ML19350D096

Environ Rept for Facility

ML19350D096
Person / Time
Site:	National Bureau of Standards Reactor
Issue date:	11/30/1980
From:	NATIONAL INSTITUTE OF STANDARDS & TECHNOLOGY (FORMERL
To:
Shared Package
ML19350D089	List: ML19350D088 ML19350D096
References
FOIA-81-113 NBSR-12, NUDOCS 8104130304
Download: ML19350D096 (47)

Text

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ENVIRONMENTAL REPORT h

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J NATIONAL BUREAU OF STANDARDS 4

NBSR 12 j

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Compiled and Edited by The Staff Research Reactor Radiation Division November 1980 1

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Submitted by *Ihe National Bureau of Standards, U.S. Department of Commerce in support of its application to increase the authorized power level of its reactor to 20 MWt and renew the license for a twenty-year period.

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Section 1.

INTRODUCTION 1-1 Section 2.

ENVIRONMENTAL REPORT 2-1 2.1 Introduction 2-1 2.2 Site Description 2-1 2.2.1 Facility Siting 2-1 2.2.2 Population Distribution 2-2 l

2.2.3 Meteorology, Seismology, Geology and Hydrology 2-4 2.3 Facility Description 2-5 2.3.1 Building Description 2-5 2.3.2 Confinement and Emergency Ventilation Features 2-6 2.3.2.1 Confinement Features 2-6 2.3.2.2 Emergency Exhaust System 2-7 2.3.2.3 Internal Clean-up System 2-7 2.3.3 Reactor Faci 11ty Description 2-7 2.3.3.1 Reactor 2-7 2.3.3.2 Reactor Upgrade 2-8 2.3.3.3 Experimental Facilities 2-9 2.3.4 Process Systems Description 2-9 2.3.4.1 Primary Cooling System 2-9 2.3.4.2 Secondary Cooling System 2-10 2.3.4.3 Emergency Core Cooling 2-11 2.3.5 Radwaste Treatment 2-12 2.3.5.1 Sources of Radwaste 2-12 2.3.5.2 Source Strengths Now and at 20 MW 2-14 2.3.5.3 Radwaste Disposal 2-16 2.3.5.4 Summary and Conclusion 2-17 2.3.6 Radiological Effects of Reactor Upgrade 2-17 2.4 Environmental Effects of Operation 2-18 2.4.1 Sources of Radioactive Releases 2-18 2.4.2 Environmental Effects 2-19 l

i 2.5 Envirenmentet Meniteries vreeram 2-20 l

2.5.1 Monitoring Methods 2-20 2.5.1.1 Soil and Grass Sampling 2-20 2.5.1.2 Water Sampling 2-20 2.5.1.3 External, Background Monitoring 2-20 l

2.5.2 Summary of Results 2-20

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2-21 2.6 Environmental Impacts of Plant Accidents 2-21 2.6.1 Small Release Outside Containment 2-21 2.6.2 Radwaste System Failure 2-21 2.6.2.1

System Description

2.6.2.2 Maximum Concentrations Expected 2-21 2-22 2.6.2.3 Accidental Release 2.6.3 Fission Product Release to Primary 2-22 System 2-23 2.6.4 Primary to Secondary Leak 2-24 2.6.5 Refueling Accidents 2-25 2.6.6 Spent-Fuel Handling Accident 2.6.6.1 Fuel Element Drop-In Pool 2-25 2.6.6.2 Heavy Object Drop Onto Fuel 2-25 I

Rack 2-25 2.6.6.3 Fuel Cask Drop 2.6.7 Accident Initiation Events Considered 2-26 in DBA 2.7 Unavoidable Effects of Facility Construction 2-27 and Operation 2.8 Alternatives to Upgrade and Operation of the 2-27 Facility 2.9 Long-Term Effects of Facility Construction and 2-28 Operation 2-28 2.10 Cost and Benefits of Facility 2-31 2.11 Conclusion 11 11 11 El El III

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I SECTION 1.

INTRODUCTION This Environmental Report was prepared in support of the National Bureau of Standards' application to amend its license No. TR-5 to increase the authorized maximum power level of its research reactor from 10 W t

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to 20 W t and to renew the license for a period of 20 years.

The National Bureau of Standards Reactor (the NBSR) is located on the Bureau's 576 acres site near Gaithersburg, Maryland. The reactor is a heavy-water moderated and reflected research reactor currently operating at 10 W.

It first went critical on December 7, 1967 and was being operated routinely at 10 W eighteen months later. The reactor is operated 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day, seven days a week with routine shutdowns every 3

six weeks for partial refueling. Additional shutdowns are scheduled during the summer and at Christmas time to accommodate staff vacations.

To date (Nov. 1980), the reactor has accumulated 640,000 W-hrs of operation.

It is normally on line between 70 and 75 percent of the time.

The NBSR is used for a vida range of programs including materials research by neutron scattering, trace analysis by neutron activation analysis, neutron radiography, neutron flux standardization, neutron dosimetry using filtered beams in the kev region, radiation effects, and isotope production. The reactor has 11 radial beam tubes, 2 through beam tubes, a thermal column, 4 built-i'. pneumatic tubes, provisions for up to ten in-core thimbles and 7 reflector thimbles.

25 major experimental instruments, most of which can be used simultaneously, are installed at these reactor facilities. The facilities are used by more than 200 scientists and technicians from 18 NBS divisions and offices, 18 other l

government agencies, and 25 universities and industrial organizations.

1' The reactor was designed for 20 W operation, but operated initially at 10 W until program demand and operating experiences were suf ficient to justify full power operation.

The only parts of the reactor system not origlually constructed for fuli 20 W operation were a few elements of the process system. During the past ten years, the process system has been upgraded as components had to be replaced.

The original aluminum heat exchanger has been replaced by two stainless steel heat exchangers.

Currently, the secondary system is being upgraded including rearrangement III h1

of pumps and piping and the replacement of the old cooling tower with a Consequently, no additional major changes to more efficient modern one.

the physical plant are required to operate at 20 MW.

The reactor and reactor site are described in detail in NBSR 9 This report addresses only those items relating to the supplements.

l The major areas in which this report present license application.

supercedes NBSR 9_fnelude:

f New population distribution estimates e

Description of water treatment e

Atmospheric dispersion calculations e

Discussion of primary to secondary leak e

References

" Final Safety Analysis Report on the NBS Reactor, NBSR 9" (1.1)

(April 1966).

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SECTION 2.

ENVIRONMENTAL REPORT

2.1 INTRODUCTION

This environmental report has been prepared following the guidance I

provided by the Nuclear Regularary Commission (2.1), (2.2).

Some areas have already been covered in detail in NBSR 9 (2.3) and are only touched

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on briefly in this addendum. The site and facility are described briefly. The only significant modifications to the original facility as L-described in NBSR 9 have been the replacement of the original aluminum I:

heat exchanger with two stainless heat exchangers and the replacement of the original cooling tower by a new, more efficient one and the associated modification of the secondary system piping. These have been described in detail in Engineering Change Notices aad reported in our semi-annual reports or in changes in the technical specifications.

The effects of a variety of postulated accidents and environmental releases are analyzed.

f Alternatives to increasing the reactor power to 20 MW and cost benefits are also discussed.

2.2 SITE DESCRIPTIE 2.2.1 FACILITY SITING. The NBS campus is shown in Figure 2.1.

The NBSR is located in the center of the 1/4 mi radius circle drawn on the figure. The NBS campus is a 575 acre site bounded on the cast by a major highway (T-270), on the north and west by Md. 124, and on the southeast by Muddy Branch Road. As can be seen from Figure 2.1, the area adjacent to the reactor building is occupied by a parkir.g lot, the reactor cooling tower, and roads. Thus, the area within a 500 ft.

radius of the reactor building stack is not readily available for the

l construction of new buildings and planning for future development of the NBS site does not include any new buildings within 500 ft. of the reactor I

stack. The concentration of radioactive argon from the reactor effluent 500 ft. (N 150 m) from the stack has been calculated following the guidelines provided by the " Workbook of Atmospheric Dispersion" (2.4).

The annual average concentration has been calculated under the following assumptions:

1.

Reactor power is 20 MW.

2.

Rate of argon-41 release is 1400 Ci/yr. This is double the present average 10 MW release rate of 700 Ci/yr.

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3.

A building taller than the, reactor stack has been built 500 ft. from the reactor so it would intercept the center line of the plum from the stack.

4.

Neutral diffusion parameters are used (Class D, wind speed l

equal 3 m/s)(2.4).

5.

The wind blows into a 22.5 sector in the direction of the bui) ding 20% of the time and meanders uniformly within the sector.

s Under these conditions, the argon-41 average concentration at the building at the same elevation as the top of the reactor stack would be only

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-8 3.0 x 10 pCi/mL which is considerably less than the 4.0 x 10 pCi/mL permitted by the (10CFR20) for unrestricted areas.

