ML20244A454

Forwards Response to NRC 890330 Communication Re Change in Operating Power & Revised Tech Specs,For Review & Approval. State of Oh Ofc of Energy Mgt Consulted W/To Formulate Answers to Questions

ML20244A454
Person / Time
Site:	Ohio State University
Issue date:	05/26/1989
From:	Redmond R OHIO STATE UNIV., COLUMBUS, OH
To:	Michaels T Office of Nuclear Reactor Regulation
Shared Package
ML20244A456	List: ML20244A454 ML20244A737
References
NUDOCS 8906120024
Download: ML20244A454 (78)

Text

[ {

r

< t

~

T H~E Engineering Experiment Station 142 Hitchcock Hall

[ 2070Neil Avenue Colum'ous, OH 4321o-1275

~

. Phone 614-292-2411-UNIVERSTTY May 26, 1989 Theodore S. Michaels, Project Manager Standardization and Non-Power Reactor Project Directorate Division of Reactor Projects -.III', IV, V and Special projects Office _of Nuclear Reactor Regulation USNRC Washington DC 20555'

Dear Mr. Michaels:

. The staff _of The Ohio State University Research Reactor, License R-75, Docket No.

50-150, has'recently completed the response to your communication of March 30, 1988 entitled " Questions Regarding Change in Operating Power". In formulating answers-to these questions the staff has consulted with the Ohio State Office of Energy

- Management,-the Office of Radiation ~ Safety, numerous other research reactors, and an engineering consultant, among others. The questions and answers hav'e also been

. reviewed by the Reactor Operations Committee and the answers were approved during

. their May 22, 1989 meeting. The Reactor staff, members of the Rea*: tor Operations Committee, the Office of: Radiation Safety, and the University Administration,are committed-to a. safe reactor and the ALARA concept. In this respect, tne answers

. to the above referenced operating power change questions and the revised Technical Specifications are submitted for further review and approval.

Please send any official communications concerning the status of this submission to my attention at this office, with copies to Dr. Don Miller, Director of the Nuclear Reactor Laboratory, Mr. Joseph Talnagi, Senior Research Associate and Mr.

Richard Myser, Associate Director of the Nuclear Reactor Laboratory. The address of the Nuclear Reactor Laboratory is 1298 Kinnear Road, Columbus, OH 43212 (phone:

614/292-6755). Questions on the technical content of the submittal should be

-directed to Mr. Talnagi.

T h ank you for your attention to this matter.

Sincerely,

& $&f,0 "

l l Robert F. Redmond Director RFR/sh i enclosures c: D. Miller 1

[ I J. Talnagi R. Myser College of Engineering -

Answers to NRC Questions Concerning 500 KW Operation 1.1 The perimeter surrounding the OSURR through which access is controlled is the Reactor Building. There is a " security" fence shown in Figure 2.2 p. 6 of the SAR which defines the " exclusion area" for the OSURR. In normal circumstances the fence serves to limit casual pedestrian traffic around the Reactor Buf1 ding, although owing to its rather isolated location this is'already very small.

The closest distance to the exclusion fence is 20 feet to the West, 40 feet to the North, and 36 feet to the East. In an emergency situation both vehicular and pedestrian traffic could also be isolated from the southerly direction.

1.2 During normal operation of the reactor, obviously depending upon the type of reactor operation and/or experiment in progress, people are allowed complete access within the restricted area. This access is unescorted only after individuals have completed 10CFR19 training through the OSU Office of Radiation Safety (ORS) and at the OSURR.

Areas that are typically controlled to limit access are the reactor pool top area (Attachment A, Figure 2) and the NE corner of the " Bay Area" (Attachment A. Figure 1) where most experimental activity takes place. If not involved in authorized experimental activities, people are expected to be in their offices or in the control room. Normal staffing is four FTE's. Distances to the " reactor" were determined in the following manner. The distances were estimated from the doorway of each office / classroom to the core of the reactor. These distances include all shielding and are found in Figures 1 and 2.

1.3 The distance between the reactor facility and the closest permanent residence is about 1200 feet as shown in Figure 2.1 of the SAR. This is to the West of the facility. In the direction of the prevailing winds (South to North Table 2.6 p. 21 of the SAR) the distance between the reactor facility and the closest permanent residence is I approximately one mile (Figure 2.3 of the SAR).

2.0 Figures 3 and 4 of Attachment A provide architectural drawings of the Reactor Building HVAC systems.

2.1 During normal operations air is evacuated through the top of the building by the exhaust fan placing a small partial vacuum on the building. Air may then enter the building through various openings (mail slot, cracks under doors, or open windows). Air flow is inward due to the small partial vacuum from the building exhaust. The fs I building internal air flow takes a suction on the bay area through return air grilles and this air is heated or cooled and distributed to all rooms.

2.2 With all circulation stopped (air conditioner off, recirculation fans off, and exhaust fan off) and the windows closed there was no measurable air flow into or out of the building. This was confirmed by measurements using a Fisher Scientific anemometer and an Alnor thermo-anemometer which indicated no flow.

1 l

L - . - _ - . _ _ _ - . _ _ _ _

2.3 From the enclosed schematic one may note that there is a return air line from the reactor bay area to each furnace. These are located i outside of Furnace Room 101 and outside of Room 104. The return line also serves the air conditioner located in Room 101. Thus, it appears that during normal operations of the ventilation system, air from the reactor bay is distributed throughout the building. Credit was taken for this when doing dilution calculations when we assumed a 70,000 cu. ft. volume for the building.

2.4 There is a single switch in the control room that will turn off all fans, air conditioners, and furnace blowers in the building. This enables the reactor operator to immediately isolate the accidental release of radioactivity from the reactor building to the unrestricted atmosphere upon warning from the alarmt, associated with the building gaseous effluent monitor and the rabbit vent system.

This action is a part of the response outlined in our procedure for nuclear emergencies. It effectively isolates the air in the building. The louvers on the building exhaust fan close as well as the louvers on the furnace in Room 101. There are no louvers on the furnace vent in Room 102. There are also air ducts open to the outside for both furnaces that allow 10% outside to mix with building air during heating and air conditioning. These air ducts do not isolate automatically but can be manually isolated. Also, the flue for the hot water heater in Room 102 does not isolate but the hot water tank can be turned off eliminating the rising hot air tnrough this pathway. Both furnace rooms also have doors that can be closed to further isolate air in the reactor building.

The systems and procedures currently existing provide appropriate isolation of accidental radioactivity releases. Our approved Emergency Plan, based in part on ANSI /ANS 1.56, identifies the Emergency Planning Zone (EPZ) as the Reactor Building. This is deemed appropriate for Research Reactors authorized to operate at 2 MW or less.

2.5 There is no air filtration system for the OSURR.

3.1 The most likely source of release of "potentially" contaminated liquid is the reactor pool water. The design of the building drain system is such that there is no hold up tank or catch basin for 11guld releases from the reactor building. In the event of a release of pool water we would depend on warnings from the ARM system, in particular the one near the process system, for a possible indication. Area surveys which are done each day of reactor operation would also alert OSURR personnel that a contaminated spill had occurred. Smear wipes are completed on at least a weekly basis.

If there has been a spill, we follow approved decontamination procedures for clean up. If pool water is spilled we can evaluate the radioactivity from the previous quarterly water sample analysis or one taken immediately after the spill is detected.

2

[l 3.2 1. There will be some small heating of the lead shielding contained within the pool walls. Appendix A of this document provides an estimation of the amount of heating and shows that the effect is not significant even under the most pessimistic assumptions.

2. Fast neutron dose will induce minor dimensional changes in the graphite contained within the thermal column extensions. Appendix B of this document estimates the amount of swelling that might occur in this graphite, using conservative assumptions for operating history, and ,

shows that the effect is very small, and, even should it occur, has no l severe safety implications.

3. Appendix C of this document discusses the effects of gamma and neutron radiation on the beam port gaskets. There are no other. materials in the high flux regions of the OSURR that would be expected to show  ;

degradation effects of radiation exposure over the life of the '

facility.

3.3 Figure 3.14 on page 55 of the Safety Analysis Report shows that the water processing system does have check valves at both the Reactor and BSF pool  ;

loop inlets. Prior (i.e., " upstream") from these points is a manually-  ;

operated valve and the. mixed bed resin exchange demineralized which serves i as the initial demineralized for makeup water to both loops. Normally, the manually-operated valve is closed except during makeup water inlet

)

operations (which are done infrequently, depending on ambient conditions),

which provides an additional barrier to reactor system water backflowing into the makeup water supply. Further, the makeup water demineralized has ,

considerable pressure drop across its active volume, making backflow of water from the reactor pool difficult, if not impossible, given the relatively low pressures generated by the static pressure head of the j reactor pool. There is reasonable assurance that the check valves are  !

quite leak tight. There have been many occasions during the 28-year l operating history of the OSURR where the city water pressure has dropped j to very low pressures (i.e., during planned and unplanned water outages) and no evidence of the pool draining back down through the process system {

has ever been detected.

On the cooling system loops (Figure 3.13, page 47 of the Safety Analysis j I

Report), the only connection of the cooling system to the city water supply is at the inlet to the secondary side of the auxiliary (or secondary) heat exchanger (bottom center of the Figure). Leakage of secondary coolant (which is not circulated around the reactor core) into the city water supply is unlikely since a liquid-to-liquid leak would have  ;

to occur in the auxiliary heat exchanger, and for plate-type heat i exchangers the most likely leakage pathway would be from the plates to the  !

floor of the building, not between primary and secondary sides of the exchanger. Even if such liquid-to-liquid leakage were to occur, reverse flow is prevented by a Watts model 909-S Reduced Pressure Backflow Preventer, which is a device substantially more reliable than any simple check valve. The vendor of this device has asserted that complete failure of this backflow preventer is extremely unlikely. Further, the drain to the city sewer system at the outlet of the auxiliary heat exchanger's 3  ;

i f

F U

p secondary side would have to be totally blocked to allow substantial K pressures to build up and cause reverse flow back into the city water supply. Such blockage is unlikely since there are no valves or other-

. constricting devices installed.at this point.

Leakage of reactor pool water from the primary coolant loop into the city water supply is even less likely since, in addition to the conditions

' listed above, a liquid-to-liquid leak of reactor pool water to secondary F loop coolant would have to' occur in the primary heat exchanger. Again, such a leakage pathway is'unlikely'for plate ' type heat exchangers.

3.4 Under normal operating conditions, the pressure is higher on the secondary side of the primary heat exchanger, so any liquid-to-liquid leakage would be of secondary loop coolant.into the reactor pool. Again, liquid-to-liquid leakage'in plate-type heat exchangers is unlikely. However, there are various abnormal conditions which could reverse this pressure

' differential, so leakage in both directions should.be considered, even though liquid-to-liquid leakage is unlikely in these heat exchangers.

Leakage of secondary loop coolant into the reactor pool would result in, contamination of the reactor pool water. The secondary coolant is a mixture of ethylene glycol and water, which could add contaminants to the reactor pool water which could result in higher concentrations of activation products in the pool. In this event, upon detection, the pool water would be assayed for radioisotope content, and appropriate measures taken to clean up or replace-the pool water, and correct the leakage problem. It is also possible that addition of secondary coolant to.the reactor pool could induce reactivity changes. However, the secondary coolant has additives which assure that if'such leakage occurs the possible reactivity effects will be negative (i.e., cause a decrease in reactor power if the reactor were critical at the time of the leakage).

Loss of secondary coolant would cause a decrease in the heat removal

-capacity of the cooling system. However, monitoring instrumentation such as flow rates, pressure drops, and coolant temperature would indicate the effects of such leakage.

Leakage of primary coolant (reactor pool water) into the secondary coolant loop would cause changes in the heat removal properties of the secondary coolant. Addition of activation products to the secondary cooling loop could also occur if the reactor were in operation at the time. The 1 secondary cooling loop is a closed system, so the only way that an offsite release could occur would be if primary-to-secondary leakage were to occur concurrent with a breach of the secondary loop pressure boundary, either at the fan-cooled heat exchanger outside the building, or across the auxiliary heat exchanger to the city water, and thence to the drain. If i

-loss of reactor pool water were great enough, a reactor trip would be ]

generated by the low pool water level trip function. ]

There is no " dedicated" leak detection system (i.e., a system whose sole function is leak detection). However, there are several ways which leaks in various parts of the cooling system could be detected. The low reactor pool level trip discussed in the paragraph above is one such way. The i heat exchangers have locally-indicating pressure gauges installed which indicate pressure drops across each heat exchanger. Flow rates in the 4

_ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____ l

primary, secondary, and auxiliary cooling loops are monitored in the control room. Temperatures at various points in the primary and secondary loops are also monitored in the control room. Leakage of coolant would likely manifest itself in abnormal readings on these monitoring channels.

