. Since 1981, those pins have been in drj, climate-cnntrolled storage with the SNM-1050 license amended for stcrage only. Each pin contains a total of about 35 gms of U-235 with a total uranium elemental weight of about 724 grams.

f l l

           ,These SPERT F-1 fuel pins are currently possessed (on loan from the De-partment of Energy) by the University of Florida licensee under NRC license         j number SNM-1050. This quality assurance program is of limited duration and is designed for the transfer of these DOE-owned pins to a DOE facility at Oak

( Ridge National Laboratories for uso in a series of reactor blanket experi-ments. The QA program is fraplemented using the existing UFSA SNM-1050 admin-istrative orginization shown in Figure 2 which is essentially the same as the administrative orginization for the University of Florids Training React- - R-56 license, f The University of Florids Radiation Control Of fice is responsible for 1 l - _____ _ ____.

v overall administration of the program, training and certification, document control and auditing. The Director of Nuclear Facilities, Dr. William G. Vernetson, as the in-dividual responsible for the direction and administration of the University of r riorida SPERT Assembly facility per Attachment I is responsible for handling, k storing, shipping, inspection, test and opera ting status, and record <eeping. prior to approving return of these fuel pins af ter un at *he Oak Ridge < { National Laboratory, the results of an evaluation and analysis will be used to determine significant by-product or transuranics build-up in the shipped fuel pins. This information will be used to evaluate whether the pins ray be ac-cepted for return to the SNM-1050 license and the racility at the University of Florida and whether a license amendment (SNM-1050) would be needed to re- { turn the fuel to the SNM-1050 license. At that time if necessary and desir-able, this QA Program will be amended to allow return of the SPERT F-1 pins. f 2. QUALITY ASSURANCE PROGRAM The Director of Nuclear Facilities as the individual responsible for the direction and administration of the SPERT Assembly facility establishes and imple'ments this QA *e rogram. Trainf ng, prior to loading and shipping, for all e QA functions is to be rwde according to written instructions with St:M-1050 f management approval. The QA Program will ensure that all defined quslity con-trol checklists, instructions, procedures, and specific provisions of the shipping container design approval are satisf) <d. The QA Program will empha-size control of the characteristics of the psekage which are critical to

   ' safety.

3. PACKAGE DESIGN COffrROL Design activities related to the shipping package are not to be ocrformed by the University of Florlds. Tho Ur Radiation Control of fico shall assur.

2 l - . _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _

k that the radioactive material shipping containor is designed and manufactured to meet the existing transportation regulations. This requirement will be satisfied by using DOT approved 6M sh% ping containers for which a Certificate of compliance will be available to assure acceptability for shipping SPERT-type fuel pins. These cor. tainers will be supplied by Oax Ridge National Labo-i ra tory. Pins will be unloaded from the storage racks and transferred from one bin at a time inte 6M container within a specially-controlled area of the fuel { storage room in sempliance with existing radiat.on control procedores. t Accidental critical'.*v will be prevented by shioping the fuel dry with no more than 65 pins per packaga. Criticality calculations to assure prevention of accidental criticality will be petformed by personne1' ht Oak Ridge National I Laboratory with rethods used and' rosults obtained supplied to University per- l f  ! sonnel to suppett loadinc .,'nd shipping of the fuel pins at the loading indi- l l { cated. l ( 4. PROCURD4ENT CONTROL Procurement activities related to 'he citipping package are not to be per-formed by the University of riorida. The proper procurement document control shall bo the responsibil. n of the supplier of the de signated shipping package. {

5. INSTRUCTIONS, PROCEDURES AND DRAWINGS Instructions. chocklists and procedures will be used to prepare all ship-ping containers prior to loading with SPTRT P-1 fuel pins. Instructions and f

checklists will be used to control the loading of all @ containers ana to assure all necessary checks are trado af ter containers are scaled. Detailed in-ventorios of all SPERT pins to be shipped will be used with the checklint con-tcolling loading of containers, f Checklists will also be uued for transf er of 6M containars loaded with e RT r - 1 fuel pic4 trom the SNM-1050 license. 3 i

m ( Instructions for radia tion and contamina *'on surveys will also be used to ( survey containers prior to loading, af ter loading and af ter loading to the shipping vehicle.

