ML20059G433
| ML20059G433 | |
| Person / Time | |
|---|---|
| Site: | MIT Nuclear Research Reactor |
| Issue date: | 06/30/1990 |
| From: | Bernard J, Kwok K MASSACHUSETTS INSTITUTE OF TECHNOLOGY, CAMBRIDGE |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9009120305 | |
| Download: ML20059G433 (31) | |
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NUCLEAR REACTOR LABORATORY: -
. AN INTERDEPARTMENTAL CENTER OF.-
MASSACHUSETTS INSTITUTE OF TECHNOLOGY 0.K. HARLING 138 Albany Street Cambrdge, Mass. 02139 i JA BERNARD, JR.
Director Telefax No. (617)253 7300 Director of Reactor Operations Teler No. 921473-MIT-CAM -
Tel. No. (617y253 42ttM202 L s
August 30,1990 '
U.S. Nuclear Regulatory Commission Washington, D.C. 20555 A*ITN: Document Contml Desk ,
Subject:
Annual Rep >rt, Docket No. 50-20, License R-37, Technical Specification 7.13.5 Gentlemen: 1
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Forwarded herewith is the Annua 1 Report for the MIT Research Reactor for the period July 1,1989 to June 30,1990, in compliance'with paragraph 7.13.5 of the !
Technical Specifications for Facility Operating License No. R-37.
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Sincerely, Kwan S. Kwok;
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. Superintendent, . j Reactor Operations e M ' Q.' !
ohn A. Bernard, Ph.D.
Director of Reactor Operations KSK/gw
Enclosure:
As stated cc: MITRSC USNRC - Region I - Chief, Reactor Projects Section 3A USNRC- RegionI-ReactorEngineer, Reactor Projects Section 3A USNRC- Senior Resident Inspector, Pilgrim Nuclear Station USNRC- ProjectManager
, Standardization and Non Power Project Directorate i h- USNRC - Acting Project Manager, I
Standardization and Non-Power Project Directorate 9009120305 900630 PDR R
ADOCK 05000020 001.54 /6 /
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MIT RESEARCH REACTOR -
- ANNUAL' REPORT-TO e
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UNITED STATES NUCLEAR REGULATORY _ COMMISSION FOR TIIE PERIOD JULY _1,1989 ; - JUNE 30,1990' 1
i BY j j
REACTOR STAFF i
August 30,1990 -
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e Table of Contents ,
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Table of Con tents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-l.
In tF0 duction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
A. Summary of Operating Experience . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 -
3
-B. Reactor Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8- .
C. S hutdowns and Scrams . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 ,
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E.- Section 50.59 Changes, Tests, and Experiments ......... . .. ...... ... ......... .... ~ 13 F. Environmental S urveys . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 G. Radiarian Exposures and Surveys Within the Facility ,........................... 24 '
H. Radioactive Effluents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 I 1
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4-MIT RESEAROI REACTOR :
ANNUAL REPORT TO q
u U.S. NUCLEAR REGULATORY COMMISSION o FOR THE PERIOD JULY 1.1989 - JUNF 30.1990 9
This report has been prepared by the staff of the Massachusetts Institute ofn Technology Research Reactor for submission to the Administrator of Region I, United States Nuclect Regulatory Commission, in compliance with the requirements ~of the Technical Sjecifications to Facility Operating Ocense No. R-37 (Docket No' 50-20),- .
Paragraph 7.13.5, which requires an annual report following the 30th of June of each year.' l The MIT Research Reactor (MITR), as ori all !
of MTR-type fuel, fully enriched in uranium 235 arb'coobconstructed,~ consisted of a core -
and moderated by heavy water 1 in a four foot diaineter core tank, surrounded by a graphite reflector. After inidal enticality 1 on July 21,1958, the first year was devoted to startup experiments, calibration, and a: '
gradual rise to one megawatt, the initially licensed maximum power. Routine three-shift operation (Monday-Fnday) commenced in July 1959. The authorized power level was1 ';
increased to two megawatts in 1962 and to five megawatts (the design power level) in 1965. i Studies of an impmved design were first undertaken in 1%7.1 The concept which was finally adopted consisted of a more compact core, cooled by I ght water, and i siirrounded laterally and at the bottom by a heavy water reflector. It is un ted for. j the purpose of maximizing the peak of thennal neutrons in the heavy water at the ends of a the xam port re-entrant thimbles and for enhancement of the neutron flux, arl the fast component, at in-core irradiation facilities. The core is hexagonal in' ape,'15i hes across, and utilizes fuel elements which are rhomboidal in cross section and which contain '
UAL intertnetallic x fuel in the form of plates clad in aluminum and fully enriched _ int q uranium-235. Much of the original facility, e.g. graphite reflector, biological and thermal shields, secondary cooling systems, containment, etc., has been retained; ,
After 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-1 was shut down on May 24, 1974, having logged 250,445 megawatt hours during nearly 16 years of operation.
The old core tank, a'ssociated piping, top shielding, control rods and drives, and some experimental facilities were disass embled, removed, and subsec uently replaced with new equipment. After preoperationa' tests were conducted on al s stems, the _U.S.
Nuclear Regula Commission issued Amendment No.10 to Facility ting License No. R-37 on July 3,1975. 'After initia.1 criticality for MITR-II on August 14,1975,and several months of startup testing, pow er was raised to 2.5-MW in December. Routine 5-MW operation was achieved in Decern%r 1976.
This is the fifteenth annual report cequired by the Technical Specifications, and it covers the period July 1,1989 through June 30,1990. Previous reports, along with the
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"MITR II Startup Report" (Report No. MITNE 198, February 14,1977) have covered the ]
startup testing period and the transition to roudne reactor operadon., This report covers the thirteenth 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 opcating experience and other activities and related statistical data .
are provided in Sections A-H of this report.
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3 A.
