Text

Annual Report: July 1, 2013 for the Maryland University Training Reactor, Docket No. 05000166, License No. R-70 (TAC ME1592), University of Maryland
	ML13303B562
Person / Time
Site:	University of Maryland
Issue date:	10/22/2013
From:	Al-Sheikhly M; Univ of Maryland - College Park
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	TAC ME1592
	Download: ML13303B562 (62)
	v • d • e

UNIVERSITY OF Building 090 MARYLAND DEPARTMENT OF MATERIALS SCIENCE AND ENGINEERING College Park, Maryland 20742-2115 301.405.5207 TEL 301.314.2029 FAX October 22, 2013 Document Control Desk United States Nuclear Regulatory Commission Washington, D.C. 20555-0001

Reference:

Annual Report: July 1, 2012 - June 30, 2013 for the Maryland University Training Reactor, Docket No. 50-166, License No. R-70 (TAC NO. ME 1592), University of Maryland Enclosed please find the University of Maryland's Annual Report for the period beginning July 1, 2012 and ending June 30, 2013 for the Maryland University Training Reactor, Docket No. 50-166, License No. R-70.

If there are questions about the information submitted, please write to me at:

Department of Materials Science and Engineering, University of Maryland, College Park, MD 20742-2115 or email me at mohamad@umd.edu. I would appreciate it if you would copy Prof. Robert Briber on any such correspondence: Department of Materials Science and Engineering, University of Maryland, College Park, MD 20742-2115; rbriber@umd.edu.

I declare under penalty of perjury that the foregoing response is true and correct.

Sincerely, Mohamad Al-Sheikhly Professor and Director Maryland University Training Reactor Enclosure cc: Robert Briber

$~c~2Th

ANNUAL REPORT: July 1, 2012- June 30, 2013 FOR THE MARYLAND UNIVERSITY TRAINING REACTOR License No. R-70 Docket No. 50-166 Department of Materials and Nuclear Engineering A. James Clark School of Engineering University of Maryland, College Park College Park, MD 20742-2115

TABLE OF CONTENTS TABLE OF CONTENTS ..................................................................................................... 1 I. INTRO DU C TION ............................................................................................................... 2 II. REACTOR USAGE ........................................................................................................ 3 III. SURVEILLANCE TESTS AND INSPECTIONS ........................................................ 4 IV. CHANGES TO THE FACILITY .................................................................................... 6 V. ENVIRONMENTAL SURVEYS OF SURROUNDING AREAS ................................ 7 VI. RADIOACTIVE RELEASE AND DISCHARGE TO THE ENVIRONMENT ............ 8 VII. ALARA REVIEW FOR FACLITY PERSONNEL AND VISITOR EXPOSURE ..... 9 VIII. UNSCHEDULED REACTOR SHUTDOWN/REPORTABLE OCCURENCES ........ 10 IX. SPECIAL EXPERIMENTS ............................................................................................... 11

x. CHANGES IN FACLITY STAFF ............................................................................... 12 A. APPENDIX A. EPA COMPLIANCE .......................................................................... 13 2

I. INTRODUCTION The University of Maryland Training Reactor (MUTR) is an open-pool type, TRIGA fueled reactor licensed for operation at 250 kW thermal power. The core is cooled by natural convection of the pool water with auxiliary cooling provided for protection of the filters and ion exchange equipment associated with reactor support piping.

The MUTR is used for academic instructions and operator training, performance of neutron and gamma irradiations, neutron activation analysis experiments, and tours and demonstrations for groups internal and external to the campus as well as for visiting nuclear power plant trainees.

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REACTOR USEAGE During the past year the MUTR operated for a total of 104 runs (Run Numbers 4185 - 4289), which are categorized below:

Operator Training/Requalification* 18 runs Tours, Labs & Demonstrations** 13 runs Calibration, Maintenance, and Surveillance 18 runs Irradiations and Activations 57 runs

Note: Some runs involved training and surveillance and may be counted in both categories.

- Note: Some of the runs in the Tours, Labs & Demonstrations category consisted of operator training. They are not included in the training category.

To perform these runs the core produced 69.827 MWh (kWh meter change from 365519 kWh to 435346 kWh), with a corresponding bumup of 3.96 Grams of U,-235.

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III. SURVEILLANCE TESTS AND INSPECTIONS All required surveillance tests and inspections were performed at the specified intervals. The required surveillance items for this reporting period include:

WATER SAMPLE TESTS AIR SAMPLE TESTS RADIATION SURVEYS CONTROL ROD DROP TEST RAM CALIBRATION SNM INVENTORIES ALARA REVIEW In addition to the above surveillance items, the following maintenance operations were performed on the indicated dates:

9/4/12 Dri-rite baked.

12/11/12 Dri-rite replaced.

3/5/13 Dri-rite baked.

6/19/13 Dri-rite replaced.

10/18/12 Makeup water system resin replaced 02/07/13 Primary water system resin replaced Additional minor maintenance was performed such as light bulb replacement and fine-tuning of equipment was performed as necessary.

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IV. CHANGES TO FACILITY There were no significant changes to the Facility during this reporting period. Changes not related to operations include ongoing cosmetic changes including the installation of new flooring, paint, and other miscellaneous minor esthetic upgrades.

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V. ENVIRONMENTAL SURVEYS OF SURROUNDING AREAS All continuous monitoring for this year was accomplished using fixed-mounted film badges throughout the interior of the reactor building itself. These badges recorded the following exposures:

Monitor Location Dose (mrem) 1 Control Room 348 2 Pool Surface 2338 3 Hot Room 192 4 W. Wall Upper 506 5 S. Wall Upper 220 6 S. Wall Lower 394 7 E. Wall Lower 477 8 Water Room 2746 9 N. Wall Lower 751 10 W. Wall Lower 863 7

VI. RADIOACTIVE RELEASE AND DISCHARGE TO THE ENVIRONMENT The Reactor Storage Sump was not discharged during this reporting period.

The only release from the MUTR consists of Ar-4 1. From Section 11 of the SER for the MUTR, a 30.0 MWh operation year would result in the generation of 100 mCi of Ar-41 for the entire year from the reactor pool tank. For this operational year, a combined 232.76 mCi of Ar-41 was released to the reactor building. This value was used in the EPA program COMPLY. The MUTR meets the EPA level 2 compliance for airborne release of radioactive materials. A copy of the output for the EPA computer program "COMPLY" is appended with this report.

