ML20214P404

Annual Progress Rept of Univ of Florida Training Reactor, Sept 1985 - Aug 1986

ML20214P404
Person / Time
Site:	05000083
Issue date:	08/31/1986
From:	Vernetson W FLORIDA, UNIV. OF, GAINESVILLE, FL
To:	Office of Nuclear Reactor Regulation
References
ORO-4014-16, NUDOCS 8612040185
Download: ML20214P404 (261)

Text

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Contract # DE-AC05-76ER04014 Report # ORO-4014-16 I

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l ANNUAL PROGRESS REPORT OF THE

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UNIVERSITY OF FLORIDA TRAINING REACTOR

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September 1,1985 - August 31, 1986

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By Dr. William G. Vernetson Associate Engineer and

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Director of Nuclear Fac111tles

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w Contract h 7E-AC05-76ER04014 Report #0RO--4014-16 I

I l ANNUAL PROGRESS REPORT OF THE.

UNIVERSITY OF FLORIDA TRAINING REACTOR September 1,1985 - August 31, 1986

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i By Dr. William G. Vernetson Director of Nuclear Facilities  ;

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I Department of Nuclear Engineering Sciences University of Florida Gainesville, Florida l

November, 1986

TABLE OF CONTENTS Page Number I. INTRODUCTION I

I-1 II. UNIVERSITY OF FLORIDA PERSONNEL ASSOCIATED WITH THE REACTOR II-1 III. FACILITY OPERATION III-1 I IV. MODIFICATIONS 'IO THE OPERATING CHARACTERISTICS OR CAPABILITIES OF THE UFTR FACILITY IV-1 V. SIGNIFICANT MAINTENANCE, TESTS AND SURVELLIANCES I OF UFTR REACTOR SYSTEMS AND FACILITIES V-1 VI. CHANGES TO TECHNICAL SPECIFICATIONS, STANDARD OPERATING PROCEDURES AND OTHER DOCUMENTS VI-1 VII. RADIOACTIVE RELEASES AND ENVIRONMENTAL SURVEILLANCE VII-1 VIII. EDUCATION, RESEARCH AND TRAINING UTILIZATION VIII-1 IX. THESES, PUBLICATIONS, REPORTS AND ORAL PRESENTATIONS r OF WORK RELATED TO THE USE AND OPERATION OF THE UPTR IX-1 APPENDIX A: UFTR FACILITY LICENSEE RESPONSE TO NRC INSPECTION REPORT NUMBER 50-83/86-01 APPENDIX B: FINAL

SUMMARY

REPORT TO NRC ON RECURRENCE I OF STICKING S-3 CONTROL BLADE PROBLEM OF SEPTEMBER 3, 1986: NOTIFICATIONS, CORRECTIVE ACTION, PREVENTIVE MAIN- '

TENANCE, TESTS AND SURVELLIANCES APPENDIX C: UFTR STANDARD OPERATING PROCEDURES:

I 1. UFTR SOP-0.3 " CONTROL AND DOCUMENTATION OF UFTR MODIFICATIONS"

2. UPTR SOP-0.4 "10 CFR 50.59 EVALUATION I AND DETERMINATION"

3. UPTR SOP-0.5 "UFTR QUALITY ASSURANCE PROGRAM"

4. UPTR SOP-E.8 " VERIFICATION OF UFTR NEGATIVE VOID COEFFICIENT OF REACTIVITY" I APPENDIX D: UFTR SAFETY ANALYSIS REPORT REVISION 2 DOCUMENTATION APPENDIX E: UFTR REACTOR OPERATOR RECERTIFICATION PLAN I

y-t I. INTRODUCTION The University of Florida Training Reactor underwent a number of opera-tional and adminstrative occurrences during the 1985-1986 reporting year that are planned to assure continued facility ava'ilability and usage along with ad-ministrative continuity. First, on September 3,1986, as the new year began, the discovery of the recurrence of the sticking S-3 control blade necessitated an extended administrative shutdown for over five months, Every effort was made to preclude this failure from recurring by performing extensive preven-tive and corrective maintenance on all control blade drive sys tems internal to the biological shield to assure restoration of control blade drive system op-erability as as near as possible to the as-built condition. Second, notifica-tion of renewal of the Department of Energy Reactor Sharing Support at an in-creased level provided continued impetus for reactor usage from educational institutions spread throughout the State of Florida. Despite the extended un-availability for the control blade drive system maintenance, significant reac-tor usage under the Rec: tor Sharing Program was maintained especially during the second half of the year with the total number of participating institu-tions increased over previous years. Third, the administration of the UFTR has been significantly stabilized as a full time Acting Reactor Manager was hired in May,1986 and a new perTnanent Director of the Facility was appointed and fully vested in May,1986 to replace the former Director who had been on an extended leave of absence and has returned to duties at the University of I' lorida in a new capacity.

The University of Florida Training Reactor's overall utilization for the past reporting year has been maintained at a high level consistent with the increased usage noted beginning in the 1983-1984 reporting year as compared to previous years. Usage continues to exceed the levels of utilization charac-I-1

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.I teriatic of the early 1980's in areas such as energy generation and experimen-tation. The total energy genera tion (KW-Hr) for this reporting year has de-creased considerably but is still at the third highest level in the last eight years despite nearly 50% unavailability for major repair and maintenance work on the control blade drive systems as well as minor repairs on other facility systems. The maintenance of a high level of utilization during this reporting year is all the more noteworthy when the extended forced outage is noted.

The major outage during the year followed the discovery on September 3, 1985 of a recurrence of the sticking control blade (Safety Blade #3) and its failure to drop and fully insert upon demand from the operating position. The maintenance and repair work associated with this failure involved a complete overhaul of all four (4) control blade drive systems internal to the biologi-cal shield as well as considerable work outside the shield as this event and efforts to correct the problem as well as to document the corrective and pre-ventive maintenance work dominated activities during the first six months of the reporting year. NRC was consulted at all major corrective steps in the recovery and performed two on-site inspections during the period in addition to receiving a number of status reports. The reactor was not returned to nor-mal operating status unti) March 3,1986. Despite this major outage, the larg-est in 15 years, interust in UFTR usage was maintained. As a result, upon re-turn to normal operating status, the UPTR was able to sustain previous high levels of usage in most areas without compromising safe operation of the fa-cility during the final six months of the year.

An analysis of facility utilization shows that the usage and energy gen-eration relative to the previous year is attributable to supporting conditions similar to last year. First, this reporting year is the third full year with complete installation of the new rabbit system and implementation of the asso-cia ted Neutron Activa tion Analysis Laboratory (NAAL) giving the staff the ca-l I I-2

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pability to promote it among University of Florida users and among researchers at other universities and colleges around the state of Florida. As its avail- ,

ability becomes better advertised, its usage continues to increase. Indeed, the point has been reached where reactor usage is limited by the NAAL capa-bilities which are under study for major upgrades in the upcoming year.

second, this reporting year was only the third ever in which the Univer-

.I sity of Florida Training Reactor was supported as part of the Department of Energy's Reactor Sharing Program. The Reactor Sharing Program is designed to increase the availability of University reactor facilities such as the UFTR to non-reactor owning colleges and universities (user institutions). Basically this grant provides funds against which reactor operating costs may be charged when the facilities are utilized by regionally affiliated user institutions for student instruction / training or for student or faculty research that is not supported by outside funding. In all, eleven (11) different academic in-stitutions around the State of Florida made use of this program to utilize the UFTR for research via neutron activation analysis to determine trace element compositions in various types of samples, to produce radioactive sources, and for reactor facility demonstrations of various aspects of operation and train-ing of students in various community college programs such as nuclear medicine technol.ogy and radiation protection technology, and finally- for support of college level courses at various institt tions. At years end, several unsup-ported research projects were still awai*lng availability of the UFTR under the Reactor Sharing Program as UPTR usage attributable to this DOE-sponsored program continues to grow. Despite considerable cost-sharing by the University of Florida and a major outage for over five months, all of 4the reactor sharing funds allocated by the Department of Energy for this supporting year will be fully utilized early in the new reporting year. For this reason, an increase in Reactor Sharing support was requested and received as the program has been I-3

I approved for renewal for the 1986-1987 reporting year.

Reactor use by University of Florida courses and laboratories continues at the substantial level established in the previous three years. Course and Department usages within the University range from the Environnental Engineer-ing Sciences Department in its graduate and undergraduate Health Physics labo-ratories to the Chemistry Department in a graduate level radiochemistry labo-ratory courses. A graduate anthropology course also used the UFTR for the first time this year; this usage may eventually lead to development of a NAA-related research project. Of course, the biggest single user department re-mains the Nuclear Engineering Sciences Department which uses the reactor 's-cility for both graduate and undergraduate laboratories, research projects and class demonstrations. Since its enrollment is growing, this usage area is ex-pected to grow as well.

The considerable test, maintenance and surveillan:. activities required by the facility license Technical Specifications or other controls also con-tributed significantly to usage. This contribution is larger than in most years because of the extensive maintenance project implemented upon recurrence of the sticking control blade to isolate and correct the problem and restore all control blade drive systems to full operability.

Although indications are that Florida Power Corporation and other utili-ties have been pleased with the UPTR staff and facilities and will continue I periodic utilization of the facilities for training its operations staff, no utility training utilization was conducted during this reporting year. This is an area where renewed efforts to obtain usage have not yet been successful.

With one training program tentatively scheduled along with continued a-vailability of the NAA laboratory and the newly improved remote sample-hand-ling " rabbit" sys tem soon to be implemented plus renewal of the Reactor Shar-ing Program support, facility utilization and energy generation for the up-I -

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coming year should be considerably augmented versus this year when the control

.I blade outage was so dominant. The latter augmentation is particularly possible because the UFTR utilizaton under the DOE Reactor Sharing 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 in-dicated a continuing interest in using the reactor facility and the functional

" rabbit" syste.n during the upcoming year. For example, the University of Cen-tral Florida used the reactor only for one tour / demonstration in this report-ing year but is planning to make extensive use of the facility in one or more coursas during the upcoming year. Several other statewide users, as a further outgrcoth of the DOE Reactor Sharing Program support, have submitted or will submit additional proposals hopefully to provide funded usage of the UFTR within the next two years. All of this provides valid expectations of con-tinued growth of reactor facility usage dependent on a continued upgrading of facility capabilities and staff expertise. For this reason the implementation of the improved new rabbit system is important as are plans to implement a neutron radiography facility, perhaps with real time imaging capabilities when this facility is implemented. Plans are also in progress to begin work on another new facility for neutron depth profiling.

As noted earlier, the facility administration was considerably stabilized I

by appointment in late May of a fully vested permanent Director of Facilities tr replace the previous Director who had been on leave of absence for nearly twa years. A pennanent full-time Acting Reactor Manager had also been ap-peinted earlier in the same month. In combination with the planned licensing of two part-time operators during the coming year, these conditions were all contributing to the considerable broad-based increases in facility usage for education and training of university students and utility operators as well as research by faculty at the University of Florida and other schools during the I-5

last four months of the reporting year. All other staff personnel are part-time employees, two of whom previously were full-time employees. Although such employees provide a good experience base for oeprations, the lack of other li-consed staff members during the current reporting year has necessitated limi-tations in the growth of some usage programs. It is hoped that these limita-tions will be removed during the upcoming reporting year as one more part-time operator is expected to be licensed early in the year with another to follow by year's end.

Several significant license-related administrative activities occurred during this reporting year. First, the facility was the subject of two (2) NRC I & E Cperations Inspections. The first was during the early stages of the maintenance to repair the control blade drive systems. No deficiencies were noted during the Inspection. A later NRC inspection during February 18-21, 1986 resulted in citations for inadequate documentation of items considered in making safety evaluations, specifically for a modification to install view-ports in the control blade shrouds to facilitate remote, ALARA-based inspec-tion of control blade integrity and freedom of movement. The facility was also asked to clarify several discrepancies and/or omissions in the Safety Analysis Report. The full facility response to Inspection Report 50-83/86-01 is con-tained in Appendix A of this report. The final report to NRC on the repair project to restore control blade operation is contained in Appendix B for re-ference purposes. Following the commitments made in the response to the NRC Inspection, the modification to the control blade shrouds has been re-eval-uated and fully documented as involving no unreviewed safety questions; in ad-di tion , the procedure cotrolling the documentation of such evaluations has been revised (included in Appendix C) and a revision of the UFTR Safety Analy-sis Report has been submitted to NRC. The revision along with explanationn is contained in Appendix D. Only the evaluation of Wigner Energy Storage remains I-6

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to be completed; work is in progress on this project which is not due for com-pletion unti1 August, 1987.

Second, as discussed earlier, there is a new Director of Facilities and a full-time SRO/ Acting Reactor Manager to provide more versatile management ca-pabilities for reactor operations. This administrative arrangement meets all regulatory requirements and has enabled the facility to meet all regulatory commitments while continuing to meet facility usage commitments.

Third, major maintenance and surveillance efforts were undertaken during the year resulting in nearly 50% unavailability of the reactor. The failure of control blade Safety #3 to drop from its normal operating position on demand involved over five months of administrative shutdown. During this time all control blade drives were restored to nomal operation with_ preventive main-tenance perfomed on all four control blade drives to include all mechanisms internal to the biological shielding. Although this maintenants was forced by the failure of a blade to drop, the work perfomed was largely part of a re-quired five year mechanical inspection of the reactor control blade system.

Some of the work performed (such as the cutting of viewports in the shrouds) will facilitate further checks of mechanical integrity. Therefore, much of the work would have been perfomed in the next several years anyway. In general, the level of maintenance activity was much higher during this reporting year than in any year since 1969, but it is expected that the efforts dedicated to maintenance should involve increased availablity in the next few years. Op-erating records for the last six months of this reporting year support this expectation. As indicated, the final report to NRC on the sticking control blade is contained in Appendix B.

Fourth, due to the extended outage for the control blade drive mainten-ance, all licensed operators were required to be recertified for knowledge and operator capabilities prior to the facility resuming nomal operations. The I I-7

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l NRC approved program used to provide the basis for this recertification is contained in Appendix E.

Finally, the facility developed a special QA Program to ' control shipment of 40 plates of High Enriched Uranium following a show cause order received early in the reporting year. The fuel was finally shipped out in February, 1986.

The UFTR continues to operate with an outstanding safety record and in full compliance with regulatory requirements. An NRC Security Inspection dur-ing the year resulted in only minor recommendations and comments with no cita-tions. All significant recommendations have been implemented. Two additional NRC (I & E Operations) inspections were conducted during the year. In addition to reviewing the treatment of the S-3 maintenance project as outlined above, the second inspection closed out the item on implementing a UFTR Quality As-surance Program as per ANSI S tandard ANSI N402-1976, " Quality Assurance Pro-gram Requirements for Research Reactors." Basically, a commitment had been made to NRC in Spring,1985 to develop a set of procedures to address quality assurance program requirements for Research Reactors as delineated in guide-lines in ANSI N402-1976. The three SOPS (0.3, 0.4 and 0.5) contained in Appen-dix C are now fully implemented in the UFTR Quality Assurance Program which has been approved by the NRC. In general, none of these NRC findings over the past two years involved any actual safety problems but rather involved a lack of supporting documentation. As indicated, the UFTR continues to operate with an outstanding safe ty record. Similarly three inspections by representatives of the American Nuclear Insurers resulted in only minor recommendations.

The reactor and associated facilities continue to maintain a high in-atate visibility and atrong industry relationships. Plans are underway to have a visit by industry executives to support utility interactions. With the DOE Reactor Sharing Program to support UFTR-related research by faculty and stu-I-8

dents at other academic institutions as well as training for various community college and university programs around the state, the reactor facility is also maintaining high in-state visibility with other institutions of higher learn-ing. The interactions of several small externally supported research programs as a result of the Reactor Sharing Work is further proof of its effectiveness.

With the renewed statewide interest, the facility is being included in several proposals to provide for funded usage of the UFTR and the NAA Labor-atory. The Reactor Sharing Program began in late 1983 and is directly respon-sible for the generation of several of these proposals. As more of these pro-posals are submitted and funded, further increases in UPTR usage can be ex-pec ted. In any case, on-campus research and service usaga of the UPTR is also increasing because of the visibility generated via the Reactor Sharing Pro-gram.

It is expected that more direct industry training will be accomplished in the upcoming year hopefully accompanied by further increases in research pri-marily through the use of the rabbit system and the associated NAAL facility both under the DOE Reactor Sharing Program and hopefully from research funded from other agencies, some of which has been developed from research begun un-der the Reactor Sharing Program. Neutron Radiography Pacility is one area where industry has indicated interest. It is also hoped that the upcoming visit by industry executives will be supportive of industry usage as the UPTR continues to experience growth in usage.

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II. UNIVERSITY OF FLORIDA PERSONNEL ASSOCIATED WITil THE REACTOR A. Personnel Employed by the UFTR I N.J. Diaz - Professor and Director of Nuclear Facilities (September 1,1985 - May 29,1986) (continued on leave of absence)

W.G. Vernetson - Associate Engineer and Acting Director of Nuclear Fa-cilities (September 1, 1985 - May 29, 1986)

- Associate Engineer and Director of Nuclear Facilities (May 30, 1986 - August 31, 1986)

P.M. Whaley - Acting Reactor Manager (3/4 time) (September,1985 -

May 4,1986)

- Acting Reactor Manager (full-time) (May 5, 1986 -

August 31, 1986)

I II. Gogun - Senior Reactor Operator (part-time) (September,1985 -

August, 1986)

G. Fogle - Reactor Operator (part-time) (Sep tember, 1985 -

Augus t, 1986)

I C.J. Stiehl - Student Reactor Operator Trainee (1/2 time)

(September,1985 - August,1986)

R. Ilanson - Student Reactor Operator Trainee (1/3 time) (January, 1986 - August, 1986)

P. Stevens - Secretary Specialist (3/4 time) (September,1985 -

August, 1986)

I B. Radiation Control Office D. Munroe - Radiation Control Officer (September,1985 - August, 1986)

II . G . Norton - Radiation Control Technician (September, 1985 -

August, 1986)

G.R. Renshaw - Radiation Control Technician (September,1985 -

August, 1986)

D.E. Perkins - Radiation Control Technician (September, 1985 -

February, 1986)

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I K. Deschesne - Nuclear Technician (1/2 time) (September,1985 -

I August, 1986)

T. Hines - Nuclear Technician (1/2 time) (September,1985 -

December, 1986)

E. Schmidt - Nuclear Technician (1/2 time) (September,1985 -

February, 1986)

S. Armistead - Nuclear Technician (1/3 time) (September,1985 -

February, 1986)

R. Hagen - Nuclear Technician (1/3 time) (September, 1985 -

February, 1986)

Basic routine health physics is perfomed by UFTR Staff; however, assistance from the Radiation Control Office is required for many operations espcially I the maintenance work performed this year in the core area to correct the sticking control blades. As a result, many radiation control office personnel are listed and though employed 1/3,1/2 or full time, only a small fraction of their work effort supports UFTH activities.

C. Reactor Safety Review Subcommittee (RSRS)

M.J. Ohanian - RSRS Chaiman, Acsociate Dean for Research, College of Engineering and Professor, Nuclear Engineering Sciences Department W.G. Vernetson - Member - Reactor Manager and (Acting) Director of Nuclear Facilities G.S. Roessle - Member (NES Department Chairman)

J.S.Tulenko{and W.E. Bolch - Member-at-large ,

D. Munroe - Member (Radiation Control Officer)

D. Line Responsibility for UFTR Administration M.M. Criser, Jr. - Precident, University of Florida W.II. Chen - Dean, College of Engineering G.S. Roessler 2 - Acting Chairnan, Depurtment of Nuclear En-I ginotering Sciences (September 1, 1986) 1985 - May 4, I J.S. Tulenko - Chaiman (May 5 - August 31, 1986)

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W.G. Vernetson 3 - Acting Director of Nuclear Facilities (September 1,1985 - May 29,1986)

- Director of Nuclear Facilities (May 30, 1986 -

August 3',1986)

P.M. Whaley4 - Acting Reactor Manager E. Line Responsibility for the Radiation Control Office M.M. Criser, Jr. - President, University of Florida W.E. Elmore - Vice President Administrative Affairs W.S. Properzio - Director, Environmental Health and Safety D. Munroe - Radiation Control Officer For line responsibility for the Radiation Control office, all personnel were employed in permanent positions fo. the full year.

I I Note 1: J.S. Tulenko currently holds the position of permanent chairman, De-partment of Nuclear Engineering Sciences replacing Dr. G.S. Roessler as of May 5, 1986.

Note 2: D r. Genevieve S. Roessler served as Acting Chairman of the Nuclear Engineering Department for most of the year until May 5, 1986 when Professor James S. Tulenko assumed the permanent position of Chair-man of the Nuclear Engineering Sciences Department.

Note 3,4: D r. N.J. Diaz was on leave for most of the reporting year. In his absence, D r. W.G. Vernetson continued in his appointment to the po-sition of Acting Director of Nuclear Facilities until May 29, 1986 with Mr. P.M. Whaley serving as Acting Reactor Manager. As of May 30, 1986, Dr. W.G. Vernetson assumed the pennanent position of Director of Nuclear Facilities with Mr. P.M. Whaley continuing as the Acting Reactor Manager.

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w III. FACILITY OPERATION Despite Induced availability, the UFTR continues to experience a high rate I of utilization in many areas when compared to the last reporting year, with total utilization continuing near the highest levels recorded in the early 1970's. This increase has been supported by a variety of usages ranging from research and edu-cational utilization by users within the University of Florida as well as other researchers and educators around the State of Florida through the support of the DOE Reactor Sharing Program.

As noted the last two years, the development of the Neutron Activation Analy-sis laboratory has improved research irradiation utilization. With successful im-pliamentation of the remote sample-handling " rabbit" facility, efforts to advertise availability and encourage usage of the UPTR (especially for research) have pro-ceeding favorably. Under the Reactor Sharing Program there has been significant usage by users from other schools with many more planned and some proposals for separate funding in progress. In addition, there have been a number of usages a-mong researchers at the University of Florida with several more noted again this year. Although there were no commercial research irradiations this year, it is hoped some commercial irradiations will be forthcoming during this next year to ,

further complement UPTR operating activities.

The level of administrative work dedicated to regulatory activities is ex-pected to be at a similar level during this next reporting year due to close o' ut of items cited by the NRC following the February 11-15, 1985 inspection where the UFTR Licensee was

• cited for failure to properly control a modification and to im-plerrent a Quality Assurance in accordance with guidelines in ANSI Stand.1rd N402-1976. Although the deficient items identified in the area of Quality Assurance were closed out during the February,1986 NRC inspection and will not require ad-ditional work, work to close out the item on Wigner Energy evaluation noted in I i I III-1 I

P this year's inspection will involve some considerable administrative effort.

Shown in Table. I is a summary breakdown of the reactor utilization for this reporting period. The list breaks UPTR utilization down into the 55 dif-ferent researt.h projects, various tests, teaching and training activities. The total reactor run-time was about 387 hours0.00448 days <br />0.108 hours <br />6.398809e-4 weeks <br />1.472535e-4 months <br /> while the various experiments and other projects used over 1232 hours0.0143 days <br />0.342 hours <br />0.00204 weeks <br />4.68776e-4 months <br /> of facility time. The run time represents a significant decrease of ~36% from last year though there were many concur-rent usages during the current year to optimize utilization of available per-sonnel. In contrast the experiment time represents only a 7.8% without accoun-ting for nearly 270 hours0.00313 days <br />0.075 hours <br />4.464286e-4 weeks <br />1.02735e-4 months <br /> of concurrent experiment time. These decreases are directly attributable, especially the run time to the low 52.3% availability of the UFTR during the 1985-1986 reporting year. The primary cause of this near 50% outage rate was the repair and maintenance activities undertaken dur-ing the first six months to restore all control blade drive systems to normal operation. Without this outage rate, run time might well have exceeded the highest level recorded in the 1983-1984 reporting year and certainly would have exceeded those of the 1984-1985 reporting year.

In summary these figures indicate continued high utilization of the UFTR facility usage over the last three years despite the large outage rate this '

year. The design and planned implementation of new facilities to include the new designed rabbit system with larger capacity and easier maintenance as well as a neutron radiography facility should provide for continued growth and di-versification of usage during the upcoming year. Of course, the Reactor Shar-ing Program is planned to continue to play a key role in encouraging facility usage as this support has again been renewed.

Table II summarizes the different categories of reactor utilization: col-lege and university teaching, research projects, UFTR cperator training, re-qualification, and recertification, testing, maintenance and surveillance ac-1 I III-2 l

I tivities, and various tours and reactor operations demonstrations which is a final category to account for all other planned usages. The absence of utility operator training is a noteworthy point; every effort is being expended to schedule such usages during the upcoming year. College course utilization in-volved 19 different courses, some more than once to account for over 130 hours0.0015 days <br />0.0361 hours <br />2.149471e-4 weeks <br />4.9465e-5 months <br /> of actual run time. The research utilization consisted of 16 projects using a-bout 190 hours0.0022 days <br />0.0528 hours <br />3.141534e-4 weeks <br />7.2295e-5 months <br /> of actual reactor run-time. Both these categories included con-siderable concurrent usage. As noted, there are increases in several areas from the last reporting year, especially in the research and training supported un-der the DOE Reactor Sharing Program. This program plus the large amount of maintenance, testing and surveillance activities are primarily responsible for the total facility utilization continuing to be one of the highest in UFTR history (considering relatively low availability) especially since grow'th in University of Florida course usage has shown only a slow increase. With one utility training program tentatively scheduled for the upcoming year and sev-eral research activities 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 previous two years. A utility training program alone could produce a substantial increase in run time by itself. With the I problem of the sticking control blades concluded to include NRC inspection ap-proval for its handling, this expected hgh usage in the upcoming year is real-istic especially in the areas of college courses and research.

Table III contains a breakdown delinea ting the 11 schools and their 54 usages of the UPTR facilities which were sponsored under the Department of En-I ergy Reactor Sharing Program grant. These sponsored usages account for about 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> of run time in Category I in Table II and over 190 hours0.0022 days <br />0.0528 hours <br />3.141534e-4 weeks <br />7.2295e-5 months <br /> of run time in Category II and have resulted in much improved visibility for the UFTR around the State of Florids and also among researchers and other users at the Univer-I III-3

P sity of Florida. Several new inquiries for involvement in this progra have been received again this year. In all, the 54 usages represent a slight de-Credse from last year; while the total of 24 participating faculty repre?ents a suostantial increase (double last year). The 99 students involved repr9sente a healthy 25% increase. Obviously this Program is the driving force behind the continued utilization and growing interest in the UFTR facility. This publicity is certainly a key factor explaining the over 500 visitors who toured the fa-cility this year. With several proposals for funding in progress and one new funded usage of the facility based on reactor sharing research results, the UFTR facility continues to build a base for long-term permanent growth of fa-cility utilization with the Reactor Sharing Program serving as the catalyst for this growth. The improved Rabbit System soon to be implemented and the planned new neutron radiography facility are simply spinoffs from the various expressed needs of those visiting the facility in conjunction with staff in-terests in diversification of capabilities.

Detailed in Table IV are the monthly and total energy generation, as well as the hours at full-power per month and totals for this past year. The UFTR generated 19.25 Mw-hrs during this twelve month reporting period, down nearly 45% from last year but still one of the larger values in UFTR operating his-tory. This decrease is primarily due to the outage for correcting the problem of the sticking control blade. Of course, there were several research usages such as for the Plasma Kinetics Diagnostics Test Project whera the usage was lengthy but at relatively low or fluctuating power levels. Finally the 52.3%

availability factor for the year did account for some considerable lost power generation and run time. Indeed, several projects were forced to find an al-terna te facility.

Described in Table V is the monthly breakdown of usage and availability data. As was noted in Section II of this report, extended forced outages for III-4

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maintenance were responsible for low availability during the first six months of the 1985-1986 reporting year. Similarly, though available much of the month of February, the reactor was limited to low power or occasional operation as extensive tests and surveillances were conducted in preparation for returning m to normal operations. Similarly, Table VI contains a detailed breakdown of days unavailable each month with a brief description of the primary cause. The over-all availability of 52.3% is one' of the two lowest documentable values in UFTR history, primarily due to the control blade drive system maintenance project.

Described in Table VII-A is the reason and date for one unscheduled trip for the reporting period. Table VII-B contains no entries for scheduled trips, primarily due to lack of utility training programs where such trips are part of the training exercise. All safety systems responded properly for the one trip.

Several reportable incidents occurred during this reporting year. Table VIII contains a descriptive log of ten (10) unusual. occurrences with brief descrip-tive evaluations of each. Again the recurrence of the sticking control blade is dominant. However, each is described in some detail as several were promptly reportable while the rest are reported in this report and in scveral cases do not need to be reported at all.

No uncontrolled releases of radioactivity have occurred from the facility and controlled releases are well within established limits. The personnel radia-tion exposures were above the usual low yearly level primarily due to the con-

. trol blade drive system maintenance project as delineated in Section VII. The one waste shipment this year was the first in some years. The shipment of 40 un-irradiated HEU fuel plates was also a unique activity. Environmental radioactivi-ty surveillance continues to show no detectable off-site dose attributable to the UPTR facility as also noted in Section VII as the facility operates within the ALARA guidelines.

I I _

III-5

I TABLE I

SUMMARY

OF FACILITY UTILIZATION (September,1985 - August, 1986)

.I-NOTE: The projects marked with a

• indicate irradiations ot neutron activa-tions. The projects marked with an ** indicate training / educational use. The projects marked with an *** indicate demonstrations of reactor operations. " Experiment Time" is total time that the facility dedicates I 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 EXPERIMENT TIME TIME PROJECT AND USER TYPE OF ACTIVITY (hours) (hours)

Sticking S-3 and Other Repair, Corrective and Pre- 2.28 189.95 Control Blades - ventive Maintenance on all UPTR (2.50)

.l Dr. W.3. Vernetson Control Blade Drive Systems Pri-3 marily Concentrating Work Internal to Biological Shield

• NAA Research - UFTR Primary Coolant. Sample 13.51 18.68 Dr. W.E . Bolch, - Selection and Evaluation With (2.13) (2.63)

Dr. W.G. Vernetson, Subsequent Irradiation of Water I R. Knecht Samples for Primary Coolant Baseline Activity Analysis and Evalua tion

• • ENU-4905/6937 - Independent Reactor Operations 160.76 334.49 Dr. W.G. Verne tson/ Laboratory Course for Under- (58.51) (95.31)

Reactor Staff graduate and Graduate Nuclear Engineering Students Cerenkov Detector Conclusion of Reactor Radiation 2.00 I

0.00 Development - Measurements to Test and Cali-Dr. E.E. Carroll brate a New Cerenkov Radiation Detector System

• • CFCC Radiation Pro- Two Reactor Operations-Based 9.50 112.53 tection Technology Radiological Control and Pro- (6.22) (5.30) tection Training Programs of I

Program - Mr. G.

S tephenson - Reac- Cooperative Work Exercises tor Sharing I

I III-6

us I TABLE I (CONTINUED)

RUN TIME EXPERIMENT TIME PROJECT AND USER TYPE OF ACTIVITY (hours) (hours)

Fire / Security System Responses to Secure Fire, 0.00 43.37 Alarms Responses - Security and Equipment Pit Water Dr. W.G. Vernetson, Leakage Alarms as Well as Make Reactor Staff Necessary Repairs Where Indicated I UPTR Task Analysis Survey - Dr. John Randall/Dr. W.G.

Interviews to Perform a UFTR Task Analysis Survey Followed by. Staff Ratings and Evaluation of the 0.00 9.50 I Verne tson Generated Task Inventory List to Produce a Final Generic List

• • Gainesville Buchholz Lecture, Tour and Demonstra- 0.00 1.25 l High School Science tion of Facility Operations W S tudents - Dr. G.

Williams

• • Rollins College Lecture, Tour and Demonstra- 0.00 2.50 Society of Physics tion of UFTR Facilities With

S tudents - Dr. J. Radiation Surveys and NAA Polley - Reactor Exercises Sharing I Irradiated Fuel Move-ments - Dr. W.G.

Verne tson Irradiated Fuel Storage Pit Consolidation, Core Unloading 7.60 41.11 and Core Reloading Following

'E Correction of Problem of l3 Sticking Control Blades

• • Licensed Operator NRC Requalification Training 2.48 36.35 Requalifica tion Requirements Including Lec-Training - Dr. W.G. tures, Demonstrations With Verne tson/ Reactor Startups, Shutdowns and Reac-S taff tivity Manipulations as Necessary

- Also Includes Recertification Training Following Extended Outage I *** Florida Foundation of Future Scien-Lecture, Tour and Demonstration of Reactor Facility Operations 1.00 7.00 tis ts - Dr. W.G. and Experimental Capabilities Verne tson for Two FFFS High School Students

• • ENU-4103 - Lecture, Tour and Demons tration 0.00 4.17

'I Dr. G.R. Dalton of Reactor Facility Capabilities and NAA Lab Use for Junior Level Nuclear Engineering Students

. Il III-7

%r 1

1 I TABLE I (CONTINUED)

RUN TIME EXPERIMENT TIME I PROJECT AND USER Waste Processing and TYPE OF ACTIVITY Preparation of Two Drums and (hours) 0.00 (hours) 6.75 l

l Shipping - Dr. W.G. Shipment of One Drum of Reactor Verne tson Radioactive Waste

• • ENU 4505L - Senior Level Nuclear Engineer- 27.16 50.16 I Dr. G.R. Dalton /

Dr. W.E. Ellis ing Laboratory Exercises and Experimen ts (15.38) (26.33)

• • • Florida Regional Ju- Three Lectures, Tours and 0.00 5.08 nior Science, Engi- Demons trations of Facility neering and Humanities Capabilities Symposium - Dr. W.G.

Vernetson/ Reactor Staff

• NAA Research - Dr. T. Evaluation of Effects of Oil- 61 .21 65.17 I D' Asaro - University of West Florida -

Reactor Sharing Related Drilling Fluids .on Sea-Grass Communities (5.30) (5.40)

• • ENV-4201 - Dr. Lectures, Tours and Demonstra- 0.40 4.70 C.E. Roessler tions of Reactor Facility In-strumentation and Reactor Op-erations 1986 Engineers' Fair Lectures, Tours and Demonstrations 0.92 5.25 I - Dr. W.G. Vernetson/ of Facility Operations Reactor S taff

• NAA Research - Dr.

NAA of Volcanic Rock Samples 39.50 41.84 Mark DeFant - Uni- For Tracts Element Identifica- (3.70) (4.12) versity of South tion of sample Origin Florida (Tampa) -

Reactor Sharing

• • ENV-6211 - Dr. Lecture, Tour and Demonstration 0.30 0.92 W.E. Bolch/Dr. W. of Reactor Facility Capabilities (0.30) (0.92)

Properzio and Reactor Operations High Enriched Uranium Preparation and Shipment of HEU 0.00 3.00

Shipment - Dr. W.G. Fuel Plates to Meet NRC Show Verr ^ r son Cause Order I *NAA Research - Dr. T.

Stocker - University of West Florida -

Evaluation of Trace Element Com-position of Archeological Sea-3.93 (1.60) 4.92 (1.89) shell Specimens for Determination Reactor Sharing of Eas tern United States Trade Routes III-8

TABLE I (CONTINUED)

RUN EXPERIMENT TIME TIME PROJECT AND USER TYPE OF ACTIVITY (hours) (hours)

• NAA Research - Dr. Cadmium Ratic Determinations 18.95 26.34 G. Smith /Dr. R. in the Special Center Vertical (9.73) (13.08)

Byrne - University Port Graphite Sample Holder and of South Florida, NAA to Evaluate Rare Earth Elements St. Petersburg - in Tampa Bay Estuary Sediments Reactor Sharing

• • Hillsborough Com- Lecture, Tour and Demonstration 0.88 2.67 munity Collegcs Nu- of Facility Operations with clear Medicine and Radiation Surveys and NAA Radiation Therapy Lectures and Training Exercises Technology Program -

Dr. M. Lombardi/Ms.

Diane Fricks -

Reactor Sharing Nuclear Chemistry Irradiation of Isotopically 2.40 3.50 Laboratory Half- Pure Samples for Use in Nuclear Chemistry Laboratory Exercises I

Life Exercise - Dr.

S. Grossman - Uni - to Determine Half-Lives of versity of South Selected Nuclides Florida (Tampa) -

Reactor Sharing

• • Santa Fe Community Lecture, Tour and Demonstration 0.97 2.67 I College Nuclear Med-icine Radiologic Technology Program -

of UPTR Operations with Radiation Surveys and NAA Training Exercises S. Marchionno -

1 Reactor Sharing

• • ANT-6128 - Dr. B. Lecture, Tour and Demonstration 0.33 0.75 I Purdy of Reactor Facility Operations and capabilities for Graduate Students in a Lithic Technology Class

• • University of Cen- Tour and Demonstration of 0.00 1.75 tral Florida Phyoics I

Facility Capabilities For Use Department - Dr. R. In a Nuclear Physics course Llewellyn - Reactor Sharing

• • S t. Augustine High Lecture, Tour and Demonstra- 0.00 3.50 School Science tion of Reactor Facility I Class - Ms. E. Doyle Reactor Sharing Operations and NAA Methods I

III-9

M I TABLE I (CONTINUED)

RUN TIME EXPERIMENT TIME PROJECT AND USER TYPE OF ACTIVITY (hours) (hours)

• • PHY-2001 - Dr. K. Lectures, Tours and Demons tra- 0.52 2.42 Eoff tions of Reactor Operations Rabbit System De- Evaluation and Removal of Failed 1.62 25.23 velopment - Dr. Rapid Pneumatic Delivery (Rabbit) (9.50)

I W.G. Vernetson System With Redesign and Implemen-tation of Improved Rabbit System I **UFTR Reactor Operator Reactor Operator Training for Training - Dr. W.G.

Verne tson/ Reactor UFTR Reactor Operator Candidate R. Hanson 29.34 (18.23) 86.12 (35.96)

S ta ff

• NAA Research - Dr. Evaluation of Rare Earth Con- 9.35 15.68 M. El Haddad tent of Egyptian Phosphorites (0.45) (1.67)

• • ENV-6211L - Dr. W. Graduate Level Health Physics 4.40 4.50 Properzio/Dr. W.G. Laboratory Exercises and Ex-Verne tson periments

• • Radiation Surveys / Radiation Surveys of UPTR Cell 15.82 17.49 RadCon Training - and Environment at Steady-State I Radiation Control /

UFTR Staf f Full Power Plus Training of Ra-diation Control Personnel (Inclu-(8.05) (8.17) ding Second Person Qualification)

~ *NAA Reeearch - Dr. Trace Element Analysis of Con- 22.03 24.00 Guy Prentice stituents of Environmental Sea- (11.37) (12.38) shells Flux Mapping - Dr. Absolute Flux Mapping of Various 10.75 18.83 W.H. Ellis/Dr. UPTR Experimental Facilities (1.15) (2.92)

W.G. Vernetson i

• • ENU-5005 - Lecture, Tour and Demonstration 0.43 2.10 I Dr. R. Pagano of Reactor Operations for Non-Nuclear Engineering Students j

• • Lincoln Middle School Lecture, Tour and Demonstration 0.20 1.83

.I Science S tudents - of Reactor Operations Ms. Sara Mesa -

Reactor. Sharing l

• NAA Research - Dr. Evaluation of Trace Elements 2.01 2.67 H. VanRinsvelt in Brain Tissue Using NAA for (0.60) (0.83)

Comparison With PIXE Analysis I III-10

M TABLE I (CONTINUED)

-I RUN EXPERIMENT TIME TIME PROJECT AND USER TYPE OF ACTIVITY (hours) (hours)

Florida S tate Uni- Irradiation of a Pure Thullium 28.33 33.50 versity Chemistry oxide (Tm 23 0 ) Sample For Use as (13.92)

I Dept. - Dr. G.

Choppin - Reactor a Beta Source (15.00)

Sharing

• NAA Research - Evaluation of Trace El' ment e Con- 6.35 7.00

.Dr. A.L. Odum - tribution to Radioactivity in (2.00) (2.12)

Reactor Sharing Purified Quartz Samples to be Subjected to Repeated Irradiations

• • Dosimetry Exercises - Evaluation of Reactor Neutron I Dr. W.H. Ellis Leakage and Its Detection for Dosimetry Calculations 1.00 (1.00) 1.00 (1.00)~

I ENU-6646/ENV-6905 -

Dr. W. Properzio/

Dr. W.G. Verne tson Graduate Level Health Physics 3.52 Practice Course Involving Reactor (2.00)

Related Exercises 6.42 (2.10)

• Florida Foundation of Summer 1986 Student Research 10.31 17.03 Future Scientists Program: Experimental Measurement (2.28) (3.17)

(NAA Research) - and Evaluation of UPTR Graphite Dr. W.G. Vernetson/ Temperature Changes During Power Derek Roberts Operations I

• Florida Foundation of Future Scientists (NAA Research) -

Summer 1986 Student Research Program: Determination of the 0.37 (0.18) 1.33 (0.67)

Lower Limit of Detection for Dr. W.G. Vernetson/ Various NBS Standard Reference Trung Lively Sources

• • CHS 5110L - Lecture, Tour and Demonstration 0.00 1.50 Dr. K. Williams of Reactor and NAA Lab Operations for Radiochemistry Research Laboratory Course

• • CHS-5110L - Radiochemistry Laboratory Project 2.00 2.83 Dr. K. Williams on Neutron Activation Analysis of Labeled Antibodies Using Irradiated Iodine as Tracer

• Neutron Radiography Neutron and Gamma Flux Measure-I Facility Develop-men t - Dr. W.G.

ments in the Thermal Column Fa-cility With Rearrangement of Ther-11.52 (1.86) 55.48 (3.25)

Verne tson, Dr. A.M. mal Column to Optimize Neutron I Jacobs and Reactor S ta ff Radiography Potential Based on a Series of Test Radiographs III-11

H I TABLE I (CONTINUED)

RUN EXPERIMENT TIME TIME PROJECT AND USER TYPE OF ACTIVITY (hours) (hours)

• Innovative Nuclear Pulsed Ionization Chamber Plasma 7.57 24.25 Space Power In- Kinetics Diagnostic System Opera- (4.40) stitute - Plasma tional Tests Kinetics Parameters Determination - Seed Project - Dr. W.H.

Ellis Maic Steel Sample Reception, Storage, Handling and 0.00 7.83 S torage - Dr. Preparation of Irradiated S teel Specimens for. Materials Science Analysis Argon-41 Effluent De- Argon-41 Stack Concentration 8.32 13.09 terminations - Dr. Measurements and Evaluation (2.32) (2.12)

I W.G. Vernetson/

Reactor Staff Miscellaneous - Dr. NRC Inspection and Enforcement 3.07 22.76 W.G. Vernetson Region II Inspections (Opera- (0.17) (0.42) tions and NMSS) , ANI Inspec-tions (3), Miscellaneous Tours Involving Facility Demonstra-tions for Potential New Staff Members, Film Interviews , Pic-I ture Taking Sessions, Visits by Representatives of the News Media I Technical Specifica-tion Requirements -

Reactor Staff UFTR Facility Component and System Tes ting, Surveillance, Calibration and Related Mea-27.70 (5.00) 59.16 (6.16) surement and Verification Activities Required By Tech-nical Specifications, Pro-cedures or NRC Commitments I

III-12

___ 1

v TABLE I (CONTINUED)

RUN EXPERIMENT TIME TIME PROJECT AND USER TYPE OF ACTIVITY (hours) (hours)

Maintenance Activities Preventive and Corrective 0.00 16.00

- Reactor S taff Maintenance and/or Replace-ment of UFTR Facility Com-ponents Excluding Minor Items and Those Listed Individually To Include System Testing as Necessary TOTAL1 ,2 560.61 1501.97 (173.45) (269.32)

, TOTAL ACTUAL 387.16 1232.65 NOTE 1: Values in parentheses represent multiple or concurrent facility uti-lization (Run or Experiment time); that is , the reactor was already being utilized in a primary run or activity for a project so a rsac-

. tor training or demonstration utilization could be conducted concur-rently with a scheduled NAA 1rradiation', course experiment, or other I. reactor run. Thus , the actual reactor run time for the 1985-1986 re-porting year is 387.16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, a significant decrease of ~36% over the I previous year (607.95 hours0.0011 days <br />0.0264 hours <br />1.570767e-4 weeks <br />3.61475e-5 months <br />). In contrast, the actual experiment time for the 1985-1986 reporting year is essentially unchanged at 1232.65 hours7.523148e-4 days <br />0.0181 hours <br />1.074735e-4 weeks <br />2.47325e-5 months <br />, a decrease of only 7.8% indicating more effective utilization of staff time. Of course, the experiment time includes extensive I reactor usage for t.he corrective maintenance actions to restore the operability of the control blade drive system but also indicates con-tinued levels of high facilitiy usage, especially as driven by the Reactor Sharing Program usages.

NOTE 2: Exp. Time is run time (total key on time minus checkout time) plus I set-up time for experiments or other reactor or facility usage in-cluding checkouts, tests and maintenance involving the reactor facility.

I I III-13

v i

TABLE II UFTR UTILIZATION

SUMMARY

Run Time Experiment Time (hours) (hours)

1. College Courses and Laboratories (17) 214.57 (83.41) 534.55 (130.96)

2. Research Activities (16) 245.69 (56.27) 351.72 (73.53)

3. UPTR Operator Training and Re-qualification (3) 31.82 (18.23) 131.97 (35.96)

4. UFTR Maintenance, Testing and Surveillance Activities (10) 63.34 (15.37) 415.15 (28.45)

5. Reactor Tours and Demonstrations (10) 5.19 (0.17) 58.75 (0.42)

I TOTAL 560.61 (173.45) 1501.97 (269.32)

I NOTE 1 : The same meaning is attached to values in parentheses as in Table I.

Values in parentheses adjacent to topic areas indicate number of en-tries from Table I that were collapsed into this utilization category.

NOTE 2: The first two categories of College Courses and Laboratories as well I, as Research Activities include significant usages sponsored under the Department of Energy UPTR Reactor Sharing Program which allowe?

eleven (11) schools to have 54 usages of the UFTR facilities as de-I lineated in Table III. This usage by 11 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 six categories is somewhat arbitrary especially for non-college tour groups for whom lectures and other training is conducted or research performed to aid facility modification or development.

NOTE 4: Console checks are excluded from this Utilization Summary. In addi-

.E tion, non specialized and usually non-scheduled tours for one or a g few persons are not normally tracked in this Utilization Summary.

These types of activities typically involve about 5-10 hours of addi-tional time per month and are offered on an as-available basis de-I' pending on staff availability.

I III-14

1 I

l l

TABLE III 1985-1986 REACTOR SHARING PROGRAM

SUMMARY

OF USAGE OF UFTR FACILITIES I.

Users School Usages

• Faculty Students Central Florida Community College (CFCC) 27 2 12 I Florida S tate University 4 4 3 I Hillsborough Community College (HCC) 1 2 14 I Lincoln Middle School 1 2 19 Rollins College 1 1 7 Santa Fe Community College (SFCC) 1 1 7 St. Augustine High School (SAHS) 1 3 20 University of Central Florida (UCF) 1 1 0 University of South Florida, S t. Petersburg (USF-SP) 2 2 1 University of South Florida, Tampa (USF-TA) 7 2 14

University of West Florida (UWF) 8 4 2 TorAL 54 24 99

• Usage is defined as utilization of the University of Florida Training Reac-tor for all or any part of a day. In many cases a school can have multiple usages but all related to the same research project or training program.

I III-15

I TABLE IV I

MONTHLY REACTOR ENERGY GENERATION (September,1985 - August,1986)

Hours at Monthly Totals KW-Hrs Full Power September, 1985 0.00 0.00 October, 1985 0.00 0.00 Novembe r, 1985 0.00' O.0d December, 1985 0.00 0.00 January, 1986 0.00 0.00 February, 1986 873.67 8.62 March, 1986 3306.76 32.90 April, 1986 3291.22 30.60 May, 1986 3897.30 38.63 ,

June , 1986 2134.79 21.25 July, 1986 2794.65 25.41 August, 1986 2989.36 29.07 YEARLY TOTAL 19,287.74 2 186.48 Note 1: Kw-Hrs yearly total for the 1985-1986 reporting year represents a 46%

I decrease over the previous reporting year while the hours at full power represent a similar decrease over the previous year. Although the 46% decrease is significant, this decrease is calculated relative to the third highest value ever recorded during the 1984-1985 report-I ing year. In actuality the energy generation was very good consider-ing the nearly 50% unavailability of the UFTR primarily due to the five plus-month outage to correct the problem of the sticking S-3 I control blade. In addition, the total run time for the facility was well above 50% of the 1984-1985 value for this reporting year indica-ting more low power usage and continued high utilization when avail-I able. The overall utilization continues to be high and with the good availability experienced since restart following return to normal operations in March,1986, a large increase in energy generation is expected for the next reporting year.

Note 2: The 19,257.74 Kw-Hrs of energy generation is the third highest one year total energy generation over the last eight years of UFTR op-I eration and, despite nearly 50% unavailability, represents the ninth highest one-year value in the 27-year history of the UFTP.

I III-16

v i

I l

TABLE V MONTHLY REACTOR USAGE / AVAILABILITY DATA (September,1985 - August, 1986) l Monthly Totals Key-On Time Exp. Time l Run Time Availability September, 1985 7.50 hrs. 40.93 hrs. 1.86 hrs. 8.3%

October, 1985 1.90 hrs. 69.00 hrs. 0.00 hrs. 0.0%

November, 1985 0.30 hrs. 57.50 hrs. 0.00 hrs. 0.0%

December, 1985 0.00 hrs. 58.62 hrs. 0.00 hrs. 0.0%

January, 19'86 1.50 hrs. 146.50 hrs. 0.00 hrs. 0.0%

February, 1986 53.90 hrs. 144.23 hrs. 44.00 hrs. 57.1%

March, 1986 68.80 hrs. 109.97 hrs. 62.75 hrs. 100.0%

April, 1986 69.60 hrs. 131.42 hrs. 61.33 hrs. 93.3%

May, 1986 67.50 hrs. 97.33 hrs. 59.92 hrs. 100.0%

June , 1986 85.92 hrs.

45.30 hrs. 38.45 hrs. 76.7%

July, 1986 117.60 hrs.

67.30 hrs. 60.68 hrs. 100.0%

August, 1986 62.50 hrs. 157.98 hrs. 58.17 hrs. 91.9%

TOTALS: 446.10 hrs. 1216.00 hrs. 387.16 hrs. 52.3%2 ,3

-l NOTE 1: Experiment Time is Run Time (Total Key-On Time minus Checkout Time) g plus set-up time for experiments, tours, or other reactor usage in-cluding checkouts, tests and maintenance involving reactor running or facility usage.

NOTE 2: Monthly Average availability is 52.3%; on the basis of days of the year, the availability is also 52.3% as indicated in Table VI. Al-though the yearly availability is the lowest in the last 12 years of UFTR opera tion , the large value of run time shows continued high uti-lization of the UFTR facility. Although the reactor was nominally a-I vallable in February, all time was devoted to surveillances and other checks with normal operation not resuming until March 3, 1986.

,I NOTE 3: Monthly average availability for the six full months following the extended outage to correct the sticking control blade problem has been 93.7% indicating the success of the various maintenance activi-ties completed during the extended outage, including preventive main-I tenance not directly related to correcting the problem of sticking control blades.

I III- _17 _

TABLE VI UPTR AVAILABILITY

SUMMARY

(September, 1985 - August, 1986)

I Month Availability Unavailable Days Primary Cause of Lost Availability I

september, 1985 8.3% 27.5 days Administrative Shutdown to De-termine Cause and Implement Corrective Action for the Sticking S-3 Control Blade I October, 1985 0.0% 31 days Maintenance and Repair of S-3 I and Other Control Blade Drive Sys tems I November, 1985 0.0% 30 days Maintenance and Repair of S-3 and Other Cvitrol Blade Drive Systems I December, 1985 0.0% 31 days Maintenance and Repair of S-3 I and Other Control Blade Drive Sys tems Janua ry, 1986 0.0% 31 days Maintenance and Repair of S-3 and Other Control Blade Drive Sys tems I February, 1986 57.1% 12 days Reload Fuel, Replace Biologi-I I cal Shield and Perform Com-plete System Checks and Sur-veillances Including Surveil-lances Overdue Because of the Extended Outage I March, 1986 100.0% 0 days Resumed Normal Operations on March 3, 1986.