SUMMARY
OF OPERATING EXPERIENCE i
- 1. Cansal
' The MIT Research Reactor, MITR-II, has traditionally been operated on a routine, .
four days per week schedule, modified as necessary to facilitate the preoperational testing!
and installation of several in-core experiments. .When apamian the reactor is normally at a: :
nominal 5-MW. However, substannal departures were made fro, m this schedule during the !
period covered by this report (July 1,1989 - June 30,1990). Starting in AuJust 1989 and continuing to late Damhar 1989, a six-week operating cycle was adoptec in which the reactor was run at full power almost 9 'tinuously (160 tours / week) for four consecutive . '
l weeks and then run intermittently (M nours/ week) for two weeks. This schedule was '
- followed in order to support a major experimental program concerning the development of methods to reduce the acdvation and transport of corrosion products in pressurized water - ;
reactor coolant. From January to July 1990, the reactor's operating cycle was snore in keeping with the tradidonal one. The period covered by this report was the thhteenth full
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year of normal operation for MITR II.- .>
The reactor averaged 48.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per week at full power compared to 40.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per-week for the previous year and 45.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> per week two years ago. As was the case in -
FY90 a lot of operation was conducted at low power in order to make measurements of the'.
medical therapy room beam. These measurements are for the ?urpose of designing an 4 epithermal neutron beam for the treatment of brain cancer (gliob astoma multiforme) and possibly skin cancer (melanoma). When neither the corrosion reduction experiments nor -
the medical beam design was in progress, the reactor was us,ually operated from late Tuesday afternoon until late Friday aftemoon, with maintenance scheduled for Mondays-Tuesdays and, as necessary, for Saturdays.
The reactor was operated throughout the year.with 25 elements in'the core. The I remaining positions were occupied by an irradiation facility used for the coolant chemistry )
loop which is designed to reproduce conditions in power reactors and by a solid aluminutn -
dummy. Compensation for reactivity lost'due to burnup was provided by three refuelings.- 0
'Ihese followed standard MITR practice which is to introduce fresh Gel to the inner pcrtion ;
of the core (the A and B-Rings) where peaking is least and and to place partially spent fuel '
in the outer portion of the core (the C-Ring). In addition, elements were inverted and
- rotated so as to achieve more uniform burnup gradients in those elements.' Twenty-one l other refuelings were performed for the purpose of makin'g accurate reactivity l
measurements and runs of the Coolant Chemistry Loop experimental facilities.
The MITR-II fuel management pogram remains quite successful. All of the original MITR-II elements (445 grams U-235) have been prmanently discharged. The average overall burnu for the discharged elements was 4i% (HQia: One element was removed prematurel because of exceu outgassing.) The maximum overall burnup achieved was 48%. - -f ive o fthe newer, higher loaded elements (506 grams U-235) I have been introduced to the core. Of them, eleven 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 1 previously to the U.S. Nuclear Regulatory Commission, been identified as snowing exces 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 i and are awaiting reuse in the C-ring. ;
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- j Protective shutdown (about Ibstem' surveillance
), an Mondays, tests 'are and on Saturdays conducted as necessary. ' on Friday evenings after .
L - As in previot.s years, the reactor was operated throughout the period without the fixed hafnium absorbers, which were designed to achieve a maximum lxaking of the thermal neutron flux in the heavy water reflector beneath the core. Tiese had been ,
removed in November 1976 in order to gain the reactivity +-smy to support more in-- :
core facilities.-- ;
- 2. Experiments 'i The MITR II was used throughout the year for experiments and irradiations in' .'
support of rescatch rd training programs at MIT and elsewhere.
Experiments and irradiarimis of the following types were conducted:
a) Pmmpt g ama activation analysis for the determination of baron-10 concentration - J in blood and tissue.' 'Ihis is bemg performed using one of the reactor's beam tubes.-
The analysis is to support our neutron capture therapy program.
b) Experimental measurements to determine the suitability of various materials to serve i 3 as a neutron filter in a medical therapy beam. These measurements are used to -
benchmark theoretical predictions.
- c) Studies of the material composition of.su wrconducting phases of various alloys were performed by activating samples and t sen identifying characteristic radiations, d) Irradiation of archaeological, environmental, engineering materials, biological. .
geological, oceanographic, and medical specimens for neutron activation analysis purposes, c) Production of gold 198, dysprosium 165, and holmium-166 for medical research, diagnostic, and therapeutic purposes. '
l 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.
i) Irradiation of geolodcal materials to determine quantities and distribution of fissile materials using solk. 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. Controllaws that i are insensitive to modeling inaccuracies are currently being explored. -
l k) Experimental studies of various closed-loop control techniques including digital
, filters to reduce signal noise.
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- 1) Studies of the uptake of boron-containing. chemicals in animal (mouse) tissue.
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m): Neutmn radiography studies to identify incipient cracking in structural compensats wuch as airplane wmss.-
n) Measurements of the energy s trum of leakage neutrons using a mechanic ,
cuyperin a radial beam port (4 HI). Measurements of the neutmn wavelength Bragg reflectti then permits demonstration of the DeBroglie relationship for physm, s courses at MIT and other universides.
- o) Gamme irradiariaa of seeds for demonstration of radiation damage effects for high .
school students.
- p) Study of radiation damage effects on magnets that are intended for use in the
- supercollider. !
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Dose reduction studies for the light water reactor industry began reactor use on a :
regular basis in 1989.: (Planning and out-of-core evaluations had been in progress for >
several years.) Dese studies entail installing loops in the reactor core to investigate the chemistry of corrosion and the transport of radioactive crud.14s _that replicate both pressurized and boiling water reactors have been built. ~ The PWR loop has been operationel and in-core since August 1989. He BWR loop has now been fully tested and experiments involving it should start in September 1990. In addition, an experiment involving irradiatian-assisted stress corrosion cracking is planned.
l Another major research project that is'now making and will continue to make extensive use of the reactor is a program to design a facility for tre treatment of plioblastomas (brain tumors) and melanomas (skin cancer) using neutron capture therapy, his is a collaborative effort with the Tufts New England Medica Center, t
! 3. Manon to Facility Desian
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Except for minor changes reported in Section E, no changes in the facility design
! were made d uring the year. As indicated in past reports the uranium loading of MITR-II-l fuel was increased from 29.7 gams of U-235 per plate and 445 grams per element (as' made by Gulf United Nuclear Huels, Inc., New 1laven, Connecticut) to a naminal 34 and !