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VII. ALARA REVIEW FOR FACILITY PERSONNEL AND VISTOR EXPOSURE A review of exposure records and all facility operations were performed by facility management as part of the annual ALARA audit. For this reporting period, all badged personnel and students received doses less than ten per-cent of their annual dose limit.

The Pocket Dosimeters recorded minimal exposure for all guests and service personnel. Calibrations of these self-reading dosimeters were performed on an annual basis by the University of Maryland's Radiation Safety Office.

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VIII. UNSCHEDULED SHUTDOWNS/REPORATBLE OCCURRENCES There were three unscheduled shutdowns during this period.

The first occurred during operation number 4228, which was performed on January 22, 2013. This shutdown was caused by a spurious trip in the period scram circuitry.

The next shutdown happened during operation number 4283 dated May 22, 2013 and was caused by a momentary loss of building power.

The final unscheduled shutdown occurred on June 7, 2013, the operation number was 4287, and as in the January 2013, scram, the cause was noise in the period circuit.

There were no reportable occurrences during this reporting period.

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IX. SPECIAL EXPERIMENTS There were two new special experiments performed during this reporting period. The descriptions and procedures are attached in Appendix B.

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X. CHANGES IN FACILITY STAFF There were no significant changes to staffing during this reporting period.

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APPENDIX A: EPA COMPLIANCE Below is the output from the EPA program COMPLY for the Ar-41 release from the MUTR:

COMPLY: V1.5d. 10/10/13 8:49 40 CFR Part 61 National Emission Standards for Hazardous Air Pollutants REPORT ON COMPLIANCE WITH THE CLEAN AIR ACT LIMITS FOR RADIONUCLIDE EMISSIONS FROM THE COMPLY CODE, VERSION 1.5d Prepared by:

University of Maryland Maryland University Training Reactor College Park, Maryland Mohamad Al-Sheikhly (301) 405-5214 Prepared for:

U.S. Environmental Protection Agency Office of Radiation Programs Washington, D.C. 20460 COMPLY: Vl.5d. 10/10/13 8:49 13

MUTR Argon 2012-2013 SCREENING LEVEL 1 DATA ENTERED:

Effluent concentration limits used.

DATA ENTERED FOR STACK 1:

CONCENTRATION Nuclide (curies/cu m)

AR-41 4. 30E-05 DATA ENTERED FOR STACK 2:

CONCENTRATION Nuclide (curies/cu m)

AR-41 4.30E-05 NOTES:

Input parameters outside the "normal" range:

None.

RESULTS:

You are emitting 12600.0 times the allowable amount given in the concentration table.

- - Failed at level 1.

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COMPLY: V1.5d. 10/10/13 8:49 MUTR Argon 2012-2013 SCREENING LEVEL 2 DATA ENTERED:

RELEASE RATES FOR STACK 1.

Release Rate Nuclide (curies/YEAR)

AR-41 4.300E-05 RELEASE RATES FOR STACK 2.

Release Rate Nuclide (curies/YEAR)

AR-41 4.300E-05 SITE DATA FOR STACK 1.

Release height 10 meters.

Building height 10 meters.

The source and receptor are not on the same building.

Distance from the source to the receptor is 15 meters.

Building width 15 meters.

SITE DATA FOR STACK 2.

Release height 10 meters.

Building height 10 meters.

The source and receptor are not on the same building.

Distance from the source to the receptor is 15 meters.

Building width 15 meters.

Default mean wind speed used (2.0 m/sec).

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COMPLY: VI.5d. 10/10/13 8:49 NOTES:

Input parameters outside the "normal" range:

None.

RESULTS:

Effective dose equivalent: 6.5E-05 mrem/yr.

Comply at level 2.

This facility is in COMPLIANCE.

It may or may not be EXEMPT from reporting to the EPA.

You may contact your regional EPA office for more information.

- - END OF COMPLIANCE REPORT **********

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Appendix B: Special Experiments 17

MARYLAND UNIVERSITY TRAINING REACTOR PROPOSAL TO INVESTIGATE NEUTRON ATTENUATION TESTING OF MATAMIC HT USING THE MUTR THERMAL COLUMN NEUTRON BEAM 18 56 Final version dated 1/23/2013 Cameron Goodwin Approved Director, Maryland University Training Reactor Mohamad Al-Sheikhly Approved Chair, Reactor Safety Committee Gary Pertmer 18

1.0 PURPOSE This procedure describes the test method, calculations, and calibration for neutron attenuation testing to be conducted on Metamic HT materials. Samples and standards will be subjected to a neutron beam to determine the transmitted neutrons, and ultimately the areal density.

Neutron Attenuation is the measure of neutrons that are transmitted through a sample to determine the B4C loading through the comparison to a developed standard curve. Standards are provided to the testing facility by Holtec with known amounts of B4C loading.

The neutron source will be either a nuclear reactor or isotope that emits a reasonable amount of neutrons such that a statistically significant amount of counts can be provided during testing in a reasonably short period of time. A neutron beam will be used, with adequate shielding to prevent scattering and minimize uncertainty in the neutron beam, which will penetrate the sample being tested and the neutrons that are transmitted through will be detected. Samples will be placed manually in the path of the neutron beam.

The beam will be normalized by monitoring the power level and correcting; or, through short testing campaigns and re-calibrations. The sample percent transmission will be compared to a developed areal density standard curve to determine the corresponding areal density value.

The MUTR thermal column will serve as a source of thermal neutrons for these irradiations. A specially-designed collimator access plug will allow only a narrow beam of thermal neutrons to stream past the reactor containment into the experiment. The nominal fluence of this beam has been measured with a NIST-calibrated fission chamber (3.6E4 cm- 2s-1).

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2.0 PRECAUTIONS AND REQUIREMENTS 2.1 Removing the existing concrete access plug from the thermal column is classified as a Special Experiment. Therefore this proposal requires the approval of the Reactor Safety Committee and the Reactor Director. The removal of the concrete thermal-column access plug does not require a License Amendment Request (LAR) as allowed under 10CFR50.59. The removal of the plug does not require a modification of the technical specifications. The proposed experiment has been evaluated and does not meet the criteria set forth in 10CFR50.59(c).

