ML20197F408

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Forwards Addl Info Requested for Renewal of License R-57, in Response to
ML20197F408
Person / Time
Site: 05000131
Issue date: 12/17/1997
From: Claassen J
DEPT. OF VETERANS AFFAIRS MEDICAL CENTER, OMAHA
To: Alexander Adams
NRC (Affiliation Not Assigned)
Shared Package
ML20197D519 List:
References
636-151C, NUDOCS 9712300212
Download: ML20197F408 (26)
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Text

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DEPARTMENT OF VETERANS AFFAIRS Medical Conter 4101 Woolworth Avenue Omaha NE 68105 December 17,1997 in nepiy nefer To: g

? Alexander Adams, Jr.

3 Senior Project Manager Non-Power Reactor and Decommissioning:

Project Directorate

- Division of Operating Reactor Support Office ofNuclear Reactor Regulations US Nuclear Regulatoly Commission.

. Washington, DC 20555-0001

. Re: Docket 50-131

^

Dear Al:

Enclosed you will fmd two copies c,f the additional information requested for the renewal of our Facility Operating License R-57. The original has been sent to the Document Control Desk.

It is my hope that I have adequately addressed the questions and comments outlined in your April 2,1996 letter. The Safety Analysis Report has been given major revision while the Technical Specifications have been changed slightly to ensure that nomenclature is more consistent and to propose changes in fuel element inspection frequency and reactor period limit. Any further changes to the Technical Specifications will best be accomplished during your site visit of our facility.

The Safety Analysis Report is enclosed as a complete document with the exception of Fig. 2.1 and Fig. 2.7. : There were numerous changes and additions to the S AR. For purposes of clarity, I have included two attachments to this letter. Attachment #1 is an itemizeel summary list of additions made to the SAR, to include answers to the first set of questions (RIA No.1) which were not previously incorporated into the license. Attachment #2 is a summary list of changes and additions to the S AR in response to the second set of questions (RIA No. 2).

Sincerely, AW.

t JOHN P. CLAASSEN Reactor Manager

Enclosures:

Reply to Questions (Attachments 1 & 2)

Tech Spec Changes Radiation Protection Program :

-SAR e.

9712300212 971217

^PDR ADOCK 05000131 p

PDR

Page 2 of 13

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ATTACHMENT #1 Additions incorporated into the license with respect to the first set of questions that had not been previously added As mentioned previously, the answers to RAI No.1 have been incorporated into a smooth revised SAR. In this attachment, is a summery of the answers that were not incorporated into the license from RAI No.1. In some cases, it was necessary to modify the some of the answers. These modifications are also listed below. A justification was added when it was appropriately necessary to clarify the reason behind these modifications.

Question #4 The following information has been assimilated into the license in SAR Section 1.2 Table 1-1:

Reactivity loss per year (averaged from ""10 to 1994) = 0.02% Sk/k/yr.

Justification:

Supplies additional information with respect to actual facility operation.

Question #6 The following information has been deleted from the license in SAR Section 3.1:

...and has been determined to be 6.9 x 10-3 mrads/yr at ground level.

Justification:

/

Doses at ground level in Appendix A have been recalculated. This value in no longer valid.

Question #7 The following information has been edited and added to the license in SAR Section 3.1:

If the exhaust fan stops while the reactor is operating the hospital HVAC engineer manning the central system will notify the maintenance man on duty to repair the fan. During of t hours the hospital maintenance crew will also respond. if the SRO on duty determines that there is a potential haa.rd; the reactor will be scrammed immediately.

Justification:

This change allows the operator more control over operatici.4. Even if the exhaust fan was inoperable; the amount of radioactive effluents generated by the reactor (see Appendix A) is very small. As a result, the loss of exhaust wotid not constitute an immediate hazard.

_ Question #8 Replace the following information in SAR Section 3.1:

There are no gaskets, packing or other matetials to prevent or inhibit air exchange between the reactor room and other spaces within the hospital to which the public has access. However, while the reactor room exhaust fan is operating there is negative pressure in the room prohibiting reactor room air nom entering the hospital basement. If the exhaust fan stops, supply air into the reactor room is shut off by means of an automatic damper in the supply duct. The operation of this damper is checked monthly.

With the Following:

De areas of potential air exchange are predominantly at the doorways (the doors are nonnally closed).

Here are no gaskets, packing or other materials to prevent or i..hibit air exchange between the reactor room and other spaces within the hospital to which the public has access. Breeches in the walls due to

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conduit, pipes, and other structures are sealed with concrete. With the ventfan off there is still a slight

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negative pressure in the reactor room from the two fume hoods.

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Page 3 of 13 Justification:

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This paragraph has been modified to provide additional clarity.

v Question #17(b)

The following information has been assimilated into the license in SAR Section 3.2.9:

The flow inlet pipe is 13 feet above the top of the core. In the event of a rupture in the cooling system the maximum amount of water lost would be to this level. However, the water level would most likely lose only a few inches before the skimmer began to suck air. This would effectively cause the pumping system to lose its prime.

Questiog#18 The following information has been assimilated into the license in SAR Section 3.2.10:

In the 38 year history of the facility there has not been any observed changes in the strength or integrity of the fuel element components, the tank or the lining material due to neutron or gamma radiation damage. Four fuel elements have been examined each quarter by removing each separately and placing it in a device that allows us to examine it in detail underwater with a 20 power telescope each element has failed to show any significant change. Likewise, we have not observed any change in the reactor tank. Consequently, there is no reason to believe that there will be a breach ofintegrity of the components during the requested license extension.

Question #19 The following information has been assimilated into the license in SAR Section 3.2.10.1.1:

ne reactor tank has never overflowed.

_ uegtlon #21(d)

Q O

Question #14 v'

RSO has been changed to SRO in SAR Section 3.2.6.2.

Question #15 Replace the following information in SAR Section 3.2.6.3:

Unauthorized use or inadvertent operation of the central thimble is prevented by the fact that the Reactor Operator is only 17 feet away and in direct visual contact with the top of the central thimble. ne central thimble irradiation device can only be inserted or removed by the reactor operator or his designee. As is always the case ALARA will be complied with.

With the Following:

Unauthorized use or inadvertent operation of the central thimble is prevented by the fact that the Reactor Operator is only 10 feet away and in direct visual contact with the top of the central thimble. The central thimble irr.diation device can only be inserted or removed by the reactor operator or his designee.

Justification:

This change allows for more accuracy and less redundancy.

The following information has been assimilated into the license in SAR Section 4.2:

e.g. the exact linear rod positions can be reproduced by 0.2%. Here is no position indicator for the safety rod. It must be either all the way down or all the way up or the interlock will not allow the shim or regulating rod to be moved.