April, 1986 93.3% 2 days Maintenance on the Wide Range Drawer Following a Failure Caused by an Electrical Tran-sient g may. 1eee 1ee.e% e eays --=-=- ----

I III-18

n__, -

I TABLE VI (Continued)

UFTR AVAILABILITY

SUMMARY

(September,1985 - August, 1986)

Days Primary Cause of Mon th Availability Unavailable Lost Availability June, 1986 76.7% 7 days Maintenance and Repair Work on the Dilutant Fan Tachometer coupling Leading to the RPM Gauge Indicator July, 1986 100.0% 0 days -------- -

August, 1986 83.9% 2.5 days General Maintenance TOTAL ANNUAL UNAVAILABILITY: 174 days TOTAL ANNUAL AVAILABILITY: 191 days = 52.3%

NOTE 1: This availability summary neglects all minor unavailabilities for g periods smaller than a half-day. In most cases these periods are for E less than an hour as some minor problem is corrected. This availa-bility summary also neglects unavailability for scheduled tests and surveillances.

NOTE 2: As indicated elsewhere, the maintenance and repair work on the Con-trol Blade Drives accounts for most (162.5 days or over 90%) of the I unavailability. Because of the prior unavailability due to mainten-ance and repair work on the Control Blade Drive Systems, a number of other repair projects were completed during the extended outage but none impacted on the critical pa th.

___ HHH-1@

TABLE VII-A UNSCHEDULED TRIPS

• Date Description of Cccurrence April 14, 1986 At 1216 hours0.0141 days <br />0.338 hours <br />0.00201 weeks <br />4.62688e-4 months <br /> during startup for the Operations Labo-ratory Course with Safety Blades all at 540 units and the Regulating Blade on the bottom, Acting Reactor

.I Manager P.M. Whaley was directing the student reactor startup when a trip occurred. All instrumentation was reading normally prior to the trip which was caused I by the momentary interruption of power to the build-ing. Lights were noted to flicker in the control room, and in the upstairs offices; a call to work Management Center confirmed there has been a large electrical power transient at the time of the trip.

Therefore, the trip was evaluated to be from a known cause - momentary interruption of power due to a power surge. All safety systems wre noted to respond properly to perform their intended safety function.

During the daily checkout prior to restart, the wide range drawer below the campbelling mode was found to be malfunctioning with the Safety / Log Calibrate I Switch introducing a too low test signal at points '1, 2 and 3. (See Maintenance Log Page #30). The source of the problem was a failed transistor caused by the ,

electrical power surge; af ter replacement, the system '

was restored to normal operation. Note that the safety function of the wide range drawer was not im-paired by this failure which was discovered during a checkout performed subsequent to the trip.

I

• All safety systems responded as intended for the trip listed in this Table.

l 1

I i

1 III-20

v 1 i

1 TABLE VII-B l SCHEDULED TRIPS i

Date Description of Occurrence There were no scheduled trips performed for training or experimental purposes during this reporting year. Part of the reason for this lack of scheduled trips was the failure to schedule eny utility operator training programs where l such trips are a designed part of the training program. '

1

'I

'I ,

t I

III

. .. _ . _ _ . _ - 2 1 ,

TABLE VIII LOG OF UNUSUAL OCCURRENCES During this reporting period there were no events which compromised the health and safety of the public. Several events, classified as unusual occurrences, are described below as they deviated from the normal' functioning of the facil-ity and are included here as the most important such deviations for the re-porting year. Unscheduled shutdowns are included here as well.

3 September 1985 - At 1030 hours0.0119 days <br />0.286 hours <br />0.0017 weeks <br />3.91915e-4 months <br /> reactor control blade (Safety-3) on the University of Florida Training Reactor failed to drop and fully insert on demand from a 64% removed positien in re-sponse to operator action. This failure (sticking about 31% removed) was discovered by a Reactor Operator as he commenced a power increase from the 1 watt critical posi-tion. The operator had accidentally raised the Safety-3 blade a few units instead of the negulating Blade for I this power increase; in returning it to the normal 640 unit position, he felt. the response was sluggish and so he attempted to drop' the blade from 640 units withdrawn to check it. Following clutch current release, the blade stopped at the 310 unit position and was cubsequently driven in with the other three blades to shut the reactor I down concluding a reactor run and constituting an un-scheduled shutdown. The Facility Director (SRO on call) was notified immediately and assisted the operator in characterizing the sticking position.

Immediate checks involving subsequent removal to various heights showed this sticking problem to be intermittant I and to center in the 290-315 unit range but with some ap-parent sluggishness in the drop from other higher and lower heights. It should be noted that this was essen-I tially a recurrence of the event reported by our facility in a letter dated January 28, 1985 and evaluted to be closed out in a report dated March 26, 1985. For this re-curring event the reactor was officially placed on ad-ministrative shutdown in a letter dated September 4, 1985, until cleared by the Reactor Safety Review Subcom-mi ttee (RSRS) and Nuclear Regulatory Commission (NRC).

I The Executive Committee of the RSRS reviewed the occur-rence on September 4, 1985, and concluded that it was a potential abnormal occurrence as defined in UFTR Techni-cal Specifications, Chapter 1. The RSRS in its Executive Committee meeting on September 4 instructed NRC notifica-tion as per Section 6.6.2 of the UPTR Tech Specs which was carried out on September 4. An interim 14-day report I was sent to Region II on September 17, 1985.

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w LOG OF UNUSUAL OCCURRENCES (CONTINUED) 3 September 1985 Essentially the NRC was kept advised throughout the work (continued) effort to restore UPTR operability. Numerous telephone calls as well as two I & E inspection visits occurred I during the total time from September 3,1985 when the sticking S-3 blade was discovered to March 3,1986 when a letter from the Director of the Facility permitted re-sumption of normal operations on March 3,1986 in concur-rence with RSRS and NRC permission. Essentially the prob-lem was determined to originate in the jack shaf t bush-ings of the control blade drive shaf t on S-3 with a simi-lar problem developing to a lesser extent in other blade jack shaf t bushings.

On November 8,1985, the cause of the sticking S-3 con-trol blade was isolated and determined to be caused by a failed west oriented bushing on the shaf t coupling of Safety-3 where it supports the blade in the core region.

The west oriented bushing had to be pried off the assem-bly while the remaining bushings slid easily off the as-sociated assemblies. This bushing was bound to the shaf t; I examination showed the bushing /shaf t contact surface to be rippled and rough indicating irregular wear on the bushing.

Replacement bushings and new stainless steel shaf ts were installed for all the control blades to prevent recur-I rence of this problem. The conversion of the blade shaft coupling assembly to stainless steel, installation of bearing collars on the control blade shaf t couplings and installation of pillow-block bolt head capture elements I were all reviewed as potential unreviewed safety ques-tions. In addition the shroud spacers on all control blades were drilled out to provide inspection ports.

On November 29, 1985, the Regulating Blade system to in-clude shaf ts, couplings, supports, blade and shrouds was  ;

I reassembled. Unfortunately the UFTR engineering drawing

(#89-31-116) of the shaf t coupling assemblies external to the biological shield was in error as the alignment pin did not engage the manufactured stainless steel (304 SS)

I shaf ts to allow the lowest position of the Regulating Blade to correspond to full insertion into the core.

I S taff and subsequent RSRS evaluation determined this oc-currence was not a promptly reportable occurrence. Paul Frederickson at NRC agreed when contacted on December 2, I 1985 for an update on UFTR s tatus. As a corrective action based on ALARA considera tions, a new ex-core shaf t coup-ling arrangement was reviewed and approved as a modifica- I tion to the facility with a followup submitted to NRC with a letter dated December 10, 1985.

I III-23

v LOG OF UNUSUAL OCCURRENCF.S (CONTINUED)

_ 3 September 1985 The new bushings along with shrouds and blade shaf ts were (continued) then installed on all control blade drive systems. The fuel boxes were completely reinstalled tmd the system de-termined to be leaktight by 23 January %36.

Af ter installing modified aluminum spacer plates below the core center island to replace embrittled material, restacking of the center island graphite as well' as upper tiers of graphite around the fuel boxes was completed by 28 January 1986. All operators were then recertified to load fuel following an NRC-approved plan (see Appendix E).

Complete fuel loading of the UFTR including adding one I half (5-plate) fresh fuel bundle to the SW fuel box was accomplished in three days (5-7 February) using a special procedure previously supplied to NRC. Af ter reloading, the RSRS reviewed the UFTR status and granted permission to restart on February 12, 1986. The next two weeks through February 25, 1986 were spent in low power physics te s ting.

I On February 25, all operators were recertified to conduct power testing which was begun on February 25 and com-I pleted on February 28, 1986. All tests conducted during the month were successful as delineated in the workplan supplied with the letter to Julian at NRC dated 6 January I 1986 (See Appendix E) and the NRC was notified on Feb-ruary 28 that system recovery was complete and the reac-tor ready to resume normal opreations. The only outstand-ing items involved final recertification of all opera-I tors. As per the C.A. Julian letter dated 6 January, all operators conducted a startup into the power range under observation of either the Facility Director or Reactor Manager and were then fully recertified.

Prior to fuel loading and then af ter fuel loading was completed on February 7,1986, all control blade drop times were checked with all graphite and one shield block in place on February 10. Af ter completion of restacking on February 12, all control blade drop times were again checked on February 13 with all times remaining below 0.5 i seconds in agreement with original system specifications l demonstrating finally that the sticking S-3 problem was '

I resolved and the corrective action effective. Subsequent checks have confirmed the effectiveness of the corrective action implemented to restore proper control blade drive i

I system ope ra tions.

I I III-24 I

v LOG OF UNUSUAL OCCURRENCES (CONTINUED) 3 September 1985 The UFTR was then approved to resume normal operations on (con tinued) March 3, 1986. The final report outlining the problem of the sticking control blade as well as the corrective ac-

-l5 tion taken to prevent recurrence was submitted to NRC with a letter of transmittal dated 3 April 1986. This en-tire report to NRC is contained in Appendix A.

19 February 1986 - At 0900 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br />, following a successful pre-orerational checkout, Reactor Operator G. Fogle commenced a reactor startup. However, at 0916 he performed an unscheduled I- shutdown from suberitical at less than one watt due to erratic operation of the blade position indicator on the regulating blade. Console indications showed the blade to I be responding prop'erly with improper indication only.

Following shutdown, the third digit nixie tube on the regulating blade was found to be loose in its socket. The I tube was reseated in its socket under MLP #15 to restore normal operation. There were no safety-related or radio-logical consequences associated with this occurrence.

20 February 1986 - At 1108 hours0.0128 days <br />0.308 hours <br />0.00183 weeks <br />4.21594e-4 months <br />, Reactor Operator G. Fogle dropped Safety-Two Control Blade from 400 units withdrawn and all other blades banked at 100 watts as part of the effort to gen-I erate new reactivity worth curves for all blades follow-ing core reassembly and fuel reloading. At 1110 hours0.0128 days <br />0.308 hours <br />0.00184 weeks <br />4.22355e-4 months <br /> as recovery to 100 watts with S-2 at 300 was underway, er~

I ra tic operation of the blade position indicator on the regulating blade caused the operator to perform an un-scheduled shutdown though there was no other problem with system operations. Following shutdown, inves tigation showed the nixie tubes to be at fault. The card for the hundredth's position nixie tube on the position indicator for the regulating blade was replaced af ter discovery of I the burned out nixie to restore proper operation of the blade position indicator for the regulating blade under MLP #17. There were no safety-related or radiological I consequences associated with this occurrence. The blade itself continued to respond properly with only the indi-ca tion a t faul t.

27 February 1986 - At 1102 hours0.0128 days <br />0.306 hours <br />0.00182 weeks <br />4.19311e-4 months <br /> following a successful pre-operational checkout, Acting Reactor Manager, P.M. Whaley, commenced a reactor startup. At 1118 while still suberitical, M r.

'l Whaley performed an unscheduled shutdown due to malfunc-E tion of the linear (red) pen of the two-pen recorder.

Upon shutdown, the cause of the fault was determined to g be apparently due to an intermittant fault caused by poor 3 contacts in the driving circuit. Af ter checkout, normal operation was resumed with no further problems noted.

There were no safety-related or radiological consequences I associa ted wi th this occurrence.

_ III-25

u LOG OF UNUSUAL OCCURRENCES (CONTINUED) 2 April 1986 - In attempting to separate a sample of irradiated man-ganese for shipping to Dr. Steve Grossman at the Univer-sity of South Florida at Tampa for use in a half-life ex-I periment in a Nuclear Chemistry Laboratory Course, some contamination was spread on the top reactor deck. Some clothes which were contaminated with the irradiated man-ganese powder (2.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> half-life) were held for decay

'I overnight. The area was decontaminated and all was re-stored to normal by April 3,1986 whereupon the clothes were returned. There were no safety-related consequences I associated this event. The contamination levels were low enough that there was no impact on the health and safety of the personnel involved.

7 April 1986 - Af ter two new rabbit capsules were manufactured and ac-cepted for usage, one was inserted empty for a test; sub-sequently, the other capsule was inserted containing a hair sample and was not able to be returned af ter a short irradiation. At the failure to re turn , the reactor was promptly shut down to limit sample fluence and radioacti-I vity buildup. Efforts to retrieve the capsule at shutdown conditions were unsuccessful so the rabbit system was re-moved from the UFTR on April 8, 1986. With radiological controls in place, increased capsule driving pressure was used on April 14, 1986 to remove the capsule in pieces whereupon the second capsule was inserted and also became I stuck. A t this point, the rabbit internal assembly was evaluated to be damaged and to be the cause of the prob-lem; because of difficulty of repairs, the system was re-moved from service with no repairs planned. Subsequently, I a new rabbit system was designed, manufactured and ap-proved for implementation.. Full implementation is scheduled to occur in Sept:mber, 1986. Because the UPTR was shut down upon failure of the capsule to return and the cystem allowed to decay prior to determining the cause of the failure and removing the capsule, there were I no radiological effects of this malfunction. Essentially there were no safety-related or radiological consequences associated with this occurrence.

14 April 1986 -A trip due to an electrical power transient occurred at 1216 hours0.0141 days <br />0.338 hours <br />0.00201 weeks <br />4.62688e-4 months <br />. All safety systems functioned as required to perform their intended safety function. See Table VII-A for details.

I III-26

ur-ll l LOG OF UNUSUAL OCCURRENCES (CONTINUED) 4 August 1986 - Af ter a startup to 1 watt was completed by a student as part of the operations laboratory cour'se, the two-pen re-corder was noted by SRO P.M. Whaley not to be operating

-I properly in that the paper drive was not functioning.

Since this function is required by UFTR Tech Specs as a limiting condition for operation, the UFTR was promptly shut down. Af ter comple ting the unscheduled shutdown, MLP

1. 43 was opened as the the press-fitted gear on the re-corder was noted to have worked loose and to be free wheeling. Af ter the gear was peened into place under MLP

1. 43, the recorder was checked out and found to be func-tioning (recording) properly. The reactor was evaluated as prepared for return to normal operations with no fur-ther problems noted. There were no safety-related or ra-

.I dialogical consequences associated with this occurrence.

I 21 August 1986 - SRO H. Gogun responded to an af ter-hours call and found the pit alarm was activated indicating water in the pit.

Since the primary coolant pump was still running from the previous day's operation, initial evaluation indicated no rupture disk breakage, though a small leak in the primary coolant line, demineralizer line or the secondary coolant line were all considered possible to. include a small leak I from a fuel box or coolant line through the trench to the pit. The SOLU Bridge showed normal primary water resisti-vity which also indicated the loop was not pumped dry.

I Since any break would be to a closed sump, all pumps were secured to await evaluation during the regular work day.

Af ter the Radiation Control Officer was notified of the I leak, the pit was uncovered and a small leak was located at the connections on the demineralizer with less than two cups of primary coolant in the pit sump. The leak was corrected under MLP #46 by reseating the demineralizer I connec tions. Following evaluation showing no contamina-tion problems and assuring the leak was fixed, the de-mineralizer pump was restarted to restore proper coolant I resistivity levels. The reactor was returned to normal operational status when a check on August 22, 1986 showed no leaking at the domineralizer connections af ter over-night operation of the domineralizer system. There were no safety-related consequences associated with this oc-currence; the radiologick. consequences were negligible.

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m LOG OF UNUSUAL OCCURRENCES (CONTINUED) 22 August 1986 - Af ter about six hours of full power operation on this date, the stack monitor recorder needle was noted not to be functioning; as a result, the recording function of the stack monitor was not operable. Since the stack re-cording function is a UFTR Technical Specification limit-ing condition for operation, the reactor was promptly shut down. Following this unscheduled shutdown, the failed stack monitor recorder was removed and a spare meter movement with meter from a spare unit was installed under MLP #47 to restore normal stack monitor function-ing. Following reconnection of the stack monitor recorder and verification of proper respenco, the UPTR was eval-uated to be restored to normal operation with no further problems noted. There were no safety-related or radiolo-gical consequences associated with this occurrence. The stack radiation monitor was indicating throughout this occurrence.

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1 1 " IV. -MODIFICATIONS TO THE OPERATING CHARACTERISTICS

- OR CAPABILITIES OF THE UPTR A number of modifications wre made to the operating characteristics or capabilities of the UFTR facility during the reporting period. These modifica-tions were all subjected to 10 CFR 50.59 evaluations, and then determinations, as necessary, to assure no unreviewed safety questions were involved.

Carried over from 1984-1985 Reporting Year:

Modification 6: Replacement of Vent System Inclined Manometers.

Modification 7: Addition of Flow Sensors (Rotameters) to Secondary and City Water Systems.

I.

1. Annex Basement Lab Installation (Permanent - Closed Item)

(1985 Modification Number 10; Evaluation Completed September 9,1985)

A section of the basement of the Annex north of the reactor building was segregated to provide a separate laboratory, space for materials testing equipment. An evaluation was performed based on the changes involved in the building abutting the north reactor cell wall. The partition is com-pleted and this item was considered closed out upon installation of a fire detection system sensor in the area adjacent to the north cell wall (1985 Modification Number 19).

A 2. Vent Fan Motor Replacement / Blower Replacement (Permanent - Closed Item)

(1985 Modification Number 11; Evaluation Completed September 9,1985)

This modification permits the use of a squirrel cage blower and a lower RPM single phase motor, as necessary, in the core vent system. The lower RPM motor has r;ot been installed at this time, but the squirrel cage blower was placed in the core vent system to provide the Tech Spec requi-site core vent flow.

Contolling Document: Maintenance Log Page 47.

3. Temporary Change to Vent Damper Operator to Document Reopening (Temporary - Item Closed)

(1985 Modification Number 12; Evaluation Completed September 6, 1985)

This modification pennitted the use of a clamped block to limit partial l-B

- re-opening of the core vent damper that occurred due to mechanical fail-ure of the shaf t/ spring assembly until a replacement damper control unit p could be obtained. The clamped block was a temporary measure which per-formed its function well until replaced with a new shaf t/ spring assembly.

Controlling Document: Maintainence Log Page 42.

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4. Temporary Replacement of Annex Siren (Temporary - Item Closed)

(1985 Modification Number 13; Evaluation Cornpleted September 6,1985)

This modification evaluated the use of a solid state siren to be used in place of the motor driven device unitl a new identical replacement could ,

be obtained for a failed (burned out) siren which occurred durng a test I run for a quarterly evacuation drill. The replacement siren performed the same alarm function' to assure emergency evacuation until a permanent re-placement siren was installed.

Controlling Document: Maintenance Log Page 43.

5. Diluting Fan Fuse and Belt Replacement (Permanent - Closed Item)

(1f45 - Modification Number 15; Evaluation Completed October 17, 1985)

I The diluting fan, during a rapid sequence of operations involving ener-gizing and de-energizing the fan motor, blew fuses repeatedly. Mainten-ance checks indicated normal run current, rpm and nomal starting cur-I rent. The only sources of the problem noted included s. light diluting fan belt slippage and installed instantaneous-blow fuses (where time delay or

" slow blow" fuses are normally used). Conference with maintenance and UFTR staff resulted in a decision to install v-notch belts and slow-blow fuses (more appropriate to induction motor starting characteristics).

Both modifications were installed to provide reliable operation of the diluting fan with an evaluation indicating no unreviewed safety questions were involved.

Controlling Document: Maintenance Log Page 57.

6. Sticking S-3 Work Procedure for Mechanical Manual Blade Manipulation (Special Proccdure - Closed Item)

(1985 - Modification Number 16; Determination Completed September 23, 1985)

I Trouble-shooting the sticking control blade (S-3) proved to be impossible without undertaking a massive program of investigation and repair; this procedure (used in work controlled by Maintenance Log Page 58) provided I the first step in that program by allowing examination of movement-inhi- l biting characteristics of the S-3 control blade in an effort to isolate i the problem to in-core or ex-core origin. Basically, this modification allowed manual mechanical movement of the S-3 control blade with audio-sensors in place to monitor sources of the sticking problem.

Controlling Document: Maintenance Log Page 58.

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7. UPTR Shroud Spacer Cutting Alteration for Ease of Blade Inspection (Permanent - Closed Item)

(1985 - Modification Number 17; Determination Completed October 4,1985)

A method was devised to visually determine if adequate clearance existed between the control blocks and the control blade shrouds, consisting of putting a viewport about 7 inches long in the trailing edge (away from the fuel) of the spacer blocks holding the two shroud plates in place as l

I indicated in Figure IV-1 (Proposed Shroud Viewport). Concern over poten-tial problems due to the open viewport in the shroud allowing foreign matter into the control blade area was briefly expressed and dismissed as

E reflector graphite covers the opening and concrete covers the graphite.

'E Concern for the potential for initiating a magnesium fire caused pro-cedures to be generated to make the cut under tight fire controls, with retention elements for magnesium chips. (A lack of documentation of the first concern lead to NRC concern that the safety functions of the shroud i

spacers had not been adequately expressed, reviewed and documented; this concern resulted in a revision to UFTR SOP-O.4 involving a system of do-lI cumeating references and specific topics of concern used in providing an evaluation or determination that no unresolved safety question is l breached by a modification.) This particular modification was review 1.d by the RSRS again in February in response to NRC Inspection Report 50-83/86-01 prior to resuming normal operations, confirming that the installed in-spection ports do not constitute an unresolved safety question. The in-spection ports proved a considerable aid to reducing dose by expediting troubleshooting and control blade alignment during core reassembly.

Controlling Documents: Maintenance Log Page #61 - Safety-3 Shroud, Radiation Work Permit (RWP) RWP-85-16-I Modify Shrouds In-Core, and Maintenance Log Page #64 - (Out-of-Core)

Sa fe ty-1, Safety-2, Regulating Blade.

8. Spent Fuel Pit Irradiated Fuel Plate Consolidation Operation (Temporary - Closed Item)

(1985 - Modification Number 18; Evaluation Completed October 17, 1985)

Defueling the reactor was required in the course of making repairs of the S-3 control blade; it was determined that the available spent fuel pit storage space was not adequate under the existing storage configuration

'I (partial bundles in plates in 10 pits) to accommodate the core inventory.

Since the spent fuel storage had never been optimized, it was determined to consolidate partial assemblies, where possible. A procedure was gen-erated to constitute a spent fuel pit loading plan for that operation.

The consolidation was successf* lly carried out with minimal fuel movement under this procedure which was evaluated not to involve any unreviewed safety questions.

Controlling Document: RWP 85-12-I .

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9. UFTR Fire Alarm System Modification - Rate of Rise Alarm in Reactor Support Facility (UFTR Basement Annex Adjacent to Cell Wall)

(Permanent - Closed Item) '

(1985 - Modification Number 19; Evaluation Completed October 17, 1985)

I This modification provided for installation of an additional fire detec-tion system sensor in the basement annex area adjacent to the north reac-tor cell wall to improve the system installed under modification number I 10 of 1985. Specifically, a rate of rise alarm was determined to be the best detector for the reactor support facility located in this area.

Evaluation indicated no unreviewed safety question was involved.

10. Conversion of Blade Shaf t coupling Assembly to Stainless steel (Permanent - Closed Item)

(1985 - Modification Number 20; Evaluation Completed November 21, 1985)

I During the maintenance work on the control blade drive system, the S-3 blade to shaf t coupling (jack shaf t) was discovered to have a rough sur-face at the shaf t/ graphite bushing interface. The surface was pitted and rusted, which prevented +he control blade from operating smoothly. To prevent future problems from that source, it was determined to manufac-ture new jack shaf ts from stainless steel. A comparison was made of the properties of the new stainless steel and the old carbon steel shaf ts based on the properties of similar carbon steels (material properties of SAE 1040 are bracketed by SAE 1035 and SAE 1045) and the properties of both potential stainless steels, SS 304 and SS 316 (indicated by the En-I gineering Machine Shop as the commercially available candidates available for use without special order). These items were verified and reviewed by the RSRS which agreed with the original evaluation that there is no unre-

=

viewed safety question.

11. Installation of Bearing Collars on Control Blade Shaft coupling (Design / Cancelled - Closed Item) k (1985 - Modification Number 21 - Design Modification; Determination Completed November 21, 1905)

Consideration and evaluations were made for the need and design for bear-I ing collars on the control blade shaf ts to limit axial movement. This modification was determined not to involve an unreviewed safety question, but in technical evaluation was determined to be unnecessary, when the point for the installation was reached.

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12. Installation of Pillow Block Bolt Head Capture Elements l

(Design / Superseded - Closed Item)

(1985 - Modification Number 22 - Design Modification; Determination Completed November 21, 1985)

During 'the control blade jack shaf t bushing removal process, a design flaw was noted in that there was no capability for grasping the free end of the retaining nut / bolt assembly under the pillow block. This fault ne-I cessitated personnel spending a large amount of time and' resultant dose removing the bolts that anchored the pillow blocks. This modification was the first attempt at designing an effective mechanism for allowing quick and/or remote bolting and unbolting of the pillow blocks.

Controlling Documents: RWP-85-21-I and RWP-85-23-I (Regulating Blade Safety 3 Blade),

RWP-86-2-I (Safety 1 and Safety 2).

13. UFTR Control Blade Drive Unit /Long Shaf t Coupling Modification (Permanent - Closed Item)

(1985 - Modification Number 23; Determination Completed December 3,1985)

Alignment problems while assembling and restoring the regulating blade control blade drive system to normal operations uncovered evidence that I the original carbon steel jack shaf ts had not been constructed per en-gineering diagrams. To permit the use of the stainless steel jack shaf ts designed to reduce the likelihood of future drive system failures a com-I mercial web and hub shaf t coupling system indicated in Figure IV-2 (Pro-posed shaf t coupling Method) capable of being orierited to various shaf t positions as opposed to the limitations of the ongonal design was deter-mined to be an adequate choice to replace the previous coupling which had I' no potential for adjustment. The original design is shown in Figure IV-3.

Controlling Document: Maintenance Log Page #68.

14. Security System Magnetic Reed Switch Replacement Modification (Pemanont - Closed Item)

(1985 - Modification Number 24; Evaluation Completed December 5,1985)

This modification is considered safeguards information and will not be described in this document.

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15. Bolt Penetration Enlargement on SE Fuel Box Primary Piping Top Flange -

(Design / Cancelled - Closed Item)

(1985 - Modification 86-1; Evaluation Completed January 13, 1986)

During replacement of the South fuel boxes af ter correcting the problem of the sticking control blades, the flange to box alignment was hampered by small clearance for the bolts in the flange penetrations. The best so-lution from an ALARA standpoint was determined to be enlarging the flange I penetrations by about 1/8 inch (which would remain within drawing speci-fica tions) for the South East fuel box. However, subsequent work resulted in sufficient space for alignment to reassemble the coolant flow path, so this modification was cancelled.

16. Jack Shaf t Duching Ifold-Down Retainer Nuts: Final Designs (Permanent - Closed Item).

(Modification 86-2; Determination Completed January 16, 1986)

Two final hold-down bolt retainer designs were made to provide capability for fastening and unfastening the jack shaf t pillow block bushings ex-I peditously and/or remotely (for ALARA considerations). The first retainer was manufactured and placed in S-3 and the regulating blade. Based on the results of time involved to insert the retainers in S-3 and the regulat-ing blade, a second design was manufactured and used on S-1 and S-2. The I two designs were determined not to involve any unreviewed safety question involved.

Controlling Documents: RWP 86-4-I , RWP 85-23-I and RWP 86-2-I.

17. Graphite Stacking (Permanent - Closed Item)

(Modification 86-3; Evaluation Completed January 16, 1986)

During replacement of graphite around the control blades, it was dis-I covered pieces specified by the stacking diagrams did not exist. Examina-tion of photographs taken during the unstacking process showed the pre-vious stack was not cons!atent with the stacking diagram. In order to al-I low stacking to continue, the construction of the graphite reflector with respect to requirements and capability for pile geometry variations was reviewed and the evaluation result indicated that no unresolved safety I question was involved as .long as a regular graphite reflector with no gaps or air spaces was constructed.

Controlling Docuemnt: RWP 86-5-I.

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18. South Beam Port Plug Support Device (Permanent - Closed Item)

(Modification 86-4; Evaluation Completed January 16, 1986)

No diagram could be located detailing the construction of the aluminum guide-chute in the South Beam Port. Consequently, a diagram of the as-built construction was made and reviewed to assure that no unreviewed safety question existed for the guide-chute as installed.

19. UFTR Center Island Graphite Bed Plate (Permanent - Closed Item)

(Modification 86-5; Evaluation Completed February 5,1986)

When the UFTR core center island graphite was removed, a 1/4" thick piece I' of phenolic, compressed or hardened fiberboard or similar material was found below the center island acting apparently as a means of leveling and evening out the surfaces of the aluminum beams on which the center island graphite rests. This piece of material was found to be fractured into thousands of pieces. Since this piece of material was not shown on any of the core drawings and no literature reference could be located re-I ferencing the material, the decision was originally made to eliminate it on restacking the center island graphite. As a result, the 1/4" drop to the aluminum beam sufaces made it impossible to align the vertical pieces of graphite or to complete insertion of the center island.

Since the phenolic-like material supplied no structural support but only served to provide an even spacing for inserting graphite, it was decided I to replace the unknown material with three identical aluminum plates.

Since this material only sarves to provide an even and is obviously needed to provide proper graphite alignment, the modification to aluminum I does not represent an unreviewed safety question, especially since the aluminum is the same material as that in the structural support beams themselves. The overall graphite dimensions are maintained in a regular I geometry as supported by the 50.59 Evaluation for UFTR Modification Num-ber 86-3.

Controlling Document: RWP 86-5-I.

20. Source Holder for the Antimony Beryllium (Experiment - Closed Item)

(Modification 86-6; Evaluation Completed February 14,. 1986)

The small Christmas tree-like holder used in lif ting the Antimony-Beryl-I lium Source from the UFTR broke off while it was out of the reactor for the extended maintenance work. Swipe checks including spectroscopic ana-lysis of the swipes has demonstrated that the source is not leaking (as expected since the thin tree-like lif ting assembly is not part of the I source encapsulation). Therefore, no unreviewed safety question is in-volved in continuing to use this source. A small aluminum holder was de-signed to hold and to transfer the source. No unreviewed safety question I is involved in usig this new holder sinco it is aluminum like most of the core structure and has little mass, low long term activation prob-abilities and insignificant reactivity effects (especially compared to 1

I the sbBe source).

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21. Changes' in Order and Location of Fuel Bundle Loading from Listing in Special Procedure (Special Procedure - Closed Item)

(Modification 86-7; Evaluation Completed February 12, 1986) '

The order of loading fuel bundles and the respective locations specified in the special fuel loading procedure reviewed and approved by the RSRS were intended only to control the loading so that core contents would be properly tracked. Some higher burned bundles were transferred to lower flux areas but the only governing control in the Special Procedure was to assure that adequate records of the loading would be maintained.

First, an error in the locations of two bundles was noted in Table 2 in I disagreement with Table 1. Since Table 1 was the original control, the correction of Table 2 to agree with Tablo 1 was implemented with concur-rence of the Reactor Manager and Facility Director.

I second, when bundle UF-32 was discovered to have excess length on its bolts to prevent its addition to the planned SE location in the NE fuel box, the decision was made to return this bundle to the spent fuel pit I until all other irradiated fuel had been loaded. To allow the loading to proceed with a minimum of unnecessary delays, fuel bundle UF-14 was switched with UF-32 in the Table 2 order controlling the loading. Again, E this decision was made to allow the loading to proceed as far as possible g before a long hold would occur. As a result, it was planned to correct the problem of the protruding bolts on UF-32 so it would be able to be added in a different location at a different loading increment, In the first case a typographical error was corrected. In the second case one bundle addition (19th fuel bundle to be added) was switched with another (21st fuel bundle) to be added. The intent of the procedure was maintained in both cases so neither change was considered to involve an unreviewed safety question.

22. Change in Partial Bundle UF-40 Constituents (Special Procedure)

(Modification 86' 8; Evaluation Completed February 12, 1986)

I During inspections prior to assembly of one partial fuel bundle, one new fuel plate specified in the controlling procedure was discovered to be bent near the bolt hole. This modification specified the use of a dif-forent fuel plate in test partial bundle which was evaluated rlot to in-volve any unreviewed safety question.

I I

I IV-8

v 1

23. Bolt Length Reduction in UF-32 Fuel Bundle (Permanent - Closed Item)

(Modification 86-9; Evaluation Completed February 12, 1986)

Clearance problems were noted in attempting to insert bundle UF-32, at-tributed to excessive length of the bolts holding the bundle together. An.

evaluation was made that removing excess length of the bolt did not in-volve any unreviewed safety questions. Af ter bolt length reduction, the assembly was inserted into the core with no further clearance problems.

Controlling Documents: Maintenance Log Page 11, RWP 86-8-I.

24. APD Motor Replacement and Striker Bar Interrupter Arm

.I (Permanent - Closed Item)

(Modification 86-10; Evaluation Completed February 27, 1986)

The recorder striker on the Air Particulate Detector Recorder was dis-I covered to be striking intermittently during the daily checkout. The problem was found to be degraded operation of the plastic operating gear which was apparently caused by worn teeth on the internal gear of the drive unit of the motor assembly. It should be noted that even with the striker arm inoperable, the APD still indicates and is capable of alarm-ing and records intermittantly; therefore, its safety function was main-I tained.

The motor removed was the original APD motor; the replacement moter was selected from the reactor use only cabinet as a standby spare already in I stock. The two motors have the same electrical characteristics but dif-ferent manufacturers. In attempting to mount the new motor, it was found that the striker bar interrupter arm used to deactivate the striker for I striker protection when changing charts would have to be installed on the hub of the new motor shaf t under a plastic gear. Since it was not desira-ble to remove the gear to mount the arm (possibility of damage to the motor as well as introducing dirt into the sealed cavity), a small sl'ot I was cut in the arm on the non-bearing side of the pivot hole. The slot was then opened and the arm slipped into the hub. The slot was then closed to assure it does not slip open during use. The motor was then installed; the APD was then started and checked for proper operation.

Standard tests of motor operation to assure proper motor installation and APD operation were conducted as per daily checkout requirements of SOP-I A.1 to assure continued ability to fulfill its safety function with eval-uation indicating no unreviewed safety questions were involved.

Controlling Document: Maintenance Log Page 20.

I I

I IV-9

25. Removal of 3/32" From Graphite CVP Tube Holder Stringer l 1

(Permanent - Closed Item)  !

(Modification 86-11; Evaluation Completed April 17, 1986)

In attempting to place the graphite multiple vial sample holder (1985  !

Modification Number 9) into its core center location, it was discoveled I

that the CVP graphite is about 3/32" undersized from the nominal 4" x 4". '

This modification provided for trimming the vial holder accordingly with an evaluation that no unreviewed safety question was involved. I

26. Replacement of Dilute Fan Tachometer (Permanent - Closed Item)

(Modification 86-13; Evaluation Completed June 26, 1986)

During evaluation of dilute fan rpm indicating circuit failure, no infor-I mation could be located regarding the operation or specifications of the tachometer. A specific tachometer was obtained (on a contingency basis) with known specifications as an approved replacement should it be needed.

Controlling Document: Maintenance Log Page 59.

27. Design, Installation, Testing of New Pneumatic Delivery Rabbit System (Permanent - Closed Item)

(Modification 86-14; Determination Completed July 18, 1986)

This modification is to replace the previous failed rabbit system with a system that can be easily dismantled to recover or replace failed com-ponents without cutting, grinding, or welding on the rabbit system de-livery system. This modification should signficantly improve ALARA impact and system availability in the event of capsule or system failures. The rabbit system is designed to operate safely and provide no impact on the health and safety of the general public.

The second part of the modification allows rabbit system operation only with positive action by the reactor operator at the controls even when the system is connected. This control is accomplished through a solenoid-

-I operated valve controlling the regulated nitrogen-gas supply. The sole-noid is controlled by a switch in the UFTR Control Room.

I The third part of this modification is to assure the rabbit system is secured to the reactor shield. This is accomplished by bolting the incore assembly to the rabbit shield plug and using a strap, bolting the shield plug to the biological shielding. The actual insertion is controlled as an experiment run. The new system will be available for normal operations early in the new reporting year.

I Controlling Documents: RWP 86-12-I, RWP 86-14-I, Modification Package 86-14, Run Request 86-39.

I IV-10

I

28. Dilutant Fan Tachometer Drive Coupling (Permanent - Closed Item)

(Modification 86-15; Evaluation Completed August 21, 1986)

This modification was the first of a series of changes implemented to provide reliable coupling for the dilutant fan RPM . indicator. This change-shif ted the coupling from an altered rubber. stopper to a machined alumi-num piece with a set screw locking the coupling device to the tachometer shaf t. This coupling change affects the indicating system only, not the diluting fan itself; the change was determined not to involve any unre-viewed safety questions.

Controlling Docuemnt: Maintenance Log Page 35.

.g u 29. Conversion of Dilutant Fan Tachometer Drive Coupling From Aluminum to 316 Stainless Steel (Permanent - Closed Item)

(Modification 86-16; Evaluation Completed August 21, 1986)

I The aluminum coupling installed through Modification 86-15 was found to be wearing too fast and wan subsequently replaced by a 316 stainless I steel coupling to assure longer coupling life. Again the change was eval-uated not to involve any unreviewed safety questions.

Centrolling Document: Maintenance Log Page 41.

I I IV-11

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3 ORIGINAL UFTR SHAFT COUPLING NETHOD l (DRIVE UNIT TO LONG SHAFT OF CONTROL BLADE ASSEMBLY)

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V. SIGNIFICANT MAINTENANCE, TESTS AND SURVEILLANCES OF UPTR REACTOR SYSTEMS AND FACILITIES Records for the 1984-1985 reporting year show extensive corrective and preventive maintenance was performed on all four control blade drive systems

' external to the biological shield. Maintenance work during the 1985-1986 re-I porting year has been 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 I finally and correct the cause of the sticking blade. This work is summarized in detail in Appendix A. This work was the predominant activity for the re-porting year. Although considerable maintenance checks were performed on the

,I external parts of the blada drive systems; nevertheless, the really extensive maintenance was performed on the control blade drive systems internal to the biological shield to include replacement of all jack shaf t graphite bushings and all four jack shaf ts and couplings as well as replacement of the right I angle coupling mechanisms external to the biological shield on all blade drives.

I Other signficant maintenance efforts were devoted to the diluting fan RPM indicating system and to designing, manufacturing and initial testing of a re-designed rapid sample pneumatic delivery (rabbit) sys tem.

In the table that follows all signficant maintenance, tests and surveil-lances of UFTR reactor systems and facilities are tabulated and briefly de-scribed in chronological order; this tabulation also includes administrative checks. Surveillance tests or other checks / maintenance required by the Techni-cal Specifications, NRC commitments or other administrative controls are de-signated with a prefix letter and a number; otherwise the items listed are I considered maintenance, though in the area of the maintenance on control blade drive systems, this work actually constitutes part of the V-1 Blade System Mechanical Checks required to be performed for the entire reactor control sys-tem every five years as specified in the L'TR Technical Specifications sur-I veillance Requirements, Section 4.2.2, paragraph 4.

It should be noted that many required surveillances were delayed during the extended outage for work on the control blade drive systems. All surveil-lances were performed, however, prior to declaration of the facility as ready for normal operations on March 3,1986 and at the lowest operable power level

,I allowed to perform each surveillance. The overall program outlining this work is detailed in Table II attached to the letter to NRC in Appendix B which con-tains the overall summary report to NRC on the recurrence of the sticking S-3

,I control blade.

Date Description I 3 September 1985 Verified withdrawal of S-2 control blade upon demand  !

following drop from 640 units withdrawn as part of weekly checkout as a temporary RSRS requirement. )

4 September 1985 Replaced S-2 control blade position indicator nixie I I tube element using on-hand spare (4 Sep 85, MLP #52). l I V-1 1

I Da te Description l 4 September 1985 Replaced shield tank recirculation pump bearings and reinstalled pump (7 Sep 85, MLP #53).

4 September 1985 Performed maintenance checks consisting of getting the S-3 blade stuck and checking gear trains, clutch housing, right angle gear box and the gear box shim alignment to try to isolate cause of sticking S-3 blade (5 Sep 85, MLP #54).

4 September 1985 Performed maintenance to clean and overhaul S-3 con-trol blade drive system external to the biological shield - no degradation noted (5 Sep 85, MLP #55).

9 September 1985 Performed S-3 maintenance consisting of uncoupling the blade shaf t from the drive unit, checking the I saddle bearing external to the biological shielding, checking the blade drive shaf t for wear, cleaning the drive shaf t and reassembling the drive shaf t system;

'E unable to iso ate cause except to verify the source E must be within the biological shielding (13 Sep 85, MLP #58).

10 September 1985 Tightened primary coolant loop flow return switch to storage tank to correct a small leak which had ac-tuated the pit sump alarm (10 Sep 85, MLP #56).

I 10 September 1985 Replaced stack diluting fan fuses blown as result of rapid actuation and securing of the vent system with I time delay fuses; also replaced V-belts on diluting fan with notched belts (13 Sep 85, MLP #57).

I 25 September 1985 Q Quarterly Calibration Check of Area Radiation Monitors (Area Monitors only performed one month ear-ly in anticipation of removal of biological shielding i on September 25 which would make accurate calibration of the area monitors impossible when scheduled in Oc tober. )

3 october 1985 Performed visual and audio inspections of S-3 control blade during manual withdrawal with reactor secured to determine source of sticking problem - results in-conclusive (7 Oct 85, MLP #59).

4 October 1985 Q Quarterly Calibration Check of Stack Radiation Monitor. (Area Monitors performed one month early in September to avoid conflict with increased cell ra-diation levels caused by removal of biological shielding to facilitate S-3 inspection and repair work.)

8 October 1985 S Semiannual Inventory of Special Nuclear I Ma te rial .

V-2

~

I Date Descrip tion 8 October 1985 S Semiannual Inventory of Security-Related Keys for UFSA.

9 October 1985 Conducted DOE-ORAU Sponsored Task Analysis of the UFTR Facility utilizing DOE Consultant Dr. John Randall.

10 October 1985 Accomplished consolidation of all irradiated partial fuel bundles within the irradiated fuel storage pits per 10 CFR 50.59 Evaluation No. 85-18.

11 October 1985 Unloaded the entire UFTR core to the irradiated fuel

.I s torage pi ts.

14 October 1985 Repaired the striker pointer on the East Area Radia-tion Monitor recorder and reset to zero (14 Oct 85, MLP #60).

15' October 1985 S Check of Sb-Be and Pu-Be Sources for Leakage.

16 October 1985 S Semiannual Inventory of Security-Related Keys for UFTR.

16 October 1985 A Annual Check and Update of Emergency Call Lis ts.

16 October 1985 Q Quarterly Radiological Emergency Evacuation Drill.

25 October 1985 Modified the top spacer material on the shroud for the S-3 control blade by cutting a short inspection

-l port out of the top of the shroud to allow visibla E inspection of blade movement over the full travel pa th. Modification was approved under 50.59 Evalua- '

I tion and Determination #85-17 for all blades to allow inspection using a camera and light enhancement to prove the S-3 blade was moving freely within the shroud over its full travel path with no apparent mechanical clearance problem (25 Oct 85, MLP #61).

1 November 1985 Performed preventive maintenance on security system ce11 monitors (1 Nov 85, MLP #62).

18 November 1985 A Annual check and update of the Emergency Call Lists.

20 November 1985 Drained and flushed primary coolant water storage tank (23 Jan 86, MLP #63).

21 November 1985 Cut shroud inspection ports out of the top of the shrouds for S-1, S-2 and the Regulating Control Blades (6 Jan 86, MLP #64).

V-3

I Date Descript. ion 26 November 1985 Conducted practical training on Emergency Response Equipment and facilities.

3 December 1985 Performed security system maintenance (3 Dec 85, MLP j

1. 65).

3 December 1985 Performed security system maintenance (4 Dec 85, MLP

1. 66).

l 10 December 1985 A Annual check and update of Emergency Call Lists (update of November, 1985 check).

11 December 1985 Performed security system maintenance (11 Dec 85, MLP

1. 67).

11 December 1985 Q Quarterly Radiological Emergency Evacuation Drill - large annuti drill involving all outside agencies.

11 December 1985 S Semiannual check of (Replacement) of Security System Batteries.

12 December 1985 Installed a modified out-of-core control blade drive unit /long shaf t coupling modification to make use of I shaf t coupling assemblies manufactured to meet UFTR Drawing 89-31-116 which was found to be in error (17 Dec 85, MLP #68).

I 13 December 1985 Installed shim materials for bushings on S-3 and RB control blade systems (9 Jan 86, MLP #69).

I 13 December 1935 Performed drop time retest checks on S-3 Safety and Regulating Blades without clutch current applied, with current applied and shielding in place (18 Feb

• t 86, MLP 770).

16 December 1995 Replaced North Area Monitor with spare af ter setting bias on detector (17 Dec 85, MLP # 71 ).

l l" 30 December 1985 Replaced fan belt on cell air handling system (UP

Mechanical Maintenance) (2 Jan 86, MLP #72).

9 January 1986 Performed drop time retest checks on S-1 and S-2 Safety Blades for various reassembly stages (31 Jan 86, MLP #1).

16 January 1986 Cleaned fuel boxes to remove foreign matter (16 Jan 86, MLP #2).

I 16 January 1986 Performed main tenance to correct primary coolant sys-tem leak from reassembled fuel boxes (23 Jan 86, MLP i

1. 3).

V-4

W I

Da te Description 27 January 1986 Installed modified aluminum spacer plates below the I UFTR core center island graphite per 50.59 Evaluation No. 86-5 (31 Jan 86, MLP #4).

28 January 1986 I Replaced makeup water domineralizer resins (28 Jan 86, MLP #5).

29 January 1986 Q Quarterly Calibration Check of Area and Stack l

5 Radiation Monitors ( Area Monitors due in December, Stack in January).

I 30 January 1986 S Semiannual Leak Check of Antimony-Beryllium Neutron Source.

30 January 1986 I S Semiannual Replacement of Secondary Deep Well Pump _ Motor Fuses.

31 January 1986 A Annual Replacement of Control Blade Clutch Cur-rent Light Bulbs.

31 January 1986 S Measurement of Control Blade Drop Times.

31 January 1986 Corrected control blade drive motor failure due to loss of power by cleaning the brush contacts for I Safety Blade S-2 and Regulating Blade (31 Jan 86, MLP

1. 7).

2 February 1986 S Measurement of Control Blade Controlled Inser-tion Times.

3 February 1986 Q Quarterly Check of Scram Functions.

4 February 1986 Performed nuclear instrumentation voltage adjustments and checked calibration.S (A-2 Surveillance) (27 Feb '

86, MLP #9).

4 February 1986 Manufactured a new SbBe source holder to facilitate remote manipulation per 50.59 Evaluation #86-6 (5 Feb 86, MLP #10).

4,27 February 1986 A Annual UPTR Nuclear Instrumentation Calibration I check and Calorimetric Heat Balance (all except calo-rimetric heat balanace and calibration check com-pleted on 4 February, remainder on 27 February 86).

7 February 1986 Removed protruding excess bolt length from fuel bun-die UP-32 per 50.59 Evaluation #86-9 (7 Feb 86, MLP

1. 11).

10 February 1986 Replaced shield tank recirculating system ceramic filter (10 Feb 86, MLP #12).

I I V-5

W ~

l Date Description 10-24 February 1986 S Annual Reactivity Measurements: Worth of Con-trol Blades, Total Excess Reactivity / Maximum Reac-tivity Insertion Rate and Shutdown Margin.

13 February 1986 Cleaned contacts to 2-pen recorder to correct red pen channel failure (13 Feb 86, MLP #13).

I 17 February 1986 Added water to the primary coolant storage tank to raise level from 28 inches to 31 inches (17 Feb 86, MLP #14).

17 February 1986 Restored free movement of 2-pen recorder chart drive gear by resoldering drive platen of 2-pen recorder (17 Feb 86, MLP #15).

~

18,27 February 1986 Primary coolant 1 gallon sample removal following milestones in reactor recovery (af ter low power run-ning and af ter first two days of full power checks).

19 February 1986 Reseated Nixie Tube in socket for the Regulating Blade Position Indicator to restore proper indication (19 Feb 86 MLP #16).

19 February 1986 Cleaned contacts for Blade Position Indicators for S-1, S-2 and S-3 control blades (19 Feb 86, MLP #17).

20 February 1986 Replaced hundredths card for the Regulating Blade Po-I sition Indicating Nixie Tube to correct failed indi-cator (20 Feb 86, MLP #18).

20 Fobruary 1986 Cleaned contacts on Safety-Three Blade Position Indi-

'I cator to restore proper indication (20 Feb 86, MLP

1. 19). .

21 February 1986 Shipped 40 unirradiated high enriched fuel plates to

-I Oak Ridge National Laboratory per NRC Show cause Order dated September 27, 1985.

25 February 1986 A Annual Measurement of UPTR Temperature Coeffi-cient of Reactivity.

25 February 1986 Replaced failed APD recorder motor and modified

'I striker bar interrupter arm per 50.59 Evaluation #86-10 (25 Feb 86, MLP #20).

25 February 1986 Corrected negative bottom position indication on Safety-Three control blade position indicator (25 Feb 86, MLP #21).

25-26 February 1986 Q Quarterly Radiological Survey of Restricted A reas.

I V-6

Da te Descrip tion 26 February 1986 Performed maintenance on spare blade position indica-I tor assemblies to provide spares in the event of BPI failures (open, MLP #22).

I 26-27 February 1986 Q Quarterly Radiological Survey of Unrestricted Areas.

27 February 1986 .

I 28 February 1986 S Measurement of Argon-41 Stack concentration.

S Measurement of Dilution Air Flow Rate.

I 28 February 1986 Replaced shield tank recirculation system ceramic filter and added 35 gallons of demineralized water (3 Feb 86, MLP #6).

18 March 1986 S Semiannual Leak Check of PuBe Nsutron Source.

18 March 1986 Hooked up and removed hot water heater for tempera-ture coefficient measurements (19 Mar 86, MLP #23).

18 March 1986 Added ~18 gallons of demineralized water to the pri-mary coolant storage tank (18 Mar 86, MLP #24).

31 March 1986 Replaced the Nixie Tube Decade Board (units) of the S-2 blade position indicator with a pretested spare II (31 Mar 86, MLP #25).

3 April 1986 Replaced failed switch in relay circuit of shield I 4 April 1986 tank pump (3 Apr 86, MLP #26).

S Semiannual Inventory of Security-Related Keys for UFTR and UFSA.

7 April 1986 s Semlannual Inventory of Special Nuclear '

Ma terial.