Exposure to personnel in these areas is further reduced by more than a factor of 4 by the fact that such buildings would usually be occupied only during the normal 40-hour work week.

If the Bureau should, at some future date, consider constructing a new building to be occupie1 on a regular basis within 500 ft. of the reactor stack, an analysis would be performed to deterniine that any radiation exposere to occupants of the building would meet the regulations applicable at that time before construction would be approved.

The NBS campus is located in upper Montgomery County as shown in a

Figure 2.2.

Circles of 1 mi and 5 mi radius have been drawn around the reactor to aid in estimating distances.

2.2.2 POPULATION DISTRIBUTION. The daytime population of the NBS campus is about 3000, all of whom are under the control of NBS.

The current and projected population for the surrounding area out to 10 mi is shown in Table 2.2-1.

The data is divided into 16 sectors of 22.5 each, with the center of the first sector being due north.

The population is shown as a function of distance from the reactor and is projected through the year 2000.

The rapid population growth that has taken place in Montgomery County was anticipated when the NBSR was first licensed. This can be seen from Table 2.2-2 which compares the projections given in the original M

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10 11 12 13 14 15 16 Torat Rg5 100 100 600 400 2900 100 200 200 300 200 100 500 100

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1930 TG 3'Y 400 600 600 500 800 100 200 300 300 200 100 700 400 5700 1985 200 300 400 600 600 600 900 200 200 400 200 200 100 800 500 6400 0-1 1990 300 300 400 600 700 700 1000 300 300 500 500 300 200 800 500 7400 1995 300 300 500 700 800 800 1000 400 400 500 600 400 200 800 500 8200 2000 1980 1200 3500 2600 1500 300 300 300 360 800 300 100 400 1200 1400 300 14500 1935 300 1800 3700 3000 1700 600 1100 900 300 1000 300 300 500 1500 1600 800 19400 1-2 1990 300 2300 3700 3000 2300 700 1300 1100 400 1500 300 500 1200 2000 1700 900 23000 1995 500 2400 3700 3100 1300 900 1400 1200 400 2000 300 600 2000 2200 1800 1000 25800 2000 700 2600 3800 3200 2500 1000 7500 1300 400 2500 400 700 2700 2500 1800 1100 28700 1980 500 800 2000 3400 1800 1500 400 400 200 900 1100 400 800 400 1000 900 16500 6

1985 1100 800 1800 3'400 1000 2200 900 700 500 1000 800 600 900 500 1200 1000 18400 y

2-3 1990 1100 800 1800 3500 1100 2300 1000 800 500 1000 800 600 1000 900 1400 1200 19300 1995 1300 900 1800 3500 1300 2400 1000 800 600 1100 900 600 1100 1100 1600 1500 21500 2000 1500 1000 2000 3500 1500 2500 1000 900 700 1200 900 700 1200 11LC 1900 1800 23300 1980 11100 4700 600 2100 1600 2900 6000 3200 800 400 200 700 400 400 1700 1800 38600 1985 12600 5100 1200 2100 1700 2900 6000 3200 1100 700 r400 900 500 700 2700 3100 44900 3-4 1990 It:300 5300 1300 2200 1800 3000 6100. 6500 1300 900 500 1000 600 1100 3500 4000 53t:00 1995 1'1700 5400 1400 2300 1800 3000 6100 6600 1500 1100 600 1100 700 1600 3600 4700 56200 2000 15500 5400 1500 2300 1800 3100 6100 6600 1700 1100 800 1200 800 2200 4600 5800 60500 1980 7200 6090 200 1900 1200 2000 9600 3800 1700 300 300 300 200 400 900 1200 37200 1935 8500 7300 300 1900 1300 2100 9700 3800 1800 500 600 500 200 700 1700 2100 43000 8 00 2100 1600 2300 9900 3900 2000 800 700 600 300 1500 3300 2500 50800 4-5 1%0 11400 7500 4

1995 12000 7600 500 2300 1700 2400 10000* 4100 2100 1000 800 700 400 1800 4600 3400 55460 f800 1000 58WO 2011 13000 7800 500 2300 1800 2500 10000 4100 2200 1100 1000 8u0 100 2000 1

19M0 2200 2/00 2700 10000 8500 4G500 7:4400 20600 25700 1800 1700 1100 1000 1400 2200 2700 175200

'1935 2900 3500 3800 13500 12000 46600 46600 21600 26500 2200 2000 1400 It:00 1700 3200 4500 193200 AtOO 5800 210100 t

t7500 (19700 22600 27400 2600 2200 1600 1500 2200 i

5-10 1990 3200 3700 4700 15000 16000 i

52700 2tG00 28600 3000 2400 2000 1900 2700 5500 8000 227600 1995 3400 4000 5100 16500 17500 ri9700 i

2000 3600 4300 5400 17500 18500 51800 55200 25900 29900 3200 2700 2200 2000 2900 6000 8900 240000 TOTAL 1980 28t900 Pu u o cts: Q.51-Q.13) 1985 324600 1990 3G3500 2000 419100

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SAR (NBSR 9) with th2 current (1980) populatien figures. Tha NBSR 9

,~J figures were conservative in that they projected about 50 percent greater population within a 3 mi radius than actually exists.

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actual figures for population within a 10 mi radius, however, are very close. This shows that, although the population of Montgomery County has grown rapidly, the growth was anticipated at the time that the reactor was approved for siting at its present location.

Table 2.2-2 Comparison of NBSR 9 Population Projections with Actual 1980 Population Radius NBSR 9 Projection Actual Population I

0-1 at 6 800 2 900 a

1-2 mi 21 250 14 500 2-3 mi 23 500 16 500 3-4 mi 30 700 38 600

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4-5 mi 28 000 37 200 5

5-10 mi 189 300 175 200 Total 299 550 284 900

.q 2.2.3 METEOROLOGY, SEISMOLOGY, GEOLOGY, AND HYDROLOGY.

The meteorology for the NBSR site was discussed extensively in NBSR 9.

It was based on many years of data collected at Washington National Airport. These data were shown te correlate well with more limited data taken at the NBSR site. The analysis showed that there were to unusual or particularly severe weather conditions that would present a problem for the siting of a reactor at the NBS site.

The meteorological conditions have not changed since the earlier analysis so the conclusions reached in NBSR 9 are still valid.

We have continued to record wind conditions at the NBSR and have used a portion of those data to reconfirm the correlation of wind patterns at the NBSR with those at Washington National Airport.

Figure 2.3 is the scatter diagram for the correlation of the average wind speed between Washington National Airport and NBS for the six month period Sept. 1978-Feb. 1979.

If the correlation were perfect, all the dots would fall on the 45 straight line.

If there were no correlation at all, the dots would be scattered at random over the whole diagram.

Figure 2.4 is a similar diagram for the daily resultant of the wind direction over the same period.

It can be seen from these two diagrams that the correlation is very good.

1 Tha calcnolcgy, g;cicgy, cnd hydrolcgy of tha cito wara digcussrd cxtensivaly in NBSR 9.

They have not changed so no, additional discussion is provided here.

2.3 FACILITY DESCRIPTION The reactor is housed in a 90 ft. x 90 ft. square concrete structure designed l

to confine the results of any credible accident which might occur.

In addition to housing the reactor, it provides space to carry out the scientific programs for which the reactor was designed..Although space is provided within the l

confinement building for both beam tube and in-core irradiation experiments, the additional laboratories and offices required to support the scientific 3

programs are not located within the confinement building.

A detailed description of the reactor and its building can be found in NBSR 9, so only a summary is provided here.

2.3.1 BUILDING DESCRIPTION.

The reactor building is shown in Figures 2.5 through 2.7.

The building has three main levels; the basement which houses the primary process equipment and the spent Iuel storage pool, the first floor which serves the reactor beam holes, and the second floor which provides access to the top of the reactor and where the control room is located.

The normal ventilation systems in these three areas are separate so that the 1

mixing of air between the different levels is minimized.

Figure 2.5 shows the basement plan view.

About half of the basement area is devoted to the process room which is surrounded by a thick concrete wall.

Access to this room is through a steel shielding door from the pool area.

Directly under the reactor in the process room is the subpile room whose 3 ft.

concrete walls and steel access door provide shiciding for the relatively high radiatior area directly under the reactor.

On the south side of the basement is the storage pool.

The pool is used R.

to store spent fuci elements until shipment.

A canal leading from the

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room allows fuel elements to be transferred directly from t'he reactor venei into the storage pool without the use of transfer casks.

On the east side of the basement are two radiological laboratories and a counting room.

They are used primarily in conjunction with the four pneumatic tubes which provide rapid sample access to the reactor.

Figure 2.6 shows the plan view of the first floor.

The experimental facilities using the beam tubes and thermal column are located on this floor.

These facilities are serviced by a 15 ton annular crane.

This floor is at the I

same level as the adjacent laboratories. Access to the reactor confinement I

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IW building fra tha 1 b:rcterisc is thr: ugh tha two dooro chown cn thn crct cid2 of the confinement building.