Question #22f aMb)

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The following information has been assimilated into the license in SAR Section 4.3.1.1 (2):

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Exampics of" unacceptable" circuit performance are listed on page 4-23 under Stack Errors. Rese errors can effect the operation of the NM 1000. Failure of stack 1-9 to take longer than 1.5 sec. will

Page 4 of 13

. cause the watchdog timer to be tripped. Internal diagnostics and self tests are performed continuously

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in the NM.1000, whether the reactor is secured or at power, to insure operation integrity. RAM, ROM 4

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and battery backup RAM are continually monitored and tested.

Question #22c Information has been assimilated into the license in SAR Section 4.3.1.1 (2).' The table has been changed and has been added to the license. Please refer to Attachment 2, #7.

Question #23 -

The following information has been assimilated into the license in SAR Section 4.3.1(6):

There is no apparent indication of any rod shadowing or flux density shifts that effect the response of the two neuvon detectors. He detecto s are individually calibrated on a yearly basis.

Question #24 The following information has been assimilated into the license in SAR Section 4.3.1.2:

They also directly compared, by measurement, the time required for detection of signals and low level, high level and period scrams of the NM 1000 with the TRIGA analog system. ne times were found to be equivalent.

QuestLQ!}.31 Additional information has been added to the license. Please refer to Attachment 2, #8.

Question #26(a)

The following information has been assimilated into the license in SAR Section 4.3.2(1):

The water monitor is calibrated so that when it u removed from the box (Fig. 3.8) and exposed to a

,/mh 100 mr/hr field, a full scale reading on the meter is equivalent to 0,1 Ci/cm' of 10-min-old fission

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. products in the water system. The alarm is set for 80% of this value. During the 38 years of operation the meter has not indkated over 2% of full scale. The water monitor reading is recorded on the daily check list before each startup.

Question #26(b)

- Section 4.3.2(3) has been reworded.

. Ques *lon #26(c)

SAR Fection 4.3.2(5)b has been changed.

Question #28 Beginnir'g of Chapter 7 has been changed.

Question #29a.

The following information has been assimilated into the license in SAR Section 7.1:

Since most of the liquid waste from our neutron hetivation procedures involve short half-lived isotopes the waste is stored until it reaches background. If release into sanitary sewemge is indicated it will be done so as to comply with 10 CFR 20.2003 and the values recorded in the Omaha VA Medical Center -

. sewerage disposal records, if assay of the wast; is required it will be done with calibrated radiation survey meters or a well Ge detector coupled to'a multichannel analyzer.

Questlo'.i #29c The following information has been assimilated into the license in SAR Section 7.1.5:

[ )T Any separation of radioisotopes that may be involved in our neutron activation procedures will be

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transferred to and come under thejurisdiction of the Omaha VA Medical Center Type A Broad Scope 4

1.icense,

Page 5 of 13 -

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t Question #30s b

The following information has been assimilated into the license in SAR Section 7.2.1:

Release of radioactive components, etc. for unrestricted use may only be done by the Reactor Manager or a delegated Senior Reactor Operator es referenced in the facility ALARA Program which is approved by the Reactor Safeguards Committee.

Question #30c The following information has been assimilated into the license in SAR Section 7.1.7:

Ambient background is define as follows: " Radiation from costr';. wirces: naturally occurring radioactive material including radon (except as a decay product one m or special nuclear material) and global fallout as it exists in the environment from the testing of nucie r explosive devices.

Background radiation does not include radiation from sources controlled or regulated by the overseeing regulatory authority".

Qy19 tion #31d Please refer to Attachrnent 2, #13.

Question #32a Please refer to Attachment 2, #14.

Question #32b The following information has been assimilated into tne license in SAR Section 7.2.3:

c Defore a new experiment can be performed in the reactor a " Request for Neutron Activation" form

' I must be completed by the requester. His form lists the phase of the sample, tne requested irradiation

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time, the concentration of all elements which are or may be present in the sample, whether the sampic contains compounds which are highly reactive with water, potentially explosive, or fissionable material. The form also asks if the sample contains uranium, thorium, plutonirm, boron, lithium er eny other known high neutron absorbing element and how the sample will be encapsulated. This completed form together with an experimenta! protocol is submitted to the Reactor Supervisor. The Reactor Supervisor then completes a separste " Check List for Reactor Experiment Appmval form making an independer.t analysis of the safety of the procedure and certifying that the experiment is in compliance witie the Omaha VA Ho9 ital Technical Specifications ne experiment is then submitted to the Reactor Safeguards Committee for Approval. After approval the experiment is rehearsed with the investigator. During the first procedure radiation levels will be measured, observations of the -

sample before and after irradiation will be noted. if applicable, and an analysis made of the results.

Each time an experiment or irradiation is to be performed it must be approved on the Daily Experiment Checklist by the SRO.

9 Qggstion #33a The following information has been assimilated into the license in SAR Section 7.4:

While CAMS are designed to be air particulate monitors, both will detect *Ar and other non-particulate airborne radioactive raaterials. The efficiency of the NMC CAM has been calculated to be 4x10-' %.

4 The efliciency of the Eberline ANS-3A Cam has been calculated to be 1.23x10 % and consequently a 4

2 count rate build-up of 180 cpm /hr is equivalent u an air concentration of 2.28x10 Ci/cm.

Since the DAC limit for "Ar is 3x10-6 Ci/cm', both CAMS can detect the DAC limit of *Ar.

Realistically, we have not seen a rise of 180 cpm /hr in either CAM, which is consistent with our calculated argon emission as shown in Appendix A. Since it is not feasible to monitor such effluents from a 20 kW reactor in real time at the point of release, calculated releases will be 4x10%.

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Consequently, a count rate build-up of 180 cpm /hr is equivalent to an air concentration of 1.5x10

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pCi/cm'.

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- Pig 3 6 of 13 Question #33b L[

The following information has been assimilated into the license in SNR Section 7.4.1:

Both detectors are calibrated with a "Tc sandard of 0.005 Ci that has a 0.29 MeV beta. Since the monitors use a thin endwindow GM tube, a calibrated source #an E, greater than 0.29 MeV would have a greater efficiency and consequently resuh in a larger epm /cm' hr. Using the efficiency determined from the *fc standard the calibration of the CAMS are as fo!!ows:

NMC Recorder rise of 180 cpm /hr = lx10* pCi/cm' Eberline Recorder rise of 185 cpm /hr = lx t0* pCi/cm' Since all of the radioisotopes that we may encounter at the facility have an inhalation DAC ofless than lx10* as listed in Table i Appendix B to 10 CFR 20.1001-20.2401, a rise of I80 or 185 epm /hr would indicate that we had a possibility of exceeding the DAC limit. The above calibration values are for the present calibration and may vary from calibration to calibration.