510 grams respectively (made by the Atomics-International Division of Rockwell Intemat.onal, Canoga Park, Califomia). With the exception of six elements (one Gulf, five AI) that were found to be outgassing excessively, performance has been good. (Please see l Reportable Occurrence Re l
50-20/86 2, and 50-20/8bmrts -1.) The Nos. 50-20/79-4, heavier 50-20/83 loading results in 41.22,w/o 50-20/85-2, U in tie core, 50-20/86-1, based on 7% voids, and corresponds to the maximum loading in Advanced Test Reactor (ATR) fuel. Atom'cs International completed the production of 41 of the more highly 1
loaded elements in I?82,36 of which have been used to some degme. Eleven with about
+
40% burnup have beca discharged because they have attained t w fission density limit.
Addidonal elements art now bemg fabricated by Babcock & Wilcox, Navy Nuclear Fuel ~,
! Division, Lynchburg, Nirginia. Nine of these have been received at MIT and are now in -
. use.
- The MITR staff has_ been folicwing with interest the work of the Reduced Enric;anent for Research and Test Reactors (RERTR) Program at Argonne National
- Laboratory, particularly the development of advanced fuels that will permit uranium i
loadings up to several times the recent upper limit of 1.6 grams total uranium / cubic centimeter. Consideration of the thermal-hydraulics and reactor physics of the MITR-II
, core desi pt show that conversion of MITR-II fuel to lower enrichment must await the successfu demonstration of the proposed advanced fuels.
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F 4. Changes in Performan e Characterisdes u .
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Performance characteritics of the MITR-II were reported in the "MITR-II Startup p Report." Minor changes have ieen described in previous reports. There were no changes 4 durmg the past year. - j
- 5. chansen in nnetine Prrmimes Relatad to Enferv i Amendment No 25 to the Facility Operating L.icense was issued on 11 December ,
1989. It provides for a temporary increase in the U-235 possession limit. Also, an . '
amendment concerning the fission density limit has been requested. Both'of these -
amendments are noted in Section E of this report. .
With respect to operatint procedures subject ont to MITR internal review and?
approval, a summary is given be ow of thoes chanps lemented during the past year.-
, Those changes related to safety are discussed in secuon E this report. a i
a)- PM 6.4.16, "lew and High levels Cooling Tower Basins No. I and No. 2" was avised to permit use of a new test apparatus that greatly reduces the work involved ' '
in calibranng the level probes. (SR #0 89-17) b) he core configuration map used to document all reactor refuelings was redrawn for' !
compatibility with newly-purchased word-processing equipment. (SR #0 8919) -
l c) PM 6.5.5, " Backup Steam Suppl Availability" was revised to permit condnued testing of the altemate steam sup to the reactor containment tuming. The steam j is used for heating the building. e revision was required because of changes in the MIT boiler system, unrelated to the reactor. (SR #0-89-22) .
d) he administrative procedures, Chapter 1 of the Procedure Manual, were revised to !
update the lists of names .md comnuttee memberships._ (SR #0 901) _
4 e) PM 6.5A1A and PM 6.5.6.2-- PM 6.5.6.4, " Calibration of System Pressure Gauges" were revised to permit use of master gauges traceable to NIST in lieu of a deadweight tester. (SR #0 90-2) i
.l f) Tables defining emergency action levels were added to several of the existing MIT Abnormal Operating Procedures. This was done to facilitate the transition from abnormal to emergency conditions. (SR #0-90-4) g) PM 4.4.4.14 and 4.4.4.15 " Excess Radiation at the Exclusion. Area (Site)
' Boundary Resulting from a Contained Source" and." Escape of. Airborne Radioactive Material from the Containment Building" were revisec to clarify phone numbers for contacting civil authorities. (SR #0-90 5) h) PM 6.5.14, " Calibration of Shim Blade Drop Timer" was revised to allo v use ofJ digital equipment for the calibration. (SR #0-9011) i) A new timer to measure shim blade dro ) times was designed, tested, and installed.
De new equipment is all digital and rep aces out-dated equipment. (SR #E-901) l l
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1 6 SurvdMancs Ta**e and inenactinne ^
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. There are many written procedures in use for surveillance tests and inspections-. '
1 reqdred by the Techaical Specifications. These procedures provide a detailed method for - l cond icting each test or inspection and specify an acceptance criterion which must be met in - .
order for the equipment or system to comply'with the requirements of the Technical {
Speciftetions. Tme tests and inspections are scheduled throughout the year with a i frmuency a least equal to that required by the Technical Specificadons.- Twenty seven suc i tests and catiorations are conducted on an annual, semi annual, or quarterly basis.' i Other surveillance tests are done each time before startup of the reactor if shutdown -I for rante 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 channel has been repaired or de-enerpized, and j
. at kast monthly; a few are on different whedules. Procedures for such survetlance are -
incorporated into daily or weeldy startup, shutdown or other checklists.
i During the reporting period, the surveillance frequency has been at least equal to y that required by the Technical Specifications, and the results of tests and inspections were , t satisfactory throughout the year for Facility Operating License No. R 37. -
- 7. Sentue of Snent Fml Shinnwnt Pursuant to Amendment No. 25 to Facility Operadng Uc'ense No. R-37, paragraph -
2.D.(2) subparagraph (b), nported herewith is the status of t te establishment of a shipping -
capability for spent fuel and other activities relevant to the temporary increase m the possession limit. ,
Relative to the ca ability of shi i following wr.s accomplirL ted in FY90: pping spent fuel from the MIT Research Reactor, the (a) The Certincate of Compliance and the Safety valysis Report of the BMI-l cask; ,
were reviewed by MIT and the cask was determined to be acceptable for shipping ?