3.5 First, it should be noted that pre-acknowledgement of the audible alarm annunciator has been a feature of the OSURR safety system since the original tube-based annunciator system was replaced by solid-state components. The first system update was in 1965, and the latest update was in 1986. In both cases, proposed design' changes were reviewed and approved by the Reactor Operations Committee (ROC), and deemed not to constitute an unreviewed safety question. Alarm pre-acknowledgement in no way compromises the operation of the safety system or any related system.

Reactor trip functions are unaffected by pre-acknowledgement of any-alarms.

All possible alarms may be pre-acknowledged. This does not mean that the trip function is defeated. The reactor trips still function, and visual indications are still provided in the annunciator panel. Only the audible annunciator is silenced. Thus, in terms of reactor safety, there is no safety significance to pre-acknowledgement of alarms. The operator on duty at the console is authorized to pre-acknowledge alarms. No' written procedures exist concerning alarm pre-acknowledgement. There are no restrictions to alarm pre-acknowledgement. It is our belief that there are no implications with regard to OSURR Technical Specifications of alarm pre-acknowledgement.

3.6 The radiation monitors available at the NRL to be used in the event of an accident are listed in section 7.3.5 of the SAR. There are two low range G-M type instruments for normal operations and a high range air Ionization chamber (0-199.9 R/hr) for " accident" conditions. We also have an air ionization instrument not listed in the SAR that ranges up to 1000 R/hr. The Office of Radiation Safety has a Keithley Model 3612100 (20 R/hr) Victoreen 450 (50 R/hr) and an additional Bonner Sphere for neutrons good up to 5 rems /hr. It is our belief that these instrument are suitable for all postulated normal and abnormal operating conditions.

3.7 There is no credible means for the city water supply to become contaminated since no experiments are cooled with city water nor plumbed into the system. An experiment utaljzing city water as a coolant would have to be approved by the Reactor Operations Committee before implementation. A part of the review would be radiological control. I Use of city water for experiments, as noted on page 123, infers use of water in ways such as a solvent for liquid samples to be irradiated (e.g., l for chemical standards to be used in an activation experiment), or {

dissolving and/or dilution of a radioactive source produced in the reactor. As noted above, such uses do not involve plumbing of city water into any experiment or facility undergoing irradiation I

i l

I l

5 I

t f

3.8 There are two demineralized systems in use. The one described on pages 125 and 156 is a cartridge ion exchanger that is used to remove ions from the reactor pool water on a daily basis. The second is a mixed bed resin exchanger that is used to clean up city water before it is used as "make up" for either the reactor or bulk shielding pool. Both of these are indicated in Figure 3.14 p. 55 of the SAR.

We regenerate the mixed bed resin exchanger ourselves. The cartridge ion exchanger is either disposed of as radioactive waste or sent back to the manufacturer for regeneration of the resins. The decision is based on whether we can detect any radioactivity in the resins counted on a high-resolution Ge(Li)-based gamma-ray spectrometer.

Since Na-24 is produced it may be detected in the resins if a sample is counted soon after removal from the process system. However, with a 15 hour1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> half-life, Na-24 does decay to " negligible" levels after several months. Assuming we held the resin for only two months, the decay of Na-24 would be about 96 half lives. As discussed on page 156, the OSURR staff is aware of the potential for buildup of longer-lived isotopes and will deal with this situation appropriately.

3.9 If an experiment or sample requires encapsulation for safety reasons, then this is done. Encapsulation "where necessary" is the criteria used. Any experiment where encapsulation for safety reasons is a question reviewed by the ROC if it has not been previously approved.

This is covered in Section 6.2.4(4) of the Technical Specifications.

In addition, Item II of the Table of Request for Reactor Operations Approvals (Administrative Procedure-04) requires the SRO to assure that sample containment is sufficient to assure the integrity of the reactor system and contain the radioactivity produced. The requirement is in Technical Specification 6.4.2, Approved Experiments, which requires an evaluation, when pertinent, of experimental integrity, physical or chemical interactions, and radiat!on hazards.

The "where possible" statement referred to on page-158 simply recognizes that for some experiments encapsulation is not necessary and/or possible.

Again, as stated above, if safety concerns dictate, encapsulation is done.

However, there are many experiments where encapsulation is not desirable and/or possible and/or necessary. For example, testing of a self-contained ion chamber (e.g., a CIC, or fission counter) does not require encapsulation of the entire device for safety concerns, nor is it desirable to even attempt such encapsulatiu .

i 3.10 The references listed on page 159 were those used as cited in the Report.

These editions are those that happen to be available to the NRL staff. We affirm our belief that the references cited are suitable and adequate for the purposes for which they were used. It is our belief that accuracy, not age, should be the determining factor in whether one uses a particular piece of information. We believe that factual information must stand or fall on its own merits, not necessarily its age or its relationship to so-i called up-to-date references. i 6 l

e ey i

4

~

4'.1. Over _ the last five years (1983-1987) the OSURH has averaged 184 effective full. power hours'(EFPH)-. It is anticipated that this.could-double.- A conservative,irealistic estimate for' future use is 400 EFPH.

. 4.2. Examination _of the operational records over the last five years (1983-1987) indicates the following percent of the yearly EFPH for each experimental fscility listed in' Table 6.3 of the SAR.

Facility

• EFPH I

Description Yeerly CIF 32.4%

Beam Port 1 25.4%

Beam Port'2 < 1.0%

Rabbit 4 2%

Thermal Column 9.0%

4" Dry tube 1.0%

2" Dry tube 1.5%

J 4.3 In principle, there is nothing that would preclude " simultaneous i.

operation" of all of the facilities, with the exception that.the 2" and 4"

! dry tubes would likely not be used simultaneously, since there is only one mounting bracket for the dry tubes. However, one must be careful in interpreting exactly what " simultaneous operation"_ implies. For example, it is possible to have an experiment (e.g., activation samples) loaded-into the CIF, while at the same time having another experiment (e.g.,.n a This does not mean, however,

.lonization chamber) loaded into Beam Port 1.

that both facilities, or either one, will be releasing any Ar-41, either continuously or as a puff release. Continuous release is not likely since both facilities would be sealed during operation, and, even if they were not, there would be no strong driving force to release the air entrapped withiri each facility. Puff releases would require opening and physically removing apparatus from the facility which would force a puff release of the entrapped air. Such operations are not likely until some time after shutdown of the reactor, during which time the Ar-41 in the facility will b 7 L ~

L have decayed, and Ar-41 levels in the building air will have been reduced by purging of the reactor building air by the building exhaust fan.

Based on previous operating history, the most likely " simultaneous operation" of experimental facilities would be operation of the Rabbit while another experiment is being irradiated in one of the other facilities which do not have continuous air purging. For example, Rabbit operation during a reactor run with experiments loaded into the CIP is done occasionally. There have been experl7en'ts performed with samples being irradiated in the CIF, and a series ef Rabbit operations, each with a 5 minute irradiation time, conducted simultaneously, over the entire time the samples in the CIF were irradiated. The total duration of this run was 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

4.4 As analyzed in Chapter 6 of the Safety Analysis Report, Ar-41 can be produced in a number of ways. Section 6.1.2.2 (pages 132-134) discusses the production of Ar-41 from air dissolved in the water of the reactor pool and subsequently released to the building air. This can be considered as a continuous release term, and is always present during reactor operations at power levels sufficient to produce detectable quantities of Ar-41. Other sources are experimental facilities from which air is purged continuously (Rabbit), or released in puffs (e.g., Beam Ports, or CIF). Ar-41 releases from these facilities are also analyzed in Chapter 6.

Ar-41 physical half-life is assumed to be 1.83 hours9.606481e-4 days <br />0.0231 hours <br />1.372354e-4 weeks <br />3.15815e-5 months <br />. As discussed in Section 6.3.4 the effective half-life of Ar-41 within the confines of the Reactor Building must take into account both radioactive decay of the isotope and purging of the building air. Using the measured volume flow rate of the building exhaust fan (2000 CFM) and the estimated building volume (based on measurements, estimated to be 70,000 cubic feet), the effective Ar-41 half-life is computed as 42.75 minutes. Now, equilibrium activity is assumed to be reached in five effective half-lives (which yields about 97% saturation activity), which is thus a reactor run of at least 213.75 minutes.

Now, assuming that the Rabbit blower is operated continuously during this postulated operation, the release term for the Rabbit must be added to the release from the pool water. Section 6.3.4.7 (page 147) of the Safety Analysis Report analyzes this situation. The Ar-41 source term is calculated to be 3.656 microcuries of Ar-41 added to the building air per

• second. Time-dependent concentration, taking into account the effective Ar-41 half-life in the building atmosphere, is described by the following equation:

-At C(t) = P(1-e gj(yy) where P is the production (source) term, V is the building volume, t is the time after the start of isotope production (reactor run time), and A l 1s the effective decay constant for Ar-41 in the building atmosphere (this equation is developed in section 6.3.4.3, page 141 of the Safety Analysis Report). The form of this equation shows that the peak Ar-41 concentration under these conditions is attained at the end of the 8

____-___ D

postulated reactor run, and is essentially the saturation concentration of Ar-41:

C(=) = P/AV,

-6 which is equal to 6.83 x 10 microcuries of Ar-41 per milliliter of building air, as noted on page 147 of the Safety Analysis Report.

The average Ar-41 concentration during the postulated run is obtained by integrating the time-dependent concentration' equation given above over the length of the reactor run:

t 2

C g

=

C(t) dt/(t 2-ty)

J ti which leads to:

C ave

=

([P/XV] (1-e~ ) dt}/(t 2-t3 )

ti

=

(P/AVt)[t + (e /A) - (1/A)],

assuming t = 0 and t = t, which, upon substitution of Appropriate 2 6 values, leads to an average concentration of 4.92 x 10 microcuries per milliliter of building air. This average Ar-41 concentration is a factor of about 2.46 above the allowable concentration of Ar-41 in a restricted area, averaged over a 13-week calender quarter, assuming a 40-hour work week. Under the interpretation that an MPC concentration inhaled over a specified time leads to an internal dose equal to the allowable limit for the specified time, and if one assumes that the allowable quarterly occupational exposure is 1250 mrem, then an MPC level concentration corresponds to an internal dose rate of about 2.4 mrem / hour. If the average concentration is 2.46 times above MPC, the implied internal dose rate is thus about 5.9 mrem / hour. For the postulated run time of 213.75 minutes, the total internal dose commitment is about 21 mrem.

-6 If one assumes a peak concentration of 6.83 x 10 microcuries of Ar-41 per milliliter of building air, the concentration is about 3.4 times the allowable MPC for Ar-41 in a restricte,d area. Using the assumptions stated above, this corresponds to an internal dose rate of about 8.16 mrem / hour once the equilibrium, saturation concentration has been reached.

For persons breathing air containing a gamma-emitting radionuclides, an external dose is delivered from the volume-distributed source, as well as an internal dose from the ingested air. Exact computation of the external dose delivered from the volume-distributed source is a complex problem, and must take into account photon transport effects such as scattering (i.e., " shine"), self-absorption, and absorption / scattering effects from surrounding structures. Such an exact analysis is a formidable task.

Appendix D of this document presents a simplified analysis for external dose from a volume-distributed source of Ar-41 mixed with air. This analysis shows that the external dose term is small compared with the internal dose, and can be neglected for this analysis.

9 )

f; p,

b 1 L

g 4.5 If the building exhaust fan is stopped during a reactor operation involving a continuous release of Ar-41, the Ar-41 concentration would increase at a rate and reach an equf13hrium value determined by the physical (radioactive) decay constant of Ar-41 alone. The analysis developed in section 6.3.4.3 of the Safety-Analysis Report is still I '

correct, if the physieni decay constant for Ar-41 is used in place of the effective decay constant calculated in section 6.3.4.2, page 140. The time-dependent concentration can be er. pressed by:

C(t) = P/1 2V + e [C,- P/A 2V],

where V = building volume, P = production rate of Ar-41, A = physical decay constant of Ar-41, 2

t = time from the point at which the building 2

exhaust fan shut off to the point at which the concentration is calculated, and

-C = concentration of Ar-41 in the building air at the time the exhaust fan was shut off.

From earlier analyses, it is seen that Cg = (P/ A3 V)(1 - e~ 1),

where t = time from the beginning of the run to the time g

that the building exhaust fan is shut off, and A = effective decay constant of Ar-41 in the building 3

atmosphere at' king into account physical decay.and purging by the exhaust fan.