.I 7 April 1986 cleaned cont.; cts of Safety #2 blade position indica-tor to corret t flickering indication (7 Apr 86, MLP I 7 April 1986

1. 27).

Removed Rabbit S*f stem assembly from UPTR, re trieved I stuck capsule, verified system failure and decided to replace system with now design (28 Apr 86, MLP #28).

I 7/14/21/27 April 1986 Withdrawal of one galfon primary coolant samples for baseline activity analysis project for developing a model of expected coolant fission product buildup and decay as UPTR power running is conducted.

14 April'1986 Replaced APD motor front bearing and side wear plates to restore air flow rate (14 Apr 86, MLP #29).

I V-7

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Date Description 14 April 1986 Replaced failed amplifier in W/R Drawer to correct problem of calibration check points not responding properly during a checkout (16 Apr 86, MLP #30).

I 25 April 1986 Q Quarterly Radiological Emergency Evacuation Drill.

5/12/19/27 May 1986 Withdrawal of one gallon primary coolant' samples for baseline activity analysis project.

5 May 1986 Replaced Safety Blade S-2 blade position indicator with a repaired and tested spare (5 May 86, MLP #22).

15 May 1986 Replaced failed resistor in 24 volt area monitor bat-tery charger (16 May 86, MLP #31).

21 May 1986 Replaced motor bearings and shock-mounting feet on APD blower (21 May 86, MLP #32).

3 June 1986 Replaced diluting fan duct flex ccupling (14 Aug 86, MLP #33).

13 June 1986 Q Quarterly check of Scram Punctions.

I 13 June 1986 S Semiannual Check / Replacement of Security System Ba tteries.

I 19 June 1986 Replaced fan belts and then replaced rubber tacho-meter generator coupling with modified aluminum coup-ling per 50.59 Evaluation #86-13 and 86-15 to restore diluting fan RPM indication (30 Jun 86, MLP #35).

26 June 1986 S Measurement of Dilution Air Flow Rato, ,

I 26 June 1986 Q Quarterly Radiological Emergency Evacuation Drill.

I 27 June 1986 A Annual check and Update of Emergency Call Lists.

2 July 1986 Repaired Pager #0740 used for emergency drills and carried by SRO's on call (12 Aug 86, MLP #36).

3 July 1986 Replaced grating over cell ventilation air intake vent (2 Jul 86, MLP #34).

3 July 1986 S Measurement of Dilution Air riow Rate.

3 July 1986 S Measurement of Argon-41 Stack Concentration.

3 July 1986 Q Quarterly Radiological Survey of Restricted Areas.

l w

I V-8

l l Date Descrip tion 8 July 1986 Performed security system maintenance (8 Jul 86. MLP

1. 37).

18 July 1986 Q Quarterly Radiological Survey of Unrestricted Areas.

25 July 1986 Q Quarterly Calibration Check of Area and Stack Radiation Monitors.

25 July 1986 Replaced failed stack monitor recorder with spare from stock (25 Jul 86, MLP #38).

28 July 1986 Adoed ~40 gallons of makeup demineralized water to the priraary coolant storage tank (28 Jul 86, MLP

1. 39).

29 July 1986 Replaced aluminum shaf t coupling on stack tachometer l fan with spare made of 316 stainless steel per 50.59 l Evaluation #86-16 (29 Jul 86, MLP #41).

4 August 1986 Replaced flexible shim on tachometer alignment brac-I ket to align tachometer and diluting fan shaf t and restore control room diluting fan RPM indication (1 Aug 86, MLP #42).

4 August 1986 Reattached small press fitted gear in two pen re-corder to correct free wheeling (4 Aug 86, MLP #43).

13 August 1986 Cleaned contacts to free a stuck chopper on the two pen recorder to restore operation of linear channel (13 Aug 86, MLP #44).

18 August 1986 Replaced ceramic filter on shield tank recirculation system (18 Aug 86, MLP #45). ,

21 August 1986 Cleaned and tightened demineralizer connections to correct small primary coolant leak from the connec-tions into the pit sump (25 Aug 86, MLP #46).

22 August 1986 Replaced broken needle on stack monitor recorder with spare needle (22 Aug 86, MLP #47).

25 August 1986 S Semiannual Measurement of Control Blade Drop Times.

25 August 1986 S Semiannual Measurement of Control Blade Con-trolled Insertion Times.

25 August 1986 Adjusted set point on East Area Radiation Monitor to alarm at 10 mR/hr (25 Aug 86, MLP #48).

I V-9

Date Description 25 August 1986 Began repairs on spare Area Radiation Monitor Re-corders (remains open, MLP #49).

25 August 1986 Began repairs on spare Blade Position Indicators (re-mains open, MLP #50).

19 September 1986 overhauled failed walkie-talkie radio and replaced battery (19 Sep 86, MLP #40).

NCYrE:

Infomation in parentheses gives the maintenance log page (MLP) nuirber and the date for closure which is nomally the date that maintenance is completed and the system in question is restored to nomal operability.

This closure date may be some time af ter work begins depending on work to be performed, availability of parts, need for system involved or operability of system affected.

I l

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VI. CHANGES TO TECHNICAL SPECIFICATIONS, STANDARD OPERATING PROCEDURES AND OTHER DOCUMENTS A. The new Technical Specifications for the UFTR were issued on August 30, 1982 and officially established on September 30, 1982. Two sets of re-quested corrections / changes to the Technical Specifications were sub-mitted to the NRC during the 1982-1983 reporting period. As. noted in the I 1983-1984 Annual Report, the UPTR facility received approval for Amend-ment No. 14 and No. 15 to the UFTR Technical Specifications during that reporting year.

During the 1985-1986 reporting year, one change request was submitted for the approved UFTR Tech Specs; this change was requested with a letter dated October 17, 1985 and was intended simply to correct an error in I numbering Section 3.5 " Limitations on Experiments" which had been incor-rectly numbered as section 3.4. This change was approved an Amendment 16 to the UFTR operating license and Technical Specifications via a letter from NRC dated Novmeber 25, 1985. No further requests for changes in the approved Tech Specs are expected for the operation of the UPTR with its present high-enriched fuel at a rated power level of 100 kWth. It is ex-pected, however, that another substantive amendment to the Technical Spe-cifications will be required before the UPTR can be converted from utili-zing high-enriched MTR plato-type fuel to utilizing low-enriched SPERT pin-type or silicide plate-type fuel. This decision will be made during the upcoming reporting year as DOE support becomes available.

D. Generation of New Standard Operating Procedures Following an NRC inspection on February 11-15, 1985 (previous reporting year) and as part of NRC Inspection Number 50-83/85-01, the UPTR Licensee was cited specifically for failure to adequately control and document a revision to the reactor control circuit when an interlock was installed ,

which provided for a trip of the diluting fan / vent fan on activation of I the evacuation alarm, and generally for a failure to adequately implement the guidelines for a Quality Assurance Program as delineated in ANSI Standard N402-1976 referenced in Chapter 17 of the UPTR Safety Analysis Report.

In responding to this Inspection Report, a number of changes and addi-tions were made to the UFTR Standard Operating Procedures. The changes in I response to commitments made in the facility licensee response to NRC In-spec tion Report No. 50-83/85-01 during this 1985-1986 reporting year in-cluded generation and implementation of two new procedures: UFTR SOP-0.3, I " Control and Documentation of UFTR Modi fications" and UPTR SOP-o.5, "UFTR Quality Assurance Program."

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I VI-1

v.

I I UFTR SOP-0.3 was reviewed and approved to assure adequate documentation was obtained and maintained to assure quality in UFTR Modifications in-cluding reviews. UPTR SOP-0.5 was reviewed and approved to implement an overall documented Quality Assurance Program at the UFTR facility, espe-cially emphasizing nuclear safety-related items. With the completion of these two procedures, the procedures committed to be developed in re-sponse to Inspection Report No. 50-83/85-01 are completed to address the I items delineated in ANSI S tandard N402-1976. During the NRC Inspection conducted on February 18-21, 1986, Inspector Barnett closed out this and all other items from NRC Inspection Report No. 50-83/85-01.

One other new procedure was implemented during the latest reporting year as UPTR SOP-E.8, " Verification of UPTR Negative Void Coefficient of Reac-tivity." This procedure was developed and implemented to control the me-thod by which the UFTR void coefficient is verified to be negative. The void coefficient is required to be determined to be negative biennially as per UFTR Technical Specifications, Section 4.2.1, Paragraph (3). This I procedure simply implements and controls a method of verification pre-viously implemented as an experiment via UFTR SOP-A.5. Since the method has been tested and demonstrated to be adequate, SOP-E.8 now implements I and documents the method for use as a regular surveillance to meet the Tech Spec requirements.

Since these three procedures are all newly generated during the latest reporting year, the full text of all three procedures is contained in Ap-pendix C for reference purposes and to meet Tech spec requirements. Since it was revised in February, 1986 af ter initial implementation in Decem-ber, 1985, only the latest revision of UFTR SOP-0.5, "UPTR Quality As-surance Program" is contained in Appendix C.

C. Revisions to Standard Operating Procedures All existing UPTR Standard Operating Procedures were reviewed and rewrit-ten into a standard fonnat during the 1982-1983 reporting period as re-quired by a commitment to NRC following an inspection during that year. ,

As committed to NRC, the final approved version of each SOP (except se-

'l curity response procedures which are handled separately) is permanently 5 stored in a word processor to facilitate revisions and updates which are incorporated on a continuing basis in the standard format.

Table VI-1 contains a completc lis t of the approved UFTR Standard Operat-ing Procedures as they existed at the end of the previous (1984-1985) re-porting year. The latest revision number and date for each non-security related procedure is listed in Table VI-1 During the 1985-1986 reporting year, a number of changes were incorporated into the UFTR Standard Opera-ting Procedures as needs and/or errors were identified. " Temporary Change Notices" were incued to correct minor discrepancies or better express the intent of several procedures to include SOP-0.2, SOP-0.5 , SOP- A.1, SOP-A.2, SOP-D.1, SOP-D.4 , SOP-E.7 and SOP-F.1.

I I

I VI-2

v Only two procedures were revised during this reporting year to include SOP-0.4, "10 CFR 50.59 Evaluation and Determination" and SOP-0.5, "UFTR Quality Assurance Program." The latter SOP was originally generated dur-I ing this reporting year in December, 1985 so only the revision is ad-dressed in this report. Both revisions were substantive and essentially to implement better the UFTR Quality Assurance Program.

I Following an NRC inspection conducted on February 18-21, 1986, and as part of NRC Inspection Report No. 50-83/86-01 dated March 27, 1986, the facility license was cited for a Severity Level IV violation for the I failure to maintain adequate records of the safety evaluation for 50.59 Evaluation and Determination #85-17 entitled, "t FTR Control Blade Shroud Spacer Cutting Alteration for Ease of Blade Inspection." The low level I citation was basically for failure to maintain adequate records of fa-cility changes as required by 10 CFR 50.59(b). Therefore, to assure ada-quate records of safety evaluation reviews of facility modifications are main tained, SOP-0.4 used to control the evaluation of 50.59 changes to I aasure no unreviewed safety questions are involved was revised. The SOP-0.4 revision consisted primarily of changes to incorporate a new fom UFTR SOP-0.4B entitled, " Supporting Material for 10 CFR 50.59 Determina-I tion." Since the NRC Inspection Report No. 50-83/86-01 cited the facility for failure to provide adequate documentation of all the information re-viewed as part of a 50.59 Evaluation and Determination, this change was implemented af ter careful review to assure items considered in evaluating

, a modification are easily and c1carly documented for later reference.

The revision to SOP-0.5 consisted of including all the quality assurance-

=

related surveillance, test and check forms in SOP-0.5 which had not pre-viously been controlled as part of any standard operating procedure.

Since the SOP-0.5 procedure describes the overall implementation of the I UPTR Quality Assurance Program, in response to the February 11-15, 1985 NRC inspection which cited lack of our adequately documented Quality As-surance Program, these special forms were incorporated into the first re-

.g vision of SOP-0.5 in February, 1986 to assure necessary documentation of I a ' Pha=es of the UFTR quality Assurance Program in the complete set of Standard Operating Procedures. .

Since those procedures were subjected to substantive revision during the current reporting year, the entire text of the currently implemented pro-cedures SOP-0.4 and SOP-0.5 are contained in Appendix B for reference purposes. Note only the la test Revision 1 of SOP-0.5 is included.

I I

lI i VI-3 u

W D. Revisions to Safety Analysis Report Following the NRC inspection conducted February 18-21, 1986, NRC Inspec-tion Report No. 50-83/EG-01 identified several errors or apparent incon-sistencies in the UFTR Safety Analysis Report. The facility was committed to revim 3AR Paragraph 7 ?.3, "Non-Nuclear Instruementation Channels" and revise ac necessary. As requested in the Inspection Report, our fa-I cility committed to update Paragraph 7.2.3 of the UFTR Safety Analysis Report "which describes cperation of the control rod inhibit system and automatic control system, which is different from the performance de-scribed in Technical Spe71fication 3.2.1. Surveillance procedures con-firmed performance in conformance to the requirements of Technical Speci-fications." As a result of the review of Paragraph 7.2.3, SAR Revision 2 I also included revisions to Sections 7.3 and 7.6 as well as a rela ted Figure 1-8 from Chapter 1 of the SAR.

Many of the changes represent simple typographical errors or omissions, I though several involve facility description discrepancies discovered as a result of the NRC inspection in February, 1986. All changes were reviewed by UFTR staff and by the Reactor Safety Review Subcommittee as required by Technical Specifications.

Revision 2 was sent to NRC with a letter of transmittal dated July 18 1986. The entire transmittal is included in Appendix D which also in-I cludes a detailed description and justification of the revision along with a summary table of the changed pages as well as the actual SAR re-placement pages that constitute Revision 2.

NRC Inspection Report No. 50-83/86-01 also required review of CORA four group noutron flux shape calculations for proper identification of mate-

E rials and revisions of SAR Figure 4-16. Initial review of these calcula-g tions revealed that the incorrectly labeled Figure 4-16 had been cor-rected as part of SAR Revision 1 submitted to NRC on May 11, 1982 and not I inserted into the SAR volumes supplied to the NRC inspector. Addi tional review of the input for the CORA calculations verified the corrections of Revision 1 so no SAR Revision was necessary for Figure 4-16. However, as ,

a result of discovering an incomplete SAR volume, all SAR volumes are being updated to preclude future uncertainties on SAR revisions.

At this time , further SAR revisions are anticipated only when modifica-

E tions arrectino s^n contents are installed. Severa1 such modifications LE are planned but not immediately. Most such modifications are planned to occur during the conversion process fmn using high-enriched to low-en-riched fuel in the core. The SAR revision at this point should be signi-

ficant and possibly quite lengthy.

I

I VI-4 C

I E. Revisions to Physical Security Plan First, no violations or deviations were noted during a routine safeguards I inspection on June 24-25, 1985 by Region II inspectors. However, as a re-sult of the inspection several minor discrepancies were noted in the Phy-sical Security Plan (PSPh Therefore, in a letter dated July 10, 1985, a I proposed Revisien G of the UFTR Physical Security Plan was transmitted updating q'tantitics of fuel stated to be on hand in several places in the PSP and incorporating the latest version of all security response pro-cedures. NRC approval and acceptance of this submission as Revision 8 of I the UPTR Security Plan was received on July 30, 1985 with only one small portion of a security response procedure deemed inconsistent and hence rejected and required not to be implemented. Since this inconsistency was only" the result of NRC interpretation, there was no problem and the in-

-I consistent portion of the Security Response Procedure in question was re-written as proposed Revision 9 to the UFTR Physical Security Plan early I in this reporting year. This preposed change was reviewed by the RSRS and submitted to NRC in a letter dated September 6,1985 as Revision 9 to the UFTR Physical Security Plan.

I This Revision 9 was installed in the UPTR Physical Security Plan on Sep-tember 10, 1985 and implemented upon official notification from NRC on October 3, 1985 tha t the proposed Revision 9 was acceptable. The instal-

'g .lation itself was transparent in its effect as no security operations g were affected, only administrative interpretations by NRC.

'E No violations or deviations were noted during a routine NRC Safeguards g inspection conducted on May 28-29, 1986 for the current reporting year.

At this time, since there were no discrepancies noted, no further revi-sions are planned for the Physical Security Plan. The Physical Security Plan (Revised) is withheld from public disclosure.

F. Quality Assurance Program Approval For Radioactive Material Package A Show Cause Order dated September 27, 1985 was received on October 3,

• 1985 giving the Facility Licensee 30 days to respond to the order to re-move all but one bundle of unirradiated HEU fuel to a secure facility within 120 days or have the order take effect as written. Our official response dated October 23, 1985 was sent by certified mail to the NRC I NMSS Office in Washington, D.C. In the response a relaxation of the time limits only was requested to provide time to check our fuel and deterTnine how much of the unirradiated HEU fuel would be needed during refueling I prior to shipping any of it. In addition, we requested being able to keep two plates for experimental purposes. We also asked for relaxation of the 120-day time limit since DOE could not guarantee meeting this limit with a shipping container and other support. It might also not have been pos-I' sible to check our fuel prior to reload following repair work on the sticking Control blades and still ship the unneeded fuel within the 120 day limit. Following extensive discussions with NMSS in Washington, it

. was determined that we should meet the 120 day limit but that the 120 day time limit would be applied from the date of our response (October 23) which set February 20 to remove the unneeded unirradiated HEU fuel. Fol-j lowing shipping from DOE Idaho in mid-December, the 6M Type B shipping 5 container arrived on 16 January 1986. Despite some relaxation of require-I VI-5

I ments, in January, the decision was made to ship out 40 unirradiated high

-I enriched fuel plates by the deadline of February 20, 1986 since the core was expected to be reloaded prior to February 20.

On 6 February 1986, it was determined that the UFTR licensee needed a specific QA program to ship fuol. A QA program was generated for review and approved at the February RSRS meeting, submitted to NRC and approved by NRC in a letter dated 19 February 1986 as Quality Assurance Program Approval for Radioactive Material Packages - Number 0578. The unirra-diated HEU fuel (40 plates containing 578.13 gms U-235/620.l 7 8 gms U) was then loaded into the 6M container on 19 February and finally picked up for delivery by Napier/ Airborne Express on 21 February (within the NRC-interpreted 120 day time limit as verified with NMSS). The shipment fi '

nally reached the ORNL Y-12 plant on 27 February 1986 af ter several un-successful transportation moves by truck to various. cities. During the NRC Safeguards inspection on May 28-29, 1986 the records and implementa-tions of the QA Program For Shipping Radioactive Material were reviewed I and determined to be satisfactory under the conditions of Approval No.

0578. The QA Program is withheld from public disclosure.

I G. Reactor Operator Recertification Plan I The unavailability of the University of Florida Training Reactor (UFTR) following the discovery of a sticking S-3 control blade on September 3, 1985 due to maintenance work and subsequent unloading of fuel in October, 1985 resulted in the inability to meet some of the requalificaiton train-ing program requirements delineated in our approved Operator Requalifica-tion and Recertification Training Program for 1985-1987. Specifically the intervals for performing startups and shutdowns as well as for performing I weekly and daily checkouts were exceeded since complete checkouts were not possible af ter the fuel was unloaded in mid-October. Therefore, as counselled by NRC, a Reactor Operator Recertification Plan was developed I to recertify that the knowledge and understanding of facility operations and administration were satisfactory for each UFTR operator prior to the individual's performance of licensed duties for normal operations. The letter of transmittal for this plan (dated January 6,1986) as well as the Recertification Plan i tself are included in Appendix E.

The certification program was divided into two primary parts because cer-I tain requirements had to be met to move irradiated fuel and reload the core (licensed activities) and only then would complete daily and weekly checkouts be possible as required by the Requalification Training Pro-I gram. Prior to commencing reactor operations, the second part of the cer-tification training was required before internally certifying operators for the lengthy task of performing a complete set of reactor operational checkouts and tests (many of which required running the reactor at part I or even full power) as required by the Tech Specs before the reactor could be returned to normal operations. Approval was received in a letter from NRC dated January 28, 1986 indicating this Plan would serve as an I adequate basis for certifying that each operator had satisfactory know-ledge and understanding of facility operation as required by 10 CPR 55.31 ( e ) . This Recertification Plan was then used to recertify all UFTR I opera tors with final recertifica tion for all completed on March 3, 1986 with the information also transmitted to NRC on that da te.

I VI-6

w I

TABLE VI-1 1

LISTING OF APPROVED UFTR STANDARD OPERATING PROCEDURES

( August 31, 1985)

O. Administrative Control Procedures 0.1 Operating Document Controls (REV 0, 2/83) 0.2 Control of Maintenance (REV 3, 5/85) 0.4 10 CFR 50.59 Evaluation and Determination (REV 0, 3/85)

A. Routine Operating Procedures A.1 Pre-operational Checks (REV 13, 6/85)

A.2 Reactor Startup (REV 11, 5/85)

A.3 Reactor Operation at Power (REV 10, 5/85)

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 Differential Reactivity Worth (REV 1, 6/85)

B. Emergency Procedures B.1 Radiological Emergency (REV 3, 5/83)

I B.2 B.3 B.4 Fire (REV 5, 5/85)

Threat to the Reactor Facility (Expanded into F-Series Procedures)

Flood (REV 1, 4/83)

C. Fuel Handling Procedures C.1 Irradiated Fuel Handling (REV 4, 2/85)

I C.2 C.3 C.4 Fuel Loading (REV 4, 4/83)

Fuel Inventory Procedure (REV 3, 2/85)

Assembly and Disassembly of Irradiated Fuel Elements (REV 0, 9/84) '

D. Radiation Controls Procedures D.1 Radiation Protection and Control (REV 3,1/83)

I D.2 D.3 Radiation Work Permit (REV 9, 5/85)

Primary Equipment Pit Entry (REV 2, 5/85)

D.4 Removing Irradiated Samples From UPTR Experimental Ports (REV 3, 5/85)

E. Maintenance Procedures I E.1 E.2 Changing Primary Purifica tion Demineralizer Resins (REV 3, 6/85)

Alterations to Reactor Shielding and Graphite Configuration (REV 2, 4/83)

E.3 Shield Tank and Shield Tank Recirculation System Maintenance (REV 2, I E.4 4/83)

Superceded E.5 Superceded E.6 Argon-41 Concentra tion Measureman* (REV 0, 1/84)

E.7 Measurement of Temperature Coef ficient of Reactivity (REV 0, 5/85)

I VI-7

I TABLE VI-1 (CONTINUED)

F. Security Plan Response Procedures (Reactor Safeguards Material, Disposi-E tion Restrictedi I F.1 F.2 Physical Security Controls (Confidential, except for UFTR Form SOP-F.1A)

Bomb Threat (Confidential, except for UPTR Fonn SOP-F.2A)

F.3 Thef t of (or Threat of the Thef t of) Special Nuclear Material (. Con-fidential, except for UFTR Form SOP-F.3A)

F.4 Civil Disorder (Confidential)

P.5 Fire or Explosion (Confidential)

F.6 Industrial Sabotage (Confidential)

P.7 Security Procedure Controls (REV 1, 9/84) l I

I l

l I

I S I

I I

I  ;

I VI-8

V TABLE VI-2 LISTING OF APPROVED UFTR STANDARD OPERATING PROCEDURES (August 31, 1986)

O. Administrative Control Procedures 0.1 Operating Document Controls (REV 0, 2/83) 0.2 Control of Maintenance (REV 3, 5/85) 0.3 Control and Documentation of UFTR Modifications (REV 0,10/85) 0.4 10 CFR 50.59 Evaluation and Detemination (REV t ,. 5/86) 0.5 -UPTR Quality Assurance Program (REV 1, 2/86)

A. Routine Operating Procedures A.1 Pre-operational Checks (REV 13, 6/85)

A.2 Reactor Startup (REV 11, 5/85)

A.3 Reactor Operation at Power (REV 10, 5/85)

A.4 Reactor Shutdown (REV 9, 6/85)

I A.5 A.6 Experiments (REV 3, 4/83)

Operation of Secondary Cooling Water (REV 1,10/83)

A.7 Determination of Control Blade Integral or Differential Reactivity Worth (REV 1, 6/85)

B. Emergency Procedures B.1 Radiological Emergency (REV 3, S/83)

B.2 Fire (REV 5, 5/85)

B.3 Threat to the Reactor Facility (Expanded into F-Series Procedures)

B.4 Flood (REV 1, 4/83)

C. Fuel Handling Procedures C.1 Irradiated Fuel Handling (REV 4, 2/85)

C.2 Fuel Loading (REV 4, 4/83) '

I C.3 C.4 Fuel Inventory Procedure (REV 3, 2/85)

Assembly and Disassembly of Irradiated Fuel Elements (REV 0, 9/84)

D. Radiation Controls Procedures D.1 Radiation Protection and Control (REV 3, 1/83)

D.2 Radia tion Work Permit (REV 9, 5/85)

I D.3 D.4 Primary Equipment Pit Entry (REV 2, S/85)

Removing Irradiated Samples From UPTR Experimental Ports (REV 3, 5/85)

I I

I VI-9

F TABLE VI-2 (CONTINUED)

E. Maintenance Procedures E.1 Changing Primary Purification Demineralizer Resins (REV 3, 6/85)

E.2 Alterations to Reactor Shielding and Graphite Configuration (REV 2, 4/83)

E.3 shield Tank and Shield Tank Recirculation System Maintenance (REV 2, 4/83) ,

E.4 Superceded E.5 Superceded E.6 Argen-41 Concentration Measurement (REV 0,1/84)

I E.7 E.8 Measurement of Temperature Coefficient of Reactivity (REV 0, S/85)

Verification of UPTR Negative Void Coefficient of Reactivity (REV 0, 12/85)

F. Security Plan Response Procedures (Reactor Safeguards Material, Disposi-tion Restricted)

F.1 Physical Security Controls (Confidential, except for UFTR Form SOP-P.1A)

P.2 . Bomb Threat (Confidential, except for UFTR Form SOP-F.2A)

P.3 Thef t of (or Threat of the Thef t of) Special Nuclear Material (Con-fidential, except for UFTR Form SOP-F.3A)

P.4 Civil Disorder (Confidential)

I F.5 F.6 P.7 Fire or Explosion (Confidential)

Industrial Sabotage (Confidential)

Security Procedure Controls (REV 1, 9/84)

I 5

I VI-10

I VII. RADIOACTIVE RELEASES AND ENVIRONMENTAL SURVEILLANCE This chapter summarizes the gaseous, liquid and solid radioactive releases

.E from the UFTR facility for thic reporting year. A rgon-41 is the primary gas-E cous release while there was no measureable liquid Telease and no solid re-lease at all. Finally, this chapter includes a summary of personnel exposures at the UPTR facili ty.

.I A. Gaseous (Argon-41)

The gaseous releases from the UPTR Facility for this reporting year are summarized in Table I. The basis for the gaseous activity release values is indicated in Table II. These values are obtained by periodic measurements of stack concentrations as required by Technical Specifications.

TABLE I UPTR GASEOUS RELEASE

SUMMARY

Month Release Average Monthly Concentration Sep tember, 1985 0.00 x 106 pCi/ Month 0.00 x 10-9 pCi/ml Oc tober, 1985 0.00 x 10 6 pCi/ Month 0.00 x 10-9 pCi/ml November, 1985 0.00 x 10 6 pci/ Month 0.00 x 10-9 pCi/ml December, 1985 0.00 x 10 6pCi/ Mon th 0.00 x 10-9 pCi/ml January, 1986 0.00 x 10 6pCi/ Mon th 0.00 x 10-9 pCi/ml February, 1986 4.54 x 10 6pci/ Mon th 1. 58 x 10-9 pCi/ml March, 1986 1.72 x 10 7pCi/ Mon th 5.98 x 10-9 pCi/ml 7

April, 1986 1. 71 x 10 pCi/ Mon th 5.95 x 10-9 pCi/ml May, 1986 2.03 x 10 7pCi/ Mon th 7.05 x 10-9 pCi/ml June, 1906 1.11 x 107 pCi/MonUi 3.86 x 10-9 pCi/ml July, 1986 1.30 x 10 7pCi/ Mon th 4.35 x 10-9 pCi/ml August, 1986 1.391 x 10 7pCi/ Mon th 4.65 x 10-9_pci/ml TOTAL ARGON-41 Releases................................ 97.07 Ci AVERAGE ARGON-41 Release Concentration....... 2.79 x 10-9 pCi/ml UFTR Technical Specifications require average Argon-41 release concentra-tion averaged over a month to be less than 4.0 x 10-8 pCi/ml. Total releases and average monthly concentra tions are based upon periodic Argon-41 release concentra tion measurements made at equilibrium full power (100 Kw) condi tions.

I The results for these experimental measurements used in calculating the gas-cous Ar-41 release data are summarized in Table II.

VII-1

TABLE II UPTR GASEOUS RELEASE DATA BASE Releases Per Unit Instantaneous Argon-41 Months Energy Generation Concentration at Full Power Sept. '85 - Jan. '86 5144.2 pCi/Kw-hr 13.24 x 10-8 pCi/ml Feb. '86 - June '86 5196.5 pCi/kw-hr 13.00 x 10-6 pCi/ml July '86 - Aug '86 4650.7 pCi/kw-hr 11.20 x 10-8 pCi/ml B. Liquid Waste from the UFTR/ Nuclear Sciences Complex There were approximately 73,950 liters discharged from the liquid waste holdup tanks to the campus sanitary sewage system during this reporting period. For this period there was only one single batch discharge as summa-rized here:

Volume Concentra tion Date (li ters) (pci/ml)

January 16-17, 1986 73,950 1.30 x 10-8 I The minimum detectable activity (MDA) is 3.66 x 10-9 pCi/ml.

The effluent discharged into the holding tanks comes from twenty labora-tories within the Nuclear Sciences Center as well as the UPTR complex. The UFTR normally releases approximately 1 liter of primary coolant per week to

• the holding tank as waste from primary coolant sampling. Because of the drain-ing of the primary coolant system during maintenance work, only 42 samples were taken during this reporting year with none available between November 25, 1985 and January 28, 1986. The average activity for this coolant was I 2.1 x 10-7 pCi/ml (6 - y) and 2.0 x 10-7 pCi/ml (a) for this 1985-1986 annual reporting period. The minimum detectable activity (MDA) on these samples is 1 x 10-l pCi/ml (6 - y) and 1 x 10-8 pCi/ml (a).

During the cleanout of the primary coolant system on November 21, 1985 approximately 655 liters of primary coolant were discharged to the waste hold-up tanks with an average activity of 1.2 x 10-7 pCi/ml (6 - y) and 1.0 x 10-8 pCi/ml (a) again 5;ith the MDA at 1 x 10-10 pCi/ml (S - y) and 1 x 10-8 pCi/ml (a). The total activity released to the tanks on this date then is estimated to be 7.9 x 10-2 pCi (6 - 7) and 6.5 x 10-3 pCi (a).

I I

VII-2

n-C. Solid Waste Shipped offsite The UPTR facility made one shipment of colid waste during this reporting year. This shipment was made on December 10, 1985 through ADco services, Inc.

and consisted of 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 sys tems. The activity of the shipment was approximately 3.125 curies with the activity primarily attributed to Cobalt-60. A similar shipment of two drums is planned for the next reporting year to remove all the products I resulting from the control blade restoration and maintenance project. No date has been set for this next shipment D. Environmental Monitoring The UFTR maintains continuous film badge as well as thermoluminescent dosimeter monitoring (new for the 1982-1983 reporting period) in areas adja-I cent to and in the vicinity of the UPTR complex. The badge and TLD cummulative totals for this reporting period from September 1985 through August 1986 are summarized in Table III.

E. Personal Radiation Exposure The maintenance work to restore proper function of all control blade drive systems necessitated higher facility personnel exposures than in most years. The doses recorded represent efforts to follow ALARA controls. The doses are actually quite low when the extent and complexity of the work in-volved is considered, especially considering the necessity to unload the full UPTR core, remove all core graphite, disconnect and remove all control blades, shrouds and fuel boxes, reconnect / replace all these items and finally reload I the core and restack the shielding. In all cases workers and work were con-trolled using Radiation Work Permits to ensure adequate monitoring of whole body as well as extremity doses using dosimters as well as badges with redun-I dant dosimeters and TLD badges to assure adequate records. UFTR-associated personnel exposures significantly greater than minimum detectable during the reporting period are summarized in the following two tables. .

Table IV lists monthly permanent badge exposures recorded above back-ground for personnel employed directly at the UFTR. Table V lists results of ring and other specialized badges and dosimeters utilized to record dose, es-pecially to the extremities, durino the lengthy process of correcting the s ticking control blade problem. Table VI lists the exposures for University of Florida personnel employed by the Radiation Control office where the exposure is attributed to radiation control work during the repair work to correct the s ticking control blade.

I For visitors , s tudents, or other non-permanent UPTR personnel, no indivi-dual had a non-zero dosimeter exposure measurement above 10% allowable for this reporting period. In mos t cases , the values of one up to five mrem re-corded dosimeter exposures are probably due to uncertainty in reading the de-vices. It should be noted that tours of reactor facilities around the core and other work areas were s trictly limited during the extended period af ter core shielding uns tacking on September 25, 1985 up to core shielding res tacking on February 7, 1986. There fore, the lack of significant visitor exposure is ex-pected and in agreement with ALARA guidelines.

I VII-3

TABLE III CUMMULATIVE RESULTS OF ENVIRONMENTAL MONITORING FOR THE 1985-1986 REPORTING YEAR Film Badge Total Yearly Total Yearly Designation Exposure (mrem)EIl TLDs{2] Exposure (mrem)IIl A1 10 1 M A2 320[3] 2 320[3]

A3 M 3 M A4 M 4 M A5 10 5 M

-lE A6 A7 M

M 6

7 M

M 8 M 9 M 10 M 11 M 12 M Note 1: M denotes mini aal (<10 mrem) meaning background only.

Note 2: The first seven TLDs are attached adjacent to the corresponding num-bered film badge monitors.

Note 3: The high exposure on Film Badge #A2 and TLD #2 was due to removal of reactor shielding from September, 1985 to January,1986 necessitated by maintenance efforts to repair sticking control blades. Film #A2 and TLD #2 are located on the reactor exhaust stack. The .high expo-sures on Film Badge #A2 and TLD #2 were expected, and proper boundary '

controls were set up by the Radiation Control Office. In addition, the higher monthly values were reported to the Reactor Safety Review Subcommittee end evaluated as expected under properly controlled areas to assure restricted access to the areas involved on the reac-I tor building roof.

I I

I VII-4

w-l TABLE IV PERMANENT BADGE EXPOSURE REPORTED ABOVE BACKGROUND l I Oc tober, 1985 l

C.J. S tiehl 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 G.W. Fogle 60/60 deep /whole body C.J. Stiehl 20/20 deep /whole body P.M. Whaley 250/250 deep /whole body January, 1986 G.W. Fogle 20/20 deep /whole body II. Gogun 250/250 deep /whole body R.K. Hanson 130/130 deep /whole body C.J. S tiehl 440/440 deep /whole body W.G. Vernetson 100/100 deep /whole body P.M. Whaley 430/430 deep /whole body February, 1986 G.W. Fogle 80/80 deep /whole body P.M. Whaley 20/20 deep /whole' body May, 1986 W.G. Vernetson 130/130 deep /whole body P.M. Whaley 20/20 deep /whole body July, 1986 P.M. Whaley 20/20 deep /whole body August, 1986 R.K. Hanson 10/10 deep /whole body C.J. S tiehl 50/50 deep /whole body P.M. Whaley 30/30 deep /whole body

!I O G I VII-5

g_

l l

I TABLE V-A l RADIATION EXPOSURE

SUMMARY

(I .

' l SAFETY-THREE MAINTENANCE PROJECT OCTOBER, 1985 - FEBRUARY, 1986 OCTOBER Whole Body Ankle Wris t Forehead Fingers I

S tiehl Personal Dosime tar 196 NR(2) 237 NR NR TLD 136 NR 227 87 RH-270(3)

LH-212 Whaley Personal Dosime ter 65 NR 70 NR NR TLD 49 NR 60 13 RH-70 '

LH-70 I4)

NOVEMBER Whole Body Ankle Wrist Forehead Fingers S tiehl I Personal Dosime ter 521 660 480 293 NR I Whaley TLD 302 623 445 224 RH-601 Personal Dosime ter 426 610 574 313 NR I TLD 206 453 395 222 RH-499 LH-679 NOTE 1: Dose in millirem.

NOTE 2: NR means dose measurement was not required.

NOTE 3: RH = right hand; LH = lef t hand.

I NOTE 4: Estimated convervatively at 70 mrem, TLD Chip lost from ring badge during anti-C clothing removal.

I VII-6

v-TABLE V-A'(CONTINUED)

RADIATION EXPOSURE

SUMMARY

.Il i

DECEMBER l Whole Body Ankle Wrist Forehead Fingers S tiehl Personal Dosime ter 200 109 58 32 NR(2)

TLD 183 80 37 27 RH-37( }

LH-27 Whaley I Personal Dosime ter 170 485 195 119 NR TLD 85 264 1 21 77 RH-249 LH-185 Fogle Personal Dosime ter 60 125 68 30 NR TLD 35 72 53 20 RH-60 LH-70 I NOTE 1: Dose in millirem.

NOTE 2: NR means dose measurement was not required.

NOTE 3: RH = right hand; LH = lef t hand.

I I

I I

I I VII--7

W TABLE V-A (CONTINUED)

RADIATION EXPOSURE

SUMMARY

(l}

JANUARY Whole Body Ankle Wrist Forehead Fingers Stiehl Personal Dosime ter 590 1143 654 383 NR(2)

TLD '347 798 384 248 RH-485 LH-496 Whaley I Personal Dosimeter 518 800 505 352 NR TLD 304 499 333 212 RH-470 LH-376 Hanson Personal Dosimeter 150 60 154 72 NR TLD 76 42 86 50 RH-99 LH-185 Gogun Personal Dosime ter 311 632 308 195 NR TLD 316 385 189 136 RH-257 Verne tson Personal Dosime ter 138 290 400 110 NR TLD 74 172 192 81 RH-286 LH-192 NOTE 1: Dose in millirem.

NOTE 2: NR means not required.

I I

I VII-8 _ _ _.

I I

TABLE V-A (CONTINUED) {

RADIATION EXPOSURE

SUMMARY

II}

l FEBRUARY '

Whole Body Ankle Wrist Forehead Fingers i S tiehl -

1 Personal Dosimeter 41 24 47 22 NR l

TLD 18 5 37 34 NR l Whaley Personal Dosime ter 21 NR NR 4 NR TLD 14 NR NR 5 NR NOTE 1: Dose in millirem.

NOTE 2: NR means not required.

I I

I I

I I

I I

I VII-9

v TABLE V-B RADIATION EXPOSURE TABULATED BY JOB (Personal Dosimeter Readings From RWP File - Whole Body Doses)

RWP No. Date Job Description Name (mrem) 86-9-I 2-12-86 Restack Core Shielding Whaley 7 Vernetson 3 S tiehl 0 Fogle 0 I 86-8-I 2-7-86 Modification to UF-32 Whaley S tiehl

-16 27 86-7-I 2-5-86 Refuel Reactor Whaley 8 S tiehl 6 Gogun 3 Vernetson 2 Fogle 77 I 86-6-I 1-31-86 Inspect Core Area, Place Neutron Sources Into Reactor Whaley Stiehl 8

0 I Verne tson 0 86-5-I 1-24-86 Restack Center Island Graphite S tiehl 166 and Remaining Graphite Whaley 1 31 I 86-4-I 1-21-86 Rework South-side (East and West) Fuel Boxes S tiehl Whaley 34 10 86-3-I 1-14-86 Connect Fuel Boxes South-side Stiehl '120 Whaley 11 Gogun 15 86-2-I 1-6-86 Remove / Replace South Side S tiehl 260 I Blade Assemblies Whaley Gogun Verne tson 265 270 138 I 86-1-I 1-3-86 Replace North Side Graphite, Whaley 90 Unstack South Side Graphite, Gogun I

12 Remove South Side Fuel Boxes I VII-10

v.-

I TABLE V-B (CONTINUED)

RADIATION EXPOSURE TABULATED BY JOB (Personal Dosimeter Readings From RWP File - Whole. Body Doses)

RWP No. Date Job Description Name (mrem) 85-23-I 12-13-85 Align RB, S-3 Shaf t and Bushings S tiehl ~300 IlI Whaley 266 Fogle 60 I 85-22-I 12-8-85 Package Radioact'ive Waste for S tiehl 32 Shipment 85-21 -I 11-29-85 Replace RB, S-3; North-Side S tiehl 170 Graphite, Begin Replacement Whaley 175 S-1, S-2 Bushings 85-20-I 11-21 South-Side Reactor Hardware Re- S tiehl 2 moval; Shaft Coupling Replace- Whaley 101 ments; R.B. View Port 85-19-I 11-13-85 Reg. Blade Removal, N.E. Fuel S tiehl 132 Box Dismount Whaley 7

'I 85-18-I 11-5-85 S-3 Control Blade Assembly, Graphite Island, Intra-Shroud Graphite, NW and NC Fuel Boxes S tiehl Whaley 181 1 41 85-17-I 10-30-85 Graphite Reflector Removal S tiehl 196 Whaley 65 85-16-I 10-25-85 Modify S-3 Shroud (Inspection S tiehl 39 Port) Whaley 0 85-14-I 10-15-85 N/A I

Remove Final Layer of Shielding, S tiehl First Graphite Layer Whaley 1 I 85-13-I 10-11-85 Unload Core to Spent Fuel Pits S tiehl Whaley 2

10 NOTE 1: Off Scale Dosimeter, TLD and Film Badge Readings Taken (estimated l

-" reading - supported by later film badge results presented in Table IV). l I VII-11 l i

Y TABLE V-B (CONTINUED)

RADIATION EXPOSURE TABULATED BY JOB (Personal Dosimeter Readings From RWP File - Whole Body Doses)

RWP No. Da te Job Description Name (mrem) 85-12-I 10-10-85 Consolidate Spent Fuel Pit Stiehl No Dose con tents Whaley No Dose 85-11-I 10-7-85 Visual, Remote Inspections and S tiehl 9 Top-Deck Work Whaley 3 I 85-10-I 9-30-85 Visual, Remote Inspections and S tiehl 17 Reactor Top-Deck Work Whaley 8 85-9-I 9-25-85 Unstack Core Shielding S tiehl 19 Whaley 6 85-8-I 9-11-85 Remove S-3 Shaft S tiehl No Dose Whaley No Dose I

I I

,I I VII-12

w 1 I 1 I TABLE VI l

PERMANENT BADGE EXPOSURE REPORTED ABOVE BACKGROUND I i

FOR RADIATION CONTROL PERSONNEL

• I Name Ti tle Dose Donald L. Munroe Rsdiation Control Officer 110 mrem Harvey G. Norton Radiation Control Technician 10 mrem

I Dale E. Perkins Radiation Control Technician 100 mrem Katherine Deschesne Nuclear Technician 10 mrem Eric Schmidt Nuclear Technician 50 mrem I Scott Armistead Nuclear Technician 30 mrem I Rod Hagen Nuclear Technician 60 mrem

• NOTE: These personnel are all employed through the Radiation Control Office and are not regularly involved in work at the University of Florida Training Reactor for routine radiation control work which is performed by UFTR Staff personnel qualified in radiation control.

I I

VII-13

I VIII. EDUCATION, RESEARCH AND TRAINING UTILIZATION NCrrE: The participating students are indicated with an *. Other participants are faculty or staff members of the University of Florida, unless specifically designated otherwise. A ** indicates those students work-ing on theses or dissertations.

Control Blade Drive Repairs - Major Overhaul of All Four UFTR Control Blade Drive Sys tems - Dr. W.G. Vernetson, P.M. Whaley, Reactor Staff. l This major repair operation to restore free movement of all four control '

blades was utilized in many areas to familiarize students with how to set up and work in high radiation fields, how to reload and conduct the necessary tests to assure system operability and to conduct many other practical outage-and restart-related reactor operations exercises.

Coolant Radioactivity Inventory Research - Development of a Model for the Ex-pected UPTR Primary Coolant Fission Product Inventory - Dr. W.E. Bolch, D r.

W.G. Verne tson, R. Knecht**, UFTR S ta ff.

Detectable but low levels of fission products in UFTR primary coolant were at-tributed to various sources of tramp uranium. To provide baseline data, sam-I ples were obtained prior to reactor running and following reference power runs. Due to the administrative shutdown and lack of reactor operation, sample taking was interrupted for six months. However, with draining of the primary system and refilling with demineralized water, periodic samples were analyzed af ter each significant power run. A good analytical model has been developed to predict the very low level fission product inventory in the UFTR coolant.

Some water samples were also irradiated to check analytical model with experi-men ts. The results of this work provide quantitative support for assuring the integrity of the UFTR fuel and the existence of low levels of fission products due to tramp uranium.

Cerenkov Noise Detector Development - Development of a Detector of Reactor Core Perturbations - Dr. E.E. Carroll, Prof. G.J. Schoessow, H. Carvajal**,

C. Levy *, N. Yunessi*, D. Lin*, Reactor S taff.

A new design Cerenkov detector is being developed and tested using the prompt-I gamma radiation deriving from the reactor core. The detector is being located in the thermal column entrance port with shielding plugs removed and substi-tuted by lithiated paraffin plugs made for the purpose of reducing the neutren I flux to acceptable values when the reactor is running at power. Samples of these lithiated paraffin plugs were irradiated to assure that no unexpected activation products would be formed were the plugs to see a large flux. O ther work has involved spectroscopic analysis of the gamma energies emitted from

~

the thennal column where the detector will be placed. The Cerenkov detector l 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 pertur-I bations 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 encourac.ing. This project I awaits student interest before further work is undertaken.

VIII-1

v-Radiation Protection Training - Reactor Operations Based Radiation Protection Health Physics Cooperative Work Training Program, D r. W.G. Verne tson , G.

Stephenson (CFCC) , R. Rawls (CFCC) , Reactor S taff.

A set of reactor operations based. radiation protection health physics coopera-tive work training exercises h' ave been developed to meet the cooperative work needs of Radiation Protection Technology students at Central Florida Community College (CFCC). Two of these courses were conducted during this reporting year with great success. Students who take these courses are well suited to work as radiation control technicians and health physics assistants at nuclear power plants. The exercises are also extremely adaptable and some of them have been upgraded and used in the graduate health physics laboratory course at the Uni-versity of Florida. The developm'ent of this course and its subsequent presen-tation to CFCC students has been partially supported under the UPTR DOE Reac-tor Sharing Program and has been a valuable resource in the effort to increase reactor utilization.

Reactor Operations Laboratory (ENU-4905/6937L) - Dr. Verne tson, Reactor Staff.

Students in the Reactor Operations Lab spent about two (2) hours weekly at the I controls of the UPTR performing reactor operations under supervision of li-censed reactor operators. The lab encompasses training in reactivity manipu-la tions , reactor checkouts, operating procedures, standard and abnormal opera-I tions and all applicable regulations. Specific exercises directed toward de-7elopment of understanding of light water power reactor behavior are included as this laboratory course serves as basic preparation for students entering the utility industry in the test and startup as well as plant operations a-I recs. When this course is not interrupted by outages, s tudents perform a

. serica of exercises designed to assure them of conducting 10 startups and 10 shu tdowns. A special effort is made to correlate UFTR exercises with various I aspects of LWR operations as was done with the extended outage during this reporting year.

I UPTR Task Analysis Survey - Dr. John Randall (General Physics, Inc. ) , Dr. W.G.

Verne tson , UPTR S ta ff.

A preliminary task analysis was completed for the UPTR. Staff interviews and I subsequent reviews of the task lists were utilized to develop final task lists for overall documentation and control of reactor operator training at the UPTR that is generic but easily amended for implementation in. future training cycles for new operators.

Service To Florida Foundation of Future Scientists - Lectures, Tours and De-mons tra tions of Reactor Ope ations - Dr. B. Abbott, Dr. W.G. Vernetaon, UPTR S ta f f. 4 A series of lectures, tout. and demonstra tions of reactor operations and nu-clear facility capabilities are conducted for a large number of student and faculty participants in the annual Junior Science, Engineering and Humanities Symposium jointly sponsored each winter by the Florida Foundation of Future I Scientists and the University of Florida for promising high school juniors and their teachers.

11 I VIII-2

NAA Research - Neutron Activation Analysis of Seagrass Community Components -

-I Dri G. Chiu (UWF), Dr. Ranga Rao (UWF), Dr. W.G. Verne tson , D. Morton* (UWF),

L. Hung *, S. Kahook*, L. Parla tore * , M. Leetzow* , Reactor S taff.

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 irradia tions to monitor the uptake of certain heavy metals, principally barium and chromium, both of which are suit-able for detection using neutron activation analysis. Reactor time for this work is supported under the DOE Reactor Sharing Program. Results to date are encouraging and work is continuing with only some rerun samples remaining to be analyzed to conclude this project.

NAA Research - Neutron Activation Analysis of a Seagrass Bed Exposed to Dril-ling Fluids, Dr. C.N. D'Asaro (UWF), Dr. T. Duke (UWF), R. Montgomery ** (UWF),

S. Macauley* (UWF), D. Morton* (UWF), S. Kahook* , L. Hung * , L. Parla tore * , M.

I Lee tzow* , Reac tor S ta f f.

This project involves moving cores from a seagrass bed to the laboratory where I they are exposed to various drilling fluids to determine possible effects on seagrass community structure and biomass. Barium and chromium are present in the drilling fluids and are known to impact negatively on animals and plants.

I However, knowing the correct concentrations of these metals is critical in or-der to correlate observed effects with metal concentrations to explain the phenomena involved. Use of the UFTR facility for the irradiation and subse-quent NAA provides an effective means of performing the chemical analyses.

Reactor time for this work is supported under the DOE Reactor Sharing Program.

A proposal to the Environmental Protection Agency may provide outside support for this project during the next reporting year as an outgrowth of the DOE Reactor Sharing Program.

NAA Research - Neutron Activation Analysis of Archeological Seashells - Dr. T.

S tocker (UWF) , Dr. W.G. Verne tson (UF), S. Kahook* , L. Hung * , L. Parla tore * ,

M. Lee tzow* .

I Under the Reactor Sharing Program, neutron activation analysis is being ap-plied 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 iden-tifying enough trace elements constituents in these seashells to develop a method for determining Indian trade routes in the Eastern United States. This research is in its early s tages.

I I

I I VIII-3

[

NAA Research - Neutron Activation Analysis of Estuary Sediments. - Dr. R. Byrne

( USF-S t. Pe tersburg) , Dr. G. Smith (USF-St. Petersburg) , S. Kahook*, L. Hung *,

L. Parla tore * , M. Leetzow* .

Under the DOE Reactor Sharing Grant, Instrumental Neutron Activation Analysis

(_ (INAA) is being applied to estuary su3iments from the Tampa Bay region of Florida to determine and quantify the spatial distribution of various rare ^

earth metals. Work to date has included preparatory work, mapping the spatial variation of ~ the flux in the UFTR vertical ports and another exercise to de-

{. termine accurate values for the cadmium ratios for ports to be used 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 many rare earth metals. The NAA work on this project has begun as sample preparations are now completede Virgin teflon tube sample holders have been demonstrated to withstand extended reactor- runs and have

( been analyzed for impurity content using NAA; several samples are currently

under analysis using INAA.

UPTR Reactor Operations and NAA Lab Exercises - Dr. W.G. Verne tson, G.

[- Stephenson/R. Rawls (CFCC), Dr. M. Lombardi/D. Pricks (HCC), Dr. S. Marchionno (SFCC), Dr. S. Polley (Stetson) , P.M. Whaley, S. Kahook* , L. Parlatore* , Reac-tor S ta f f.