MITR spent fuel. Arrangements have been made with bot 3 the cask owner, 4 Cintichem, Inc. and DOE for MIT to use this cask. ,
(b) The University of Missouri Research Reactor (MURR) basket was reviewed and found to be suitable for use with the MIT fuel elements in the BMI-l cask. MURR !
! has agreed to make their basket available to MIT for the required shipments.
(c) A quality assurance propam for MITR-II spent fuel shipment was prepared and ' I
, approved under the safety review program at MIT. . The Q/A prograrc was ' ;
submitted to NRC for approval on 10 August 1990. l l
(d) 'Ihe decay heat load of each spent element was determined by k snember of the .;
MITR staff and found to be within the limits specified in the Certificate of 1 CanaHaac* for the cask. Radiatiori shielding calculations were also performed and I
<. radialion levels associated with the loaded cask are estimated to be within allowed l limits. Criticality calculations are near completion and are being wrformed using '
the Monte-Carlo Code KENO-V which was obtained from the Raciation Shielding Information Center of the Oak Ridge National Laboratory.- Preliminary results show that degree of suberiticality of a cask fully loaded with MIT fuel elements is within specification.
We anticipate shipping spent fuel during IW91.
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a' i B. REACTOR OPERATION i
Information on energy generated and on reactor operating hours is tabulated below: . ]
\
a
,. Quaner 1 2- 3 4 Total
- 1. Energy Generated (MWD):
a) MITR-II(MITFY90) 150.9 148.9 116.7- 93.8 510.3 (normally at 4.9 MW) b) MITR-Il -
- 10,113.8-1 (MIT FY76-89) ,
l c) MITR-I 10,435.2 -
(MIT FYS9-74) d) Cumulative, MrIR-I & MTIR-II 21,059.3 -->
- 2. MITR-II Operation (Hrs): l (MIT FY90) a) At Power
(>0.5-MW) for 829.9 802.1 '646.5 - 499.1 ~ 2,777.6 Research t
b) Low Power
(<0.5-MW) for 121.9 44.2 113.1: 324.5' 603.7.
Training 0)and Test l c) TotalCritical 951.8 846.3- 759.6- 823.6' 3381.3 l
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- 0) These hours do not include reactor operator and other training c' onducted while the l
reactor is at full power for research purposes (spectrometer, etc.) or for isotope production. Such hours are included in the previous line.
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.9 l C. SHU'IDOWNS AND SCRAMS L
During the period of this report there were 16 inadvenent scrams and 4 unscheduled 1 l power reductmas. .
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l 'Ihe term " scram" refers to shutting down of the reacts through protecdve system l 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 suberidcal by thel u
reactor operator in response to an abnormal condition indication. Rod drops and electric power lost, without protective system acdon are included in shutdowns.-
'Ibe following summary of scrams and shutdowns is provided in appraria=*aly the j same format as for preymus years in order to facilitate a comperson. . -
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- 1. Nuclear Saferv Svntem Scrams ,Tgtal te a) Channel M power supply. 9 b) Channel M power supply. ,) ChannelM circuit board failure.- 1 d)- Irxxxtect ECP calculation resulting in too short a period. 1 q
e) Malfunction of signal validation during . ;
digital control experiment.- -1 l
.j O Period channel level signals offscale low during power decre.ase. 1 1
. Subtotal 14-II. Pmeens System Scrame a) Incorrect starting of secondary pumps. 1-b) Incorrect changeout of primary flow recorder chart. 1 J
! Subtotal 2 l I !
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v III. Unscheduled Shutdowns or Power Reductions a) Shutdown due to coolant conosion loop malfunction, 'l i b) Shutdrmn due to loss of offsite electricity, _1._ .
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Subtot
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Total 20
. Experience during recent years has been as follows for scrams'and unscheduled ;
shutdowns: '
EmaQ2ar Hanbar 86 27 87- 21 88 21
- 89 18 90 20 .
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. MAJOR MAIN'IENMCE 1
Major maintenance projects during FY90, including the effect if any, on safe = .
operation of the reactor are described in this section.
M4ar maintenance items were continued in FY90 to support the requirements of the 3 dose reduction projects for light water reactors. These projects are the Pressurized Water:
Reactor Coolant Chemistry I. mop (PCCL), the Boiling Water Reactor Coolant Chemistry .
loop (BCCL), and the Irradiation Assisted Stress Corrosion Cracking (IASCC) Loop.
Two shield plugs were constructed and installed in the upper shield access ring to provic e penetrations for these experimental facilities. The penetrations were designed with.
serpentine benth so as to minimize radiation streamin 5. Dey are used for coolants, gases, L ' electrical cables, and thermocouples. An equipment p atform for the IASCC (in addition to l the existing PCCL platform) was setup on the reactor floor for data loggers, a data.
1 acquisition computer, and control equipment. Equipment racks were also installed on the l far side of the reactor top for components of the IASCC experiment. Addidonally, i supporting equipment and ccwonents of the PCCL underwent modification near the end - -
of FY90 in anucipation of the installation of the BCCL experiment in early FY91.
= In parallel with the above preparations and modifications, floor space available to experiments on the reactor top, reactor floor, and basement areas was increased by removal .
of oki expwiments, consolidation ofinfrequently used equipment, and shielding blocks that l are no longer in use. De surface of the reactor top and the top of the Blanket Test Facility - ,
(BTF) were strip ped and painted with multiple coats of epoxy paint so as to improve space . '
utilizadon, overa l appearance, and to facilitate decontanunanon, shouki it be needed.
Painting was also performed on sections of the containment building. Sections of -
the exterior steel shell showed rust spots at the base where rain water could accumulate. -
ne original bitumastic coatin,g had dned and cracked at below the grs,de level. Earth at the interface next to the steel shel. was removed to an average depth of two feet. - Additionally, spot checks as deep as five feet below grade were made to ensure coating inte ty on the ,
unexposed section of the shell. All degraded sections were sand-blasted with ve grit (SSPC-SP 10). Two coats of KOP COAT Bitumastic Number 300 M Coal Tar Epoxy-were put on the prepared surface.
l Another problem associated with the containment building was the nesting of-pigeons and starhngs in the covered space where the truck lock outer door operating hoist - ,
is located. A professional company was contracted to clean and install deterrent devices 1 (Nivalite)in the space where nests were found. In addition, screens were cut and fitted to 1
close all openings to the truck lock hoist area. His action has been effective in keeping the I pigeons and startings from retuming to their original nesting place.