Radionuclides would be identified by analyzing the filter with the calibrated Ge or GeLi detectors. Tbc Omaha VA Medical Center Reactor SOP lists the maximum weight of isotopes which could off-gas, sublime, or volatilize and produce an activity such that if 100% of the gaseous activity or radioactive aerosols escaped to the reactor room or the atmosphere, the airborne concentration averaged over a year would not exceed the limit of Appendix B of 10 CFR 20.1001 20.2401. Experiment approval is dependent on not exceeding these weight limits. The facility uses the Canberra ASCs Peak Search and Isotope Identification software together with the SAMPO 90 Interactive Gamma Spectrum Analysis

-package.

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' Pag? 7 of 13 3

_ ATTACHMENT #2 =

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L Answers to General Comments and Questions from RAI No. 2 General

1..Where applicable, I have incorporated all the answers to questions, substitute pages,-_

' appendices, and inserts into a smooth revised SAR.' Information, in the form of answers to--

c questions, witich should have been incorporated but did not assimilate into the license in i our March 1,1995 response to RAI No.1, has been added to the SAR and documented in f

. Attachment #1 to this letter.)

2a. Calculations for coolant rate through the core.have been revised in Appendix A of the SAR.

The temperature rise across the core was determined by experiment and thus leads to an Emore precise value for the in-core flow rate. The coolant rise time value was calculated

- theoretically for 20KW based on the value determined in ' Hazard Report for Torrey Pines

- TRIGA Reactor", GA-722. As a result, the two values are reasonably consistent with each-other.--

2b~ A values for release fraction of fission fragments have been changed to 1.5 x 108 These-changes have been revised in Appendix A and other applicable sections.

2c. The derived doses determined at the roof exhaust have been deleted. Calculations for the Most Exposed Worker (MEW), Most Exposed Member of the Public (MMP), and Nearest Permanent Resident (NPR) have been calculated for airbome emitters. Doses due to :

', f routine operation such as "Ar were calculated using NUREG - 0851. Doses due to fission 4

product gases were determined using Regulatory Guides 1.109 and 1.145 73.

Additional clarity has been made to the Technical Specifications with respect to applicable' sections of the SAR.

i Questions About Responses to RAI No.1 1.

Question #9 L

The following information has been assimilated into the license in SAR App. D, Section D.3:

It is not readily apparent that the activated products within the soil about the reactor tank would have any _

L method of transport to the water table. The reactor tank within a poured concrete basement floor. The closest area of exposed soil is a minimum of 28 feet radially form the top of the tank. :It is assumed that all the soil ~

activity was removed by the water.

The following information has been added to the license as SAR Section 2.4.1:

M.4.l - Groundwater liydrology _

l Groundwater generally moves in horizontal and latent directions. In determining the subsurface move.ne st of water, the actual trails are assumed to t svel smooth pathways known as streamlines.

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Thus, the water molecules are taken to travel directly through matter. The laminar flow rate v of L

l underground water is given by Darcy's equation.-

V= KX (Hi-H2)

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L

,\\v]-

where ; y - Laminar flow rate

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,, +.. - -,,

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1 iPage 8 cf 13 K = Ilydraulic conductivity.

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Hi = Total head measured by a piezometer at site I H = Total head measured by a piezometer at site 2 -

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2

. L = Distance between the ends of the piezometers

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'Ihe hydraulic conductivity of soil can be affected by temperature, ionic composition of the water,

and the presence of entrapped air. The density and vis.
osity of water changes with temperature. K values are normally expressed at 20*C. The ionic composition of water can change the K value via ion y exchange when exposed to porous material containing clay. In addition, the pores of those clays can j-be so small as to produce size exclusion for some of the larger ions. Entrapped air in the soil generally -

causes the K value to be less. Air can become trapped within the soil by a rise in the water table. It

. may also occur when colder outside water enters an aquifer.

For the purposes of our calculations the hydraulic gradient of flow will be assumed to be unity.

Therefore.

-v=K.

From the site of the reactor the ground water will flow to the south-west. Traveling downward by gravity through the relatively impermeable loess until it reaches the level ofimpermeable glacial till.

' As seen in the area map (Fig. 2 2), the Big Papillion Creek runs in an south-easterly direction approximately two miles west and four miles south-west of the site. With the water table troughing.

along this creek the underground water would migrate along the creek until it returns to the Missouri

, river south of Offut Airforce base. Once within the Missouri river the water would be readily available to members of the public for ingestion.

Glacial Till has a hydraulic conductivity in the range of lx10 " to 2x10 meters per second.

4 4

6435m + 2x10 m/s = 102.03 years.

f;k For comparison, the slowest rate is 6435m + lx10'" m/s = 204,052,511 years.

Thus, any radioactive isotopes produced via soll activation would decay out well before it was made available to the public for consumption.

References added to the end of SAR Section 2:

l 1.

Bouwer," Groundwater Hydrology", Mcgraw-Hill.1978.

2.

Todd," Groundwater Hydrology" John Wiley and sons.1980.

2. Question #10 l~

' The following information has been assimilated into the license in SAR Section 3.2.1:

The spent fuel storage pits are designed with sufficient spacing to ensure that the array, when r fully loaded, will be substantiairy subcntical. For comparison, actual measured multiplicatioli in an array of five fully loaded (19 elements each) storage pits of similar design yields a k, of 0.45 (dry);-

3. Question #12
k. has been changed to k,,in SAR Section 3.2.3..

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Pags 9 of 13

4. Question #16 &

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5. Question #21(a)

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The SAR Sections 3.2.8, 4,2, (4.3 Table 4.1)and 4.3.1.1 have been changed in order to provide more consistent nomenc!ature. Please substitute the following to replace TS Sec. 3.2.3, Table 1:

Tablei RequiredScrams Safety Channel Set Point Minimumnumber required Percent power 100% licensed power i

Linear power **

100% licensed power i

lon chamberpower supply

  • Loss ofnigh voltage 1

Fissioncounterpower supply **

Loss ofhigh voltage Console scram button Manual i

Magnet currentkey switch Manual I

Watchdogtimer Key software tasks takelonger i

than 1.5 sec

  • lonchamberanitlogsystem

" Fission counterdigitalsystem

,A and TS 3.2.4 Table 2:

1 Table 2 RequiredInterlocks SafetyChannel Minimum Number Function Required Startup I

Preventswithdrawalof allcontrol rodsif countrateC cps)

(Neutran count rate)

Log N (Period) 1 Prevents withdrawalof all control rods when periodis less than 3 seconds Siraultaneousmanualwithdrawal 1

Prevents withdrawal of two rods' Withdrawalof shim or regulating i

Prevents withdrawal rod with safetyrod not allthe way out or seated Withdrawalof safety rod with I

Prevents withdrawal shim or regulatingrod not seated

6. Question #21(b)

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SAR Section 4.2 has been changed so that the servo-system is not described in a way that Q

would indicate the need for a interlock. Failure of the recorder would cause the rod to drive out

Page 10 of 13

,7 l but the reactor safety' systems would respond with a scram at 100% power. Adding a servo

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interlock would introduce a " double" failure design.

