ML20247H704
| ML20247H704 | |
| Person / Time | |
|---|---|
| Site: | MIT Nuclear Research Reactor |
| Issue date: | 06/30/1989 |
| From: | Bernard J, Kwok K MASSACHUSETTS INSTITUTE OF TECHNOLOGY, CAMBRIDGE |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 8909200073 | |
| Download: ML20247H704 (34) | |
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- 2 NUCLEAR REACTOR LABOR' TORY.
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AN INTERDEPARTMENTAL CENTER OF HEM /
MASSACHUSETTS INSTITUTE OF TECHNOLOGY O K, HARLING 138 Albany Street, Cambndge, Mass. 02139 J.A. BERNARD, JR.
- Director Telefax No. (617)253-7300 Director of Reactor Operations Telex No. 92-1473 MIT-CAM
-Tel. No (617) 253-4211 August 30, 1989
- t z U.S. Nuclear Regulatory Commission ATTN
- Document Control Desk Washington, D.C. 20555 l-Subj ect: Annual Report,. Docket No. 50-20, License R-37, Technical Specification 7.13.5 Dear Sirst Forwarded herewith is the Annual Report for the. MIT Research Reactor for the period July 1, 1988 to June 30.1989, in compliance with paragraph. 7.13.5 of the Technical Specifications for Facility Operating License R-37.
Sincerely, M<
Kwan S. Kwok Superintendent, Reactor Operations 0
John A. Bernard, Ph.D V
Director of Reactor Operations KSK/gw
Enclosure:
As stated cc: MITRSC USNRC - Region I - Chief, Reactor Projects Section No. 3B USNRC - Region I - Project Engineer, Reactor Projects Section No. 3B USNRC - Senior Resident Inspector, Pilgrim Nuclear Station p
USNRC - Project Manager, I
l Standardization and Non-Power Reactor Project Directorate 8909200073 890630 PDR ADOCK 05000020 R
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MIT RESEARCH REACTOR ANNUAL REPORT TO UNITED STATES NUCLEAR REGULATORY COMMISSION FOR THE PERIOD JULY 1, 1988 - JUNE 30, 1989 BY 4
REACTOR STAFF August 30, 1989
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.. TABLE OF CONTENTS Page Seetion Stenber Table of Contents................................................i Introduction.................................................. 1 A.
Summary of Operating Experience............................
3 B.
Reactor Operation..........................................
9 C.
Shutdowns and Scrams......................................
10 D.
Maj o r Ma i n t e n an c e......................................... 12 E.
Section 50.59 Changes, Tests, and.........................
15 Experiments F.
Environmental Surveys.....................................
26 G.
Radiation Exposures and Surveya Within....................
27 the Facility H.
Radioactive Effluents.....................................
28
e._
A.
MIT RESEARCH REACTOR ANNUAL REPORT TO UNITED STATES NUCLEAR REGULATORY COMMISSION FOR THE PERIOD JULY 1.
1988 - JUNE 30, 1989 Introduction This report has been prepared by the staff of the Massachusetts' Institute of Technology Research Reactor for submission to the Admin-istrator of Region 1, United States Nuclear Regulatory Commission, in compliance with the requirements of the Technical Specifications ' to Facility Operating License No.
R-37 (Docket No.
50-20), Paragraph 7.13.5, which requires an annual report following the 30th of June of each year.
.The MIT Research Reactor (MITR), as originally constructed, con-sisted of a core of MTR-type fuel, fully enriched in uranium-235 and cooled and moderated by heavy water in a four-foot diameter core tank,.
surrounded by a graphite reflector. After initial criticality on July 21, 1958, the first year was devoted to startup experiments, calibra-tion and a gradual rise to one megawatt, the initially licensed maxi--
mum power.
Routine three-shift operation (Monday-Friday) commenced in July 1959.
The authorized. power level was increased to two megawatts in 1962 and five megawatts (the design power level) in 1965.
Studies of an improved design were first undertaken in 1967.
The concept which was finally adopted consisted of a more compact core, cooled by light water, and surrounded laterally and at the bottom by a heavy water reflector.
It is undermoderated for the purpose of maxi-mizing the peak of thermal neutrons in the heavy water at the ends of the beam. port re-entrant thimbles and for enhancement of the neutron flux, particularly the fast component, at in-core irradiation facili-ties.
The core is hexagonal in shape, 15 inches across, and utilizes fuel elements which are rhomboidal in cross section and which contain UAL intermetallic fuel in ths form of plates clad in aluminum and x
fully enriched in uranium-235.
Much of the original facility, e.g.
graphite reflector, biological and thermal shields, r 'ondary cooling systems, containment, etc., has been retained.
Af ter Construction Permit No. CPRR-118 was issued by the former U.S. Atomic Energy Commission in April 1973, major components for the modified reactor were procured and the MITR-I was shut down on May 24, 1974, having logged 250,445 megawatt hours during nearly 16 years of operation.
The old core tank, associated piping, top shielding, control rods and drives, and some experimental facilities were disassembled, re-moved and r2bsequently replaced with new equipment.
After preopera-tional tests were conducted on all
- systems, the U.S.
Nuclear i
4 2
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Regulatory Commission. Issued Amendment No. 10.to Facility Operating License No. - R-37 on July 23, 1975.
After initial ' criticality for MITR-II on August'14th, 1975, and several months of startup testing, power ' was1 raised to 2.5 MW in December.. Routine 5 MW operation-was achieved in December 1976.
This is the f ourt eenth i annual report required by the Technical
- Specifications, and it covers the period. July 1, 1988 through June.'30, 1989..
. Previous reports, along with the "MITR-II Startup Report" -
(Report - No. MITNE-198, February 14, 1977) have : covered the startup testing period and the transition to routine reactor operation.
This report covers the twelfth full year of routine reactor operation at the. 5 MW licensed power level.
.It was: another year in which the safety and reliability of reactor operation met. the requirements of reactor users.