2.2 The reactor will not be operated for a minimum of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to removing the concrete thermal-column access plug and replacing it with the neutron-beam collimator access plug. This will allow activities of short-lived fission products to decay. If the collimator access plug is installed prior to this experiment, this precaution is not necessary.

2.3 The neutron-beam collimator access plug allows only a 2" diameter beam of radiation to propagate beyond the reactor containment. A brick of lead and BorAl serves as a shutter to block the neutron and gamma radiation in the beam. This shutter is attached to a pulley system and is operable from the southeast corner of the reactor, outside of the beamline. When the shutter is open, a warning light visible from the southern half of the reactor building illuminates, signifying that neutrons are passing through the experimental setup.

2.4 The Maryland University Training Reactor has previously been operated at a number of power levels with the thermal column neutron-beam collimator access plug in place. Dose rate measurements at several power levels are listed in Tables 1-3. Dose rates during the proposed experiment are not expected to exceed those listed in Tables 1-3. If changes are made to the experimental configuration that are expected to increase dose rates to personnel, surveys should be completed before proceeding.

2.5 No dose rates listed in Tables 1-3, with the exception of the neutron doses within the beam, exceed 100 mrem/hr, therefore, we do not expect the need to establish a high radiation area outside the beam. However, if at any time during the experiment dose rates exceed 100 mrem/hr outside of the thermal-column neutron beam, the operation will be paused, and a high radiation area will be established and marked, allowing control over individual access to the area in accordance with 10 CFR 20.1601. A high radiation area will be established only under the approval of the Reactor Director. Posting and surveys of the area will be completed by RSO staff.

2.6 Video monitoring of the thermal column and the south face of the reactor containment will be implemented. This video monitoring will be conducted from the reactor control room by the Reactor Operator.

2.7 During the experiment, personnel will communicate with the reactor operator on the console via radio handsets. No changes in plug position will occur without the Reactor Operator's knowledge.

2.8 Neutron detectors and gamma/beta detectors will be used to monitor the radiation levels throughout the reactor facility during course of this experiment.

2.9 In addition to whole body TLDs, personnel handling the access plugs or working with equipment in the proximity of the beamline will wear ring and wrist badges.

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2.10 Personnel shall not place any part of their body in the beamline when the beam shutter is open.

Borated polyethylene bricks around the experiment will restrict access to the beam. Personnel will obey ALARA principles at all times during this experiment.

2.11 The materials used in this experiment are aluminum and boron carbide. No significant activation is anticipated, see section 3.3. The neutron-beam collimator plug is made of carbon steel and is tack welded with nickel-free welding rod. The collimator plug is filled with a mixture of steel shot, paraffin wax, and boron carbide. Beyond the collimator plug, components in the beam line include the beam shutter (lead and BorAl), the target (Metamic HT), the target holder (aluminum), and the beam stop (borated polyethylene and BorAl). Fixed activities of these components following irradiation are expected to be < 1 uCi combined, with radionuclide half-lives < 300 s.

2.12 After irradiation, components from this experiment will be surveyed by an RSO staff member prior to removal from the MUTR facility. These components will be tagged with indications of their radionuclide concentrations.

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3.0 REQUIRED MATERIALS AND EQUIPMENT

- Neutron detectors, gamma/beta detectors.

- Ring, wrist, and whole body TLDs and EPDs.

- Radio handsets for communication with Reactor Operator

- Lever hoist for removal of access plugs - Columbus McKinnon Model 5318, rated at 3000 lbs

- Thermal-column neutron-beam collimator access plug.

- Lift tables rated at 2000 lbs to support the concrete access plug and the neutron-beam collimator

- Kimwipes and methanol for cleaning the neutron-beam collimator access plug

- Swipes to check access plugs for contamination

- Pulley, lever,. cable, 80/20 track, lead/boral brick, light, and microswitch for shutter

- Borated polyethylene bricks for beam stop

- Fission chamber reference detector and associated electronics

- Metamic HT targets

- 80/20 support table for holding the targets 3.1 PROPERTIES OF METAMIC HT Metamic HT is developed by Holtec International Inc. The material is constructed from aluminum mixed with natural boron carbide. The properties for Al, B and C can be seen below.

Material Thermal Cross Section (barns)

Scattering Capture Fission Al 2E0 2E-1 N/A B-10 2E0 2E+3 N/A C (natural) 5E0 2E-3 N/A 4.1 PROCEDURE

1. Allow core to sit cold for a minimum of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . Skip to 5 if the collimator is already installed.

While monitoring dose rates with neutron and gamma detectors, remove concrete access plug from thermal column using lever hoist, cable and lift table.

2. Swipe concrete access plug to check for removable contamination. Transfer swipes to RSO staff.

3. After cleaning, insert the neutron beam collimator access plug into thermal column.

4. Attach lead/boral brick and shutter track assembly to pulley and cable. Attach shutter microswitch circuit. Verify operation. Return shutter to closed position.

5. Startup reactor to a maximum of 200kW.

6. Develop calibration curve.

6.1 Turn on computer, counting equipment, and detectors and check that they are operating properly.

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6.2 A standard will be selected and its details recorded on Table 4 or similar.

6.3 Place the standard in the holder in the line of sight with the neutron beam. Run the neutron beam for 30 seconds or until 15,000 counts are achieved, record counts. Time and counts may be limited for standards that have higher areal density so that testing time is reasonable, such as 2 minutes.

6.4 The counts shall be divided by time to get counts per minute (CPM) and recorded on Table 4 or similar. Note: This may be done during post processing.

6.5 The measurement results will be corrected for background, reactor power, and counting errors (may be 3 sigma) and recorded.

6.6 Repeat the above steps until all standards have been measured and results recorded.

6.7 Background count will be measured by blocking the neutron beam, black coupon. Because it is expected to have a low neutron count, the counting time will be limited to a maximum of 2 minutes.

6.8 Using the known areal density of the standards and the measurement results just taken, a calibration curve may be generated. Counts will be plotted against the known areal density and an exponential interpolation fit will be added to the plot. Figure 1 is an example of a calibration curve. Note: This calibration curve is valid for 1 run (a maximum of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ).

Curve may not actually be created but used in tabular format.