' 7. Question #22(c)

The following section was rewritten for more consistency nomenclature. Please refer to SAR section 4.3.1.1(2)

The followingpre the Performance Specifications of the NM 1000

- Sensitivity Linearity Function.

{sps/nV)

EnDES

( %.) -

' Log / linear 0.2 2x10-' % to 120%

2 Percent power :

0.2 1% to 110%

2 8.- Question #25 When the NM1000 is used in a pulsing system the computer is dedicated to gathering other-

. data and the 1.5 sec. is a designed value. During development, when all the code had been-completed this value was used to allow all subroutines to update the programmed data and '

thus prevent spurious trips during pulsing. SAR Section 4.3.1.3 has been reworded.

9, Question #26(b)

Please delete TS 2.1 and replace with the following:

2.1 Safety Limits 2.1.1 This specification applies to the temperature of the reactor fuel.

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Objective:

The objective of this specification is to define the maximum fuel temperature that can be permitted wi*.h confidence that no damage to the fuel element will result.

Soccifications: The temperature in any fuel element in the OVAMC TRIGA reactor shall not exceed 500 *C under any conditions of opere1an.

Basis:

'A loss in the integrity of the fuel element cladding could arise from a buildup of excessive pressure between the fuel ard the cladding if the fuel temperature exceeds the safety limit. The pressure is caused by the heating of air, fission product gases, and hydrogen from the dissociation of the fuel-moderator. The magnitude of this pressure is determined by the temperature of the fuel element and by the hydrogen content. Data indicate that the stress in the ::ladding due to hydrogen pressure from the dissociation of ZrH. will i

remain below the ultimate stress provided that the fuel temperature does not exceed 1050 *C and the fur! cladding temperature does not exceed 500 *C, When the cladding temperature can equal the fuel temperature the fuel temperature design limit is 950 *C, (M. T. Simnad, G.A. Project No. 4314 Report e-Il7-833,1980).

Experience with operation of TRIGA fueled reactors at power levels up to 1500 kW shows no damage to the fuel due to thermally-induced pressures.

The thermal characteristics of aluminum-clad TRlJA fuel elements using ZrHu

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'2387, PNL4028, Credible Accident Analyses for TRIGA and TRIGA fueled moderator have been analyzed (S. C. Hawley and R. L. Kethren, NUREG/CR-(

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Reactors,1982). A loss-of-coolant analysis showed that in a typical graphite-reflected Mark i TRIGA reactor fueled with 60 aluminum-clad fuel elements (Reed ev,

Page 11 of 13 gN-College) the maximum fuel temperature would be less than 150 *C following

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infinite operation at 250 kilowatts terminated by the instantaneous loss of water.

V' These temperatures a.e well below the region where the a+ p + y to a + 6 phase change occurs in ZrHo (560 *C).

2.1.2 Limiting Safety system Settings (LSSS)

Anolicability: This specification applies to the reactor scram setting which prevents the reactor fuel temperature from reaching the safety limit.

Obiective:

The objective of this specification is to provide a reactor scram to prevent the safety limit from being reached.

Specification: The LSSS shall not exceed 20 kW as measured by the calibrated poveer channels. The LSSS which does not exceed 20 kW provides a considerable safety margin. One TRIG A reactor (General Atomics, Torrey Pines) showed a maximum fuel temperature of 350 *C during operation at 1500 kilowatts, while a 250-kilowatt TRIGA reactor (Reed College) showed a maximum fuel temperature ofless than 150 *C (reported by S. C. Hawley, R. L.

Kathren, NUREG/CR-2387, PNL4028 (1982), Credible Accident Analyses fur TRl(GA and TRIGA-Fueled Reactors). A portion of the safety margin could be used to account for variations of flux level (and thus the power density) et vanous parts of the core. The safety margin should be ample to compensate for other uncertainties, including power transients during otherwise steady-state operation, and should be adequate to protect aluminum-clad fuel elements from cladding failure due to temperature and Q

pressure effects.

}

.d SAR Section 4.3.2(3) has been rewcrded. Please delete TS 2.2.1 since the administrative shutdown at pool temperature 35'C is based on ion-exchanger resin degradation rather than fuel element integrity. Please renumber TS 2.2.2 Thermal Power to be TS 2.2.1. Also, renumber TS 2.2.3 Poolwater Level to be TS 2.2.2.

10. Question #26(c)

Procedures are available for determination of the CAM's efficiency for radiciodine.

Specifically referenced instrumentation in SAR Section 4.3.2 Sb has been omitted.

Please substitute the following to replace SAR section 4.3.2 Sb:

b.

A calibrated, continuous airbome radiation monitor (CAM) located in the reactor laboratory near the top of the reactor. The monitor can detect both gaseous and partic iate radioactivity. The morator also contains a charcoal filter to provide the capability of monitoring for radiciodine. The charcoal filter is not required for reactor operation. Procedures are available for determination of the CAM's efficiency for radioiodine.

11. Question #30f a)

SAR Section 7.2.1 has been supplemented to address this question. Release of radioactive components, etc. are authorized by the Reactor Manager or his delegate. This authonty is s,tecified in the facility Radiation Protection Program (Al. ARA) which is approved by the Reactor Safeguards Committee.

12. Question #31(a)

C The requirement for SCBA's at our facility has beer < 1oved. On March 25,1995, a letter was s.

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sent stathg that this requirement was not nec%sary. A letter was sent by representatives of the NRL. on 9/26/95 (TAC # M92036) acknowledging our changes. The reference in section 7.2.1 to SCBA was deleted in our March 1,1995 changes to our license.