6. DOCtMENT CONTROL A complete detailed inventory of all SNM-1050 material will be performed prior to implementing this Quality Assurance Program. Detailed checklists will be used to control and document which SPERT P-1 fuel pins are shipped and which remain under the SNM-1050 license at the University of Florida. A com -

plete separate inventory with identitication numbers for all fuel pins will be maintained for h'>th sets of f uel pins - those shipped to Oak Ridge National Laboratory and those remaining in storage under SNM-1050 at the University of l Florida. These inventories will be generated prior to ieplertenting this QA Program to assure proper documentation of identification and control of Spe-l cial Nuclear Material under the SNM-1050 license. The detailed inventory of SPERT fuel pins in the shipmant will also be provided to the carrier for transmittal to the receiver. Documents related to specific packages and shipments are to be retained by the UP Radiation Control Of fice. This documentation shall include radiation and contamiratier. surveys, checklists to control loading and transferring con- l l [ l tainers, cargo rtanif ests, notes concerning labeling and sealing. Form 741, and l all other inforrution related to the control and accountability of the radio- l l l active materials to include detailed inventory checklists for the pins loaded into each shipping container and the total quantity shipped. Procedures and checklists, and changes thereto, are to be approved by the SNM-1050 Pacility Director and by the Safety Review Subcommittee or their re-spective designates. Documents and notes relating to securicy provisions will be retained by 4 L - - - - - - - - _ _ _ _ _ _ _ _ _ _

the UFSA SNM-1050 management.

7. CONTROL OF PURCilASED MATERIAL, EQU?.PHENT, AND SERVICFS No special * .>ose materials or equipment are to be purchased for this activity. Services such . ' container of f-loading, container weighing and con-tainer on-loading will be procured via nortnal University proced .res. Carrier transport services will be obtained directly via Oak Ridge National Laboratory and will serve only as the carrier to transport the shipment to the Oak Ridge National Laboratory site where it will be transferred.

8. IDENTIFICATION AND CONTROL OF MATERIALS, PARTS, COMPONENTS l No msterials, parts or components are to be identified or controlled for this activity. Replacements other than serviceable items will be performed under other approved programs.

9. SPECIAL PROCESSES No special processes are to be undertaken under this program.

1

10. INSPDCTION CONTROL 10.1 Receipt Inspection. --- Prior to use, each shipping container will be opened to determine operability of closures, to visually inspect the in-tegrity of the structure, and to provide cecess for interior swipes. Inade-quately identified packaging, or packaging which deviates significantly from certifications, will not be used unless or until corrected. All containers will be checked to assure they meet Certifit 2te of Compliance requirementar nonconforming containers will be returned to the supplier. All fuel pins will be checked for danige visually and tracked per the inventory and packaging checklists. Visuil inspection for container structural integrity as well as radiation and contamination surveys will ba prformed prior to transfer of each i

container to the shipper. f ( 5 l _ _ _ _

10.2 Maintenance. --- Mait .enance other than prescribed servicing will not be performed by the University. 10.3 Final Inspection. --- Checklists will be established to ensure that:

1. Packages are properly assembled.

2. Moderators and/or neutron absorbers are present if required.

3. Shipping papers are properly completod.

4. Packages and transport vehicles are conspicuously and durably marked as and if required by DOT.

l l S. Pre-loading and Post-loading radiation surveys have been completed. I

6. Final inspection has been completed. l l

Inspection is to certified by the racility Director or Manager and by the Radiatien Control Officer or their designated alternates. l 1

 #1,  TEDT CONTROL 11.1 Use of Packages. --- Tests permitted, recommended, or specified by package licensee will be used to establish a QA checklist.

l 11.2 Radiation Survey. --- Radiation and contamination survey results will be compiled and records m? intained by the Radiation Control Of ficer.

12. CONTROL OF Mr.ASURING AND TEST EQUIPMENT As a user, the University of Florida does not expect to use gauges, fix-tures, reference standards, or other devices used to measure product (con-tainer) cha rac teris tics. Radiation survey and monitoring equipment shall be raintained and calibrated in accordance with normal procedures.