Ep.,,,tiaJIntrpaI!ati-rnFit to Ccupon Data 400

~j 200 100 0 J 0 JO 10 25 MuminalMinuo 3 Si;rna Ara~eet B-10 eminkr2)

Figure 1: Example calibration curve for the transmission count rate corrected for reactor power and background plotted against Pmin, the measured nominal standard areal density minus 3o error.

7. Coupon testing 7.1 A coupon will be selected and its details recorded on Table 5 or similar.

7.2 Place the coupon in the holder in the line of sight with the neutron beam.

7.3 Run the neutron beam for 30 seconds or until 15,000 counts are achieved, record counts.

Time and counts may be limited for samples that have higher areal density so that testing time is reasonable, such as 2 minutes.

7.4 The counts shall be divided by time to get counts per minute (CPM) and recorded on Table 5 or similar. Note: This may be done during post processing.

7.5 The measurement results will be corrected for background, reactor power, and counting errors (may be 3 sigma) and recorded.

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7.6 Repeat the above steps until all coupons, or coupons tested within an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months window, have been measured and results recorded. Note: This calibration curve is valid for I working day or a maximum of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

7.7 Using the calibration curve, determine the areal density of the coupon.

8. Coupon data recorded.

8.1 The coupon ID, count time, raw sample counts, raw power count, corrected sample counts, nominal areal density, time, and date will be recorded.

9. Close the beam shutter and shut down the reactor.

10. No components will leave the MUTR without prior survey by RSO staff.

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FIGURES AND TABLES

Con~d Rod

Iflrwwnft~ld Rod IEJRabbt a01 9 8 7 6 5 4 3 2 1 Through Tube \//

I Figure 2. Plan view of MUTR core layout showing the inner face of the thermal column.

TWPFMAL COLOMIN LINERIT -N TANK WALL Aur.

CO&UN

~ CAM PS, GRA028 3 6vtpn r,&

Figure 3. Elevation view of the MUTR thermal-column. Access plug cavity is highlighted in red. Core begins at the right-hand side of the diagram. Edge of concrete reactor containment appears at the left-hand side of the diagram.

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Figure 4. Model of the neutron-beam collimator access plug.

Figure 5. The collimator access plug after filling with wax, steel shot, and boron carbide mixture.

26

A variety of extruded shapes produced from Metamic Figure 6:Extruded shapes of Metamic. Samples would be approximately 3 in by 3 in.

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except mapping locations, setup. All dose experimental including the containment face of the reactor south 1 and 2.

view of the as they appear in Table

7. A top down south balcony, are labeled for those taken on the Figure 28

MUTR thermal column neutron beam dose rates with beam shutter closed (measured 10/25/11) access plug skirt behind beam stop personnel bench time power level neutron gamma neutron gamma neutron gamma 11:46am LPC (24 mW) 11:53 100mW 11:56 500 mW 11:59 1W 12:04pm 10W 12:07 100W 12:12 500 W 12:17 1 kW 0.05 12:21 10 kW 20 (0.36) 0.3 0.07 0.2 12:34 SOkW 80 (1.46) 3.5 0.3 1.6 12:47 75 kW 100 (1.82) 3.5 0.4 1.3 12:58 100 kW 100 (1.82) 4.0 0.5 10(0.18) 2.0 1:07 125 kW 120 (2.18) 5.0 20 (0.36) 0.6 10 (0.18) 2.5 1:20 150 kW NA 6.0 20(0.36) 0.7 10(0.18) 3.0 1:35 175 kW NA 7.0 40 (0.72) 1.0 10 (0.18) 3.5 1:46 200 kW NA 8.0 40 (0.72) 1.3 10 (0.18) 3.0 1:59 250 kW NA 10.0 60 (1.09) 1.4 10 (0.18) 3.5 Table 1. Dose rate measurements as a function of power level, taken in the vicinity of the thermal column neutron beam with the beam shutter closed. See Figure 11 for measurement locations. Highlighted columns show the highest dose rates experimental personnel must work in when the beam shutter is closed. All gamma measurements are in mR/hr (measured with meter # 9884). All neutron measurements are fast neutron measurements in cpm (calibrated conversions in parentheses in mR/hr) (measured with meter # 14358). Blank entries signify background dose rates. Gamma background = 0.03 mR/hr. Neutron background = 0 cpm.

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MUTR thermal column neutron beam dose sured 10/26/11) 10 cm from apparatus in beam line access plug skirt time power level beamline boundary neutron gamma neutron gamma neutron gamma neutron gamma 10:04am LPC (24 mW) 100mW 500 mW 1W 10 W 10:24 low 100 W 500 W 0.05 10:36 1 kW 40* 0.07 10 kW 10 (0.18) 0.2 10 (0.18) 0.4 10(0.18) 0.18 3.5e4* 0.7 50 kW 20(0.36) 1.1 20(0.36) 2.0 10(0.18) 1.1 1.6e5* 3.0 75 kW 80 (1.46) 2.0 100 (1.8) 2.0 20 (0.36) 2.0 NA 5.0 100 kW 110 (2.0) 3.0 140 (2.5) 4.0 40(0.7) 3.0 NA 7.0 11:35 125 kW 140 (2.5) 3.0 180 (3.3) 5.0 60 (1.1) 3.5 NA 8.0 150 kW NA 3.5 240(4.4) 6.0 60(1.1) 3.5 NA 10.0 NA 120 (2.2) 4.0 NA 30 175 kW NA 4.0 6.5 NA 4.5 NA 8.5 160 (2.9) 5.0 NA 35 12:15 250 kW NA 6.5 NA 11.5 180 (3.3) 7.0 NA 50 Table 2. Dose rate measurements taken in the vicinity of the thermal column neutron beam with the beam shutter open. See Figure 11 for measurement locations. Highlighted columns show the highest dose rates experimental personnel must work in when the beam shutter is open. All gamma measurements are in mR/hr (measured with meter # 9884). All neutron measurements are fast neutron measurements in cpm (calibrated conversions in parentheses in mR/hr) (measured with meter # 14358). Blank entries signify background dose rates. Gamma background = 0.03 mR/hr. Neutron background = 0 cpm. NA signifies that measurements were not taken for this location and power level.

These measurements are for thermal neutrons. The neutron meter was not calibrated for thermal readings and therefore these numbers are only relative.

Measurements in this location (in the open beam) were discontinued for neutrons due to high dose rates.