- i A$acent to the truck lock, there are two for:y-foot riser pipes for the cooling towers l that required rhment. Rese were eight inch diameter schedule-forty steel pipes. They?
had been in use for the past thirty years and had been weakened by corrosion both above and below the grade level, ne portions of pipes below grade were replaced with the same '
schedule steel pipes. He exposed surface was first painted with bitumasdc coal tar for -
protection agenst corrosion and then wrayped in a closed-cell polyurethane foam for.
thermal insulation. He fcrty foot sections a x ve grade were replaced with a higher weight schedule eighty PVC pipes which are rated for an operating pressure of 200 psig at 73 F.
The exposed PVC pipes were painted with a black satin paint so as to protect against .
ultraviolet radiation danage to the polymer. '
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. . . . . - . - - - - - , - , . , , , , , . . - . - - . . . - . , . - - - - - . . - - ~
During the pipe removal and installation pmcess, both cooling tower pump sheds which house t w booster pumps and valving at the riser pipes were removed to allow access -
to the underground pipes. Two new pump sheds with upgraded design were constructed on reenforced concrete foundation pads with floor drains to large drywells. The sheds are constructed with pressure treated wood on a large frame for mam support. : There are built-
, in removable side panels to allow access for maintenance and improvement of air flow to' the tower during severe hot weather operation. Additionally, new electrical services were -
installed at each shed with provisions for future expansion.
= While performing the 1990 Annual Containment Pressure Test, one of the exterior I vacuum breakers was found to be leaking during the inidal While this leakage was within acceptable levels,it the didneed indica
- pressurization for the !test.
for maintenance.
Both exterior breakers were isolated and rebuilt. Data collected during the subsequent l pressure test showed leak rates well whhin the acceptance criteria of the test. - ! -
The reactor basemer.t personnel lock inner door gasket was alad replaced immediately prior to the 1990 Annual Containment Pressure Test. -The gasket had ,
a small hole which allowed compressed air to leak from the seal to the inside of E developk.
the air.oc Normal seal pressure was maintained and any buildup inside the airl '
- bled into the reactor containment as designed. - Containment integrity was never .j cowevudsed. A new gasket was ordered from the original manufacturer and installed. i The personnel lock was retumed to normal service after sadsfactory completion of the 1990 y Annual Containment Pressure Test.
L The operating mechanism of the thermal column steel door failed. This door allows thermali=d neutrons originadng from the reactor core to enter the hohlraum and Blanket j '
Test Facility (BTF). A shear pm in one of the gear operators had broken. : Entry to the hohlraum was gained throuJh the BTF with the reactor shutdown. The shear pin was replaced, and the door was acjusted to hang parallel to the floor and tested satisfactorily.
1 Two proximity switch tubes in the core tank developed leaks and were replaced. l The proximity switches indicate the full in and 80% insertion positions of the shim blade absorbers by responding to permanent magnets which are mounted on the connecting rods-of the shim blades. 'Ihese eight foot long 5/8-inch OD x 0.035-thick 6061-T6 aluminum- '
tubes were replaced with 5/8-mch OD x 0.062 thick tubes so as to prolong the useful life of these tubes The shaft seal on one of the primary coolant pumps, MM-1A, developed a small leak and was replaced. The pulicy system of the graphite gasholder developed worn spots and caused spurious alarms. The pulley system was resurfaced and. lubricated. All ,
diaphragms in the diaphragm valves of the make-up water system were replaced as a preventive measure. , Fina ly, as part of a general MIT program, all PCB-containing transformers associated with the reactor were replaced.
Many' other routine maintenance and preventive maintenance items were performed throughout the year, l
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.. - .13 t E. SECHON 50.59 CHANGES. 'IESTS. AND EXPERIMENTS -
1 1
This section contains a descripdon of each change to the facility w procedures and of the conduct of tests and experiments carried out under the condidoas of Seedon 50.59 of ;
10 CPR 50, together with a summary of the safety evaluation in each case. - -
1 The review and approval of changes in the facility and in the s as described in the SAR are documented in the MITR records by means of " Review l Forms". These have been paraphrased for this report and are Mentified on the '3- u" ages for ready reference if further information should be required with reprd to any item.
kOm pages in the SAR have been or are being revised to reGoct these c unges, and they either have or will be forwarded to the Director, Standardizadon and Non Power Reactor -
Project Directorate, OfHce of Nuclear Reactor Reguladon, USNRC.
The conduct of tests and experu' nents on the reactor ar normally documented in the :
ex . ts and irradiation files. For experiments carried out uncler the provisions of 10 50.59, the review and approval is documented by means of the Safety Review Form..
- All other expen'ments have been done in accordance with the descriptions provided in -
Seedon 10 of the SAR, " Experimental Facilities".-
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-t Pr. .. -i.~t er ainne enemeinn i nnn @cen l SR #0-86 9 (04/21/88),0 88-4 (07/28/88),0 88-5 (09M9/88),0 88-14 (12/07/88),0 89-2 l (014 6/89), 0 89 3 (01/19/89), 0-89-6 (01/24/89), 0-89-9 (06 02/89), 0 89-14 (06/19/89),- )
0-46 (03/2W90),0 B7 (03/2W90),0-48 (03/2@90),0-B9 (03/2W90). :
4 L This project involves the design, installadon, and operation of a pressurized light : !
water loop in the MITR core for the purpose of studying the production, activation, and 1 transport of corrosion to determine the optimum method for reducing the creation of i activated corrosion products (crud) and thereby reducing radiadon fields associated with !