13. ItANDLING, STORAGE AND SilIPPING The radioactive roterial is low enriched uranium-dioxide clad with 304 stainless steel in the form of fuci pins that are approxinutely 1.06 meters in length (see rigure 2). The maximum U-235 in any pin is less than about 35 gms.

I 6 L _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

Written instructions will be used to assure proper placement of fuel pins in the inner container and for proper blocking to maintain each pin in proper placement with respect to the liner as well as for securing the containers and assuring their integrity following closure. 13.1 Handling and Storage. --- No special handling and lif ting equipment l wil' be used in accordance with equipment specified or provided by the package licensee, and according to conditions identified in a Certificate of Compli-ance as well as instructions provided by the packaging licensee. See Sections 4, 5 and 6. Containers supplied by the receiver will be used promptly (within one week) and retured as designated by the supplie'; they will not be placed in storage at the University of riorida. l l l 13.2 Preparation for Relmse and Shipment. --- Measures will be insti-tuted to ensure that:

1. Cavities are dry.

l

2. Specified operr* ions, inspections and tests are verified by check-l list.

3. The Radiation Control Officer is responsible for the observation of NRC and DW requirements, and for the prepara rion of the shipping papers.

4. Quality Assurance will be performed and documented with checklists.

! 13.3 Transportation Safeguards. --- For the Special Nuclear Material ad-f dressed by this QA Program *amper Indicating Devices (TIDs) 3r s ea ls .till be applied to all containers of SNM of fered for transport so that the package cannot be opened without breaking the seal. TIDs shall be designed so that tr ' y cannot be breached, broken, or otherwice removed and reapplied without obvious visible evidence that *ampering with the package has occurred. TIDs will be uniquely identifiable. Seal numbers shall be included in the shipping papers, be provided as part of the advanced notification, and appear on the i 7 r

f Nuclear Material Transportation Report (Form 741). Unused TIDs will be kept in a secure area. A TID invet tory will be maintained and the use of TIDs logged and documented. TIDs will be applied with two persons verifying the applica-tion and initialling the application record. Before transport, the University of Florida, as shipper of record, shall

a. Verify name and address of consignee,

b. Obtain approvals for the shipment.

I

c. Provide information to consignee concerning material to be shipped, packaged, carrier, and shipping arrangements,

d. Verify receiver's license status by obtaining a copy of the license.

At the time of shipment, the shipper of record shall: h a. Notify consignee of shipping time, shipping date, and estimated time l of arrival.

b. Confirm in writing this informtion and the TID seal numbers.

c. Remind consignee to notify shipper if shipment has not arrived in reasonable time.

14. INSPECTION , TEST AND OPERATING STATtJS Status is to be tracked by a mater checklist that acknowledges check-off of individual checklist completion.

15. CONTROL OF NONCONFORMING MATERIALS, PARTS, OR COMPONENTS l

No applicable. Rework, repair, maintenance, or modification are not to be l undertaken by the University of Florida.

16. CORRECTIVE ACTIONS 16.1 Reporting. --- It is the responsibility of the Radiation Control Of-ficer University (c/QA to report conditions detrimental to quality to the pa ckage licensee.

16.2 Closecut. --- The University as a user will deem closcout completed 8 I _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _

A f upon (a) correction of the condition by the package licensee, or (b) package f licensee's - withdrawal of the container from service meaning return of the package unused to the supplier.

17. QUALITY ASSURANCE RECORDS A record shall be generated for each shipment of SNM showing the certifi-cate of compliance nanber for the package, authorization to use the package, package inspection repc bs, radiation level and contamination records for package, type, and quality 'of SNM in each package, date of shipment, instruc-tions to carrier, bill of lading showing requirements for signature service, name and address of receiver, and prior notification to receiver of shipment.

DOE /NRC Forms 741 and 742 shall be retained for the lifetime of the fa-cility. All other forms shall be retained for five years. Written procedures, checklists, equipment lists, drawing and radiological { survey and exposure data will be retained by the UFSA SNM-1050 management. Records are to be retained by the Radiaiton Control Of fice, which is also 1 responsible for maintaining all University ree:ords related to personnel expo- i sures, radioactive material releases and shipment, and radiation protection matters related to the University Reactor Facility. l {

18. AUDITS The activity covered by this QA Program is a short-term ef fort. Af ter the shipment is completed, the LTSA Safety Review Subcommittee will appoint a representative tr "eform a closeout audit, an audit in accordance with writ-ten checklists to masure proper accounting for all SNM-1050 fuel and adequacy of the records generated under this program.