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MUTR dose rates with thermal column beam shutter open (measured 10/26/11) south balcony time power level gamma neutron left center right left center right 100 kW 0.15 0.2 0.14 0 0 0 11:35 125 kW 0.19 0.3 0.15 0 0 0 150 kW 0.2 0.36 0.17 0 0 0 175 kW 0.27 0.43 0.23 0 0 0 200 kW 0.34 0.52 0.29 0 0 0 12:15 250 kW 0.44 0.7 0.41 0 0 0 Table 3. Dose rate measurements taken on the south balcony of the reactor building with the thermal column neutron beam shutter open. All gamma measurements are in mR/hr (measured with an ion chamber). All neutron measurements are fast neutron measurements in cpm (measured with meter # 14358). These dose rates did not fluctuate when the beam shutter was opened and closed repeatedly.

Standard B-10 Areal B-10 Areal Thickness Count Time Fission Fission CPM Date Time ID Density Density (in) (s) Chamber Chamber (g/cm2 ) Error Counts Error 0.0133 0.014 0.0186 0.020 0.0238 0.025 0.0271 0.030 0.0367 0.040 0.0461 0.050 0.0553 0.060 Blank Al

Background

(Black coupon, 100mg)

Table 4. Calibration Curve Test 31

Coupon ID Count Time Fission Fission CPM Nominal Date Time (s) Chamber Chamber Areal Counts Error Density (g/cm2)

Table 5. Coupon Tests 32

APPENDIX A 50.59 Changes, tests and experiments.

(a) Definitions for the purposes of this section:

(1) Change means a modification or addition to, or removal from, the facility or procedures that affects a design function, method of performing or controlling the function, or an evaluation that demonstrates that intended functions will be accomplished.

(2) Departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analyses means:

(i) Changing any of the elements of the method described in the FSAR (as updated) unless the results of the analysis are conservative or essentially the same; or (ii) Changing from a method described in the FSAR to another method unless that method has been approved by NRC for the intended application.

(3) Facility as described in the final safety analysis report (as updated) means:

(i) The structures, systems, and components (SSC) that are described in the final safety analysis report (FSAR) (as updated),

(ii) The design and performance requirements for such SSCs described in the FSAR (as updated),

and (iii) The evaluations or methods of evaluation included in the FSAR (as updated) for such SSCs which demonstrate that their intended function(s) will be accomplished.

(4) Final Safety Analysis Report (as updated) means the Final Safety Analysis Report (or Final Hazards Summary Report) submitted in accordance with Sec. 50.34, as amended and supplemented, and as updated per the requirements of Sec. 50.71(e) or Sec. 50.71(f), as applicable.

(5) Procedures as described in the final safety analysis report (as updated) means those procedures that contain information described in the FSAR (as updated) such as how structures, systems, and components are operated and controlled (including assumed operator actions and response times).

(6) Tests or experiments not described in the final safety analysis report (as updated) means any activity where any structure, system, or component is utilized or controlled in a manner which is either:

(i) Outside the reference bounds of the design bases as described in the final safety analysis report (as updated) or (ii) Inconsistent with the analyses or descriptions in the final safety analysis report (as updated).

(b) Applicability. This section applies to each holder of a license authorizing operation of a production or utilization facility, including the holder of a license authorizing operation of a nuclear power reactor that has submitted the certification of permanent cessation of operations required under Sec. 50.82(a)(1) or a reactor licensee whose license has been amended to allow possession of nuclear fuel but not operation of the facility.

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(c)(1) A licensee may make changes in the facility as described in the final safety analysis report (as updated), make changes in the procedures as described in the final safety analysis report (as updated), and conduct tests or experiments not described in the final safety analysis report (as updated) without obtaining a license amendment pursuant to Sec. 50.90 only if:

(i) A change to the technical specifications incorporated in the license is not required, and (ii) The change, test, or experiment does not meet any of the criteria in paragraph (c)(2) of this section.

(2) A licensee shall obtain a license amendment pursuant to Sec. 50.90 prior to implementing a proposed change, test, or experiment if the change, test, or experiment would:

(i) Result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the final safety analysis report (as updated);

(ii) Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the final safety analysis report (as updated);

(iii) Result in more than a minimal increase in the consequences of an accident previously evaluated in the final safety analysis report (as updated);

(iv) Result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the final safety analysis report (as updated);

(v) Create a possibility for an accident of a different type than any previously evaluated in the final safety analysis report (as updated);

(vi) Create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the final safety analysis report (as updated);

(vii) Result in a design basis limit for a fission product barrier as described in the FSAR (as updated) being exceeded or altered; or (viii) Result in a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analyses.

(3) In implementing this paragraph, the FSAR (as updated) is considered to include FSAR changes resulting from evaluations performed pursuant to this section and analyses performed pursuant to Sec. 50.90 since submittal of the last update of the final safety analysis report pursuant to Sec. 50.71 of this part.

(4) The provisions in this section do not apply to changes to the facility or procedures when the applicable regulations establish more specific criteria for accomplishing such changes.

(d)(1) The licensee shall maintain records of changes in the facility, of changes in procedures, and of tests and experiments made pursuant to paragraph (c) of this section. These records must include a written evaluation which provides the bases for the determination that the change, test, or experiment does not require a license amendment pursuant to paragraph (c) (2) of this section.

(2) The licensee shall submit, as specified in Sec. 50.4, a report containing a brief description of any changes, tests, and experiments, including a summary of the evaluation of each. A report must be submitted at intervals not to exceed 24 months.

34

(3) The records of changes in the facility must be maintained until the termination of a license issued pursuant to this part or the termination of a license issued pursuant to 10 CFR part 54, whichever is later. Records of changes in procedures and records of tests and experiments must be maintained for a period of 5 years.

[64 FR 53613, Oct. 4, 1999, as amended at 66 FR 64738, Dec. 14, 2001]

GP-903 ATTACHMENT 1 IOCFR50.59 Screening Form Submitted By: Layla Shahamat Date: 1/15/2013 Telephone: 301-405-1904 Email: laylas@umd.edu Activity to be performed: These series of experiments will focus on neutron attenuation testing of Metamic HT materials made of various amount of aluminum, carbon and boron-10. Samples will be exposed to a neutron beam to determine the transmitted neutrons, and ultimately the areal density. The MUTR thermal column is utilized as a thermal neutron source for these experiments.