And the other terms as defined previously. The form of the time-dependent concentration equation shows that as t becomes large, C(t) approaches a 2

constantdeterminedbyP/A}V,whichigtheequiljbriumsaturation concentration. Using A = 1.052 x 10 seconds , V = 70,000 cubic feet, andPascalculatedear$1er(section6.3.4.7),geequilibrium concentration of Ar-43 is found to be 1.75 x 10 microcuries of Ar-41 per milliliter of building air. Using the interpretation for MPC related to internal dose rate as stated in an earlier response, this concentration, which is about 8.8 times the allowable restricted area MPC, an internal dose rate equivalent of about 21 mrem / hour is obtained.

4.6 Appendix E of this document presents an alternate analysis for Ar-41 releases from the OSURR under various conditions.

10

4.7 It is our understanding that an " airborne radioactivity area" as defined

-in 10CFR20.203(d).is an area that exceeds MPC (10CFR20 Appendix B. Table 1, Column 1) or 25% of MPC for the hours the area is occupied. We must

, assume that the Reactor Building will be occupied 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per week.

Calculations shown in the response to question 4.4 above indicate that the Ar-41 concentration may exceed 25% of MPC. Therefore, we must look to 10CFR20.103(b)(2) for guidance. This section stipulates that intake shall be limited as far as possible below that which would result from inhalation of such material for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> at the concentration specified in Appendix B, Table 1, Column 1. We therefore intend to add a Technical-Specification to 3.6.2 (Radioactive Effluents) which limits Ar-41 release in a restricted area. In view of the results of the answers to questions 4.5 and 4,6, we intend to impose no additional specifications.

4.8 The entire Reactor Building is defined as a Restricted Area. Personnel access is normally controlled by doors locked from the outside. To gain access, an individual must either be admitted by a member of the OSURR staff, and go through appropriate access procedures (badging, sign-in, etc.), or have their own key to the buiJding. Keys are only issued to individuals who have been trained in Radiation Safety (in accordance with 10CFR19 provisions) and in security and general access procedures. Some of this information 's'noted in section 5.7, page 124, of the Safety Analysis Report.

During reactor operations, outside doors are normally kept closed. A check is made during prestart checkout procedures to assure that the doors are closed and are capable of being properly closed and locked. Windows to the outside are normally closed at all times, although they can be opened on an occasional, non-routine basis. Doors leading to offices and classrooms are normally closed during warmer months, since adjacent offices and classrooms have air conditioning while the reactor room does not. Doors may be opened during cooler months. There are no restrictions on positioning of these doors during reactor operations. As noted in the response to question 2.3 earlier, it is assumed that air circulates freely through a)) volumes with the Reactor Building. This is normally the case, even during warmer months, since return air for the HVAC system is taken from the reactor room and distributed through the system.

The " classroom" referred to (room 201) is in fact a modest-sized meeting / conference room. Seldom are " classes", as such, held in this room. It serves more as a staging area for experimenters, meeting space for staff, and has in the past been used as a multiple-occupancy office.

l On occasion, individuals, small groups and larger groups may be present in this room during reactor operations, but thit would be for a small percentage of the total operating hours of the OsURR. The presence of students in the Reactor Building is, obviously, one of ti,e primary reasons for the existence of university-owned research reactors.

I The worst-case accident scenario for persons occupying this classroom area would be the same as for any other person in the Reactor Building at the l time the accident occurred. The possible accident cases are analyzed in l Chapter 8 of the SAR. However, if anything, the potential consequences of accidents to persons in room 201 is less severe than for persons in other l

i areas of the Reactor Building, since room 201 is located away from the 11

(

____-_ _ _ i

e h

h ps

[ pool top area,;and rapid egress from room'201 and the Reactor' Building is

, possible,'since room 201 is located'near the stairway _ leading to the front

' door of the building.

-The' definition of radiat! ion worker versus member'of the public is determined by the nature of an individual's presence at the facility.

Obviously, if a person receives instruction in radiation safety as per 10CFR19 requirements, it is assumed that this person-is a radiation worker. Such individuals are generally persons involved in ongoing experiments and~other work at the facility, regular < facility staff, and students and faculty involved in course instruction utilizing the OSURR on a regular basis. Occasional visitors,;non-regular experimenters,; persons-touring the facility, one-time classes. involved in instruction and demonstrations, custodial staff, vendor representatives, and members of j the news media are considered members of the public, and would not receive-instruction as-required by'10CFR19. However, these individuals are escorted and/or supervised during their time at the facility, and such supervision would be the primary means of keeping' exposures and doses as l low as reasonably achievable. .i 4.9 First, it must be assumed ~that the operator will not simply open the Rabbit carrier tube and breathe.the air contained within it directly.

'This is a' reasonable' assumption since operators of the Rabbit system will ,

L have some knowledge of how the system operates and will be aware of the .]'

need to avoid direct inhalation of the air within the Rabbit carrier tube.

l Rather, a more reasonable assumption is that the operator opens the Rabbit'  ;

carrier' tube at a typical working-level, then is surrounded by a' mixture l of Ar-41 and air. It is assumed for this analysis that this volume'of air  !

takes-the form of a spherical bubble with a radius equal to the' distance 'l between the carrier tube and the worker's head; .-Typically,_the carrier-j tube'is opened on a work surface at about waist level. Thus, for an-  !

averageperson,thecarrierisabout3feetawayfromtheirgead. The j volume of a sphere with'a radius of 3 feet is about 3.2 x 10 milliliters.

Table 6.3, page 131 of the Safety Analysis Report indicates that the saturation Ar-41 activity in the Rabbit carrier tube is about-1.11 millicuries. Again, this is conservative in that it assumes that a saturation concentration of Ar-41 is present in the carrier,-which is almost always not the case, since Rabbit irradiations are generally 20 l

, minutes or less. However, assuming,forconservagism,thatareleaseof -)

1.11millicuriesoccursintoavgumeof3.2x10 milliliters, a f concentration of about 3.47 x 10 microcuries of Ar-41 per milliliter of air results. Thisisafactorofabout173.5abovetherestricgd-area

_MPC for Ar-41. Assuming that the restricted-area MPC of 2 x 10 ,

microcuries per milliliter results in a dose equal to the radiation worker i dose limit of 1250 millirem per calender quarter, a dose rate of 2.4 millirem per hour is implied, assuming a 40-hour work week. Thus, an exposure 173.5 times above this implies a rate of 416.4 millirem per hour.

Now, one sust next estimate the duration of the exposure. For a person working with the Rabbit carrier tube, the nature of their postulated work makes reasonably sure that they will move out of the area of high gas concentration within about one minute. This is because they will likely be working with the isotopes produced in the Rabbit tube in some manner, either in a counting experiment of other type of manipulation. The one minute exposure time allows enough time to open the carrier tube and 12

perform necessary radiation surveys. Assuming a one minute exposure time, the dose commitment:from breathing the air from the carrier tube is estimated to be about 6.9 millirem.

It should be noted that this estimate is extremely conservative for.

several reasons. First, it assumes that saturation concentrations of Ar-41 aie present in the Rabbit carrier. tube. This requires a long irradiation in the Rabbit facility,-and generally irradiations are on the order of 20 minutes or less. Also, it assumes that there is no dispersion of the Ar-41 contained in the 6-foot diameter bubble surrounding the carrier tube after it is opened, for the duration of the one-minute exposure time postulated above. In fact, significant dispersion occurs in the air of the building since the building atmosphere is continuously purged to the outside air by the exhaust fan. Natural dispersive effects also occur within the building air. Thus, the concentrations of Ar-41 in the immediate vicinity of the system operator are likely much less than those estimated above.

Table 6.3 of the SAR may also be used to estimate dose from a more typical, conservative estimate of Rabbit system operation. Assuming a 20 minute irradiation at 500 kilowatts, a source term of 131.2 microcuries of Ar-41 is estimated for the carrier tube. Using the spherical bubble _godel developed above, the concentration of Ar-41 would be about 4.1 x 10 microcuries per milliliter. This is about 20.5 times above restricted-area MPC, which implies a dose rate of about 49 millirem per hour.

Assuming again a one minute exposure time, the total dose commitment would be about 0.82 millirem.

A more typical 5-minute Rabbit irradiation would result in a carrier tube activityinventoryofabout34microcurieso{Ar-41,andresulting spherical bubble concentration of 1.07 x 10 microcuries per milliliter.  ;

The equivalent dose rate would be about 12.8 millirem per hour, which for a one minute exposure leads to a dose of 0.23 millirem.

Such exposures are kept as low as reasonably achievable. By this, we .l imply that unnecessary Rabbit operations are avoided. The carrier tube is i opened at ralst level or further away, and the air from the carrier tube is not breathed directly. Dispersion occurs within the building air by use of'the building exhaust fan. Reactor operations do not take place unless the fan is operable. Unused volume in the carrier tube is taken up l by cotton wadding, thereby reducing the effective volume and, therefore, the activity produced in the carrier tube, i

4.10 The effluent monitoring system must be in operation during all reactor operations. The output of tFxs system is continuously recorded. Reactor l trips occur if the system is turned off or the recorder is turned off.

Checkout and calibration of this system (using a check source) is performed as part of the reactor prestart checkout, performed before all operations. Stripchart recordings of the effluent monitor signal are kept. A visual signal is provided to the operator if the output exceeds the setpoint associated with restricted-area MPC levels of Ar-41 in the l effluent. Since these indications and recordings are made continuously, comparisons are made for each operation, and for each release that occurs.

I l

1 l l

13 j

i Currently, comparisons are made to unrestricted area MPC on a semi-annual basis. With 500 kilowatt operation, a check will be made at the end of the reactor run to compare average effluent concentration readings with unrestricted area MPC, and 10 times unrestricted area MPC for a one day average.

4.11 The thermal column extensions which extend from the pool walls to the west and south' edges of the core are clad in aluminum. There is no access to these volumes from the outside of the pool. Thus, no releases of Ar-41 would be made from these regions, even if there were trapped air in this volumes (which is not at a)) certain that there is even any air in them).

Section 6.3.4.5, page 144 of the Safety' Analysis Report, estimates Ar-41 concentration for a puff release from the main graphite thermal column, which is the accessible region for the thermal column facility.

5.1 The staff of the OSURR is most aware there are several sources of direct radiation that may require administrative controls to maintain doses ALARA. Specific procedures and~ controls already in place include but are not limited to the following:

1. AP-02 General Rules - outlines the rules experimenters are expected to be familiar with and follow while within the NRL facility

2. AP-04 Procedures for Approval of Requests For Reactor Operation

- assures compliance with ANS 15.6 Standard for Review of Experiments for Research Reactors prior to performance of each experiment

3. RS-09 Area Radiation Surveys - assures that routine and supplementary surveys are completed as deemed necessary by the SRO on duty or by the conditions of the experiment

4. RS-15 Radiation Safety Instruction - assures each individual is instructed as required by 10CFR19.

These procedures assure prior training of experimenters and staff and evaluation of experiments and doses prior to their performance. We anticipate developing additional procedures as the need arises.

These are reviewed and approved by the Reactor Operations Committee as described in the SAR. We also anticipate that all testing operations al 300 KW will be considered new experiments and shall require prior ROC approval. During these start-up tests we wi))

establish access control from actual measurements. For example it is likely that pool top access may be limited.

We have also held discussions with representatives from the Office of Radiation Safety about obtaining additional personnel for health physics activities during and after an increase in power. We anticipate ORS will be actively involved in advising and developing procedures and controls to maintain doses ALARA.

14 i

p i

6 Other items under consideration are addition of area radiation monitors,

^

alarming personnel dow1 meters, and a plug for the central irradiation facility. We are also installing an alarm panel in the control room that L may be used to warn of pool top access or intrusion into other controlled areas. ,

The combination of existing procedures which require training, review and approval of radiation safety procedures and all reactor experiments, and appropriate surveys; testing to be completed at 500 KW; Office of Radiation Safety participation; and.other as needed items will be used to maintain personne1~ doses well within 20CFR20 limits.

5.2 The attached document, "As Low As Reasonably Achievable" Commitment and Implementation, which is a part of The Ohio State University License Application for Renewal submitted July 28, 1986 indicates the-basis of the ALARA policy and the University's firm commitment to it.

It is signed by the Vice President for Health Services. As indicated in Figure 9.1 p. 231 of the SAR there is a direct line of

' responsibility from this Vice President through the Office of Radiation Safety (ORS) to the Nuclear Reactor Laboratory for matters regarding radiological safety. In addition, the Director of the Office of Radiation Safety is an ex-officio member of the Reactor Operations Comm!ttee (ROC). As described in Chapter Nine of the SAR, both the ORS and-the ROC shall review and audit operations to assure the facility operates in a manner consistent with public safety.

Thia includes the ALARA policy.

5.3 There is a spelling error. The correct spe))ing for the word meaning "to be complementary to" or " serving to fill out or complete" is complement. As you suggest, implement, meaning "to carry out" is perhaps a better choice of words but complement is certainly acceptable.