Each set of doors consist of two double doors sep.arated by a short passageway.

In an emergency, a sliding steel door closes and seals the building shut with an inflatabic gasket.

The reactor, about 20 ft. in diameter, is located in the center of the floor. The biological shield runs up to the ceiling and serves as support for the inner rail of the annular crane which services the area.

A plan view of the second floor of the confinement building is shown in Figure 2.7.

The top of the reactor shielding in flush with the floor and utility and access trenches under the floor provide access to the radiation facilities that go into the core and reflector from the top of the reactor.

The area is served by a 20 ton crane which is used to handle shielding casks for radioactive sampics removed from the reactor.

'Ivo large square hatches in the floor provide access to the floor below making it possible to move heavy l

equipment from one level to the other.

The control room is also located on this floor and looks out over the reactor top.

All the process instrumentation as well as the reactor instrumentation is located in the control room so all aspects of reactor operation can be monitored from this single location.

2.3.2 CONFINEMENT AND EMERGENCY VENTII.ATION FEATURES 2.3.2.1 Confinement Features.

The building housing the reactor is designed to confine any radioactive material released in an accident so that it may be exhausted in a controlled manner through an emergency exhaust system (Section 2.3.2.2) which filters out the radioactive materials before the air is exhausted to the environment.

Because the exhaust air is filtered and released in a controlled manner, the building does not have to be as leak tight as a total containment building.

The NBSR confinement building, however.

is designed to be as tight as possible with a conservative upper limit on the

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allowed leak rate of 24 cfm for a pressure differential across the walls of 1" of water.

Any release of radioactive material into the confinement system is detected by redundant detectors in the normal ventilation system and initiates a building closure. The sliding steel doors are closed automatically and scaled by inflatable gaskets and all normal ventilation ducts are sealed shut "o isolate the confinement building.

The emergency exhaust

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fan is automatically started and maintains the building at a negative internal pressure differential across the walls of about.25" H O so that any

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2 leakage is into the building. At the same time, a large internal I

cican-up system (Section 2.3.2.3) of 5000 cfm capacity can be activated w

to circulate air within the building through filters to clean it up and minimize the release of radioactive iodine to the environment.

2.3.2.2 Emergency Exhaust System. The emergency exhaust system 3

consists of two redundant sub-systems A and B cach of which contains an exhaust fan and identical filters and controls.

Either sub-system can draw air from the normal exhaust system duct work at a rate up to a maximum of 100 cfm. The fan operates to maintain a negative differential pressure of about 0.25" H O between the inside and the outside of the 2

building.

Both of the redundant emergency exhaust fans are provided with both an AC and a DC motor to assure continuous availability even when AC power fails. The exhaust fans were siz=d to assure that a negative pressure could not be generated inside the building that would impe r the buildings structural integrity.

Each of the emergency exhaust subsystems has separate sets of filters consisting of pre-filters, absolute filters, and activated charcoal filters.

The charcoal filters are backed by a fourth particulate f11ter. The absolute f11ters have an efficiency greater than 99.9%

for DOP aerosol of 0.3 pm (0.3 micron) size and the charcoal filters have an efficiency greater than 99% for the extraction of iodine.

2.3.2.3 Internal Clean-Up System.

During emergency operation, the air in the reactor building can be recirculated and filtered by means of a separate system. Air in this system is drawn from all areas of the reactor building and circulated at a rate of 5000 cfm through an absolute filter and an activated charcoal filter bank similar to the filter system for the emergency exhaust system (Section 2.3.2.2).

This system is designed to remove particulate and gaseous activity such as iodine with an yproximate time constant (time required to reduce concentration by 1/e) of two hours.

2.3.3 REACTOR FACILITY DESCRIPTION 2.3.3.1 Reactor.

An elevation drawing of the reactor is shavn in I

Figure 2.8 and a plan view in Figure 2.9.

The reactor is cooled, moderated id 2-7

e cnd raficcted by h::vy w:tcr. Th2 fu21 cicmenta cra crrcnged cn osv:n inch c:ntors which cllowa r:om for vertical incora experim ntal fecilitics and for use of semaphore type shim arms. The reactor vessel is made of I

aluminum as is the rest of the core structure.

The large volume above the core is provided so that spent-fuel elements may be removed from the core, transferred to the transfer chute and lowered into the storage pool below - all without having to remove the spent-fuel from the reactor shielding. A large annular tank.is also located in this region and is always filled with D 0 because its open top is just below the normal 2

water level for reactor operation.

It serves as the first step of the emergency core cooling system (Section 2.3.4.3).

The reactor is cooled by heavy water flowing up through the fuel elements and returning down through two outlet pipes at the bottom of the reactor vessel.

The heat is removed from the primary system through a heat exchanger to a secondary (H O) cooling system which disperses its heat into the atmosphere via a 2

cooling tower.

M 2.3.3.2 Reactor Upgrade.

No modifications will have to be made in the core structure or fuel element configuration.

(The number of fuel elements in the core remains the same.) Reactor nuclear and process instrumentation will have to have their ranges adjusted for the higher power and greater coolant flow. Additional instrumentation will be added i

for monitoring the additional heat exchanger in the primary system and the increased flow in the secondary system.

The only part of the reactor system not originally constructed for full 20 MW operation were a few elements of the process system.

During the past ten years, as components needed to be replaced, they were upgraded in accordance with current technology and sized for 20 MW operation. The original 10 MW aluminum heat exchanger has been replaced by two 10 MW stainless steel heat exchangers; the original cooling tower is being replaced by a more efficient 20 MW cooling tower; and the-secondary proc.ess pumps and piping have been modified to increase their efficiency. Therefore, no additional major construction is needed to

'iperate at 70 $f.

11.

i... n = 11. l u nes-d to increase the inventory of fuel has been analyzed including the need to store spent-fuel for a longer time before shipment 3

for reprocessing.

Since the present storage time is determined more by scheduling conditions than by cool down requirements, it is anticipated that no signLticant increase in the inventory of fuel will be needed.

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2.3.3.3 Exparimental Fecilitiec. Ths rccctor wss danigned to mest a variety of needs based on the broad, spectrum of activities of the National Bureau of Standards.

Consequently, the reactor has a large y

number of facilities of varying kinds.

These include 11 radial beam tubes, 2 through beam tubes, a thermal column, 4 built-in pneumatic tubes, provisions for up to ten in-core thimbles and 7 reflector thimbles.

Experimental facilities presently available at the reactor include 10 fully automated neutron scattering facilities, 3 monoenergetic filtered beams in the kev region, a radiography facility, 2 standard fast-flux fields, 5 pneumatic tubes, 2 in-core irradiation facilities, and a s

capture y-ray facility.

In addition, a small angle neutron scattering facility, a vertical radiography facility and a fast neutron pneumatic tube are under development. Most of these instruments are used 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day, 7 days a week and are highly automated to meet this demand. All the reactor beam tubes are now utilized so the addition of a significant number of experimental facilities to meet increasing demand is not feasible. Doubling the reactor power, however, will make it possible to do experiments more quickly and of a more complex nature on the existing facilities thereby making it possible to meet the great demand for access to these facilities.

2.3.4 PROCESS SYSTEMS DESCRIPTION 2.3.4.1 Primary Cooling System.

To minimize the loss of heavy water used as the primary cooling fluid, the primary loop is completely closed and sealed. The D 0 surface in the reactor vessel is swept by 2

helium gas which passes through a catalytic recombiner to recover any deuterium produced by radiolysis of the water. At 10 W, two pumps drive the water through the reactor and a single 10 W heat exchanger. A third primary pump is installed in the system such that any two of the three may be used. A second 10 W heat exchanger has also been installed and checked out to serve as a backup at 10 W.

To preserve it in its new, unused condition, the connecting piping was disconnected.

It will be used at 20 W operation to provide the additional heat exchange capability required. Although there will no longer be an unused backup heat exchanger available at 20 W, the reactor power can always Se reduced if it should become necessary to take one heat exchanger out of service.

A fourth primary pump and its connecting piping is also on hand.

Thus, the only modifications h1 1

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th:t mu t ba mada to thm primary cycten in to r: connect tha c:cond heat l

exchanger and install the fourth pump.

The increased volume of the system would require an increase of about 1000 galicas in the heavy water inventory, but no significant increase in the loss of heavy water is anticipated.

g At 10 W, the primary water inlet temperature is about 100 F and the outlet temperature about 112 F., At 20 W, these temperatures are expected I

to increase by no more than 10 F or 20 F since the primary flow will be significantly increased.

2.3.4.2 Secondary Cooling System.

The secondary cooling system is in the process of being upgraded and modernized. The present cooling tower is approaching the end of its useful life and is unable to provide adequate cooling during the summer.

It is being replaced by a larger tower with three cells instead of two which will provide greater flexibility in choice of operating mode.