[

Mini-courses- (including lectures, tours, demonstrations , reactor operations, l NAA of unknown and standard samples, e tc.) have been developed and presented I

( as part of the UFTR DOE Reactor Sha' ring Program to provide practical reactor operations and health physics training as well as NAA laboratory experience for groups of students from Central Florida Community College Radiation Pro- j tection Technology program, Santa Fe Community College Nuclear Medicine Tech-

[ nology/ Radiologic programs and the Hillsborough Community College Nuclear Me-

}

dicine/ Allied Health Technology programs and Stetson University's Society of {

Physics Students. Other participants in all or part of such mini-courses this

[ year include several Buchholz High School Science Classes, Li'ncoln Middle School and St Augustine High School. l

( NAA Research - Neutron Activation Analysis of Volcanic Rock Samples - Dr. M.

DeFant (USF-Tampa), Dr. W.G. Vernetson, S. Kahook* , L. Parlatore* , L. Hung *,

M. Lee tzow* , UFTR S taf f.

Under the DOE Reactor Sharing Program Neutron Activaiton Analysis is being ap-plied to various volcanic rock samples from widely dispersed geographic loca-tions ranging from Central America to both North and South America. The re-

[ search is directed toward trace element identification of rare earth nuclides in the samples. Eventually information on geologic origins of such samples is expected as this project continues.

[ Nuclear Chemistry Laboratory - Irradiation to Produce Nuclides for Half-Life Experiments - Dr. S. Grossman (USF-Tampa) , Dr. W.G. Vernetson, UFTR S taff.

Under the DOE Reactor Sharing Program, various pure nuclides have been irra-diated to provide radioactive sources for use in nuclear chemistry laboratory experiments directed toward measuring and analyzing half-life properties of

[ radioactive nuclides. This activity is expected to continue on an annual cycle.

[

VIII-4

{

v NAA Research - Rabbit System Remote Handling Facility Development and Imple-menta tion - Dr. G.S. Roessler, D r. W.G. Vernetson, Reactor Staff.

Radiation and contamination surveys are performed in the radiochemistry labo-ratory where the new NAA Instrumentation and Counting Facility has been uti-lized. Periodic checkouts are conducted of the " Rabbit" facility to assure ef-ficient rapid transfer 'for remote sample insertion and removal from the UPTR core region especially when new rabbit capsules are first utilized. To handle the sample volume more efficiently for Neutron Activation Analysis, efforts are currently directed to obtaining and implementing a better computer-based analyzer system. In addition, failure of the rabbit system sample delivery ,

system during this year has resulted in a redesigned improved system for which considerable testing has been completed. Implementation of this improved sys-tem is expected early in the next reporting year.

Nuclear Engineering Laboratory I - (ENU-4505L) - Dr. G.R. Dalton, Dr. W.H.

Ellis, Reactor S taf f.

ENU-4505L is the nuclear engineering laboratory for undergraduate senior level students in Nuclear Engineering Sciences. The UFTR is used for a variety of exercises and experiments, including NAA exercises, radiation dose measure-ments, measurement of induced radioactivity and reactor physics parameters as well as operational measurements. During this year the students also partici-I pated in analyzing and monitoring the reloading of the UFTR core following completion of repair work as well as many of the tests and surveillances re-quired following core reloading.

UFTR Operator Training and Requalification - Dr. W.G. Vernetson, Reactor S ta f f.

Lectures and hands-on operations on the reactor are necessary to license oper-ators for the UFTR. The requalification program establishes a required number of startups, weekly checks, daily checks, drills, practical exercises and lec-I tures for each operator. Operator participation is mandatory in order to main-tain assurance of proficiency levels and to be abic to demonstrate the requi ~

site operator skills. Operation proficiency is assured by written and oral I tests as well as observed practice exercises. Extensive fuel handling training as well as recertification training for all operators following the extended outage for the control blade failure resulted in more than normal training during the last reporting year. The same program in an accelerated mode is I used to train UFTR reactor operator license candidates.

I I

I

'I I VIII-5

v-I Reactor Operations Demonstrations - Reactor Operations Instruction and Demon-

• strations for Various Courses within The University of Florida - Dr. W.G.

Verne tson , Reactor Staff.

The following courses are identified where one or in some cases as many as four or five classes or labs in a course would be conducted using the UPTR facility. All would begin with the lecture, tour and reactor operations and facility capabilities demonstra tion with later classes, where needed, devoted to more detailed lab instruction in one or more areas of UFTR facility opera-tions. Courses include:

Course Ins true tor

.l

.5' ANT-6128 ENU-4103 Dr.

Dr.

B. Purdy G.R. Dalton ENV-4201/5206 Dr. C.E. Roessler I ENV-4241 ENU-5005 CMS-5110L Dr.

Dr.

Dr.

C.E. Roessler R. Pagano K. Williams ENV-6211 Dr. W.E. Bolch I ENV-6211L PHY-2001 Dr.

Dr.

W.E. Properzio/W.E. Bolch K. Eoff ENU-6646/ENV-6905 Dr. W.E. Bolch/W.E. Properzio NAA Research - Neutron Activation Analysis of Egyptian Phosphorites - Dr. M.

.El Haddad, Prof. J.S. Tulenko, Dr. W.G. Vernetson, Dr. G. S. Roessler.

Neut-on activation analysis has been applied to analyzing the rare earth ele-mental composition of various Egyptian phosphorites' from several locations.

The results obtained with the Egyptan phosphorites are being evaluated and I compared with the rare earth elemental composition of phosphate ores in va-rious locations in Europe and the United States to detemine the advisability of economic separation of rare earth elements from the Egyptian phosphorites.

Reactor Facility Research - Absolute Flux and Fast / Thermal Flux Ratio Deter-minations - Dr. W.H. Ellis, Dr. W.G. Vern'e tson , P.M. Whaley, L. Parlatore**,

W. Choi, UPTR S ta f f.

An extensive effort has been undertaken to measure and validate the thermal flux profiles, the absolute thermal and fast flux and various fast / thermal flux ratios in all the UFTR experimental ports. Ports completed include the thermal column for several configurations, all vertical ports including the special CVP graphite multiple sample holder as well as several positions on I the South and other beam ports. This work will be continued to detemine these operational, experimentally-necessary parameters for all experimental ports including the shield tank. I NAA Research - Trace Element Evaluation of Seashells - Dr. Guy Prentice (UF) ,

D r. G.S. Roessler, S. Kahook*, L. Hung, L. Parla tore , M. Lee tzow, UPTR S ta f f.

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 llm (nuclides) can be identified ac signatures for shells from various loca tions.

The work continues as its purpose is being reevaluated as the work progresses.

I VIII-6 i

v I

NAA Research - Neutron Activation Analysis of Brain Tissue - Dr. H. Van Rins-velt, Dr. G.S. Roessler, A. Kinyua, P.M. Whaley, UFTR S ta f f.

I Neutron activation analysis is being applied to identify trace elements in brain tissue samples as a complementary method with previous results obtained using PIXE analysis. Depending upon the extra information obtained in several preliminary NAA analyses , a larger ef fort may be forthcoming to perform a com-plete evaluation of the complementary features of NAA versus PIXE analysis for trace element identification in brain tissues.

Source Production - Act!vation of a Thullium Oxide Beta Source - Dr. G. Chop-pin (FSU), J. Rink * , UFTR S taff.

Under the DOE Reactor Sharing Program with some other support, a pure thullium oxide (Tm2 0 3) source was irradiated to produce a relatively long-lived beta source to support various experimental nuclear chemistry research experiments.

NAA Research - Neutron Activation Analysis of Purified Quartz - Dr.

~

A.L. Odum (FSU) , J. Rink * , Dr. W.G. Verne tson.

Neutron activation analysis was performed on several purified quartz samples in anticipation of later irradiations to examine microscopic structural ef-fects of irradiations on such samples to various fluences. The research to this point is to identify the trace element composition of the quartz to pre-dict radiation control on later samples irradiated to various larger fluences.

Radiation Protection and Control Health Physics Practice - (ENU-6646/ENV-6205)

- Dr. W.E. Bolch, Dr. W.E. Properzio, Dr. W.G. Verne tson , D. Munroe, H. Nor-ton, T. Hines, Reactor Staff. '

This course provides students in various disciplines with practical experience in radiation protection and control such as performing radiation surveys in and around the UPTR cell and environs, calibrating area radiation monitors, determining effluent levels, setting up emergency exercises, etc. These exer-cises also serve as training for radiation control technicians.

I NAA Research - Neutron Activation Analysis of Various NBS Standards - Dr.

Vernetson, P.M. Whaley, L. Parlatore*, T. Lively **.

W.G.

various NBS standard reference source samples in various dilutions are being I irradiated for neutron activation analysis to determine the NAA lower limit of detection for the various standards. This work formed the basis for training a high school student in research methods under the 1986 Florida Foundation of Future Scientists Summer High School Student Research Program. Limited results have been obtained to date though a good report was prepared by the student.

I I

I I VIII-7

v -

Reactor Characteristics Determination - Determination of Thermal and Wigner Energy S torage in Grapaite - Dr. W.G. Verne tson , P.M. Whaley, D. Roberts ,

Reactor S taf f.

A graphite temperature monitoring system had been developed to measure the temperature rise at vairous locations in the UPTR graphite during extended I power operations. Work concluded to date' includes design and development of the temperature monitoring system as well as the labyrinth shielding required to allow electrical leads to run from thermecouples in high radiation areas to recording devices outside the reactor. Results to date and future results will be used in ovaluating the storage of Wigner energey in.UFTR graphite (response to NRC Inspection Report 50-83/86-01) as well as to support comments on the NRC petition for ru' iemaking relative to Wigner erargy storage and graphite fire potential. This work formed the basis for training a high school student in research methods under the 1986 Florida Foundation of Future Scientists Summer High School Student Research Program.

Nuclear Chemistry Laboratory (CHS-5110L) - Neutron Activation Analysis of Iodine Labeled Antibodies (Labeled With S table Iodine) - Dr. K. Williams, G.

Sanchez, Reactor Staff.

This project was a study of the feasibility of using a stable iostope tracer (I-127) in antibodies and determining the distribution of the traced material via neutron activation analysis to evaluate the amount of the activated tracer (I-128). Suspect quality assureance in preparation of the antibodies rendered the results of this experiment inconclusive to date.

Neutron Radiography Facility Development - Flux Measurements and Thermal Column Alterations to Optimize Neutron Radiography Capabilities - Dr. A.M.

Jacobs, Dr. W.G. Verne tson , P.M. Whaley, Reactor S taff.

Extensive thermal neutron flux measurements and thermal neutron flux to gamma dose level measurements have been made for various thermal column configura-tions directed toward neutron radiography. Implementation of several designs with a borrowed radiograph plate indicates excellent small cross section ra-diographs can be obtained with good depth of field; however, checks to deter-

-lW mine possibility of producing real time radiographs in several configurations will be conducted in the near future. This developmental project is ongoing and a major enterprise for utilizing staff time and design efforts in the next reporting year. -

Basic Physics Research - Development of Pulsed Ionization Chamber Plasma Kine-tics Diagnos tics Capabili ties - Dr. W.H. Ellis , D r. E.T. Dugan, W. Choi*, M.

A lam * .

Experimental measurements are being made with several pulsed ionization cham-E ber ae i 9as to detenaine pie ma xinetic propertiea inctuains first eaa secona E order recombination coefficients as well as ion number densities in a fission-ing plasma. Work to date has been confined to helium plasmas. This work is ongoing as part of the Innovative Nuclear Space Power Institute research ef-I forts in the Strategic Defense Initiative for supporting the development of space nuclear power sources.

I I VIII-8

um I

UFTR Risk Assessment - Dr. W.G. Vernetson, R. Grif fith*.

{

A preliminary probabilistic risk assessment of the University of Florida I Training Reactor has been conducted. This project has determined an estimate of the probability of occurrence of a set of postulated maximum credible UFTR acciden ts. The results will be used to show that the UFTR poses no significant )

I risk to the general population and environment around the UFTR ard has demon-strated proficiency in PRA analyses as additional PRA projects are undertaken.

Specifically, research is continuing to obtain better data for the maximum credible accidents and extend the methodology to examine risk associated with less serious but high probability UPTR-related accidents. This project is re-latively inactive at present awaiting further student interest.

NAA Research - Trace Element Analysis of Human Blood Serum and Bone Marrow Samples, Dr. G.S. Roessler, Dr. W.E. Bolch.

I Blood and serum samples have been analyzed for trace element concentrations from sick as well as healthy patients relative to Leukemia. Results have also been compared with standards. The objective is to correlate trace element con-centrations (high or low) with certain diseases. The initial project in this series has been completed and a proposal was submitted to support continuing work during the last reporting year; future studies in this area are planned with the level of effort dependent on response to the proposal.

UFTR Core Redesign (LEU Program) - Neutronics Analysis for UFTR Core Redesign

- Dr. E.T. Dugan, Dr. W.G. Vernetson, P.M. Whaley.

As part of the DOE Low Enriched Uranium Conversion Program, investigations have been performed on the UFTR to detemine the feasibility and desirability I of replac-ing the 93% enriched MTR plate type fuel with 4.8% enriched, cylin-drical SPERT fuel pins. For this redesign, the only permanent structural modi-fication is the insertion of new grid assemblies into existing fuel boxes. Ac-ceptable neutronic criteria (Possible keff range, maximum flux and degree of I undermodera tion) have been determined using industry-accepted, 4-group cross sections in one, two and three-dimensional diffusion theory calculations of k 9ff, flux profiles , power peaking factors and coefficients of reactivity.

I First order perturbation calculations have been used to determine key kinetic parame ters. Neutronic results to date indicate that the UFTR/SPERT core rede-sign can be accommodated to meet requisite neutronic criteria with an actual I increase in peak thermal flux levels which will be very useful for NAA and other research projects requiring high thermal flux levels. Since the HEU con-version regulation has been passed by NRC, the UFTR expects to receive support during FY-1987 to begin the conversion process beginning with a decision on I whether to go with SPERT or plate-type fuel.

I I

I WRRR-@

' UFTR Core Redesign (LEU Program) - Thermal-hydraulic Analysis for Core Rede-sign - Dr. E.T. Dugan, Dr. W.G. Vernetson, Dr. N.J. Diaz.

I As part of the DCE LEU Conversion Program', thermal-hydraulic analysis related to redesign of the UfTR core using SPERT fuel rods has been performed. Com-puter analysis has been undertaken to evaluate the UPTR/SPERT design for steady-atate conditions as well as transients arising in response to a step insertion of reactivity, a loss of coolant flow, and a loss-of-coolant acci-dent. Results to date indicate required safety margins and transient response conditions can be maintained with the UFTR/SPERT core design. Since the HEU I conversion has been mandated, the decision on whether to go with SPERT or plate fuel will be made in the near future with conversion e:.pected to begin in FY-1987.

UPTR Trancient Analysis - Implementation of DSNP Program Language to Analy2.e UFTR Operational Transients - Dr. E.T. Dugan, Dr. W.G. Verne tson , J. Samuels**.

The Dynamic Simulator for Nuclear Power (DSNP) Plants programming language is being implemented to analyze selected UPTR heat up and cooldown transients.

Results from DSNP calculations are being compared and evaluated relative' to existing and new transient UFTR output recorded on various output devices.

This analysis will serve as a teaching aid for the DSNP programming language and will hopefully allow fast-running analysis of UFTR transients for class exercises and other similar applications within the Nuclear Engineering Sciences Department.

I Gaseous Release Determinations - Argon-41 S tack Measurements - Dr.

Verne tson , Dr. W.E. Bolch, P.M. Whaley*, Reactor S taff.

W.G.

A cobalt-60 Standard Sample .has been applied in standardized controlled mea-I surements of radioactivity (Ar-41) in stack effluent. A direct detailed stan-dard operating procedure (UPTR-SOP-E.6: Argon-41 Concentration Measurement) has been developed and approved as the best practicable evaluation of Ar-41 I releases from the UFTR facility as required by UFTR Technical Specification on Effluents Surveillance in Section 4.2.4, Paragraph (2). Application of this SOP continues to obtain a statistically significant number of data points and I eventually to investigate the effect of variable mre vent flow on total Ar-41 releases. Other commitments during this report!Ag ' ear have limited progress on this project; with the expectation of e"W 213 raising power levels plus the decreased Ar-41 release limit in the 3rr , o Je- 0 CFR 20 revision, this work should be moved to a higher priority la th. .txt reporting year.

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I VIII-10 i

ns Reactor Operations - (ENU-5176L) - Dr. E.T. Dugan, Dr. W.G. Verne tson, Reactor S ta f f.

Students in the reactor operatiors course spend about two hours weekly at the controls of the UFTR performing reactor opera tions under supervision of li-censed reactor operators. The lab encompasses training in reactivity manipula-tions, reactor checkouts, operating procedures , s tandard'and abnormal opera-tions and all applicable regulations. Specific exercises directed toward de-velopment of understanding of light water 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. '

A special effort is made to correlate UPTR exercises with the classroom lec-tures on various aspects of LWR operations. This course was not offered during the current reporting year. However, work was concluded to have the laboratory part of this course approved as a separate stand-alone course to assure proper s tudent credit for this popular lab course in the future.

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IX. THESES, PUBLICATIONS, REPORTS AND ORAL PRESENTATIONS

.I OF WORK RELATED TO THE USE AND OPERATION OF THE UFTR

1. " Fall Semester Reactor Operations Laboratory Manual for ENU-4905/6937L,"

W.G. Vernetson, Sep tember, 1985.

I 2. " Annual Progress Report of the University of Florida Training Reactor for September 1, 1.984 November, 1985.

- August 31, 1985 Reporting Year," W.G. Verne tson,

3. " Final Report on the Spring Semester Reactor Operations-Based Health Phy-sics Cooperative Work Training Program," conducted for Radiation Protec-tion Technology Program Students at Central Florida Community College, W.G. Verne tson, December, 1985.

4. " Spring Semester Reactor Operations Laboratory Manual for ENU-4905/6937L,"

W.G. Verne tson, January, 1986.

5. " University of Florida Reactor Sharing Program," W.G. Verne tson, proposal submitted to Department of Energy, March, 1986.

6. " Reactor Operations-Based Radiation Protection Health Physics Laboratory Course," paper presented at the Health Physics Society Florida Section Spring Meeting held in Gainesville, Florida, April 4-5, 1986.

7. " Final Report on the Spring Semester Reactor Operations-Based Health Phy-I sics Cooperative Work Training Program," conducted for Radiation Protec--

tion Technology Program Students at Central Florida Community College, W.G. Verne tson, May, 1986.

8. " Summer Semester Reactor Operations Laboratory Manual for Edu-4905/6937L,"

W.G. Verne tson, May, 1986.

9. " Proposal to Extend Evaluation of Drilling Fluids on Seagrass Communities Uptake of Barium, Chromium and Lead," C. D'Asaro, R. Rao, e t. al. , propo-sal submitted to Environmental Protection Agency, June, 1986.*

10. " Determination of the LLD for NBS Standards," T. Lively, oral presentation on FFFS Summer Research Project, University of Florida, August 5,1986.

11. " Graphite Temperature Measurements in the University of Florida Training Reactor During Power Operation," D.M. Roberts, oral presentation on FFFS

Summer Research Project, University of Florida, August 5, 1986.

I i

12. " Stored Energy in Reactor Graphite," D.M. Rober ts , summer research project report submitted as a participant in Florida Foundation of Future Scien-I I tists 1986 Summer Research Program (prepared also for use as a High School Science Fair Project), Nuclear Facilities Division, University of Florida, August 5, 1986.

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13. " Determination of the Lower Limit of Dectection for NAA," T. Lively, sum-I mer research project report submitted as a participant in Florida Founda-tion of Future Scientists 1986 Summer Research Program (prepared also for use as a High School Science Fair Project), Nuclear Facilities Division, University of Florida, August 6, 1986.

14. " Determination of Parameters for a UFTR Primary Coolant Fission Product '

Activity Model," R. Knecht, Masters' Thesis Project in Environmental En-gineering Sciences Department, University of Florida, August, 1986.

I 15. " Physical Basis' of Heat and Radiation-Induced Color Changes in Topaz:

AL2 (SiO4 )(OH,F)2," J. Rink, Masters' Thesis in Physics Department, Florida State University, Augus t, 1986.*

16. " Nuclear Seeded MHD Plasma Diagnostic Experiments With PIC," to be in-cluded in University of Florida INSPI Quarterly Report, INSPI-UF-86-004, September, 1986.*

17. " Reactor-Related Research Projects Interests and Developments," W.G. Ver-netson, scheduled to be presented at joint NES/ANS seminar on September 18, 1986.

18. " Distribution of Rare Earth Elements In Egyptian Phosphorites," Mervet El Haddad, scheduled to be presented at joint NES/ANS seminar on September 19, 1986.

19. " Feasibility Study of UFTR Real Time Neutron Radiography Capabilities,"

J.D. Cox, Internal Report, General Imaging Corporation, in preparation for submittal for October, 1986.

20. " Revitalization of Reactor Usage Through Reactor Sharing," W.G. Verne tson, paper accepted for presentation at the Winter Meeting of the American Nu-clear Society to be held in Washington, D.C., November 16-20, 1986.

21. " Effects of Drilling Fluids on an Experimental Seagrass (Thalassia testudinum) Community: Potential for Bioaccummulaton of Barium and Chromium," Dana Morton, Masters' Thesis in Biology Department, University of West Florida, Pensacola, degree expected December,1986*.

22. " Nuclear Seeded MHD Plasma Diagnostics," W.H. Ellis, oral presentation planned for the 4th INSPI IWG Meeting to be held in Washington, D.C.,

November 14, 1986.*

23. " Trace Metals and Alzheimer's Disease," Antony Kinyua, Master's Thesis in I Nuclear Engineering Sciences Department, University of Florida, Gaines-ville, degree expected May,1987.*

I 24. " Implementation of DSNP (Dynamic Simulator of Nuclear Plants) and Applica-tion to the Analysis of Transients for the UFTR," J. Samuels, Mas ters' Thesis Project in Nuclear Engineering Sciences Department, University of Florida, degree expected May, 1987.

It is expected that the results of this work will be published in a jour-nal article at a future date.

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APPENDIX A

] UFTR FACILITY LICENSEE RESPONSE TO NRC INSPECTION REPORT NUMBER 50-83/86-01 J

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u m u . = 0. 4,. ct. NUCLEAR FACILITIES DIVISION gA uuctun nacron suitomo . y q G4petSVILLE,FLO4f04 32618 UNIVERSITY OF FLORIDA '. *

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e PH0fet (904)392-8429 TELEA$6330 .

, April.17, 1986 United States Nuclear Regulatory Commission Region II 101 Marietta Street, N.W.

Atlanta, GA 30323 -

Attention: Roger D. Walker Director, Division of Reactor Projects

• t Re: Inspection Report No. 50-83/86-01 (Inspector Paul Barnett)

Dear Mr.. Walker:

Inspection Report No. 50-83/86-01 cites our facility with a Severity Level IV

- violation for failure to maintain adequate records of the safety evaluation (50.59 Evaluation #85-17 dated'4 October 1985) made to support modification of the UFTR control blade shrouds which involved cutting of inspection ports in the shroud spacers.

_ Admission of the Alleged Violation Unreviewed Safety Question Evaluation / Determination Number 85-17, "UFTR Shroud I Spacer Cutting Alteration For Ease of Blade Inspection" is considered to have insufficient documentation that the protective (safety) function of the shrouds as delineated in Paragraph 4.1.1 of,the Safety Analysis Report and the I Technical Specifications in Paragraph 3.2.1 (1 ) had been adequately considered in concluding that the modification'did not involve an unreviewed safety ques-tion. Information provided by Mr. Whaley (Raactor Manager), Dr. Ohanian .

(Chairman, RSRS) and Dr. Vernetson (Facility Director) shows that the proper SAR and TS, references were considered as part of the 50.59 review, specifical-

' ly, the protective function of the shrouds was carefully considered at the RSRS meeting where 50.59. Evaluation / Determination Number 85-17 was approved.

I Nevertheless, we acknowledge not all of this information had been provided as part of the supporting documentation for this modification review. Therefore, the violation citing lack of complete documentation is admitted though it is clear the proper reviews were made, simply not fully documented. The UFTR staff and RSRS members all agree there is no unreviewed safety question in-volved in this modification; the violation addresses only the failure to do-

~

cument all items considered as part of the review - namely, that the protec .

tive ( safe ty) function of the shroud is not compromised.

Reasons For The Violation The failure to provide sufficient documentation of the shroud spacer modifica-tion is attributed to lack of sufficient detail about required documentation in the requirements of UFTR SOP-0.4, "10 CFR 50.59 Evaluation and Determina-tion" used to control and document reviews of potential unreviewed safety questions. This inadequacy led to insufficient notes from which the documenta-I tion of review was generated. As a result one part of the evaluation process was not adequately documented.

I Roger D. Walker April 17,1986 Page Two Corrective Steps Taken/Results Achieved The 50.59 Evaluation / Determination Number 85-17 was reviewed again at I!he first RSRS meeting on February 27, 1986 held following the NRC inspection which concluded on February 21, 1936. At that time the Subcommittee considered the Inspector's concerns expressed in the exit interview and reaffirmed its earlier finding that no unreviewed safety question was involved in the modifi-cation and documented that pai-t of- the basis for the decision is that the shrouds do' continue to fulfill their protective (safety) function.

In a memorandum dated March 31, 1986, a supplement to 50.59 Evaluation / Deter-mination No. 65-17 was generated. This supplementary documentation was re-I. viewed at the RSRS meeting on April' 17, 1986 and is considered to meet the 10 CFR 50.59(b) requirements that the licensee maintain records of changes made in the facility pursuant to 10 CFR 50.59(a).

Corrective Steps To Be Taken To Avoid Further Violations UFTR Management and the RSRS are committed to providing proper documentation of 50.59 evaluations / determinations. The incomplete, documentation resulting in I the cited violation is considered to be a relatively rare event considering

-the-dozen or more 50.59 reviews inspected during the February 18-21' inspec-

' tion. Nevertheless, corrective action is in progress to assure this lack of I documentation of items considered does not recur. In addition to the recon-sideration of Safety Review No. 85-17 by all parties and extensive discussions of the violation and its basis cited in Inspection Report No. 50-83/86-01, a revision of SOP-0.4 "10 CFR 50.59 Evaluation and Determination" is being de-veloped. This revision will include a requirement that a simple list of items considered in the 50.59 review be attached to each 50.59 review package to ,

document all items considered. Therefore, SOP-0.4 will be revised to require I that this list be~ generated for all 50.59 reviews calling specific attention to pertinent sections of the UFTR Safety Analysis Report, the Technical Speci-fications and/or Standard Operating Procedures to assure complete documenta-I tion of 50.59 reviews. ,

Date When Full Compliance Will Be Achieved The specific violation cited is lack of full documentation for 50.59 Evalua-tion / Determination Number 85-17. With the generation of the documentation dated March 3, 1986 as a supplement in the original 50.59 evaluation, the necessary documentation now exists indicating that the protective iunction of the shroud was considered in the safety review of the modification. As to the corrective steps to be taken to avoid further violations, the revision of SOP-0.4 will be generated, approved and implemented by June 15, 1986. This revi-f sion will be entirely adequate to assure this situation does not recur. There-fore, full compliance will be achieved by June 15, 1986, though the supplement.

to Evaluation / Determination Number 85-17 has already assured compliance in the one problem area.

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Roger D. Walker April 17,1986 Page Three Exit Interview Commitments As committed in the exit interview:

1.

SAR Paragraph 7.2.3.on Non-Nuclear Instrumentation is being rewritten and will be revised as necessary' by August 31, 1 986.

2. The CORA diffusion. theory calculations for UPTR core fluxes will be re-I, viewed for proper identification of material regions and the SAR revised as.necessary by Au, gust 31, 1986,

3. The storage and potential release of Wigner Engery is not addressed in the UFTR SAR. This isn 'ot considered to be of signficant concern for the fluxes in the UFTR graphite regions. Nevertheless, the UFTR management is committed to evaluating the storage and potential release of Wigner Ener-gy in this graphite by May 31, 1987.

We trust this response satisfies the' requirements delineated in the inspection

. report. If there are further questions, please adviso. -

I Sincerely,

,/6 William G. Vernetson Acting Director of Nuclear Facilities .

-WGV/ps cc: P.M. Whaley .

G.S. Roessler

-I Reactor Safety Review Subcommittee (RSRS)

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4 APPENDIX B I FINAL

SUMMARY

REPORT TO NRC Oli RECURRENCE OF STICKING S-3 CONTROL BLADE PROBLEM CP SEPTEMBER 3,1986:

NOTIFICATIONS, CORRECTIVE ACTION, PREVENTIVE MAINTENANCE, TESTS AFD SURVEILLANCES I ,

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s au. mo.

m.,_._.._.. NUCLEAR FACILITIES DIVISION NUCLEAR REACTOR BUILDING rh$t%

CAINESV8LL FLO.104 32618 PHON 4(504))93-8429 TELEX 5633e ,

UNIVERSITY OF FLORIDA g*

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x April 3,1986

- I- . Nuclear Regulatory CoLission /

Suite 2900 101 Marietta Street, N.W.

Atlanta, Georgia 30323

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. Attention: J. Nelson Grace '

~I, ' Regional Administrator, Region II

~

Re: University of Florida Training Reactor I- Facility License: R-56 Docket Number: 50-83 Gentlemen: -

Pursuant to the reporting requirements of paragraph 6.6.2(3)(c) of the UFTR Technical Specifications, a description of a potential abnormal occurrence as I. defined in the UFTR Technical Specifications, chapter 1 is described in this final report to include NRC notification, occurrence scenario, evaluation and general work and corrective action undertaken to correct the sticking blade as well as actions taken to prevent recurrence of this problem. The potential ab-normal occurrence involved the failure of one of the UFTR ' control blades (Safety Blade #3) to drop on demand from approximately a 31% withdrawn posi-I tion.

Note that the capability to shut the reactor down and maintain adequate shutdown margin was never lost.

-NRC NOTIFICATION The Executive Committee of the Reactor Safety Review Subcommittee reviewed t this occurrence on September 4, 1985 and concluded that it is a potential ab-normal occurrence, as defined in UFTR Technical Specifications, Chapter 1. The ,

RSRS then instructed NRC notification as per Section 6.6.2 of the UFTR Tech specs. This notification was carried out by both telephone and a following telecopy (Attachment I) on September 4, 1985. An interim report representing the 14 day followup report as required in UFTR Toch Specs, Paragraph 6.6.2 (3 ) (c) was submitted on Septobmer 17, 1985 (Attachment II). Additional I update reports were supplied on November 18 and December 10, 1985. The current final report is considered to close consideration of this occurrence since the problem leading to the occurrence has been shown to be fully resolved.

It should be noted that NRC Region II through Paul Frederickson and Inspector Keith Poertner have been kept fully updated as to the progress on the resolu-I tion of this problem through telecoms on September 4,16,18 and 25, October 4 and 19, December 2, January 10 and 23 and February 5 and 18. In addition, NRC Inspector Keith Poertner conducted an inspection related to the sticking safe-ty blade on October 9-11, 1985 and indicated satisfaction with efforts under-I way to correct the problem. Finally, on February 18-21, 1986, Inspector Paul )

Darnett conducted a general facility inspection concentrating on the corree- J tive maintenance for the sticking control blade and preventive actions taken to prevent recurrence with general satisfaction with action taken but with

• c- 80WAL 0*FOAfWeffv/a,Mau4ftvt ACN

I J. Nelson Grace

-lB April 3,1986 Page Two several specific concerns which have been addressed since the inspect.on. This report is a summary of key points of a more detailed report from the UFTR Reactor Manager included as Attachment III to this report for reference pur-I- poses.

-OCCURRENCE SCENARIO

. As indicated in the telephone conversation with Mr. Paul Frederickson of NRC, Region II, , and a following telecopy on 4 Septemer 1985 (Attachment I), the-

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LUniversity of Florida Training Reactor (UFTR) was found to have one of its ,

reactor control blades (Safety-3) which failed to drop on demand from a 64%

removed position. This failure (sticking about 31 % removed) was discovered by I a Reactor Operator as he commenced a power increase from the 1 watt critical position. The operator had accidentally raised the Safety-3 control blade in-stead of the Regulating Blade for this power increase; in returning it to the normal 640 unit position he felt the response was. sluggish and so he attempted

to drop the blade from 640 units withdrawn to check it. Following clutch cur-rent release, the blade, stopped at the 310 unit position and was subsequently driven in with the other three blades to shut the reactor down.

Immediate checks by the Acting Facility Director involving subsequent removal to various heights showed this sticking problem to be intermittant and to cen-

.E ..ter..in.the 290-315 unit range.but with some apparent sluggishness in the drop g from other higher and lower heights. This occurrence was noted to be essen-tially a recurrence of the event reported by our facility in a letter dated I January 28, 9,

1985 with subsequent followup in an interim report dated February 1985 and assumed closed out in a report dated March 26, 1985. Subsequent checks in the ensuing six weeks showed this stickirig phenomenon occurring over a relatively narrow range from about 25-354 withdrawn and not to occur on some I occasions though the problem continued to worsen as time progressed. However, despite the maintenance performed in response to the previous occurrence as summarized in the March 26, 1985 report, the problem discovered on September

.I 3, 1985 continued with drop times from full out eventually exceeding the 1 second maximum tech spec limit as maintenance on the control rod drive system I was undertaken external to the biological shield. It should be noted, however, I

that the capability to manually insert the Sr.fety-3 control Blade using the drive system was always available along with sufficient shutdown margin to as-sure the- reactor remained shut down.

EVALUATION The problem of the sticking S-3 control blade was . initially suspected to be I related to one or more of three previous maintenance projects conducted at the University of Florida Training Reactor (UFTR):

I 1.

2.

3.

Safety Blade Three Drive System Repair (February / March,1985).

Puel Inspection (January,1985) .

Thermocouple Repair (September, 1 984).

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m J. Nelson Grace April 3,1986 Page Three These projects involved peripheral' activities that it was thought could possi-bly have ' created inadequate clearance between the control blade and the con-I trol blade housing (shroud) by forcing the graphite stacked around the shroud into contact with the shroud, or misaligning the control blade shaft.

These projects also provided current information about the work environment in the reactor core region. The preferable course of action from a radiation ex-posure standpoint was to repair the problem with fuel iri the core if possible since fuel I '

~- exposurecurring oc. removal as part and replacement of the involves troubleshooting considerable and repair dose processes. In commitment trospect, the commitmsnt to seek a solution consistent with ALARA considera-re-tions significantly contributed ~ to delays in successful completion of this

.I latest maintenance program; however, this commitment aided evaluation of pro-posed actions at each step of the maintenance program from the standpoint of

• ' efficacy and dose commitment.

Staff' analysis indicated that a problem not severe enough to have previously prevented a successful scram insertion of S-3 had been developing for some time, perhaps initiated or exacerbated by maintenance in the core area. Other problems in tandem with the primary problem had previously created enough friction in the S-3 drive system to inhibit successful scrams; these other

. problems were corrected by the control. blade drive system maintenance per-

.I - . fonned in. February / March, 1985 on the S-3 drive system (and subsequently on the other three drive systems) external to the biological shield. The primary problem had developed to such a magnitude that it became sufficient to prevent full trip insertion for control blade Safety Three by September 3, 1985.

From the beginning, because of the recurring nature of this occurrence, the

.UFTR staff was committed to in-core inspections to establish control blade I clearances and clearly close out this problem. Although correspondence with similar research reactors produced .no probable root causes to the UFTR stick-ing control blade problem, the input from other Argonaut reactors provided ad-I ditional checkpoints to be examined during work progress and performance of preventive maintenance on the control blade driva systems. The degradation in 1

drop times at other facilities seemed to be due to the use of grease-lubri-cated bearings (which tend to de-polymerize in a radiation field); the UFTR control blade in-core bearings are constructed of graphite to avoid that spe-cific problem.

As initial investigations were undertaken, four (4) sources of the sticking problem were identified for investigation:

1. Blade drive system bearings or gears.

2. Blade shaft to shaft housing clearance.

3. Control blade to control blade housing clearance.

4. Blade shaft bearings.

Safety Three control blade drive unit gears and bearings at the exterior of E the bio ogical shielding were examined first. Although it was decided to re-g place one set of bearings in the unit because of somewhat rough operation, no

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J. Nelson Grace April 3,1986 Page Four I significant problems were detected clearly, 'the drive system external to the biological shield was not the source of the problem. The long-shaft and the I shaf t penetration were examined to determine if any clearance problems could be detected, but again examination indicated adequate clearances and no source of the problem.

At this point, prior to removal of biological shielding, the four original areas of investigation were effectively reduced to the two within the biologi-cal shielding: control blade / shroud clearances or blade-shaft bearings / bush-ings within the biological shielding. Tests and inspections clearly indicated that the problem source was centered in the reactor core area in agreement with earlier staff commitments to perform in-core inspections and maintenance actions-as necessary to correct the. problem.

GENERAL INSPECTION WORKPLAN AND CORRECTIVE ACTION I Specific test equipment and procedures were used to aid the S-3 inspection and repair program. Af ter the biological shielding was removed, a closed circuit television (CCTV) and a boroscope were used to examine the shaft, shaft pene-tration and bearing assembly of Safety Three and the Regulating Blade. No problems ccould be identified, but some graphite powder was noted under the control blade shaft on the bearing (bushing) support pedestal for control I. - blade Safety Three. One possible explanation for the presence of the material with significance to the safety Three problem was excessive wear or abrasion of the graphite bushing.

At this point, the control blade was manually manipulated subject to the con-ditions of a, special controlling procedure (reviewed at NRC prior to implemon- l tation) with fuel in place while visual inspection was performed to determine 1 if the control blade was experiencing any shifting along the axis of the shaft. A microphone was mounted on the control blade shroud 'to attempt detec-I tion of shroud and control blade contact. No further specific indication of the source of the problem was detected in'these inspections.

It was decided that the control blade shroud should be modified to permit vi-sual inspection of the control blade with the fuel in place. To this end, a UPTR modification to remove part of the magnesium shroud was reviewed and ap-proved as involving no unreviewed safety question and indicating the shroud I continued to maintain abili'ty to perform its safety function. Originally it was planned to examine the control blade with the CCTV camera during manipula-tions. Although approved for implementation, attempts to modify the shroud as I indicated were delayed awaiting fuel removal to reduce radiation fields to al-low sufficient worker time in the core area to perform the inspection and any necessary corrective maintenance. This decision was based on the fact that the S-3 bushing was suspected to be the source of the problem and corrective ac-I tion would clearly require long work residence time for removal of necessary components and ro-installation of the control bl'ade system with new bushings.

When fuel was unloaded on October 11, 1985, radiation Icvels originally mea-sured in the range of 8-15 R/hr at the bushing assemblies were significantly reduced to less than 1.5 R/hr in most areas. Some of this unloading operation f I .

I 1 J. Nelson Grace l

April 3,1986 l

Page Five I was observed by NRC inspector Keith Poertner'during the October 9-11., 1985 in-spection.

Though corrective act. ion was directed initially to restoring proper operation of the Safety Three control blade, all four control blade drive systems inter-I nal to the~ biological shield were subjected to the same maintenance work. For each of the four control blade drive systems, the following key tasks were accomplished following unstacking of biological shielding and complete fuel removal:

1. Disconnection of the right-angle gear box from the control blade shaft; only the gears and bearings of the safety-3 blade external to I the biological shielding were disassembled and checked with no sig-nificant problems noted showing the maintenance of February / March, 1985 to be effective.

2. Removal of sufficient moderator / reflector graphite to allow access to each control blade shroud and control blade.

3. Removal of a seven (7) inch section of shroud spacer material from the top of the shroud to provide for remote visual inspection of control blade integrity (this modification was performed out of core

.. . .. .for all shrouds except S-3).'

4. Removal of fuel boxes, control blade shrouds and control blades.

5. Removal of bush'ing-pillow block bolt retainers (simply locking nuts) involving considerable dose commitment.

6. Modification of bushing-pillow block bolt retainers for ease of re-mote bolt removal or replacement in the future for ALARA purposes. ,

7. Removal of original 1040 carbon steel control blade jack shaft coup-ling and graphite bushing. Note: it was the bushing improperly bound to the S-3 control blade that was determined to be the cause of the sticking problem.

8.

Replacement of control blade jack shaft (coupling) with a new coup-I ling made of Type-316 stainless steel detennined to represent no un-reviewed safety question.

9. Installation of a modified ex-core shaf t coupling modification to I allow use of stainless steel jack shaf t couplings due to an error in facility drawings as communicated to NRC, Region II.

10. Reinstallation of control blades and shrouds.

11. Replacement of graphite moderator material around control blade shrouds.

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l J. Nelson Grace April 3,1986 Page Six ,

W 12. Re-installation of all six fuel boxes.

I 13. Replacement of center island graphite af ter replacement of bottom spacer phenolic-like material with aluminum spacer plates of the same dimensions.

14. Leakage check and flushing of primary system. 9 15.

Resetting of control blade position indicating potentiometers to"ap-proximately zero for a low level blade position. ,

16. Resetting of all control blade limit switches.

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Reloading 'of fuel into core to include 21 irradiated bundles plus one-half bundle (5-plates) of unirradiated fuel to ' provide needed replenishment of excess reactivity. ,

18.

Restacking of remainder of graph 3,te and biological shielding.

TE:ST AND OIECKOUT PROGRAM At various points during the completion of the previously referenced tasks for I '

- all of the control blades, ~ a series of drop time tests were performed to as-sure correction of the problem of the sticking S-3 blade and proper operation of the blade drive control systems. The drop times of'all control blades were I initially measured by manually withdrawing each control blade until the top limit switch engaged to provide a top-end fall out signal; then the manual withdrawal device released so that the control blade fell to the bottom posi-I . tion where the bottom limit switch provided the full in signal. All control blades exhibited drop times of about one-third of a second during this mecha-nical test. This time' was expected to be somewhat less than the times that would be obtained when the disengagement of the magnetic clutch would be used .

to initiate the control blade drop. Results of subsequent drop time measure-ments for the on all fourfour following (4) control blades are tabulated in Table 1 of this report (4) conditions:

1. Without use of magnetic clutch (mechanical check).

I 2.

Prior to fuel loading, all drive assemblies functional (using magne-tic clutch).

3.

I Following completion of fuel loading to include replacement of the first layer of biological shielding.

4.

Following complete restacking of all permanent concrete biological shielding.

,g Drop times for all control blades for all checks including those following re-placement of all biological shielding were less than 0.5 secondo as summarized E in Table 1.

These times are essentially comparable to those drop times ex-f I ~

J. Nelson Grace April 3,1986 Page Seven pected to be measured when the system .was first installed. These results clearly demonstrate that the maintenance actions taken to correct the problem I of the sticking S-3 blade is effective and that the preventive maintenance performed on the other control blade drive systems is also effective,; drop times for all blade systems are restored to originally as-installed values.

I The number of measurements summarized in Table 1 also demonstrates the consis-tency of these results. Finally comparison with historical data for control blade drop times presented in Table 2 demonstraten that the sticking problem and any indications of it are removed and the problem fully resolved.

I RECOMMENDATION I

The UFTR Management presented the results of this work completed on February 27, 1986 to the RSRS for approval at its March 27, 1986 meeting to declare the

" sticking safety blade (S-3) occurrence as a closed issue with resumption of

- normal UFTR operations allowed cubject- to satisfactory completion of several checkouts and training items remaining to be completed as committed to NRC.

All modifications involving potential unreviewed safety questions were re .

~ viewed and evaluated and/or determined not to constitute an unreviewed safety question by facility management as well as the RSRS. Therefore, the RSRS ap-proved restart of the UPTR on March 27 subject to completion of the minor items delineated above. A memorandu'm from the Director of Nuclear Facilities I ,

. . officially removed the UFTR from Administrative Shutdown and allowed resump-tion of normal operations on March 3, 1986 following telephone communications with NRC on February 28, 1986 indicating completion of all tests with only some training remaining to be completed.

PREVENTIVE ACTION I The UPTR Technical Specifications Surveillance Requirements, Section 4.2.2 Pa-ragraph (4) states, "The mechanical integrity of the control blades and drive system shall be inspected during each incore inspection but shall be fully I checked at least once every 5 years." This requirement dates from UFTR reli-consing in 1982 and is considered sufficient to provide reasonable assurances that this binding in the gear / clutch system will not recur. This assurance is I further validated since graphite bushings are used and would be expected to last at least 20 years. The gear box and right angic bearing systems over-hauled in February / March, 1985 and the remainder of the control blade drive I systems overhauled as part of the current project have not previously under-gone such maintenance. Therefore, inspections at five year intervals along with careful tracking of results of controlled drop time tests are considered adequate and sufficient to assure this problem does not develop in the future.

CONSEQUENCES I As concluded by the Reactor Safety Review Subcommittee, this potential abnor-mal occurrence did not compromise the health and safety of the public. In ad-dition, with RSRS approval of this report conveyed to NRC, this problem is

, considered to be fully resolved.

I ~

F I ~

J. Nelson Grace April 3,1986 Page Eight I POSTSCRIPT

• I It should be noted that several surveillances which were not performed during the administrative shutdown beginning on September 3,1985 following discovery of the sticking blade. In addition, various other checks and surveillances re-lated to unloading 'and reloading fuel were required to be completed before I. _

normal operations i:ould resume. Finally, because of the extended _ shutdown, the

. UFTR operating staff was required to be recertified to perform licensed duties such as reloading fuel, operating the UFTR for testing, etc.~A plan to re'cer-

-tify the operators was submitted to C.A. Julian at NRC, Region II, in a lotter dated January 6, 1986 with an attachment delineating the work plan of checks and surveillances required to be completed prior to. resumption of normal. op-

' _erations This package for recertification was approved by NRC in a telephne call on January 23 and by a letter dated January 28, 1986. For completeness and reference purposes the package transmitted to C.A. Julian on January 6, I 1986 is enclosed as Attachment IV. All required recertification training as well as UFTR checks, calibrations and surveillances were performed as required

. prior to resuming normal operations.

These checks and surveillances included I- completion of the annual reactivity measurements prior to beginning normal op-erations. For these reactivity measurements, the generation of control blade integral worth curves was accomplished using the blado drop method in incre-ments of 100 units resulting in up to 10 additional drops for each of the con-

.lW , Jtr.ol blades. All drops were successful. In addition, operations since March 3 1986 have been normal for over seventy hours of reactor operation. All of these points further support closing out the issue of the sticking UPTR con-trol blade.

/ 8 i // N

' Willioni G. Vernetson 'Date I* Acting Director of Nuclear Facilities

/

WGV/ps Enclosures cc: Reactor Safety Review Subcommittee P.M. Whaley, Acting Reactor Manager I -

I  ;

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W TABLE 1

SUMMARY

OF UFTR BLADE DROP TIME CHECKS February, 1986 I Following maintenance and overhaul of all drive systems, measurements of con-trol blade drop times have been made at each step of reassembly of the UFTR' core to include complete restacking of the concrete shield blocks. All results I '

are averages of three drop time measurements .except for Reactor Condition D.

The results of four sets of drop time measurements have been recorded to in-clude:

A. Drop Time Measurements With No Magnetic Clutch Acting (Mechanical Check) Following Connecting of All Drive Components.

I, B. ' Drop Time Measurements Prior to Fuel Loading With All Drive Assem-blies Functional Including. Magnetic Clutch.

I C.

Drop Time Measurements Following Completion of Fuel Loading to In-clude Replacement of the First Layer of Concrete Shield Blocks. .

D.

I- Drop Time Measurements Following Restacking of All Permanent Con-crete Shielding Blocks.

The results of these drop time measurements are recorded in Table 1.

TABLE 1

. Reactor Condition Control Blade Drop Time (Seconds)

I A S-1 S-2 S-3 0.475 0.425 0.483 RB 0.408 B S-1 0.433 I .

S-2 S-3 RD 0.400 0.467 0.400 C S-1 0.450 S-2 0.450 S-3 0.483 I. RB 0.408 D S-1 0.450 I S-2 S-3 RB 0.450 0.467 0.41 7

' Since each data point in Table 1 represents an average of all drop time mea-surements and since all values are less than one half second, the problem of I the sticking S-3 control blade is clearly corrected and the preventive main-tenance on S-1, S-2 and the regulating blade has clearly significantly reduced their respective drop times. Current values are characteristic of the original y 3

installed drop tiraes recorded for the blades on these drive systems. ,

ATTACIDENT I .

I mus.0.u.acu=

C.C.YERMETSCM, REACTOR MAbtACER . NUCLEAR FACILITIES DIVISION e

NUCLEAR REACTOR BUILDING l UNIVERSITY OF RORIDA CAIMilvtLLt.FLORIOA 326tl .

• PHONS (304) 393 8429 itLEA 14130 seret Nuclear Regulatory Commission Suite 2900 101 Marietta Street, N.W.

Atlanta, GA 30323 J. Nelson Grace I' .

Attention:

Regional Administrator, Region II Re: University of Florida Training Reactor (UPTR) -

Facility License: R-56, Docket No. 50-83 '

I As per telephone call of 4 September 1985, we are reporting the failure of one of the reactor control blades (Safety-3) on the University of Florida Training Reactor to drop on demand from a 64% removed position. This failure (sticking I

about 31% removed) was, discovered by'a Reactor Operator as he commenced a pow-er increase from the 1 watt critical position. The operator had accidentally raised the Safety-3 instead of the Regulating Blade for this power increase; I in returning it to the normal 640 unit position he felt the response was slug-gish and so he attempted to drop the blade from 640 units withdrawn to check i t. Following clutch current release the blade stopped at the 310 unit posi-tion and was subsequently driven in with the other three blades to shut the reactor down.' _..

Immediate checks l'nvolving subsequent removal to various heights show this I sticking problem to' be intermittant and to center in the 290-315 unit range but with some apparent sluggishness in the drop from other higher and lower heights. It should be noted that this is essentially a recurrence of the event I reported by our facility in a letter dated January 28., 1985 with subsequent followup in an intierim report dated February 9,1985 and closed out in a re-port dated March 26, 1985.

TheExecutiveCommhtteeoftheReactorSafetyReviewSubcommittee(RSRS)has reviewed the occurrence and concluded that it is a potential abnormal occur- .

ence as defined in UPTR Technical Specifications, Chapter 1. The RSRS has in-structed NRC notification as per Section 6.6.2 of the UFTR Tech Specs.

Analysis of the current problem is underway with corrective action to follow based uoon inspection results and RSRS recommendations. Based on previous ex-perience with the January 28 event and subsequent corrective maintenance, in-dications are that the sticking problem may be due to a binding in the S-3 I clutch possibly due to moisture or other effect reducing clearance. Such bind-ing was fcund to be part of the problem in the January 28 ovent. If this pre-liminary evaluation is correct, this binding can be corrected by increasing

,g the clutch clearance, a check of which has been approved by the RSRS Executive 3 Committee.

/ b William G. Vernetson Q 'W ,

I cc: RSRS Committee Acting Director of Nuclear Facilities September 4, 1985

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Preomet (906) 392 8429 TELEA 5430 September 17, 1985 i

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' Nuclear Regulatory Commission

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-Atlanta, Georgia 30323 .

' Attention: J. Nelson Grace - ,

Regional Administrator, Region II '

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Re: University of Florida Training Reactor Facility License: R-56, Docket No. 50-83 . .

, . Gentlemen:

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Pursuant to the reporting requirements of paragraph 6.6.2(3)(c) of th'e'UFTR

' Technical sp' ecifications, a description of a potential abnormal occurrence as .

defined in the UFTR Technical Specifications, Chapter 1 is described in this interim, report to include NRC notification, occurrence scenario and proposed

. solutions. The potential abnormal occurrence involved the failure of one of the UFTR control blades (Safety Blade #3) to drop fully into the core on de-

, , mand from a 64% withdrawn position.

j NRC Notification The Executive Committee of the Reactor Safety Review Subcommittee reviewed this occurrence on September 4,1985 and concluded that it is a potential ab-normal occurrence as defined in UFTR Technical Specifications, Chapter 1. The RSRS then instructed NRC notification as per Section 6.6.2 of

• hn UFTR Tech Specs. This notification was carried out by both telephone to Mr. Paul Frederickson and a following telecopy on September 4, 1985 (See Attachment I).

This interim report represents the 14 day followup report as' required in UFTR f Tech Specs, Paragraph 6.6.2(3)(c).