5.4 There are two individuals who have worked at the OSURR for approximately 10 years each. They have completed the very large majority of experimental work in and around the reactor facility and should be a good indication of typical doses received at the reactor. (

Their average is less than 1% of that allowed per year for whole body doses. If power is increased by 50 times, their doses might increase to as much as 50% of the allowed limits. We do not anticipate this large of an increase. Prudent monitoring practices should allow us to maintain doses significantly lower than 50% of the allowed limits.

Realistically, based on previous experience with effective implementation of the University ALARA policy, an increase to about {

10% or less of allowed limits is possible.  !

)

15 l

l L- --____

h.

5.5 "According-to 10CFR20.4 C(3) the doses tabulated-in Table 7.5, page 171 of the Safety Analysis Report should be increased'by-a factor of

10. .Then multiplying by 500,000 watts, one should obtain the total b ,

res/hr for the several open experimental facilities listed below. .

CIP -

2170 Rem /hr BP*1 - 132,000 Rem /hr BP#2' - 90.500 Rem /hr TC -

4800 Rem /hr In order to make these doses low enough to be' acceptable, the-experimental facilities.will have.to be closed, plugged,~or otherwise shielded during operations at 500 KW. This is always the case.

During startup testing for 500 KW operation the various' experimental facilities shall be monitored for neutron as well as gamma doses.

Appropriate shielding, administrative controls, and access control-E will be implemented during the startup testing to assure all doses are ALARA.

6.1 The basis.of this statement is the Safety Analysis Report submitted for.

the University of Lowell Nuclear Reactor (ULR), License No. R-125, Docket No. 50-223,- an MTR-type facility licensed by the NRC to operate at a steady-state power level of 1 megawatt. This facility.has been relicensed in the past two years. , Quoting from the ULR SAR submitted in September, ,

' 1973, section 9.1.5, page 9-29, regarding total loss of water: l

" Work was'done_by Wett 12 in investigating the surface temperature of Oak  !

Ridge Research Reactor (ORR) fuel elements under natural. convection air  !

cooling. This work, along with experimental results from the Low IntensityTesting' Reactor (LITR)gndtheLivermorePool-TypeReactor (LPTR) was correlated by Webster , whose conclusions are pertinent ,

because the ORR, LITR, and LPTR are all light-water moderated research

- reactors with solid plate-type fuel (commonly called MTR or BSR type l elements) which are very similar to the.LTIR design. Webster states.

"This analysis indicates that the LITR could be operated continuously at 3 l MW and could lose the cooling water through a rupture'in the reactor tank

! l

? '

.' without danger of melting the fuel, even without core spray." The

~

indication that a total loss of water from the pool of the LTIR operating i at 1 MW would lead to no melting of the fuel is clear."  !

In the above quotation, the "LTIR" cited is the former name of the ULR )

facility, the Lowell Technological Institute Reactor. .Also, references 12 ,

and 13 noted above are reproduced in the references to this document as l references 4 and 5.

Using a similar line of reasoning, the indication that a total loss of l~ water from the pool of the OSURR operating at 500 kW would lead to no melting of the fuel is clear. It should be noted here that the primary considerations when comparing such analyses are the heat source term and ]

possible heat removal mechanisms. For isolated fuel elements cooled by  ;

natural convective circulation of air, reactor-specific features such as l l heat sinks, experimental facilities, etc., are irrelevant. ]

16 l

The LITH was constructed at Oak Ridge National Laboratory as a hydraulic-mock-up facility for the MTR. Etherington [1] states that the LITR was "in many respects identical in physical configuration wJth the MTR." The LITR initially operated at 500 kW, but was upgraded to 3'MW early in its operating history. It used MTR-type fuel, with 16 curved fuel plates per fuel assembly. The uranium was 90% enrichment, in the form of UAly, with a uranium concentration of 18.5%. The core was surrounded by Be reflectors. The fuel elements were about 3"x3"x49.4", with the fuel plates being 2.8"x0.05"x24.63". The fuel plate pitch in each element was 0.18". The core dimensions could range up td 15"x27"x24.63", which, at 3 MW power, leads to a power density of about 18.4 watts /cc of active core, assuming the maximum core size. LITR power density could be higher, especially at the beginning of life of a core configuration, if the core were made smaller. The 500 kW OSURR core, as shown in Figure 4.4, pagg 93 of the Safety Analysis Report, has an active volume of about 77,871 cm ,

which, at 500 kW operating power, leads to a power density of about 6.42 watts /cc of active core. This is about one-third the power density of the LITR. Given the lower power density, it is likely that the average heat source term for the OSURR fuel elements would be smaller than for the LITR fuel elements. The specific heat of the LITR fuel plates-is very close to that of the OSURR fuel plates, given that the material composition is similar (aluminum cladding, uranium fuel). The fuel plate geometry is also very similar (0.05" thickness, 24" length, 2.8" width). LITR heat generation rate was higher than that predicted for the OSURR since the LITR operating Iawer-was a factor of 6 higher. This more than offsets any small differences in total peaking factor. The LITR operated at essentially room temperature and pressure, as does the OSURR, so fuel plate temperatures during normal operation are similar. Fuel plate >

temperature under conditions of total loss of coolant conditions would be higher for the LITR since its operating power and heat generation rate are higher than the OSURR. Heat conduction from the uncovered core would be similar for the LITR and OSURR, since both are surrounded by reflectors.

If anything, the heat conduction from the uncovered OSURR core would be higher than for the LITR, since the LITR Be reflectors are not in as close

-proximity to the edge of the core as are the OSURR reflectors. )

All of these considerations lead us to believe that a comparison between the LITR fuel elements cooled by natural convection of air and the OSURR fuel elements cooled by natural convection of air is valid, and, if

[

anything, likely conservative as far as safety implications are concerned.

6.2 These effects are estimated in calculations shown in Appendix F of this document. The estimates include both direct radiation from the core at the pool top area, and floor-level scattered radiation (i.e., " shine")

from scattering off of the building ceiling.

l 6.3 Current OSURR procedures and Technical Specifications require annual inspection of five fuel elements and triennial inspection of all fuel elements. There is no requirement for inspection of control rods.

Our proposed Technical Specifications for 500 KW operation require visual inspection of the shim safety rods and regulating rod (Section l 4.2.1(3)). There is no requirement for fuel element inspections. It ,

is planned that we will revise Procedure OM-07, Fuel Element Maintenance, to reflect these changes. We anticipate inspection of l 1

17 I

__ a

l-fuel elements, if at all, to take place while they are still well below the water surface. In this manner doses will remain very small. It may also be possible to inspect using a shielded fuel element cutting facility that is being developed to trim our highly enriched fuel before shipment. The control rod inspections will be conductec na;n the rods below the surface of the water. The water and pool wall will provide shielding to maintain doses ALARA.

6.4 The PARET code used in the OSURR accident analyses is the newer version developed at Argonne National Laboratory for honpower reactors.

6.5 The internal dose affectivity factors used irs this calculation are taken from the Safety Analysis Report for the University of Lowell research reactor, an MTR-type facility operating at 1 megawatt steady-state thermal power. The factors used are very similar to those used by the University of Texas at Austin reactor, Docket No. 50-602, a TRIGA reactor licensed to operate at up to 1 megawatt steady-state thermal power, and 1400 megawatts pulse power. Note that the factors used in the OSURR analysis are more conservative than those used for the University of Missouri et Rolla research reactor, License No. R-79, Docket No. 50-123, a 200 kilowatt MTR facility.

7.1 Section 8.4.4.1, pages 211-212 of the Safety Analysis Report discusses the scenario assumptions for radiciodine release from a damaged fuel plate.

Many of the assumptions in this analysis are conservative, e.g., the assumption of an infinite power history before the accident, and the location of the damaged fuel plate being at the peak flux location in the core. Other sections of Chapter 8 discuss scenario assumptions for other gaseous radionuclides. For conservatism, we can assume that, at the time of the damaged fuel plate accident event, there are persons in the building that are not classified as radiation workers, and thus lower dose limits would apply. Further, exact location of persons within the building will not be considered, since the building air circulates freely within the building and it is assumed, in some cases conservatively, that instantaneous mixing with the building air occurs.

For persons within the building at the time of the accident, the total dose commitment will depend in part on the exposure time. Experience based on emergency evacuation drills, as required by our NRC-approved Emergency Plan, indicate that a reasonable evacuation time for a member of the general public is about 1.5 minutes. Using methodology presented in section 8.4.4.3, page 217 of the Safety Analysis Report, the total thyroid dose commitment from radiolodine resulting from a 1.5 minute exposure under the postulated conditions is about 88.2 millirems. For a member of the OSURR staff having responsibility for assuring that the building was evacuated, the total exposure time, based on past emergency drills, is about 5 minutes. The total thyroid dose commitment from this exposure time is about 293.5 millirem. This would be considered the maximun credible worst-case thyroid dose from radiciodine exposure.

For inclusion of other gaseous radionuclides with the radiciodine, Table 8.12, page 224 of the Safety Analysis Report allows estimation of the whole-body immersion dose, using methodology described in section 8.4.4.5, pages 220-226 of the SAR. Table 8.12 provides the most conservative 18

_ _ _ _ _ .D

d estimates of whole-body. dose. For times shorter than 5 minutes, a linear-

. interpolation of the data shown in Table 8.12:is reasonable, since,the

~

total dose function varies slowly over time for times relatively'soon after the postulated event occurs. Thus, for a member of the general-public, being evacuated from the building in l'.5 minutes, the total. dose commitment is about 398.6 millirems. For a person having evacuation supervision responsibilities, entailing a 5 minute exposure time, Table 8.12 estimates a total whole-body dose of about 1.2 rems. Again..it should be noted that the assumptions used in the analysis which. derives these estimates are conservative.

• For doses external to the Reactor' Building, whole-body dose estimates would yield essentially no dose, assuming that the evacuation plan is followed and the building ventilation systems are turned off. With the building purge fan deactivated. the louvers on the exhaust aperture port close under their own weight, effectively sealing the interior of the building from the outside environment. Any miscellaneous leakage that might occur from the building would be greatly reduced in activity over that which would occur of the fan were still operated, and in any case would also be greatly diluted by atmospheric dispersion. Recall'that-earlier estimates for Ar-41 dispersion indicate a dilution factor in the wake of the Reactor Building of about 71, under rather pessimistic 6 assumptions for atmospheric dispersion, and could be as high as 590 under most likely conditions. Such dilution of gaseous radionuclides leaking from the building would lead to rather low external doses.

It should be noted that even if the Emergency Director should overlook

' turning off the building ventilation system from-inside the building prior to. evacuating the building, it is possible.to turn off building power from outside the Reactor Building, which would accomplish the task of isolating the interior of the building from the outside environment.

7.2- The proposed dose " guidelines" during the postulated accident conditions are, once again, guided by the ALARA principle. This requires prompt evacuation of non-essential personnel from the building at the time the emergency requiring evacuation is declared, as dictated by the NRC-approved Emergency Plan for the OSURR. Similarly, the Emergency plan as.

approved by the NRC requires certain tasks to be performed by designated OSURR personnel during an emergency. These tasks shall be completed as expeditiously as possible under the conditions encountered, again keeping doses as low as possible while still completing necescary, required tasks.

For persons located in the nearest occupied facility external to the Reactor Duilding, ALARA is once again is the operating principle.

Unnecessary doses will be avoided. Given the nature of the location of the Reactor Building relative to the nearest occupied facility (Van de Graaff Laboratory, about 150 feet to the south of the Reactor Building),

only the simplest of measures need be taken to avoid essentially all possible dose. For example, prohibiting access to the Reactor Building, or limiting approach to the building during the emergency, are likely all that will be required to elimintte possible doses to Van de Graaff Laboratory personnel. Monitoring personnel' access to the vicinity of the Reactor Building is a relatively simple matter, since OSURR personnel would be required to report to the Van de Graaff Laboratory entrance in 19

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

i ,

the event of an. emergency requiring" evacuation, and commun1' cations to

~

appropriate personnel and/or. agencies would.be accomplished from.the Van.

de Graaff' Laboratory,;as, required by'the Emergency Plan.

~

g 7.3 As noted in-the. response to question 7.1;;the likely. exposure time within the. Reactor Building-for a. member;of.the OSURR staff having responsibilities for' assuring building evacuation is 5 minutes, based on

-Emergency Planfdrills conducted in the past several years for scenarios u ' requiring building evacuation.- For a person:not having evacuation.

assurance responsibilities,.the'likely upper limit on occupancy time is

~

'about-1.5 minutes. ,

7.4 Results are contained.in the. response to question 7.1 above.

It is our belief that providing estimates for a variety of exposure times should not be viewed as."nonpertinent information". Rather, providing a-range of possible exposure times and resultant doses, even if the postulated times are outside the~rango of credible exposure times, makes possible a." feel" for how sensitive-the calculated parameter is to the other parameters.used in the. calculation. Such sensitivity analyses.are' particularly useful for establishing " envelopes" within which various activities can be planned and: carried out.