The three cell tower is 21 ft. high, 27 ft, wide, and 65 ft. long. The capacity of the basin is 75,000 gal. At 10 W the average daily water makeup is about 50,000 gal. The blowdown is routed to the sanitary sewer.

The average blowdown rate is about 11 gpm.

g The blowdown and water makeup rates should be approximately twice these numbers at 20 W.

The secondary water treatment services for corrosion and microbiological control, are provided by the Nalco Chemical Company, Chicago, Illinois.

A principal consideration in the selection of the treatment is that the chemicals used are environmentally acceptable.

In this regard, some time ago, NBS changed from a chromate treatment which is a superior corrosion inhibitor to organic treatment strictly to obtain improved environmental results. The types, amounts and functions of the chemicals used in the treatment are outlined below.

Corrosion Control - Product Identification:

Nalco 8376 contains zinc for corrosion, polyester (organic phosphate) for scale inhibitor and

, dispersant, and liquid sulfonate as a disperant.

This product is added at a rate of one to two gallons per day to maintain 30-50 ppm in the system with zine concentrations maintained at 1-2 ppm.

j Microbiological Control a) Product Identification: Nalco 7328 is a mixture of quaternary and organo-tin compounds of n-alky dimethyl benzyl, ammonium chloride w

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(12.5%) cnd Bio (tric-n-butyltin oxide) 2.5%.

This product is cdded ct a rate of five to six gallons per week to maintain 40-200 ppm in the f

system.

l b) Product Identification:

Nalco 7320 is an organo-bromine labeled 2, 2 dibromo 3 nitrilo propionamide.

This product is added at a rate of 2 liters per week to maintain 1-12 ppm in the system.

pH Control - Concentrated sulfuric acid is added to the system at a rate of one to two liters per day to maintain the pH in the range of 7.8-8.3.

Blow Down - As stated previously the blow down is released to the sanitary sewer at a rate of about 11 gpm and contains the chemical concentrations given above. These are further diluted by more than a factor of 20 by dilution in the Bureau's sanitary sewer system before leaving the Bureau site.

Doubling the reactor power to 20 MW will not require any change in the chemical concentrations in the secondary cooling water and, at most, will require a doubling of the blow down rate to 22 gpm.

I 2.3.4.3 Emergency Core Cooling.

It is highly unlikely that the NBSR would suffer a loss of coalant accident.

The only mechanism is a rupture of the reactor vessel itself or of one of the main cooling pipes.

Since the reactor primary cooling system is a low temperature, unpressurized system, the possibility of a major rupture is extremely

{

remote. The possibility of damaging the piping in the primary system by y

external forces is also extremely remote since the portion of the basement containing the piping is locked during reactor operation and heavy

, m equipment access to that area is not possible.

Nevertheless, as an extra b

precaution, the NBSR is equipped with emergency core cooling.

In fact, there are two emergency systems. One is a completely passive system a uW which operates automatically with no electronic or moving parts and provides emergency cooling for the first half-hour af ter shutdown.

The other system requires operator action and provides a longer period of cooling.

1 As was shown in Figure 2.8, there is a large volume of water in the reactor vessel above the core.

This space was provided to facilitate 7i fuel element transfer, but it also provides the opportunity for a unique U

emargency core cooling facility.

The system consists of an annular water tank, open at the top, which is placed in the reactor vessel above m

W 2-11

lt Th2 cpsrcting watcr 1;v 1 in tha r;;cter c ra la cb:v2 tha t:p th] cara.

, cf th] canulce tank co tha trnk 12 c1wsy filled with w2 tar.

Tha ecnk has two openings at the bottom Icading to a distribution pan which directs water into each individual fuel element.

If water is drained from the reactor vessel for any reason, the water held up in the annular tank drains into the distribution pan and cools the fuel.

No valves have l

to be opened, no operator action is required and there is no signal to fail or give a false indication.

The tank takes about one-half hour to drain providing ample time for the initiation of additional action.

A second system provides a supplement to the one just described and y

provides cooling for a longer period of time if necessary.

Since the D 0 2

'l moderator is very expensive, light water which would degrade the heavy 9

water should not be used for emergency cooling.

Therefore, the second system consists of a 3,000 gal. tank filled with D 0 located near the top 2

q of the reactor building. The operator can open a valve which permits water from the tank to be forced up through the core by gravity.

The plenum piping is designed to minimize the possibility of a rupture in the plenum region.

If, however, the plenum is ruptured in such a way that emergency cooling water would be lost and not forced up through the

~

fuel, then the operator can simply switch valves and use the water from 1

the emergency cooling tank to replenish the water draining out of the annular tank in the reactor vessel.

The water spilled into the basement I

is collected in sumps and can be used to refill the 3,000 gal. emergency tank to provide continued emergency core cooling if necessary. H O is 2

I also available if it appears undesirable to reuse the D 0.2 2.3.5 RADWASTE TREATMENT 2.3.5.1 Sources of Radwaste.

Each of the sources of radioactive waste at the NBSR and the means for controlling the production of such material and/or reducing its release to the environment is outlined below.

Argon Radioactive argon-41 is produced by neutron capture in l

the stabic argon normally present in air when the air is introduced into voids mainly around the reactor vessel.

An extensive CO system has been 2

41 installed at the NBSR that serves to reduce Ar production by maintaining

~

a positive pressure of CO in the void area between the reactor vessel 2

and the thermal shield, thereby significantly reducing the amount of air pre ent in that area.

CO is als used in all pneumatic facilities 2

41 further reducing Ar production.

G 2-12

C cous Tritium - Rcdio ctiva tritium is prcduced by neutron capture in the D 0 moderator. Caseous tritit m, in the form of water vapor, is 2

l the result of evaporation of tritiated D 0.

The primary D 0 system 2

2 including the cooling and moderating water is a closed system to prevent the escape of the water or its vapor. All the free D 0 water surfaces 2

are swept by helium gas that passes through a recombiner and returns the condensed vapor to the primary system storage tank. Thus, the tritiated water vapor is contained and only very small amounts escape to the environment.

D 0 Spills and Leakage - Tritiated D 0 may be transferred, spilled 2

2 or leaked from the closed primary system as a result of fuel transfer operations, routine maintenance activities, failure of components, and leakage from the primary side of the heat exchanger to the secondary side. The means for preventing or minimizing the D 0 leakage from each 2

of these sources are listed below, a)

Fuel transfer operations - The release of tritium during fuel transfer operations is minimized by drying and purging the_ fuel transfer lock with helium prior to opening the isolation valve between the Reactor vessel and the storage pool.

The transferred elements because of their decay heat are essentially dry and only a few drops may be transferred to the storage pool.

i I

b) Routine maintenance activities - Release of tritiated D 0 2

during routine maintenance activities is prevented or reduced by appropriate operational procedures and practices, tet.porary containment of the material, and radiological safety controls and monitoring. Most of the small amounts of D 0 involved is retrieved and stored for future reprocessing.

2 c) Failure of components - Leak detectors, located on valves and flanges throughout the D 0 system, are employed to indicate immediately 2

any component failure that could release tritiated D 0.

A tritium 2

monitor constantly monitors the air within the process room and that exhausted from the confinement building. A leak in the D 0 system would 2

be quickly detected by this monitor. Other reactor instrumentation, such as the reactor vessel level indicator, provides a prompt indication of D 0 losses from the system as a result of component failure.

2 2-13

Ev n if cuch fcitures vara to eccur, the epilled D 0 will b2 contained 2

within the process room and is retrieved by installed sump pumps with little or no loss of D 0.

2 d) Heat exchanger leaks - The secondary system is monitored continuously by an on-?ine N-16 monitor which would detect very small D 0 leaks to 2

that system during reactor operation.

This monitor is augmented by water level instrumentation and by periodic sampling and evaluation of the secondary system water for the presence of tritium, using an independent counting system.

Since the installation of the new stainless steel heat exchangers there have been no failures.

The amount of leakage permissible is strictly limited by NBSR Technical Specifications and would result in average concentrations substantially below allowable limits.

Solid Waste - Solid radioactive waste is produced at the facility from three principal sources:

1.)

Experiments 11.) Routine maintenance iii.) Fuel reprocessing and cutting operations f

The amount of solid waste produced by each of the first two sources is minimized by careful planning prior to the initiation of the activity, procedural controls, and adherence to accepted radiological safety practices.

In general these are very small amounts.

The solid waste generated by the third source is from spent-fuel as a result of reactor operations.

This has been significantly reduced by improved fuel element design and increased fuel loading.

2.3.5.2 SOURCE STRENGTHS NOW AND AT 20 MW Argon Over the last three years an average of approximately 1

41 L

700 Ci of Ar were released annually from the reactor stack.

This is equivalent to an average annual concentration of 1.6 x 10-pCi/mL at h

the stack.

Using neutral atmospheric dispersion conditions (Class D, u

-11

= 3m/s), this is reduced to < 8 x 10 pCi/mL at the site boundary averaged over the year.

Even if it is assumed that ' Ar releases will l

double at 20 MW, actual concentrations at the boundary will be several r

hundred times below MPC.