9 l - - ---

Q f: L i. l LEVEL 1 Director, Environmental UF President Health and Safety h Dean, College of En-Chairman, Radiation g Control Comittee ar h-f gineering Sciences Dept. LEVEL 2 Facil.ity Director SPERT Facility Safety , for SNM-1050 Redew Ssomhtee License LEVEL 3 Manager for SNM-1050 Radiation Control License Officer l LEVEL 4 Radiation Safety - Operating Staff Specialists FIGURE 1. UNIVERSITY OF FLORIDA SPERT ASSCiBLY (SNM-1050 LICDISE) QA PROGRAM ORGANIZATIONAL CHART 1 l 1 l L .. _ _____

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1 ss **" allows for the Expansion } i of Fission Cases s MNi C i w, . r @b

                                ~.                                             UO2 Fuel Pellet
                                                                                                                         )

s

                                                                                                                      .l' V

Stainicss Steel Cladding

         -                     +

SPERT FUEL ROD CMRACTERISTICS Length of F-1 Fuel rod 106.05 cm g Cladding Outside Diameter 1.184 cm Cladding Thickness 0.051 en Active Fuel Length 91.44 cm Fuel Pellet Outside Dianoter 1.067 cm Weight of U-235 Per Rod 35.2 gm Urantun-235 Enrichment 4.81 'ut% UO2 Fuel Pellet Density 10.03 g/cc

                 %L)

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                                /

Figure 3 Isometric of the SPERT Fuel llod Containing Uraniun-Dioxide Fuel Pellets

l I i I l I l l l APPENDIX II CORRECDON PAGE FOR TIIE 19851986 ANNUAL REPORT 1 k l 1

r TABLE IV PERMANENT BADGE EXPOSURE REPORTED ABOVE BACKGROUND oc tober, 1985 C.J. stiehl 210/210 deep /whole body P.M. Whaley 70/70 deep /whole body November, 1985 C.J. s tiehl 400/400 deep /whole body P.M. Whaley 400/400 deep /whole body December, 1985 W.M. Cason Cancelled

c.W. Pogle 60/60 deep /whole body i C.J. Stiehl 20/20 deep /whole body P.M. Whaley 250/250 deep /whole body

( January, 1986 G.W. Pogle 20/20 deep /whole body H. Gogun 250/350 deep /whole body { R.K. Hanson 130/130 deep /whole body C.J. Stiehl 440/440 deep /whole body W.G. Vernetson 100/100 deep /whole body l P.M. Whaley 430/430 deep /whole body February, 1986 c.w. rogle 80/80 deep /whole body P.M. Whaley 20/20 deep /whole body May, 1986

               . ':;-- .; t: : :      13?!?3?                 h ;f' Fr!= br&,

( . P.M. Whaley 20/20 deep /whole body July, 1986 j P.M. Whaley 20/20 deep /whole body ( Augus t, 1986 R.K. Har.cen 10/10 deep /whole body C.J. S tiehl 50/50 deep /whole body f P.M. Whaley 30/30 deep /whole body NOTE 1: Doses recorded in mrem. L m ------ -- ---

NUCLEAR ENGINHRING SCIENCES DEPARTMENT Nuclear Reactor Fac!:ity University of Florida a v~~ 01EI,N"..m,

                     -,..m...

1 j November 30,1988 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attn: Document Control Desk Re: Facility License R-56 Docket No. 50-83

Dear Sir:

In compliance with our Technical Specifications reporting require-ments, enclosed is one copy of the 1987-1988 University of Florida Training Reactor Annual Progress Report. This document complies with the requirements of the UFTR Technical Specifications, Section 6.61 Please advise if further information L needed. Sincerely, [t5hvL LON William G. Vernetson Director of Nuclear Facilities WGV/ps Enclosure ec: P.M. Whaley Acting Reactor Manager OY N Apqo S 41r1 dl

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