Page 12 of 13 m

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13. Question #31(d) d Presently the OVAMC facility has two CAMS that read particulate radiation. The alarms are based on setpoints of activity (e.g.1000 cpm). They are not designed to alarm based on rise of activity (e.g.1 DAC = lx10" pCi/cm' = 150 cpm rise in one hour). The alarm setpoint mentioned 8

in TS 3.7.1 Table 4 (6x10"*pCl/cm per 1 hr particulate) corresponds closely to the 1000 cpm alarm setpoint during reactor operation of approximately one work day (6 - 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />). Currently both CAMS can be viewed from the reactor console and are periodically observed by the reactor operator during operation. Please add to section 7.2.1:

Ali of the radioisotopes that may be encountered at the facility have an inhalation DAC ofless than lx 104' as listed in Table 1 Appendix B to 10 CFR 20.100120.2401, sensitive to one DAC change in one hour is defined as being able to detect lx10"' pCi/cm) in one hour. At the. present time, lx10 '

d pCi/cm' is equivalent.a CAM tise of greater than 150 cpm in one hour or both of our CAMS.

14. Question #32(a)

A current copy of are Radiation Protection Program (ALARA) has been submitted as an enclosure to this letter.

15. Question #32(b)

SAR Section 7.2.3 has been reworded to clarify the term " effects".

The term " effects" is interpreted to mean abnormal physical alteration of the sample and its containing vial.

16. Appendix A page A-13 (N

The calculations involving the " building cross section area" were deleted in favor of the method 5

used in NUREG-0851 for determination of off-site doses. Relevant changes are documented in Appendix A, Section A.2 and summarized in Table 1.

17. Appendix A pagos A-13 and A-14 The calculations involving the annual doses of the MMP and NPR due to routine release ofAr were calculated by the method used in NUREG-0851. The changes are documented in a new section A.4 and summarized in Table 1.
18. Appendix A e pace A-19 All previous values have been recalculated. Additionalinformation summarized in Table
1. Clarification has been made regarding point of release, total activity released per year, and doses to the MEW, MMP, and NPR.
b. Daae A 19 MEW values are now given and summarized in Table 1.
c. Daae A 19 Section A.3 has been deleted. Final calculations are based on theory rather than using a experimentally determined correction factor. In addition, it has been noted in Section A.4 that previous attempts at measuring Argon-41 at our facility have had little success which would be expected in comparison to the theoretical value.
19. Appendix A pace 8

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a. Daae A-8 Equation 21 is no longer used.

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b. Daae A-8 A basis has been provided in Section A.3 of Appendix A to the SAR. The

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cose rate, from N-16, calculated at the top of the reactor tank is 7.00 x 104mr/hr.

Thus, the dose is several orders of magnitude less than the limits specifico in 10 CFR 20.

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20. Section B.2, Appendix _ B, pages B-5 and B-6

. a. L Where applicable, intemally consistent values have been used._ The release fraction l

of 1.5 x 10-5 is used consistently throughout this section.

' b. SAR Section B.2, Appendix B has been recalculated with methods and assumptions presented in regulatory guides 1.109 and 1.145. These calculations were based on more conservative values of atmospheric stability, windspeed, etc. as would be expected from -

postulated accident events, NUREG - 0851 contained nomograms for several isotopes of xenon and krypton.

However, not all of the pertinent isotopes were represented. As a result, a direct comparison of whole body doses to NUREG - 0851 would seem to be inappropriate.

21. Table B - 4 Appendix B Typographical errors have been corrected. The table was changed to display comparisons with -

the derived air concentration DAC.

22. Apaendix B A list of references has been provided for SAR' Append;x B.
23. Analysis in Anoendix C Additional calculations were added to include members of the public on the first floor above g

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the reactor. The summary in Table C-1 also reflects this addition. Appendix C was also reworded to address potential dous to members of the public outside the building.

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Page 1 of 4 PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS

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1. Chanae of the specification for the minimum period setooint. The minimum period setpoint of 7 seconds was established for this facility in the original SAR in 1959. Experience with this and other TRIGA fueled reactors, some of which are licensed to pulse, indicates that this specification is extremely conservative and that the TRIGA-fueled reactors may be safely operated at periods considerably less than 7 seconds. This experience is summarized and analyzed in the report " Credible Accident Analyses For TRIGA and TRIGA-fueled Reactors, SC Hawley and RS Kathren, NUREG/CR-2387 (PNL-4028),1982.

The major concern with minirnum period is that the fuel not be allowed to approach the Safety Limit, which for our facility is 500 C for our mostly aluminum core. With our relatively low power (20KW) and our maximum core excess of $1.00 it is not feasible that temperatures would even approach the Safety Limit.

Even in the case of an inadvertent transient involving fullinsertion of maximum core excess Kathren and Hawley conclude that TRIGA-fueled reactors would not endanger the public welfare. They considered the General Atomic Torrey Pines TRIGA SAR ($2.00 aluminum clad fuel), the Reed College Pines TRIGA SAR ($3.00 aluminum clad fuel), THE Washington State University MTR-TRIGA conversion SAR ($3.75, stainless steel clad fuel),

the Advanced TRIGA Prototype Reactor, and destructive tests with SNAPTRAN reactors.

Under our conditions of operation:

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=> 56 aluminum clad fuel elements

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=> 1 stainless steel clad element

=> $1.00 maximum excess reactivity there is no credible method for our reactor to approach the Safety Limit. In retrospect, similar reactors at much higher power levels are licensed for periods in the millisecond range.

The OVAMC reactor typically operates at 18KW and for purposes of control a limit on the minimum period would seem to be advisable in order to prevent overshoots and possible activation of the Lirniting Safety System Setting on power level, which must not exceed 20KW. A reduction in the minimum period setpoint would allow attainment of steady state power levels more quickly than is now the case, and would help to reduce the number of inadvertent shutdowns caused by random electronic noise at low power levels in the period measuring circuit.

Therefore, we are requesting the deletion of the specifications requiring a scram at our limiting pericd and substitute an interlock which would prevent rod withdrawal when the period is less than three seconds.

Such a period interlock system has been evaluated at other TRIGA reactors and is in use specifically at the Geneml Atomics 250KW TRIGA Mark I reactor. For this reason, the capability already exists to convert our facility by physically bypassing the period scram and configuring a console to give a waming at the three second period time. This waming

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automatically inhibits control rod viithdrawal until the period becomes greater than the C/

setpoint.

Page 2 of 4 m

This interlock would prevent any operation which might exceed the Safety Limit, since

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periods of the order of milliseconds must be used to pulse this or similar reactors to such V'

power levels. When the reactor power level is approaching the Limiting Safety System Setpoint the rate of change of the fuel-moderator temperature is such as to prevent inadvertent overrun of the maximum desired power level.