A summary of operating experience and other activities and re-lated statistical data are provided in the following Sections A-H of this report.
4
A.
SUMMARY
OF OPERATING EXPERIENCE 1.
General During the period covered by this report (July 1, 1988 - June 30, 1989), the MIT Research Reactor, MITR-II. was operated on a routine, four days per week schedule, modified as necessary to facilitate the preoperational testing and installation of several in-core experi-ments. When operating, the reactor was normally at a nominal SMW.
It was the twelfth full year of normal operation for MITR-II.
The reactor averaged 57.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per week at full power compared to 64.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per week for the previous year and 80.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per week two years ago.
The reactor is normally at power 90-100 hours / week, but holidays, major maintenance, long experiment changes, waste ship-
- ping, etc.,
reduce the average.
During the past year it was again reduced more than usual because of the installation of several major experiments concerning the production, activation, and transport of corrosion products in pressurized water reactors.
Also, a lot of operation was conducted at low power for the purpose of making meas-urements on the medical therapy room bean.
The reactor usually operates from late Tuesday afternoon until late Friday afternoon, with maintenance scheduled for Mondays / Tuesdays and, as necessary, for Saturdays.
Starting in FY90 (Av. gust 1989), it is anticipated that the 1
reactor will adopt a six week operating cycle in which it is run at full power almost continuously (160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br /> / week) for four weeks for ex-periments and then run intermittently (*24 hours / week) for two weeks.
The reactor was operated throughout the year with 25 elements in the core.
The remaining positions were occupied by an irradiation facility used for a coolant chemistry loop which is designed to reproduce conditions in power reactors and by a solid aluminum dummy.
Compensation for reactivity lost due to burnup was provided by a single refueling that had been performed late in the previous fiscal year.
That refueling had introduced 2600 mS of reactivity which pro-vided sufficient excess margin to compensate for che reactivity loss due to burnup in FY89. Also a refueling was perfomed to rotate seven elements in the C-ring so as to achieve more uniform burnup gradients in those elements.
Seventeen other refuelings were performed.
One was to move spent fuel from the fuel storage ring in the core tank to the spent fuel storage pool.
The other sixteen were for the purpose of making accurate reactivity measurements and trial runs of the Pressurized Coolant Corrosion Loop (FCCL) experimental facility.
The MITR-II fuel management program remains quite successful.
All of the original MITR-II elements (445 grams U-235) have been per-manently discharged.
The average overall burnup for the discharged elements was 42%.
(Note: One element was removed prematurely because I
of excess outgassing.)
The maximum overall burnup achieved was 48%.
Forty-two of the newer, higher loaded elements (506 grams U-235) have been introduced to the core. Of them, eight have attained the maximum allowed fission density.
However, these may be reused if that limit is increased as would seem warranted based on metallurgical studies by DOE.
Another five have, as reported previously to the U.S. Nuclear
(
t Regulatory Commission, been identified as showing excess outgassing and have been removed from service.
As for the other twenty-nine higher loaded elements, they are either currently in the reactor core or have been partially depleted and are awaiting rense in the C-ring.
The capability to ship spent fuel for reprocessing continues to be a major concern.
It had been previously assumed that a DOE-owned shipping cask would eventually be available.
However, it is now understood that NRC will not license the DOE casks.
Accordingly, l
plans are now in progress to utilize one that is commercially avail-able.
Arrangements for this are proceeding and we hope to begin the first of several shipments of spent fuel in April-May 1990.
In the interim, a request has been filed for a temporary increase in the MITR-II's U-235 possession limit.
This is needed to permit uninter-rupted reactor operation.
Please refer to our letters of 14 February 1989 and 24 August 1989 for further information.
Protective system surveillance tests are conducted on Friday evenings after shutdown (about 1800), on Mondays, and on Saturdays as necessary.
As in previous years, the reactor was operated throughout the period without the fixed hafnium absorbers, which were designed to achieve a maximum peaking of the thermal neutron flux in the heavy water reflector beneath the core.
These had been removed in November 1976 in order to gain the reactivity necessary to support more in-core facilities.
2.
Experiments The MITR-II was used throughout the year for experiments and irradiations in support of research and training programs at MIT and elsewhere.
Experiments and irradiations of the following types were conducted:
a)
Prompt gamma activation analysis for the determination of boron-10 concentration in blood and tissue.
This is being performed using one of the reactor's beam tubes.
The analysis is to sup-port our neutron capture therapy program.
b)
Experimental measurements to determine the suitability of various materials to serve as a neutron filter in a medical therapy beam.
These measurements are used to benchmark theoretical predictions.
c)
Tb9 production of M5ssbauer sources by the irradiation of Gd-160 and Pt-196 for studies of nuclear relaxation of Dy-161 in Gd and for the investigation of the chemistry and siructure of gold compounds, d)
Irradiation of archaeological, environmental, engineering materi-als, biological, geological, oceanographic, and medical specimens for neutron activation analysis purposes.
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of gold-198, dyspronium-165, and holmium-166 for E
e)
Production.
medical research,. diagnostic and therapeutic purposes.
f)'
Irradiation of tissue; specimens on particle track detectors for plutonium radiobiology.
g)
~ Irradiation of. semi-conductors to determine resistance to high doses of fast neutrons.
h)
Use.of the facility for reactor operator training.
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Irradiation of geological materials to determine quantities and distribution of fissile materials using solid state nuclear track detectors.
j)
Closed-loop direct digital control of reactor power using a shim blade as well as the regulating rod during some steady-state and transient conditions.
Control laws for adjusting a reactor's neutronic power in minimum time were developed and demonstrated.
These laws are now being extended to permit power increases from suberitical.
k)
Experimental studies of various closed-loop control techniques including digital filters to reduce signal noise.
1)
Design and experimental evaluation of feedback methodologies for-use in conjunction with the minimum time laws.
m)
Development of a technique for the control of core average tem-perature.