Caseous Tritium - The average release of gaseous tritium for the last three years was less than 300 Ci per year. This corresponds to an h4 average annual concentration of 7.6 x 10-pCi/mL at the stack corresponding

?

2-14

~11 to chout 3 x 10 pCi/mL ct thm bsundary cssuming neutral ctmospheric dispersion conditions. The amount of release of tritiated water vapor j

is not expected to increase appreciably at 20 MW; however, tritium

{

concentration in the moderator increases with reactor operations.

Even if the present concentrations were to triple, the actual concentration of tritium at the boundary would be several orders of magnitude below

-7

~

the allowable limit of 2 x 10 pCi/mL.

D 0 Spills and Leakage - The present tritium concentration in the 2

D 0 is approximately 1000 pCi/mL (June 1977).

The expected maximum 2

activity in the future is 2000 pCi/mL. The typical average annual releases to the sanitary sewer is less than 2 Ci of tritium and less than 2 mci of other S-y activity. The average annual concentration of tritium in the liquid effluent released from the reactor when diluted by the Bureau's total annual discharge to the sanitary sewer system is

-6 about 7 x 10 pCi/mL which is typically 0.007% of MPC for tritium.

8-y concentrations are similarly insignificant 1y small.

Therefore, even if the concentration at 20 MW were to be triple that at 10 MW, it would still be several orders of magnitude below allowable limits.

j D 0 spills and leakage have been minimal and easily controlled in 2

the past and should not be affected by an increase in the power level.

Similarly, the probability of leakage through the heat exchanger should not be affected in view of the installation of reliable stainless steel heat exchangers and the quick detection of such leakage.

Solid Waste IE l

a.)

The estimated average activity of the low level, radioactive waste generated by experiments and the major part of reactor maintenance u

activities is about 350 mci per year.

b.)

The estimated average activity of high level radioactive waste generated by specific reactor maintenance activities such as replacement of filters and resins is about 2.5 Ci per year.

The amounts of low and high level solid wastes from routine experimental and reactor maintenance operations are expected to only increase slightly as a result of operation at 20 MW.

c.)

To date (Nov.1980), approximately 220 spent-fuci elements have been shipped for reprocessing which covers eight years of 10 MW operations.

Improved fuel element design and increased loading over the l

years have reduced the annual number of elements used by more than a factor Ill i

2-15

ef twa frco th t criginally picnn2d.

Furth2r improvements cre picnned l

I such that the number of spent-fuel e,lements used per year at 20 MW will be less than that originally planned for 10 MW.

The same holds true for I

the number of non-fuel sections of spent elements shipped as solid waste. The amount of high level solid waste, from fuel cutting operations (i.e. non-fuel sections), shipped from the facility was 84 Ci involving a total of four shipments.

This is the total quantity of high level waste produced from fuel cutting operations since the reactor became operational in 1967.

It is anticipated that one shipment each of high level waste and spent fuel will be made cach year as a result of operations at 20 MW.

2.3.5.3 RADWASTE DISPOSAL Storage - All low level solid waste and limited amounts of high level solid waste, such as drums of used resin, are stored and prepared for shipment in a separate waste storage building located at the rear of the reactor facility.

Spent-fuel and high level waste resulting from fuel cutting operations are stored in the reactor spent-fuel storage pool. When a shipment of this material is necessary, it is loaded under I

water into a shipping cask and transferred directly to the carrier.

Liquid effluent from the reactor is stored in a 1,000 gallon retention tank or a 5,000 gallon hold-up tank.

Small amounts of liquid waste, which contain activity other than tritium and cannot be released to the i

sanitary sewer system, are collected by health physics, absorbed in a suitabic material, and then treated as solid waste.

The disposal of this solid waste follows procedures approved by the Nuclear Regulatory Commission and acceptable for a licensed burial ground.

Control - All radiological control of solid and liquid waste'is the general responsibility of the reactor's health physics section, af ter the material has been collected by them.

Radiation monitoring of solid waste material is performed by health physics personnel.

All gaseous and liquid effluents released from the facility are monitored by in-line monitors, health physics personnel, or both.

Q Monitoring of the gaseous effluent relaased from the reactor stack is accomplished by both a radiation monitor mounted in the stack itself and sampling of the effluent air by health physics personnel.

Liquid effluent released from the facility is sampled and evaluated for activity content by health physics prior to release, id 2-1e 1

Disposal - Salid waste is transferred to o disposal fccility for land burial.

Spent-fuct is shipped in special casks for reprocessing.

Low level liquid effluent is diluted and released to the sanitary j

sewer system.

Gaseous activity is amply diluted by a high volume stack discharge rate prior.to its release and subsequent dispersion in the atmosphere.

2. 3. 5. 4 Summary and Conclusions.

The amount of radwaste released l

as a result of the operation of the NBSR is small and is carefully monitored and controlled.

Environmental monitoring and sampling of areas around the reretor covering a period of many years, before and I

after reactor startup in 1967, showed no significant deviation from background levels. As is shown above, the anticipated radiological releases at 20 MW will be no more than twice the 10 MW 1evels and will result in radioisotope concentrations no greater than 1% of MPC at the site boundary.

2.3.6 RADIOLOGICAL EFFECTS OF REACTOR UPGRADE.

A discussion of the radiological effects of reactor upgrade includes the following:

Radiation source terms at 20 MW Radiological releases to the environment Radwaste handling Dose to in-plant personnel The first three items are covered in section 2.3.5 on radwaste treatment and section 2.4 on the environmental ef fects of operation.

The offeet on in-plant doses will be discussed here.

The primary sources of radiation exposure to plant personnel are tritium in the heavy water, direct radiation from various components of the primary process system and the small amounts of radiation f rom the vicinity of the beam tube experimental facilities.

As new beam tube facilities are built and old ones modified, the l

radiation shielding is constantly being upgraded.

This upgrading will k

assure that the radiation dose to the users of the experimental facilities at 20 MW operation will not increase significantly above the very small exposures they presently receive.

Operating personnel do not receive significant exposures in the performance of their routine duties. When working in radiation areas or on radioactive or contaminated equipment, their exposures are carefully ta) 2-17

I limited by health physics monitoring procedures.

These procedures will, of course, continue to assure that all personnel exposures will remain within NRC regulations and be as low as reasonably achievable.

The increased reactor power will increase radiation levels in some maintenance situations, but in no case more than a factor of two above that which would exist at the present power level.

The effects of the increased levels will be mitigated by the use of increased shielding, careful decontamination, remote handling and shorter exposure times.

Furthermore, improvements and modifications to the primary system will make its maintenance (presently the major cause of operating personnel exposure) easier which will further compensate for the somewhat higher source strengths resulting from 20 W operation.

The reactor biological shield was designed originally for 20 W operation and is adequate to reduced radiation levels around the reactor to acceptable values.

Fuel transfer operations and the storage of spent-fuel will not be significantly affected by the power increase.

The fuel element being transferred is always either within the reactor. ielding itself or in transit through the heavily shielded subpile roou until it enters the canal leading to the storage pool. The handling within the pool is through many feet of water so no personnel exposure results from the fuel transfer procedure.

This will not change at the higher power.

The fuel element cutting, cask loading and packaging of non-fueled components all take place under water so no significant dose is received now and this won' t change at 20 W.

The waste handling has been covered in section 2.3.5 and will not result in any dose increases or other complications.

In summary, the radiological effects of increasing the reactor power on plant operations and personnel will be small.

Routine operating procedures will not have to be changed and maintenance procedures will only have to be changed to accommodate the 20 W modifications of the process system and to reduce exposure rates where appropriate.

In-plant personnel radiological doses will not increase significantly.

2.4 ENVIRONMENTAL EFFECTS OF OPERATION 2.4.1 SOURCES OF RADIOACTIVE RELEASES.

Releases of radioactive materials resulting from reactor operation are described in Section i

2.3.5, "Radwaste Treatment." As discussed there, the following typical releases may be anticipated at 20 MW.

M.r U U Tf 2-18 l

f Czssous Rniceses:

Argon-41 1400 Ci/yr Tritium 900 Ci/yr Liquid Releases to Sanitary Sewer:

Tritium 6.0 Ci/yr Other B-y

< 4.0 mci /yr 2.4.2 ENVIR0te! ENTAL EFFECTS.

These releases are spread almost uniformly through the year and so the concentrations in the atmosphere resulting from these releases can be averaged over the year. This also indicates the use of neutral dispersion coefficients in the atmospheric dispersion calculations.

These are the same as those used in Section 2.2.1.

The concentrations have been calculated for the point on the site boundary where they would be the largest based on proximity to the source (reactor stack) and prevailing winds.

In the case of liquid e f fl uents, the concentrations are diluted by the total Bureau sanitary sewer effluent and are given for the point where the Bureau system joins the county system.

The results for 20 IW operation are shown in Table H

2.4-1 which also includes the concentration at the stack and the MPC levels as determined from 10 CFR 20, Appendix B, Table II.