2. Chance of the specification for examination frecuency of the fuel elements. The reactor fuel element inspection frequency was established for this facility in the originai SAR in 1959.

Experience with this reactor and with other TRIGA-fueled reactors indicates the fuel elements in non-pulsed reactors are dimensionally stable over long periods of time (see the Hawley and Kathren report cited above). No damage to fuel elements has been detected at this facility, inspection of the fuel elements involves physical removal of each element and visual inspection for sweiiing, cracks, corrosion, and pitting. This is accomplished by placing the element in a periscope permanently attached to tha side of the reactor tank. The viewing is aided by the use of a 20 power rifle scope attached to the top of the periscope. The elements are positioned approximately 8 feet below water level. Exposure of operators to radiation from the radioactive fuel elements is very low - it has not been detected at this facility. Furthermore, such handling involves risks of damaging the fuel, such as by dropping an element from the insertion tool, and always involves bum,ning the fuel element during removal and replacement.

In order to meet the goals of the ALARA policy, reduce the risk of fuel damage, and in light

/O of the years of experience with inspection of TRIGA fuel elements, we propose to reduce

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the frequency of handling of fuel elements. Instead of inspecting 4 or 5 elements each quarter (3.5 year basis); the fuel would by inspected on a five-year basis by inspecting at least 20% of the fuel elements annually. In the unlike'y event that a leaking element is confirmed; all the fuel elements would be inspected in order to facilitate locating the damaged element.

As previously mentioned in Attachment #2, please substitute the following to replace TS Sec. 3.2.3, Table 1:

Tablei RequiredScmms Safety Channel Set Point Minimum namber required Percent power 100% licensed p >wer i

Linearpower" 10% license (power I

lon chamberpower supply' Loss of mgh voltage 1

Fissioncounterpower supply" Loss of high voltage I

Consolescram button Manual 1

Magnetcurrent key switch Manual 1

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Watchdogtimer Key sot 1 ware taskstake longer 1

than 1.5 sec j

  • lonchamberanitlogsystem

" Fission counterdigitalsystem

Page 3 -

4 and TS 3.2.4, Table 2:

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Table 2 RequiredInterlocks SafetyChannel Minimum Number.

Function Required Startup i

Preventswithdrawalofall controlrods ifcountrate <2 (Neutron count rate) c.;3)

Log N(Period) 1 Ireventswithdrawalofall controlrods when periodis less than 3 seconds Simultameousmanualwithdrawalof I

Prevents withdrawal two rods' Withdrawalofshim orregulating i

Prevents withdrawal rod with safety rod not allthe way out or seates Withdrawalofsafetyrod with shim i

Preventswithdrawal or regulatingrod not seated

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Table 3 Required Minimum Measuring Channels MeasuringChannel Numb [rOperable Function Startup 1

Monitorsubcritical multiplicationfor startup Power Level (NM1000,fissicn i

Input for safetypowerlevel chamber) scram and to digitaldisplayunit and recorder Log N (NM1000, fission 1

Wide range powerleveland chamber) displayon digitalunit and on recorder Period (NM1000, fission 1

Input for period display on digital chamber) unit and periodinhibit Per Cent Power (lon chamber) 1 Input for powerlevelscram and displayon analogmeter Poolwater Temperature 1

Display on antilogmeter PoolLevel 1

Alarms when waterlevelisless than 12 ft abovetop ofcore g

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Page 4 of 4

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Please replace TS Sec. 4.1.5.1 with the following:

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d 4.1.5.1 Fuelelementinspection Apolicability: This speciGcationapplies to the inspection requirements for the fuel elements.

Q)jssj.iy_n The objective is to verify the continuing integrity of the fuel element cladding.

Specifications: Each fuel element shall be examined for physical damage by visual inspection at least once each five years, with at least 20 percent of the fuel elements examined each year. Observation will include inspection for swelling, cracks, corrosion and pitting.

He reactor shall not be operated with damaged fuel except to detect and identify damaged fuel for removal. A fuel element shall be removed from the core if a clad defeet exists as indicated by the release of fission products.

Basis: The frequency of examination allows each element to be inspected ever 3.6 years. Previous inspection experience has shown that this frequency of inspection is adequate and thus reduces the risk of accident or damage due to handling.

Various sections of the SAR have been altered to reflect these proposed charges.

SAR Section 4, Figures 4.1 & 4.3 have been redrawn.

SAR Section 4.2, Table 4.1 has been editad.

SAR Section 4.2 has been reworded.

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A.'J. BLOTCKY REACTOR FACILITY RADIATION PROTECTION PROGRAM e

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TABLE OF CONTENTS -

1.0 T I ntrod uction..............................................................................

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4 2.0 ' _ _

Management and Administration...............................................

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21. Radiation U nits L..........................................................................

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22 Radiation Limits.;......................................................................

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3.0.

2 Training...................................................................................

.4-4.0 '

S u rveilla nce................................................................................

5

. 4.1 Radioactive Materials Accountability..........................................

6

= 4.2 : Effluent Monitoring.....................................................................

5 --

4.3 : C ontamin ation Surveys...............................................................

5 4.4. Environs Monitoring...............................................................

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4. 5 - Personnel Expos ure..............................................................

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5.0 ALARA Program....................................................................

7-5.1 _ Policy and Objectives.................................................................

7-5.2 I m plemen tation..........................................................................

.7-5.3 Elements of the ALARA Review and Report..............................

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6.0 R eferences.................................................................................

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1. Introduction w

This Radiation Protection Program has been prepared by personnel of the A.J. BLOTCKY TRIGA Mk i Nuclear Reactor Facility in response to the requirements of Title 10 Part 20.1101, Code of Federal Regulations (10CFR20). The goal of the Program is the liniitation of radiation exposures and radioactivity releases to a level that is as low as reasonably achievable (ALARA) without seriously restricting operation of the Facility for purposes of education and research.

The Facility is operated under License R-57 (Docket 50-151) issued by the U.S. Nuclear Regulatory Commission (NRC). The Program is executed in coordination with the Radiation Safety Department, Omaha VA Medical Center. It has been reviewed and approved by the Reactor Safeguards Committee for the Reactor Facility. Certain aspects of the Program deal with radioactive materials regulated by the OVAMC Broad Scope Materials License #26-00138-10.

This program is a pmt of the Operations Manual (completion pending) for the Reactor Facility, although it is published separately. A closely related part of the Operations Manual, also published separately, is the Emergency Plan. The Radiation Protection Program is designed to meet requirements of 10CFR20. It has been developed following the guidance of the American National Standard Radiation Protection at Research Reactor Facilities [1 } and Regulatory Guides issued by the NRC [2-7).

