ML20216H657

Operating Rept for Univ of Ma Lowell Reactor for Period 960701-970630

ML20216H657
Person / Time
Site:	University of Lowell
Issue date:	06/30/1997
From:	Bettenhausen L MASSACHUSETTS, UNIV. OF, LOWELL, MA (FORMERLY LOWELL
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9709170011
Download: ML20216H657 (22)

Text

h tlttiverstttj of Nssocitt4setts Loweff Rndiation Laborntortj liittiversittj Avetitle LowcR, tinssnchtssetts, 01854 S08 934-336S 9 September 1997 Document Control Desk U. S. Nuclear Regulatory Commission Washington, D.C. 20555 License No. R '.25 Docket No. 50-223 UNIVERSITY OF h1ASSACilUSE1TS LOWELL RESEARCil REAC1T)R Sirs:

The at: ached report is submitted as required by Section 6.6 of the Technical Specifications for the license cited above.

Sincerely yours, c'~$>4'%DDr$w-Lee liiettenhausen Reactor Supervisor /

cc: Regional Administrator, Region i T. Dragoun, Region i T. S. hilchaels, Senior Project hianager Reactor Safety Subcommittee hiembers /,p 9709170011 970630 PDR ADOCK 05000223 R PDR k!\,kkk ..

ce m

OPEltATING 11EPOltT IO R Tile UNIVEllSITY OF MASS. LOWELL llEACTOlt IOR Tile PERIOD JULY 1,1996 TO JUNE 30,1997 t

Docket No. 50 223 License No. R-125

^

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,' [ CON 11NIS A. Introduction B. Function C Operating Experience

1. lixperlinents and Facility Use

2. Changes in Facility Design

3. Performance Characteristics

4. Changes in Operating Procedures Related to Reactor Safety

5. Results of Surveillance Tests and inspections

6. --Staff Changes

7. Operations Sununary D_ .

Energy Generated 11 Inadvertent- and - Emergency Shutdowns F. Major -Maintenance G Facility Changes Related to 10 CFR 50.59

-Il Environmental Surveys

1. Radiation Exposures and Facility. Surveys

1. Personnel Exposures

2. Radiation Surveys

3. Contamination Surveys -

J. Nature and Amount of Radioactive- Effluents

1. Liquid Wastes

2. Gaseous Wastes

3. Solid Wastes OP97-2

.' A. INTRODUC110N -,

In the late 1950's the decision was made to build a Nuclear Center at what was- then Lowell Technological Institute. -Its stated aim was to train and educate nuclear scientists, engineers and technicians, to serve as a multi disciplinary research center for LTI and all New England academic institutes, to serve the Massachusetts business community, and to lead the way _in the economic revitalization of the Merrimack Valley. The decision was taken to supply a nuclear reactor and a Van-de Graaff accelerator as the initial basic equipment.

Construction of the Center was started in the :,ununer of 1966.

Classrooms, offices, and the Van de Graaff accelerator were in use by 1970. Reactor license R 125 was issued by the Atomic Energy Commission on December 24, 1974, and initial criticality was achieved on January 1975.

The name of the Nuclear Center was officially changed to the "Pinanski-Building" in the spring of 1980. The purpose was to reflect the change in emphasis of work at the center from strictly nuclear studies. At that time, the University of 1.owell Reactor became part of a newly established Radiation Laboratory. The Laboratory occupies the first floor of the Pinanski lluilding and performs or coordinates research and educational studies in the fields of physics, radiological sciences, and nuclear engineering. The remaining two floors of the Pinanski 13ullding are presently occupied by various other University departments.

On February 14, 1985, the University of Lowell submitted an application to the Nuclear Regulatory Commission for renewal of the facility operating license R-125 for a period of 30 years. On November 21, 1985, the license renewal was granted as Amendment No,9 of License R.

125 in accordance with the Atomic Energy Act of 1954.