IS Table 2.4-1.

Concentrations of Radionuclides at Site Boundary (20 MW)

Caseous Releases at South-East Boundary (pCi/mL)

Stack Boundary 10CFR20(1977)

MPC

-6

-10

-8 Argon-41 3.2 x 10 1.6 x 10 4 x 10

-6 Tritium 2.3 x 10 9 x 10-2 x 10-Liquid Releases to Sanitary Sewer

]

-5

-1 Tritium 2 x 10 1 x 10

-9

-5 Other B-y 1 x 10 9 x 10 All of the above releases are well below 1% of MPC.

Furthermore, more than five years of reactor stack effluent sampling has revealed no

, n d

E evidence of the release of iodine 131 (instrumental sensitivity better

-13 than 10 Ci/mL). Therefore, the environmental effects of radioactive releases during normal reactor operation at 20 MW will be negligible.

y l

2-19 w

E l

l 2.5 ENVIR0to! ENTAL MONITORING PROGRAM 2.5.1 MONITORING METHODS. The environmental measurement and monitoring program at the NBSR includes a variety of sampling and analysis activities to detect any changes in environmental radioactivity levels and the radiation background as a result of NBSR operation.

Samples of soil, grass, and water are collected and analyzed for activity.

Measure-ments of the ambient radiation level at the site perimeter are also

)

made.

{

2.5.1.1 Soil and Grass Sampling.

Samples of soil from five (5) designated areas on the NBS site are collected not less frequently than quarterly, except during the growing season.

During the growing season, t

5.5 grass samples are collected instead.

Both soil and grase samples are radiochemically analyzed for mixed fission products following collection.

The sites from which soil and grass samples are collected are shown in Figure 2.10.

2.5.1.2 W.ater Sampling.

Samples of water in the vicinity of the NBS Reactor facility are collected monthly from four (4) surface streams and from ground water in three (3) residential wells.

These samples are analyzed for gross gamma activity and tritium content.

The location of these sampling sites are shown in Figure 2.11.

2.5.1.3 External, Background Monitoring.

The ambient background radiation level at the NBS site is measured by more than 50 thermoluminescent dosimeters (TLD's) placed around the site perimeter fence and on NBS buildings as shown in Figure 2.10.

Four control monitors are kept at locations 3 to 20 miles from NBS.

2.5.2

SUMMARY

OF RESULTS. The environmental measurement and mcaitoring program at the present NBS site was begun in 1963 when the 1

I nonthly collection of water samples from the surrounding area was initiated.

Soil and grass saepling and analysis at the site was started in 1965.

l Measurements of the ambient background were initiated in 1966.

In December 1967, the NBS Reactor went critical and in February 1969, the Reactor achieved 10 MW operation.

1 No significant changes in the activity levels present in the soil, grass, and water samples collected nor the external radiation background at t.ie site have been observed since the start of the environmental monitoring program. Minor fluctuations in levels were noted both before 2-20

&%,..~

i cnd of:cr ctmm:ncement of NBSR operation. These varied from month to month and year to year, but none of the variations could be correlated with reactor operation.

2.6 ENVIRONMENTAL IMPACTS OF PLANT ACCIDENTS 2.6.1 SMALL RELEASE OUTSIDE CONTAINMENT.

Regulatory Guide 4.2, g

Revision 2 (2.2), visualized releases of this type resulting from steam relief valves and other systems' handling radioactive material external to the reactor containment which might leak small amounts of radioactive materials. The NBSR has no such systems external to the confinement t

building and so no significant mechanism for this type of release exists.

This is substantiated by the fact that in ten years of operation of the NBSR, there have been no such releases.

2.6.2

_RADWASTE SYSTEM FAILURE 2.6.2.1 System Description. All liquid effluent from the controlled areas of the reactor building is drained into a 1,000 gallon retention tank.

When a predetermined level is reached, liquid in the 1,000 gallon tank is pumped into a 5,000 gallon hold-up tank for sampling and subsequent discharge to the Bureau's sanitary sewer system.

The line running from the reactor building to the 1000 gallon tank is monitored and trips an alarm if the liquid effluent from the reactor to the tank contains significant amounts of radioactivity.

The 5000 gallon tank is monitored by drawing a sample from the tank to a monitoring location in the reactor building where its analyzed for gross 8-y activity and tritium. The contents of the 5000 gallon tank is discharged to the sewer only after l

• the analysis has shown the activity to be at an acceptable level.

The two tanks are located in concrete vaults, approximately twenty-feet underground, outside of the building. These vaults are accessible and the tanks can be inspected visually.

2.6.2.2 Maximum Concentrations Expected. Typical annual liquid effluent releases expected at 20 MW from the NBSR are 6 Ci of tritium and less than 4 mci of gross 8-y activity.

During a year, the 5,000 1

t gallon hold-up tank is filled and discharged more than 20 times (20 daily discharges a year).

The average activity contained in each discharge, therefore, is an order of magnitude lower than the total for the six-month period. An upper limit on the concentration of the liquid discharged in any one day averaged over the day (i.e. the 2-21 L

l l

maxicum daily c:necntrcti n) c:n ba datcrmin:d to followa. A:Cuming cna half the annual total activity was contained in enn 5,000 g:11cn tcnk of l

water, the maximum !! concentration in the tank would then be about

-4 0.16 uC1/mL while that for S y activity would be 1 x 10 pCi/ml. When urther diluted by the Bureau's daily water discharge, the upper limit en

-4 3

the daily maximum concentrations are about 9 x 10 Ci/ml for 11 and

-6 3 x 10 C1/mi for S-y.

These concentrations are at least a factor of l

10 below allowabic daily limits. 'The use of the activities accumulated over a six-month period in a single release allows for any possibic fluctuations.

j 2.6.2.3 Accidental Release.

Because of the small activity and the l

low concentrations, listed in 2.6.2.2, the consequences of any spill or accidental release would be insignificant even if the entire 5,000 gallons in the hold-up tank were involved.

In case of damage to the hold-up tank, the spilled effluent would be contained in the underground vault and can be disposed of by either pumping into the sanitary sewer i

or into suitable containers.

Further, the concentrations are so small ao as to represent no hazard.

The accidental release of the entire contents of the hold-up tank into the sanitary sewer would result in concentration substantially below allowable daily limits due to the i

large dilution by the Bureau's daily effluent discharge.

2.6.3 FISSION PRODUCT RELEASE TO PRIMARY SYSTEM.

Normally, the fission product inventory of the primary system is negligible.

The only mechanism for significant fission products getting into the primary system is from a leaky fuel elcuent.

The NBSR is not normally operated with faulty fuci elements.

If a fission product release is detected in the primary system, normal operation is terminated and a proce dure is commenced to detect the faulty fuel element and subsequently remove it from the reactor. This has happened only once during the ten years of operation of the NBSR and the resulting release of fission products to the primary system was so small that it was difficult to be sure that a faulty fuel element existed. The small release into the primary system did not result in any measurable release to the environment.

The normal water g

treatment system for the primary system readily took care of the fission

]

products and once the fuel element had been removed, the primary water was casily cleaned up.

Since the fission product monitor in the primary 1 I 2-22 a _ed

~

cyctea c n r::dily d;tcet ficcica producto 1ccking from a fumi elem:nt and since the reactor can be quickly shutdown, large fission product releases to the primary system can be prevented.

Therefore, although small amounts of fission products may be introduced to the primary system from faulty fuel elements, these would not escape from the primary system to the environment.

Therefore, fission product releases of this type are not expected to have any significant impact on the environment nor in the work area.

Due to the conservative heat transfer design of the NBSR, even the

s most severe transients would not cause significant fuel element failure.

These severe transients are discussed in Section 3 of the SAR (2.15) submitted in support of the present application.

The "offdesign transients" appropriate for consideration in this report would not be severe enough j

to cause any fuel element failure and so would not result in release of fission products to the primary system.

I 2.6.4 PRIMARY TO SECONDARY LEAK.

Every precaution is taken to prevent the heavy water in the primary system from mixing with the light water in the secondary.

Any leak is quickly detected by a detector located in the secondary system which senses the N-16 activity in any

. 8 1

heavy water that might leak into the secondary.

If this detector alarms, I

l l

1 the secondary water is sampled for tritium.

D 0 levels in the primar,y 2

system are also checked.

If these checks confirm that a leak has developed, l

the reactor is shut down and steps are taken to isolate the heat exchanger.

If the N-16 monitor indicates an N-16 level very much higher than the f

alarm set point, the reactor is shut down immediately without waiting for additional confirmation.

The NBSR does not operate with faulty fuel elements. Consequently, there are no fission products in the primary water during normal operations.

If an element does develop a leak, it is quickly detected and appropriate action taken. Other than the very short lived N-16 activity, the only significant radioactivity in the primary system is tritium.