7.5 Asnoted[above,.themitigativefeaturesoftheReactorBuilding(closing of purge fan. exhaust. aperture louvers after turning off the ventilation systems).would lead _to essentially =no airborne radionuclides concentrations in the unrestricted area. Thus, there would not be a significant f immersion-type dose or inhalation dose outside the Reactor Building.

7.6 'Again, assuming that the mitigative features function properly, there would be essentially no offsite release that would lead to significant whole-body or. thyroid dose commitments l.

7.7' The calculated building doses, when compared with " accepted guidelines",

are acceptableLsince it is assumed that the operative " guidelines" during.

an emergency as postulated in the analysis will be the ALARA principle and 10CFR20.11mits. The Emergency Plan, dictates that steps be taken to' keep personnel doses as low as possible during an emergency, while still accomplishing necessary, required tasks. 10CFR20 limits are met for all dose estimates presented in the response to question 7.1 above, except for the quarterly whole-body immersion dose for a non-occupational exposure during the assumed 1.5 minute exposure time. However, if we assume that this is a worst-case, one-time exposure for the entire year, the yearly limit of 0.5 rem non-occupational exposure is met.

Offsite dose estimates are within acceptable limits since the mitigative features of the Reactor Building and required actions to be taken in the event of an emergency will result in no offsite airborne radionuclides.

Technical Specifications 3.4 and 3.5 address Containment Isolation and Ventilation Systems, respectively. Specification 4.4 addresses surveillance requirements related to the containment isolation features.

.The specific operating procedures associated with these Specifications provides for assurance that the ventilatfor systems will be shutdown on 20

y

+ s ,

commands and that the building isolation louvers will close when the purge j fan is turned off. Technical Specification 4.6.1 addresses requirements related to operation of the Effluent Monitor. This system must be i operable and operating during all reactor operations and will provide indications of airborne radionuclides, if visual observation of fission gas release has not already been reported, or if other monitoring systems have not detected the presence of airborne activity.

i Technical specifications address operational limitations and conditions, and their purpose does not include specifying individual emergency procedures or specific features of the NRC-approved facility Emergency Plan. Specific emergency procedures are written by the OSURR staff and reviewed and approved by the university's Reactor Operations Committee (Figure 9.1, page 231 of the Safety Analysis Report notes the role of this committee and its lines of communication within the University). The facility Emergency Plan is approved by the NRC and reviewed annually by the OSURR staff. Thus, building evacuation, which would be required to keep personnel doses within the limits estimated above, is addressed by the Emergency Plan, and emergency operating procedures, and not in explicit Technical Specifications. It remains our belief that specific operating procedures and/or features of the Emergency Plan should not be incorporated into the Technical Specifications, since unnecessarily restrictive and redundant limitations might he inferred, or unreasonable requirements imposed.

7.8 The equation cited does contain a typographical error, and a " slash",

denoting division, should be placed between the symbol for density and the symbol for molecular weight of water. This typographical error does not affect the calculations and results which follow, since the correct equation was used for these.

l l

l l

l l

21

f.

4 8' 0 OSURR Technical Specifications Related to Power Increase 8.1.1 We have added a statement to T.S. 2.2(1)'that the steady state power level shall not exceed 500 KW thermal.

8.1.2 The bases for the Limiting Safety System Settings are.

found in the SAR (4.8.2 and 8.4.3). 4.8.2 discusses the fuel clad temperature at steady state operation of 500 KW. This temperature is 56.5*C. In 8.4.3.3 one assumption is a reactor trip occurs at 750 KW after steady state operation at 600 KW with core inlet temperature of 91* C. All fuel clad temperatures are well below the blister threshold of 550*C.

8.2 We have added information to the bases for Specification 3.1.1(1) which addresses safety considerations as well as the already referenced operational considerations.

8.3.1 We do not intend to have a scram at 0.5 MW power level.

That is intended to be our steady state power level..

Specification 2.2 Limiting Safety System Settings discusser and justifies a scram setting of 600 KW.

8.3.2 One must recall that our core is always cooled by-natural convection. The choice of 120 KW to have the pumps on with primary coolant flow is based very simply on keeping core inlet temperature below 35*C during extended runs at higher power levels. Procedurally, the pumps may be on providing flow for primary' cooling at power levels much lower than 120 KW. The quantative justification for 3.2.3 Items 4 and 5 and 3.3.1 is the 35"C Limiting Safety Syster. Setting and the preferred -

normal core inlet temperature range of 20-25* C.

8.3.3 As stated in Specification 3.3.2 Coolant Level, Section 7.1.1.4 of the SAR discusses the shielding provided by the water. The bases for the entire Specification 3.2.3 which is a table of scrams were somewhat cursory.

Often, as in the case cited above the bases were elaborated upon in other specifications.

8.3.4 Historically we had operated with magnet currents set at 60 ma. Measured release times for the magnets had always met our previous technical specification of 60 milli-sec release time with 60 ma of current and 600 msec insertion time. Our previous scram on this only occurred if two magnet currents reached 80 ma simultaneously. Our new technical specification simply allows each individual rod to drop if magnet current goes above 60 ma. Magnet current settings are briefly 22

I.

discussed in section 3.3.9.1 of the SAR. We have also always maintained records of our magnet control rod performance. Our current Specification, 3.2.1 Control Rod Drop Times, has a limit of 600 m sec. Our records indicate that this has always been met for magnet

~

current readings of 60 ma. We now normally. set our magnet currents at about 40-45 ma. For these settings T.S. are also met. The lower current setting is simply uan attempt to extend magnet life. Specification 4.2.2(7) already provides surveillance for this item.

8.4 As indicated in 7.1.1.4 doses at the pool top directly-  !

over the core may be relatively high. There are numerous procedural arrangements that may be implemented in order to maintain doses ALARA. Some that have been considered are the following:

1. Contralled access

2. Motion detector with control room alarm  ;

3. Additional radiation alarms i

The important aspect in that access will be controlled procedurally when actual measured doses are available 1 1

to determine what is ALARA.

8.5 The Staff of the OSURR realizes that its likely that impurities will be activated much more significantly at ,

500 KW. Of particular concern is when these are trapped in the demineralized that is a part of the  !

process cleanup system. Although this demineralized is '

somewhat inaccessible it is most likely that additional shielding or limiting of access to this area will be required. If other areas exhibit significant radiation levels due to trapped radioactivity they shall be j posted and access limited to maintain doses ALARA. l Our real concern remains for the production of Na-24 from the large quantities of Al in the core area. l After extended operations the 15 hour1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> half-life Na-24 l 1s often trapped in the demineralized which we have I indicated may have to be shielded or isolated. There f is an Area Radiation Monitor near the demineralized.

In addition, it is surveyed weekJy or after each days operation whichever is more frequent. The intent is to l address the concern with existing procedures rather than add to the Specifications.

8.6 The detection system in Specification 3.3.4 is automatic and is the same system that initiates a scram if pool water level drops below 20 feet.

I 1

l l

1 l

23 1

-l j

i' F

b .

8.7 We suggest that this Specification (3.3.5) be revised

.and we-have added a specification for primary coolant-water radioactivity. A specification would also be added to the Surveillance Requirements in section 4.3.2 Coolant System Radioactivity to monitor for secondary coolant radioactivity.

8.8. As described above a quarterly surveillance requirement or a sample immediately prior to or after a release would assure regulatory control for-liquid effluents. l In terms of gaseous effluent releases, Specification 3.4 , Containment Isolation, was written to assure thut ,

i the building can be isolated if necessary to limit effluent releases. The building ventilation fan and  !

all others are turned off by a single switch in the {

control room. If all windows and doors are shut then ]

further isolation is assured. -l 8.9 The word operable as used in Specification 3.4 means

.that all doors, windows, and vents shall be able to be closed in order to isolate the building. Doors may occasionally be open during operation at 500 KW to allow normal entry and exit from the reactor buildirg.

Air flow through any open window or door is measured to be inward due to the building exhaust fan. Windows are occasionally open for climate control. Measurements made using a Fisher Scientific anemometer, an alnor thermo-anemometer and visual observation of air patterns indicate that airflow is through a contro))ed pathway into the building through open doors, windows or cracks and out via the building exhaust' fan. Before each days reactor operation the prestart checkout requires one to verify that the building exhaust fan is operating.

8.10.1 For Specification 3.6.1(1) we have revised Operating Procedure RS-17, Ar-41 Release Calculation, to include an alarm for our effluent monitoring system. It is currently set to alarm at MPC for a restricted area.

The operator is then instructed to determine if continued operation will exceed 240 MPC hours for a single day. If this is or will be exceeded the reactor is to be shut down because this may cause us to exceed 10 x MPC for an unrestricted area when averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. (Reference ANSI /ANS-15.1-1982 "The development of technical specifications for research reactors" 3.7.2(3)).

24 1

, = _ _

Qr 01 ys [ ~;

+

• Ai E . ,

3 r

p n:.

r c4 I.'. t r '-' -

i

.' 2' S Specification 3.6'.1(2). applies 10 theLyet'to.be-2 installed Ar-41_ monitoring system for the, rabbit. An.

,- alarm:for;this maylbeiset.at:MPC.for a restricted area'.

This.would: alert the operator?to.then' closely monitor

-c Lthe~ building' effluent monitoring! system' described, above.

Alarm' set points as, discussed lip Specification 3;6J1(3)_

for-the Area Radiation Monitors-(ARM's)1are variable' Ds

< _from 1'mr/hr to11000:mr/hr. For operations at 10 KW we

.normally have'.them' set-to alarm at 10 mr/hr. It is likely we shall maintain the alarm set points th'e:same even with operations at 500 KW. -This will alert thei operations staff to the higher but. anticipated doses at."

E ,' the pool top and perhaps inithe' process. system-area during normal ~ operations. These'are discussed <in 8.4 and 8.5 above. Once these ARM's alarm they'will.have

~ to be. reset to higherclevels and. appropriate. procedures followed to limit access'and/or exposure.

8.10.2'The' word " operating"'has been substituted for the word "on" in Specifications 3.6.1(1), (2), and (3).

8.11- For. Specification 3.6.2(1).concerning liquid. releases we do not think any changes are necessary since liquid rreleases~are made at or below MPC. -The additional-

.three million gallons of water released from the site (The Ohio State University) each' day certainly meets the ALARA. concept by diluting already legal releases.to insignificant' levels. Specification 3.6.2(2) concerning Ar-41. releases to the. unrestricted area has

.already been discussed in 8.10'1-and in the answer to-

~

DAR 4.6. The release specification, as'already g'

indicated, is based on ANSI /ANS-15.1-1982'"The development of technical;specificationsafor research reactors"..

w It is appropriate that we add a specification to limit Ar-41 concentrations inside tlur Reactor Building (i;e.

the Restricted Area). However, the easiest method of accomplishing this is to determine the concentration in C1/ml as related to counts measured by the gaseous effluent monitor. We would do this by' adding a scaler to the monitoring system. This scaler would provide us with a total number of counts. This count would be compared to the total 10CFR20.103(b)(2) counts allowed from

+, operating at-MPC for a week (40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />). As indicated previously there shall be alarms at restricted MPC concentrations for both the rabbit monitoring system and the building effluent monitoring system. To meet the ALARA concept no unnecessary-irradiations shall take place and no operations shall be allowed if it is likely we shall reach weekly limits.

25 1

r

.8.12i1 We have inserted the word " absolute" before the word

.value in Specification 3.7.1.

t 8.12.2 There are differences between 3.7.1(4) and (5). 3.7.1(4)

e. ; is for insertion rate. 3.7.1(5) is for oscillated experiments where the frequency may vary but there is a limit on the total peak to peak reactivity. Currently these two specifications are needed since we-do perform oscillator experiments to measure the reactor transfer E function and may utilize experiments with moving parts.

g The original specification for our HEU core was written in such a way as to combine both specifications. After review of other reactor's specifications.it was decided to make these separate.

8.12.3 The definition of " movable" in section-1.3 ijt consistent with ANSI 15.1. Essentially all of our experiments are e

movable. Therefore reactivity limits in 3.7.1(2) and (3) are more conservative than for a potential secured experiment as described in 3.7.1(1).

8.12.4 The recommended change was made and we again referenced Section 8.4.3.2 of the SAR.

8.13 The recommended change has been made.

8.14 The recommended change with slight modification has been made. We think this is more realistic since releases to the tertiary system via the secondary are most unlikely but would be the actual pathway to the sanitary sewer system.