OP97-3

11. FUNCTION The Radiation Laboratory is a major research focal point of the University, h1 ore than 200 graduate students have used or are using the Laboratory's services; the comparable number for the faculty is in excess of 25. The University departments utilizing the facility include Illology, Chemistry, Earth Sciences, Physics, hiechanical Engineering, Plastics Engineering, Radiological Science and Chemical / Nuclear Engineering. The University's Amherst campus and hiedical Center have active research programs at the Radiation Laboratory. hiuch research is concerned with safety and efficiency in the nuclear and radiation industries, including pharmaceuticals, medical applications, health effects, public utilities, etc.;

however, much research is also done by workers in other fields who use the unique facilities as analytical tools, in addition, the Laboratory's facilities a e .ased in the course work of various departments of the University, it also provides these services to other campuses of the hiassachusetts system, other universities in the New England area, government agencies and, to a limited extent, industrial organizations in hiassachusetts and the New England area, as well as numerous school science programs in the hierrimack Valley.

C. OpERAT!NG EXPERIENCE

1. Experiments and Facility IJse The major uses of the reactor during this fiscal year were activation analysis, dosimetry studies, calibrations, specialized isotope production, neutron effects studies, teaching and personnel training.

Activation techniques were used to study geologic composition of rock samples and constituents of forest soils. The evaluation of the neutron to the gamma ratio and detailed neutron spectral mapping for in-core experiments is continuing.

Dosimetry studies and calibrations utilized N-16 production for high energy gamma fields and reactor facilities for mixed neutron and gamma dosimetry, OP97 4

I l

Isotopes were produced for calibration standards, medical research use, and lab practicums.

Reactor operating time used for teaching purposes included a reactor operations course emphasizing control rod calibrations, critical approaches, period measurement, prompt drops and calorimetric measurement of power and preparation of students and staff members for NRC licensing examinations. Freshman laboratories for reactor principles and activation analysis were conducted for chemical / nuclear engineering students.

Radiological science students utilized the facility for performance of

, radiation and contamination surveys. Senior students participated in a laboratory that required locating and identifying an unknown isotope of low activity in a mockup power plant environment. The isotope was provided for the students in an isolated area in the reactor pump room during non-operating hours. During the practicum, the students were supervised by faculty and staff. The reactor served as a source of neutron and gamma radiation for various radiological science and biology laboratories.

A number of activation and decay experiments were performed for both university and non-university students alike. For the seventh consecutive year, activation and decay experiments were provided for local school science classes involving more than 2,000 students who observed the experiment at the reactor or in their classrooms via interactive cabic T.V.

The major outside uses for the reactor facility is neutron and gamma damage studies of electronic components, characterization of neutron detectors, and neutron effects upon materials.

2. Changes in Facility Design None made. Changes are pending to nuclear instrumentation and radiation monitoring systems.

3. Performance Characteristicji Overall, the performance of the reactor and associated systems has been normal over the past year, except for the ventilation valves discussed in Section F, Maintenance.

4. Changes in Onerating Procedures Related to Reactor Safety OP97-5

We still await the Department of Energy to provide a firm schedule for fuel element fabrication, Submittals m all requests and answers to questions were completed in July,1997 aad an NRC order to effect the change to LEU fuel was issued on July 31, 1997. Changes to operating procedures will then be needed to implee'nt the new fuel use.

Provisional changes to operating pu eJures have been made and approved for new nuclear instrumentation. They will be implemented as the equipment is installed. The same process will be used for new radiation monitoring equipment,

5. Results of Surveillance Test and Inspeelinns All Technical Specification Surveillances required during the fiscal year were performed in a timely manner. The results of each requirement have been reviewed by the Reactor Supervisor and Chief Reactor Operator.

Almost all surveillance test results were found to be within specified limits and surveillance inspections revealed no abnormalities which would jeopardize the safe operation of the reactor, One as-found surveillance test showed that one rod drop time was slightly greater than specification; investigation found an oil film on the magnets which delayed breakaway by about 100 milliseconds. The magnets and rod heads were cleaned and ,

drop times were reduced to we!! within specification. Each required calibration was also performed.