(Experimental evaluation is planned.)
n)
Measurements of the energy spectrum of leakage neutrons using a mechanical chopper in a radial beam port (4DHI). Measurements of
.the neutron wavelength by Bragg reflection then permits demon-stration of the-DeBroglie relationship for physics courses at MIT and other universities.
o)
Studies of fast neutron damage to liquid crystal display materials using a delayed neutron detector.
A major research project that will make significant use of the reactor in subsequent years began operation in 1989.
It is a dose reduction study for the light water reactor industry and involves the installation of a pressurized loop in the reactor core to investigate the chemistry of corrosion and the transport of radioactive crud in PWRs.
A gimilar project involving the study of boiling water reactor chemistries and another concerning factors affecting irradiation-assisted stress corrosion cracking are planned.
Both should start in the next few months.
Another major research project that is now making and will con-tinue to make extensive use of the reactor is a program to design a facility for the treatment of glioblastoma (brain tumors) and melano-mas (skin cancer) using neutron capture therapy. This is a collabora-tive effort with the Tufts New England Medical Center.
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'3.
Chances to Facility Desfrn Except for minor changes reported in Section E, no, changes in the facility design were made - during the year.
As indicated in past reports the uranium loading of MITR-II' fuel was increased f rom 29. 7 grams of U-235 per plate and 445 grams per element (as made by Gulf United Nuclear Fuels, Inc., New Haven, Connecticut) to a nominal 34 and 510 grams respectively (made by the Atomics International Division of Rockwell International, Canogs Park, California).
With the excep-tion of six elements (one Gulf, five AI) that were found to be-outgas-sing excessively, performance has been good.
(Please see Reportable Occurrence Reports ~Nos.
50-20/79-4, 50-20/83-2, 50-20/85-2, 50-20/
86-1, 50-20/86-2, and 50-20/88-1. )
The heavier loading results in 41.2 w/o U in the core, based on 7% voids, and corresponds to.the maximum loading in Advanced Test Reactor (ATR) fuel.
Atomics Inter-national completed the production of 41 of the more highly loaded elements in 1982, 36 of which have been used to some degree.
Eight with about 40% burnup have been discharged because they have attained the fission density limit.
Additional elements are now being fabri-cated by Babcock & Wilcox, Navy Nuclear Fuel Division, Lynchburg, Virginia.
Six of these have been received at MIT and are now in use.
The MITR staff has been following with interest the work of the Reduced. Enrichment for Research and Test Reactors (RERTR) Program at Argonne National Laboratory, particularly the development of advanced fuels that will permit uranium loadings up to several times the recent upper limit of 1.6 grams total uranium / cubic centimeter.
Considera-tion of the thermal-hydraulics and reactor physics of the MITR-II core design show that conversion of MITR-II fuel to lower enrichment must await the successful demonstration of the proposed advanced fuels.
4.
Chances in Performance Characteristics Performance characteristics of the MITR-II were reported in the "MITR-II Startup Report".
Minor changes have been described in pre-vious reports. There were no changes during the past year.
5.
Changes in Operating Procedures Related to Safety There were no amendments to the Facility Operating License during the last year.
However, two amendments have been requested.
These are noted in section E of this report.
With respect to operating procedures subject only to MITR inter-nal review and approval, a summary is given below of those changes implemented during the past year. Those changes related to safety are discussed in section E of this report.
l l
a)
PH 2.7.4,
" Removal of Spent Fuel," requires discharge of MITR fuel once a certain percent of the fission density limit, as defined by the Technical Specifications, has been attained.
The purpose of this internal limit, which is more conservative than that given in the technical specifications, is to allow for uncertainties in the U-235 loading, flux distribution, and power
4 level.
Each of these factors can affect the calculation of the fission density.
The limit in PM 2.7.4 was revised to reflect a more accurate determination of the uncertainty in the U-235 load-ing for MITR fuel elements.
(SR #0-88-12) b)
The administrative procedures, Chapter 1 of the Procedure Manual, were revised to update the lists of names and committee member-ships.
(SR #0-89-1) c)
PM 6.1.3.11,
" Emergency Power Transfer Test," was revised to include an explicit requirement for the notification of those l
affected by the test.
(SR #0-89-7) d)
PM 4.4.4,
" Emergency Operating Procedures," was revised to include a requirement to notify civil authorities of an emergency l
condition within fif teen' minutes of such a condition's being declared.
(SR #0-89-10) e)
A " Nuclear Instrument Connection Verification Worksheet" was formally issued as part of the quality assurance program for servicing nuclear instrumentation.
This was issued as part of a corrective action to ROR #89-1.
The worksheet provides a formal mechanism and permanent record for documenting changes to the nuclear instrumentation.
(SR #0-89-11) f)
PM 1.10.3,
" Operator Qualification Sheet for Radial Beam Port Use," was issued to provide a formal means for training students on the use of the radial beam ports.
(Similar procedures exist for other facilities.
This one became necessary to facilitate student use of a port previously used only by senior experi-menters.)
PM 1.10.3 also serves to provide a written record of student training on the facility.
(SR #0-89-12) l j
g)
PM 1.2, " Security and Visitor Control," was revised to include a i
provision for documenting the fact that packages being brought out of the facility by visitors were inspected.
(SR #0-89-13) h)
A procedure was issued for the performance of axial gamma scans in in-core sample assemblies installed in the core of the MIT Research Reactor.
(SR #0-89-15) i)
Miscellaneous minor changes to operating procedures and to equip-ment were approved and implemented throughout the year.
6.
Surveillance Tests and Inspections There are many written procedures in use for surveillance tests and inspections requirec by the Technical Specifications.
These pro-cedures provide a detailed method for conducting each test or inspec-tion and specify an acceptance criterion which must be met in order for the equipment or system to comply with the requirements of the Technical Specifications.