35

Screening questionnaire If all questions are answered "NO," then submit the proposed activity to the full RxSC for review and approval unless the submission only addresses a modified routine experiment.

If ANY of questions 1-4 are answered "YES," a 50.59 review MUST be performed.

If question 5 is "YES", the activity requires a License Amendment Request (LAR) before activity is implemented.

YES NO

1. Does the proposed activity involve a change to a SSC that adversely affects a FSAR (as F1 updated) described design function?

2. Does the proposed activity involve a change to a procedure that adversely affects how

[] FSAR (as updated) SSC design functions are controlled or performed?

3.. Does the proposed activity involve revising or replacing a FSAR (as updated) described evaluation methodology that is used in establishing the design basis or used in F1 the safety analysis?

4. Does the proposed activity involve involve a test or experiment not described in the FSAR (as updated), where an SSC is utilized or controlled in a manner that is outside the reference bounds of the design for that SSC or is inconsistent with analyses or descriptions in the FSAR (as updated)?

F1 5. Is a change to the Technical Specifications incorporated in the license required?

Conclusion The activity may be conducted without performing a written 50.59 evaluation.

The activity requires a written 50.59 evaluation.

The activity may be performed only after submitting a License Amendment Request (LAR).

Provide justification for conclusion: The $3.00 total value of in-core experiments is not applicable in this case since the experiments are performed outside of the reactor core. There are no proposed changes to the structure of the reactor as a result of performing these experiments. These experiments do not require any changes to the Technical Specifications.

RxSC Reviewed by: Chair Date:

Approved by: Director Date:

For guidance with questions, see section 4.2 of NEI 96-07.

APPENDIX B 10 CFR 20.1601 36

High Radiation Areas 2.1 Options for Access Control Of the options for access control provided in 10 CFR 20.1601(a), the most widely used procedure at nuclear power plants is keeping high radiation areas locked. Although licensees have the option to control high radiation areas with the use of a control device to reduce radiation levels when an individual enters the area or the use of an alarm to alert the individual and the supervisor to an entry into a high radiation area, experience has shown that these options have limited practical application at nuclear power plants. In addition to the provisions of 10 CFR 20.1601(a), a nuclear power plant licensee may apply for Commission approval of alternative methods for control under 10 CFR 20.1601(c). See Regulatory Position 2.4 below.

2.2 Positive Access Control Positive control over each individual entry is required by 10 CFR 20.1601(a)(3) when access is required to a high radiation area that is normally controlled by being locked. In a large facility such as a nuclear power plant, appropriate positive access controls can be instituted through the use of radiation work permits (RWPs) or an equivalent program. Such a system ensures appropriate supervision through specific procedures that establish requirements for control and delegate responsibility to qualified individuals. Procedures for establishing positive control over each entry should provide for:

1. Surveys that identify the radiation hazards in the area should be made and the results documented;

2. An appropriate level of supervision to determine that exposure of the individual to the hazards is warranted;

3. Communication of the nature and extent of the radiation hazards to each individual entering the area;

4. Protective measures (e.g., shielding, time limits, protective clothing, monitoring) to protect the individual from excessive or unnecessary radiation exposure; and

5. Permission for only authorized individuals to enter the high radiation area with all entries and exits documented.

2.3 Direct or Electronic Surveillance Direct or electronic surveillance is identified in 10 CFR 20.1601(b) as a substitute for the controls required in 10 CFR 20.1601(a). The direct or electronic surveillance should have the following capabilities as a minimum.

1. Detect attempted unauthorized entry,

2. Warn individuals that their attempted entry is unauthorized, and

3. Alert the proper authority about an unauthorized entry so that action can be taken to correct the situation.

2.4 Alternative Methods for Access Control 37

The requirements in 10 CFR 20.1601(a) for access to high radiation areas may, in some instances, cause unnecessary restrictions on plant operations. According to 10 CFR 20.1601(c), licensees may apply to the Commission for approval to use alternative methods for control. The following method is acceptable to the NRC staff as an alternative to the requirements in 10 CFR 20.1601(a) for the control of access to high radiation areas.

Each high radiation area as defined in 10 CFR Part 20 should be barricadedw and conspicuously posted as a high radiation area, and entrance thereto should be controlled by requiring issuance of a radiation work permit (RWP) or equivalent. Individuals trained and qualified in radiation protection procedures (e.g., a health physics technician) or personnel continuously escorted by such individuals may be exempted from this RWP requirement while performing their assigned duties in high radiation areas where radiation doses could be received that are equal to or less than 0.01 Sv (1.0 rem) in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (measured at 30 centimeters from any source of radiation) provided they are otherwise following plant radiation protection procedures, or a general radiation protection RWP, for entry into such high radiation areas. Any individual or group of individuals permitted to enter such areas should be provided with or accompanied by one or more of the following:

A radiation monitoring device that continuously indicates the radiation dose rate in the area,

A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rates in the area have been determined and personnel have been made knowledgeable of them,

" An individual qualified in radiation protection procedures with a radiation dose rate monitoring device. This individual is responsible for providing positive radiation protection control over the activities within the area and should perform periodic radiation surveillance at the frequency specified in the radiation protection procedures or the applicable RWP.

In addition, areas that are accessible to personnel and that have radiation levels greater than 0.01 Sv (1.0 rem) (but less than 500 rads at 1 meter) in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at 30 cm from the radiation source, or from any surface penetrated by the radiation, should be provided with locked doors to prevent unauthorized entry, and the keys should be maintained under the administrative control of the shift supervisor on duty or health physics supervisor. Doors should remain locked except during periods of access by personnel under an approved RWP that specifies the dose rates in the immediate work areas and the maximum allowable stay time for individuals in that area. In lieu of a stay time specification on the RWP, direct or remote continuous surveillance (such as closed circuit TV cameras) may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.

Individual high radiation areas that are accessible to personnel, that could result in radiation doses greater than 0.01 Sv (1.0 rem) in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and that are within large areas where no enclosure exists to enable locking and where no enclosure can be reasonably constructed around the individual area should be barricaded and conspicuously posted. A flashing light should be activated as a warning device whenever the dose rate in such an area exceeds or is expected to exceed 0.01 Sv (1.0 rem) in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at 30 cm from the radiation source or from any surface penetrated by the radiation.