8.15 The recommended change has been made.

26

Attachment A Figures Referenced in the Preceding Section l

l 27

Il

= ~O g

(

l R O i o R

.n l

a f i f .n 1

1

{

_ O 2

i c

e 1

0 7 l C

4 i O l

~

a gTte M 1

-g 1 ir I

f 6

e e

ni g a l

1 b

t 0 l

N o

r t

R h Ce oa rc et o

r B a

R R

• y

.ni e E hO l

a r F i

e g a

O '

D o u o r R r e o

l 1 l

U

• O p

S D

o o

r a ,

U R

R G

r o

u n _

d .

F r

o }== '

' F l

n  : o o

t r 3

D 2 P o _ 2

- _ l o _ 4 a r I f

r 2

~

f e n e 4 e e t R

t f e O R C i e f i

o n.

f n. u R o R t i i n 1

0 i

i r f i c o t t e f

.n " i 9

< i e 3 n 1

- c Ol g 0

1 e

c e 3 A l 1 e I

1 0 Ob 1 ~

lI '

, ' 8 t ,,

- .-=

t.,.,,.... .

,a

. ... . .. .. -. . . ... . .. . ,.-.~.g. ,.

• i',,

J .

i a. .

.i - -

. .. .i

. .. .i: - -

. ,. .i  :

. .. .i  :

. .. n .i  :

. .. Ce .i  :

. N x

oa .i  : o

. .- re .i  :

r

. . et  : ,

t

. .. a . :

r .i  : h -

. . - * . .. .i~ :

. . ... . . - . .. ... . . .. . ... .... . .. . , .... . .i

... ,.. . .i:

_. .i  : F

• .4

.!  : i g

u

...i: r

. . .... . .., ......... ,...: i e

...i

2

...i

...i  :

,s,..,......a, l

w. ..i a . - .

.,,.,a ...i

.i

.i S

O U

R 0 R 2 3 S 2 e 0 4 0

'g I. c f f f o e e e n e e d e e t t t

F Co C goC R O R l o

ln R on f o

af ot .m f m r se m

. mro 2 i .

c sr 2' re l 0 e 0 P l

on / 5 9 a oc l I

n me/

I

~

y

n. -

1 l i .!1 lll \, -11. .,, 1

' i4i, f. . .a9.?.I 0o e.

L__ '> r -

.i , ,; e

$ 9. 3 l^

n d ut il 'k a $h..

a s

M 9't pe o

?

. sgi ,$

g Y A w

,I a 2G 94 jf . Oj l! @f .

4 d)e 9 8 ifE $ 't i j.e *i;:.(o( st +

i y sq 2

s 3 j l.

8 Og b lad 9gd 13g]-

e a

l" 5

bhiihi!j.3e

$ s, l .. hof,! .hb ip :n. 3 A m th y n it i le da h4qQac 'r -

n:L a i i

! @FHV i

w mtbfaf.

_ - 9., @gA M o Ls e r ,

m m i_ m n -:[- %j_% , nft ne3jE Jo e- - - - _ _ _ ._7 .1

- - L cz

v. _ _v.

fkIff;i hk o v

. lS. i j o n b

e ?Im Kgi

':3 =e o a c pa l,) { f) l 0l  !! k T' *

, _, g- a- , , - y r + ,

lE

%4 .v a y soa" h,w wi!ey! . n (j

'4 E'  ; S i <--L '

o> c, l B38 j.'aJQ [ h u MO 319E

'i i; f8y - .

I=

j}

.4 .

!!@I'fj!M!no'j. ao E

@$m

~

l 3

7'n l4 I ' $ ) i FI ~

18  %~ *N,k.

I &i-i

.f'J, i" F! lx aoff ,

~

']; ' 0' A

0

'r Ie/ J_ "

dem -

o 9sz' u

. s d,,e l a- ,,

e- rn J, s a

gt" .

- t @ g g,.

M p a

\ c.s i,.! r .'

p'rw

. e e

. tu7 3 i ,

. c. -J 0N i t ii

/01> lr < uu -

3 /

t . 1 4, a

3 ma a e; p31 g3 t g: cab 21 ;i eq p,!e gi o '- 9 ly a!.*Id k bI h hi s. %%{ r 4

30

(.

~ ' ,

y ,.p -

4s s- q m ,

4, - ., o r d[; },, - 2

.o g k . gr .

-[', , t.h'.,

1 W

l p .' #1 ' ~.y i h

-.<- 7{ {

,e:9 ,;i y 3's. (4 l ;n,__ j y .

3'.

^

4m. j  :' j 5.- '.,gsr l_ 10. 4; . al .. d g **

Z

,py .. -

9.y( . g y ,. ,

.j ,

. ~o 3 l y -t u 9, g,g 4 m;Hw yt 1;+ . .. 3 .

. :

n /z v -d go

' )}dqo(:.4._ y p ;;

'c f .y .

( , s, <C

-i

_'_1 '

ye w _u H: -i

[i-

-] 1 m

y~ i y@, vgis  % W gj

. i

-' a L. .N,i - +

gi g j (gj r g o, )

(/

y'~ .5 l,- h ._ _

_v_

• , , ] .

f

.g 4' -

,6 '

W ni-

,=

~ { .)r

[-w- - T,f;..

. LY E J

']. d

.D .l

/[' l (p

) <,. r u >

.t. N ar

~,' ,<. J ' g ';

'. (O hl w

j

• E 'I a 4

> n  % r 1 .!+.= - =#- 2 r ,

N *

'( u Y,f_._f.#.

r 9 .. f' ' l vl g d a

,c d 8

• /at >'

gi sg i

[ s, y }s.,

, Rf l

g 1 j 'd ll '

l

, .. n  :

a p f.i #g '

.f . b;ih t- #o "

y y y}'hL ]  ?- (s ti 8

5; t

l0 -.Q . .. . .

e0 3 J

'1' .

3 TO b 4 t,.$ . .

A " n.

h h;'.I t .N i g[.]+ ,g n 0 h.Ul:

w .e,

, ef y

5 ,gs P -

g  :.. I

'A  ; (9 t) kh f

+

" M . FCr61- -

C M - rr =4p-- >p5 < d y:h.

,, s v, P1 y

=J

! si ist m if J!

w!.l m.

@ mre a

s .

.s W@ m! i -

~, .  : o, gg4 ly t f1 -'

$4, g4 --

. f;l l y $Th_ bibbi.Ldti[]; ,

~li f in '

p%e J -

gi L' a t 5 g 5 2 *z I e

{ m.

d

.\. .-. ; Lg '

a g p; e

l' w _,

3

, !una m

31 2

_.--...-..=-!=.=

~

l

={- . .

J

7;.7 7 _ - ---

h  !

y n:

F---

(

l I.

,V l

l

,. 1 l.

l -{

- i, 1

Attachment B

~

Statement of the ALARA Policy 1

l l

i 1

i i

1 i

l, 1

l 32

'v-4 L

L As Low As Reasonably Achievable w Commitment and Implementation The administration of The Ohio State University remains committed to the ALARA pnilosophy, a p,hilosophy which has always been a major part of tne University's Radiation Safety Programs. The Vice President for Health Services maintains an Office of Radiation Safety, whose-responsibility it is to develop and monitor ALARA concepts, and to which the autnority to enforce such concepts throughout the University has been delegated. .The Office of Raciation Safety is supported in these efforts by tne Vice President's Office, and by tne University Radiation Safety Committee and the Human Use Radionuclides Committee. A tnorough description of this organization is presented in Item 7 of this application.

Implementation of the ALARA philosophy at The Ohio State University completely incorporates all elements of tne model ALARA program published in USNRC Regulatory Guide 10.8, Proposed Revision 2, August, 1985, Appendix G. Most elements are explicitly discussed throughout this Application for Renewal of License No. 34-00293-02. The remaining elements are embocied within local standards and procedures.

By his signature to'the NRC Form-313M, the The Vice President for Health Services affirms tne continuing commitment of tne administration of The Ohio State University to ALARA ss exprassed in 10CFR20.1(c).

The Ohio State University ALARA 33 License No. 34-00293-02 Application for Renewal Page ALARA-1 of ALARA-3 kM8m RB 1925 __. _ _ _ _ - _ _ _ _ _ _ -

o' Appendix A Gamma Heating in the Lead Shielding of the Main and BSF Thermal Columns 34 u__--__-

l a

l l

In order to estimate the effects of gamma heating on the lead shielding located within both the main and BSF thermal columns, an estimate of the gamma photon l' flux must be obtained. On page 103 of the Safety Analysis Report in Sectior.

l 4.3.2, it is stated that the gamma dose rate at the edge of the HEU-fueled 5 OSURR cots cperating at 10 kilowatts is estimated to be approximately 5 x 10

, R/hr, based on measurements made at low power using a dose-callbrated ionization chamber placed in an air-filled tube (Beam Port 1) at the edge of the core. Assuming that gamma dose rate is linear with operating power (as stated in this section) and that the LEU-fueled OSURR core will exhibit similar properties, thedoserateattgeedgeofthecoreoperatingat500kilowattsis estimated to be about 2.5 x 10 R/hr.

Now, Etherington [1] reports that gamma dose rate in air can be converted to gamma photon flux using the following:

1 D = (5.767 x 10-5)(p / )E gr R/hr.

Assuming that the average gamma photon energy is 1 MeV, and using the following data as reported by Etherington:

V /p = 0.0280 cm / gram (1 MeV gammas in air) we obtain:

f = 1.55 x 10 photons /cm /second.

Now, this gamma photon flux interacts with the graphite thermal column extensions before it reaches the lead shielding. As shown in Figures 3.22 and 3.23 on pages 84 and 85, respectively, of the Safety Analysis Report, the graphite thickness is 11.25" between the edge of the core and the lead shieldit_ issuming that the gamma flux emitted by the core can be described approximately as collimated plane source, the total collided and uncollided photon flux exiting the graphite extensions and interacting with the lead shielding can be estimated from; f = B rr (0)e~ ,

where r F') = photon flux at core edge estimated above, h,,= dose build-up factor for plane collimated source, x = graphite thickness, and p = total mass attenuation coefficient for graphite.

Again from Etherington we obtain:

U/P = 0.0636 cm2 / gram (using data for carbon at 1 MeV),

using a density of 2.62 grams /cm for carbon, we can calculate:

35

h

!L l

1 p= 0.166632 cm~ .

Using x = 11.25" = 28.575 cm, we obtain the number of relaxation lengths as:

Ex = 4.762 for 1 MeV gamma photons in carbon. Using Etherington once again for the values of B , we btain: ,

r B (px) = 7.508 (using values of B for water assuming 1 MeV gammaphotons,witlilinearinterpolation). 4 Data for water was used in the estimate of B above since the Z for water is  !

close to Z for carbon,- and build-up factors are smooth functions of Z. Using  !

all of these data.together with the earlier estimate for' gamma photon flux at l i

the edge of the core, we obtain:

f = (1.55 x 10I3)(7.598)(e~ * ) l 2 2

= 1.007 x 10 photons /cm /second.

This is the incident flux of 1 MeV gamma photons to the lead shielding.  ;

Etherington reports that the rate of heat deposition in the shield from  ;

collided and uncollided gamma photons can be obtained from:  !

1 l

h=(1.602x10-6)(p /p}EgaB ( t)f ergs /second/ gram, with  ;

E = gamma photon energy (1 MeV),

9 p, = energy absorption mass attenuation coefficient, l t = shield thickness, i Ba(F at ) = energy absorption build-up factor, and

~M p T =f (0)e , with )

i p= total mass attenuation coefficient,  !

I

)

f (0) = incident flux on shield being heated. I h

For the lead shield, Figures 3.22 and 3.23 show a thickness of 3", or 7.62 cm.  !