2. Management and Administration d

Preparation, audit, and review of the Radiation Protection Program is the responsibility of the Reactor Manager of the Nuclear Reactor Facility. The activities of the Reactor Managar and annual audits prepared by the Reactor Manager, are reviewed by the Reactor Safeguards Committee chaired by the Chief of Staff for the OVAMC Medical Center. Records required by the Radiation Protection Program as well as audit reports by the Facility Reactor Manager are examined by tne Reactor Safeguards Committee during their annual audit. Training, surveillance and record keeping are the responsibility of the Reactor Manager. ALARA activities, for which record keeping is the particular responsibility of the Reactor Manager, are incumbent upon all radiation workers associated with the Reactor Facility.

Substantive changes in the Radiation Protection Program require approval of the Reactor Safeguards Committee. Editorial changes, or changes to appendices, may be made on the authonty olthe Facility Reactor Manager. Cha',ss made to operating or emergency procedures apply automatically to the Radiation Protection Program and corresponding changes may be made in the Program withcut further consideration by the Reactor Safeguards Committee. As with procedures, the Reactor Manager may override elements of the Program on a temporary emergency basis so long as the emergency changes are brought promptly to the attention of the Safeguards Committee.

2.1 Radiation Units The traditional units of Curie, rad, rem, and roentgen are to be used in record keeping. SI units

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of becquerel, gray, and sievert may t,e used in calculations, dose assessments and reports, so

[Q long as final results, conclusions, etc., are given in traditional units as well.

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73 Extemai exposure is to be recorded in terms of deep or shallow dose equivalent (index).

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According to the ICRP [8), the deep dose equivalent (in rem units)is within 4 percent of the free-field exposure rate (in roentgen units) for gamma rays with energies between 0.6 and 8.0 MeV. Therefore, survey or area monitoring instruments calibrated in roentgen units may be used for assessment of deep dose equivalent in routine surveillance.

The total effective dose equivalent (TEDE)is the sum of the deep dose equivalent for extemal exposure and the committed effectivo dose equivalent for intemal exposure. Intemal exposure associated with the Reactor Facility has never been a threat to workers or the public. Should it be considered as a potentiality, in connection with planned special Exposures or in the conduct of ALARA reviews, its evaluation should follow the guidance of 100FR20, Regulatory Guides

[3-7] or the ICRP [9-11].

2.2 Radiation Limits Occupational dose !imits (except for planned special exposures), given by 10.CFR20.1201 are as follows. The annual limit for adults, in summary, is the more Emiting of the following:

5 rem total e*fe% 0 dose equivalent (TEDE), or 15 rem to the lens of the eye, or

. - 50 rem shallow riose equivalent to the skin or to an/ extremity, or 50 rem combined dcap dose equivalent and committed dose equivalent to any organ other than the lens of tne eye.

Dose limits for individual members of the public, given by 10CFR20.1301, are in summary as

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follows:

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0.1 rem total effective dose equivalent (TEDE) in one year, and 0.002 rem TEDE in one hour.

3. Training implementation of training for radiation protection is the responsibility of the Reactor Manager.

Re-training for active researchers n ust be administered biennially exc2pt for Roactor Operators and Senior Reactor Operators takir g pad in the annual Reactor Facility Requalification Program. Internal exposure monitoring is required only for adults likely to receive in 1 year in excess of 10% of the applicable annuallimits of intake for ingestion and inhalation or for minors or pregnant women likely to receive in excess of 0.05 rem committed effective dose in one year.

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4. Surveillance Surveillance requirements related to radiation protection and imposed by the Reactor Safeguards Committee are as follows:

Daily (operational)

Environmental surveillance of the reactor area before and after each days operation Monthly Wipe test reactor laboratory Environmental surveillance (survey mater)

Quar'erly Source inventory and leak test (OVAMC RSO)

Semi annually Emergency equipment inventory 4.1 Radioactive Materials Accountability

. Materials accountability and release of radioactive materialis the responsibility of the Reactor-1 Manager or his delegate. Procedures are in place to assure that applicable regulations and procedures are in complicance. The responsibility for the facility nuclear fuel inventory is delegated to the Reactor Manager by the Hospital Director. A written statement of this responsiblity is updated every two years. The fuel inventory is calculated semi-annually ond the fuel report is sent to the Department of Energy and the Nuclear Regulatory Commission.

t 4,2 Effluent Monitoring Monitoring of liquid effluents from the Facility is performed to assure compliance with 10CFR20.2003.

4,3 Contamination Monitoring and Surveys Personnel shall monitor hands and feet for contamination when leaving known contaminated i

areas or restricted areas that are likely contaminated. If contamination is detected, then a check

- of exposed areas of the body and clothing should be made. Monitoring control points shall be established for this purpose. Materials, tools, and equipment shall be monitored for contamination before removal from contaminated areas or restricted areas likely to be contaminated 4.3.1 Limits for Removable and Fixed Contamination:

Acceptah a Surface Contamination Levels for Unconditional Release Nuclide' --

Average fixed"

~ Maximum fixed" Removable" U nat,2"U, and 5,000 dpm /100 15,000 dpm /100 1,000 dpm /100 4

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associated decay cm'[8000 Bq m )

cm2 [25000 Bq m ]

cm'[1700 Bq m )

products

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Tranwanics 'Ra, 100 dpm/100 300 dpm/100 20 dpm/100 cm'[30 82 22 era'"Th 2'Th, *Pa, cm'170 Bq m ]

cm'[500 Bq m ]

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r'Ac, *l, "*l 2 5

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Th-nat, '82Th. "Sr, 1,000 dpm/100 3,000 dpm/100 200 dpm/100 8"Ra,224Ra,222U s2:1, cm'[1700 Bq m)

cm'[5000 Bq m)

cm [300 Bq m-2) 2 mg, m; 8

Beta gproma 5,000 dpm - /100 15,000 dpm -/100 1,000 dpm - /100cm emitters (nuclides with cm'[8000 Bq m )

cm'[25000 Bq m)

[1700 Bq m *]

2 decay modes other than alpha emission or spontaneous fission) except Sr-90 and others noted above

  • Where surface contamination by both alpha and beta-gamma o nitting nuclides exists, the limits established for alpha-and beta-gamma-emitting nuclides should apply independently.
  • As used in this table, dpm (disintegrations per minute) means the rate of emission by radioactive material as determined by correcting the counts per minute observed by an appropriate detector for background, efficiency and geometric factors associated with the instrumentation.