ressurized water reactors (PWRs). 'Ihe ultimate goal is to reduce radiation exposures to
- MR maintenance personnel. <
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Approval for the PCCL was given by the MITR Staff and the MIT Reactor i Safeguards Committee on 04/2W88. It was determined at that time that no unreviewed ,
safety question existed because no failure or accident ansacia*M with the PCCL could lead to an accident or failure involving reactor components. Details of that determination, i together with safety review #0 86 9, were submitted to the U.S.' Nuclear Regulatory i Commission on 04/21/88. j Subsec uent to the determinadon that no unreviewed safety specific procec ures for PCCL operation were prepared. -These included: question exisi 3
Procedure for Ex-Core Testing ,
Supplement to the Safety Evaluation Report l PreoperadonalTest Procedure? ;
Abnormal Operating Procedures for the PCCL >
Procedures for PCCL Startup/ Shutdown Procedures for PCCL Installation / Removal Experiments using the PCCL began in April 1989 and have been quite successful. During the riod covered by this report, several changes were made to the ex core portion of the j
These were:
l !
l Addition of a heated autoclave to serve as an auxiliary pressurizer (SR #0-%6)
L -
Installation of uninterruptible power supplies for the PCCL main circulating pump andinstrumentation. (SR#0B7) :
Replacement of the main circulating pump's variable frequency power supply !
with one that would restart automatically following -a line voltage dip.
l (SR #0 90-8) ,
i- -
Replacement of compression seals with either O rings or welds. (SR #0-90-9) g None of the.e changes involved an umeviewed safety question. All served to increase the reliability of the experinent. >
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Railine Ennlant Onrrrwinn I nnn (BCCD
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SR #0 89-14 (Oti/19/89),0-89-20 (12/2W89)
.. This n
- of a boiling light-water o
loop in the hR U_ojoct core for theinvolves purpoac ofthe studying design,' the la=*=11=%. andand tunsport
'on, activation, .
l of corrosion products. The effect of various water chemistries will be examined to-L determine the um method for reducing the creadon of activated corrosion products ,
(crud) and reducing radiariant fields associated with boiling water mactors (BWRs).- 1
'- The ultimate g is to reduce radiation exposures to BWR maintenance personnel. U In 1988 and 1989, the Reactor Staff made a determination that boiling within an in-core facility is not w--/ to the technical specifications provMed that reacuvity limits for .
movable experiments are not exceeded. It was also concluded that boiling in the experiment volume would not significantly affect reactor operation.- ly, a s carefully controlled experiment was proposed to demonstrate that boiling within an core facility would not adversely affect reactor operadon. Following both a determination thatE -
no unreviewed safety question was involved and approval by the MIT Reactor Safeguards ,
Committee, this expenment was conducted. The results were as expected.
The final safety evaluation report for the BCCL was completed on 8 March 1989 i and appoved by the MITR Staff. On 12/2W89, the MIT Reactor Safeguards Committee .
, deternuned that there was no unreviewed safety question involved in the conduct of the BCCL experiment and approved the BCCL SER. On 9 March 1990, a copy of the BCCL.
SER together vAth the safety analysis prepared by the MITR Staff were forwarded to the ~
U.S. Nuclear Regulatory Commission pursuant to 10 CFR 50.59(b)(2), ,
Experiments using the BCCL are scheduled to begin in September 1990.
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I SR #089-4 (01/23/89),0 89-8 (03/01/89) . j l . ,= .. l l In coq the U.S. Depe) nmentunction of En y,with MIT the Tufts or theMedical New En@ingCenter and with'the an epithermal support neutron beam of 1 is design
- treatment of brain cancer lastoma). Dermal seams have been used successfully for .
this treatment in Japan ~ reason for designing an epithermal beam is to allow tumor 'I treatment without having to sub' the patient to surgery involving mmoval of a portion of , i the skull. . Also, an epithermal gives greater penetration. Thus far, the research has i consisted of simulanon studies using Monte Carlo codes and experiraents using the MIT Reactor's medical beam to verify those studies. Two facility changes were previously.
made. These were:
(1)- Installation of a liner and a support plate in the medical therapy beam.
Dese were installed to permi,t the subsequent installados of candid
- filter '
materials for producing an epithermal beam. ~ (SR #0-89-4)
-(2) Installation of a candidate filter. Currently, filters contain135 sulphur and l aluminum with small quantities of lithium and cadmium ar, pear to give the best results. One such filter was installed on a trial basM icr the purpose of : -
confirming the results of the mimniation studies. (SR #0 89-8)-
Neither of these design changes was judged to involve an unseviewed safety question.
a Filter design studies were continued during the present reporting period and DOE -
support for the project was renewed.' Animal studies usmg unce were begun in June 1990 : i and the preliminary results show that the mice treated by neutron capture therapy are outliving the controls (mice with tumors, conventional treatment). This suggests that the therapy may be effective. However, definitive results will not be available for at least ,
several months, i
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Diaital Onmnuter Onntml of Demetort UnAar Standy. State and Tranelant Panditinna 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 (0343/82), E 82 5 (04/14/82), E 82-6 (07/13/82), 0 83 5 (02/03/83), E-831L (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 (12/06/84), 0-84 21'(11/01/84), 0-85-11 (05/09/85). 0-85 13 (06/28/85),-0 85-16 (07/12/85), 0-85-20 (08/16/85),. 0-85-25 (12/01/85), 0-85 26 (12/01/85), 0-86-11 (10/17/86), 0 86-13 (11/28/86), 0 87-11 (06/01/87), 0 87 17 .
(12/24/87), 0-88-10 (12/01/88),
i
. The project involving computer analysis, signal validation of data from reactor ' ,
instruments, and closed loop control of the MIT Reactor by digital computer was- !
continued. A non-linear supervisory algorithm has been developed and demonstrated. It functions by restricdag the net reactivity so that the reactor infinite by reversing the direction of control rod motion. period
- It, combined can with be rapidly made 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 closed-loop digital control and signal validation methods to ,
demonstrate the unprovements that can be achbved in reactor control. . .,
Each new step in the program is evaluated for safety in accordance with standard review procedures (Safety Review numbers listed above) and approved as nac- y by the -
MIT Reactor Safeguards Canmittee.