1 Occurrence Scenario As indicated in the telephone conversation with Mr. Paul Frederickson, Section Chief, Region II, and a following telecopy on 4 September 1985 (Attachment I),

I one of the reactor control blades (Safety-3) on the University of Florida Training Reactor failed to completely insert on demand from a 64** removed po-sition. This failure (sticking about 31% removed) was discovered by a Reactor ,

Operator as he commenced a power increase from the 1 watt critical position-where a complete set of readings are required to be entered into the daily

• operations log. The operator had accidentally raised the Safety-3 about 20 units instead of the Regulating B?ade for this power increase; in returning it to the normal 640 unit position he felt the response was sluggish and so be attempted to drop the blade from 640 units withdrawn to check it. Following clutch current release the blade stopped at the 310 unit position and was sub f I sequently driven in with the other three blades to shut the reactor down. The ~

Facility Director and then the Reactor Manager were notified immediately of this occurrence.

g.: -

Nuclear Regulatory Commission '

September 17,' 1985 Page Two It should be noted that the most recent blade drop times performed on June 21, 1985 showed a slightly increased drop tima from the values determined in March .

following previous maintenance work. However, the blade was dropped four times

.I with consistent and successful drop times on each check. In addition, several trips, both un' scheduled and for training, showed proper S-3 Crop response over the several months prior to this occurrence on September 3, 1985.

Immediate checks (with all other control blades fully inserted) involving sub-sequen't removal to various heights showed this sticking problem to be inter-mittant and'to center in the 290-315 unit range but with some possible, slug - ~

gishness in the drop from other higher and lower heights. It should be noted j

that this is essentially a recurrence of the event reported by our facility in E a letter dated January 28, 1985 with subsequent followup in an interim report dated February 9, 1985 and closed out in .a report dated March 26, 1985.

As indicated to Mr. Frederickson on' September 4, 1985 and again on September 16, 1985, the need to formulate plans, make various checks and, as entry into the core region is required, to let the core and structure cool for a period, prevents a final report on this occurrence at this time. However, Mr. ,

Frederickson did advise the submission of an interim 14-day report and recom-

• mended including an update'on the status of the problem. This update is pro-vided in this report.

Evaluation Evaluation and determination, of the methods for alleviating this problem of a sticking control blade as well as preventing recurrence 'were discussed by the UPTR staff on September 4, 1985.

Essentially this event represents a recurrence of the previous sticking blade

I event so the staff reviewed and expanded upon those items considered at the January 31 staff meeting following the original S-3 problem where it was de -

cided that potential blade drag points would include:

1. Inside gear boxes and/or bearings (previously identified as the cause of the January 28 problem),

I 2. Inside the blade shrouds perhaps due to failed rivets, buckling or warpage of the shroud or the control blade,

3. Shifted blade shaft / pedestal or bearing,

4. Shifted blade shaft / drive unit or' bearing,

5. Mechanical drag of the blade shaft in its guide channel. '

I It was agreed that all of these possiblities should be investigated in a sys-tematic program until the cause of the sticking blade is isolated, corrected and prevented from recurring. J'y I ~

Y ,

.g Nuclear Regulatory Commission .

g September 17, 1985 Page Three ,

The Executive Committee of the Reactor Safety Review Subcommittee (RSRS) was apprised of this occurrence on the day it happened and met to evaluate it on September 4, 1985. The decision was also made to put the UFTR on administra-tive shutdown with limitations noted in Attachments II and III. As indicated, they recomrrended reporting the event. The entire RSRS considered the event in I more detail at its regular meeting on September 6,1985. In both cases, the RSRS concluded in agreement; with the facility administration that this poten-tial abnormal occurrence did not compromise the health and safety of the '

I public. The blade has always responded properly to drive in to allow reactor

. shutdown. Required shutdown margin has always been available.

The RSRS agrees that all the possibilities listed above should be investigated in a systematic program to assure the cause of the sticking blade is isolated, corrected and prevented for recurring.

I Work Progress To Date I rollowing staff and RSRS evaluations,' this sticking S-3 blade problem is being addressed in a series of planned maintenance / inspection checks beginning with the right angle. gear box, drive motor, magnetic clutch, et.c. external to the I biological shield (designated ex-core meaning essentially environmental back-

. ground radiation levels) and working in toward the core regions where rela-tively high radiation levels are expected. Each platined maintenance / inspection activity or series of activities is described in a proceduro or instruction discussed by the UPTR staff and administration prior to performance. It is then being reviewed and approved by the Executive Committee of the RSRS con- .

sisting of the RSRS Chairman, the Radiation Control Officer and the Reactor Manager prior to the , start of work.

As of this date, the following maintenance has been performed:

1.

Right angle drive gear box inspection has been performed with all compo-nents found to be functioning properly.

2.

Dearings in right angle drive unit were inspected and replaced since a small amount of rough operation was noted in one bearing. This roughness was not considered sufficient to be the source of the problem.

3.

The shaf t and connecting. bearing were uncoupled from the blade drive unit and removed from the control blade. Both were inspected "for scar marks with no significant problems noted.

4.

The shaft penetration was cleaned - some oxidation and carbon products I were removed but not considered sufficient to have been the cause of the problem. .

5 5.

3 The bearing and shaft were reinstalled and recoupled, the potentiometer was repositioned and blade drop and timing checks made.

I ~

..a I , .

Nuclear Regulatory Commission I September 171 1985 ,

Page Four 64 Preliminary checks indicated the sticking problem is not? cleared.

Conclusions To Date

• The' problem has been isolated to within the biological shielding in the vicinity of the core reflector; most likely causes are a problem with the bearings supporting the blade / shaft coupling or with the magnesium shroud housing-to-blade clearance either warpage, misalignment or loose rivets.

consequences As concluded by the RSRS Executive Committee, the full RSRS Committee and UFTR administration, 'this potential abnormal occurrence did not compromise the.

^

health and safety of the public. This occurrence was discovered at a low power condition. The Safety-3 blade drive system was always functional; and even with the S-3 blade at ~30-35% withdrawn, the UFTR core has a shutdown margin of . ~3.7k Ak/k.

Followup .

' Sfnce further work will involve considerable radiation dose commitment, the core and structure has been allowed to cool while the above checks and. main-I' tenance efforts were completed. A preliminary procedure to address the remain-ing in-core maintenance checks is nearly complete and will be presented to the RSRS Executive Committee for approval on September 18. Work to check and in-spect for the possible in-core sources o. the problem and 'to perform main-tenance where necessary and approved is expected to begin later this week.

Commitroent The UFTR administration with concurrence of the RSRS is committed not only to I clear 'the sticking blade problem but also to obtain a significant reduction in the S-3 drop time. This reduction is considered necessary to preclude recur-rence of this event. In addition, the UPTR administration has committed to the I RSRS to clear any restart with NRC Region II prior to removing the facility f rom the current administrative shutdown.

/ O d' 2:2h%% lf $

William G. Vernetson D6te

Acting Director of Nuclear Facilities WGv/ps

.I Attachments cc P.M. Whaley f I

1 Reactor Safety Review Subcommittee -

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'l I Nuclear Regulatory Commission Suite 2900 ATTACIDIENT I September 4,1985 101 Marietta Street, N.W.

• Atlanta, GA 30323 ,

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I Attention: J. Nelson Grace ~

Regional Administrator, Region II

.Rc: University of Florida Training Reactor (UFTR) . .

Facility License: R-56, Docket No. 50-83 I As per telephone call of '4 September 1985, we are reporting the failure of one of the reactor control blades (Safety-3) on the University of Florida Training Reactor to drop on demand from a 64% removed position. This failure (sticking I about 31% removed) was discovered by a Reactor Operator as he commenced a pow-er increase from the 1 watt critical position. The operator had accidentally i

I raised the Safety-3 instead of the Regulating Blade for this power increase; I in returding it to the normal 640. unit position he felt the response was slug-gish and so he atitempted to drop the blade from 640 unitis with' drawn to check -

i t. Following clutch current release the blade stopped at the 310 unit posi-tion and was subsequently driven in with the other three blades to shut the

- ' reactor down. .

Immediate checks involving subs'e quent removal to various heights show this sticking problem to be intermittant and to center in the 290-315 unit range but with some apparent sluggishness in the drop from other higher and lower heights. It should be noted that this is essentially a recurrence of the event I reported by our facility in a letter dated January 28, 1985 with subsequent followup in an interim report dated February 9,1985 and closed out in a re-port dated March 26, 19S5.

The Executive Committee of the Reactor Safety Review Subcommittee (RSRS) 'has reviewed the occurrence and concluded that it is a potential abnorntal occur-I ence as defined in UPTR Technical Specifications, Chapter 1. The RSRS has in-structed NRC notification as per Section 6.6.2 of the UPTR Tech Specs.

I Analysis of the current problem is underway with corrective action to follow based upon inspection results and RSRS recommendations. Based on previous ex-parier.ca with the January 28 event and subsequent corrective maintenanco, in-dicaticas are that the sticking problem may be due to a binding in the S-3 clutch p::ssibly due to moisture or other of feet reducing clearance. Such bind-Ing was fcund tc. be part of the problem in the January 28 ovent. If this pre-liminary evaluation is correct, this binding can be corrected by increasing I the clutch cicarance, a check of which has been approved by the RSRS Executive Committee.

/ J William G. Vernetson d4f W l I cc: RSRS Committee Acting Director of Nuclear Facilities September 4, 1985

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ATTACIC4ENT II -

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September 4,1985

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MEMORANDUM ,

TO: Reactor Staff l%" *

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FROM: W.G. Vernetson,, a -

I. Acting Director of Nuclear Facilities

SUBJECT:

. Administrative Shutdown of the UFTR '

Because of the need to evaluate and fix the sticking safety blade

1. 3, the UFTR is placed on Administrative Shutdown until further notico.

• I .

This administratkive shutdown precludes all reactor operations in-

,,_ volving . operating the reactor except for 1) performing weekly and

" daily checkouts as far as possible and 2) withdrawal of Safety 3 up to 50 units. ' Daily checkouts performed for training and' non '

• I$ '

operational. uses of the facility such as for crCC students are also permitted. Withdrawal of Safety-3 more than 50 t'. nits is pre- -

cluded except as specifically authorized in RSRS-. approved test and maintenance. procedures.

The only other activities allowed during this administrative I ,.

shutdown are routine administrative work (updating training re-cords, e tc. ) , tours of the facilities (no operations) as well as housekeeping and maintenance on the non-nuclear-safety-related I equip::.ent in the reactor room provided none of work in progress or cause unnecessary exposure.these is also possible provided approved by p' roper levels.

affect the Major maintenance .

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. ,' September 9,1985 MEMORANDUM - .. .. , ,y.-Q

'10: Reactor Staff . ..

I FROM:

. W.G.VernetsonQ -

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SUBJECT:

Allowable Control Blade Manipulations During the Cur-rent Administrative Shu~tdown

• Two p,ointe should be noted relative to performance of weekly and, -

4aily checkouts during the administrative shutdown to correct the I . . . .

problem of the sticking S-3 blade:

~

First, the RSRS has agreed t' hat the full removal of the S-3 control. blade is alloUable during the weekly checkouts. To -

do this will require my authorizing signature. Note thgt re- .

moval of S-3 up to 50 units has been allowed with the usual limitations.' obviously this full. or partial removal will only be possible when the blade drive is co' nnected.

• I Second, care should be exercised shenever the S--3 blade drive is disconnected to assure .that ,the rupture disk is not .

broken. Recall that in the dump valve. logic whenever the electrical connections to a blade are disconnected, this is I like a partially removed blade. Each of the scram checks of one blade actually look like two blades and result in dump of primary coolant to the storage tank. Therefore, a three minute wait is required prior to reset.

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. ATTACHMENr III x

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'_ MEMORANDUM ns -

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From: .

W. G. Vernetson, ' Facility ,Diry t P.-M. Whaley, Reactor Manager \

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Subject:

Final Report on C'ontrollBlude' Repairs Date:'

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March 5, 1986 , -

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[. ' HISTORICAL PERSPECTIVE -,

control Binda Driva checkn ti . Safety-control. blade three (S-3) '

{u. was discovered on during a reactor scram. September 3, 1985, to be not fully inserting be. -The pcoblem was initially suspected to 1

related to one or more of three previous' maintenance programs

{

performed at the University of Florida Training Reactor (UFTR) :

~

1. Safety Blade Thr'ee Drive System Repair ~

(February / March, 1985)

[. . . . 2. Fuel Inspection (January, .1985) s

, , ,,,,3. Thermocouple . Repa,i rl '_(September, 1984) '

( These programs involved peripheral activities that could have created inadequate ,

clearance l between the control blade and the control blade housing (shroud) by forcing the graphite stacked around

( the shroud into contact with the shroud, or misaligning the control blade shaft. . *

{ These programs provided recent information about environment in the reactor core region. The preferable the course of work action from a radiation exposure standpoint was to repair the r Problem with fuel in .the corer . fuel removal and replacement L

involves part considerable dose commitment above exposure occurring as of the troubleshooting and repair ' process.

the committment to seek a solution In retrospect,

( considerations significantly contributed to delays inwith consistent ALARA completion 'of this latest maintenance program; successful however, this committment aided evaluation of proposed actions at every. step of

( the maintenance program from the standpoint of efficacy and dose committment.

{ correspondence and contact With other Organizations.

Contact with the staff of other Argonaut reactors was made to

{

determine if they had useful information.

UCLA. The UCLA Reactor Staff indicated that they had seen i

7 i PAGE 2 similar symptoms; the problem source at UCLA was warped control blades initiated by differential thermal expansion of dissimilar metals in the control blade assembly. The UCLA Reactor Staff reported that they mechanically straightened the control. blades.

University _qf, Washington. The University of Washington Reactor.

Staff reported that the blade drop times of all control blades at

!I UW has le'nghtened over years of operation; that trend seems unrelated to the UFTR S-3. problem.

Universities Research Reactor. Universities ,qf, Manchester. The Universities Research Reactor ~ (URR)

Liverpool .andI Staff. .

reported two specific' control blade problems : the control blade

' shaft bearings fail at approximate yearly intervals (due to' radiation hardening of bearing grease) and the rivets holding the control material in one blade partially failed <:ausing the clearances in the shroud to become inadequate.

Results. Inspection of the control blades through inspection I ports previously established in already being. considered to establish The degradation in drop times at the other the control blade shrouds.was blade / shroud. clearances.

facilities seemed to i be.due to the'use of grease-lubricated bearings de-polymerize in a in-core bearings are radiation field);

(which tend to the UFTR control blade constructed of graphite -to specifically avoid that problem. Although correspondence produced no probable

.from root causes to the UFTR sticking control blade problem, the input other Argonaut reactors provided additional checkpoints to be examined during work progress and performance of preventive maintenance on the control blade drive systems.

Results of Recent Maintenance Activities ,

Staff analysis was that a problem not severe enough previously to have prevented a successful scram of S-3 had been developing for a significant amount of time. The problem may I ,

have been initiated or exacerbated by maintenance area. Other problems in tandem with the in the core primary problem had ,

previously crcated enough friction in the S-3 drive system to inhibit successful scrams; these other problems were corrected by the previous control blade drive system maintenance performed in February / March, 1985. The primary problem developed to such a I magnitude that in September, 1985, it became sufficient to reduce scram capabilitics for control blade Safety Three.

PROGRAMMED TESTS AND INSPECTIONS Ini ti al Inves tiga ti ons. As initial investigations were undertaken, four (4) sources of the sticking problem were identified for investigation :

I 1. Blade drive system bearing's or gears,

2. Blade" shaft to shaft housing clearance, -

]

I

PAGS 3

,I I' 3. Control' blade to control blade housing clearance, and

4. Blade shaf t bearings I Safety Three control blade drive unit gears and bearings at the exterior of the biological shielding were examined first.

Although it was decided to replace one set of bearings in the unit because of rough operation, no significant problems were I detected.

to determine The long-shaf t and the shaf t penetration were examined if any clearance problems could be detected, but examination indicated adequate clearances. _-

At this point, prior to removal of bi'ological shielding,' the four.: .,-'

original areas of investigation were effectively reduced to the I two within cicarances the or biological shielding  : control blade / shroud-blade-shaft bearings within the . biological shielding. Tests and inspections had clearly indicated',that the problem source was centered in the reactor core area.

Test and In spec ti on' Eauipment. Specific test equipment ,and I procedures were proposed to significantly aid the S-3 inspection and. repair program. A microphone suitable,for manipulation in the reactor core. region was purchased with the expectation that, if the shaft bearings were failing or if l S -3 was striking the shroud, some audible indication might be electronically ^ magnified to locate the source of the problem ~ or check that the problem is -

resolved. A procedure was generated and approved to allow manual I "

icactormanipulation secured ofcondition S-3 under controlled conditions that assu. red a was m'aintainted; a manual operator appropriate to that activity was located. The construction of a I shielded platform was considered to provide a shielded area from which to work while in the core area, but judged unnecessary from ALARA considerations as wel1 as ease of performing the necessary I inspection on and maintenance activities.. A boroscope was obtained loan from Florida . Power and inspection within the blade housing Light company to attempt and other high radiation I areas within the biological shield.

Initial Preparations. Biological' shielding was removed fo'r access to the control blade ' shroud and the region adjacent to the I reflector where the shaft bearings supporting the incore blade assembly are mounted. Visual inspection was conducted of,all comp)nents that could be seen without physical intrusion into the I core area. A closed circuit television (CCTV) and a boroscope (en loan from Florida Power and Light Company) were used to examine the shaft, shaft penetration and bearing assembly of I Safety Three and the Regulating Blade. No problems could be identified, but some graphite powder was noted under the control blade shaft on the bearing (bushing) support pedestal for control blade Safety Three. One possible explanation for the presence of

'I the material with significance to the Safety Three problem was excessive wear or abrasion of the graphite bushing.

I Inspection with. Puel in Core. During the remote inspection I

PAGE 4 process with the boroscope and a cl.osed circuit television camera, the name of the bushing manuf acturer was discovered to be written on top of the bearing assemblies; attempt to obtain at least one set of it was decided to bushings to Provide '

replacement on a contingency basis. No records were ava.ilable as to dimensions of the in-core shaft bearings other than inner '

diameter. Contact with the manufacturer (Graphite Metallizing . .

Corporation) indicated that their records were not sufficient to provide the UFTR with order information; local investigation and '

contact with Argonaut. UTR's at UCLA,

. Washington, and the Virginia Polytechnic the University of . . -

' Institute did not.

provide the information necessary to order replacement parts..

The only other method of determining bearing size was actual .

measurement. The bushings are graphite . lubricated greased; unfortunately, graphite lubricated bearings are not a versus-stock item. Preliminary investigation indicated a minimum week delay from three time of order to receipt o.f graphite bearings.

'Therefore, it w,as. decided to measure the bushing to identify the correct replacement only if. removal was later deemed necessary.

The control blade was' manually manipulated conditions of the special controlling procedure (reviewed subject to the NRC prior to implementation) with fuel 'in place while by the inspection was performed visual expe'riencing any shifting along the axis of the shaft. wasA to determine if the control blade microphone was mounted on the control blade shroud to attempt detection of shroud and control

'Ehe source of the problem was detected. blade contact. No indication of It was by decided removing that 7the about in.control blade shroud should~ be ' modified of material from the edge of the shroud to permitTovisual inspection of the control blade with the fuel.in place. this end, a UFTR modification to remove part of the l magnesium shroud was reviewed and approved unreviewed safety question. The review decision as involving no was based on the facts that 1) protection from foreign material in the control blade operating area was not an issue (the shroud-edge is covered by graphite stringers); and 2) the cutting operation was to occur. with special safety measures to assure that ignition .

conditions did not occur in the (magnesium) shroud material.-

Originally it was the new inspection planned ports to examine the control blade'through manipulations.

with the CCTV camera during Although approved for implementation, attempts to modify the shroud as indicated were delayed until the fuel was  ;

removed to the spent fuel pits.

. Bushing Removal 11gd. Removal .

assemblics Radiation levels (8 - 15 R/hr) at the bushing were such that reduction of the field was considered necessary work to allow sufficient stay times in the area to perform tasks such as drilling the approved shroud inspection ports and removing the control blade shafts and bushings. Accordingly,,

s

. . .. ~

PAGE 5 I complete ' unloading of the reactor consolidation of'various partial (irradiated) was accomplished followir$g stored in the irradiated. fuel storage pits.. Consolidation ' fuel assemblies was I necessary to provide enough free unloading October of the UFTR. The pits to consolidation allow complete fuel was com'leted 10, 1985, and unloading of the fuel started and completed p on on October 11, 1985. .

The result was a significant reduction in the radiation that previous .to field in the core work' area. It should be noted this point, the decision had been made to

. perform similar preventive maintenance on all other

~

control _."

blades as would be found necessary to correct the problem for the:

S-3 control blade drive system. .

Graphite Removal. Prior, to removal o5 the bushing, partial -

removal of the reflector graphite was necessary as follows I

1. The top layer of. gra.phite was remov'ed f rom the reflector for visual and physical a.ccess to.the control blade shrouds on the north'and south sides of the reflector; -

2. The north-west (south-west, for control blades safety One and Safety Two) section of graphite was partially removed for access to the control blades; I 3. The center island graphite was nearly completely removed to allow removal of the north fuel box ~and

""4. . The graphite between Saf ety Three and the Regulating Blade (and later, Saf ety One, and Safety Two) was removed to. permit I removal of the control blades and shroud assemblies on the north side of the reflector.

I Shroud Modi fication_. The control' blade Safety Three shroud was modified on October 25 by removing.about 7 inches of material from the top edge of the magnesium shroud to permit visual inspection of the control cut. blade after the inspection port was Special procedures to assure incendiary magnesium conditions for were not achieved included rigging a water hose in the work area, CO2 fire extinguishers in the woth area (to reduce the I metal temperature if it appeared to be rising abnormally'high) and temperature indicating tape attached to the metal as it was being cut. Sample pieces of magnesium were also cut outside the I UFTR to observe the behavior of the metal during processing. The other three control maintenance project;blade shrouds were modified much later in the the special procedures were not deemed necessary to the process based on the experience gleaned in I cutting the S-3 shroud, and were not imposed shroud modifications performed out of core. for the remaining The control blade was examined with the CCTV camera during manipulations over its jI full range of movement.

No shroud to blade contact was detected.

The inspection showed no apparent warpage or deformity of the control blade such as had occurred at the UCLA reactor. The cadmium absorbers' appeared to be intact and properly in place (n I ~

l . [

I i

PAGE 6 l

3 loose rivets or warping). The end result is that the inspection 3 port provides excellent visual access to the control blades and will be used for future regular blade and shroud inspections controlled sheet.

by a standard operating procedure or surveillance data Eouipment Disansembly. As noted, due to construction features,

' it was necessary to remove fuel boxes, control blade shrouds, and control blades to remove the bushings.

fastener-hardware removed exhibited radiationEach piece of E two Roentgen per hour on contact levels of one to -

!E- The boxes (indicated by a. teletector).

fuel were stored in the shield tank, most 'of the-[

fastener hardware support brackets for was stored in the fuel transfer cask, the fuel- ,

the north-west and the north-central fuel.

boxes of were stored in a tank of water near the south-west corner the reactor cell, and the control blade, shroud, bushings and remaining hardware were stored between the shield tank and concrete shielding on the south s'ide of. the shield tank.

Replacement hardware (all stainless steel material) was purchased -

from ' Southern Distributors, Inc. and removed hardware disposed I of for ALARA considerations. .A waste shipment was made by the UFTR in mid-December; of the core was shipped. all hardware removed from the ~ north side The reactor waste from the south side of the core is temporarily stored and is scheduled for the next waste shipment.

I _ushing-Pillow B' lock Bolt process of disconnecting Retni ngra Modi f i ca ti'on. During the the bushing pillow blocks from their pedestals, it was discovered that the bolts holding the pillow I blocks in place.(mounted so that the bolts were on the under-side of the pedestal and the nuts on top) had nothing to prevent them from turning af ter the nuts were loosened.

This feature of the pillow block assembly caused extra time to be spent by personnel I in the highest radiation field present in the core area.

prevent a recurrence of this difficulty, a slotted retainer would To that partially capture the bolt head was constructed for the I north side control blades.

the Participation in planning for work on south, side by a UFTR Staff member with extensive experience in metals and machining processes resulted in'a modified appr'oach I for single the securing south. side control blade assemblies; on the south side, aluminum blocks each side of were the threaded bushings to . accept both bolts to provide a single ~ nut I assembly for each set of bushing hold down bolts that is easily manipulated. In each case, this modification allowed work to be accomplished quickly and/or remotely during reassembly; in addition, this feature will greatly reduce time and exposure committments during any future maintenance requiring removal of the bushings or control blades.

Both bolt retainer designs were reviewed and evaluated not to involve any unreviewed safety questions. -

Bushinn-Pillob Block Removal. On November 8, 1985, tne west-oriented bushing on the shaft-coupling of Safety Three had M

PAGE 7 I"

to be pried off of the shaft coupling assembly (jack shaf t) ,

finally providing positive indication of the source of the sticking control blade Safety Three. The remaining bushings literally slid off of the shaf t assemblies, indicating. that the west bushing on Safety Three had been imporperly bound on the' shaft. The interior of failed the S-3 bushing was found to be~

rippled and rough indicating some sort of irregular pressure and wear pattern on the bushing. This damage is considered to be the cause of the sticking S-3 control' blade. The bushings were' measured and replacements ordered as planned. A three to five .'.-

week delay was anticipated for receipt of the bushings. -

Blade-Shaft Coupling Modification. The removed shaft-blade.

couplings were made of a plain carbon steel (SAE number 1040) not-i stainless steel (according to Engineering Machine Shop Personnel a's well as the applicable engineering diagrams). The shaft coupling attached to safety Three was not servicable; the shaft coupling was rusted and scarred. It is probable that the rust on the shaft, some manufacturing. irregularity of the shaft coupling, I previous control blade problems or some combination of those had created or contributed to the failure of the control blade Safety Three bushing. The decision was made to manufacture new coupling assemblies (jack shafts) constructed of Type-316 stainless steel;

, this was contracted to the University of Florida Engineering Machine Shop. The. negative (unreviewed safety question) evaluation. for this modification was based on th'ree facts; I ....;. Type-316 stainless steel.is virtually corrosion proof, is topgher and has a higher value of hardness than the carbon steel. ,These facts mean that the jack-shaft surface is now less subject to abrasion and wear. The in-core coupling assembly change'from carbon steel to stainless steel was approved and evaluated as a change that .will reduce the probability of failures similar to

'the recent sticking of control blade Safety Three. '

REPAIR AND REASSEMBLY PROCESS Bushing and Control Blade Instal l ati on I Bushing Installation. The north side placed in the mounting pads.

bushing The retainers were regulating blade re-installation was commenced before the work on Safety Three for space considerations; access to the north side of the core was accomplished via removal of graphite f rom the north-west corner and (for the Regulating Blade) from between Safety Three shroud and the Regulating Blade shroud. Replacing Safety Three first would block access to the Regulating Blade. South side work was similarly accomplished. The control blade shafts, coupling assemblies and bushings were loosely bolted in place. The shroud was placed over the control blade and loosely fastened. The .

whole assembly was visually aligned and the bolts securely fastened. During the installation of the Regulating Blader it became apparent that the west bushing would require a shim; securing the bushing locked the shaft, while loosening then J

-( "

PAGE 8

[,

bushing allowed the shaft and. blade tc operate freely. While a '

shim in was being prepared preparation for blade drop timing for the bushing, the shaft was re-coupled

{ ' re-coupling process, it became apparent that the coupled checks. During the could not be rotated. to a blade because the drive fully inserted or bottom' position mechanism w a s' holdin'g the blade in an' b intermediate position (about 30  % withdrawn). Investigation revealed that ther.e is no significant adjustment that can be made for the bottom position of the control blade.

[._ '

Analysis nf. Misal'ignmenE.

The control. blade drive train consists".

of gears and shaf ts that are splined together in one orientation,

[' with no potential'for adjustnient. The critical parameter for- .

determining the- angular position of fixed position of the. drive . train the control blade with a-and long shaft is the 1 orientation of a steel

[ assembly (that the long shaft pin in the center bore of the coupling engages to move the control blade) with respect to- the control blade.

The potential sources for

.s this position control blade mismatch were seen as either the' orientation of the with respect to mounting on the coupling assembly or manufacturing of the coupling assemblies not consistent with the specifications of the engineering

[ engineering drawing of the shaft coupling drawings. assembli~s e The UFTR (General Engineering ' Nuclear Corporation. drawing 89-31-116, Control Blade Attachment Detail) was compared to the newly manufactured

{ stainless steel shafts; long shaft engages was proper. orientation of the steel pin that the The original shaft coupling

' assemblies removed from the core, however, did not match the drawings. The original shaft coupling

{ manufactured to 'specifica.tions. .

assembly had not been

.The coupling as'sembly and the drive unit could not be matched to allow the lowest position of the control blade to correspond to a fully inserted position.

Although the original jack shaft did not match the engineering diagram, this condition

[ would never have been known except for .

- the changes not in required to repair control blade Safety Three and is Three. any On way related to the problem of the sticking of Safety discovery, this condition was reported to Paul

{ .

Fredrickson at NRC Region II in a telephone conversation on December 2, 1985. This aspect of the system not matching the (i.e. jack shaft affected the operation engineering specifications) had not adversely

[ or safety of the control system; consequently, this drawing error is not considered to constitute an unreviewed safety question. However, a change to the jack shaft was necessary to .make the

[ system reflect the approved engineering diagrams using the manufactured replacement stainless steel shafts.

[ Ex-core Shaft Coupling Bodification. Because the new jack shaf ts were to be used based ALARA considerations, on a combination of cost, engineering and a difforent (consisting of a web-coupling unit) ex-core coupling arrangement

( with the potential for adjustment was selected to couple the drive unit and the long/

[

[ - - - -

a PAGE 9 shaft, replacing the previous star coupling. This modification was also reviewed relative to the potential 'for involving an unreviewed safety question with a negative 10CFR50.59 I determination.

orientation of This modification permits changing the drive unit / drive train with respect to the the control blade and allows the lowest position of the control blade to correspond to full insertion.

Control shroud, Blade Shroud Installation. As. With the Safety Three the remaining . control blade shrouds were modified with - ,

inspection ports drilled in the top of each shroud to permit visual inspection of the control blades with the shrouds in place. The control blades and their shrouds were placed back, I into* the core, and the assembly fasteners loosely made up.'

Consideration wa,s given to placing bearing collars on the shaft coupling to prevent axial movement as the only physical mechanism I. presently in place to prevent axial movement is the bushing; the bushings are not designed as thrust bearings and they only supply radial support. However, the decision was made not to use bearing collars as the collars.could provide extra friction and wear surfaces.

Fuel Enx Installation. Prior to replacement of center island and top-core level graphite, the fuel boxes were re-installed.

New (unirradiated) fastener'harduare was used at the flange, and new nuts'at'the bottom fuel box flange.top fuel box I _ _ ;2 box flange studs were.never removed and were consequently about 1 Roentgen per hour on contact producing the highest

. Bottom fuel still in-coge field.

Some difficulty was encountered while placing the I south-east fuel box in position; by securing that box first and.

working towards the south-west box, the south fuel boxes were finally. secured. This difficulty of . reassembly. was not I. encountered with the reinstallation of the north fuel boxes.

Center Inland Graphite Support Modification. During the fuel box '

removal, a phenolic-type material was discovered under the center I island in.

graphite, positioning the center island graphite about 1/4 above the core support assembly. The material had apparently been embrittled during reactor I operation.

material had broken into pieces, and was consequently removed during maintenance and cleanup operations. To provide the proper The geometry, three 1/4 in'. aluminum plates were placed under the I center island graphite. This change was approved and evaluated per SOP-0.4 not to constitute an unreviewed saf ety question.

SYSTEM CHECKS FOLLOWING REASSEMBLY Pre-Puel Load itaintenance Program

_Primar,yt Syst.ca Plushing.and Leak Checks. When the fuel boxes were reinstalled and the connections secured, the primary coolant storage tank was filled (having previously been drained to the waste holding tank) with city-water. The primary coolant systemi I \

I 1

t

be w PAGE 10 was used to circulate water through the piping and fuel boxes.

~

The system was then leak-checked. The initial check indicated gross leakage from the primary coolant system. To provide for leak isolation, .a set of three expandable plugs was used to prevent fluid from draining. from the fuel boxes (the plug, inserted to about six inches below. the -inlet piping flange) .

Leakage was isolated to the south-west and south-east fuel boxes; the lower flange surfaces of the boxes'were then cleaned and dressed and the boxes re-installed. .

The final leak check indicated some -water loss (an estimated -.-

one-gallon loss over ad approximate 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period), but investigation demonstrated that this. water. loss was from, I, "

evaporation. . 'Th e . total interf ace area along which the flowing water exposed to moving air (from the core vent system) was conservatively estimated at 20 square feet. A simple experiment was conducted measuring water loss from evaporation using a (static) pan of water. The results indicated that 0.62 gallons of. water could be lost from a 20 square foot surface area over~

I the same time frame.

secured as much as transport modes, is and As a final check, the core vent system was physically possible to restrict- mass the primary system circulated overnight.

I The resulting water loss (perceptable, but not measurable with the primary' indicated tnat coolant storage

'the ~ previous tank water level losses' detection system) had ^been due to I - ;-

evaporation.

Following a long period of circulation, the primary coolant storage tank was drained and visua,lly inspected. The tank was determined to be clear of debris; the primary ' coolant . storage tank was refilled with demineralized water, and the purification system activated. Resistivity values stabilized at greater than I 1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

~

megohm-cm '(a typical normal operating value) within a few

~

I Retesta Prior.to-Connectina Control Blade Assemblies.

times of all control blades.were initially measured'by manually withdrawing each control blade until the top limit switch engaged The drop I to provide a top-end fall out signal; device position where the bottom then the manual withdrawal was released so that the control blade fell to the limit switch provided the full in bottom signal..

I one-third All of control blades exhibited drop times of about a second. This time was expected to be somewhat less than the times that would be obtained when the disengagement of the magnetic clutch would be used to initiate the control I blade drop. Results of drop time measurements on all four (4) control blades are tabulated in Appendix A of this report for the following four (4) conditions :

1. Without use of magnetic clutch (mechanical check),

I 2. Prior to fuel loading, all drive assemblies functional (using magnetic clutch),

}

s, - ,

. PAGE 11

3. Following completion of' fuel loading to include replacement of the first layer of biological shielding, l 4. Following complete restacking.of all permanent concrete biological shielding.

I Retest Prior in Fuel Load. As delineated in the letter of January 6, 1986, to C. A.

'Julian (NRC, Region II), a number of checks, tests and .surveillances were conducted as required ' prior _.

I to fuel-reloading.

loading.

The results of all. required checks, tests _and; surveillances, as delineated in Appendix B, were favorable for .

~

Fuel . Load Trai ni ng. Fuel load training wa's held on' February 4, 1986, for radiation control personnel, the UFTR staffr and two Nuclear Engineering data aquisition and Department. classes (expected t'o help in the Training included (fuel analysis during the approach to critical) .

load) standard operating procedures, special fuel load procedure, I -

radiation as work practical permits and training radiation on radiation control procedures, control procedures as well fuel handling equipment.

Recertification training had previously been held on I 28, 1986, on the reactor. startup, operating, procedures as well ,a s the pre-operational chec~k procedure with 28 and January shutdown associated practical training.

~

- - %ntimonv-neryllium Source Modification. While preparations were being made for the fuel loading (preparation of counting systems, I radiationwas source workdiscovered permit requirements, etc.), ,the Antimony-Beryllium to device by which the source was handled.

have lost the permanently attached Investigation determ'ined that the source was leaking, but a special container was I necessary such as required to permit the leak not source checks.

to be manipulated' for . activities The container was designed, reviewed and approved with a 10 CFR 50.59 evaluation performed to I- assure no unreviewed safety question was involved. The container was constructed and the source inserted into

. port. the west vertical Fuel Loading Process I Fuel Loading.

training.

allow Fuel loading began on the day following fuel load Final preparations and personnel schedules did not the procedure to begin until late'in the day, but four fuel bundles were (4) loaded on February 5, 1986. Fuel loading proceeded on the February 6 with ten (10) more fuel bundles added. Fuel loading was completed on February 7 as the last sevenpartial(7) irradiated fuel bundles were returned to the core and I one (5-plate) bundle of fresh fuel was added to provide necessary excess reactivity. The Boecial_ EMR1 Load Procedure controlling the fuel loading sequence was modified on February 7 to allow a change in loading sequence, and a change in the I.

s- . .

e .

PAGE 12 constitueht fuel plates. for the fresh partial (S-plate) fuel bundle. l I During the 1.oad process, fuel bundle UF-32 was discovered to have been constructed with about protruding past the surface of 1/4 in. of the assembly-bolts 1

clearance the nut. This eliminated the between that bundle' and the bundle next to which UF-32

-was to be loaded. Consequently, UF-32 was loaded into a . later-position, allowing the UFTR to proceed with core loading and then to trim -enough of the bolts off to allow the bundle to fit. l I Special Fuel Load Procedure was altere.d to refle'c~t the required.

changes, and.a 10 CFR 50.59 evaluation for~the bolt modification The _. '

l on UF-32 completed per SOP-0.4. A record of the fuel loading I, sequence is contained in Appendix C including a Summary of Fu'l"

' Loading Movements, Table 1 depicting fuel bundle ' locations e in-core (update of Table 1 f rom t.he Special Fuel Load Procedure)

.g 'and Table 2, showing loading increments (update of Table 2 from 3 Special Fuel Load Procedure).

Physics Parameter Retests. As fuel loading progressed, subcritical multiplication data was collected from four (4) independent neutron monitoring channels as delineated in the Special Fuel Loading Procedurn. This data was used to generate I inverse multiplication (1/M). plots which were constructed during the fuel load process following each fuel load increment.

The reactor on achieved criticality at 393 units of the Regulating Blade the penultimate increment (original 21 Irradiated bundles

'lo'aded), and at 599 units. on Safety Three .on the final (partial plate, fresh bundle loaded) . increment. The estimated I

~

excess reactivity loaded into the core based on the subcritical multiplication data and the mass coefficient of reactivity as presented in Appendix D was estimated to- be 1.4 % 4k/k conservatively.

Biol ogical Shielding Restack. Biological shielding was restacked '

I on February 12, 1986.

Pre-operable Testing -

I Spacial Tests. Drop times for all control blades were following replacement of biological shielding; conducted all drop times were measured to be less than 0.5 seconds as indicated in I Appendix A to assure the sticking problem is resolved. A limited series of control blade worth de termina tions determine the was made total worth of each control blade, excess to reactivity (based on critical control blade I positions) reactivity worth of the last increment (1/2 fuel bundle).

Subsequently, complete sets of reactivity data were aquired to and the show all Technical Specification limits and conditions are met and that the core excess reactivity loading is 1.12 % Ak/k.

Radiation surveys were conducted at 1 watt and 100 watts during the control blade worth determina tions; surveys conducted during power ascension at 1 kW, 10 kWwere later and 100 kW

, PAGE 13 l -

levels.

Su rv ei l l an ce s . Several surviellances had exceeded the maximum accepted period since their last performance either because the

' conditions of the reactor did not permit their performance during the extended outage or they had no meaning during the outage.

3 Prior to restoring the reactor to an operable condition, the g overdue surveillances were all satisf actorily completed to meet the requirements of the UFTR Technical . Specification.

Observation of the performance of and the data obtained for these-.,

surveillances was used to support Nuclear Engineering laboratory -

courses. 'The surveillances were all successfully completed by' February 28, 1986, as conveyed to'NRC Region II by telephone on February .28, 1986. After all reactor operator certification '

requirements were completed on March 3, 1986, the reactor was returned to full operability on March 3, 1986.

ALARA CONSIDERATIONS ,

Radiation Exposure Limits. . Radiation exposure limits (normally iI 75 mrem / week) were extended to 300 mrem / week for the control blade repair period (see Appendix E for Memorandum). Radiation exposure limits were' restored to. normal Univer.sity of Florida limits on February 7, 1986.

UPTR Staff' Exposure Summ4rv. Exposure tabulated by radiation I ._. ., control includes personal dosimeters and thermo-luminescent dosimeters used to detect doses to whole body, right ankle, right wrist, forehead, and fingers. Dosimetry was specified to provide I dose information based on specific tasks to be~ accomplished and character of the work area during the tasks. Results' for the principal UFTR workers (based on the accummulation of significant doses for each month) are tabulated by month in Table 1.

Exposures for the principal UFTR workers (based

{I of significant dose during controlling Radiation Work Permit (RWP) and each task) on' accumulation are- tabulated by the task description in Table 2.

1 l

l 5

u ,

L l

i 1

. TABLE 1 RADIATION EXPOSURE

SUMMARY

Il }

SAFETY-THREE MAINTENANCE PROJECT OCTOBER,1985 - FEBRUARY,1986 OCTOBER .

"Whole Bod'y ' Ankle Wrist Fingers

. Forehead -

I Stiehl Personal Dosimeter 196 NR I2) 237

• NR NR

.TLD '136 NR 227 87 REi-270(3)

{ LH-212 i

Whaley Personal Dosimeter 65 NR 70 NR

• NR TLD .

49 NR ,

60 13 RH-70 LH-70 I4)

. :.f;..-

NOVEMBER Whole Body Ankle Wrist Forehead Fingers

Personal Dosimeter 521 660 480 293 NR l '

l I Whaley TLD 302 623 445 224 RH-601 Personal Dosimeter 426 61 0 5 74 313 NR l

l TLD 206 453 395 222 RH-499 LII-679 1

NOTE 1 : Dose in millirem.

NOTE 2: NR means dose measurement was not required.

NOTE 3: RH = right hand; LH = left hand.

NOTE 4: Estimated' conservatively at.70 mrem, TLD Chip lost from ring badge during anti-C clothing removal.

I

v- y ,

TABLE 1 (CONTINUED)

RADIATION EXPOSURE

SUMMARY

III DECEMBER Whole Body ' Ankle Wrist Forehead . Fingers ,

stiehl '

Personal -

Dosimeter 200 10'9 ~58 32 NR '

TLD 183 80 37- -

27 RH-37(3)

Whale'y LH-27 Personal Dosimeter 170 485 195 119 NR I Fogle TLD 85 264 1 21 77 RH-249 LH-185 Personal Dosimeter 60 125 68 30 NR TLD 35 72 53

- 20 RH-60 LH-70 NOTE 1 : Dose in millirem.

NOTE 2: NR means dose measurement was not required.

NOTE 3: RH = right hand; LH = left hand.

I 4 e

O e

I .

l l

e I

il -

1 i _ _ _ _ _ _ _

- s TABLE 1 (CONTINUED)

RADIATION EXPOSURE

SUMMARY

(I}

JANUARY Wholo Body Ankle Wrist Forehead Fingers Stiehl Personal Dosimeter 590 ' 1'143 654 383 NR(2}

TLD '347 798 384 248

. RH-4 85 i Whaley LH-496 Personal .

Dosimeter 51 8 800 505 352 *NR TLD '304 499 333 21 2 RH-470 Ilanson LH-376 Personal Dosimeter 150 60 154 72 NR .

TLD 76 42 86 50 RH-99 I .

Gogun Personal LII-185 Dosimeter. 311 632 308 195 NR TLD 31 6 385 189 136 RH-257 Vernetson LH-272' Personal Dosimeter 138- 290 400 110 NR TLD 74 1 72 1 92 81 RH-286 .

LH-192 FEBRUARY Whole Body Ankle Wrist Forehead Fingers Sticht Personal Dosimeter 41 24 47 22 'NR TLD 18 5 37 34 NR Whaley Personal Dosimeter 21 NR NR 4 NR TLD 14 NR NR 5 NR NOTE 1: Dose in millirem.

/

NOTE 2: NR means not required.

~

. .~

1

' TABLE 2 l RADIATION EXPOSURE TABULATED DY JOB D

(

l (Personal ' Dosimeter Readings From RWP File - Whole Body Doses)

RWP No. Date. Job Description Name' (mrem) 86-9-I 2-1 2-86 Restack Core Shielding Whaley 7 .

Vernetson 3

. Stiehl 0 Fogle 0 86-8-I 2-7-86 Modification to UF-32 Whaley 16 Stiehl- 27 86-7-I -

2-5-86 Refuel Reactor Whaley 8 Stiehl 6 Gogun 3 Vernetson 2 Fogle 77

-- : 6-I IU31-86 Inspect Core Area, Place Whaley 8 Neutron Sources Into Reactor Stiehl 0 Vernetson 0 I 86-5-I 1-24-86 Restack Center Island Graphite and Remaining Graphite Stiehl Whaley 166 1 31 86-4-I 1 86 , Rework South-side (East and Stiehl 34 West) Fuel Boxes Whaley 10 86-3-I 1-14-86 Connect Fuel ' Boxes South-side Stiehl 1 20 -

Whaley 11 Gogun 15 86-2-1 1-6-86 Remove / Replace South Side Stiehl 260 Blade Assemblies Whaley 265 Gogun .270 I Vernetson 138 86-1-I 1-3-86 Replace North side Graphite,

- Whaley 90 Unstack South Side Graphite, Gogun 12 Remove South Side Fuel Boxes

r..

I l

TABLE 2 (CONTINUED)

RADIATION EXPOSURE TABULATED BY J'OB (Personal Dosimeter Readings From RWP File - Whole Body Doses)

I RWP No. Dato Job Description Name (mrem)

{ .

,85-23-I 12-13-85 Align RB, S-3 Shaft and Bushings Stiehl ~300 II ) --

( Whaley' 266 . .

Fogle 60~

t 85-22-I 12-8-85 Package Radioactive Waste for , Stiehl - 32' I

l l

Shipment 85-21-I 11-29-85 Replace RB, S-3; North-Side. Stiehl 1 70 Graphite, Begin Replacement Whaley 1 75 1

S-1, S-2 Bushings 85-20-I 11-21-85 South-Side Reactor Hardware Re-l Stiehl 2 E moval; Shaft coupling Replace- Whaloy 1 01 g . s. . ments; R.B. View Port I

85-19-I 11-13-85 Reg. Blade Removal, N.E. Fuel Box Dismount Stiehl Whaley 132 7.

85-18-I 11-5-85 S-3 Control Blade Assembly, Stiehl 1 81 j

I Graphite Island, Intra-Shroud Whaley 141

• Graphite, NW and NC Fuel Boxes

{ 85-17-I 10-30-85 Graphite Reflector Removal Stiehl 196 Whaley 65 1

I 85-16-I 10-25-85 Modify S-3 Shroud (Inspection Port)

Stiehl Whaley 39 0

85-14-I 10-15-85 Remove Final Layer of Shielding, Stiehl N/A First Graphite Layer Whaley 1 85-13-I 10-11-85~ Unload core to spent Fuel Pits Stiehl

• 2 Whaley 10 NOTE 1: i Off Scale Dosimeter, TLD and Film Dadge Readings Taken (estimated  !

reading).

w. .

m .

TABLE 2 (CONTINUED)

RADIATION EXPOSURE TABULATED BY JOB I . (Personal Dosimeter Readings From RWP File - Whole Body Doses)

RWP No. Date Job Description Name (mrem)

I 85-12-I 10-10-85 Consolidate Spent Fuel Pit Stiehl 'No Dos'e- ;..

Contents Whaley No, Dose ,

.I

• 85-11 -I 10-7-85 Visual, Remote Inspections and Stiehl 9 s.

Top-Deck Work Whaley' 3 I 85-10-I 9-30-85 Visual, Remote I.nspections and 'Stiehl 17 Reactor Top-Deck Work Whaley 8 85-9-I 9-25-85 Unstack Core Shielding Stiehl 19 Whaley 6 I " "85-8-I 9-11-85 Remove S-3 Shaft Stiehi Whaley No Dose No Dose I .

I ~

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APPENDIX A

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1

SUMMARY

OF UFTR BLADE DROP TIME CHECKS '

, February, 1986

'I .

Following maintenance and overhaul of all drive systems, measurements of con- 1 trol blade drop times have been made at each step of reassembly of the UFTR

,I core to include complete restacking of the concrete shield blocks. ' All results are averages of three drop time measurements except for Reactor Condition D. i The results of four se,ts of drop time measurements have been recorded to in-clude:

A.

Drop Time Measurements With No Magnetic Clutch Acting (Mechanical I

Check) Following Connecting of All Drive Components. ,

B.' Drop Time Measurements Prior to Fuel Loading With All Drive Assem-

~

blies Functional Including Magnetic Clutch.

I C.

Drop Time Measurments Following Completion of Fuel Loading to In-clude Replacement of the First Layer of Concrete Shield Blocks..

D. ~

Drop Time Measurements Following Restacking of All Permanent Con-crete Shielding Blocks.

The results of these drop time measurements are recorded'in Table 1.

TABLE 1

, Reactor Condition- Control Blade Drop' Time (Seconds)

I A S-1 S-2 0.475 0.425 S-3 0.483

I RB 0.408 B S-1 .

l 0,433, S-2 0.400 S-3 0.467 RB 0.400 i C S-1 0.450 S-2 0.450 ,

S-3 0.483 RB 0.408 D S-1 -0.450 I S-2 S-3 RB 0.450 0.467 0.417 Since each data point in Table 1 represents an average of all drop time mea-surements and since all values are less than one half second, the problem of I the sticking S-3 control blade is clearly corrected and the preventive main-tenance on S-1, S-2 and the regulating blade has cicarly significantly reduced

~

their respectivo drop times. Current values are characteristic of the original I y

, installed drop times recorded for the blades on these drive systems.

gjw

~' ~

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p .

L L'

L F

G I

E

~

APPENDIX B E

[ .

4

[ .

[ .

{-

[

[ ]

L .. .

- ---.-- ~

. . , ,.

• s - *

. ~ . . . . . . .

~

ATTACIDENT II .

WORK PLAN FOR,'RESIORATION OF UFTH TO OPERABLE CONDITION -

' . ~ .- .

2. .

Complete Repair end Haintenance Work on Contrsal Blade Driv e Systent (F-1)

.) -l II. .ruel Reload: - ' .

~

~'. . , . I.'

A. Loading Plan '

' , . V, 0. *

.s.,- ..' * .

1. . Develop The Plan

?s. ' . .

l,.: :ty. .

.'f..

.~. 2 .

Receive Approval by RM, FD, RSRS B ,. ,

./."

Staff Recertification

• . Program '

.-r- .

1' .

Conduct Training Required to Reload Fuel .

E '. : , '

2. ..

.- ' Recertify Operators by Pacility Dire'ctor

"[ ,

, , . . 3.' .

- ,' T. ... .

Receive Approval of Recertification from NRC Region II Operato Licensing . .

.~ . . .* :. . - ~

E , ' ,'. h. . .'. ."...

.- .::::,, . ' C.

. Inst rumentation Setup .

D. '1. .~ *

• Fuel Reload Performance

~

~ ,; . . ' , . 1.

l

.a . -

Inspect All Irradiated Fuel Elements (D-2) '

- r: '

2. ,

, . c. . . . . . -

Reload All Approved Elements With All Operators Assisting '

3.

- ~

Document' Reload III. Test and Checkout Progra

) .

A. - -

Staff Recertification Program

l , 1. Complete Practical Training' .

a.

Conduct Weakly Checkout (Group)

\ -

I ,

D.

• Precritical Tests b.

Conduct Individual (observed) Daily Chechouts '

1. .

Perfom (Q-2) Quarterly Calibration Check of Arca and Stock Monitors I

2.

Perfom Annual Replacement of Control Blade Clutch Current f

Light Bulbs (A-4) .

\

l. 3

.Heasure Control Blade Removal Times (Weekly Checkout) l

, w - - - ~ _ _ . . . .

~

, c. -

. 1

-I.

4. .

Measure

., Control Blade Controlled Insertion Times (S-5) .

5. .

Measure Control Blade Drop Times (S-1) . .

6.