2.5 Controls for High Radiation Areas (Control Points and Barriers) 38

Controls (e.g., locked doors, access control, and posting) for high radiation areas may be established at locations beyond the immediate boundaries of the high radiation areas to take advantage of natural or existing barriers. The use of one locked door, or one control point where positive control over personnel entry is exercised, to establish control over multiple high radiation areas is acceptable provided the following conditions are met:

1. The individual high radiation areas are barricaded and posted separately to identify the actual areas of concernA

2. Control points are established sufficiently close to the high radiation areas that adequate supervision of access to the areas can be assured, and

3. The required protective measures and other requirements for entering the high radiation areas (e.g., dosimetry, monitoring) are enforced at the control pointLJ&

2.6 Control of Keys The shift supervisor or the radiation protection manager (or their respective designees) should administratively control the issuance of keys to personnel requiring access to high radiation areas and the return of the keys.

MARYLAND UNIVERSITY TRAINING REACTOR PROPOSAL FOR CHARACTERIZING THE THROUGH TUBE PORTS USING IC-18 ION CHAMBER DETECTORS 39

AISIEýY>

18 56 Final version dated Monday, October 21, 2013 Submitted by Slavica Grdanovska Approved Director, Maryland University Training Reactor Mohamad Al-Sheikhly Approved , Chair, Reactor Safety Committee Gary Pertmer Table of Contents 1.0 PURPO SE .................................................................................

2.0 PRECAUTIONS AND REQUIREMENTS ......................................................

3.0 REQUIRED MATERIALS AND PREPARATIONS ...................................... 43 4.0 PROCEDURE ..............................................................................

5.0 10 CFR 50.59 REVIEW ......................................................................

APPEN DIX .................................................................................................................................. 47 I. FIGURES ................................................................................................................................ 1 II. TEXT OF 10 CFR 50.59 .................................................................................................... 9 40

§ 50.59 Changes, tests and experiments. ........................................ 9 III. TEXT OF 10 CFR 20.1601 ............................................................................................ 12 High Radiation Areas .................................................................................................................... 12 41

4.0 PURPOSE The purpose of this proposal is to provide a detailed procedure on how to characterize the through tube of the Maryland University Training Reactor (MUTR). The purpose of this set procedures is to characterize the port for future experiments and shielding calculations. The design of the Through Tube and its plugs are shown in the Appendix under Section I (Figures). Future experiments involving fiber optics require knowledge of the neutron and gamma dose rates at the opening and inside the through tube ports with the reactor at power.

5.0 PRECAUTIONS AND REQUIREMENTS 2.1 Removing the through tube plugs is not classified as a Special Experiment. However, operating the reactor with the Through Tube with the modified "Through Tube Plugs" installed has not been performed in the past and is considered a "Special Experiment."

Therefore this proposal requires the approval of the Reactor Safety Committee and the Reactor Director. This proposal must also comply with the guidelines set forth by 10 CFR 50.59 (See Appendix III).

2.2 The MUTR has not previously been run at power with the through tube ports open, thus the exposure of individuals working in the reactor building needs to be measured and understood BEFORE any experiments can be conducted. To characterize the Through Tube, the access plug (Figure 6) will be removed and modified plugs (Figure

7) installed in order to measure the doses outside of the reactor vessel for future experiments. The modified plugs are made of concrete with a thin steel covering. The dimensions and design of the modified plugs are shown in Figure 8.

2.3 If at any time during the characterization the radiation level exceeds 100 mrem/hr outside of either through tube port, the characterization will be suspended and a high radiation area will be established and marked, providing positive control over individual access to the area in accordance with 10 CFR 20.1601. The establishment of a high radiation area, if necessary, will be completed with the approval of the Reactor Director. Postings and surveys of the area will be completed by Radiation Safety personnel. If a high radiation area is required, there will be video monitoring of the area. The monitoring will be conducted from the reactor control room by the reactor operator.

2.4 Neutron detectors and gamma/beta detectors will be used to monitor the radiation levels at the face of the east and west through tube ports while loading, irradiating, and unloading the detectors in the beam port. All operations performed with the port open will conducted with radiation levels observable by the reactor operator or another person in the control room. Operation will be suspended and a high radiation area will be established if radiation levels exceed 100 mrem/hr.

42

2.5 No materials will be irradiated during the characterization with the exception of radiation monitors and other items as required to ensure the safety of those within the reactor confines and the general public.

2.6 During the dose rate measurements, personnel will communicate with the reactor operator on the console via radio handsets. No changes in plug position or reactor power level will occur without the Reactor Operator's knowledge and the approval of the SRO.

2.7 The reactor building wall across from the West Through Tube Port is also an exterior wall. It will be ensured that the wall is marked appropriately during the experiment and will perform surveys as necessary. Results of surveys will be recorded and attached to the experimental proposal for use in additional experiments using the modified through tube plugs (Figure 9). It will also be ensured that no personnel are in the room if it becomes marked as a Radiation Area.

2.8 Entrance to the bottom floor of the reactor will be secured during this procedure. All personnel will enter from the top floor entrance. The area in front of and around the east and the west post will be roped off so that personnel are not exposed to higher doses of radiation than necessary. At least one person will be equipped with an direct reading alarming dosimeter capable of monitoring both gamma and neutron dose rates.

2.9 In addition to whole body TLDs, personnel handling the access plugs will wear wrist and ring badges. They will also wear gloves to minimize the spread of contamination, should it be present.

6.0 REQUIRED MATERIALS AND PREPARATIONS Two IC-18 ion chamber detectors Neutron detectors and gamma/beta detectors to monitor radiation.

Radiation monitoring as recommended by the Radiation Safety Officer or their appointed representative.

A concrete shield (Figure 10)

Direct reading electronic dosimeter capable of monitoring both neutron and gamma radiation.

7.0 PROCEDURE 7.1 Verify that the reactor has not been operated for a minimum of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months by examining the operations log.

7.2 The beam port access cover (Figure 4) will be removed while monitoring radiation levels while the reactor is shutdown.

43

7.3 While monitoring the radiation levels with neutron and gamma detectors, the beam port access plugs will be removed and replaced with the "Modified Through Tube" plugs while the reactor is in a shutdown condition. The modified plugs are made of concrete with a thin steel covering. The outer diameter of the plugs is 6.5 inches and each is 31 inches long. There are two modified plugs that are inserted when the access plugs are removed. The concrete acts as a radiation shield and therefore radiation is only present in the 0.625 inch hole that leads to a rotational tube through the modified plugs (Figure 8). The rotational tube has been designed to prevent radiation streaming.