Initial tests of this digital closea-loop controller were enad=*d in 1983-1984 using.
- the facility's regulating rod which was of relatively low reactivity worth (0.2% AK/K).
Following the successful completion of these tests, facility operating license amendment No. 24 was obtained from NRC (April 2,1985)c It permits:
(1) closed loop control of one or more shim blades and/or the regulating rod provided that no more than- 1.8% AK/K could be inserted were all the
- connected control elements to be withdrawn, (2) closed-loop control of one or more shim blades and/or the regulating rod -
provided that the overall controller is designed so that reactivity is 1 constrained sufficiently to permit control of reactor power within desired or authorized limits.
A successful experimentation program is now continuin g under the provisions of this license amendment. A protocol is observed in which the bRC-licensed supervisory .
controller is used to monitor, and if necessary override, other novel controllers that are still in devalaa==* Tests of novel controllers are conducted under the provisions of technical l specification #6.4 which requires that reactivity be constrained to ensure " feasibility of .I control". . Signal implementation is accomplished using a variable-speed step sing motor.
This motor is installed prior to the tests and removed upon their comp;etion. An indegndent hard-wired circuit is used to monitor motor speed and preclude an overs conc ition. This arrangement for the conduct of these Msts has been approved by the J Reactor Safeguards Committee.
Research on digital control was ham?cred during the present reporting period by the ,
need to make the reactor available for the cose reduction studies and the medical therapy I program. Experiments to evaluate the design of robust control laws (laws that are l
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)p hin August 1990. Also,~ comparadve studies of:
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19 l'5 :
Revinian of Fiamiaa twseity Limie i il SR #0 88-12 (1101/88) !
The fisdon density limit for the UAl fuel s used by the MIT Research Reactor is l l.8 1021 flasions/cc. Rc warch conducted by the Idaho National Engineering Laboratory i (Nucl. Tach.. 49,136149, June 1980) shows that a limit of 2.3 1021 fissions /cc is !
technically justified. Analysis of the MITR fuel cycle showed that increasing the MITR 1 fission density limit to 2.3 1021 fission /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 Safeguards Committee, submitted to the U.S. Nuc1 car -
l Regulatory Commission (F.C) op 13 February 1989. On 27 November 1989, NRC mquested additional information. 'that material was forwarded on 6 July 1990. Approyu. - l of this proposed change is still pending. :
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Te.c..-. ras la===* in U-2M Thee*= elan 13mit SR 488-13 (12/20/88)
De DOE 4wned shipping casks that MIT had been planning to use were not licensed by NRC. According y. preparations are now being made for use of a commercial cau. In the interim,it was to obtain an increase in the U 235 possession limit.
A safety analysis was performed following al by the MIT Reactor Safeguards Committee submitted to the U.S. Nuclear Regu Commission on 14 February 1989.
A request for further information was received on 1 April 1989. His was answered on 24 August 1989. Approval of the request by NRC was given on 11 December 1989. '
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SR 80-89-16 (08/18/89)
A differendal (HTM iTO) tritium air samplir g system was set up in parallel with the existing sampling devices, for the purpose of making comparative analyses. No changes were made to any exist'Ag reactor systems. However, a safety analysis was required because the enerimemers would be working with tridum and H2/N2 carrier ps mixtures.
(Hols: The H2 content of these carrier ras was less than 1%.) No unrenewed safety questions were identified. A report ositte results of the comparative analyses will be available in Septanber 1990.
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Mht of Annenline on FWie Cea..v.=a IndimeM in Imw and Mwlium Fast Flum SR #0 89-18 (09/D/89)
This series of experiments involves the study of radiadon damage effects that ml@t be incurred in electmnic components operadnj in space. Da nase associated with sudc en bursts of radiation has been extensively studiec . However, therchas been litde research on l
- the efforts of low level, steady state r=<tiatie=i fields such as might be encountered on manned voyages to Mars. These experiments am being conducted in the reactor's fast flux facility. No unreviewed safety questions were identired.
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23-F. ENVIRONMENTAL SURVEYS
! Environmental surveys, outside the facility, were conducted using radiation munitors and analmatry devices. De monitors located within a quarter mile radius of the facility consist of GM-based detectors with associated electronics. The dosimetry devices are film badges. During FY90, makr construction at one of the sites (west),9recluded the collection of data from that site's radiation monitoring system. The fiscal y,w total for 1990 vs 1989 indicates a decrease by a factor of 3.3. His decrease is not um .asonable because Ar 41 releases for the year also demonstrated a decrease from the previous year by
, a facts of 2.8. De detectable radiadon levels per sector due to Ar-41 are presented belcw.
The quarterly monitors (film badges) indicated no detectable radiation for the period in quesdon. '
33 Exposum (07K)1M19-0MMV90) ;
North 0.07 mR/ year East 0.02 mR/ year South 0.12 mR/ year i
Wes:
- 0.10 mR/ year Green (east) 0.06 mR/ year FiscalYearTQiah:
1990 0.3 mR 1989 1.1 mR t 1988 1.2 mR t :
I 1987 1.2 mR 1986 1.8 mR l
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1
G. RADIAIIONIXPOSURE3 AND SURVEYS WITHIN ' DIE FACILITY A namunary of radiadon exposures roosived by facility penonnel and experimenters is g;ven below:
hlrl19R-lunedL1990 Whole BodyExnopure Rany (Rems) Number of Personnel No measurable ....................................................................103 Measurable - < 0.1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 59 0.1 - 0.25 ..................................................................... 9 0.25 - 0.5 .................................................................... 9 0.5 - 0.75 ..................................................................... 3 0.75 - 0 ..................................................................... 7 i
TotalMan Rem = 9.87 TotaManonnel= 185 From July 1,1989 through June 30,1990, the Reactor Radiation Protection Office -
provided radiadon protection services for the facility which included power and non-power operational surveillance (performed on daily, weekly, monthly, uarterly, and other frequencies as required), maintenance activities, and experimental support. Specific examples of these activities include, but are not limited to, the foll g:
- 1. Collection and analysis of air samples taken within the containment building and in the exhaust / ventilation systems.