The tests and inspections are scheduled throughout the year with a frequency at least equal to that required by the Technical Specifications. Twenty-seven such tests and calibra-tions are conducted on an annual, semi-annual or quarterly basis.
r-
q 8-Other surveillance tests are done each time before startup of the reactor if shut down for more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, before startup if a chan-nel has been repaired or de-energized, and at least monthly; a few are i
on different schedules.
Procedures for such surveillance arc incor-porated into daily or weekly startup, shutdown or other checklists.
During the reporting period, the surveillance frequency has been at least equal to that required by the Technical Specifications, and the results of tests and inspections were satisfactory throughout the year for Facility Operatirig License No. R-37.
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a B.
REACTOR OPERATION
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Information on energy generated and on reactor operating hours is tabulated belows i
Quarter 1
2 3
4 Total
- 1. Energy Generated (MWD):
a) MITR-II (MIT FYB9) 142.4 114.4 56.6 119.7 433.1 (normally at 4.9 MW) b) MITR-II (MIT FY76-88) 9,680.7 c) MITR-I (MIT FY59-74) 10,435.2 d) Cumulative, MITR-I 20,549.0
& MITR-II
- 2. MTTR-II Operation (Hrs):
(MIT FY89) a) At Power 769.7 570.1 292.1 642.3 2,274.2
(>0.5 MW) for Research b) Low Power 130.8 233.5 164.7 168.6 697.6
(<0.5MWhjor Training and Test c) Total Critical 900.5 803.6 456.8 810.9 2,971.8
(
Note: These hours do not include reactor operator and other training conducted while the reactor is at full power for research purposes (spectrometer, etc.)
or for isotope production.
Such hours are included in the previous line.
6 4
1
, I C.
SHUTDOWNS AND SCRAMS During the period of this report there were 10 inadvertent scrams and 8 unscheduled power reductions.
The term " scram" refers to shutting down of the reactor through protective system action when the reactor is at power or at least critical, while the term " reduction" or " shutdown" refers to an unscheduled power reduction to low power or to suberitical by the reactor operator in response to an abnormal condition indication.
Rod drops and electric power loss without protective system action are included in shutdowns.
The following summary of scrams and shutdowns is provided in approximately the same format as for previous years in order to facilitate a comparison.
I.
Nuclear Safety System Scrams Total a)
Channel #6 noise due to oxide layer on signal connector.
4 b)
Channel #4 due to improper closing of medical shutters.
I c)
Channel #1 and 3 off scale low due to noise on channel #1.
I d)
Withdraw Permit Circuit open due to weak relay.
1 Subtotal 7
II.
Process System Scrams a) nigh reactor outlet temperature recorder due to low battery voltage.
I b)
High reactor outlet temperature recorder due to defects in new battery.
I c)
Operator error when servicing primary flow /AT recorder.
1 Subtotal 3
l
- l.,
III. Unscheduled Shutdowns or Power Reductions a)
Shutdown due to main exhaust damper failed closed.
I b)
Shutdown due to high reading on auxiliary core l
purge monitor which later was determined to be malfunctioning.
I j
c)
Operator shut reactor down to:
1)
Investigate channels #4 and 5 being OOC.
(ROR #89-1.)
1 11)
Investigate loss of D,0 gasholder level indication.
1 d)
Operator lowered power to investigate or corrects i)
High core purge reading.
2
- 11) Low level D 0 dump tank alarm.
(Alarm was 2
improperly set.)
1 111) Low flow auxiliary pump caused by a clogged filter.
1 Subtotal 8
Total 18 Experience during recent years has been as follows for scrams and unscheduled shutdowns:
Fiscal Year Number 85 10 86 27 87 21 88 21 89 18
o D.
MAJOR MAINTENANCE Major maintenance proj ects during FY89, including the effect, if any, on safe operation of the reactor are described in this section.
Major maintenance items were continued to be performed in FY89 in anticipation of supporting the necessary requirements of the upcoming dose reduction projects for light water reactors.
These proj ects are the Pressurized Water Reactor Coolant Corrosion Loop (PCCL), the Boil-ing Water Reactor Coolant Corrosion Loop (BCCL), and the Intragranular Radiation Assisted Stress Corrosion Cracking (IASCC) Loop.
The upper travel limit on the three-ton crane hoist was raised by eighteen inches so as to allow sufficient clearance for the PCCL thimble to be inserted into anc removed from the core tank to the shielded assembly /
disassembly area.
The shielded assembly / disassembly area was con-structed with high density concrete blocks.
A manual hoist which has a one thousand pound capacity was designed and installed above the shielded area so as to allow the installation and removal of the PCCL's internal structure without occupying the containment crane.
The hoist was constructed from high strength aluminum I-beams and with a rope and pulley system.
The hoist can be stowed away when not in use without any interference with the containment crane.
In addition to the monitoring equipment for the FCCL in the control room, areas on platforms located a few feet below the level of the reactor top were used to set-up the main control panel and equip-ment, both mechanical and electrical, for the PCCL.
The main control panel and the charging system are on the platform on one side of the reactor top and the let-down and chemistry monitoring systems are on another platform.
The main circulating pump is located on the reactor top so as to minimize the piping length from the pump to the reactor core.
Lead shielding is constructed a vund the pump to shield the N-16 radiation emanating from the PCCL fluid.
In FY89, all tests and certification runs for the PCCL that were required by the MITR Safeguards Committee were completed successfully.
Production runs for data collection began in the last quarter of FY89.
The design stage of the BCCL is close to completion.
An initial in-core trial fit-up and measurements of reactivity effects due to void formation in the BCCL were performed.
The reactivity measure-ments showed that the bubble formation effects in the BCCL were undetectable by the reactor instrumentation. Additionally, an in-core gamma flux measurement was performed to provide data for the design of the IASCC project.
The upper shield access ring is a lead filled steel weldment sup-ported by the upper shield ring situated above the upper core tank.
The inner cylindrical surface of the upper shield access ring is clad with a thin layer of 304 stainless steel and the other surfaces are protected by epoxy paint.