A five-year containment integrated leak rate test was performed satisfactorily on May 28-30, 1997. The test report is attached,

6. Staff Changes As of June 30, 1997 the operations staff consists of three Trainees, two part time student Reactor Operators, one part time student Senior Reactor Operator, a former student who maintains a Reactor Operator license, a staff Health Physics Technician / Reactor Operator, and three staff Senior Operators, including the Chief Reactor Operator and the Reactor Supervisor, OP97-6

7. Operations Summary -

During the course of the fiscal year 1996-1997 the reactor was critical a total of 546.27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br />. The utilization is broken down as follows:

Operating flours Critical hours 546.27-llours at full power 439.49 Megawatt hours 458.42 Experimental Utilization-Sample hours 1115 (includes multiple samples)

Number - of irradiations 511 Number of training hours 141 D. ENERGY GENERATED Total energy generated (MWD) 19.1 Number of hours reactor was critical 546.27 Total cumulative energy output (MWD) 228.15 E. INADVERTENT AND EMERGENCY SilUTDOWNS

- There were 14 inadvertent scrams. Six of these scrams were due ta aging instrumentation - which is being replaced. The eight scrams not related to instruments were operator errors in upranging and downrr aged picoammeters.

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F. MAJOR MAINTENANCE Another series of problems occurred with ventilation valves in this reporting period. The facility exhaust valve, Valve E, and he emergency exhaust valve, Valve D, both failed to close. Disassembly revealed that the threads on the piston head failed, apparently from lack of full engagement.

A thread repair was made, the valve seals rebuilt, the quick-release rebuilt, air leaks eliminated, and the valve and containment returned to service after closure time testing. Since this problem with thread failure has been recurrent, the manufacturer was consulted. New enerator pistons are now on order; the new piston head threads are twice the diameter and twice the length of engagement compared to the old pistons.

G, FACILITY CllANGES RELATED TO 10 CFR 50.59 There have been no facility changes to date which pose an unreviewed safety question. A review of the changes for new nuclear instruments was made in accordance with 10CFR50.59 l1. ENVIRONMENTAL SURVEYS Surveys of the environs external to the reactor building have continued to show no increase in levels or concentrations of radioactivity as a result of reactor operations. Air particulate samples collected at a continuously monitored site on the roof of the Pir.anski building have shown no reactor produced radioactivity. Thermoluminescent dosimeters are used to monitor unrestricted areas outside of the Reactor. The results of these measurements show that doses in these areas were indistinguishable from background radiation levels during the period of July 1,1996 to June 30, 1997..

Analysis of water samples collected from the Merrimack River upstream and downstream of the reactor location have continued to yield no radioactivity associated with reactor operations.

I. RADIATION EXPOSURES AND FACILITY SURVEYS OP97-8

l, j'ersonnel Exposures Personnel exposures were maintained at the lowest reasonable levels. Doses received by individbals concerned either directly or indirectly with operation of the reactor were within allowed limits.

Twenty individuals were monitored by film badge during the year. Only four received measurable external deep dose equivalents ranging from 20 to 80 mrem.

2. Radiation Surveys Radiation levels measured in the reactor building have been typically less than 0.1 mrem /hr in general areas. Experiments have been conducted in which transient levels at specific-locations have been in excess of 100 mrem /hr. Doses in these instances have been controlled by use of shielding and/or personnel access control. The pump room remains designated as a high radiation area during reactor operation and access is controlled. Dose equivalent levels in the order of 10 mrem /hr are present adjacent to the closed beam ports during maximum power operation.

3. Contamination Surveys General area contamination has not been a problem in the reactor building. Contamination has occurred at specific locations where samples are handled and particular experiments have been in progress.

Contamination in these areas is controlled by the use of easily replaced plastic-backed absorbent paper on work surfaces, contamination protection for workers, and restricted access.

J. NATURE AND AMOUNT OF RADIOACTIVE WASTES

1. Liquid Wastes

- Liquid wastes are stored for decay of the short lived isotopes and then released to the sanitary sewer in accordance with 20 CFR 2003. A total of 13.6 pCi were released over the 12 rnonth period. The principle isotopes released were Na-24 and corrosion products, i.e. Mn-54, Co-60, OP97-9

2n 65 'and Sb 124,

2. Gaseous Wastes Argon-41 continues to be the only significant reactor produced radioactivity identifiable in the gaseous' effluent. Following are the monthly stack release data for Ar4 1 for the reporting period:

Month Ar-41 Released llours Curies July 1996_ 0.6-August 1996 0.2 September 1996 1.4 October -1996 1.4 November 1996 1.2 December 1996 0.6 January 1997- 1.1 February 1997 4.3 March- 1997 2.0 April 1997 3.4 '

May- 1997 0.4 June 1997- 0.6 Total 17.2 This . release represents a 12 month dose of 0.4 mrem to the nearest member of the public using the EPA Comply code.