In addition to the N-16 monitor, a leak into the primary system can be detected by a change in the level of the D 0 storage tank and by periodic sampling 2

of the secondary water for tritium. The sensitivity of these methods are such that a leak of about 35 gal. in one day or 50 gal. in one week can be detected. Under either of these conditions, the reactor would be

~

shutdown and the leak corrected.

2-23 x

l As uming that tha tritium concentraticn ct 20 MW hto racched a level of 5 mci /mL (a very conservative number since it is planned to maintain the concentration below 2 mci /mL), the 35 gal. release in one

'l day would result in a maximum concentration of.luci/mL (%1MPC) in the I

sanitary sewer.

Based on a 100% release into the atmosphere, the maximum

~

concentration at the site boundary would be about 6 x 10 pCi/mL or q

about 3 MPC for that day. When averaged over a year, the concentrations jd would be less than 1/500 MFC including changing wind conditions.

j The 50 gal. leak in one week would, of course, give lower daily concentrations than those above but would give a slightly higher concentration of about 1/350 MPC at the site boundary when averaged over a year.

These leaks are of such a magnitude that they can be easily detected and the faulty tube in the heat exchanger located and plugged.

It is conceivable, however, that a leak might be so small that it could not be located in the heat exchanger. The rate would have to be less than 1/2 gal./ day. A leak of this size could still be detected through the tritium sampling of the secondary water, but it might not be possible to locate it in order to repair it.

It is very unlikely that any such Icak would remain small for a long period of time.

If, however, it should not be possible to locate and repair it for a whole year,180 gal, of D 0 would be released to the secondary during the year.

This would give 2

an average concentration at the site boundary of no more than 1.7 x 10 pCi/mL or less than 1/100 MPC.

These calculations are based on a tritium level of 5 mci /mL in the primary system.

It is anticipated that the tritium concentration will be maintained at a much lower level which would result in much smaller releases in the event of a leak.

The only releases directly from the i

}

primary system to the environment result from the unusual occurrence of a leak in the heat exchanger and even then releases would be only a very i

small fraction of MPC.

I 2.6.5 REFUELING ACCIDENTS. The top shielding plug of the reactor i

is never removed while there is fuel in the core.

Thus, it is not credible to assume that a heavy object can fall on the core.

The fuel is moved within the core or removed from the core by hand operated pickup tools built into the top plug. The elements, weighing about 25 lbs. each, are moved one at a time.

Even if one should fall from the di 2.

i trcn3fcr mechanica, it is n:t h2cvy enough to drmag2 th2 core which ic licItcd und:r a h2cvy tcp grid picts.

Sincn the cisment would only drcp a few feet at most, it would not be very seriously damaged.

Since the fueled plates are completely enclosed by unfueled side plates, there would be no damage to the fueled plates and no release of fission products. The most severe consequence would be some damage to the end fittings or unfueled side plates which would render the element unfit for future use.

I 2.6.6 SPENT-FUEL HANDLING ACCIDENT 2.6.6.1 Fuel Element Drop-In Pool.

The NBSR fuel element weighs only 25 lbs. in air and about 16 lbs. in water.

Thus, if it were dropped l

in the pool during handling, it would not fall very hard and at most would dent the end fittings or side plates (un-fueled). There would be no release of fission products and no personnel exposure.

2.6.6.2 Heavy Objeu,: Drop Onto Fuel Rack.

Following the guidelines set forth in Regulatory Guide 4.2, the following assumptions are made:

(a) The void activity (1% of total noble gas and halogen activity) of a typical fuel element is released.

(b) Thirty-day decay time.

[

(c)

Iodine decontamination factor in water is 500.

(d) Charcoal filter efficiency for iodines is 99%.

Following the recommendations of AEC Regulatory Guide 1.3, Revision 1, for the appropriate dispersion model and atmospheric conditions and including the above assumptions, the following doses are calculated for a person standing at the nearest site boundary (400 m) for 30 days or more:

Whole Body Dose

-6 Direct and Sky Shin 4.2 x 10 rad From Cloud 4.7 x 10-rad Thyroid Dose 3.8 x 10-rem These extremely small doses clearly present no threat to the environment i

or the general public.

2.6.6.3 Fuel Cask Drop.

It is highly unlikely that the fuel cask for shipping the NBSR elements could be dropped in such a way as to injure the contained elements when the cask is not within the secled confinement building. When the truck dcor is open (no containment), the transfer cask is only lif'.ed above the floor enough to place it on the (19

~

j L--

2-25

~ ~ ~ ~

truck. A drop from this elevation would not injuro the contents of th2 i

cask. Nevertheless, it is assumed th'at such an accident does occur releasing some fission product gases.

The assumptions are:

(a) Ground level puff release external but adjacent to reactor

{

building.

(b) 1% of total nobic gas activity in fully loaded cask is released.*

(c) Fuel cooling perfed is 120 days.*

(d) Procedures for calculating dose are those given in ANSI /ANS 15.7-1977, "Research Re~.ctor Site Evaluation," and Regulatory Guide 1.4, Revision 2, 1974.

The exposure is then given by

• h

\\ ) Ci-s I

Q y

I " uno o exP 2a 20,

,3 The total dose a person would receive from exposure is the same that he would get from standing in a semi-infinite cloud with a concentration numerically equal to $ for one second.

It is assumed the individual is directly down wind from the ground release at the site boundary, 400 m l

away. Thus'both y and h are zero and the equation becomes:

$ = no o u

yg where Q equals the total number of curies released in the puff.

For the 85 conditions cited above, Q = 120 Ci of noble gases.

(Essentially Kr).

-2 This yields a dose of 2.3 x 10 rad following Regulatory Guide 4.1.

In view of the conservative nature of the calculation (Pasquill condition F, u= 1m/s), and the incredible nature of the assumed release, this small dose does not represent a threat to the public.

2.6.7 ACCIDENT INITIATION EVENTS CONSIDERED IN DBA.

The accident which could initiate the DBA is the flow blockage of a fuel element.

Other than the DBA itself, this would not lead to any release of radioactivity

~

1 outside the primary system. Other accidents considered involve the rupture of primary system components.

Since all primary system components are located completely within the confinement building, no rupture would release primary water directly to the exterior of the building with the exception of leaks in the heat exchangers which would not initiate a DBA I

and are treated in Section 2.6.4.

Thus, the DBA associated accidents to

• Based on Regulatory Guide 4.2, Rev. 1 (Jan. 1975).

2-26

b] consider:d cra licit d to tha cpilling cf primary cyctem water into the containment building. Th::a are diccu;ced in NBSR 9. Addendum 1

}l (2.15) and shown not to result in significant doses to the general public.

)

2.7 UNAVOIDABLE EFFECTS OF FACILITY CONSTRUCTION AND OPERATION As discussed in Section 2.3.3.2, no additional major construction l

is required to operate the NBSR at 20 MW.

The radiological and environmental ef fects of operating the NBSR at the increased power have been discussed in previous sections. The 20 MW operation will result in an increase use of several resources.

Electricity 6

0 use will increase from 3.2 x 10 KWH/yr to a total of 5.4 x 10 KWil/yr at 20 MW operation. Water consumption in the cooling towers will double to 4 x 10 gal /yr and U-235 burnup will double to 6.4 kg/yr. Although the increase in electricity and water use is not completely negligible, it is still small compared with t.ne total use of the whole laboratory.

ii.e %.

it will not present any special supply or environmental p roblems. The 3.2 kg/yr increase in U-235 use is down by two orders of magnitude from the increase resulting from the addition of just one full size nucicar power plant to the U.S. grid.

Therefore, the increased use of U-235 at 20 MW is a negligible additional burden on U.S. resources.

In suramary, the unavoidable ef fects of facility construction and operation are small and will have no significant adverse effects on the environment.

2_. 8 ALTERNATIVES TO UPGRADE AND OPERATION OF THE FACILITY, There is only one alternative available to provide the increased research reactor sersices needed by NBS. That would be to place the additional Bureau work and that needing higher flux at some other existing research reactor. There are, however, only two other laboratories in the country that could provide experimental facilities and neutron fluxes comparable to those availabic at a 20 MW NBSR. These are Brookhaven National Lab and Oak Ridge National Lab.

They are both far away from NBS and do not have many of the specialized facilities available at the NBSR.

Furthermore, they are both fully utilized now and could not take on an additional work load of the magnitude of that anticipated by NBS.

Therefore, this is not a viable alternative.

If the NBSR power is not

.. y increased, the extensive NBS needs for increased reactor service will not be met.

m) 2-2,

i

,J 2.9 LONG-TERM EFFECTS OF FACILITY CONSTRUCTION AND OPERATION t

The cost of upgrading the physic ~al plant to permit operating the NBSR at 20 W is small because the reactor was originally designed for 20 W and the few major items that were not originally sizad for 20 W have been added over the last ten years of plant operation. Therefore, if at a future date, it should be desirable to reduce the reactor power back to 10 MW, the capital expenditure put in at this time to achieve 20 MW would not be a significant itretrievable commitment.