- 2. Collectw.. and analysis of water samples taken from the cooling towers, D2O system, primary system, shield coolant system, heat exchangers, fuel storage facility, waste storage tanks, and expenmental systems.
- 3. Performance of radiation and contamination surveys, radioactive waste collection, calibration of area radiation monitors, calibration of effluent !
monitors, calibration of radiation survey instruments, and establishing /'
posting radiological control areas.
- 4. Provide radiation protection services during fuel movements,'in core experiments, sample irr diations, beam port use, ion column removal, etc.
The results of all surveys and surveillance conducted have been within the guidelines established far the facility.
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H. RADIOACTIVE EFFI1 TEN'I3 his section summarises the nature and amount of liquid, gaseous, and solid radioactive wastes released or discharged from the facility.
- 1. Liquid Wane Liquid radiometive wastes generated at the facility are discharged only to the sanitary sewer serving the faellity. There were two sources of such wastes during the year: the cooling tower blowdown and the liquid waste storage tanks. All of the uid volumes are measured, by far the largest being the 4,293,900 liters discharged durin FY90 from the cooling towers. (Lar quantides of non-radioactive waste water are scharged to the sanitary sewer system other parts of MIT, but no credit for such diludon is taken since the volumeis not ly measured.)
All releases were in accordance with Technical Specification 3.81, including Part 20, Title 20, Code of Federal Reguladons. All actMdes were substandally below the limits specified in 10 CFR 20.303, but the monthly tritium releases are reported in Table F-3 in accordance with Technical Specification 3.8-1 because its concentration emadad 3x104pCL/ml.
- 2. Gaseous Waste Oaseous radioactivity is discharged to the atmosphee from the containment building exhaust stack and by evat . Son from the cooling tc wers. All gaseous releases likewise were in accordance with the Technical Specinceeg.d and Part 20, and all nuclides -
were below the limits of 10 CFR 20.106 after the authorized tiludon factor of 3000. Also, all were substudally below the limits of 10 CFR 20, A9pendix B. Note 5, with the excepdon of Ar-41, which is reported in the following Table H 1. De 542.6 Ci of Ar 41 were released at an average concentration of 0.14 x10 8 Ci/ml for the year.- This represents 3.5% of MPC (4x104 pCi/ml) and is significantly less than the previous year's release of 1529 Ci. De decrease is due to continued success of our program to idendfy and eliminate sources of Ar-41.
- 3. Solid Wa=ta Only one shipment of solid waste was made during the year, information on which is provided in the following Table H 2, 4,
linuid Discharse to the Ranitmev Sewarnee Svetem Total gross beta acdvity in the liquid effluents (cooling tower blowdown, waste storage tank discharges, and engineering lab sink discharges) amounted to 0.022 Ci for FY90. De total tritium was 0.059 Ci. The total effluent water volume was 4,293,900 liters, giving an average tritium concentration of 16.2x104 pCi/ml.
The above liquid waste discharges are provided on a monthly basis in tne following Table H 3.
1 ,
. . _ _ _ . - _ _ . . ~ . _ . _ . . . . _ . . _ _ . _ _ _ _ _ _ .. _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . . _ _ _ . . _ _ . _ . _ _ _ . _ _ ___
e 26 IABLE Hd i
ARGON-41 STACK RELEASES l B1C6L YEAR 1990 1 3
)
I Ar4i Average Discharged ConcentradonW j (Curies) (uCihn!)
July 1989 21.4 0.07 x 10-s August 57.7 0.15 l tapand=r 38.5 0.13 l
! October 42.1 0.14
'l November 92.5 O.32 !
December 18.5- 0.05 l January 1990 41.2 0.14 !
I i February 49.1 0.16 i March 55.1 0.15- !
April 24.8 0.08 l May 63.2 0.21 l
> June 38.5 0.10 i i
Totals (12 Months) 542.6 0.14 x 10-s MPC (Table II, Column I) 4 x 104
% MPC l
3.5%
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27 TABIE R2
SUMMARY
OF MrrR.Il RADIOACTIVE SOIin WAS*IE SHIPMENTS BSCAL YEAR 1990 Description Units Shipment #1 Total l Volutne Cubic Feet 192.6 192.6 Weight -
Pounds 6232 6232 Activity (l) Curies 0.035 0.035 Date of shipment June 15,1990 Disposition tolicensee for burial U.S. Ecology, Inc. f (1) Radioactive waste includes: dry active waste comprised of irradiated components, contaminated items, and solidified wastes. The principal radionuclides are acdvation products such as800, SICr,65Zn,12sSb, and etc.
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TABLE H-3 LIOUID WASTE DISCHARGES FISCAL YEAR 1990 f Total Volume Avenge' [
Gross Beta Total ofEffluent Tridum -
less Tridum Tridum WaWI) Concentration .
, t (x104 Ci) (t103 Ci) (x108 liters) (x104 pCi/ml) ~.
l July 1989 NDAG) '1.25 11.4 11.0 Aug. 207.8 10.74 80.95 13.2 ,
Sept. NDAG) 5.98 43.6 13.7 Oct. NDAG) 1.41 36.6 3.85-Nov. 3153.3 11,79 68.97 17.14' Dec. NDAG) 1.95 10.3 'l 8.9 [
l l Jan.1990 NDAG) 0.93 11.2 8.3 Feb. NDAG) 0.76 41.3 1.84 Mar. 17,500 12.39 33.03 37.15 l
Apr. 1090 "
4.65 8.79 52.9 May NDAG) 0.26 41.8 0.62 June 50.6 6.40 41.45 15.46 ;
12 months 22,001.7 58.51 429.39 16.17--
U) Volume of effluent from cooling towers, waste tanks, and NW12139 Engineering lab sink. Does not include other diluent from MIT estimated at 2.7 million ga lons/ day. ,
G) No Detectable Activity; less than 1.26x104 pCi/ml beta for each sample. ;
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