'Mecsurements of average contaminant should not be averaged over more than 1 m'. For objects of less surface area, the average should be derived for each such object.

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'The maximum contamination level applies to an area of not more than 100 cm'.

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'The amount of removable radioactive material per 100 cm' of surface area should be determined by wiping that area with dry filter or soft absorbent paper, applying moderate pressure, and assessing the amount of radioactive material on the wipe with an appropriate inntrument of known efficiency. When removable contamination of objects ofless surface area is determined, the pertinent levels should be reduced proportionally and the entire surface should be wipod.

The average and maximum radiation levels associated with surface contamination resn!Gng from bs'a gamma emitters should not exceed 0.2 mrad /h at 1 cm and 1.0 mrad /h at 1 cm, 8

respectively, rneasured through not more than 7 mg/cm of total absorber.

4.4 Environs Monitoring Environs monitoring is required to assure compliance with IOCFR20, Subpart F and with Technical Specifications for the Facility operating license. Technical Specifications require the following monitoring during reactor operations:

a Area radiation monitor at the top of the tank. Calibration of area radiation monitors is required annually.

b.

Continuous air monitor Calibration cf the continuous air monitor is required annually and is performed following the facility standard operating procedures.

T 4.5 Personnel Exposure

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Regulation 10CFR.20.1502 requires monitoring of workers likely to receive, in one year from sources external to the body, a dose in excess of 10 percent of the limits prescribed in 6

10CFR20.1201. The regulation also requires monitoring of any individuals entering a high or I. O

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very high radiation area within which an individual could receive a dose equivalent of 0.1 rem in s

one hour. According to Regulatory Guide 8.7 [2), if a prospective evaluation of likely doses indicates that an individual is not likely to exceed 10 psrcent of any applicable limit, then there are no requirements for recordkeeping or reporting. Likewise, Regulatory Guide 8.3413]

indicates that, if individual monitoring results serve as confirmatory measures, but monitoring is not required by 10CFR20.1502, then such results are not subject to the individual dose recordkeeping requirements of 10CFR20.2106(a) even though they may be used to satisfy 10CFR20.1501 requirements. As a result, the following procedures are implemented:

a.

When the reactor is in operation, no person may enter the reactor room without the permission of the Reactor Operator on duty at the Console.

b.

Unescorted visitors will not enter the reactor room.

c.

Except as indicated below, no person may enter the reactor room without gamma-ray personnel dosimetry.

d.

Escorted individual visitors must have personnel dosimetry prior to reactor room

entry, Escorted visiting groups will be monitored by not less than two self-reading e.

pocket dosimeters or two film or TLD badges for every fifteen members of the group.

5. ALARA Program 5.1 Policy and Objectives

/m Management of the Reactor Facility is committed to keeping both occupational and public V) radiation exposure as low as reasonably achievable (ALARA). The specific goal of the Al. ARA

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program is to assure that actual exposures are no greater than 10 percent of the occupational lirMs and 50 percent of the public limits prescribed by 10CFR20, namely, ALARA goals of:

Workers:

o 5 500 mrem annual TEDE

=> $ 5 rem annual dose equivalent to any organ except the lens of the eye

=> $ 1.5 rem annual dose equivalent to the lens of the eye o 5 5 rem annual dose equivalent to the skin

=> 5 50 mrem dose equivalent to the fetus during pregnancy Pubhe:

=> < 50 mrem annual TEDE 5.2 Implementation of the ALARA Program Planning and scheduling of operations and experiments, education and training, and facility design are the responsibilities of the Reactor Supervisor. Any action which might lead to as r.ch as half the annual ALARA dose limit (Section 5.1) to any one individualin one calendar quarter requires a formal ALARA review and report. Any staff member or experimenter, or any member of the Reactor Safeguards Committee may call for an ALARA review of a proposed action. Under any of these circumstances, it is the responsibility of the Reactor Supervisor to conduct an Al. ARA review and report. Only with the approval of the Reactor Supervisor or the

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Chairman of the Reactor Safeguards Committee.

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p 5.3 Elements of the ALARA Review and Repcrt

\\b The following topics shall be considered, if applicable. The report shall include discussion of how these topics affect personnel exposure and specific actions recommended, categorized by topic:

Features for External Radiation Control:

Shielding and construction materials Radioactive material storage and disposal e

Monitoring systems e

Facility layout e

Control of access to high and very high radiation areas Contamination Control e

e Ventilation and filtration Containment of contamination Construction materials to facilitate decontamination e

Liquid effluents e

Effluent monitoring e

Operations and Operations Planning Assessment of potentialindividual and collective exposures Application of shielding, time, and distance for dose reduction e

Use of ventilation and decontamination to reduce collective dose Provision of special personnel training and practice

-e Provision of special supervision and surveillance e

Provision of special clothing or other protective gear

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6.0 References 1.

American National Standard Radiation Protection at Research Facilities ANSI /ANS 15.11 (Final Draft), American Nuclear Society, La Grange Park, Illinois, October,1992.

2.

Instructions for Recording and Reporting Occupational Radiation Exposure Data Regulatory Guide 8.7 (Rev.1), U.S. Nuclear Regulatory Commission, Washington, D.C.,1992.

3.

Monitoring Criteria and Methods to Calculate Occupational Radiation Doses Regulatory Guide 8.34, U.S. Nuclear Regulatory Commission, Washington, D.C.,1992.

4.

Air Sampling in the Workplace Regulatory Guide 8.25 (Rev.1), U.S. Nuclear Regulatorv Commission. Washington. D.C.1992.

5.

Planned Special Exposures Regulatory Guide 8.35, U.S. Nuclear Regulatory Commission, Washington. D.C.1992.

6.

Radiation Dose to the Embryo / Fetus Regulatory Guide 8.38, U.S. Nuclear Regulatory Commission, Washington, D.C.,1992.

7.

Interpretation of Bioassay Measurements Draft Regulatory Guide 8.9 (DG 8009), U.S Nuclear Regulatory Commission, Washington, D.C.,1992.

8.

Data for Use in Protection Against Extemal Radiation Publication 51, Intemational Commission on Radiological Protection.1987.

9 Limits for intakes of Radionuclides by Workers Publication 30, Intemational

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Commission on Radiological Protection,1979.

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10.

1990 Recommendations of the Intemational Commission on Radiological Protection Publication 60, Intemational Commission on Radiological Protection,1991.

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11. -

Limits for intakes of Radionuclides by Workers Based on 1990 Recommendations of the Intemational Commission on Radiological Protection Put* cation 61, intomational Commission on Radiological Protection,1991, 3

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