The gasket which seals water from entering the interface between the upper shield access ring and the upper core tank had deteriorated and allowed rust to form on the non-stainless surface of the shield ring.
The rust particles were in-turn carried
away by the surge action of the primary coolant at the overflow level.
The end result was that the depletion rate of the ion column in the pnmary cleanup-system was accelerated from about one column for every eight weeks to about three to four weeks per column.
The remedy for this situation was to remove all penetrations and attachments on the upper shield access ring, lift the top shield lid and upper shield access ring from the reactor top, and remove all rust from all sur-faces with a grinder and wire brush.
All non-stainless surfaces on the upper shield ring, upper shield access ring, and the top shield lid were then primed and repainted with epoxy paint.
The performance of the ion-columns after the refurbishment has been fourteen to six-teen weeks per column.
Efforts to reduce the production of Ar-41 f rom activation of air
.were continued in FY89. Methods of detecting gas leaks as reported in the FY88 annual report were used and numerous seals were installed especially around the biological shields and port plugs.
The purge gas in the graphite region was converted from helium to CO,.
The rate at which the purge gas was introduced into the system was increased from two to four cycles per day when using helium to tec1ve to fourteen cycles per day when using CO,.
The reduced leakage combined with the higher cover gas purge rate provided a lower air content in and around the graphite region and the biological shields.
The over-all result is that the production of Ar-41 has been decreased such that the monthly averaged effluent from the reactor stack is now at about ten percent of the limit as specified in the technical specifi-cations compared to about 20 to 30 percent previously.
While installing gas seals around the 6RH1 rotary port, neutron and gamma shields were rearranged with additional layers of shielding materials. This reduced the neutron background in the area-by as much as a factor of 10.
As a result, shadow shields which were used for nearby experiments were no longer necessary.
The cadmium curtain which prevents neutrons from entering the thermal column region is moved vertically by two stainless steel air-craft cables attached to a rotating shaft.
One of the two cables broke after prolonged use.
The cables were buried in heavy shielding near the primary inlet and outlet pipes. The cables were replaced and the operating mechanism which included the clutches and pulleys were rebuilt.
A new limit switch was also installed to provide physical position indication of the curtain at the end of its travel.
l The cooling jacket in the 3GV6 facility developed a leak and allowed shield coolant water to leak into the space between the upper annular ring and the lower annular ring.
The Jeak was not realized until the level in the shield coolant storage tr.nk showed a decrease.
The faulty cooling jacket was removed, the water was pumped out, and a new jacket was installed.
A leak probe made of a pair of conductive wires mounted on a plastic plate was installed in the space between the upper annular ring and the lower annular ring.
The leak probe will provide early detection of leaks in the 3GV facilities in the future.
The air leak located at the expansion joint of the containment
sf vehicle (truck) lock was repaired with a silicone rubber sealant. The truck lock passed the 1989 containment pressure test.
The root cause of the failure was determined to be accumulated rain water on the top exterior portion of the expansion joint.
Entry points for the rain water have been sealed and periodic inspections are performed.
The condition of the expansion joint will be monitored routinely.
The D O transfer pump (DM-2) developed a small leak at the shaft seal.
The pump motor, shaft sleeve, and shaft seal were replaced with new components. The main containment isolation damper hydraulics were drained, flushed, and refurbished with new pumps, valves, and filters to increase system performance and reliability.
A new dual-pen level recorder was installed for the graphite reflector's purge gas and for the D0 reflector's cover gas.
The 3
Channel One neutron-sensitive ion chamber was replaced and another neutron sensitive chamber was installed in the vertical 3GV3 facility for future use.
Many other routine maintenance and preventive maintenance items were performed throughout the year.
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_35_
E.
SECTION 50.59 CHANGES, TESTS, AND EXPERIMENTS This section contains a description of each enarse to the f acili-ty or procedures and of the conduct of tests and experiments carried out under the conditions of Section 50.59 of 10 CFR 50, together with a summary of the safety evaluation in each case.
The review and approval of changes in the f acility and in the procedures _as described in the SAR are documented in the MITR records by means of " Safety Review Forms".
These have been paraphrased for this report and are identified on the following pages for ready refer-ence if further information should be required with regard to any item.
Pertinent pages in the SAR have been or are being revised to reflect these changes, and they will be forwarded to the Director, Standardization and Non-Power Reactor Project Directorate, Office of Nuc1 car Reactor Regulation, USNRC.
The conduct of tests and experiments on the reactor are normally documented in the experiments and irradiation files.
For experiments carried out under the provisions of 10 CFR 50.59, the review and ap-proval is documented by means of the Safety Review Form.
All other experiments have been done in accordance with the descriptions pro-vided in Section 10 of the SAR, " Experimental Facilities".
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a e.
Pressurized Coolant Corrosion Loop (PCCL)
SR #0-86-9 (04/21/88), 0-88-4 (07/28/88), 0-88-5 (09/09/88), 0-88-14 (12/07/88), 0-89-2 (01/06/89), 0-89-3 (01/19/89), 0-89-6 (01/24/89),
0-89-9 (06/02/89), 0-89-14 (06/19/89)
This proj ect involves the design, installation, and operation of a pressurized light-water loop in the MITR core for tne purpose of studying the production, activation, and transport of corrosion products.
The effect of various water chemistries is being examined to determine the optimum method for reducing the creation of activated corrosion products (crud) and thereby reducing radiation fields asso-ciated with pressurized water reactors (PWRs).
The ultimate goal is to reduce radiation exposures to PWR maintenance personnel.
Approval for the PCCL was given by the MITR Staff and the MIT Reactor Safeguards Committee on 04/21/88.
It was determined at that time that no unreviewed safety question existed because rio f ailure or accident associated with the PCCL could lead to an accident or failure involving reactor components. Details of that determination, together with safety review #0-86-9, were submitted to the U.S. Nuclear Regula-tory Commission on 04/21/88.