Etherington reports that for Icad:

p/p - 0.0684 cm / gram, 3

assuming 1 MeV gamma photons. Using a density for lead of 11.4 grams /cm , we obtain:

p = 0.77976 cm~ ,

j i

36 ,

I I _________J

>re , in which case we have: 2 b=[(Sr)(7_,-pec)f(4Pca y 2 x g) This condition is met for a point located at the top of the reactor pool, which is about 15 feet above the top of the core, and 15 feet >> 7.5 inches. The 60 , value for pc. the core attenuation factor, accounts for absorption of gamma rays by the core materials. For a completely dry OSURR core, the only absorbing materials.will be aluminum and uranium. For conservatism, wey will is i.e., p = 0.166 cm at 1 assume MeV. that This isp"consequal ervativetobecausethat for pure aluminum, it neglects the strongeE absorbing effects of the uranium and silicon contained in the fuel meat, and assumes all gamma rays are gmitted at 1 MeV. Now, using a flux-to-dose conversion factor of K = 2 5.77 x 10 photons /cm -second per rad / hour, assuming-1 MeV photons, and substituting other appropriate values, we find that: D = 3451 rem /hr (at the pool top) converting rad /hr to rem /hr using a quality factor of 1 for gamma radiation. This analysis can be repeated for combinations of decay times and reactor operation times. Table F.1 of this document shows these results. The estimate of dose rate at the time the core is uncovered is the most conservative of all estimates. Obviously, direct exposure to the uncovered core leads to a potentially high dose commitment, and such exposures are to be avoided by appropriate action during such an event as total loss of pool water (i.e., evacuation of building). F.4 Scattered Radiation If the OSURR core is uncovered, the emitted radiation will exit the top of the pool and scatter off the metal ceiling of the building. This scattered radiation, sometimes called " shine", can give rise to radiation doses at other points within the building. Although such indirect exposure rates are much lower than direct exposure rates, it is more likely that individuals will be exposed to this indirect source, since under accident conditions personnel will likely move away from the pool top area and into regions exposed to the scattered component of the emitted radiation. Therefore, the analysis to follow will attempt to quantify this indirect exposure term. First, the nature of the radiation source must be specified. As before, for conservatism, we can assume that the reactor has operated at 500 kilowatts for an infinitely long operation, and that a loss of coolant event occurs that drains the pool in the characteristic drain time of 199 seconds. Further, it is assumed that the reactor is shut down on a low pool water level trip signal at the start of the loss of coolant event. Thus, 199 seconds is considered the decay time. After the core is uncovered, we assume that all the radiation is emitted from the top of the pool at an angle of 180 degrees to the floor of the building, i.e., a " beam" of gamma radiation is directed vertically upward towards the ceiling. We assume that the emitted gamma photons have an average energy of 1 MeV. The nature of the scattering material has a major effect on the magnitude and energy of the scattered photons. In the interests of conservatism, assume that the ceiling of the building is formed of an infinitely thick ordinary concrete slab, which will tend to enhance scattered effects. Normal ceiling materials will give rise to smaller scattered doses, since they are primarily (relatively) thin metal sheets. 61

p Operating Times ,

Decay' .< Times 1 Hour 10 Hours 30 Hours Infinite 199 Sec. 1541 2237 2477 3451 j 10 Min. 894 1554 1793 2766 60 Min. 251 738 963 1931 i 10 Hrs. 23 158 296 1215 24 Ifrs. 8 ~69 154 1019 I 48 Hrs. 4 33 83 886 7 Days 0.8 8' 22 .688 30 Days 0.1 1.4 4.2 512 ] 60 Days 0.06 0.6 1.9 445 l 90 Days 0.04 0.4 1.1 410 l i i .) l .i Table F.1 j Dose Rates In Rem /Hr From Direct Exposure To The Uncovered OSURR Core As A Function Of Operating History and Decay Time  ! {; 62 L_z--_______-. The location of the receptor must also be specified. For this analysis, we will assuma two different locations. The first location will be at the second floor of the building, about 21 feet from the ceiling. The second location will be at ground floor level, 35 feet from the ceiling. In both cases, we will assume minimum angular separation between the receptor and the point on the ceiling dira,ctly above the core. This assumption is also conservative, since locations removed from this point will have greater separation and thus lower dose rates. The point located 21 feet frca ,the ceiling is representative of a position an Individual might have when exiting the control room, or room 201, as they move towards the exit of the building. The ground floor location is representative of a person's position if they were located in the reactor bay area, moving towards either the front (south) or rear (north) exit. The scattered dose rate can be estimated from: D=N (Z/A)(19CEQ,)/[k(E)x ] A where p = density of scattering material (for ordinary concrete, this is taken to be 2.3 grams /cc), Z/A = ratio of the average atomic number of the scatterer to the atomic mass of the scatterer (0.5), I gG = product of the incident photon current and the cross section of the scatterer beams, in photons /second, k(E) = photon current-to-dose conversion factor, at the energy of the scattered photon, E - energy of the scattered photon in MeV, and x = distance from the scatterint, point to the dose receptor location, in cm. Now, in the above equation, Q, is given by: Q = 1/[p + p (cos0 /cos0 g )](6a/60) 3 where p g =attenuationcoeffgientinthescattererfortheincident photons (0.146 cm ), l p = attenuation coefficient in g e scatterer for the 3 scattered photons (0.292 cm ), 0 = incident photon current angle measured from the normal to the scatterer surface (assume O degrees), 0 = scattered photon current angle measured from the normal 3 to the scatterer surface (assume 25 degrees), 6c/60=di{ferentialKlein-Nishinascatteringcrosssection,in cm per electron-steradian [8]. 63 mm cl= . y; U?ll ~ ?% s 5 9 t 'n [Likewise, in the previous equation for b the.energyrof the scattered photon,. > E,.is given by: L .. ..E =.E o /[1+E,(3-cos0)/0.51]. [ . H,, , where- .E o = incident photon energy,"and t 6~=7 scattering angle =x-O n -03 Now, assuming that tiie emitted photons exit:the top of the reactor pool and are normally incident ~upon'the. ceiling at a point directly over the core, we can assume that B o = 0, and the expression for l oC can be " simplified" to - lC=SW o o , where' = Sy wrf/pc' S o with pothe core attenuation factor,given earlier S the-volume-distributed y source strength,'and rc the equivalent radius of-the cylindrical-equivalent n OSURR core, ,Also, we have.the geometry dependence, W, expressed as: s W = (w/2'- G)/2n, where y2(r2 _x 2 )+r 2 (r 2 2x )' G = arcsin (r2 x2 ) (72 y2) _ with x 'the half-width of the pool, yo the half-length of the pool, and ot the o . distance from the core to the top of the pool. Substituting appropriate values 'into the formula, which for the OSURR facility are: r o= 15 feet = 4.572 meters xn= 22.5 inches = .5735 meters yo = 64 inches = 1.6256 meters we have G = arcsin(0.9965467) = 1.487667 radians (85.24 degrees) 64 ?' L --- . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ g n-From this, we get: l W = 0.0132305 Now, using Sy calculated earlier, with the cylindrica}-equivalent core, we find that 3 I C = Sg W = 8.803 x 10 photons /second Next, the energy of.the scattered photons must be determined. The scattering angle of the scattered photons is given by 0=w-0 -0 but earljer we assumed that 0 was O degrees and 0 was 25 degrees. This 3 forces 0 to be 155 degrees. OsinF this in the equation for E, the scattered photon energy: E = Eg /[1+E,(1-cos0)/0,51] and assuming that the incident photon energy, Egis 1 MeV, we have E calculated to be about 0.211 MeV. Next, the Klein-Nishina differential scattering cross-section is estimated. This parameter is given by the following equation [8]: 60/60 = (re /2)[E/E, - (Esin0/Eg ) + (E/E9 ) ] -3 where r is the classical electron radius, assumed to be 2.818 x 10 cm. Substit0 ting 0 = 155 degrees, E = 1 MeV, E = 0.211 MeV, and r* given above, we U obtain -27 l 6c/60 = 8.44 x 10 cm falectron-steradian. The term Q can now be evaluated from the expression given earlier: Q = 1/[ p + g(cos0 9/cos9 3)](So/60) l  ! assuming that 0 and 03 are 0 and 25 degrees, respectively, and that 6a/60 is that estimated Ebove, we have: Q,= 1.803 x 10~ cm / electron-steradian. 65 1 It remains to find the flux-to-dose conversion factor,.k(E), at the energy'of-l ' l -the scattered photon,.E. For photons of engrgy 0.211 MgV, the flux-to-dose conversion factor is taken to be 4.956 x 10 photons /cm -second per rem / hour. Using all of the values estimated above, the scattered dose equation can now be l used to estimate doses at various elevations, given an initial source term dependent on operating history and decay time. The worst-case assumptions.- once again, are for an infinite operating time, following by loss of pool water taking 199 seconds to drain. For these condjtfons, the-following estimates-are . provided: D = 0.41 rem /h'our at 35 feet (ground floor level) b'=1.14ren/hourat21 feet (second floor level) l-b=2.24 rem /hourat15 feet (pooltop-level) Table F.2 of this document shows dose rate estimates for other combinations of operation and decay times. Using the most conservative estimates shown above, it is clear that occupational exposure limits can be maintained if an l individual does not expose himself to direct radiation from the core, and removes himself from areas of relatively high dose in a reasonable time. Even if a person were located at the pool top during the' loss of coolant event,.they would have a little over 30 minutes of exposure at the initial exposure rate of 2.24 rem / hour before they exceeded their allowed quarterly, dose limit. This is a conservative estimate'because the dose rate would be decaying during the exposure time and thus would be lower after the postulated time. But, even neglecting this offect, 30 minutes is much more than sufficient time for a-person to remove himself from the pool top area under normal circumstances. Typically, it takes just a few seconds to move from the furthest point in the ~ pool top area away from the' stairway to the steps at a reasonable pace (there is also a ladder located at the northwest corner of the pool top area which could be used if a person did not want to walk to the stairs). Assuming a 10 second exposure time at the highest dose rate estimated above, a total whole-body dose commitment of about 6 millirem would result. For persons at other points within the Reactor Building, exposure rates would be much less. For example, a person stanning at the second floor level outside the control row would be exposed to a dose rate of about 1.14 rem / hour. Now, assuming that this person were a member of the general public, they would have about 6 minutes of exposure time before they exceeded their allowable quarterly exposure. Based on evacuation drills conducted regularly at the Reactor Building, a typical evacuation time from the second floor to the first floor is about 60, seconds. This results in a dose commita:nt of about 19 millirem whole-body exposure. Once the first floor lev: !s reached, the dose rate drops to about 0.41 rem / hour. Again, assuming Inat evacuation time from the first floor to the outside takes about one minute, a total dose of 6.8 millirem of whole-body exposure results. Adding this to the dose total for moving from the second floor to the first floor, a total whole-body dose of 25.8 millirem 66 p i h-I' Operating Times p. ! Decay Distance 1 10 30 Hour Hours Hours Infinite. Times From Ceiling 35 feet 0.184 0.267 O.295 0.411 190 sec. 21 feet 0.510 0.740 0.820 1.14 15 feet 1.00 1.45 1.61 2.24 35 feet 0.107 0.185 0.214 0.330 10 min. 21 feet 0.296 0.515 0.593 0.915 15 feet 0.580 1.09 1.16 1.79 35 feet 0.030 0.088 0.115 0.230 60 min. 21 feet 0.083 0.244 0.319 0.639 15 feet 0.163 0.479 0.625 1.25 35 feet 2.75x10-3 1.89x10-2 3.53x10-2 0.145 10 hrs. 21 feet 7.64x10-3 5.24x10-2 9.80x10-2 0.402 15 feet 1.50x10-2 0.103 0.192 0.788 35 feet 9.95x10-4 8.23x10-3 1.83x10-2 0.121 24 hrs. 21 feet 2.76x10-3 2.29x10-2 5.09x10-2 0.337 15 feet 5.42x10-3 4.48x10-2 9.97x10-2 0.661 1 { l 35 feet 4.38x10-4 3.95x10-3 9.85x10-3 0.106 48 hrs. 21 feet 1.22x10-3 1.10x10-2 2.14x10-2 0.293 15 feet 2.39x10-3 0.15x10-2 5.36x10-2 0.575 Table F.2 Dose Rates (rem /hr) From Scattered Radiation At Various Positions in the Reactor Building 67 i l results. This is a reasonable estimate of dose commitment for evacuation from the second floor of the building to the outside under loss of coolant conditions. This dose total is within the allowable quarterly exposure limits for non-occupational exposure. For persons outside the Reactor Building', the radiological consequences of the postulated loss of water accident are less severe than for persons within the building. The distances between the scattering surface are larger, which attenuates the source strength of the scattered radiation. For example, for a person at a distance of 50 feet from the scattering surface, at an angle of 50 degrees from the normal to the scattering surface, assuming an infinitely long reactor run at 500 kilowatts and a 109 second decay time, the dose rate is on the order of 150 millirem per hour. Such a relatively low dose rate allows ample time to clear the immediate area of the Reactor Building and to limit access to the area until the termination of the emergency, as per guidelines in the approved facility emergency plan. F.5 Direct Radiation Through the Pool Walls Loss of water in the reactor pool reduces the effective shielding provided by the water-concrete composite shield around the core. An estimate is made, in this section, of the dose rate expected from the uncovered OSURR core through the concrete pool walls. The source term estimated for the OSURR core, as calculated in sections F.1 and F.2, will be used once again. For the volume-distributed cylindrical core, we have: Sy = 3.82 x 10 MeV/second/cm , assuming 1 MeV gamma photons. Now, the dose rate for a cylindrical source shielded by an infinitely long slab shield is given by [8]: 2 -b b=[k(E)ES,R/4] Aeg 2i[K(033,b21)*K(021,b21)]/(a+Z ) i=1 where k(E) = flux-to-dose conversjon factor at photon energy E. S = volume-distributed source term, y R = cylinder radius, A = Taylor-form buildup coefficients, 3 K,Z = functions to be defined later, a = distance between edge of cylinder and the point at which the dose is received, 68

m. .. ...i ... . .