Perform Quarterly Check of ,Screra.'Punctions (Q-1) ,

C. Critical Checks I ','.

- ~

. [.~ -

1. ,

. Establish 1 watt critical position consisting of five startups , ,

to 1 watt followed by shutdowns (used also as part of operator.

recertification ' training).

. 2.

I.., ...

• Certify each Direct'or) . operator' to resume licensed duties (Facility '. ), . .

~~

4.;', .

3. -

. Perform Annual Reactivity Measurements (S-2): .

I . :f.,,. ..71 -

a. , Worth of Each Control Blade s

,- b.

Reactivity Insertion Rate of Each. Blade Total Excess Reactivity

c. '

di Shutdown Reactivity (Shutdown Margin) . -

s.

.s:_., .

D. Power Checks I _'. .[ '.'. ...-

. l"> ~..

T .fl 1.

G: Perform Environmental' Celi. Surveys During Power Ascension, I'

" ..E. 2. ' Perform' Annual Measurement of UPTR Temp'eratm o Coefficient of

' Reactivity-(A-3) - -

I',-. . 3.

' Perform Quarterly Radiological Survey of Restricted Areas (Q-5) s

4. .

. Perform Quarterly Radiological Survey of,Unrestiicted Areas (Q-4) .

5. '

s Measure Argon-41 Stack Concentration (S-4)

~

6 Measure Dilution Air Flow Rate (S-4 from A-1)

~1.

. - Perform TIFTR Nuclear Instrumentation Calibration Check and Calorimetric Heat Balance (A-2) -

8 Perform Verifications Pollowing Calorime6 tic *

a. Temperature Coefficient

b. Blade Worths

, IV. Review and Approval of All Results A. Reactor Manager B ". Facility Director I C. Reactor Safety Review Subcommitteo - )

l D. Transmittal'of Final Report to NRC Region IT

a u a e gn , . -2 O O

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j APPENDIX C 4

a 5

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d e

I $

k

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a L '. . ..

, ' TABLE 1- . , ,- .

E'- - '

UFTR CORE ret.0ADIrlG PLAfl DIAGRAff Cor'e Map' Showing Loading Movements Used to Reload UFTR Core From Storage P E.

L With More Highly Burned Fuel Moved Generally to the East side of the Core.

r .

flW Box NC Box -

__ ~

u -

23 11E Box .-C 13 -

4- 3 g,

' pit 13 pit'18 -

~ 15 20 .~ -

pit 19 pit 14 R . UF-32. UF-29 pit 3 Shld'Tnk  :.-

L . . :.

• UF-22 UF-17 UF-19 UF-34D '

?

~ 17 9 '2 i t' l' 11 21  ;-

( , r& F .-

pit 4 UF-27 pit 22 UF-28 pit 21 pit 15 pit'9 pit 5

. UF-23 UF-16 UF-11 UF .

( p, . . .

L_. . . v' -

.O -

SW Box. SC Box

[ ' , I. ~22 12 SE Box 5 6. 10 18 I'O.,- f pit l' pit 16 UF-26 UF-24 pit 17 UF-25 pit 11 pit 6 pit 12

, UF-12 UF-20 UF-33 1 -

' i'h '

24/25 16

' 8 7 14 19 l '.ish - '

NEW pit.2 pit 8 pit 10 pit 7 Shld Tnk UF-40/ UF-18 UF-10 UF-13 UF-21 UF-36D

-I

-UF-340 .

" Note 1: Upper numerals in each bundle location give the number of the fuel or .

dummy tiundle

.y.. movement (1 to 25) and the ordered sequence of reload-b '

'ing fuel bundlesas per SOP-C.2, " Fuel Loading," to control the '

loading' pattern as.shown.

l.,. .-' ' Note 2:

l Pit #'s' iVidicate irradiated fuel storage pits from which fuel was ~

+

removed to the shield tank for visual inspection and then placed into

,.., ~the core. - '

}

tiote 3: The final fuel bundle to be loaded was placed .in the SW corner of the SW fuel box. Based upon subtritical multiplication data and the I. desire to limit excess reactivity to about 1.5% Ak/k or less, a partial (5-plate) fresh fuel bundle (UF-40) was added in this location along with a newly constituted partial (5 plate) irradiated dummy bundle (UF-39D). An excess reactivity of about 1.4% Ak/k.is estimated to .be in effect. '

Note 4:

The fuel bundle aluminum wedging pin was inserted whenever a fuel bo.x '.

was completely loaded to prevent reactivity variations from shifting -

fuel bundles. Therefore aluminum weding pins were inserted after the ~ .-

following movement numbers filled fuel boxes as noted:

NC Box - ' Af ter Movemen.t #4  !!E Box SC Box - After Hovement !8 kfter Movement #21 SE Box - Af ter Hovement #19 tiW Box - After Movement #23 SW Box - Af ter tiovement #25 /

M

~ - ~

~

TABLE 2 -

• UFTR CORE FUEL RELOADItiG FROM'

• IRRADIATED FUEL STORAGE PITS I- .

. (includes insertion of dummy bundles)

I- .

Movement flumber Initial Pit Location Fuel Bundle (gm U/gm U-235)

Previous

ore Location Final New Core Location Initials

'RX Manager.

Confirming '. .

l. '

• Final .

P1acement *'

1.. '21 . -

UF-23(172.65/160.75) NW of SE l'9.'T .:.

-}'3 .2 15 14 UF-16(170.49/158.75) SE of NC SW of NC SE of flC Y -

. UF-17(170.50/158.80) NE of NC NE of NC-

- ~

j 4 19 UF-22(170.47/158.71) SW of SE I'. . .,'I. NW of NC

• 5 17 UF-25 NW of NE 6

7 11 UF-12((172.56/16'0.64) 171.31/159.51) NE of SC NW of SC NE of SC '

.10 '- UF-13 171.44/159.64) SW of SC SE of SC

~

8 -8

~ UF-10(171.61/159.79) SE of SC SW of SC

-9'- 22 I,"~4 ~

6 UF-28(173.61/161.66) NE of SE SE of NW -

.'10 .

UF-20(170.26/158.55) SE of NW

. NW of SE g' . } 11 9 3 c,' 16 UF-11(171.49/159.64) NW of SC SW of NE

. '12

/ . ;. '

-

• UF-24(172.53/160.62) SW of flE NE of SW

' 13 .. 18 UF-29(173.30/ 61.35)

I. ,

14 ,g- 7 ' -

UF--21(170.41/158.68)

SE of NE NE of NW NE of NW SW of SE

~

15 3

,- .16, UF-l'9(170.46/158.74) NW of NE NW of NE

. . "l 2 UF-18(170.45/158.71) NE of SW SE of SW '

. 17 .4 18 UF-27(173.85/161.89) SW of NW SW of NW 12 19 -

UF-33(170.38/158.65) SW of NC NE of SE UF-35D(11 plate dummy) --

- SE of SE ,

20 -

UF-34D(11 plate dummy)

I. 21 5 NE of NE

~

UF-14(171.46/159.63) NW _of NW SE of NE '

22 1~ UF-26(173.28/161.34) NW of SW NW of-SW 23 13 UF-32(17.23/158.53) NW of NC NW of NW

~

24 I. 25 NEW NEW(i rrad)

UF-40 (78.34/72.95)

UF-39D(5 plate dummy)

N/A N/A SW of SW SW of SW Note 1:

Spaces indicate reloading configurations where subcritical multiplication dath was taken as required by UFTR SOP-C.2, " Fuel Loading," to predict critica limits on fuel to be loaded in the next increment. ~

Note 2: '

The. Reactor, Manager / Facility Director initials in the final column indicate all data addition hadofbeenmore evaluated fuel . and the next fuel bundle insertion approved prior to the those configurations for which subtritical multiplication data w plots.

I Note 3:

mThea w new =~partial bundle UF-40 loaded into the SW corrmafLfdbwaputurJuke

._# . m - - -. - - - - -

g 44 2,g ,

• O I

lI g .

I .

,I i

^

4 APPENDIX D i

I I .

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L;. fy_; ! -

r .

u~ . -

ESTINATION OF UFTR EXCESS REACTIVITY '

[. ,', .

FOLLOWItiG FUEL LOADING i.U Analytical Estimation of Excess iteactivity ~

' . .i .

? u,

.c'.-

> N P:nultimate Loading (7 Feb 86)' Final Loading (7 Feb 86) i .* -

. ;:6; .

"21 Fuel Bundles

• :.i? .* '

21 Fuel Bundles .

.- t :

L. - c.E'h.2 Dummy Bundles [in NE . 1 Partial (5 plate) Fuel Bur.dle (SW o.f SW Box [

". 'Posi. tion in NE Box and 1 Partial (5 plate) Dummy Bundle (SW of 8W Bos 3%,fSEPositioninSEBox] 2 Dummy Bundles'(NE of NE Box & SE of $E Box 3 p'..;.- . .

~

L." 5. Critical' Po'sition for Per ultimate Loading . Critical Position for Final Loading '

{ . . . .

t .55-1 at top (1004 units) . S-1 at top (iOO4 units) '

".S-2 at top..( 995 units)

?

S-2 at top ( 995 units) -

iC.'S-3 at top ( 995 units) S-3 at 599 units 3 R:g. Blade.at 393 units P.eg. ' Blade fully inserted ' ~

t ." .'."; . .

.c .

~ assume here that all the old control blade worth curves remain valid. .Since the fuel in tR

'nultimate loading is merely exchanged in various positions but the same amount is loaded inG the same fuel box locations, the assumption of- constant control blade worth curves is expected D ba nearly valid. For the final loading with one additional partial (5 plate) fuel bundle Eided.to the-SW location in the SW fuel box (west side of'the S-2 blade), the blade worths of-S-1f S-3 an'd :the Regula' ting Blade are not expected to chan.ge niuch as the S-2 worth will take a .

gost of the effect which cancels out in both the penultimate and the. final loading critical psitions since.S-2'is fully removed in each case. -

(.. .c ; -

~.- Excess Reactivity Estimation for the Penultimate Loading -

l MaximumfAvailable Reactivity Reactivity for Critical Configuration 6i- ,

.S-1 at top: .0150 Ak/k S-1 at top:

0.0150 Ak/k g L:S-2 at top:" .0133 Ak/k S-2 at top:

l

'S-3 at top: /- 0.0133 Ak/k'

.0220 Ak/k- S-3 at top: 0.0220 ' Ak/k I RB at topi

. 0098 Ak/k

. Total: .0501 ok/k RB at 393 units: _

Totah 0.0556 Ak/k 0.0053 Ak/k ,

f. control blades worth curves are still valid for penultimate loading, the excess reactivity oaded then is only:

= *

( Ak/k) excess ' (0.0601 '- 0.0556) Ak/k -

=

0.0045 Ak/k ,

=

0.45% Ak/k .

' i ...

ince one of t'he aluminum wedging pins was found not fully inserted following the determinatie f this critical position, this value is probably closer to 0.6% Ak/k. This evaluation; is based on the positive worth of inserting the aluminum vedging pins. 1 .

m - __ _ _ _ _ _ _ _ . _ _ _ _ . _ . _ _ _ _ _ . _ _

cp .* , ,

e - . .

j .

2

,I,.

Excess Reactivity Estimation for the Final Loading:

Maximum Av&ilable Reactivity-

_ Reactivity for Critical Configuration y-S-2 at top:

0.0150 Ak/k _S-l'at top:

3 S-2 at top: 0.0133 Ak/k '0.0150 Ak/k. J . '.

E ,S.3 at top:

S-2 at top: "

0.0133 Ak/k '.'.

0.0220 Ak/k S-3 at 599 units:

~

'0.0183 Ak/k RB at top: 0.0098 Ak/k RB fully inserted: .'

Total: 0.0601 Ak/k 0.000 Ak/k -

I.', -

Total: 0.0466 Ak/k S- -.

'.- ~.e lEstiin;ated TOTAL Excess Reactivity Loaded in the Final Configurat

~

.C . .

- [,,3 c .

(Ak/k) excess , = ~~

(0.0601 - 0.0466) Ak/k ,-

~ N.b = "

(Ak/k) excess 0.0135 Ak/k , ,'

. .;f, X *

=

.' (Ak/k) excess 1.3% @ .

' . With this calculated change in excess reactivity of about 0.35% 'from the'old core loadins

' en if the worth curves are all changed by 0.1% evenly, it is c1' ear that the. excess reacti-ty loaded is less than 1.5% Ak/k, which is well within the 2.3% Ak/k limit 'of the UFTR Technical Specifications and the 2.1% Ak/k . limit set by the Special procedure used to control el ~1oadi.ng. ,, -

II .Exp'erim' ental Estimation of ExcessR ' eactivity . -

Analysis of the subcritical multiplication data and best 1/M plots' generated in reloadint the'UFTR and Mon.itoring indicates- Channel that the 4. two(2) most consistent 1/H plots are those from monitoring Channel The predictions on critical loading for these two monitoring char is are yery consistent from loading to loading and show good agreement. The results of the redicted critical loading for three configurations are summarized in Table 1.-

(- .

. Table 1 -

Prediction of Critical Configurations .

Monitoring Config'uration Monitoring .

Channel 1 Channel 4 S1 and S2 out 22.5 bundles S1,S2 and S3 out- 22.5 bundles 21.0 bundles 21.0 bundles '

S1,S2,53 and RB out 210 bundles 20.75 bundles -

Since the critical configuration 'with the final core loading (21.5 bundles) is with S1 ar Iaoved) fully removed but S3 only 60% removed (approximately two-th we have the following data summary:

.}.

I .

m

f ,.3,

[

I. -

lI' ,

1.

Interpolating.from loading for the actual Tablecrit

~

1, the estimated predicted critical -

., a t 60%) i s. . . . . . . . .. . .. . . ........... . ical bl ade removal (S1,S2.out , S3

~ ................... 21.50 bundles.

, 2.

i From Table 1, the esti~ mated critical load.ing for all blades out iis.. 20.75' bundles' ,

' $.3.. .. Bundle difference rep' resenting loading of excess reactivity.......

I Therefore, the 0.75 bundle difference representing .. .

.5 bundles are loaded, 3/4 bundle represents approximatelythe excess reactivity loading . fsinc.

, 0.75/21.5 = 0.035 = 3.5% change 235" loading.

lealculate an ex' cess Using the mass reactivity reactivity loading as follows:coefficient value of=a" 0.4% Ak/k/% U-235, we no; I ;"_. 3. .s

$A k) re

". ,l "

I *

(% Ak/k) excess=

'= d (% Ak/k/%U-235) x %U-235 change (

I excess '

p.~ . -

d.[

(% Ak/k) excess '=. l A% Ak/k excess reactivity.

hl. Sumary l - ~

~

}. a .

~

hlying upon the. experimentally determined critical data but u value of the mass coefficient, yield results in close agr6ement.

~

It should be noted that the ss coefficient should remain relatively unchanged for this new core loading. The excess ree tivity loaded into the UFTR is estimated to be 1.35'-1 A0% Ak/k which is well within the 2 3

~

dure used to control fuel loading.k limit of the UFTR Technical specifications and .

.~

Reccmmendatiori 1 - .

Ice'After all training requirements are met and all necessary checks' are conducted in a bnerating status, .the UrTR is ready to operate.at the critical I wat rmal ! critical position and give UFTR operators the necessary startups for recertification.

Ien the UFTR will be ready for running at'100 watts to obta for the S-2 Annual R'eactivity 14easurements Surveillance.

Enning stored to to nonnal complete operating surveillance status. will be undertaken after which the UFTR will be able t toercontinueproblems develop. work toward operating ' status with no need to check back with th Y[

F (

h L. O Ohw /9-l5$ 50 Reactor Manager / '

Date -

l

/ !N . J p- l$ febr u >, N Dir~ector of'ttuclear facilTties Date -

l R. . -

. . . _ _ _ m ,

• O e

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• 1 1

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APPENDIX E I ,

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7 .

UNIVER'SITY OF FLORIDA

g. .

Offica cf Administrativs Affdrs . .

, Gainesville,32611 ..

• Environmental H9alth Er Safety Division -

(904) 392 - 1591 Nuctrr Sciences Center -

n .

F .

( . October 8,1985

.. 7'

t. . .

u- ,

.. q, M E M 0 R A N'D U M ~- 1 r,,' -

L. ' ~

~..

TO: William G. Vernetson .

{- Acting Director of Nuclear Facilities - .'

-FROM:

Donald L. Munroe %f h1MA -

U- ,

Radiation Control Officer . --

SUBJECT:

Approval of Increased Weekly' Dose Limi'ts e -

- ? . _. . : . The Radiation Control Coimmittee has approved your request for - -

increased exposure limits during UFTR control blade repair

[ work. 'The allowed exposure limits are: .

whole body 300 mrem / week 10b0 mrem /qtr' - -

extremities 2000 mrem / week 6000 mre~ m/qtr

[ , .

g E . .

T- .

.f

)km

M . w icro .

Imu.

WC.vtansyson.csAcToa MusAcs3 NUCLEAR FACluTIES DMSION - -

puctua aucroa sun.omc

' i UNIVER50Y OF RORIDA IC48Mf5VfLLE.FL0aeO4 32483 MtOME hog 3,2 543 TELEX $4M4 I

January 6, 1986 l 1

! - l

^

Caudle A. Julian, Acting Chief Operatio'ns Branch -

M Division of React'or Safety

• U.S. Nuclear Regulatory Commission, Region II 101 Marietta Street, N.W. '

• P .

Suite 2900 '

L.~ Atlanta, coorgia 30323 -

Dear Sir:

I

~

As explained in a telephone call to Mr. Bruce Wilson on De--

l cember 16, 1985, the unavailability of the University of Floridh Training Reactor (UFTR) since September 3,1985 due to mainten-I~ ance work' and subsequent unloading of fuel in October,1985 has resulted in our inability to meet some of the requalification

} training program requirements delineated in our approved program.

I Specifically the 3-month interval (with one month grace) for startups and shutdowns will have been exceeded for all operators before we are prepared to-reload our fuel. The same is true of the interval for performing weekly and daily checkouts for most of our o[erators, especially since complete checkouts'have not' been possible since the fuel was unloaded in mid-October. There-I

{

fore, as informed by' Mr. Wilson, we will need to certify that the knowledge and understanding of facility operations and admini-stration are satisfactory for each o.f our operators prior to the individual's performance of licensed duties.

I. The enclosed pages of Attachment I describe Lne classroom

{

and practical training that will be used as the basis for certi-I l

fying that each reactor operator at the University of Florida Training Reactor facility has demonstrat.ed that his knowledge and understanding of facility operation and administrationn are sa-I tisfactory pursu' ant to 10 CFR 55.31 Conditions of the Licenses as delineated in Paragraph (e). Attachment I contains a description

} of the scenario requiring these actions, a review of the license-I

)

l maintenance related activities conducted during our extended shutdown, a proposed program for certifying our operations staff, and a final summary of all the Operator Requalification Actions.

The certification program is divided into two primary parts because certain requirements must be met to move irradiated fuel and reload our core (licensed activities) and only then will com '

plete daily and veeldly checkouts be possible as required by our requalification training program. The design of our reactor will require at least one day and probably several days following re-loading to replace shielding and perform overdue checks, tests and surveillances as required by the UFTR Technical Specifica- 1 7

tions before startup operations can be commenced f or measurement of blade worths, etc. With this in mind, prior to commencing op-g _ _ _ . _ _ _ _ _ _ _ _ - _ . _ - - - - - - - -

f.

~

h.' caudie A. Julian January 6, 1986 -

. Page Two ~ '

I a .

erations, we will then require the second part of our certifi.cah

^ '

{.. tion training before internally certifying our. operators for the .

E-

-length 9 task of performing a complete set of reactor operational

' chec1 gouts and tests (many of which require running the reactor at .

part or even full. power) as required by the Tech Speca 'before the - .

reactor can be returned to. normal operations. ': .

r '

- r -

.The work plan for returning.the UFTR to full operable status' *-l L ",, .

. is' enclosed for information purposes in Attachment II. We' expect -

to complete the maintenance.. work by Januar.y 10 barring unforeseen

. .'... problems. We then expect to be ready to reload fuel by January .

16, 1986 af ter completing the recertification training. There-fore, we'are requesting that you review the attached plan and let us know if it is satisfactory. If it is possible to forgo the NRC

{' approval of our certification, since we are committed to meeting -

all of the training requirements delineated in the Recertifica- . .

,#- tion Planj please advise us of this fact. If furt.her information "

{ s is. needed, please let us know.

Sincerely,

• ddlak.A U W -

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William G. Vernetson

• LL Acting Director' of '

• Nuclear Pacilitics ..

WGV/ps Enclosures '

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ATTACIDfENT I

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UFTR REAC'IUR OPERATOR ,

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.RECERTIFICATION PLAN

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January, 1986 '

lI-I. Scenario:

lIi The University of Florida Training Reactor has been in administra?.ive shutdown!

for four (4) months to effect: repairs following.the discovery of a sticking control blade on September 3, 1985. As a result all licensed' Senior Reactor j . ;.y .- operators have' exceeded the maximum allowed interval between performing start -

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ups and o'ther reactivity manipulations as required by the approved Requalifi-cation Training Program. The reactor operator who performed the startup and g.. '

shutdown on September 3,1985 is also likely to be past the allowed interval 3' . prior to restart. Since an SRO must also direct the reloading of the reactor

• core and at least witness the restart following the maintenance performed, it

I is necessary to prepare and certify the UPTR operations staff prior to fuel reload and subsequent 1/M approach to critical as well as restart of opera- .

l . tions for checkout and testing.

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"II . License-Maintenance Related Activities Conducted During Shutdown:

_During the four month shutdown all UFTR staff have continued to participate in

~-all regularly scheduled requalification training. Such training c'onducted dur-ing this period included:

1. ' Emergency Equipment Training, *

2. Review of Annual kctivities Report with special emphasis on all

.I- unusual occurrences for the reporting' year, ,

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3. Participation in Two Emergency Drills. ,

{ In addition special training.was conducted on Standard Operating Procedures l I

' required for fuel movement for the core unloading performed in mid-October. In addition, although all Atems of weekly and daily checkouts could not be per-formed, some members of the staff have performed considerable numbers of. ab-- .

breviated weekly and daily checkouts,ob' th to assure control of the facility and to assure continuing familiarization with equipnent as available. Limited weekly checkouts continue to be performed primarily as relates to radiation protection and control surveillance requirements.

The core unstacking, unloading and core-reflector area work for implementing '

corrective actions to prevent recurrence of the sticking control blade problem has also required staff involvement and intimate familiarity with the design of the system. In addition, considerable staf f time has also been expended in tours and other educational descriptions of the facility to train students.

All of these activities assure that our operations staf f has maintained a pa-tisfactory level of knowledge and understanding of facility operations and ad '

ministration.

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[ III. Proposed Program For Certifying Operations Staff: .

To support the certification that staff personnel have demonstrated that the.ir

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knowledge and understanding of facility oparation are satisfactory, the fol-

• lowing certification program has been developed. A preliminary condition for certification of the UFTR Operations Staff is that the maintenance ' work has I been completed satisfactorily as far as can 'be checked prior to loading fuel.

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A. To Reload Fuel-

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Prior to f0el reload the following training will be completr;d for all li-;

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consed personnel to be certified: "

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1.

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I Review of Work Plan for Restoring UPTR to Operable Con' tachment . II).

• d ition (At-

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2 Training on S'tandard Operating Procedures required within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of beginning fuel movement to effect reload to include practical exercise with fuel handling tools and review of assigned work sta-tions. Subject procedures for training will be:

l a. SOP-C 1, " Irradiated Fuel Handling"

b. SOP-C.2, " Fuel -Loading" .

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3.

c.

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SOP-D.2, " Radiation Work Permit" Classroom review of the fuel loading plans and associated loading diagrams to include planned sequences ~ of fuel element movements and expected defector responses. Classroom review of the instrumentation .

setup for monitoring the approach-to-critical as well as the theory involved in reloading for the approach-to-criM. cal.

4. Classroom review of SOP-A.1, " Pre-operational checks" will be con-I ducted along with a walkthrough checkout of all available console I

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checkouts. Note that 'some checks cannot be performed without fuel

' loaded. To assure operator f amiliarity, all operators will partici-pate in a group performance of a weekly and daily checkout to assure I

j that all operators' are certified in this area of the training pro-gram as far as possible prior to commencing fuel addition to the core.

5. Classroom revicu of Standard Operating Procedures for startup and shutdown to include:

a. Lecture on SOP-A.2, " Reactor Startup," SOP-A.3, " Reactor Opera-tion at Power," and SOP-AA , " Reactor Shutdown."

b. Console walkthrough demonstration of as much of these SOP's as is possible prior to commencing reload. The emphasis here will be on expected instrument responses during the fuel loading and during the subsequent restart.

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L .ctura/Dircutcion en expected observations during a startup to full power followed by shutdown.

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7. Examination on key training points.

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f B. Recertification L ,

, q Each operator successfully' completing the requirements of section' A will'

'be' certified by the Facility Director to resume licensed activities and '

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hence to participate in UPTR fuel reload and approach-to-critical to in _,'

clude fuel movement and console operation. This internal operator recer- !

tification shall then be communicated to.NRC Region II -for approval. This *.

{ approval must be *communir:ated from Region II prior to the certification -

becoming effective to perform licensed activities. ,

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C. Fuel Inspection and Reload: -

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Following NRC approval of recertification of all operators to resume li-

/ censed operations, all certified operators will participate substantially

.in reloading the UPTR core. In addition, all fuel will be ' inspected prior to being added to the core.

I, D. To Commence Reactor Operations:

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, Following completion of fuel reload and replacement of all shielding,. the following training items will be perfonned for each previously cer.tified operator:

I.- 1. .Walkthrough group performance of a complete weekly checkout; by all' l operators with fuel in place. ,

2.

Individual performance .of a complete independent daily checkout by l all operators monitored by another operator.

Note that both'of these items are now licensed activities since fuel will be present in the core.

A classroom review lecture will be held on the specific tests, surveil-lances and checks with associated maintenance procedures and instructions to be accomplished prior to declaring the UPTR fully operable. This re-view lecture will assure all operations staf f are cognizant of the limits to UPTR operations until various Tech Spec tests, checks and surveil-lances are completed and the UPTR is returned to full operablo status.

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E. 'Startup, Reactivity Manipulation and Shutdown: '

Each certified' operator will perform an observed startup to the 1 watt critical' position followed by a controlled shutdown. As shown in the work-

I~ plan ( Attachment II), these startups and shutdowns will serve,the dual purpose of establishing a good cold critical position and checkout of.the system as well as training experience. Note that all checks -tests and I. ' surveillances required by the UFTR Technical Specifications prior to this.

startup as delineated in the Work Plan (Attachment II) will be completed ,

prior to permitting startup of the UPTR and also prior to power ascen-

~s ions where Tech Spec limitations are involved.

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Upon successful completion of all items delineated 'in Sections A - E, ,each op-- *

' erator will be certified by the Facility Director to resume conducting such '

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' .3 ' startup' and power operations as are required and approved to return the UFTR, , '

to -full operational status under the conditions of the respective operator li-censes.~ Complete records of each step will be maintained as part of the re-qualification training program files. In the course af conducting all tests' required to return the UFTR to full operability, each certified operatoi will ,

also conduct at least one startup into the popwer range followed by 'a shutdown I1" -

under direct observation of the Reactor Manager or Facil'ity Director.

Complete documentation of all requalification training will be. maintained as I[- }, ,part of the requalification training program files. .

, .. ,.Iy. Summary of Operator.Recertification Actions:

I A. Prior Condition - Completion of Outage Maintenance Activit;ies.

B. Conduct training required in Section III to allow operator recerti--

fication prior to fuel movement, inspection and core reload.

I- C. Certify all operators to resume licensed act;ivities..,

D. Receive NRC approval of Certification.  ; .

E.

Assure all certified operators , participate substantially ih the core reload activities.

F. Assure all operators perform complete weekly and daily checkouts, some steps of which are licensed activities when fuel is present.

G. Certify each operator (internally by Facility Director) to resume licensed startup and shutdown activities. .

H. Assus e that each certified operator performs a startup/ reactivity ~

manipulation and controlled shutdown under direct observation of the Reactor Manoger or Faci-lity Director. '

I. Maintain complete documentation of all requalification training as part of the requalification training program files. .

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ATTACHMENT II .

(.- WORK PLAN FOR RESIORATION OF UFTR TO -

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~, OPERABLE CONDITION * '

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,, I. Complete Repair and Maintenance Work on Control Blade Drive Sysiten (F-1). '.

(.r ;f II. - Fuel Reload:

A. . Loading Plan .

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1. Develop The Plan *

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', 2. Receive Approval by RM, FD, RSRS

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. B. Staff Recertification Program '

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1.- . Conduct Training Required to Reload Fuel

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2. Recertify Operators by Facility Director ,

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Receive Approval of Recertification from NRC Region II Operator

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• Licensing' ,

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g ;. C. ' Instrumentation Setup *

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Fuel Reload Performance r

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1. Inspect All Irradiated Fuel Elements (B-2)

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2. . Reload All Approved Elements With All Operators Assisting h 3. Document ' Reloac{ .

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III. Test and Checkout Program A. Staff Recertification Program -

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1. Complete' Practical Training "

a. Conduct Weekly Checkout (Group)

'b.' Conduct Individual (Observed)' Daily Checkouts B. Precritiical Tests

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1. Perform Quarterly Calibration Check of Area and Stack Monitors (Q-2)

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2. Perform Annual Replacement of Control Blade Clutch Current Light Bulbs (A-4) .,'

. 3. Measure Control Blade Removal Times (Weekly Checkout)

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4.

Measure Control Blade Controlled Insertion Times (S-5)

5. Measure Control Blade Drop Times (S-1) .

6 Perform Quarterly Check'of Scram Punctions (Q-1)

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C .' C'-itical checks 1.-

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.E:.tablish 1 watt critical position consisting of five startups -

to 1 watt followed by shutdowns (used also' as part of operator .

recertification .

training). *

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Certify each operator to resume licensed duties (Pacility -*

Director). .

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3. Perform: Annual Reactivity Measurements'(S-2): ,

a. Worth of Each Control Blade

' b. Reactivity Insertion Rate of Each Blade c.

Total Excess Reactivity ,

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Shutdown Reactivity (Shutdown Margin) ,

D'. Power Checks *

1. Perform Environmental Cell Surveys During Power Ascension l

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- +C 2. _ Perform Anr.ual Measurement of.UPTR Temperature Coefficient of Reactivity (A-3)- -

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Perform Quarterly Radiological Survey of Restricted Areas (Q-5) l 4. '

Perform Quarterly Radiological Survey of Unrestricted Areas (Q-4) ,

5. Measure Argon-41 Stack concentration (S-4) .

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Measure Dilution Air Plow Rate (S-4 from A-1) ,

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Perform UFTR Nuclear Instrumentation Calibration Chec( and Calorimitric Heat Dal'ance (A-2)

8. Perform Verifications Following Calorimetric

a. Temperature Coefficient

b. Blade Worths IV. Review and Approval of All Results A. Reactor Manager H. Pacility Director C. Reactor Safety Review Subcommittee D. Transmittal of Pinal Report to NRC Region II

h APPENDIX C UFTR STANDARD OPERATING PROCEDURES ORIGINALS AND REVISIONS

1. UFTR SOP-0.3, " CONTROL AND DOCUMENTATION OF UPTR MODIFICATIONS" (REV 0)

2. UPTR SOP-0.4, "10 CFR 50.59 EVALUATION AND DETERMINATION" (REV 1)

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3. UPTR SOP-0.5, "UPTR QUALITY ASSURANCE PROGRAM" (REV 1)

4. UPTR SOP-E.8, " VERIFICATION OF UFTR NEGATIVE VOID COEFFICIENT OF REACTIVITY" (REV 0)

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w-I SOP O.3 PAGE 1 OF 15 UFTR OPERATING PROCEDURE O.3 1.0 Control and Documentation of UFTR Modifications 2.0 Approval ,

Reactor Safety Review Subcommittee . . . . ( gh b2 Date' Facility Director . . . . . . . . . . . . jfktd]l* [ f/ /f)

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I REV 0, 10/85

pr I SOP-0.3 PAGE 2 of 14 3.0 Purpose and Discussion 3.1 This procedure shall be used to control and document all nuclear safety related modifications to the UFTR facility to include reactor safety and control system, reactor protection s y s t e m ,-

radiation monitoring system, and other systems delineated in the I UFT1 Safety Analysis Report. This procedure may be used to docu-ment any other modifications to the facility.

I 3.1.1 Routine preventive maintenance or surveillance operations conducted in accordance with approved procedures or instruc-tions are considered part of routine reactor operations and a r e n o .- intended to be controlled by this procedure.

3.1.2 Procedures and equipment described and used only for specific experiments are controlled by SOP- A.5, and are not required I to be controlled by this procedure: the controls of this pro-cedure may be required by the UFTR administration.

3.2 Specific structures, systems and components whose modification is controlled by this procedure include structures, systems, components whose intended functions are to:

3.2.1 Prevent accidents that could cause undue risk to the health and safety of the public; or 3.2.2 Mitigate the consequences of accidents that could cause undue risk to the health and safety of the public.

I 3.3 Identification of Quality Assurance functional organization responsibilities.

3.3.1 Technical Staff Support (TSS) is a general designator for I technical support of the UPTR staff that includes, but is not limited to, UFTR Staff, University of Florida Radiation Con- ,

trol personnel, Engineering Machine Shop personnel, Nuclear I Engineering Sciences Electronics Technicians, e tc. ; area of TSS expertise is defined by the appropriate job classifica-tion or specialty of the individual. TSS personnel will func-tion to:

3.3.1.1 Identify potential nuclear safety related needs and deficiencies.

I 3.3.1.2 Propose modifications (or independently evaluate proposed modifications to address problems within areas of exper-tise as directed by the Reactor Manager / Facility Director).

3.3.1.3 Perform authorizvd maintenance, surveillance, or Quality

! Assurance functions under the supervision of a licensed UPTR Reactor Operator.

3.3.1.4 Manufacture and/or install equipment for use in planned and approved modifications of the UFTR facility.

, REV 0, 10/85 t

SOP-0.3 PAGE 3 of 14 I 3.3.2 Licensed Reactor Operators will function to:

I 3.3.2.1 Perform authorized maintenance, surveillances, or Quality Assurance functions.

3.3.2.2 Authorize, supervise, and direct approved Quality As-surance activities of.TSS personnel.

3.3.3 Senior Reactor Operators will function to:

E 3.3.3.1 Authorize, supervise, and direct approved Quality As-surance activities of TSS personnel and Reactor Operators.

3.3.4 Reactor Manager / Facility Director will. function to:

3.3.4.1 Indicate specific codes, standards, and regulations to be used or referenced in the phases of the modification.

3.3.4.2 Evaluate the proposed modification to assure that the I modification does not represent an unroviewed safety ques-tion using UFTR SOP-0.4, UFTR Form .S O P -0. 4 A .

I 3.3.4.3 Evaluate results of completed modificaton work and convey results to the Reactor Safety Review Subcommittee.

3.3.5 Reactor Safety Review Subcommittee functions in two separate categories:

3.3.5.1 Category 1, (TSS functions):

3.3.5.1.1 Perform functional evaluation of proposed modification.

I 3.3.5.1.2 Propose solutions to problems whose solutions are con-trolled by this procedure.

4 3.3.5.2 Category 2, (Administrative Functions):

I 3.3.5.2.1 Ensure that proposed modifications have been subjected to sufficient analysis prior to submission for 10 CFR 50.59 evaluation or determination relative to unre-

-I viewed safety questions.

I 3.3.5.2.2 Review evaluations indicating that modifications pose no unreviewed safety question, 3.3.5.2.3 Determine that proposed modifications do not represent unreviewed safety questions, 3 3.3.5.2.4 Review the results of Quality Assurance actions rela-tive to modifications, 3.3.5.2.5 Review and audit records for modifications.

I REV 0, 10/85

SOP-0.3 PAGE 4 of 14 3.4 QA Documentation for Modifications 3.4.1 Documentation detailing Material Procurement shall be main-tained (as applicable) to include:

3.4.1.1 The University of Florida Purchase Order Form and attach-ments.

j 3.4.1.2 Material Specifications, as available from purchase orders.

3.4.2 Materials Control Documentation shall be maintained detailing (as applicable)

3.4.2.1 Material storage method.

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3.4.2.2 Material condition at time of use, general condition and performance capability (functional testing as applicable).

3.4.3 Process control documentation shall be maintained to include references to applicable process control standards and in-structions 'in the procedure designed to control and implement the modifications.

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3.4.3.1 Specific techniques and methods to be used in the work process to accomplish standard tasks shall be indicated in the work procedure.

( 3.4.3.2 Applicable process control standards and instructions I

l generated in house shall have at least two separate, inde-pendent reviews by persons cognizant and capable.of per-forming a critical evaluation of the process control stan-dards and instructions, as indicated by dated signature on the appropriate document.

3.4.4 Documentation of nuclear safety related modifications (de-sign, equipment fabrication, and installation) under the jurisdiction of this procedure shall include:

3.4.4.1 Documentation that the proposed action has undergone necessary review and approval certifying the adequacy of the proposed modification to perform its intended func-tion.

3.4.4.2 Documentetion certifying that the proposed action has un-dergone necessary review to ensure the action involves no unreviewed safety question as specified in UFTR S O P - 0. 4 and recorded on Form S OP-0.4 A.

3.4.4.3 Documentation that the proposed action has undergone necessary review to ensure that it complies with the ap-propriate procedures, codes, standards, and regulations.

REV 0, 10/85

SOP-0.3 PAGE 5 of 14 3.4.4.4 Approved test and check out procedures for assuring the proper implementation of the proposed modification.

3.4.4.5 Approved, drawings and schematics clearly identifying the changes involved in proposed modifications.

3.4.4.6

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Approved procedure or instructions for making the mod'ifi-cation.

I 3.4.4.7 Records and description of tests to be performed upon com-pletion of the modification to assure'that it will perform its intended function.

3.4.4.8 Records of the results of tests performed upon comp 1.etion.

3.4.4.9 Appropriate Reactor Operations Log entries.

.I 3.4.4.10 Appropriate Maintenance Log entries (if applicable).

3.4.4.11 tocumentation certifying that the Reactor Safety Review

.bbcommittee has reviewed the applicable test results and process / procedure performance record for the modification.

3.4.5 UFTR Form SOP-0.3A, "QA Document Checklist for Modification Packages" contained in Appendix A shall be used to assure all necessary QA documentation of modifications is supplied in the modification package.

3.5 Documentation for tests of components or systems under the jurisdiction of this procedure shall include:

3.5.1 Approved test procedure or instruction with documented review indicating that:

I 3.5.1.1 The test procedure or instruction will provide the in- .

tended information about the performance of the modifica-tion.

3.5.1.2 The test procedure poses no unreviewed safety question I (The procedure for testing the system / component following completion of the modification may require an evaluation via S O P- 0. 4 , UPTR Form SOP-0.4 A ) .

3.5.1.3 The test procedure complies with applicable procedures, codes, standards, and regulations.

'l 3.5.2 Appropriate Daily Operations Log entries noting test perfor-E mance and results.

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I SOP-0.3 PAGE 6 of 14 L

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3.5.3 Appropriate Maintenance Log entries.

NOTE: Preventive Maintenance and Surveillances performed

- in accordance with approved Standard Operatign Pro-cedures and instructions may require a Maintenance Log Entry, if used to document the proper perfor-mance of a modification.

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3.5.4 Documented test results.

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3.5.5. Records of Review and Evaluation of Test Results by:

u 3.5.5.1 TSS Personnel.

3.5.5.2 Reactor Manager.

3.5.5.3 Facility Director.

3.5.5.4 Reactor Safety Review Subcommittee.

3.5.6 UPTR Porm SOP-0.3A, "QA Document Checklist for Modification Packages" contained in Appendix A shall be useo to assure all necessary QA documentation for tests of components and.sys-tems is supplied in the modification package.

3.6 Documentation of External Reporting Requirements under the jurisdiction of this procedure shall include:

3.6.1 Record in the Monthly Activities Report indicating successful completion and closeout of the proposed modification.

3.6.2 Record in the UPTR Annual Report submitted to the Nuclear Regulatory Commission.

4.0 Precautions and Limits '

4.1 Material Controls:

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4.1.1 Material obtained for e dedicated use that may suffer degra-dation in an uncontrolled environment will be protectively

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packaged and stored in either the Reactor Use Only locker, or in another access-controlled location with the location tracked in the appropriate system file.

4.1.2 Material shall be examined for condition before use in the UFTR systemr; where possible, performance of individual com-ponents shall be tested prior to installation.

I 4.1.3 Material specifications shall be examined to assure the mate-rial is suitable for use in the UFTR systems, such evaluation noted on the appropriate Maintenance Log Page prior to use.

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I REV 0, 10/85

SOP-0.3 PAGE 7 of 14 4.2 Unreviewed Safety Question Evaluation and Determination I

4.2.1 Nuclear safety related systems shall not be modified without completing a negative evaluation (and declaration if neces-sary) as to whether an unreviewed safety question is involved as per S O P- 0. 4 , " 10C FR 5 0. 5 9 Evaluation and De te rmina tion."

4.2.2 If an unreviewed safety question evaluation of a modification indicates the potential for an unresolved safety question E

g exists, no action shall be performed without a complete de-termination through the Reactor Safety Review Subcommittee that no unresolved safety ques tion exists.

4.2.3 Modifications controlled by this procedure may be performed only if:

I 4.2.3.1 In the judgement of the Reactor Manager and the Facility Director, no potential for involving an unresolved safety question exists (as indicated by completion of the evalua -

tion section of UFTR Form SOP-O.4 A ) , and 4.2.3.2 Proper authorization is granted by the Facility Director, 4.2.3.3 If the proposed modification is evaluated to possess the potential to involve an unreviewed safety question, then further review and approval must be obtained from the I Reactor Safety Review Subcommittee to include a full nega-tive determination as documented on UFTR Form SOP-0.4 A.

I 4.3 All required tests shall be referenced as retest or evaluative requirements in the modification package to ensure proper review prior to and following completion of the modification.

5.0 References 5.1 UFTR Safety Analysis Report 5.2 UPTR SOP-0.2, " Control of Maintenance" 5.3 UFTR SOP-0.4, "10CFR50.59 Evaluation and Determination" 5.4 10CPR50.59, " Changes, Test and Experiments" 5.5 ANSI Standard N402-1976, " Quality Assurance Program Requirements for Research Reactors" 6.0 Records Required 6.1 Reactor Operations Log E 6.2 UPTR Form SOP-0.2A, " Maintenance Log Page" 6.3 UFTR Form SOP-0. 3 A , "QA Document Checklist for Modification Packages" REV 0, 10/85

m SOP-O.3 PAGE 8 of 14 6.4 UFTR Form SOP-O.4A, "Unreviewed Safety Question Evaluation and Determination" 7.0 Instructions 7.1 Proposing a modification to the facility I 7.1.1 Prerequisites I 7.1.1.1 Modifications may be found necessary or desirable for various reasons:

I 7.1.1.1.1 To restore operability following equipment failure with no available identical replacement.

7.1.1.1.2 To improve reliability of the UFTR. facilities.

I 7.1.1.1. 3 To expand the experimental or training capabilities of the UPTR facilities.

7.1.1.2 When the need or desirability'of,a modification is identi-fled, potential solutions should be generated as a TSS function using available technical resources.

7.1.1.3 Potential solutions should initially be evaluated for feasibility and effectiveness by the personnel proposing I 7.1.2 the modification.

Performance W 7.1.2.1 A general outline of the proposed modification shall be prepared to include:

7.1.2.1.1 Justification of the modification, 7.1.2.1.2 Diagrams of the affected system as existing prior to I modification and as proposed following the implementa-tion of the modification, I 7.1.2.1.3 Relevant material requirements and specifications for the affected components and systems, 7.1.2.1.4 Installation procedure detailing points of interface

'I between the proposed modification and the present sys-tem (s), how they will be altered, safety precautions, check points (such as temporary jumper logs, fuse re-

moval logs, e tc. ) , Radiological Control requirements, etc.

I 7.1.2.1.5 Expected post-modification system parameters and speci-fications to be used as verification of the proper im-plementation of the modification, 7.1.2.1.6 Proposed means or method to check performance following the modification.

! REV 0, 10/85 l

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F I SOP-0.3 PAGE 9 of 14 I 7.2 Processing a modification to the facility.

7.2.1 Prerequisites 7.2.1.1 The Proposed Modification shall be submitted to the Reac-I tor Manager or Facility Director for processing and shall be accompanied by a complete detailed description of the proposed modification.

7.2.1.2 The Reactor Manager and/or Facility Director shall examine the description and supporting documentation for the pro-posed modification to determine its consistency with ap-I plicable required codes, standards, technical specifica-tions, and license requirements.

7.2.1.3 The Reactor ~ Manager and/or Faci'lity Director will identify

.I required ' codes and standards that must be incorporated.in-to the installation procedure.

7.2.1.4 The Reactor Manager and/or Facility Director will examine material requirements and specifications to determine that the proposed modification will meet the specific require-ments of the UFTR and the UFTR license.

7.2.2 Performance 7.2.2.1 The Reactor Manager and Facility Director or a designated UPTR staff member shall prepare the Modification Package m (MP) using supplied information; the Modification Package g shall include:

7.2.2.1.1 Overview of the Modification Package, explaining the I purpose of the modification and the method for imple-menting it, g 7.2.2.1.2 Engineering diagram of the affected system (s) as ex-g tant, and a detailed engineering diagram of the modj-fled system, 7.2.2.1.3 Relevant material requirements and specifications for affected systems and components, I 7.2.2.1.4 Detailed installation procedure or instruction in-cluding:

7.2.2.1.4.1 Applicable process control standards and instruc-tions to be used at specific points in the work process; 7.2.2.1.4.2 Specific check points and acceptable inspection and process control standards to be applied during the installation procedure or instruction; REV 0, 10/85

SOP-0.3 PAGE 10 of 14 7.2.2.1.4.3 A list of specific points where installation de-tails, material details, or test results may vary

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I 7.2.2.1.5 at the vising discretion of the Reactor Operator super-the installation.

General outline of a Test Procedure for verifying ac-ceptable performance of the installed modification.

( 7.2.2.2 The Reactor Manager and Facility Director shall consider the Modification Package to evaluate the potential of all aspects of the modification and the Modification Package to represent an Unreviewed Safety Question, using UFTR I 7.2.2.2.1 S O P -0.4 , "10CFn50.59 Evaluation and Determination"; UFTR Form SOP-0.4 A shall be used to document the evaluation.

If the proposed modification is evaluated not to iI possess the potential to involve an unresolved safety question, 7.2.2.2.1.1 The modification may be initiated immediately under i

the following conditions:

1. Completion of the 10CFR50.59' evaluation, and

2. Approval of the Faciity Director to proceed.

7.2.2.2.1.2 The Modification Package (including the UFTR Form I SOP-O.4 A ) shall be submitted to the Reactor Safety Review Subcommittee at its next regularly scheduled meeting (within 60 days).

7.2.2.2.2 If the proposed modification is evaluated to possess the potential to involve an unresolved safety question, the Modification Package shall be submitted to the Reactor Safety Review Subcommittee for a full deter-l mination by the Reactor Manager, Facility Director, and ,

I the Reactor Safety Review Subcommittee as to whether an Unresolved Safety Question is involved. This determina-tion shall be completed before the modification is ap-provad for implementation.

7.2.2.2.3 If the Reactor Manager, Facility Director, and the Reactor Safety Review Subcommittee all determine that I no Unresolved Safety Question is involved, then the modification may be approved for implementation; if the Reactor Manager, Facility Director, or the Reactor Safety Review Subcommittee determines that an Unre-I solved Safety Question may be involved, then Nuclear Regulatory Commission approval shall be required prior to implementation of the modification.

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SOP-0.3 PAGE 11 of 14 I 7.3 Implementing the Modification 7.3.1 Prerequisites 7.3.1.1 An approved Modification Package shall be completed prior I to beginning installation of any modification. UFTR Form SOP-0.3A, "QA Document Checklist.for Modification Pack-ages" in Appendix A shall be used to assure all required QA documentation is supplied with the modification.

7.3.2 Performance / Implementing a Modification I 7.3.2.1 The modification shall be implemented using the proce-dures, instructions, and guidelines of the approved Modi-fication Package to control and document the installation of the modification.

7.3.2.2 When the procedure for installation is completed, or (if required) at the proper check points, the task performance.

shall be documented via a Maintenance Log Page entry.

7.3.2.3 All required tests will be performed in accordance with I UFTR SOP-O.2, " Control of Maintenance," and documented with a maintenance log page.

7.3.2.4 Entries shall be made in the Reactor Operations Log I referencing the Maintenance Log Page used to document the test results as follows:

7.3.2.4.1 When modification work commences, 7.3.2.4.2 As significant steps or results are accomplished.

7.3.2.4.3 When the modification is completed.

7.3.3 Post modification evaluation.

7.3.3.1 The modification shall not be considered successfully com-pleted until the required tests have been conducted to as-sure the proper performance of the modification.

7.3.3.1.1 Prior to commencing tests or maintenance of modifica-I tions controlled by this procedure, the Maintenance Package must contain an independently evaluated pro-cedure for testing the performance of the mo d i fi ca ti on indicating that:

7.3.3.1.1.1 The procedure will accomplish its intended function.

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7.3.3.1.1.2 The procedure complies with applicable codes, stan-dards, and regulations.

7.3.3.1.1.3 The procedure specifies acceptance tests and cri-teria for successful completion of the action.

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I SOP-O.3 PAGE 12 of 14 I 7.3.3.1.1.4 The procedure itself does not involve an unresolved safety question as documented l

7.3.3.1.1.4.1 In a properly completed UFTR Form O.4 A , when the procedure in part or whole proposes tests not previously conducted on the system, or 7.3.3.1.1.4.2 Referencing the test, procedure, or instruction documentin.g that the steps to be performed have previously been conducted.

7.3.3.2 Properly documented test results shall be submitted to the I Reactor Mananger and Facility Director for review follow-ing completion of the tests and shall consist of:

7.3.3.2.1 Test data 7.3.3.2.2 Comparison of the Data with expected results.

7.3.3.2.3 Evaluation of the performance of'the modification.

7.3.3.2.4 Completed UFTR Form S OP-0. 3 A , "QA Document Checklist for Modification Packages" 7.3.3.3 When the Reactor Manager and the Facility Director have I reviewed the data, the test results shall be submitted to the Reactor Safety Review Subcommittee for final review and closcout of the modification.

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SOP-O.3 PAGE 13 of 14 J l

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APPENDIX A

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UFTR FORM SOP-O.3A QA DOCUMENT CHECKLIST i

i FOR MODIFICATION PACKAGES I

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SOP-0.3 PAGE 14 of 14 UPTR FORM SOP-0.3A QA DOCUMENT CHECKLIST FOR MODIFICATION PACKAGES Thio checklist provides a convenient record to assure quality assurance documentation is cupplied with a modification as per ANSI N402-1976, " Quality Assurance Program Requirements for Research Reactors." Note that all items are required to be completely performed or the non-performance justified. Sections I, II and III must be completed prior to beginning I modification work except Section IB which may be postponed depending upon the 50.59 evalua-tion conducted under SOP-0.4.

SECTION I: Administrative Approvals E. Post-Modification Test Procedure to Include:

A. UFTR Form SOP-0.4A , "10CFR50.59 Eval-g uation and Determination" completed 1. Functional Reviews. . . . . . . . .

for: 2. Code /Lic/Std. Compliance...

3. Safety Precautions.........

1. Modification............... 4. Process Controls...........

2. Modification Instal- 5. Check Points / Conditions....

lation Procedure........... 6. Rad Con Requirements . . . . . . .

3. Modification Test / Check 7. Expected Test Results. . . . . .

Procedure / Ins truc tion . . . . . .

SECTION III: Installation-Related Material B. RS RS Approval / Authorization. . . Procurement and Control Requirements C. Enabling Authorization of A. PO with Attachments...........

the Facility Director.........

B. Record of Material Specs......

SECTION II: Description of Proposed Modification C. Acceptable Material Storage Method................