Subsequent to the determination that no unreviewed safety ques-tion existed, specific procedures for PCCL operation were prepared.
These included:
Procedure for Ex-Core Testing Supplement to the Safety Evaluation Report Preoperational Test Procedure Abnormal Operating Procedures for the FCCL Procedures for PCCL Startup/ Shutdown Procedures for PCCL Installation / Removal Experiments using the PCCL began in April 1989 and have been quite successful.
I
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t Boilina Coolant Corrosion Loop'(BCCL)
SR #0-89-14 (06/19/89)
This project involves the design,' installation, and operation of a boiling light-water loop in the MITR core for the purpose of study-ing the production, activation, and transport of corrosion products.
The effect of various water chemistries will be examined.to determine the -' optimum method for' reducing the creation of activated. corrosion products _(crud) and-thereby reducing radiation fields associated with boiling water' reactors (BWRs).
The ultimate goal is to reduce radia-
-tion exposures to BWR maintenance personnel'.
This facility is currently in the design process and approval has not. yet been requested of the MIT Reactor Safeguards Committee.
Preliminary design reviews have, of course, been held.
Thus far, the Reactor Staff has made a determination that boiling within an in-core facility is not contrary to the technical specifications provided that
- reactivity limits for movable experiments are not exceeded.
It was also concluded that boiling in the proposed experiment volume would not significantly affect reactor operation.
Accordingly, a carefully controlled experiment was proposed to demonstrate that boiling within an in-core f acility would not adversely affect reactor operation.
Following both a determination that no unreviewed safety question was involved and approval by the MIT Reactor Safeguards Committee, this experiment was conducted.
The results were as expected.
Details of the BCCL's design and its safety evaluation will be forwarded once they are finalized.
{-
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a; Experiments Related to Neutron Capture Therapy SR #0-89-4 (01/23/89), 0-89-8 (03/01/89).
In' conjunction with. the Tuf ts New England Medical Center and with the support of the U.S. Department of Energy, MIT is designing an epi-
. thermal neutron beam for the treatment of brain cancer (glioblastoma).
Thermal beams have been used.successfully for this treatment in Japan.
The reason for designing.an epithermal beam is to allow tumor treat-ment without-having to subj ect the patient to ' surgery involving removal of.a portion of the skull.
Also, an epithermal _ beam gives greater penetration.
Thus far, the research has consisted ~of-simula-tion studies using Monte Carlo codes. and experiments. using the MIT Reactor's medical beam to verify those studies.
Two facility changes have.been made. These are (1) ~ Installation of a liner and a support plate in the medical therapy beam. These were installed to permit the subsequent installation of candidate filter materials for producing an
' epithermal beam.
(SR #0-89-4)
(2) Installation of a candidate fliter.
Currently, filters con-taining sulphur and aluminum with small quantities of lithium and cadmium appear to give the best results.
One such filter was installed on a trial basis for the purpose of confirming the results of the simulation studies.
(SR
- 0-89-8)
Neither ' of these design changes was judged to involve an unreviewed safety question.
Experimental results achieved using this filter appear quite promising and we are hopeful of obtaining funding to permit a permanent installation once the design is finalized.
l t-
DIRital Computer Control of Reactors Under Steadv-State and Transient Conditions SR#-M-81-3 (11/17/81),
M-81-4 (12/10/81),
E-82-2 (01/08/82),
E-82-3 (02/24/82),
E-82-4 (03/03/82),
E-82-5 (04/14/82),
E-82-6 (07/13/82), 0-83-5 (02/03/83), E-83-1 (02/08/83), 0-83-12 (04/23/83),
0-83-20 (07/20/83), 0-84-11 (06/25/84), 0-84-12 (07/12/84), 0-84-16 (12/6/84), 0-84-21 (11/1/84), 0-85-11 (5/9/85), 0-85-13 (6/28/85),
0-85-16 (7/12/85),
0-85-20 (8/16/85),
0-85-25 (12/1/85),
0-85-26 12/1/85), 0-86-11 (10/17/86), 0-86-13 (11/28/86), 0-87-11 (6/1/87),
0-87-17 (12/24/87), 0-88-10 (12/01/88).
The proj e ct involving computer analysis, signal validation of data from reactor instruments, and closed-loop control af the MIT Reactor by digital computer was continued.
A non-linear supervisory algorithm has been developed and demonstrated.
It functions by restricting the net reactivity so that the reactor period can be rapidly made infinite by reversing the direction of control rod motion.
It, combined with the signal validation procedures, ensures that there will not be any challenge to the reactor safety system while testing closed-loop control methods.
Several such methods, including decision analysis, rule-based control, and modern control theory, continue to be experimentally evaluated.
The eventual goal of this program is to use fault-tolerant computers coupled with clnsed-loop digital control and signal validation methods to demonstrate the improvements that can be achieved in reactor control.
Each new step in the program is evaluated for safety in accor-dance with standard ; eview procedures (Safety Review numbers listed above) and approved as necessary by the MIT Reactor Safeguards Commit-tee.
Initial teste of this digital closed -loop controller were con-ducted in 1983-1984 usf rg the f ac;11t y's regulating rod which was of relatively low reactivity worth (0.2% AK/K).
Following the successful completion of these tests, f acility operating license amerMment No. 24 was obtained from NRC (April 2, 1985).
It permits:
(1) closed-loop control of one or more shim blades and/or the regulating rod provided that no more than 1.8% E/K could be inserted were all the connected contrcl ele-ments to be withdrawn, (2) closed-loop control of one or more shim blades end/or the regulating rod provided that the overall ccatroller l
is designed so that reactivity is constrained suffi-ciently to permit control of reactor power within desired or authorized limits.
A successful experimentation program is now continuing under the provisions of this license amendment.
A protocol is observed in which
)
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.j
this controller is used to monitor, and if necessary override, other novel controllers that are still in development.