Personnel will wear gloves while handling the plugs to minimize the spread of any contamination present. A contamination survey will be conducted when moving any of the through tube plugs or access plugs.

7.4 The IC-18 ion chambers will be placed at the end of the 0.625 inch hole of the modified plugs BEFORE the reactor is brought to power. One at a time. One measures gamma and neutron doses and one just measures neutron doses.

7.5 The reactor will be brought to low power critical while monitoring the radiation levels.

7.6 The reactor will be raised in power to 100 mW, 500 mW, 1W, 10 W, 100 W, 500 W, 1 kW, 10 kW, 50 kW, 75 kW, 100 kW, 125 kW, 150 kW, 175 kW, and finally 200 kW. The radiation levels for both neutron and beta/gamma levels will be recorded (see Figure 9). and verbal approval will be given by the on-site SRO after consultation with the RSO or their designee to increase power to the next level. Radiation levels will be taken as explained in Figure 9. Radiation levels measurements and performed surveys will be presented to the reactor director for review. A concrete shield (Figure 10) will also be used directly outside of the plug to minimize the radiation received.

7.7 The reactor will be brought subcritical and secured for a minimum of ten hours before removing the first chamber.

7.8 The second chamber will be installed in place of the first and steps 4.3 through 4.6 will be repeated.

7.9 Upon completion of the experiment, the port access plugs will'be fully inserted into the beam port AFTER the reactor is in a shutdown condition for at least 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

8.0 10 CFR 50.59 REVIEW 8.1 The removal of the east beam port does not require a License Amendment Request (LAR) as allowed under 10 CFR 50.59 for the following reasons:

44

8.1.1 This removal of the plug does not require a modification of the technical Specifications. (10 CFR 50.59(c)(1)(i) 8.1.2 The proposed experiment has been evaluated and does not meet the criteria set forth in 10 CFR 50.59(c)(2)

GP-903 ATTACHMENT 1 (10CFR50.59 Screening Form) 45

46 APPENDIX 47

MARYLAND UNIVERSITY TRAINING REACTOR 2012-2013 ANNUAL OPERATING REPORT I. FIGURES Figure 2:

1

MARYLAND UNIVERSITY TRAINING REACTOR 2012-2013 ANNUAL OPERATING REPORT

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a c6..'f ,.'oMs %o. Sousa oc.?,

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'C'S'COPE/SI 'a ,/j, Pr S A0S~,4 509025,r AC TEE aA:jaSM0 Figure 4: Through tube plugs 2

MARYLAND UNIVERSITY TRAINING REACTOR 2012-2013 ANNUAL OPERATING REPORT Figure 5: Outside of West Portsi 3

MARYLAND UNIVERSITY TRAINING REACTOR 2012-2013 ANNUAL OPERATING REPORT i piastiic access cover remouvvu 4

MARYLAND UNIVERSITY TRAINING REACTOR 2012-2013 ANNUAL OPERATING REPORT Figure 7: Through Tube access plugs Figure 8: Modified plug to be inserted into Through Tube 5

MARYLAND UNIVERSITY TRAINING REACTOR 2012-2013 ANNUAL OPERATING REPORT 6~58.465

@0 U0.01 All

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II C14 01 I, 04 10 aS

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'A Top II Figure 9: Modified plugs design 6

MARYLAND UNIVERSITY TRAINING REACTOR 2012-2013 ANNUAL OPERATING REPORT Figure 10: Radiation measurements will be taken at the fourteen spots noted above. Positions 1 and 3 are 30 cm away from the west and east port of the through tube at 450 angle from the center. Position 2 is 30 cm away from the center of the west and east port of the through tube. Positions 4 and 6 are 174 cm away from the west and east port of the through tube at 450 angle from the center. Position 5 is 174 cm away from the center of the west and east port of the through tube. Position 7 is the interior point of the west and east wall, 348 cm away from the west/east port of the through tube. All referenced positions are at a height of 91.44 cm.

7

MARYLAND UNIVERSITY TRAINING REACTOR 2012-2013 ANNUAL OPERATING REPORT r igure i i: concrete smeoaing to De piacea in wront ot ping 8

MARYLAND UNIVERSITY TRAINING REACTOR 2012-2013 ANNUAL OPERATING REPORT II. TEXT OF 10 CFR 50.59

§ 50.59 Changes, tests and experiments.

(a) Definitions for the purposes of this section:

(1) Change means a modification or addition to, or removal from, the facility or procedures that affects a design function, method of performing or controlling the function, or an evaluation that demonstrates that intended functions will be accomplished.

(2) Departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analyses means:

(i) Changing any of the elements of the method described in the FSAR (as updated) unless the results of the analysis are conservative or essentially the same; or (ii) Changing from a method described in the FSAR to another method unless that method has been approved by NRC for the intended application.

(3) Facility as described in the final safety analysis report (as updated) means:

(i) The structures, systems, and components (SSC) that are described in the final safety analysis report (FSAR) (as updated),

(ii) The design and performance requirements for such SSCs described in the FSAR (as updated), and (iii) The evaluations or methods of evaluation included in the FSAR (as updated) for such SSCs which demonstrate that their intended function(s) will be accomplished.

(4) Final Safety Analysis Report (as updated) means the Final Safety Analysis Report (or Final Hazards Summary Report) submitted in accordance with Sec. 50.34, as amended and supplemented, and as updated per the requirements of Sec. 50.71(e) or Sec. 50.71(f), as applicable.

(5) Procedures as described in the final safety analysis report (as updated) means those procedures that contain information described in the FSAR (as updated) such as how structures, systems, and components are operated and controlled (including assumed operator actions and response times).

(6) Tests or experiments not described in the final safety analysis report (as updated) means any activity where any structure, system, or component is utilized or controlled in a manner which is either:

(i) Outside the reference bounds of the design bases as described in the final safety analysis report (as updated) or 9

MARYLAND UNIVERSITY TRAINING REACTOR 2012-2013 ANNUAL OPERATING REPORT (ii) Inconsistent with the analyses or descriptions in the final safety analysis report (as updated).