I A.

B.

Reasons for modification. . . . . .

Diagrams and Drawings.........

SECTION IV: Requirements for Modification Installation

1. Extant System.............. A. Record of Installation:

2. System Incorporating Modification............... 1. OPS Log En tries . . . . . . . . . . . .

.I C. Delineation of Material

2. MLP Number (s)..............

! Regmnts and Specs . . . . . . . . . . . . . B. Records of Post-Modification Test Procedure Performance:

-I D. Installation Procedure to Include:

1. OPS Log En tries . . . . . . . . . . . .

1. Integrated Instructions for 2. MLP Number (s)..............

Implementation:

C. Records of Post Modification

n. Functional Review. . . . . . . Te s t Re s u l t s . . . . . . . . . . . . . . . . . .

b. Code / License Review. . . . .

c. Modification Interfaces. D. Review and Evaluation of Post-

d. Rad Con Requiremen ts . . . . Modification Test Results:

e. Check Points............

1. TSS Personnel..............

2. Task Instructions: 2 . Cogn iz an t S R0. . . . . . . . . . . . . .

a. Safety Precautions......

b. Material controls.......

c. Process Controls: IN Mgr Review / Approval Date

1. Task Techniques......

2. S tds/ Re f s . . . . . . . . . . . .

d. Functional Review . . . . . . .

e. Code / License leview..... Fac. Dir. IMview/ Approval Date TCN: 2/86

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SOP-0.4 PAGE 1 of 10 I

L UFTR OPERATING PRCCECURE O.4 1.0 - 10 CFR 50.59 Evaluation and Determination

[ . i 2.0 Approval y

(' Reactor. Safety Review Subcommittee . . . . . . .

., fwi J' T  !'

Da'te

[ Director, Nucle,ar Facilities . . . . . . . . . . [, j' '~ g.['" g f- - 8/22; /

, nate '

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SOP 0.4 PAGE 2 of 10 I 3.0 This procedure (UFTR S OP-0.4 ) addresses the proper review of proposed changes in equipment, systems, tests, experiments or procedures. This procedure assures the proper reviews are obtained in evaluating and making determinations relative to proposed actions to determine whether or not they involve unroviewed safety questions as described in 10 CFR 50.59 and the UFTR Technical Specifications..This. review eni evalua-tion will be referred to as a "10 CPR 50.59 Determination" or "Unreviewed Safety Question Eval.uation and Determination."

4.0 Limits and Precautions 4.1 The originator of a proposed chan'ge in equipment, systems, tests, experiments or procedures should assure that the proposed action is described in sufficient detail to allow proper evaluation as to whether an unreviewed safety ques-tion is involved.

4.2 This procedure is intended to address only changes in equipment, systems, tests, experiments or procedures. This procedure does not address normal maintenance operations to include replacement of failed systems or components with identical items.

4.3 Prior to implementation, proposed changes in equipment, systems, tests, experiments or procedures require review by:

6 4.3.1 Only UFTR Management (Level 2 and Level 3) provided the answers to all questions in Section 7.3 are nega-tive for two reviewers., both Senior Reactor Operators.

4.3.2 UFTR Level 2 and 3 Management as well as the Reactor Safety Review Subcommittee if the answer to any eval-uation question in Section 7.3 is positive.

4.4 UPTR Form S OP-0.4 A , "Unreviewed Safety Question Evaluation I and Determination" shall be used for making all 10 CFR 50.59 Determinations.

4.5 UFTR Form SOP-0.4B, " Supporting Material For 10 CFR 50.59 Determination" shall be used for documenting what support material was used and which items and issues were consid-cred in making a 50.59 Determination.

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I SOP-O.4, PAGE 3 of 10 4.6 The 10 CPR 50.59 Determination shall not be considered complete until:

4.6.1 All required signatures are obtained on Form SOP-0.4 A along with answers and bases for the answers. (Bases used to support the change are indicated on UFTR Form SOP-0.4B.)

4.6.2 Supporting material used in performing the 10 CFR 50.59 Determination is documented on UPTR Form' SOP-O.4 B. Technical references used in the Determination are listed in Section I; specific items and issues considered in making the Determination are listed in Section II 4.7 A positive 10 CFR 50.59 Determination requires submission of.an application to the Nuclear Regulatory Commission for license amendment as per 10 CFR 50.90 before the proposed action can be implemented.

4.8 All outstanding negative 10 CPR 50.59 Evaluations made by the UF'JR Level 2 and 3 Management shall be reviewed within three (3) months by the RSRS to be considered closed is-sues (completed 50.59 Determinations).

5.0 References 5.1 10 CPR 50.59, " Changes, tests and experiments" 5.2 10 CFR 50.90, " Application for amendment of license or construction permit" 5.3 UFTR Safety Analysis Report 5.4 UPTR rechnical Specifications 5.5 UFTR Standard Operating Procedures 5.6 ~UPTR Drawings and Technical Manualc 6.0 Records Required 6.1 UFTR Form SOP-0.4 A (Unreviewed Safety Question Evaluation and Determination) l 6.2 UFTR Form S O P -0.4 B (Supporting Material For 10 CFR 50.59 Determinations) 6.3 Reactor Safety Review Subcommittee Minutes I REV 1, 5/86 I i E

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I SOP-0.4, PAGE 4 of 10 7.0 Instructions 7.1 Requirements for a 10 CPR 50.59 Evaluation and Determina-I tion are contained in the documented responses to two sets of questions as delineated on Form SOP-0.4 A to determine whether a proposed action involves an unroviewed safety question.

7.2 Answers to all questions require addressing the basis for the response whether positive or negative. Note that all I questions must be answered as affirmative or negative; if any doubt exists, the answer shall be affirmative.

7.3 Questions to be answered for making the 10 CFR 50.59 Evaluation are:

7.3.1 Does the proposed action represent a change in the UPTR as described by the Safety Analysis Report? (Al-I- tering the UFTR facilities, systems, or components enumerated, described, or diagrammed in the UFTR Safe-ty Analysis Report) 7.3.2 Does the proposed action represent a change in the procedures described by the Safety Analysis Report?

(Access and Key Control in the Reactor Cell, Standard Operating Procedures, Test and Maintenance Procedures, security Procedures) 7.3.3 Does the proposed action represent a test or other ex-periment not described in the Safety Analysis Report I and not previously performed? (A new experiment, new surveillance).

7.4 Questions to be answered in making the 10 CFR 5 0.59 Deter-I. mination are:

7.4.1 Does the proposed action pose an increase in either the probability of or the severity of an accident or malfunction previously evaluated in the Safety Analy-sis Report? (Failures and malfunctions of components and systems important to safety, nuclear excursions during operation, nuclear excursions during fuel load-ing, safety-control blade system malfunctions, loss of coolant accident, fission product releases) 7.4.2 Does the proposed action pose the creation of a pre-viously unidentified accident?

'I 7.4.3 Does the proposed action result in the reduction of a safety margin as defined in the bases for the UFTR Technical Specifications?

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SOP-O.4, PAGE 5 of 10 7.5 UFTR Form S OP-0.4 B , " Supporting Material For 10 CFR 50.59 Determinations" shall be used to document that the neces-sary specific technical references have been considered for the 50.59 De termination and to document the list of specific items considered in making the 50.59 'De termina-tion. UPTR Form SOP-0.4 B shall be completed prior to.com-pletion of the 10 CFR 50.59 Determination.

7.5.1 Technical References listed in Section I of Form SOP-O.4 B shall include a listing of applicable reference material on which the 10 CFR 50.59 Evaluation and De-termination is based to include:

7.5.1.1 FSAR references

7. 5.1. 2 Technical Specification references

7. 5 .1. 3 Standard Operating Procedure References 7.5.1.4 UFTR Drawing references

7. 5 .1. 5 Technical literature references which could in-clude technical manuals, manufacturer specifica-I tions and data sheets, handbook data tables, or other materials required to suppcrt a 50.59 Deter-mination 7.5.2 Items entered in Section II of Form S OP - 0.4 B shall in-clude a listing to document the specific items and is-sues considered in making the 50.59 Determination and I would include such items as the need for a modifica-tion, material content for in-core component modifica-tions, safety functions of equipment changes under I consideration, protective measures to be implemented prior to making an approved modification and other items necessary to demonstrate that the review for the 50.59 Determination is complete. These items and is-I sues considered would be categorized under three gen-eral areas:

7.5.2.1 Reactor Safety issues, 7.5.2.2 Personnel Safety issues, 7.5.2.3 Facility Safety issues, i l

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SOP-O.4 PAGE 6 OF 10 5 7.6 If al answers t the 10 CPR 50.59 Evaluation in Section E 7.3 are negative, then the 10 CFR 50.59 Determination is negative. A positive (yes) response to any of the Evalua-tion questions in Section 7.3 requires that Section 7.4 be I completed; a positive (yes) response to any of the 10 CPR 50.59 Determination questions in Section 7.4 then indi-cates that the propcsed action does present an unroviewed safety question.

7.7 If a proposed action is determined to involve an unre-I viewed safety question or a change in the Technical Speci-fications, then the proposed action cannot be approved and cannot be carried out as proposed without NRC oermission.

In this case, the Licensee shall submit an application for amendment of the license pursuant to 10 CFR 50.90, "Appli-cation for Amendment of License or Constructien Permit" I

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I SOP-O.4 PAGE 7 of 10 I

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APPENDIX A FORMS FOR DOCUMENTING UNREVIEWED SAFETY QUESTION EVALUATION AND DETERMINATION I UFTR FORM SOP-0.4A UPTR FORM SOP-0.4B I

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w SOP 0.4 PAGE 8 OF 10 UFTR FORM SOP-0.4A UNREVIEWED SAFETY QUESTION EVALUATION AND DETERMINATION Title. Number:

.E '- "e=9onaea to 0"e= tion = "e9' ired for 'o cr" 50 s9 Evaluation 3 (See Section 7.3 ) -

Response Basis for Response 7.3.2.

7.3.3.

Reactor Manager Date 7.3.1.

7.3.2.

Facility Director Date 7.3.1.

7.3.2.

7.3.3.

I RSRS Chairman Date I II. Responses to Questions Required for 10 CFR 50.59 Determination (See Section 7.4):

Response Basis for Response 7.4.1.

7.4.2.

Reactor Manager Date 7.4.1.

7.4.2.

7.4.3.

Facility Director Date 7.4.1 7.4.2.

7.4.3.

RSRS Chairman Date I REV 1, 5/86

SOP 0.4 PAGE 9 OF 10 UFTR FORM SOP-0.4A 7.0 INSTRUCTIONS 7.1 Requirements for a 10 CFR 50.59 Evaluation and Determination are con-tained in the documented responses to two sets of questions as delineated on Form SOP-0.4A to determine whether a proposed action involves an unre-viewed safety question.

7.2 Answers to all questions require addressing the basis for the response whether positive or negative. Note that all questions must be answered as I affirmative or negative; if any doubt exists, the answer shall be affir-mative.

7.3 Questions - to be answered for making tihe 10 CFR- 50.59 Evaluation are:

7.3.1 Does the proposed action represent a change in the UPTR as de-scribed by the Safety Analysis Report? (Altering the UPTR facili-ties, systems, or components enumerated, described, or diagrammed in the UFTR Safety Analysis Report) 7.3.2 ~ Does the proposed action represent a change in the procedures de-scribed by the Safety Analysis Report? (Access and Key Control in the Reactor Cell, Standard Operating Procedures, Test and Main-tenance Procedures, Security Procedures) 7.3.3 Does the proposed action represent a test or other experiment not I described in the Safety Analysis Report and not previously per-formed? (A new experiment, new surveillance) 7.4 Questions to be . answered in making the 10 CFR 50.59 Determination are:

7.4.1 Does the proposed action pose an increase in either the probabi-lity of or the severity of an accident or malfunction previously I evaluated in the Safety Analysis Repcrt? (Failures and malfunc-tions of components and systems important to safety, nuclear ex-cursions during operation, nuclear excursions during fuel loading, I safety-control blade system malfunctions, loss of coolant acci-dent, fission product releases) 7.4.2 Does the proposed action pose the creation of a previously uniden-tified accident?

7.4.3 Does the proposed action result in the reduction of a safety mar-gin as defined in the bases for the UFTR Technical Specifications?

7.5 If all answers to the 10 CFR 50.59 Evaluation in Section 7.3 are nega-tive, then the 10 CFR 50.59 Determination is negative. A positive (yes) response to any of the questions in Section 7.3 requires that Section 7.4 be completed; a positive (yes) response to any of the 10 CFR 50.59 Deter-mination questions in Section 7.4 then indicates that the proposed action does present an unreviewed safety question.

7.6 If a proposed action ic determined to involve an unreviewed safety ques-I tion or a change in the Technical Specifications, then the proposed ac-tion cannot be approved and cannot be carried out es proposed without NRC permission. In this case, the Licensee shall submit an application for l

I amendment of the license pursuant to 10 CPR 50.90, " Application for Amendment of License or Construction Permit" 1 REV 1, 5/86

W I SOP-0.4, PAGE 10 of 10 I UFTR Form SOP.-O.4B I Supporting Material For 10 CFR 50.59 Detennination I. Technical References A. Safety Analysis Report References..........

B. Technical Specification References.........

C. Standard Operating Procedure References....

I D. UFTR Drawing References....................

I E. Technical Literature References............

I F. Other References...........................

NOTE: If a tecb .ical reference is not applicable for the 50.59 Determination, then not applicable (N/A) should be. indicated.

II: Items / Issues Considered for Evaluation / Determination:

A.

B.

C.

I D.

E.

.P.

G.

II.

NOTE: Attachments should be referenced for ease of evaluation.

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I SOP-0.5 PAGE 1 of 28 UFTR OPERATING PROCEDURE O.5 1.0 Quality Assurance Program 2.0 Approval

• 4 Reactor Safety Review Subcommittee . . . . . . .

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/ Date Director, Nucfear Facilities . . . . . . . . . . ,,) ) L4, -l>-- z/2 7/sl.

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SOP-0.5 PAGE 2 of 28 I

3.0 Purpose and Discussion '

3.1 Purpose l 3.1.1 Delineate' requirements of the Quality Assurance program at the UFTR.

I 3.1.2 Delineate licensee responsibilities towards the Quality As-surance program at the UFTR.

3.2 Discussion

.3.2.1 General description of Quality Assurance program of the UFTR.

~

3.2.1.1 Scope - the Quality Assurance program at the UFTR controls:

I NOTE: Routine preventive maintenance or surveillances con-ducted in accordance with approved procedures are considered routine reactor operations, and are not intended to be governed by this procedure.

3.2.1.1.1 All replacement's, modifications, or changes to systems having a nuclear safety related function; 3.2.1.1.2 Material procurement, material maintenance, and mate-rial use for systems having a nuclear safety related function; 3.2.1.1.3 Documentation and control of tests and procedures for systems having a nuclear safety related function 3J2.1.1.4 Documentation of Modifications 3.2.1.2 Applicability -

3.2.1.2.1 The Quality Assurance program applies to physical structures, systems, components whose inter.ded fur.c-tions are:

I 3.2.1.2.1.1 Prevention of accidents that could cause undue risk to the health and safety of the public, or 3.2.1.2.1.2 Mitigation of the consequences of accidents that could cause undue risk to the health and safety of the public.

l 3.2.1.2.2 Specific equipment includes reactor safety and control a system, reactor protection system and radiation mon-itoring systems.

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SOP-0.5 PAGE 3 of 28 3.2.1.2.3 This procedure is not intended to govern the require-monts for Quality Assurance and control of' activities that occurred prior to the inception of this program; however, it should be recognized that documentation and controls that occurred before the inception of this pro-gram meet the intent of the Quality Assurance program.

3.2.2 Licensee responsibilities 3.2.2.1 The primary responsibility of the UPTR administration is the establishment and implementation of this Quality As-surance program, including identification of:

3.2.2.1.1 Bounds of this Quality Assurance Program 3.2.2.1.2~ Specific activities governed by this QA (Quality As-surance) procedure 3.2.2.1.3 Organizations supporting this procedure and their functions 3.2.2.1.4 QA functional organization 4.0 Precautions and Limitations 4.1 Routine preventive maintenance and surveillances conducted in I

accordance with approved procedures are considered routine reac-tor operationi, and are not intended to be specifically governed by this procedure.

I 4.2 This procedure is not intended to govern the requirements for Quality Assurance and control of' activities that occurred prior to the inception of this program; however, it should be recog-nized that documentation and controls that occurred before the inception of this program meet the intent of the Quality Assur-ance program.

5.0 References 5.1 UFTR Safety Analysis Report 5.2 UFTR Technical Specifications 5.3 UPTR Standard Operating Procedures 5.4 UFTR Emergency Plan 5.5 UFTR Physical Security Plan REV 1, 2/86 I

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SOP-0.5 PAGE 4 of 28 5.6 UPTR Operator Training and Requalification Certification Plan 5.7 ANSI Standard N-4 02-1976, " Quality Assurance Program Requirements for Research Reactors" 6.0 Records Required 6.1 Operations Log 6.2 Maintenance Log 6.3 UFTR Form SOP-0.5A, " Material Control Documentation Index" 6.4 UFTR Form SOP-0.5B, " Procurement Document Package coversheet" 6.5 UFTR Form SOP-0.5C, " Process Control Instrtction Coversheet" 6.6 UFTR Form SOP-0.5D, "Special Test Control Coversheet" 6.7 UFTR Form SOP-0.5E, " Annual QA Audit Checklist" 6.8 UFTR Surveillance Data Sheets 7.0 Instructions 7.1 QA functional organization and responsibilities 7.1.1 QA Level 5 - Technical Staff Support (TSS) 7.1.1.1 Definition - Technical, engineering support such as Radia-I tion Control Technicians, NES Staff Engineers, non-li-censed UFTR staff, Physical Plant Engineers, etc.

7.1.1.2 Functions and responsibilities -

7.1.1.2.1 Identify needs and deficiencies and bring them to the attention of the Reactor Manager / Facility Director; 7.1.1.2.2 Propose solutions (or independently evaluate proposed solutions as directed by the Reactor Manager / Facility Director) within areas of expertise; I

7.1.1.2.3 Perform maintenance, surveil _ances, or other QA func-tions under the supervision or at the direction of a licensed UPTR operator; NOTE:

I Areas of expertise are defined as the appro-priate job classification or specialty.

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SOP-0.5 PAGE 5 of 28 7.1.2 QA Level 4 - (UPTR Reactor Operators) 7.1.2.1 Definition - Licensed Reactor Operator.

7.1.2.2 Functions and responsibilitica:

7.1.2.2.1 Identify needs and deficiencies and bring them to the attention of the Reactor Manager / Facility Director; 7 .1. 2 . 2 . 2 Propose solutions (or independently evaluate proposed i

solutions as directed by the Reactor Manager / Facility Director);

7.1.2.2.3 I Perform maintenance, surveillances, or QA functions as authorized; 7.1.2.2.4 Authorize, supervise, direct QA activities of level 5 personnel; NOTE: Areas of expertise include working knowledge of the SAR and standard operating procedures; oper-ational characteristics of the UFTR, associated equipment, and interfacing systems; Title 10, I Code of Federal Regulations; areas defined by the individual's experience.

7.1.3 QA Level 3 - (UFTR Supervisory Personnel) 7.1.3.1 Definition - Senior Reactor Operator I 7.1. 3. 2 7.1. 3. 2 .1 Functions and responsibilities:

Identi fy needs ar deficiencies and bring them to the attention of the steactor Manager / Facility Director; 7.1. 3. 2. 2 Propose solutions (or independently evaluate proposed solutions as directed by the Reactor Manager / Facility Director);

7.1.3.2.3 Perform maintenance, sur 11ances, or QA functions in accordance with the requirements of SOP-0.2, SOP-0.3, or other applicable procedures; I Authorize, supervise, direct QA activities of Level 4 7.1. 3. 2. 4 and 5 personnel.

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SOP-0.5 PAGE 6 of 28 7.1.4 QA Level 2 - (UPTR Administration) 7.1. 4.1 Definition - Reactor Manager / Facility Director 7.1. 4. 2 Funactions and responsibilities:

7.1.4.2.1 Identify needs and deficiencies; 7.1. 4. 2. 2 Propose solutions (or independently evaluate proposed solutions);

7.1.4.2.3 Perform maintenance, surveillances, or QA functions in accordance with the requirements of SOP-0.2, SOP-0.3, or other applicable procedures; 7.1. 4. 2. 4 Authorize, supervise, direct QA activities of Level 3, 4, 5 personnel; 7.1. 4. 2. 5 Indicate specific codes, standards, and regulations to be used or referenced in the QA action; 7.1.4.2.6 Evaluate results of QA action and convey results to Reactor Safety Review Subcommittee.

7.1.5 QA Level 1 - (UPTR Upper Level Administration) 7.1.5.1 Definition - Reactor Safety Review Subcommittee 7.1.5.2 Functions and responsibilities:

7.1. 5 . 2.1 Technical and Staff Support Function:

7.1.5.2.1.1 Perform functional evaluation of proposed actions; 7.1.5.2.1.2 Propose solutions to problems that fall within the range of applicability of this procedure; 7.1. 5. 2. 2 Review and Audit Function:

7.1. 5. 2. 2.1 Evaluate proposed actions to assure the action in-volves no unroviewed safety questions; 7.1.5.2.2.2 Ensure the proposed action has all requisite analy-sis and evaluation prior to performance; 7.1.5.2.2.3 Review the results of the action; 7.1.5.2.2.4 Review and audit the QA program.

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SOP-0.5 PAGE 7 of 28 l l

7.2 QA Program functions and responsibilities 7.2.1 Quality Assurance Documentation 7.2.1.1 Design changes 7.2.1.1.1 Design changes will be controlled by QA procedure SOP-0.3 with attendant required documentation; 7.2.1.1.2 Evaluations and determinations of design changes rela-I tive to whether unreviewed safety questions are in-volved will be controlled by SOP-0.4.

7.3 Tests and Maintenance 7.3.1 Prior to commencing tests or maintenance governed by this procedure, the following documentation must be in place:

7.3.1.1 Procedure with an independent evaluation in.dicating that:

7.3.1.1.1 The procedure will accomplish its intended function 7.3.1.1.2 The procedure complies with applicable codes, stan-dards, and regulations 7.3.1.1.3 The procedure specifies acceptance tests and criteria' that would indicate successful completion of the action 7.3.1.2 An operations log entry and a maintenance log pa'e g ini-tiated for any malfunction or failure.

7.3.2 Procedures that are Standard Operating Procedures do not re-quire a Special Test Control Coversheet (UFTR Form S OP-0.5 D ) ;

I tests and maintenance procedures that have no control under Standard Operating Procedures and that do not represent unre-viewed safety questions are required to have an evaluation as per UFTR Form SOP-0.5D, "Special Test Control Coversheet" contained in Appendix II.

7.3.3 The Reactor Manager (Level 2) will review the procedure and either 7.3.3.1 Reject the procedure, 7.3.3.2 Return the procedure to the originator for alterations or expanded detail, or 7.3.3.3 Perform a 10 CFR 50.59 evaluation per UFTR SOP-O.4 REV 1, 2/86

SOP-O.5 PAGE 8 of 28 7.3.4 The Facility Director will review the procedure and either 7.3.4.1 Reject the procedure, or 7.3.4.2 Return the procedure for further clarification or altera-tion, or 7.3.4.3 Perform a 10 CFR 50.59 evaluation per UFTR SOP-0.4 l 7.3.5 The RS RS (Level 1) will review the 10 CFR 50.59 evaluation and determination with the procedure, (per UFT R SOP-0.4 ) and 7.3.5.1 Accept the procedure as written, 7.3.5.2 Accept the procedure with revisions, 7.3.5.2 Return the procedure to the Facility Director for altera-tion, or 7.3.5.4 Reject the procedure on a specified basis.

7.3.6 When the procedure is approved for performance per UFTR SOP-O.4 (QA Level 2) may. assign the task (to tevel 2, Level 3 or Level 4 personnel) via a Maintenance Log Page entry.

7.3.7 An entry in the daily operations log shall be made refer-encing the Maintenance Log Page.

7.3.7.1 When work commences, lE 7.3.7.2 As significant steps or results are accomplished, and g

7.3.7.3 When the work is completed.

7.3.8 When the work is completed and the results evaluated by Level j 2 personnel, all specified documentation will be submitted to I the RS RS for review, and the action will be considered com-pleted when the proper information is entered in:

7.3.8.1 The maintenance log page, with all appropriate informa-tion appended, I

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SOP-0.5 PAGE 9 of.28 7.3.8.2 If applicable, in the modifications file.

NOTE: ' Preventive Maintenance Surveillances performed in accordance with approved Standard Operating Proce-dures: 1) do not require a Maintenance Log Entry; SOP-0.2 specifies those actions requiring a Main-tenance Log entry; and 2) are not intended to be ad-dressed by this procedure except in general QA con-trols, having alread.y received the appropriate con-sideration and control action in SOP-O.2.

7.4 Material Controls 7.4.1 Material Procurenent - documentation prepared by Level 2, 3, or 4 7.4.1.1 As the need is identified for specific m'aterial to be pur-chased or obtained, a Procurement Document Package will be generated detailing: 1) date; 2) storage location; and 3) a Material Control Code that indicates ultimate material use.

NOTE: The Material Control Code should be formatted as per the Material Procurement Document Abbreviations listed in Table I of Appendix I. A documentation in-dox of Material Control Codes and order / billing dates are maintained using UFTR Form SOP-O.5 A (Mate-I rial Control Documentation Index). UFTR Form SOP-0.5B (Procurement Document Package Coversheet) should be used to control the various pieces of documentation required for procurement records.

7.4.1.1.1 For some materials, such as those manufactured in house or on campus, a general description of the material composition, quality, and processing or manufacturing may be sufficient.

I 7.4.1.1.2 For materials purchased from commercial sources, the package should contain 7.4.1.1.2.1 An attached or written catalog description plus per-tinent operational and material specifications; 7.4.1.1.2.2 Copy of the University of Florida purchase order completed for the item (if a PO is used);

7.4.1.1.2.3 Justification for potential or possible usage; REV 1, 2/86 I

r SOP-0.5 PAGE 10 of 28 7.4.1.1.2.4 As the material arrives, the package will be ex- i panded to include the following:

7.4.1.1.2.4.1 Additional specifications, as available from l packing, packaging mate' rial or personal communi- ]

cations with a vendor representative; 7.4.1.1.2.4.2 The material, on arrival, will be tagged or marked to bear the designation Month-Day-Year L (material control code) to be placed with the l package; 7.4.1.1.2.5 When the material is used, the purpose and location of the use will be indicated on the package in such a way that the capability exists of finding the ma-terial from the description alone.

7.4.2 Material Control - documentation prepared by Level 2, 3 or 4 l 7.4.2.1 Material obtained for a dedicated use that may suffer de-gradation outside of a controlled environment, such as electronic components, will be maintained in the Reactor j Use Only locker or equivalent storage location.

7.4.2.2 Material in the Reactor Use Only locker will be maintained i in a dust inhibiting container (s'uch as plastic wrapping),

l and the container snall be marked as indicated in 7.4.1.1.2.5.

l 7.4.2.2.1 Material with a non-dedicated material control code I shall have an indicated storage location marked on UFTR Form SOP-0.5 B , and shall be maintained in a manner that will minimize potential for material degradation within reasonable limits 7.4.2.3 As material is used (or committed to use) in a system or component, a copy of the material procurement package will be placed in the appropriate file in the Equipment a r.d Systems section in the reactor cell file cabinet 7.5 Process Control - prepared by Level 1, 2, 3, 4 or 5 7.5.1 Applicable process control standards and instructions will be referenced in the procedure designed to accomplish the pro-posed action and shalI he reviewed by Level 2 personnel as part of the Maintenance Log Page RE V 1, 2/86

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SOP-0.5 PAGE 11 of 28 7.5.2 Applicable process control standards and instructions may be industrial standards, military standards, or procedures, in-structions, checklists, checkpoints, inspection specifica-tions either selected from sources external to the UFTR or generat~ed in-house 7.5.2.1 Applicable process control standards and instructions to be used at speci.fic points in the work process shall be indicated in the work procedure 7.5.2.1 Applicable process control standards and instructions generated in-house shall have at least two separate, in-dependent reviews by' persons cognizant and capable of per-forming a critical evaluation ef the process control scan-I dards and instructions, as indicated by signature on the appropriate document 7.5.3 Process control standards / instructions should be. maintained in a Process Control Instruction Notebook with a Precess Con-trol Instruction Coversheet (UFTR Form SOP-0.5C), maintained I to assure each process control instruction is complete in-cluding review and approval by Level 5. UFTR Porm SOP-0.5C is contained in Appendix II.

7.6 Test control 7.6.1 The monthly report to the Reactor Safety Review Subcommittee shall contain entries delineating 7.6.1.1 Surveillances performed during the reporting month, 7.6.1.2 Surveillances postponed from the previous reporting period, 7.6.1.3 Surveillances due to be performed during the next report-ing month.

I 7.6.2 For the purposes of this Quality Assurance program, test will be classified as:

7.6.2.1 Technical Specification Surveillances, 7.6.2.1.1 For information purposes, an inventory of scheduled (required) UFTR surveillances, tests and checks with file folder designations (letter-number) is contained in Appendix IV.

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W-SOP-O.5 PAGE 12 of 28 J

7.6.2.2 Other Commitments 7.6.2.2.1 Externally Generated Requirements ( N RC , RS RS , etc.);

7.6.2.2.2 Internally Generated Requirements ( Reactor Manager memo, Good Practice, e tc.) .

7.6.3 Test control procedures are classified according to the con-trolling document as follows:

7.6.3.1 Standard operating Procedures (SOP) 7.6.3.1.1 Standard Operating Procedures represent integrated test procedures receiving RS RS approval; NOTE: Routine preventive maintenance or surveillances conducted in accordance with approved procedures are considered routine reactor operations, and are not intended to be governed by this pro-cedure, although it should be recognized that the Standard Operating Procedures fulfill all the re-quirements of the Quality Assurance program.

7.6.3.1.2 Standard Operating Procedures are maintained in the Standard Operating Procedures Manuals.

7.6.3.2 Auxiliary Operating Instructions I 7.6.3.2.1 Auxiliary Operating Instructions are instructions for specific tasks, operations and tests considered to be routine and commonly performed but requiring specific documentation for:

7.6.3.2.1.1 Re f e re n c e purposes, 7.6.3.2.1.2 Specific acceptable methods of task performance.

7.6.3.2.2 Auxiliary Operating Instructions are reviewed and ap-proved by the Facility Director following an indepen-dent review by QA Level 5 personnel.

l 7.5.3.2.3 Auxiliary Operating Instructions will be maintained in I a notebook on the control console j 7.6.3.3 Special Test Procedures 7.6.3.3.1 Special Test Procedures is a general term for pro-cedures recorded and performed as part of Maintenance Log Pages, Radiation Work Permits, Special Facility Me-moranda, etc. l 1

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SOP-0.5 PAGE 13 of 28 7.6.3.3.2 Special Test Procedures not previously implemented l shall have a 10 CFR 50.59 evaluation per SOP-0.4 made  ;

prior to implementation.

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7.6.3.3.3 Special Test Procedures include guidance relative to:

7.6.3.3.3.1 Identification of items that are nuclear safety re-lated, 7.6.3.3.3.2 Documentation of performance and performance evalua-tions (as applicable),

7.6.3.3.3.3 Specifications delineating:

7.6.3.3.3.3.1 Required test frequency, 7.6.3.3.3.3.2 Parameters required for successful completion, 7.6.3.3.3.3.3 Prerequisites, cautions and warnings.

7.6.3.3.4 All Special Test Procedures are maintained in appro-priate files (special tests with limited applicability will be maintained with the controlling maintenance log page or RWP, special tests with general or repeated ap-plicability will be maintained with the Auxiliary Op-erating Procedures) with attached Special Test Control Coversheet (UPTR Form SOP-0.5D) which is used to assure the Special Te'st Control Procedure is complete includ-ing proper review and approval to include Level 5 and the RS RS. UPTR Form SOP-0.5D is maintained in Appendix II.

7.6.3.4 Surveillance Data Sheets 7.6.3.4.1 Surveillance Data Sheets (SDS) are independent docu-ments that are used to direct and record the perfor-mance of a specific surveillance or task.

7.6.3.4.2 Surveillance Data Sheets use previously approved or ac-cepted methods to direct task performance, and provide I an integrated consistent and/or performance of the format for recording the data task (s).

I 7.6.3.4.3 All current approved / accepted UPTR Surveillance Data Sheets are maintained in Appendix V.

7.7 Control of Measuring and Test Equipment 7.7.1 All required test equipment will be available REV 1, 2/86

SOP-0.5 PAGE 14 of 28 7.7.2 All test equipment to be used in procedures governed by this program shall:

7.7.2.1 Be checked for operability and accuracy, in accordance with a method generated in-house - the method shall be documented as a procedure or instruction with at least two I separate, independent reviews by persons cognizant and capable of performing a critical evaluation of the equip-ment, equipment use, and method of checking the accuracy and operability of the equipment as indicated by signature I on the appropriate document, or 7.7.2.2 Be~ operationally checked against an independent source or device, or 7.7.2.3 Have a calibration performed at intervals specified or recommended by the manufacturer by methods specified or recommended by the manufacturer.

7.8 Audits 7.8.1 QA audits will be performed yearly by the RS RS at intervals not to exceed 15 months.

I 7.8.2 Documentation of the performance of the audit shall consist of the inclusion of the QA audit in the previously es-tablished audit format.

7.8.3 Audit procedure 7.8.3.1 The following areas shall be specifically examined by re-viewing all required documentation for randomly selected QA action items:

7.8.3.1.1 Facility Operations to include 7.8.3.1.1.1 Tech Spec Surveillance Requirements 7.8.3.1.1.2 Documentation of Experiments 7.8.3.1.1.3 Health Physics Records 7.8.3.1.1.4 Fire Protection Records 7.8.3.1.1.5 Special Nuclear Material Records 7.8.3.1.1.6 Maintenance Records 7.8.3.1.1.7 Control of Modifications RE V 1, 2/86 Il 4 1

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'I SOP-0.5 PAGE 15 of 28 7.8.3.1.1.8 Procurement Control Records 7.8.3.1.1.9 Material Control Records I 7.8.3.1.1.10 Process Control Records l

7.8.3.1.1.11 Special Test control Records 7.8.3.1.1.1.2 Records for Control of Measuring and Test Equipment 7.8.3.1.2 Quality Assurance Program 7.8.3.1.2.1 Program Implementation 7.8.3.1.2.2 Document Control Records 7.8.3.1.3 Requalification Training Program Records 7.8.3.1.4 Facility Emergency Plan to include 1

7.8.3.1.4.1 Emergency Response Plan 7.8.3.1.4.2 Implementing Procedures 7.8.3.1.4.3 Records of Implementation 7.8.3.1.5 Facility Security Plan to include 7.8.3.1.5.1 Physical Security Records 7.8.3.1.5.2 Safeguards-Type Records 7.8.3.1.6 Response to previous audit findings 7.8.3.2 As each area is a'udited, the auditor should record results on UPTR Form SOP-O.5E.

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v SOP-O.5 PAGE 16 of 28 lI

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APPENDIX I I

MATERIAL CONTROL DOCUMENTATION I

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SOP-O.5 PAGE 17 of 28 TABLE I MATERIAL PIOCUREMENT DOCUMENT ABBREVIATIONS Standard material procurement abbreviations used to specify the .naterial con-trol code may be formatted in three parts as follows:

FOINAT - (Debignation for Type of Use, General Use Category, System for Intended Use)

Applicable abbreviations for the three parts of the format statement are de-fined below. Note that the abbreviations listed here are not considered to constitute a complete list.

I. DESIGNATION D - Dedicated Use ND - Non-Dedicate'd Uce C - Consumable II. GENERAL USE A. Dedicated /Non-Dedicated Codes:

PS - Power Supply REL - Iblay BAT - Battery I AMP - Amplifier SW - Switch DET - Detector COM - Component REC - Recorder WIR - Wiring D. Consumable Codes:

R - Radiation Control Supplies H - Ilardware L - Lab Supplies P - Plumbing Supplies J - Janitorial Supplies I. M - Machine Shop Supplies E - Electrical Supplies 0 - Other (should be specified)

I III. SYSTEM PCS - Primary Coolant System INS - Radiation Monitoring System I SCS - Secondary Coolant System STS - Shield Tank System RSS - Ihactor Structure System TRS - Temperature Recorder System PRS - Pneumatic Rabbit System RPS - Ibactor Protection System I RVS - Reactor Vent System ESC - Emergency Support Center FSS - Facility Security System CBDS - Control Blade Drive System PSS - Physical Security System BPI - Blade Position Indicator Svstem NI-SC1 - Safety Channel 1, Nuclear Instrumentation NI-LC - Nuclear Instrumentation (Linear Channel)

'I REC-LIC - Two Pen Recorder (Linear Channel)

REC-LOC - Two Pen Recorder (Log Channel)

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SOP-0.5 PAGE 18 of 28 F

L UFTR FOIN SOP-0.5A l e

Material Control Documentation Index F

L Date of Order / Billing Date Material Control Code 7 1.

L 2.

L 3.

4.

5.

6.

7.

[ 8.

9.

10.

, 11.

l 12.

l 13.

14.

l 15.

16.

17.

l 18.

19, 20.

21.

22.

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v SOP-0.5 PAGE 19 of 28 UPTR FORM SOP-0.5B

~ Procurement Document Package Coversheet All blanks except IIIC should be completed prior to storage; indicate NA where non-applicable applies.

I. Item Designation

'I A. Material Control Code................................

I B. Description of Item (s)...

C. Specifications

1. Required . . . . . . . . . . . .

2. Desired.............

I D. Intended Use.............

II. Order Information A. Purchase Order Number........................

I B.

C.

D.

Catalog Information..........................

Packing List.................................

Vendor Communications........................

I E.

F.

Estimated / Actual Cost........................

Previous Supply Parameters 1.

Purchase Date...........................

/

2. Previous Cost...........................

3. Previous Quantity Ordered...............

4. Previous / Alternate Supplier............. /

III. Controls Af ter Receipt A. Acceptability................................

I B. Storage Requirements 1.

2.

Location................................

Environmental controls.................. -

C. Date of Use/ Removal Proa Inventory. . . . . . . . . . .

Originator Date HM/FD Review / Approval Date I REV 1, 2/86

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SOP-0.5 PAGE 20 of 28 P!

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lI APPENDIX II ll i

PROCESS AND TEST CONTROL COVERSHEETS 2

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SOP-0.5 PAGE 21 of 28 UPTR FOR4 SOP-O.5C Process Control Instruction Coversheet One copy of this Process Control Instruction Coversheet and approved material is to be maintained in the Process Control Notebook.

I. Title / Designation of Process Control Instruction:

II. Purpose of Process Control Instruction:

I III. Origin of Process Control Instruction:

f A. Type of Source (Check One)

1. ' Equipment Manufacturer......

2. System Manufacturer.........

I 3.

4.

S.

Industrial Standard.........

Military Standard. . . . . . . . . . .

Generated In-House..........

D. Type of Instruction (Check One)

1. Material Included as Copied From Reference . . . . . . . . . . . . .

2. Material Included as written...........................

3. Manual Pages Included..................................

IV. Evaluation of Process Control Instruction A. Applicability of Process:

D. Evaluation and Comments:

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V. Process Control Instruction Review and Approval A. Level 5 or above Originator Review / Approval Date B. Level 2 Rx Mgr/Fac Dir Review / Approval Date REV 1, 2/86 I

SOP-0.5 PAGE 22 of 28 1

UFTR FORM SOP-O.5D l Special Test Control Coversheet I I. Title / Designation of Test Procedure:

I II. Reason for Generating Test Procedure:

l NOTE: If test procedure is generated due to a failure, occurrence of

'E failure must be recorded in operations log.

III. Results of Unreviewed Safety Question Evaluation

( UFT R Fonn SOP-O . 4 A ) : . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

IV. Test Procedure Evaluation Categories:

I A.

B.

C.

Functional Evaluation Compliance With Codes, Standards, and Regulations Specification of Acceptance Criteria V. Test Procedure Review and Approval:

I Functional Evaluation Compliance Specification of

/teceptance Crit.

A.

Originator 'Date B.

Rx. Manager Date C.

I D.

Fac. Director Date RSRS Chairman Date REV 1, 2/86 I

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APPENDIX III l l

ANNUAL QA AUDIT CHSCKLIST I i I

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SOP-0.5 PAGE 24 of 28 UPTR FOR4 SOP-0.5E Annual QA Audit Checklist E- Audit Area Date Performed Auditor Signature I. Facility Operations

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A. Tech Spec Surveillance Requirements B. Documentation of Experiments I

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D.

E.

Health Physics Ibcords Fire Protection Records Special Nuclear Material Records F. Maintenance Records ~

G. ' Control of Modifications

1. Unreviewed Safety Question l Evaluations and Determina- l tions'(Form SOP-0.4A)

, 2. QA Document Checklist I for Modification Packages (Form SOP-0.3A)

H. Procurement Control Documents

( I. Material Control Documents I J.

K.

L.

Process Control Documents Special Test Control Documents Correspondence / Commitments to NRC II. Quality Assurance Program A. Program Implementation B. Document Control Records t

l III. Requalification Training Program Records IV. Facility Emergency Plan A. Emergency Response Plan B. Implementing Procedures C. General Implementation l V. Security Plan A. Physical Security Records B. Safeguards-Type Records VI. Response to Previous Audit Findings Reactor Manager / Facility Director Review Date i IGIG Review / Approval Date REV 1, 2/86

SOP-0.5 PAGE 25 of 28 I '

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l APPENDIX IV INVENTORY LIST OF SCHEDULED UFTR SURVEILLANCE / TEST / CHECK DESIGNATIONS I

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SOP-0.5 'PAGE 26 cf 28 UFTR SURVEILLANCE / TEST / CHECK DESIGNATIONS

- Daily Preoperational Checks

-- Weekly Preoperational Checks (

[ Q-1 Quarterly Check of Scram Functions.

Q-2 Calibration Check of Area and Stack Radiation Monitors.

{

Q-3 Quarterly Radiological Emergency Evacuation Drill.

b Q-4 Quarterly Radiological Survey of Unrestricted Areas.

Q-5 Quarterly Radiological Survey of Restricted Areas.

S-1 Measurement of Control Blade Drop Times.

[ S-2 Annual Reactivity Measurements (Worth of control Blados, Total Excess Reactivity, Reactivity Insertion Rate and Shutdown Margin).

S-3 Semi-Annual Inventory of Special Nuclear Material.

S-4 Measurement of Argon-41 Stack Concentration.

S-5 Measurement of Control Blade Controlled Insertion Times.

S-6 UFTR Semi-Annual Security Plan Key Inventory.

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S-7 Semi-Annual Check (Replacement) of Security System Batteries.

S-8 Semi-Annual Leak Check of Neutron Sources.

S-9 Semi-Annual Replacement of Well Pump Fuses.

A-1 Measurement of Dilution Air Flow Rate (now part of S-4).

A-2 UFTR Nuclear Instrumentation Calibration Check and Calorimetric Heat Balance.

A-3 Annual Measurement of UFTR Temperature Coefficient of Reactivity.

A-4 Annual Replacement of Control Blade Clutch Curren.t Light Bulbs.

A-5 Emergency Call List Check Form.

[- B-1 Biennial Check to Assure Negative UPTR Void Coefficient of Reactivity.

B-2 Biennial Inspection of Incore Reactor Fuel Elements.

V-1 Five Year Surveillance Inspection of Mechanical Integrity of Control Blades and Drive Systems.

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APPENDIX V INVENTORY OF APPROVED SURVEILLANCE DATA SHEETS INCLUDING Q-1. . . 4/84 S-1... 6/84 A-2... 3/84 Q-2... 9/85 S-5... 3/84 A-4... 2/86 Q-3... 12/83 I Q-4. . . 10/28/82 Q-5.... 11/82 S-8.... 10/85 S-9... 1/86-A-5... 11/85 B-2.... 1/86 I .

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I UFTR QUARTERLY #1 (Q-1 )

SCRAM CHECKS D;te: Date of Last Checks:

WARNING: When any of the following checks opens the Dump Valve or results in shutting off the Primary Coolant Pump, the Dump Valve will be opened (UFTR SOP A.4, Section 7.10.2 ) , Primary Coolant Pump will be shut off, and the system permitted to com-pletely drain (3 minutes) before proceeding further with these checks. 'Ihis de-lay is to preclude breakage of the rupture disc.

NOTICE: For the scram checks, jumper connections are made at the small terminal box ac-cessible at top left of console rear center panel af ter rear door has been re-moved. Make and unmake connections in the order listed to minimize probability I of electrical shorts, or shocks to personnel. Use the special short jumper leads. Replace tertainal box cover upon completion of checks.

A. Procedures and Results:

1. CORE VENT FAN power loss: Raise any blade about 25 units. Shut off Vent Fan for scram. Restart core vent fan. (scram)

2. DILUTION FAN power loss: Insert scram check test adapter under Relay K-

11. Use switch on adapter to bypass core vent fan scram by shunting con-tacts 6 and 7. Raise any blade about 25 units. Shut off Dilution Fan for scram. Restore Relay K-11 to normal. Restart Dilution Fan. (scram) 3.

I PRIMARY COOLANT PUMP power loss: Jumper TB 2-4 to TB 1-4 to bypass PC flow scram. Jumper TB 2-3 to TB 1-3 to bypass PC low level scram. Raise blade about 25 units. Shut off PC Pump for scram. Cycle console power-on switch to open dump valve to permit system to drain. Depress PC Pump switch. Remove jumper connecting TB 1-3 to TB 2-3. Leave jumper connect-ing TB 2-4 to TB 1-4 in place. (scram)

I 4. PRIMARY COOLANT LEVEL loss: Insert test adapter under relay K-8. Shunt contacts 6 and 7 to bypass PC Pump scram. Raise blade about 25 units.

Shut off PC Pump to initiate scram. Cycle console power-on switch to open

• I dump valve and permit system to drain. Depress PC Pump switch. Remove jumper TB 1-4 to TB 2-4. Leave test adapter in place. (scram)

5. PRIMARY COOLANT FLOW loss (inlet line sensor): Jumper .TB 12-2 to TB 1-4 to bypass return line flow scram. Jumper TB 2-3 to TB 1-3 to bypass pri-mary coolant low level scram. Raise any blade about 25 units. Raise red primary coolant flow scram set point on console PC Flow Meter to flos I point for scram. Restore flow scram set point to correct setting (30 gpm). Remove jumper TB 1-4 to TB 12-2. Leave test adapter in place.

(scram)

PRIMARY COOLANT FLOW loss (return line sensor): Jumper TB 12-2 to TB 12-1 to bypass fill line flow scram. Raise any blade about 25 units. Shut off PC Pump for scram wl'ich occurs in about 40 seconds , when return line has I drained. Open the dump valve by cycling console power-on switch. Depress PC Pump switch. Remove all jumpers and restore relay K-8 to normal.

(scram)

(time in seconds)

I (Rev. Apr. 84)

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I 6. NEUTRON CHAMBER HIGli VOLTAGE RE. DUCTION:

a. 10% Drop in Neutron Chamber High Voltage (W/R Drawer):

Raise 2 blades about 25 units. Pull W/R Drawer forward about 12 inches and depress W/R Drawer High Voltage Test Switch for scram and water drop. Open dump valve by cycling console power -on switch to drain system, 'and depress PC Pump switch. Re-insert W/R Drawer.

(water dump and scram)

b. 10% Drop in Neutron Chamber High Voltage (Safety Channel #2):

Open right rear console door. Raise 2 blades about 25 units. Reach-over rear swinging panel and depress Safety Channel #2 High Voltage

'l W Switch for scram and water drop. Restore rear panel. Open dump valve by cycling console power-on switch. Depress PC Pump switch.

(water dump and scram) 7 .- SIIIELD TANK LOW WATER LEVEL:

a. Remove hooks from crane sling. Attach sling to lifting lugs on

I shield tank shield block by using the shackles. Remove shield block and place on southeast corner of concrete reactor structure (not on-to the steel bridge). Remove shield tank aluminum cover.

b. Raise any control blade 25 units. Mark water level on switch body as a reference. Loosen clamp (7/16" wrench is required) and slowly I raise assembly out of the water. Check that water level on switch body at scram corresponds to level on detector. (scram)

c. Restore switch to normal.

NOTE: Check water level at this time and make up demineralized water if needed. Enter start time of water makeup into operating log.

CAUTION: Do not overfill tank. One inch of water equals 14.7 gallons '

I of water, and at 1 gpm, takes 14.7 minutes. Enter stop time of water makeup into operating log when water makeup is completed.

8. CONTROL CONSOLE ELECTRICAL POWER loss:

a. Raise all control blades about 25 units. Turn off console power by I depressing the console power ( power -on ) green lighted switch. Re-store power and verify that reactor is in scram condition (all scram lights illuminated), and that all control blades are at bottom limits.

b. Restore Power (all rods on the bottom). (water dump and scram)

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'l B. Completion of Checkout and Restoring Reactor to Operating Condition:

1. Replace aluminum cover ~on small terminal box......................

2. Replace all control console rear doors............................

3. Replace shield tank cover and shield block............... ........

4. Record quantity of water added to shield tank.....................

C. Comments:

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Performed by Date Acknowledged / Reactor Manager Date (Rev. Apr. 84)

l UFTR QUARTERLY #2 (Q-2)

Area and Stack Monitor Calibration, Date: Date of last check:

l Source Isotope

. Serial Number Activity I

Procedure: Calculate the distance from the center of the source to the I center of the detector for radiation fields of 1 mr/hr, 2.5 mr/hr, 10 mr/hr and 25 mr/hr. If needed, make calibration ad-justments at 1 mr/hr and 25 mr/hr. Note that calibration ad-justments require a maintenance log page and Reactor Manager authorization.

Set trip #2 alarm set point at 2.5 mr/hr and trips #1 alarm set point at 10 mr/hr. (These settings are conservative relative to trip settings allowed in UFTR Technical Specifications, Table 3.3.)

I

References:

Gulf General Atomic Radiation Monitoring System Manual Radiation Control Procedures UFTR Technical Specifications: Section 3.4.1, Table 3.3 and Section 4.2.4 Paragraph (1)

RESULTS Distance Expected East North South Stack (Horizontal) Reading Detector Detector Detector Detector 1.0 mr/hr XXXX 2.5 mr/hr XXXX

• 10.0 mr/hr XXXX 25.0 mr/hr I

XXXX NA 85-90 cps XXXX XXXX XXXX COMMENTS:

I Performed B3-Date I Acknowledged:

Reactor Manager / Facility Director Date 9/85

UFTR QUARTERLY 3 (Q-3)

QUAICERLY EVACUATION DRILL IECORD DATE: DATE OF IJiST DRILL: TIME:

DRILL INITINIOR:

DETAILS OF SIMULATED EMERGENCY (SEE FILE CARD):

,J l

MAJOR ACTIONS OF KEY PERSONNEL IN EVALUATING AND IEDUCING SIMULATED EMEFCENCY:

(NOTE: IDG OF DRILL EVENTS IS ATTAC!! MENT I)

I IECOMMENDATIONS FOR IMPIOVEMENT '(SEE ATTACilMENT II FUR DETAILS):

I-I .

POST-DRILL EMEIEENCY KIT INSPECTION IESULTS (SEE SOP-B.1A APPENDIX II):

COMPIETED BY:

CIIECK OF OPERABILITY OF EMEIEENCY EXITS:

UFTR IEAR DOOR: RADIO-CIIEMISTRY IJJ) EMETEENCY DOOR:

CIIECKED BY:

IEVIEWED: ACKNOWIEDGED:

lI RADIATION CONTIDL OFFICER IEAC'IUR MANAGER / FACILITY DIIEC'IOR IEFEIENCES: SOP-B.1 A TECilNICAL SPECIFICATIONS F OR UPTR, SECTION 4.2.6(3)

I 12/83

{ Q-4 ENVIRONMENTAL MONITORING - UNRESTRICTED AREAS

(

.Date: Qua'rter:

(1 .UFTR Power Level: kw Survey Instrument:

Surveyor (s) :

{ RADIATION LEVELS (lR/hr) i Survey Outdoor Survey Location Location 634 131 24 184 30 557 G-A 1 G-B 2 G-C

{ 3 1-A 4 1-B 5 1-C 6 2-A 7

[ 2-B 8 2-C 9 3-A 10 3-B 11 3-C 12

( 4-A 13 4-D 14 C .