3. Solid Wastes Solid wastes, primarily paper, disposable clothing, and gloves, along with other miscellaneous items have been disposed of in appropriate containers. Most of the activity from these-wastes consisted of short lived induced radioactivity. These wastes were held for decay and then released-if no activity remained. The remaining long lived waste (< 10 cubic feet)

OP97-10

was collected and stored in a designated long lived waste storage area hwalting ultimate disposal at llarnwell.

OP97-11

Containment Leak Rate Test of May 28-30,1997:

On May 28,1997 a Containment Leak Test was begun as -described in Procedure S.P.3.

The test was performed in three . sections. The first was the initial pressurization of- the building during which input air Venturi throat data were taken. These data are used to calculate the building volume.

The' second section measured the building pressure decrease over -a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period. For this part of the test, the compressor hose was disconnected (see Fig.1) and the inner (reactor side) doors were closed and the outer doors open. Data were taken- using an internally referenced manometer and an externally referenced manometer and pressure gauge, j The building was. then repressurized.- The air lock doors were equalized, and the_ inner doors opened. The-leak rate test was repeated for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with the outer airlock door seals as pressure barriers. This constituted the third section of the test.

Summary of Results:

The ' leak rate calculated for .a 2 psig overpressure from- testing the inner containment doors at an actual- test pressure .of 0.5_ psig was 2.70 % per day based . upon - internal manometer reference and 2.94 % per day based upon external manometer reference. 'The leak rate calculated for a 2 psig overpressure on .the outer containment doors from an actual test- pressure of 1.0 ps_ig was 2.85 % per day based upon internal and 2.92 % per day based upon external manometer . data- This meets the test acceptance

- croterion of 7.5 e per day and the Technical Specification of 10.% per day.

- Building Volume Calculation The building was . pressurized with a rented compressor.

Pressurization to 0.5 psig took -101 minutes. Using Flieger's formula 0 0 Kent. Mechanical Engineer's Handbook,~ John Wiley & Sons. Inc. 1954 i

---

____m.__

The air flow through the throat is calculated:

w =.53 c pi

' (.225)* x '

(V460 + T ,

w = mass flow in Ib/sec c al P t = inlet. pressure to the throat in psia (psia = psig + 14.7. Psig is the reading off the Venturi's upstream . pressure gauge, as seen in Fig.1)

A = area of the throat in inches 2 (The throat is

.255 inches in radius.)

T = temp of inlet air in F (inlet temp in Rankin, R = 460 + F is accounted for in formula.)

The mass Flow Rate, w, was calculated for each pressure data point.

The flow rate varied from 0.1697 to 0.1762 lbs/sec. The- average was 0.1722 i .0467 lb/sec. The error was calculated using the conservative assumption that the uncertainty in pressure was 6 psi and the uncertainty in temperature was i 7 F.

NOTE: Appendix II of an earlier report lists the equation and solution to calculate -the error associated with flow.

therefore the total mass, wa, may be calculated w a= (0.1722 lb/sec)*(60 sec/ min)*101 min wa = 1043 i 283 lbs The pressurization resulted in a change in pressure of 14.0-inches of water. Therefore, the weight of air in the building at test conditions and building temperature of 68F, wb, may be calculated:

wb /1043 = 411.47/14,0 wb = 3.07 x 104 4.66 x 103 1bs

-NOTE: The weight of air is found by:

L absohite pressure w, intemal pressure absolute pressure = barometric- pressure (as re'ad off the barometer) x 1.133 x 12

.. barometric pressure of- 30.30 =

30.30 x 1,133 x 12 = 411.47 to calculate the error associated with wb see cited Appendix III.

The volume may be calculated:

V = (2.52 x10') 359 ft' / mole y460 + 75i2'R'

( 29 lb / mole A 460 + 32 R s ,

-V = 3.74 x 105 ft3 Building Leak Rate First Building Pressure Test-The building' was pressurized approximately 0.5 pounds (14.0 in water). The compressor was then disconnected from the building and the decrease in pressure- over -a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> was measured. The decreasing building pressure was= measured by both an -internally _

1" Reactor Containment Building Leak Rate Test for March 29-31, 1974, for LTI Reactor", submitted to USAEC on July 3,1974

referenced manometer and an externally referenced manometer (Figure 2), .