Tests performed during this reporting period primarily involved evaluation of various -
feedback methodologies in conjunction with the MIT-SNL Period Gener-ated Minimum Time Control Laws.
These - laws adjust the rate of change i
of reactivity so that the reactor period is maintained constant at a.
specified value.
As a result, power changes are accomplished in mini-mum time for a reactor limited to the'specified period.
These tests were conducted under the provisions of technical specification #6.4 using our now standard protocol in which reactivity is constrained by a supervisory controller that maintains " feasibility of control".
Signal implementation is accomplished using a variable speed stepping motor.
This motor-is installed prior to the tests and removed upon their completion.
An independent hard-wired circuit is used to moni-tor motor speed and preclude an overspeed condition.
The conduct of these tests was approved by the MIT Reactor Safeguards Committee.
Proportional, proportional-integral, and proportional-integral-derivative feedback schemes were-tested extensively.
The second of these was found to give superior results.
For the above tests, it was concluded that no unreviewed safety question' existed.
The experiments were. conducted under the supervi-sion of a licensed senior operator.
Also, use of the now-standard supervisory algorithm and the independent hard-wired circuits limited the possible envelope of operating conditions.
In particular, there was - no possibility of control mechanism withdrawal such that the allowed rate of insertion of positive reactivity could be exceeded.
Hence, for neither experiment was there an increase in the probability of an analyzed accident, a possibility of a new type of accident, or a decrease in a safety margin defined in the basis of any technical specification.
Plans currently call for experimental evaluations in which the MIT-SNL laws will be used to perform power increases from suberitical and in which techniques for the control of core average temperature will be demonstrated.
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e Installation of Waste Tank External Filter SR #0-88-7 (09/29/88), 0-88-11 (12/01/88)
An external filter was installed on the discharge of the liquid waste effluent system in order to provide further assurance that only soluble material is discharged.
The proposed change was judged not to involve an unreviewed safety question and was approved by the MIT Reactor Safeguards Committee.
The filter was subsequently installed and PM 3.6, " Waste Storage Tank Dump Procedure" was revised. Affected portions of the MITR Safety Analysis Report were revised and submitted to the U.S. Nuclear Regulatory Commission as Revision No. 36.
Addi-tional information may be obtained from that submission.
- s e
Revision of Fission Density Limit SR #0-88-12 (12/01/88)
The fission density limit for the UA1 fuel used by the MIT x
Research Reactor is 1.8 10: 8 fissions /cc.
Research conducted by the Idaho National Engineering I.aboratory (Nucl. Tech., 49, 136-149, June
- 80) shows that a limit of 2.3 1088 fissions /cc is technically justi-fled.
Analysis of the MITR fuel cycle showed that increasing the MITR fission density limit to 2.3 1088 fissions /cc would eventually reduce the overall number of elements in the cycle.
Accordingly, a safety analysis was prepared and, following review and approval by the MIT Reactor Safe 5uards Committee, submitted to the U.S. Nuclear Regulatory Commission on 13 February 1989.
Approval of this proposed change is still pending.
Additional information may be obtained from the sub-mission.
Temporary Increase in U-235 Possession Limit SR #0-88-13 (12/20/88)
As was noted in section A of this report, the DOE-owned shipping casks that MIT had been planning to use were not licensed by NRC.
Accordingly, preparations are now being made for use of a commercial cask.
Pending establishment of this capability, it will be necessary to obtain an increase in the U-235 possession limit.
A safety analy-sis was performed and following approval by the MIT Reactor Safeguards Committee submitted to the U.S.
Nuclear Regulatory Commission on 14 February 1989.
A request for further information was received on 17 April 1989.
This was answered on 24 August 1989.
Approval of the request is therefore still pending.
Additional information is avail-able in both the original submission and in our letter of 08/24/89.
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Security Plan Revisions SR #0-88-8 (11/04/88), 0-88-9 (11/08/88)
Various security procedures were revised to reflect administra-tive ' and other minor changes.
Revision No. 39 to the MITR Safety l Analysis Report was submitted to reflect these changes.
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Change of Graphite Reflector, Cover Gas frcm Helium to Carbon Dioxide SR #0-87-20 (03/08/86)
The cover gas used in the reactor's graphite reflector region was changed from helium to carbon dioxide early in 1988.
This was done as part of the program to reduce percentage of air in the graphite region and hence reduce argon emissions.
An evaluation of the effectiveness of. this conversion remains on-going.
The results to date indicate that emissions have been reduced and that no adverse chemical reactions have occurred.
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F.
ENVIRONMENTAL SURVEYS Environmental surveys, outside the facility, were performed using area monitors.
The syr.tems (located approximately in a 1/4-mile radius from the reactor site) consist of calibrated G.M.
detectors with associated electronics and recorders.
Environmental monitors using film badges on a quarterly basis indicate minimal detectable results for all sectors each year.
The detectable radiation levels due to argon-41 are listed below:
Site July 1, 1988 - June 30, 1989 North 0.12 mR/ year South 0.2 mR/ year
- East 0.44 mR/ year West 0.13 mR/ year Green (East) 0.17 mR/ year Fiscal Yearly Averanes:
1979 1.5 mR/ year 1980 1.9 mR/ year 1981 1.9 mR/ year 1982 2.5 mR/ year 1983 2.3 mR/ tear 1984 2.1 mR/ year 1985 2.2 mR/ year 1986 1.8 mR/ year 1987 1.2 mR/ year 1988 1.2 mR/ year 1989 1.1 mR/ year
- Estimated value based on average weight of previous years and sectors while system was not in service.
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G.
RADIATION EXPOSURES AND SURVEYS WITHIN THE FACILITY A summary of radiation exposures received by f acility personnel and experimenters is given below:
Period 7/01/88 - 6/30/89 Whole Body Exposure Range (Rems)
No. of Personnel No Measurable....................................
43 Measurable - Exposure less than 0.1..............
98 0.1 0.25......................................