15 5-A

{ \

5-B 10/28/82: bis

_ _LC _ _ _ _ - - _ -- l P'P 3 f 31

g-5 I ENVIRONMENTAL MONITORING - RESTRICTED AREAS Date: Quarter:

UFTR Power Level: kw A)

Survey Instrument (s): B)

C)

(Make and Model) (Serial #)

. Surveyor (s):

SURVEY RADIATION LEVELS (mR/hr)

LOCATION A B C 1

I 3 I

4 5

6 7

8 9

10 6 11 12 13 14 15 16 I 17 18 19 20 21 22 23 I 11/82:bls

f b UFTR SEMIANDUAL #1 (S-1)

Blade Drop Time Checks Date: Date of last check:

Procedure: 1. Set up dual channel Brush amplifier atop reactor console with controls facing to the rear. Set up a two-pen high speed re-corder atop anplifier. Connect the recorder's input cables to amplifier. Place both anplifier and recorder power switches in the "off" position, and connect both power cords into.a source of 120 vac.

(NOTE)

Do not attempt to set up or operate amplifier or recorder without being familiarized with the equipment's operating instructions.

2. Check calibration of amplifier and recorder as per instruc-tions delineated in the operating manuals. Put about two droppers of purple ink into each ink reservoir and prime pens. Run recorder at low speed (gear change knob on side of recorder pushed all the way in) to check flow of ink through pens.

3. Set each amplifier channel on "2 volts per chart line."

4. Using fabricated junction box with its attached test cable and leads, make the following connections:

a. Connections made by mating small flat 3 prong plug to socket.

b. Connect left red and black banana plug pair to channel one input with red to input #2 and black to input #1.

c. Connect right red and black banana plug pair to channel two input with red to input #2 and black to input #1.

I d. Remove appropriate rod drive cable plug either J-7 (Regu-lating Blade), J-8 (Safety 3), J-9 (Safety 2), or J-10 (Safety 1). Insert test plug into the now opened jack, and insert the rod drive cable plug into the jack on the test box.

5. Perform a satisfactory daily pre-operational check as per UP1R SOP-A.1 (Pre-operational Checks) Part iib - DAILY CHECK-LIST if not previously completed this day.... Satisfactory Checks .

R.O. Initial i

6. Raise selected control blade to its upper limit.

7. Start recorder at highest speed. When the chart paper is run-ning, drop the control blade and stop the recorder.

REV 1, 6/84

f (S-1 )

L Page Two F

L 8. Repeat steps 4. d) , 6. and 7. until remaining blades are taken through the procedure, r

( 9. Determine blade drop time:

Count the number of cycles between the upper and lower limits I

of blade travel from the recorder chart paper. Divide the number of cycles by 60 cycles per second to compute blade drop time.

I L

References:

UPTR TECHNICAL SPECIFICATIONS, Brush operating Instructions; Amplifier Models RD 562100, RD 562101 and RD 562102.

F l Results: Upper Blade Lower Blade Drop Time Position Position (Seconds)

Safety #1 Safety #2 l

! Safety #3 Reg. Blade l

Comments:

I l

I l '

1 1

Perfomed by: Acknowledged:

Reactor Manager REV 1, 6/84

UFTR SEMIANNUAL #5 (S-5)

BLADE CONTROLLED INSERTION TIME CHECKS I Date: Date of Last Check:

As part of the UFTR Technical Specifications under Section 4.2.2 Reactor Con-trol and Safety System Surveillance, Paragraph (2) requires that control blade full controlled insertion time shall be measured semiannually, at intervals not to exceed 8 months. The procedure is as follows:

1. Perform a satisfactory daily pre-operational check as per UFTR SOP-A.1 (Pre-operational Checks) Part 7.2 - DAILY ' CHECKLIST if not previously I completed this day. . . . . . . . . Satisfactory Checks R. O. Initials

2. Raise the selected control blade to its upper (full out) limit.

3. Drive the raised blade from its top (full out) position to its bottom (full in) position where the bottom light illuminates, interrupting the down drive while timing the entry using the digital timer available.

4. Repeat Step 3 for each of the three remaining blades.

5. Record data and results in the following table:

Controlled Insertion Upper Blade Lower Blade Time Position Position (Minutes) (Seconds)

Safety #1 Safety #2 safety #3 Reg. Blade I

Performed By Date Acknowledged - Reactor Manager Date I REV 1: 3/84

'I S-8 SEMI-ANNUAL LEAK CilECK OF NEUTRON SOURCES I. Leak Check of Pu-Be Source Date Last Performed: Date of This Check:

i Results of Check:

Performed By:

Name Signature II. Leak Check of SS-Be Source Date Last Performed: Date of Thir Check:

Resalts of Check:

I III. Performed By:

Name Signature Reactor Manager /Pacility Director Acknowledgement Date I REV 1, 10/85

e I

S-9 SDEIANNUAL REPLACEMENT OF DEEP WELL SECONDARY PUMP FUSES Note: All three well pump fuses located in a box on the outside East wall of the building (behind the podocarpus bush) should be replaced with equi-valent 250 volt /60 amp rated fuses. A record of such replacements should be recorded on this continuing form and on the control room status board.

I Manufacturer and Model/

Type / Class Number I Date Fuses Replaced By Designation of Replacement Fuses RM/FD Acknowledgement I

I I

I

I lI I

REV 0, 1/86 l

TCN, 3/86 i L_

Pega 1 ,

UPTR ANNUAL #2 (A-2)

UFTR NUCLEAR INSTHJMENT CALIBRATION CIIECK & IIEAT BALANCE Date Checks Perfortned A. Calibration check of Primary Flow Meter:

Should be 56.6 gpm on test; reads gpm on test; adjusted to gpm.

B. Temperature System Calibration Check:

. 1. Slidewire cleaned & checked (yes or no).

2. Vacuum tubes checked (yes or no); replaced 12AX7, 12BH7

3. Thermocouple / recorder calibration check (TC #7 and #8):

TC *C TC *F Rcdr *F Details of any adjustments made I

I C. Nuclear Instrument Voltage Checks:

Safety Channel #2 Wide Range Drawer Power Supply +15v -15v +1tv +15v -15v +HV -CIC Measured Adjusted To D. lieat Dalance: Water Density:

To = *F 'F lb/ gal *F lb/ gal 86 8.309331 140 8.205414 I Ti=

AT =

• F

• F 95 104 113 8.295353 8.280698 8.263982 149 158 167 8.183332 8.160074 8.135688 I T av = *F 122 131 8.245806 8.226253 176 185 8.110218 8.083688 gal / min x 'F x lb/ gal x 1 Btu /lb ' I I KW = .

56.88 BTU /KW-min x 0.99 m

= KW E. Comments and significant adjustments made:

I Performed By Date Reactor Manager Acknowledgement Date REV 3/84

)

UFTR ANNUAL #2 (A-2) l WIDE RANGE LOG CHANNEL OPERATING VOLTAGE CHECKOUT AND ALICNMENT l

l Instrument Used: Date:

Table 1 l Calib Switch Red Voltage @ TP3,A4, Calib Switch Red Voltage @ TP3,A4, Position Mark # Log Amp & Summer Position Mark # Log Amp & Summer

1. 1 1st #4 4th

1. 2 2nd #5 5th I #3 3rd #6 6th I

Table 2 Performance Check Assembly' Test Calib Switch Voltage Readings Point Position Previous Present A-1, Campbell Rectifier #1 #4 A-1, Campbell Rectifier #1 #5 A-1, Campbell Rectifier #1 #6 A-2, Discriminator & Driver #1 #1 A-2, Discriminator & Driver #2 #3 A-3, Log Count Rate Pumps &

DC Amp #1 #1 A-3, Log Count Rate Pumps &

DC Amp #1 #2 A-3, Log Count Rate Pumps & ,

DC Amp #1 #3 A-4, Log Amplifier & Summer #1 #4 A-4, Log Amplifier & Summer #2 #4 A-4, Log Amplifier & Summer #2 #5 A-4, Log Amplifier & Summer #2 #6 A-4, Log Amplifier & Summer #3 See Table 1

.I A-4, Log Amplifier & Summer #4 #4 NOTf;: Une TP-3 on A2 as Ground Return for Voltage Measurements.

COMMENTS:

I 00 OY Da(e Reactor Manager Acknowledgement- Date REV 3/84 L

F I A-4 REPLACEMENT OF CONTROL' BLADE CLUTCH CURRENT LIGilT BULBS Discussion To prevent failure of control blade clutch current light bulbs during reactor operations and resultant dropping of a control blade as a partial trip, all control blade clutch current light bulbs shall be replaced annually at inter-vals not to exceed 14 months as per the February 7,1984 commitment to the NRC relative to the January 25, 1984 bulb failure. This replacement is considered

-l5 na3or maintenance since the bulbs are part of the reactor control system. This major maintenance is controlled as a surveillance (A-4).

Instructions

1. Record date of last replacement............................

I 2. Record the A-4, replacement operation in the daily operations 1og......................

3. Replace the S-1, S-2 and S-3 clutch current bulbs with UPTR Type A (385) or approved equivalent; Record bulb number.........................................

4. Replace the Regulating Blade clutch current bulb with UFTR Type C (382) or approved equivalent; Record bulb number.........................................

5. Check and record clutch (DC) voltage values for all blades (points 11 and 12 in terminal box under control blade drive units /in pedestals):

a. S-1 c. S-3

b. S-2 d. RB Items 6 - 9 require only the initials of the individual performing and attest-ing to the checks and that the results of the checks are acceptable.

6. Check blade drop times for all blades (see S-1 Instruction)

7. Check controlled removal times for all blades (See UFTR SOP-A.1, Part I)......................

8. Check controlled insertion times for all bladee (See S-5 Instructions)......................

I 9. Indicato satisfactory completion of A-4 surveillance in the Daily Operations Log and on the Status Board. . . . . . . .

Performed By Date RM/FD Acknowledgement Date REV 1, 2/86

V I A-5 EMERGENCY CALL LIST CIIECK FORM An annual check (conducted as preparation for the Annual Drill) is required to I ensure that the UPTR Emergency Call Lists are updated and posted to reflect the current personnel situation. This check should be comple+ed within 60 days prior to the annual drill involving interactions with outside agencies. This form properly completed', signed, dated and acknowledged is the official record that this annual check has been conducted.

Call List 1 (See Attachment 1)

UPTR SOP Manual (Console Copy)

Reactor Staff Room Emergency Response Center (Room 108 NSC)

I Emergency Response Auxiliary Supply Room (Room 106 NSC)

Radio-Chem Lab (near the phone outside the NAA Lab)

Call List 2 (See Attachment 2)

UFTR SOP Manual (Console Copy)

Reactor Staff Room UFTR Control Room

• Emergency Response Center (Room 108 NSC)

Emergency Response Auxiliary Suppfy Room (Room 106 NSC)

Call List 3 (fke Attachment 3)

Upstairs door connecting the Rt. actor building to the NSC Downstairs door connecting the Reactor building to the NSC Security controlled door downstairs accessing the buffer area adjacent to the Reactor Cell Security controlled door upstairs accessing the buffer area adjacent to the Reactor Cell Checked By Acknowledged By Reactor Manager / Facility Director Date REV 2,11/85

B-2 F

L BIENNIAL INSPECTION OF INCORE j REACTOR FUEL ELEMENTS '

I L

DATE: TIME:

~

L In performing the incore inspection of reactor fuel elements, the -

l provisions .of UFTR SOP-C.1, " Irradiated Fuel Handling" must be followed. If more than two bundles are inspected, additional num- i bered copies.of this form should be used.

{.

1. Fuel Bundle Inspected:

Number...................................

Designation..............................

Location.................................

Conene nts:

E F

1 I

l Inspectors:

(SRO in Charge):

(Others)*

j 2. Fuel Bundle Inspected:

l Numder...................................

Designation..............................

Location.................................

Comments:

1 Inspectors:

(SRO in Charge):

(Others):

RM/FD Review and Approval Date REV 1, 1/86

L L

SOP-E.8 PAGE 1 OF 8 I -

t UFTR OPERATING PROCEDURE E.8 w

1.0 Verification of UPTR Negative Void Coefficient of Reactivity L

2.0 Approval ,

Reactor Safety Review Subcommittee . . . . . . . f

(' Ms R 5 8P pay'e Director, Nuclear Facilities .

L

~/

/b[6 d

' Dste l

l l

I I

)

l l

l REV 0, 12/85

SOP-E.8 PAGE 2 of 8 3.0 Purpose and discussion

~'

3.1 This procedure is intended to provide a standard method of de-termining that the void coefficient of reactivity for the UFTR is negative.

I 3.2 The void coefficient of reactivity is required to be determined to be negative biennially at intervals not to exceed 30 months I as per UFTR Technical Specifications, Section 4 . 2.1 , Paragraph (3).

4.0 Limits and precautions 4.1 Hydraulic pressure transients of about 7 psig will cause the primary rupture disk to break; care should be taken to limit both the rate of pressure increases and the maximum pressure.

The maximum pressure required for successful completion of the check is less than 2 psig.

4.2 Reactor Manager approval is required prior to implementation of the test conditions controlled by this procedure 4.3 A senior reactor operator must be present when this procedure is accomplished.

4.4 For the purposes of accomplishing this procedure,. personnel are not allowed in the primary equipment pit when the reactor is not secured.

4.5 The controls and requirements of UFTR SOP-D.3, " Primary Equip-ment Pit Entry" shall be applicable throughout the implementa-tion of this procedure.

4.6 The compressed gas supply shall be physically secured to prevent compressed gas container handling accidents. The preferred gas is nitrogen, but air may be used.

4.7 Technical Specifications require that primary coolant resistivi-ty be greater than 0.4 Megohm-cm after 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of operation.

4.7.1 Since the in-line monitoring system is in the demineralizer loop, the primary coolant demineralizer system shall not re-main secured for greater than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

l 4.7.2 This procedure shall not be implemented unless the inlet and I outlet resistivity readings on the domineralizer system are both above 0.8 Megohm-cm.

i 4.8 Care should be taken to ensure that no contamination results from making and removing connections.

REV 0, 12/85 I

w SOP-E.8 PAGE 3 of 8 4.9 To assure observed transient reactivity effects occur cnly from the introduction of voids (gas bubbles) into the core, the r e a c --

tor shall have a power history of:

4.9.1 Less than 50 kw-hr for the previous 2 days, and 4.9.2 No operation above the point of adding heat (1 kw) for the previous 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 4.10 The summary checklist data sheet contained in Appendix A should be used to track and (with appropriate operating log entr,ies) docu-

ment the implementation of this procedure.

5.0 References 5.1 UFTR Technical Specifications 5.2 UFTR SOP-A.1 " Pre-operational Checks" 5.3 UPTR SOP-A.2 " Reactor Startup" 5.4 UFTR SOP-A.4 " Reactor Shutdown" 5.5 UFTR SOP-0.3 " Primary Equipment Pit Entry" 6.0 Records required 6.1 UPTR Operations Log entry (including Peactor Manager approval) 6.2 UFTR Form SOP-E.8 A , " Summary Checklist Data Sheet: Negative Void Coefficient Verification" 7.0 Instructions 7.1 Preliminary Operations To Establish Test Conditions 7.1.1 Check.and assure the inlet and outlet resistivity on the de-mineralizer loop both exceed 0.8 Mogohm-cm 7.1. 2 Secure the primary coolant pump 7.1.3 Secure the primary purification loop 7.1.3.1 De-energize the resistivity bridge at the control panel; 7.1. 3. 2 Secure the primary demineralizer pump; I 7.1. 3. 3 Close the inlet and outlet primary purification loop iso-lation valves; REV 0, 12/85 I

E SOP-E.8 PAGE 4 of 8 7 .1. '3 . 4 Record the time on'UPTR' Form SOP-E.8A, " Summary Checklist Data S h'e e t : Negative Void Coefficient Verification" 7.1.4 Install the gas (preferred gas is nitrogen) injection system:

7.1.4.1 Position and physically-secure a compressed gas bottle

[_ near the equipment pit; 7.1.4.2 Connect a hose to the' primary sample line fitting; i

{. 7.1 ~. 4 . 3 Connect the free end'of the hose to the regulated com-I pressed gas supply stationed outside the primary equipment pit.

{

7.1.5 Establish gas' injection pressure

[ 7.1. 5.1 Opon the primary purification system inlet isolation valve, check shut the primary coolant purification outlet isolation valve; 7.1.5.2 Establish primary coolant flow;

{ 7.1.5.3 open the primary sample line valve; 7.1.5.4 Slowly increase regulated pressure setpoint to supply gas to the primary coolant loop; 7'.1.5.5 Check sJas flow through the purification flow rotameter; h 7.'.5.6 Record downstream ges pressure required to establish gas flow on UFTR Form SOP-E.8A.

{ 7.1.5.7 Secure,the gas-pressure regulator.

7.2 Verification of Negative Void Coefficient 7.2.1 Perform a reactor startup per SOP-A.2 7.2.2 Bring the reactor stable and critical at 1 watt; then perform

{, the following steps:

7.2.2.1 Establish regulated gas flow through the primary coolant cystem using the gas pressure parameters as noted in Step

{= 7.1.5.6; 7.2.2.2 Observo log / linear recorder until a decreasing power level trend is established; 7.2.2.3 Secure regulated gas flow; REV 0, 12/85 E

c _ ----

SOP-E.8 PAGE 5 of 8 7.2.2.4 Observo log / linear recorder until a stable power level is

established; 7.2.2.5 Record Reactor Operator and Senior Reactor Operator ini-tiala on the chart trace; '

7.2.2.6 Record date and time on chart trace. .

21 r,

f 7.2.3 Repeat the steps in Section 7.2.2 a minimum of two times to provide three separate independent determinations that.the

_ UFTx void coefficient is negative.

7.2.4 Shut down and secure the reactor per SOP-A.4

7.3 Securing Test conditions - Perform The Following Steps .

t d

7.3.1 Close the primary coolant sample valve; 7.3.2 Secure primary. coolant flow; 7.3.3 Remove the regulated gas-pressure supply system; 7.3.4 Ectablish primary coolant flow; 7.3.5 Return the domineralizer system to normal operation

"",",T""

w- ,-

p z .;

7.3.5.1 Open the primary coolant sample valve; '?;

T5 7.3.5.2 1

~

Open the primary coolant purification inlet isolation 21 .- ; -

valve; 44 fk .

,D'-q p :

7.3.5.3 Vent the purification loop; close the sample valve; j$~1)*

.; p 7.3.5.4 Open the purification loop outlet isolation valve; m r ~

y  %; f,.'. -

d 7.3.5.5 Secure primary coolant flow; I;[-[l, '

,.,i, a 7 7.3.5.6 Start the domineralizer pump; *

.,J m

? bia .

7.3.5.7 Check demineralizer flow rate to be normal (about 1 gpm); },' v

'!Eh '.M 3 7.3.5.8 Energize the resistivity bridge; ,y ..

=,% s,

_ 7.3.5.9 Check and record primary coolant inlet and outlet resis-tivity from the resistivity bridge on UFTR Form SOP-E.8 A ;

4 if 3.?- ' h 7.3.5.10 Record the time and assure time elapsed since commence- g.{L.f. -

.,ff,;pl 's?

ment of Step 7.1. 3. 4 has not exceeded 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. ~

REV 0, 12/85 1

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' ' . .. SOP-E.8 PAGE 6 of 8 .h U 4L , 7.4 Quality Assurance Documentation of Surveillance Results ,[% .l y.

~, ' .

7.4.1 ifyR j' Assure that all operations log entries are complete for this surveillance test for negative void coefficient verification. yg f.'. -

i,e..:4. W.1-A 7.4.2 Obtain copies of the log / linear recorder chart traces illus- h ,, '. ' M A '

trating changes in power level corresponding to the estab- jgg - .,

lishment and then the securing of gas (void) injection into Ji~'r

') I the primary coolant system. D.O '5

• .. g d ?

7.4.3 Copies of the log / linear recorder chart traces shall be an- .NT/.*

.q .

W notated, dated and initialled by the reactor operator of Y.~d record and by the senior reactor operator directing the op-eration; r g*^,.g O

f A

I 7.4.4 Copies of the chart traces shall be attached to UPTR Form SOP-E.8 A (Summary Checklist Data Sheet: Negative Void Coeffi-y h< -

p.i-Q , -

Ll.? l ? ."

V.

g eient Verification) to provide a record of the surveillance M, d g' test.

.h 2

if ;l 7.4.5 The reactor operator and the senior reactor operator should kt i ',

I- include appropriate comments on UFTR Form SOP-E.8 A (Summary Checklist Data Sheet: Negative Void Coefficient Verifica-

[],Nj6

.i. ,1 Y tion). .

..' ' 5 : .,

7.5 Transmit the surveillance records to the Reactor Manager or- j ', g ,.i Facility Director for final review and approval to include

• J.

W I 7.5.1 Annotated log / linear recorder chart traces (three sets);

g. . .. .,. h. , .-

g, . :4, hjs.,

gc ? .

. 7.5.2 Completed UFTR Form SOP-E.8A. .l .y . f . "

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SOP-E 8 PAGE 7 of 8 t.

I APPENDIX A UPTR FORM SOP-E.8A t

SUMMARY

CHECXLIST DATA SHEET:

I NEGATIVE VOID COEFFICIENT VERIFICATION I

I I

I REV 0, 12/85

w SOP-E.8 PAGE 8 of 8 I UPTR F rm SOP-E.8A

SUMMARY

CHECKLIST DATA S!!EET NEGATIVE VOID COEFFICIENT VERIFICATION Date: Date of Previous check:

I. Establishing Test Conditions III. 'ecuring S Test Conditions A. Establish Proper Flow Conditions A. Remove Gas Injection System....

1. Check inlet / outlet resis- 1. Close PC sample valve.......

-I tivity on demin system......

2. Secure PC pump..............

2. Secure PC flow..............

3. Remove gas injection system.

3. Secure resistivity bridge... 4. Establish PC flow...........

4. Secure primary domin pump...

5. Shut inlet & outlet primary B. Return Demineralizer System purif loop isolation valves.

6. Record time.................

• To Normal Operation............__

1. Open PC sample valve........

B. Install Gas Injection System 2. Open purification inlet I 1. Position / physically secure gas supply..................

valve.......................

3. Vent purification loop via sample line; close sample

2. Connect hose to primary I sample line fitting.........

3. Connect hose to regulated gas supply..................

valve.......................

4. Open purification outlet valve.......................

5. Secure PC pump..............

6. Start demin pump, check C. Establish Gas Injection Pressure flow to normal (~1 gpm) . . . . .

7. Energize the resistivity

1. Open primary purif system I inlet isolation valve.......

2. Check shut PC purif outlet bridge......................

8. Read and record PC inlet resistivity (Mehohm-cm).....

• isolation valve............. 9. Read and record PC outlet I 3. Establish PC flow...........

4. Open primary sample line resistivity (Mehohm-cm).....

10. Record time / elapsed time valve....................... since I-A(6)................

• I 5. Inject gas, check flow thru purif flow rotameter........

6. Record required gas pressure

'IV. QA Documentation

7. Secure gas-pressure I regulator...................

A.

B.

Assure Operating Log Entries Are Complete...................

Obtain Copies Of Chart Traces..

II. Verification of Negative Void Coefficient C.

I A. Perform Reactor Startup. . . . . . . .

D.

Annotate Copies Of Traces......

Attach Annotated copies of Recorder Traces For Three Void Coefficient Checks.............

B. Stabilize @ 1 Watt and E. Comments:

1 Establisa regulated gas flow

2. Note log / linear decreasing I power trend.................

3. Secure gas flow.............

4. Note stable power level.....

I 5. Record RO & SRO initials on chart....................

6. Record date & time on chart.

RO Signature Date I C. Repeat Part II-B, Steps 1-6, Two Times......................

SRO Signature Date RM/FD Review & Approval Date D. Shut Down, Secure Reactor......

REV 0,12/85

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APPENDIX D UFTR SAFETY ANALYSIS REPORT REVISION 2 DOCUMENTATION

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COLLEGE I 2 J' '

s UNIVERSITY OF FLORIDA DEPARTMENT OF NUCLEAR ENGINEERING SCIENCES _

202 NUCLEAR SCIENCES CENTER

'I' ENGINEERING cAiNESviLLE, rLoniO4 32si -

AREA CODE 904 PHONE 392-1401.

@ a.

_t DEPARTMENT OF NUCLEAR ENGINEERING SCIENCES July 18, 1986 -I R

I .

Office of Nuclear Reactor Regulations Division of PWR Licensing-B I U.S. Nuclear Regulatory Commission Washington, D.C. 20555 6

g Attention: Herbert N. Berkow, Director ig.

Standardization and Special Projects Directorate Re: University of Florida Training Rea~ ctor Facility License. R-56 Docket No. 50-83

Dear Sir:

The enclosed package contains Revision 2 pages for the UFTR Safety Analy-sis Report dated January,1981 submitted as part of our relicensing effort.

This Revision 2 is submitted as a result of an NRC inspection of our I facility conducted on February 18-21, 1986, and as cited in Inspection Report No. 50-83/86-01 as Inspector Followup Item IFI 083/86-01-02. As requested in the Inspection Report and committed in our reply, our facility has committed to update Paragraph 7.2.3 of the UFTR Safety Analysis Report "which describes -

operation of the control rod inhibit system and automatic control system, which is different from the performance described in Technical Specification 3.2.1.. Surveillance procedures confirm performance in conformance to the I requirements of Technical Specifications." As a result of the review of Paragraph 7.2.3, this revision package also includes revisions to Sections 7.3 and 7.6 as well as Figure 1-8 from Chapter 1.

I All text changes are clearly indicated by vertical lines in the margins, The two figure changes can be noted by comparison with original figures. Many I of the changes represent simple typographical errors or omissions, though several involve facility description discrepancies discovered as a result o/

the NRC Inspection in February,1986. All changes have been reviewed by UFTR staff and by the Reactor Safety Review Subcommittee as required by Technical Specifications. A more detailed description of the revision is included as Attachment I. A summary table of the changed pages, along with the Revision 2 pages, is enclosed as Attachnent II.

l l

FLORIDA *5 CENTER FOR ENGINEERING EDUCATION AND RESEARCN _

_ znun zue.Lorum.ur neanarumar v rasramua riw r a e ria . - - ---- N

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Letter to Office of Nuclear Reactor Regulations i Page Two July 18, 1986 I The enclosure consists'of:

I 1. Three (3) signed originals and nineteen copies of this letter of transmittal.

2. Twenty-five (25) copies of the attachments containing a description of Revision 2 to the UFTR Safety Analysis Report as well as the Revision 2 pages.

If further information is required, please let us know. ~

Sincerely,

.. .. . . . ) , A 5 i "

William G. Vernetson

! 8,. ' y ) r i . . ... Associate Engineer and l

Director of Nuclear Facilities

$$ Xt.A n u -

i No'ta ry.,,P ubl ,ip. '

i WGV:lmc Enclosures i

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'I I

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ATTACHMENT I g

DETAILED DESCRIPTION =-

UFTR SAR REVISION 2 1

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As noted in NRC Inspection Report No. 50-83/86-01, Paragraph 7.2.3 (Non- j Nuclear Instrumentation Channels) describes operation of the control rod in- 4 hibit system and automatic control system, which is somewhat different from -

the performance described in Technical Specifications Paragraph 3.2.1. It was ,

i also noted that surveillance procedures confirm performance in conformance a with the requirements of the' Technical Specifications. As a result of IFI j 083/86-01-02, Paragraph 7.2.3 and Sections 7.3, 7.4, 7.5 and 7.6 as well as I Figure 7-3 referenced in Section 7.2 and Figure 1-8 (duplicate of Figure 7-3) have been reviewed and revised as necessary. 3 First, Figure 7-3 in Section 7.2 and Figure 1-8 were noted to indicate a BF3 proportional chamber detector for the source range detector in Nuclear f

Instrumentation Channel 1. This detector is described as a B-10 proportional j detector in the text of Section 7.2, Paragraph 7.2.2.1.1; in addition, the I B-10 detector, not a BF detector was in place when the SAR was submitted. To correct the improper in$1 cation of the source range detector from a BF detec-3 v

9

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tor to a B-10 detector, a corrected Figure 1-8 is attached. The same change is made on Figure 7-3. __

Next, Paragraph 7.2.3 was reviewed and revised to include the following clarifications: j I 1. Much of Section 7.2.3 was reworded to delineate more clearly the non- $

nuclear instrumentation system as existing and as described in the SAR, 7 no substantive changes are included in this rewording. -

2. Paragraph 7.2.3.2 describing the Control-Blade Withdrawal Inhibit System [

I was rewritten to correspond exactly to the description on the UFTR Technical Specifications with some clarifications for ease of understand-ing. No substantive changes are included in this rewritten paragraph.

J-

3. Paragraph 7.2.3.3 containing the description of the Automatic Control System was rewritten for clarification, again with no substantive changes. p I 4. A new Paragraph 7.2.3.4, " Process Monitoring and Control Systems" which expands upon the paragraphs following the description of the Automatic i

Control System was added. Again, the contents are not substantively -

I changed; however, the section is now labelled and individual subsections ~

are delineated in more detail as follows:

7.2.3.4.1, " Primary Coolant System" I 7.2.3.4.2, " Secondary Coolant System" 7.2.3.4.3, " Shield Tank System"

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I Again this delineation of Paragraphs makes locating specific items easi-er; in addition, the inforniation contained in these sections is better written and organized.

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4 Section 7.3 Reactor Trip System has also been reviewed and one change made to correct an SAR error in classifying trips. Section 7.3.1 " Nuclear In- $

7 i

duced Trips (Full Trips)" currently delineates four (4) trip conditions while  ;

3 Section 7.3.2 " Process Instrumentation Induced Trips (Rod-Drop Trips)" current-3 ly delineates thirteen (13) trip conditions. The last trip listed under Sec-3 g tion 7.3.2 is "AC power failure (fail-safe criterion)." This trip condition y

, has always caused a full trip (this is one of two methods used to dump the --s T i primary coolant to the storage tank) with two ' control blades partially removed -5 so the rewritten Section 7.3 lists this trip under Section 7.3.1 so there are g five trip conditions listed under Section 7.3.2 and only twelve (12) trip con- 7 ditions listed under 7.3.2. Section 7.3.2 is also relabelled as Process In-

• strumentation Induced Trips (Blade Drop Trips) tio reflect UFTR use of control

blades, not control rods. Again, the change simply reflects an accurate de- hg E scription of a preexisting and licensed system that was incorrectly described g in the Section 7.3 of the SAR submitted in 1981. 2 3,

p The review of Sections 7.4 and 7.5 does not show the need for any V changes. The same is true for Sections 7.7 and 7.8. However, the review of r Section 7.6 on Safety-Related Display Instrumentation showed one incorrect f!

pj reference at the end of the first paragraph. The reference to "Section -

5 I 7.1.1.1" is changed to the proper reference which is "Section 7.1.2." In addition the remainder of Section 7.6 is rewritten to provide a better, more accurate description of the Safety-Related Display Instrumentation. The only

}a 1

r. g change in the description is in the final paragraph of 7.6 where the location 2 E of the portal monitoring system is changed from "in the airlock" to "outside 4 L

the airlock" to reflect the current and pre-existing location of the portal J g monitor. This location is the preferred location to facilitate monitoring of g g personnel exiting not only from the airlock but also directly from the control ;ji.

room if necessary. Again, the revision of this Section is not considered to

- j>

involve any substantive changes from the system as relicensed but only to re-  ?=

flect a more accurate and clearer description of the Safety-Related Display Instrumentation Systems.

g In summary, this proposed revision simply corrects a number of typographi- $

' g cal or other inadvertent non-substantive errors and also provides a better, M clearer and more accurate description of systems that are unchanged since sub-

" a mittal of the UFTR Safety Analysis Report in 1981. The UFTR SAR Revision 2 -

g along with a list of changed pages is included as Attachment II. i s

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ATTACHMENT II I

FINAL SAFETY ANALYSIS REPORT UNIVERSITY OF FLORIDA TRAINING REACTOR FACILITY LICENSE R-56; D0CKET NUMBER 50-83 H Revision 2 r

L The attached Revision 2 pages revise the University of Florida Training Reactor Final Safety Analysis Report as of July,1986. Revision 2 pages y should be substituted to replace original pages as indicated in the following L table.

Original Pages Revision 2 Replacement Pages 1-21 1-21 7-7 7-7 7-9 7-9

[ 7-11 7-11 7-13 7-13 7-15 7-15

[ 7-16 7-16 and 7-17 Due to added material there is one more page of text in' Chapter 7 than

[ previously. Replacement of pages as noted here will put all pages in the proper position.

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NI CHANNEL 1 (LOG N, SAFETY I and PERIOD) LOG N

" LOG N" RECORDER PRE LOG AMP AMP B-10 -

TEST "P E RIO D_"

a CAL W" PERIOD COMPUT

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M UFTR Nuclear Instrumentation Channel 1 Schematic (Log N, Safety #1 g Figure 7-3.

and Period Channels). .

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2-ei servo amplifier, the signal is compared with the signal from the servo flux con-trol.

-= 7.2.2.2.2 Safety Channel #2. As indicated in Figure 7-4, the safety channel receives a signal from an uncompensated ion chamber and consists of the ion chamber (with an independent high voltage supply), an operational amplifier, an adjustable a

=

bi-stable trip, and a meter ranging from 1% to 150% rated power. The Safety Chan-nel #2 system initiates a reactor trip at 125% power. Safety Channel #2 also ini-tiates a reactor trip whenever the high voltage applied to the chamber drops by 10%. The channel also generates test signals to check the functioning of the channel.

, 7.2.3 Non-Nuclear Ins'trumentation Channels The UFTR is supplied with several process instrumentation channels to monitor the normal operation of the various systems, to aid in maintaining a steady-state 3 power level and to trip the system whenever a potentially unsafe situation occurs or an instrument fails. ~ 0ther channels supply information needed to safely operate the reactor but do not have protective functions. These Non-Nuclear Instrumentation Channels are described in the next four subsections.

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7.2.3.1 Control-Blade Drive System. The control-blade drive circuit is shown in

_ Figure 7-5; it consists of switches and indicating devices used in operating the four ,h ll!

" control blade drives. The twelve backlit push button switches are arranged in the l center of the control panel in three rows of four vertical sets, one set for each control blade. Each set of switches contains a white DOWN switch, a red UP switch,

] and a yellow ON (magnet on) switch. yo When the white DOWN light is illuminated, the control blade drive motor power y%}

j circuit is prevented from drive action via the down backlit pushbutton switch. When i.n: ; .(

Al the red UP light is illuminated, control blades in manual control are similarly

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prevented from up motion. The yellow ON light is series-connected in the magnetic E}i L clutch power circuit so that if the yellow light is on, the magnetic clutch is  %

d5 energized; if the yellow ON light is off, the magnetic clutch is deenergized, fgj;d f rg:

When any ON push button switch is depressed, magnet current is interrupted by E :.2d j actuation of the backlit switch, and the ON light remains extinguished for as long as .@

1. the switch is depressed. If the control blade is above its down limit, the blade a

will gravity fall back into the core. Turning off the reactor key has the same effect. In the event of a loss of power, these blades fail safe, falling into the V@Q g

g core by gravity. gj 7.2.3.2 Control-Blade Withdrawal Inhibit System. The Control Blade Withdrawal n.W 2 '.37 q Inhibit System is depicted in Figure 7-5; this Inhibit System is part of the reactor %y

_5 protection system and functions to prevent blade withdrawal for the following .[:M

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. 1. Insufficient neutron source counts to assure the proper functioning of W the source level instrumentation. A mininum source count rate of 2 L W[

cps (as measured by the wide range drawer operating on extended range) is required by the technical specifications. iM ffp

2. A reactor period of 10 seconds or faster.

7-9 REV 2, 7/86 [.M w

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3. Safety Channels 1 and 2 and wide range drawer Calibrate (or Safety 1 Trip Test) switches not in "0PERATE" or "0FF" condition. This inhibit condition assures the monitoring of neutron level increases and pre-vents disabling protective functions as control blades are raised.

4. Attempt to raise any two or more blades simultaneously when the reac ~

I tor is in manual mode, or two or more safety blades simultaneously when the reactor is in aiitomatic mode. This multiple blade withdrawal interlock is provided to prevent exceeding the maximum reactivity j

addition rate authorized by the UFTR Technical Specifications.

5. Power is raised in the automatic control mode at a period faster than 30 sec. The automatic controller action is to inhibit further regula-I ting blade withdrawal or drive the regulating blade down until the period is greater than or equal to 30 seconds.

I 7.2.3.3 Automatic Control System. The UFTR Automatic Control System is' used to hold reactor power at a steady power level during extended reactor operation at power and may be used to make minor power changes within the maximum range of the switch setting. While the automatic mode of reactor control is selec-I ted, the manual mode of operation is disabled; the control mode switch must be placed back in MANUAL before the regulating blade will respond to its UP or DOWN control switches. The neutron flux controller shown in Figure 7-6 com-I pares the linear power signal from the pico-ammeter with the power demand sig-nal and moves the regulating blade in order to reduce any difference, thereby maintaining a steady power level.

7.2.3.4 Process Monitoring and Control Systems:

7.2.3.4.1 Primary Coolant System: A primary coolant flow monitor, with a I ser.sor located in the primary fill line, indicates flow at the control console and prevents reactor operation, or trips the reactor, if flow is below the set point of 30 gpm (normal flow is about 40 gpm).

I A coolant flow switch, located in the return line of the primary coolant system to the primary coolant storage tank, initiates a reactor trip in case of I a loss of return flow. This flow switch serves as a backup for the low flow reactor trip in the fill line and activates only after the return line has been drained of water or the flow is reduced to less than about 10 gpm.

I A sight glass, attached to the north wall of the reactor room, at the east i side of the primary equipment pit, shows the water level in the core allowing a  ?-

visual check of the primary coolant level. A float switch, located behind the i-sight glass, is wired to the reactor protection system. It prevents reactor operation, or activates the reactor trip system, when the water level in the (.

g core is below pre-set limits. 4, Type "T" (copper-constantan) themocouples are located at each of the fuel box discharge lines to monitor water temperature from each fuel box to the pri-I mary coolant storage tank, and 2 thermocouples monitor the temperature of the bulk primary water going to and exiting from the core. The temperature infor-mation is sent to the 12 point recorder in the reactor control room. If any monitored temperature point exceeds preset levels, an audible alarm occurs at 150*F, and the reactor trips at 155 F.

_V .

7-11 REV 2, 7/86

A resistivity meter mounted on the east wall of the control room enables monitoring resistivity of the primary coolant and functioning of the primary coolant purification demineralizer system. The meter annunciates if system resistivity drops below an adjustable, preset value.

To monitor water leakage from any source into the primary equipment pit, a -

level switch in a small sump at the lowest point of the pit floor will activate an alarm upon collecting as little as two cups of water. .The primary equipment pit sump alarm annunciates at a control unit mounted on the east wall of the Control room.

7.2.3.4.2 Secondary Coolant System: The principal source of cooling water to remove reactor heat is the deep well, nominally rated at 200 gpm. . A

- reduction of flow to 140 gpm will illuminate a yellow warning light on the right side of the control console. A reduction of flow to 60 gpm will illumi- --

nate a red scram warning light on the right side of the console, and will 1 illuminate a red warning light on the secondary flow scram annunciator light..

Ten seconds later, the trip will occur. When using city water for reactor cooling, a low water flow of 8 gpm will trip the reactor. In either instance, the trip function is active only when reactor power is 1% or higher. A key operated switch on the console rear door switches secondary scram modes between well water or city water modes of operation.

7.2.3.4.3 Shield Tank System: A water level switch in the top of the reactor shield tank will trip the reactor when the water level drops below a preset value. This switch prevents reactor operation because of shield tank

" water loss due to evaporation or leakage. The purification loop on the west side of the shield tank has a flow indicator to monitor proper functioning of the loop.

E 7.3 Reactor Trip System The UFTR facility is provided with two types of reactor trips, both initiating the gravity insertion of all the control blades into the core.

These reactor trips can be classified into two categories:

1. Nuclear Instrumentation Induced Trips, which involve the insertion of '

the control blades into the core and the dumping of the primary water into the storage tank (this type of trip will dump primary water only if 2 or more control blades are not at bottoia position);

2. Process Instrumentation Induced Trips, which involve only the inser-tion of the control blades into the reactor core (without dumping of =-

the primary water). Figure 7-7 shows a schematic diagram of the Mf Protection System provided for the UFTR.

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-4 m g 7.3.1 Nuclear Instrumentation Induced Trips (Full Trips) df h One of five conditions must exist for the initiation of the Reactor Trip System with dump of primary water (Nuclear-Type Trip); these five conditions gd 3

include: ggg E 4 I@5% L%

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.s.- # Xhy[tu il 1. Fast Period (3 seconds or less), 4p;a.f

2. High Power, safety channel #1 (125%) or safety channel #2 (125%), $ - W. 2

3. Reduction of high voltage to the neutron chambers of 10% or more,  :$bi 1' 4. Turning off the console magnet power switch. 9?i ".

Y 5. A.C. power failure (fail-safe critcrion).

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[g 3 7.3.2 Process Instrumentation' Induced Trips (Blade-Drop Trips) yd'"

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The cor.ditions which must exist for the initiation of the Reactor Trip 3 R[ $ .

j, System without dump of primary water (process type trips) include

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. ts 5 1. Loss of power to the Reactor Vent Blower System. @w . .

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2. Loss of power to Reactor Vent Diluting System.

. nf. L 3. Loss of power to the secondary system deep well pump when operating at  ? ;-@.

i.?; or above 1 Kw and using this system for secondary cooling,

9. 4. Dropping of secondary flow below 60 gpm (normal flow 200 gpm, alarm at ba Eg Y[ga,

.', g' 1; 140 gpm) when operating at or above 1 Kw when using the well water system for secondary cooling. IM., .

. 5. Dropping of secondary flow below 8 gpm when at or above.1 Kw when Mf .3

%, a using city water for secondary cooling.

Ci B 6. Drop in water level of the shield tank (about 4 in.) 4 A:f. W@

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8.

Loss of power to primary coolant pump.

Reduction of primary coolant flow (normal 40 gpm, trip at 30 gpm);

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1.f flow sensor is located in the fill line.  :.4.M t iw

9. Loss of primary coolant flow (return line). 2 .AT .[,.

-t 10. Reduction of primary coolant level. ED 2

!f . 11. High temperature primary coolant return from the reactor (alarms at 40 h ~

.l 150*F, trips at 155*F). 0..' bf.: ;

12. Manual reactor trip button depressed.

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6.f A set of annunciator lights located on the left side of the control con- g. 'y&

_h sole indicates all scrams and 3 interlock conditions. In case of high reactor *, . af f temperature, an audible alarm is set off at 150 F and the reactor trips at gfg -

155*F. The alarm continues to sound until the indicated temperature drops 6 T^[

i below 150 F. g $i-l

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A red rotating beacon located in the reactor cell together with three

" reactor on" lighted signs located on the outside of the east side of the 30

. E f Reactor building on the second floor level, on the entrance hallway leading b. ;,@

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.g a to the control room, and on the north outside reactor building wall, are all energized whenever the console key switch is turned to the "0N" position.

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1 7.4 Engineering Safety Feature System g,;iQ.

(.i Es As explained in Chapter 6, there are no separate Engineered Safety (+g." .%

Tj Features required in the UFTR aside from those built-in into the facility. ..O d;-

yj Therefore, no instrumentation or control system relative to this system is M.L M!

ed.N.;"

present.

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, .a 7.5 Systems Required for Safe Shutdown hk py 4 a.

47 The only system required for normal safe shutdown is the safety-control  % Q.

l'Q, blades drive instrumentation channels allowing the operator to insert the Q .,

<y blades into the core to shut the UFTR system down. Proper blade movement can LJ, (n

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E be observed at the display panel where the four blade position indicators are located. In addition, the nuclear instrument channel read-outs provide another way for determining proper decrease in power for reactor shutdown. Neverthe-

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less, the only system really necessary for reactor shutdown is the control blade drive system. In case of failure of this system on a loss of power, the control blade system is designed to fail safe; the blades _ drop by gravity into -

, the system to shut the reactor down. A semi-annual measurement is made of

, blade drop times which must be less than 1 second. Normal times are about 0.5 second. If the control blades do not function properly and the core overheats, the negative void and temperature coefficients will cause the core to go sub-2 critical and shut down even without insertion of the control blades. There-fore, instrumentation is not an absolute necessity for shutting the UFTR down because of its inherent safety features. In addition, the reactor can be made _

-- subtritical and power reduced by the operator-initiated action of dumping the primary coolant.

7.6 Safety-Related Display Instrumentation Readouts from all of the nuclear instrumentation and non-nuclear instru-

_ mentation channels displayed at the reactor console are described in Section

7.2.1. l The reactor vent system effluent monitor consists of a GM detector and g preamplifier, which transmits a signal to the control room to monitor the gama activity of the effluent in the downstream side of the absolute filter, before dilution occurs. The stack monitoring system also consists of a log rate meter-circuit and indicator, a strip chart recorder, and an auxiliary log rate meter with an adjustable alarm setting capability for monitoring the gross ac-tivity concentration of radioactive gases in the room effluent air entering the stack. If the activity reaches the preset (fixed) alarm level or if activity

reaches the auxiliary alarm setpoint (operator adjusted to the highest power level permitted or expected during the operation) the monitor will actuate an audible alarm in the control room.

A complete area radiation monitoring system consisting of three indepen-dent area monitors with remote detector assemblies and interconnecting cables, and strip chart recorders and count rate meters is available. The signals from these detectors are sent directly to the log count rate meter and recorder, monitoring the gamma activity in the reactor room. Each detector has an energy compensated Geiger Counter with built-in Kr-85 check source which can be oper-ated from the control room. Two levels of alarm are provided. The high level latches in the alarm mode to preclude false indication if a high level of radi- e ation saturates the detector. Any two of the monitors seeing a high level of radiation will automatically actuate the building evacuation alarm. Actuation g

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of the evacuation alarm automatically trips the reactor cell air conditioning system and both the diluting fan and the vent fan.  %@.

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&- The sti.ck monitor and 3 area monitor modules in the control room are kW equipped with test switches and green 'N0 FAIL" lights that go out if the modules do not receive signal pulses from the detectors. Floating battery -

g packs supply power to the units in the event of electrical power loss.

4

, 7-16 REV 2, 7/86

The reactor cell air monitoring system is equipped with a flow indicator l (LPM), a strip chart recorder and an audible and visible alarm setting. The monitor is a lead-shield, compact airborne particulate Geiger Counter detector. l I The portal monitoring system outside the airlock leading from the reactor cell is a Beta-Gamma Portal Monitor (Model PCM-4A) console and portal frame. ~

J I It contains (eight) channels of Geiger tube detectors providing complete head to foot coverage of beta-gamma radiation plus individual alarm lights for each channel. An audible alarm will be activated any time the preset (count rate limit) is exceeded.

7.7 All Other Instrumentation Systems Required for Safety

, I There are no other instrumentation systems required for the safe operation of the UFTR; all the necessary instrumentation has been covered in previous sections of this chapter.

7.8 Control Systems Not Required for Safety i

lan There are no control systems in the UFTR facility which do not have safety j related functions as considered in this Safety Analysis Report. Consequently, all UFTR control systems have already been described in the preceding sections.

Even those controls which do not have a safety operational function do have a safety function in the sense of providing information on safe UFTR operation through read-outs supplied by the appropriate monitoring control.

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I APPENDIX E {

UPTR REACTOR OPERATOR RECERTIFICATION PLAN I JANUARY, 1986 I .

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I I 4

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4 ATTACHMENT II WORK PLAN FOR RESTORATION OF UFTR TO OPERABLE CONDITION I. Complete Repair and Maintenance Work on Control Blade Drive System (F-1) .

II. Fuel Reload:

t A. Loading Plan

1. Develop The Plan 2.- Receive Approval by RM, FD, RSRS B. Staff Recertification Program h 1. Conduct Training Required to Reload Fuel

2. Recertify Operator's by Facility Director

[ 3. Receive Approval of Recertification from NRC Region II Operator Licensing C. Instrumentation Setup D. Fuel Reload Performance

1. Inspect All Irradiated Fuel Elements (B-2)

{ 2. Reload All Approved Elements With All Operators Assisting

3. Document Reload

[~ III. Test and Checkout Program f

A. Staff Recertification Program

1. Complete Practical Training

{ a. Conduct Weekly Checkout (Group)

b. Conduct Individual (Observed) Daily Checkouts B. Precritical Tests

1. Perform Quarterly Calibration Check of Area and Stack Monitors

[: (Q-2)

2. Perform Annual Replacement of Control Blade Clutch Current

( Light Bulbs ( A-4)

3. Measure Control Blade Removal Times (Weekly checkout)

[

[ __ --

=

E

_ 4. Measure Control Blade Controlled Insertion Times (S-5)

5. Measure control Blade Drop Tines (S-1) je; , .

_; " > :t,1.i g

6. Perform Quarterly Check of Scram Functions (Q-1) {-['bg.:

l ..]

C. Critical Checks f,/f if vi . - *

1. Establish 1 watt critical position consisting of five startups to 1 watt followed by snutdowns (used also as part of operator recertifica tion training).

2. Certify each operator to resume licensed duties (Facility

, Direc tor) .

i

~

3. Perform Annual Reactivity Measurements (S-2):

, a. Worth of Each Control Blade

=

b. Reactivity Insertion Rate of Each Blade

c. Total Excess Reactivity

, g d. Shutdown Reactivity (Shutdown Margin) 15 D. Power Checks =

h 1 Perform Environmental Cell Surveys Durin'g Power Ascension

2. Perform Annual Measurement of UPTR Temperature Coefficient of Reactivity ( A-3)

3. Perform Quarterly Radiological Survey of Restricted Areas (Q-5)

-m

4. Perform Quarterly Radiological Survey of Unrestricted Areas (Q-4)

5. Measure Argon-41 Stack Concentration (S-4) fm

6. Measure Dilution Air Flow Rate (S-4 from A-1) ,

2 7. ?erform UPTR Nuclear Instrumentation Calibration Check and Calorimetric Heat Balance (A-2) q 8. Perform Verifications Following Calorimetric

a. Temperature Coefficient v ,;.7 -

b. Blade Worths $g;ij$

3 5 $$h?,

IV. Review and Approval of All Results j sp{$q yf, :g a n .4 g A. Reactor Manager  : p .

B. Facility Director

.f*. g

.i.;;

E NE E

C. Reactor Safety Review Subcommittee --

D. Transmittal of Final Report to NRC Region II an.

ee ium

• 5 8 "^' *"c'

• NUCLEAR FACluTIES DIVISION

. tavammenom.a: Acron urnacEn NUCLEAR REACTOR BUILDING UNIVERSITY OF FLORDA .

GA4NESVILLE.FL0alDA 32611 aves

, PHONE (904l 392-1429 TELEX $6330 November 29, 1986 Office of Nuclear Reactor Regulations Standardization and Special Projects Branch Director, Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Re: Facility License R-56 Docket No. 50-83

Dear Sir:

In compliance with our Technical Specifications reporting requirements, enclosed is one copy of the 1985-1986 University of Florida Training Reactor Annual Progress Report.

This document complies with the requirements of the UFTR Technical Specifications, Section 6.6.1.

Please advise if further information is needed.

Sincerely, A d i William G. Vernetson Director of Nuclear Facilities WGv/ps Enclosure cc: P.M. Whaley Acting Reactor Manager l

l l q EQUAL OrromfithfTYl AFFlant4TivE ACTION EterLOYER I