The radiological effects of 20 MW operation have already been dis-cussed and should be no different in the long term than in the short term.

The long-term beneficial effects of increasing the NBSR power are extensive. The NBSR is used not only by the Bureau, but also by more than thirty other government and private institutions.

The areas of application include materials science, non-destructive evaluation, trace analysis, isotope production, neutron standards and dosimetry, nuclear physics and quantum metrology. Some specific examples where the power increase will be particularly valuable include:

the study of the micro-n scopic behavior of hydrogen in metals; the structure of polymers, biomolecules, and other macromolecules; properties of new amorphous magnetic materials, trace analynis of solar cell materials; and the establishment of high-flux standard neutron fields for the calibration of reactor pressure vessel dosimeters.

In summary, the long-term detrimental effects of facility construction and operation are insignificant, whereas the benefits to be realized are great.

2.10 COST AND BENEFITS OF FACILITY The NBSR was originally designed for 20 MW operation.

The long range plan called for initial operation at 10 MW while operating experience was being obtained and user programs were developed. Thus, the reactor was built with all the difficult-to-add features for 20 MW built-in and only minor modifications remain to be made.

The increased operating costs are estimated to be $400,000 per year in FY 80 dollars compared to the present operating cost of.spproximately

$2,000,000.

The benefits are much more difficult to quantify since the dollar figure cannot be placed on the value of scientific research.

Therefore,

)

2-28 1

th; cxisting cnd pstcntici progrcms will b2 cummarized to provida a qualitative basis for judging the benefits of doubling the reactor power.

The NBSR is a national resource used by 200 engineers and scientists.

It is used by 18 other federal agencies and 25 universities and industrial laboratories as well as by many Divisions and Offices of the National Bureau of Standards.

In spite of the large number of outside users, about 75% of the use and funding comes from NBS. Most of the other financial support comes from other federal agencies. More major experimental facilities are installed at the NBSR than at any other research reactor f

in the U. S.

These instruments are highly automated to take advantage of the around-the-clock, seven day week operation of the reactor.

Clearly, it is a very heavily utilized facility.

l The major research areas include:

e Materials Characterization e

Trace Analysis e

Non-Destructive Evaluation e

Neutron Dosimetry e

Radiation Standards e

Isotope Production l

l It is not possible to describe all the programs here. More detail may be found in the most recent annual (2.14) report of NBSR activities--

l l

a 200 page document summarizing all the programs.

One or two examples l

will be given here from each area to indicate the benefits of the program.

The work in the area of materials characterization is largely I

I fundamental research in which neutrons are used as a probe to determine the atomic structure of materials and to study the forces holding the

/

atoms together. The detailed information gained in this way contribute to our knowledge of how hydrogen causes steel and other structural l

metals to become brittle and fail.

It contributes to our understanding of how catalysts work to accelerate chemical reactions. And work done at the NBSR by scientists from NIH has determined the complex structure l

of a protein and shown how a vital step in the digestive process is L

accomplished.

I I

' I Neutron activation analysis is a powerful technique for identifying trace impurities in many materials.

It is used by NBS to characterize i

standard reference materials (SRM) used for calibration references in 2-29 W+n' es c.%,

e3 e de

-m-4Ag%s-Wm-W-s-wh4 t--*Mh=4

'-NNNM M-m-

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=

ev

IE medicci clinicci 1cboratories, etcal production, cgriculture, environrental pollution studies, and many other are*as.

The technique is used to measure impurities that decrease the efficiency of solar cells and to l

study environmental pollution in the waters of Chesapeake Bay and the

{

air of Washington, D. C.

The Food and Drug Administration uses the NBSR to measure impurities in foods and drugs.

For example, many of the measurements on mercury in fish and other foods were made by the FDA using the NBSR.

The U. S. Geological Survey uses this techniquc to examina minerala in their search for new energy resources.

These are only a few exampics of the many projects using the NBSR for trace analyals.

Fautron radiography is a powerful non-destructive evaluation method.

Neutron radiography is similar to x-ray radiography but can image different materials.

Neutrons can penetrate lead to "see" plastic materials for I

example.

It has been used by NBS to study pacemaker batteries, ancient chinese artifacts in collaboration with Smithsonian Institute, rare paintings in collaboration with the National Callery of Art, jet engines, and many other items. The Bureau is developing standards for the neutron radiography industry.

The reactor is used to develop and calibrate neutron dosimeters for Da monitoring personnel radiation exposurca.

The program also includes developing methods to predict radiation doses to structural materials such as reactor pressure vessel walls to assure the accurate prediction of safe lifetimes for materials exposed to radiation.

Standard radiation ficids are maintained in the NBSR. These form the basis for a nationwide system of radiation measurement.

The fields serve as national standards for the calibration of radiation measuring equipment used to determine the strength and character of a wide range of radiation ficids.

s Although NBS does not compete with private industry in the production of radioisotopes that are commercially available, there are many isotopes that decay too rapidly for commercial production or have too specialized and limited use to warrent commercial production. An example of such an isotope is flourine-18.

It is used by hospitals to locate malignant bone tumors.

Because flourine-18 decays by a factor of two in only two hours, it must be produced near where it is used.

About 3000 patients l

in the Washington area have had bone scans using flourine-18 prepared at l

the NBSR.

l.

)

2-30

Th2 cbsva cummary chowa th2 cxtensiva une end many binafits resulting f rom the operation of the NBSR.

Thes'e programs have grown rapidly and now the demand for use of the experimental facilities exceeds their capacity. Many of the facilities have two or three month backlogs and there are now many complex experiments that require more neutron intensity than is availabic at 10 W.

Doubling the reactor power to 20 W will provide the increased intensity needed for the more sophisticated experiments and greatly reduce the waiting time for access to the experimental facilities. Experiments can be done in half the time at the higher 1

power. 'Ihus, for only a 20% increase in operating cost, the capacity of the facility and the extent of the research programs can be nearly doubled. The cost of doubling the reactor power and the resulting benefits have been carefully reviewed by the Office of Management and Budget and approved by Congress.

2.11 CONCLUSION Increasing the power of the NBSR from 10 W to 20 W will have no significant adverse impact on the environment or p:pulation.

This is based on:

(1) During the construction period and the past 10 years of operation at 10 W, there has been no detectable adverse impact on the environment based on environmental monitoring since before the reactor was operational.

l (2) The modifications necessary to double the reactor power are small.

(3) The radiological releases resulting from routine operation result in concentrations in the nearest unrestricted areas no greater LR -

than two orders of magnitude below MPC.

(4) The effects of any of the many postulated accidents analyzed earlier in this section do not result in exceeding MPC in populated areas.

(5) The benefits from the proposed power increase are great.

Productivity should increase by 50%, the current log jam of experiments will be eased, and more sophisticated experiments can be undertaken.

D 2-31 alb

^

R2ferencts (2.1)

Draf t Guidelines for Sttndard Format and Content of Environmental M

Reports for Research Reactors.

l (2.2)

Preparation of Environmental Reports for Nuclear Power Stations, Reg. Guide 4.2, Rev. 2 (July 1976).

(

(2.3)

" Final Safety Analysis Report on the NBS Reactor, NBSR 9" (April 1966).

(2.4)

D. B. Turnar, " Workbook of Atmospheric Dispersion Estimates,"

PHS-999-AP-26 (1967).

(2.5)

Cooperative Forecasting - Summary Report; 1976, J. C. McClaire, Jr., Metropolitan Washington Council of Governments, Washington, DC 20036.

(2.6)

Population, Household and Employment Growth Forecast, 1974-1984, National-Capital Park and Planning Commission, Silver Spring, MD.

(2.7)

Montgomery County Census Update Survey Summary Report - 1977, National-Capital Park and Planning Commission, Silver Spring, MD.

(2.8)

Housing Survey,1976, City of Gaithersburg, Planning Department, Gaithersburg, MD.

(2.9)

Planning, Staging and Regulating, Fif th Growth Policy Report of the Montgomery County Pinnning Board, the Maryland-National Capital Park and Planning Commission, Silver Spring, MD, (June l

1979).

(2.10)

Long Range Forecast, People, Jobs and Housing, The Montgomery County Planning Board, Maryland-National Capital Park and Planning Commission, Silver Spring, MD, (September 1979).

(2.11)

Fifth Growth Policy Report Summary, Planning, Staging and Regulating, Maryland-National Capital Park and Planning Commission, Silver Spring, MD, (September 1979).

(2.12)

Montgomery County Census Update Survey, Summary Report, Montgomery County Planning Board, Maryland-National Capital Park and l

Planning Commission, Silver Spring, MD, (September 1979).

(2.13)

The Maryland Weekly, The Washington Post, (June 12, 1980).

(2.14)

NBS Reactor:

Summary of Activities July 1978 to June 1.979, NBS TN-1117.

(2.15)

" Final Safety Analysis Report on the NBS Reactor, NBSR 9, Addendum 1,"

(November 1980).

2-32

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