32 0.25 - 0.5.......................................
13 0.5 0.75............
......................... 3 0.75 -
1.0........................................
0 Total Personnel = 189 Total Man Rem = 9.71 Summary of the results of radiation and contamination surveys from July 1988 to June 1989:
During the 1988-1989 period, the Reactor Radiation Protection Office continued to provide radiation protection services neces-sary for full-power (5 megawatt) operation of the reactor.
Such services (performed on a daily, weekly, or monthly ochedule) include, but are not limited to, the following:
1.
Collection and analysis of air samples taken within the con-tainment shell, and in the exhaust-ventilation system.
2.
Collection and analysis of air samples taken from the cool-ing towers, D,0 system, waste storage tanks, shield coolant, heat exchangers, fuel storage facility, and the primary system.
3.
Performance of radiation and contamination surveys, radio-active waste collection, calibration of reactor radiation monitoring
- systems, and servicing of radiation survey meters.
4.
Providing of radiation protection services for control rod removal, spent-fuel element transfers, ion column removal, etc.
The results of all surveys described above have been within the guidelines established for the facility.
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H.
RADIOACTIVE EFFLUENTS This section summarizes the nature and amount of liquid, gaseous and solid radioactive wastes released or discharged from the facility.
1.
Mc;uld Waste Liquid radioactive wastes generated at the facility are dis-charged only to the sanitary sewer serving the facility.
There were two sources of such wastes during the years the cooling tower blow-downs and the liquid waste storage tanks.
All of the liquid volumes are measured, by far the largest being the 4,217,000 liters discharged during FY 1989 from the cooling towers.
(Larger quantities of non-radioactive waste water are discharged to the sanitary sewer system by other parts of MIT, but no credit for such dilution is taken since the volume is not routinely measured.)
All releases were in accordance with Technical Specification 3.8-1, including Part 20, Title 10, Code of Federal Regulations.
All activities were substantially below the limits specified in 10 CFR 20.303, but the monthly tritium releases are reported in Table H-3 in accordance with Technical Specifications 3.8-1 because its concentra-tion exceeded 3x10-* pCi/ml.
2.
Gaseous Waste Gaseous radioactivity is discharged to the atmosphere from the containment building exhaust stack and by evaporation from the cooling towers.
All gaseous releases likewise were in accordance with the Technical Specifications and Part 20, and all nuclides were below the limits of 10 CFR 20.106 after the authorized dilution factor of 3000.
Also, all were substantially below the limits of 10 CFR 20, Appendix B, Note 5, with the exception of argon-41, which is reported in the following Table H-1.
The 1529 Ci of Ar-41 were released at an average concentration of 0.39 x 10-'
Ci/ml for the year.
This represents 9.8% of MFC (4 x 10-' pCi/ml) and is significantly less than the pre-vious year's release of 2627 C1.
The decrease is due to a combination of factors including the sealing of leaks and the temporary reduction in operating hours.
3.
Solid Waste Only one shipment of solid waste was made during the year, infor-mation on which is provided in the following Table H-2.
4.
Liould Discharge to the Sanitary SeweraRe System Total gross beta activity in the liquid effluents (cooling tower blowdowns and waste storage tank discharges) amounted to 0.0034 Ci for FY1989.
The total tritium was 0.106 C1.
The total effluent water volume was 4,217,000 liters, giving an average tritium concentration of 25.3 x 10-* pCi/ml.
l The above liquid waste discharges are provided on a monthly basis in Table H-3 attached.
e d-TABLE H-1 ARGON-41 STACK RELEASES ESCAL YEAR 1989 Ar-41 Average Discharged Concentration (Cu: ries)
(pCi/ml)
July 1988 408 1.34 x 10-'
August 412 1.36 September 332 0.87 October 83 0.27 November 72 0.19 December 15 0.05 January 1989 14 0.05 February 34 0.11 March 36 0.09 April 26 0.08 May 60 0.15 June 37 0.12 I
Totals (12 Months) 1529 0.39 x 10-*
MPC (Table II, Column I) 4 x 10-8
% MPC 9.8%
( }
Note:
After authorized dilution factor (3000).
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, TABLE H-2 i
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SUMMARY
OF MITR RADIOACTIVE SOLID WASTE SHIPMENTS FISCAL YEAR 1989 Units Shipment #1 Total 1.
Solid waste packaged Cubic 135 135 Feet 2.
Weight Founds 4195 4195 3.
Total activity Ci 0.053 0.053 (irradiated components, ion exchange resins, etc.)
soCo, ' 1 Cr,
'*-Fe
Zn, etc.
4.
(a) Date of shipment 05/08/89 (b) Disposition to licensee for burial U.S. Ecology, Inc.
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4 i
- TABLE H-3 LIQUID WASTE DISCHAPGES FISCAL YEAR 1989 Total Volume Average Gross Beta Total of Effluent Tritium Less Tritium Tritium Water (I)
Concentration
~0
~0 (x10 C1)
(x10~ C1)
(x10 liters)
(x10 pCi/ml)
July 1988 NDA(
7.30 40.1 18.2 Aug.
108 21.4 41.9 41.9 Sept.
NDA 5.51 50.3 10.9 Oct.
86.8 11.6 56.6 20.5 Nov.
NDA 3.44 37.0 9.3 Dec.
NDA 0.84 7.8 10.8 Jan. 1989 NDA 0.91 12.2 7.5 Feb.
2334 24.0 15.8 15.2 Mar.
NDA 1.99 19.9 10.0 Apr.
NDA 1.08 17.6 6.1 May 313 25.8 43.1 59.9 June 510 2.84 79.4 3.6 12 months 3351.8 106.7 421.7 25.3 Note _s (1) Volume of effluent from cooling towers and waste tanks.
Does not in-clude other diluent from MIT estimated at 2.7 million gallons / day.
(2) No Detectable Activity; less than
- 1. 2 6 x 10-6 pCi/ml beta for each sample.
_ -