An unweighted-linear regression analysis of the form P = A + Bt was performed. The equation- for the internal manometer readings resulting

- from the least- squares fit analysis was P = 425.48 -0.318t (3)

The standard deviation values for this equation were

= 0, = 0.794 ob = 0.022 The correlation coefficient value r^2 = .9009 Using the following formula, the leak rate was calculated e 8 leak rate = Pi - P2 (100) - (4) s Pi s This formulaz was derived -in an earlier leak rate report 2 The test started with a barometric pressure equivalent to 411.47 inches of water and the overpressure was 13.8 inches- at the start of the test for an absolute pressure of 425.27 inches of water. This was used as the starting pressure, pl and p2. were found using the least' squares fit

,_ with the intercept. of 425.48 at t- = 0.

thus, pt = 425.48 0.79 .

p2 = . pl - 0.318(24) p2 = 417.82

. leak rate = (425.48 - 417.82)/425.48 *(100%) =

m -

observed: - internal _ reference

,' '. 1.35 % . building volume- per day For. the externally referenced manometer, p = 425.053 i 1.539- 0.366t i 0.043 with r^2 = 0.76201 and the leak rate calculation is:

leak rate = (425.274 412.018)/425.274 = 1.47% per day 4

observed external reference Second Building Pressure Test

- The -air lock doors were reversed, that is, the outer door closed and the inner door was opened. The method for calcult. ting the leak was the same as the first test. The building was-repressurized to 1 psig. The test was run for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 1.

Linear regression analysis resulted in the equations:

p = 436.163 0.596 - 0.344ti.017 with r^2=0.9496 ,

internal reference an'd p =- 433.520 1.260- 0.360t .035 _w ith r^2=0.8218 -

external . reference The test started with a barometric pressure . equivalent to 410.795

.. inches of water and- the overpressure was 24.4 inches of water for an absolute pressure of 435.195 inches of water externally referenced, where S

pi =: 435.195

. p2 = 426.388 Leakrate = (435.195 - 426.388)/435.195 =

observed ext ref 2.02%- building volume per day.at 1 psig n -.

Leakrate = (433.171 - 424.785)/433.171*100% = 1.949 per day calculated ext ref The results of the second test using the internally referenced manometer were:

Leakrate = (436.895 - 428.195)/436.895*100% = 1.99 % per day observed int ref and Leakrate = (435.819 - 427.563)/435.819*100% = 1.89% per day calculated int ref Calculation of Leak Rate Adiusted to 2 PSI overnressure We can compare these results with a leak rate at 2 psi overpresuure by the method prescribed in 10CFR50, appen. J., III 4 iii.

I r p 'i Lg = L, J (P, 3 where Ls = leak rate at specified (2 psi) overpressure La = leak rate at reduced overpressure Pa = internal overpressure for the reduced pressure test (in. H2 O)

Ps = internal overpressure for the 2 psi case (in, of H2 0)

Rearranging,

'pd N Ls = L, (Paj 1st Test (Inner door closed) (olnitial = 0.5 osle)

Ls = 2.70 (cale int reO: 2.94 (cale ext reO % per day

I 2nd Test (Outer door closed) (ninitial = 1.0 nsig)

Ls = 2.85 (cale int reO: 2.92 {cale ext reO % per day Thus, the Technical Specification of 10% per day and the test acceptance criterion of 7.5 % per day were met, The temperatures within the reactor. containment building remanied between 66F and 71F throughout the test. The maximum diurnal difference was 4F at any single monitored location.

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CONTEKIS

d. Introduction B. Function C. Operating Experience

1. Experiments- and Facility Use

2. Changes in Facility Design

3. Performance Characteristics

4. Changes in Operating Procedures Related to Reactor Safety

5. Results of Surveillance Tests and Inspections .,

6. Staff Changes

7. Operations Summary D. Energy Generated E Inadvertent and Emergency Shutdowns F. Major Maintenance G Facility Changes Related to 10 CFR 50.59

11. Environmental Surveys I. Radiation lixposures and Facility Surveys

1. Personnel lix posures

~2. Radiation Surveys

3. . Contamination Surveys' J. Nature and Amount of Radioactive Effluats

1. Liquid Wastes

2. Gasecus Wastes

3. Solid Wastes OP97-2

, _ _ _ _ _ _ _ _ _ _ _ - _ _ .