0 = angular separation between dose receptor location and edges of cylinder . , t b21 = b3+u11 s 2 and b) = total optical thickness of the shield, p = attenuation coefficient of the source material. s Now, to estimate the attenuation coefficient of the uncovered core, the composition of the core must be considered. The dry OSURR core is a heterogeneous mixture of metal and air. If we assume that the metal portion is pure aluminum (which is a reasonable assumption, given the relative amounts of aluminum versus non-aluminum components), it remains to find the ratio of aluminum to air in the dry core. If the problem is viewed at the level of an individual fuel assembly, a gamma ray passing through a fuel element from one side to another perpendicular to the fuel plates would encounter a series of aluminum plates and air gaps. The plates are 0.05" thick, while the gaps have a nominal thickness of 0.116". Thus, the gamma-ray would "see" a total aluminum thickness of 0.9" (18 plates each .050" thick), and a total air thickness of 2.088". However, the aluminum fuel elements also have aluminum side plates. The nominal total cross-sectional area of a standard fuel element is 3" by 3". Of this 9 square inches of cross-sectional area, the gap between fuel plates is a nominal 0.116" thick by 2.624" wide, or a total gap area of about 0.304 square inches. If there are assumed to be 18 gaps in each element, the total air gap area per element is 3.521 square inches. The aluminum volume ratio is thus about 61%,-while the air volume is 39%. Assuming a linear attenuation factor of 0. M cm -3 for aluminum (at 1 MeV) and 7.66x10-5 cm -1 for air (at 1 MeV), the efft attenuation factor for the dry core can be estimated by: Ya

• sal "Al

• I air Unir' where the f's are the volume fractions noted above and the linear attenuation factor for the pure materials as noted above. This leads to an estimate for the core:

ps = 0.101 cmd. Now, we must determine the value of the function Z g . This function has a form dependent on the distance between the edge of the cylindrical source and the point at which the dose is received, denoted as "a" in the above analysis. For this case, we will assume that the dose is received at a point at the outside edge of the pool wall. The wall thickness is 6 feet, plus 1 foot, 10.25 inches from the inside of the pool wall to the edge of the core. Thus, we have: a = 6' + 1' + 10.25" = 94.25" = 239.395 cm l Now, the ratio of the cylinder radius, Roto the separation distance a is thus: a/Ro= 239.395 cm/19.05 cm = 12.57 69 i: Knowing this, values for Z3 can be obtained from standard tables, which list values of p Zs 3 as a function of pu 1Eo. Now we have psi given by: o 9s1 " II + C 1)Vs' where pg is as defined and calculated earlier, and ai are Taylor-form buildup coefficients. For conservatism, we will assume that the'a's are those for pure aluminum at a gamma photon energy of 1 MeV: , a 3 = -0.0682 a 2 = -0.02973 Using these we calculate: -I 9s1 = 9.411 x 10-2 cm cm -1 Ms2 = 9.800 x 10-2 Now, looking up the functional values for 23 as a function of ps1Ro, using an Ro of 19.05 cm, we have: pas 1 = 0.9465 p sZ 2 = 0.977 Dividing these by the estimates for ps1 and ps2 given above, we obtain estimates for the functional values of Z3 : Z3 = 10.06 cm Z2= 9.97 cm These values can be thought of as equivalent depths of line sources imbedded within the 'a ly of the cylindrical source giving rise to the observed dose rate at the specified location. These are almost the same value. If we round off to Z3=Z2 = 10 cm, we can add this to the value for a and estimate the angle of separation. We have the angle expressed by: I 0 = arctan[h/(2(a+Z))] where h is the height of the cylindrical source, and a and Z as defined above. Using h = 77.62 cm as estimated in section F.2, and a+Z = 249.395 cm, we have 0 = 8.85 degrees This is reasonable, given the relatively small size of the cylindrical-equivalent core and the relatively large separation distance. 70 Now, we must specify tho values for'the other Taylor-form buildup factors. [ . denoted by A3. Generally, one looks up the value for A3 in standard tables, and calculates A2 from: L A2=1-A3 For barytes concrete with 1 MeV gamma rays, we have: A3 = 23.014 , A2 = -22.014 Now, the total optical thickness of the shield, denoted by b , is 3 given by: b3 = pt, where t is the shield thickness (6 ft.=182.88 cm), and p is the linear attenuation coefficient for barytes concrete. Looking up the linear attenuation coefficient in tables, we have: < p = 0.214 cm~I, assuming again 1 MeV photons. This leads to an optical thickness of by - (0.214)(182.88) = 39.136 fr m the earlier definition: Now we can calculate b21 and b22 b21 = b3+UsiZ i .Using appropriate values obtained earlier. We have: ~ b21 = by+Vs1 21 = 39.136 + (9.411x10-2)(10.06) = 30.136 + 0.947 = 40.08 Similarly, we have for b22: D22 " D1+Es2 Z2 = 39.136 + (9.8x10-2)(9,97) = 39.136 + 0.977 = 40.11 71 _-_ - ______ __ - h p f. ~ . :: Dow. if the dose. receptor po' int'is located at the midline'of the cylindrical- ' source, the dose rate equation reduces to. 2' -b ib='[k(E)ESR/4] y9 Ae j 2i[2K( 03,b2i Il/I"*E I) i=1 'since the argular dependencies on O g have been replaced by: O =0 lj 21 " i f Further since Z is approximately equal to 2 , we can assume that 0 -=0 , and 2 3 l' .both are about 8 85 degrees. .Likewise, since b 3 is approximately equal o i b 22,andbothareabouttequalto40.1,andthelunctionK(0,b)isaslowly-varying function of b-for a given 0, we.can rewrite the dose rate equation-as follows: 2 ~h b=[k(E)ESR/4][4K(0,b) y9 Aej 21/(a+Zj ) 1=1 2 ~ = k(E)ES Rg /K( 0,b) 'Ag e bi/(a+Zj ) i=1 with 0 = 8.85 degrees, and b = 40.'1. Assigning values to these two parameters allows estimation of the value of the function K(0,b). The formal definition of:K(0,b) is as follows: b K(0,b) = e F( 0,b), where F(0,b) is~the Slevert secant integral [8]: 0 l' F(0,b) = e-b sec e' de ' JO ~ But, since the Slevert secant integral is a function of two variables with a functional dependence in exponential form, it is difficult to tabulate in a reasonably abbreviated form-in a mesh fine enough to allow linear interpolation 'between dependent variable values. Thus, the function K(e,b), which varies such more slcwly, is defined as above and listed in standard tables. Thus, looking up the value of K(0,b) for these two parameters in standard tables, it is seen that a good estimate of K(0,b) would be about 0.132, which interpolates the angle between 8 and 10 degrees, and predicts a value at a b of 40.1 by linear regression of values of K(0,b) at 8.85 degrees at values of b at 36 and

38. Again, this is a reasonable interpolation since K(0,b) varies quite slowly within~the range of tabulated values, q 72 e

n - - - - p h-A. Finally, we must specify a val'ue for k(E),Lthe flux-to-dose' conversion factor. Assuming 1 MeV photons, the tables list a value for k(E) of 1.98x10-6 res/ hour , per MeV/cm 2 -second. This is 34kely a conservative assumption since_ gamma-photons penetrating thezpool. Is:and-exiting into the regions-surrounding the pool will have an energy: spec .s which will be considerably degraded'from the average assumed' fission product energy of 1 MeV. We can'now evaluate the dose rate equation directly using: values derived to- -this point. Substituting appropriate values into the equation we find that for' the postulated conditions, D = 1,49 x 10-11 rem / hour = 15 pico-rem / hour. Obviously, direct radiation from the uncovered core penetrating the pool walls wil.1 not be of concern under-the postulated circumstances. This conclusion- .should be obvious,'since-it was shown in Chapter 7 of the Safety Analysis Report that direct radiation from the core operating at 500 kilowatts penetrating the pool walls was not of concern, and certainly the source term = for the shutdown reactor core is smaller than that for an operating core, although.in the case of the uncovered core no credit can be taken for gamma i absorption by water in and around the core. j i i I 73 1 d r. p r a. l' L / Appendix G References I- ?. I l u, 74 i l 3 - - - - - - - - _ _ _ _ . _ _._,.._m._, __ h !E ~ Fu [R ): [3 . Harold Etherington, Nuclear Engineering Handbook,,First Edition, McGraw-E Hill Book Company, Inc...New York...NY (1958) '

2. R. Van de Voorde, " Effects-of Radiation on Materials and Components".,

CERN-70-5, European Organization for Nuclear'Research (1970): I= 3. R. Van de Voorde and J.:Rustut, " Selection Guide to:0'rganic~M'terials a for Nuclear Engineering", CERN-72-2,; European Organization for Nuclear .s _ y r. Research (1972)

4. J. F Wett, Jr., " Surface Temperature of Irradiated ORR Fuel Elements b, Cooled in Stagnant Air", USAEC. Report thn.ORNL-2892, Oak Ridge National'

- Laboratory, Oak Ridge, TN (March 23, 1960);

5. C.-C.1 Webstera " Water-Loss Tests in. Water-Cooled andiModerated Research(

Reactors", Nuclear' Safety, Vol.:8,'No. 6 (Nov.-Dec', 1967) . ~ L .6. . John R. Lamarsh, Introduction j:0, Nuclear Engineering,' Addison-Wesley-Publishing Company, Inc., Reading, MA (1975) 7, David H. Slade, Meteor 91ogy and Atomic Ene;*v 1968, USAEC Technical Information Center, Oak Ridge, TN (July, .0*Is) -8. Anthony Foderaro, The Photon Shielding Manual, Second Edition, The - Pennsylvania ~ State University, University Park, PA (July,' 1976) s s 'h 75 ___.___..L._m - _m __ _____.__m_ - _ _ _ _ _ _ ___m._-_u___._u-______--._____m__-____m-m__. __. __- __ - __ -__ _ - _ - _ _ _ . _ _ _ _ _ . _ _ _ _ _ - ,, W .b;i ,

k. .

7,; i 1 e.; g. n 1 .,-.- .i l l i i ) Appendix H Bibliography ) l't . 'N 1 76 _ _ _ - - _ _ _ - - _ - _ _ _ _ - _ _ _ _ _ - _ _ _ _ - _ - _ - - - _ _ = 1 Reed Robert Burn, Research t Trainingt Test, and Production Reactor Directory, Third Edition, The American Nuclear Society, La Grange Park, IL (1988) Harold Etherington, Nuclear Engineering Handbook, First Edition, McGraw-Hill Book Company, Inc., New York, NY (1958) 1 " Final Safety Analysia Report for the Lowell Technological Institute Reactor", J The University of Lowell, Lowell, MA (September 1973) Anthony.Foderaro, The Photon Shielding Manual, Second Edition, The Pennsylvania State University, University Park, PA (July, 1976) E. E. Lewis, Nuclear Power Reactor Safety, John Wiley & Sons, Inc., New York, NY (1977) John R. Lamarsh, Introduction to Nuclear Engineering, Addison-Wesley Publishing Company, Inc., Reading, MA (1975) l' " Revised Safety Analysis Report for The Battelle Research Reactor", Battelle Memorial Institute, Columbus, OH (March 26, 1974) " Safety Analysis Report and Technical Specifications for The Ohio State University Research Reactor", The Ohio State University, Columbus, OH (September, 1987) " Safety Analysis Report for the University of Missouri-Rolla Reactor", The l University of Missouri, Rolla, MO (September 27, 1984) i David H.' Slade, Meteorology and Atomic Energy 1968, USAEC Technical Information Center, Oak Ridge, TN (July, 1968) "TRIGA Reactor Facility Nuclear Engineering Teaching Laboratory The University of Texas at Austin", The University of Texas at Austin, Austin, TX (November, 1984) R. Van de Voorde, " Effects of Radiation on Materials and Components", CERN 5, European Organization for Nuclear Research (1970) .R. Van de Voorde and J. Rustut, " Selection Guide to Organic Materials for j ' Nuclear Engineering", CERN-72-2. Eur.opean Organization for Nuclear Research (1972) l C. C. Webster, " Water-Loss Tests in Water-cooled and Moderated Research L ll Reactors", Nuclear Safety, Vol. 8, No. 6 (Nov.-Dec. 1967) J. F. Wett, Jr., " Surface Temperature of Irradiated ORR Fuel Elements Cooled in Stagnant Air", USAEC Report No. ORNL-2892, Oak Ridge National Laboratory, Oak Ridge, TN (March 23, 1960) l-i 77 f - _ _