ML20202F447

Forwards Addl Info in Support of possession-only-license Amend Application Dtd 970807.Revised TS Re Rev to Facilitate Decommissioning,Encl

ML20202F447
Person / Time
Site:	Neely Research Reactor
Issue date:	02/04/1998
From:	Davidson J Neely Research Reactor, ATLANTA, GA
To:	Mendonca M NRC (Affiliation Not Assigned)
Shared Package
ML20202F451	List: ML20202F447 ML20202F462
References
NUDOCS 9802190202
Download: ML20202F447 (6)

Text

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- College of Engineering Office of the Dean February 4,1998 Docket No. 50-160 United States Nuclear Regulatory Commission 11555 Rockville Pike Rockville, MD 20852 2738 Attention: Mr. Marvin M. Mendonca, Senior Project Manager Non Power Reactors and Decommissioning Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation h

Subject:

Possession-Only-License Amendment Application- AdditionalInformation

References:

1. USNRC Letter,

Subject:

Request for Additional Information (TAC No.

99370) dated December 12,1997

2. Submittal: Possession-Only License Amendment Request for Georgia Institute of Technology Research Reactor, dated August 7,1997

} 3. Georgia Institute of Technology Research Reactor USNRC License No. R-97

__ 4. NRC Docket No. 50-160

Dear Mr. Mendonca:

Georgia Tech is hereby submitting additional information in support of our Possession-Only License Amendment in response to your request per Reference 1. We would appreciate your timely review of this submittal and the processing of our Amendment request. If you have any questions regarding this subasittal, please contact our project manager, Mr. Bill Miller, at 404-894-4610.

Sincerely,

/ .

J. Narl Davids Associate Dean Chair, Technical Safety Review Committee

Enclosures:

1. Response to USNRC Comments Dated December 12,1997 '

/

2. Appendix 1 and 2 (Added) 3,,m

3. Technical Specifications (Revision - 2/4/98) - V f ()

j Cc (Enclosures Withdrawn): Dr. G. Wayne Clough, President v Members, Technical Safety Review Committee

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RESPONSE TO USNRC COMMENTS DATED DECEMBER 12,1997 CONCERNING THE GEORGIA INSTITUTE OF TECHNOLOGY'S REVISION TO TECIINICAL SPECIFICATIONS TO FACILITATE DECOMMISSIONING

1. Technical Specification (TS) 2.2 should specify what monitors arg required wht.

activities related to radioactive material are beingperformed. Alternatively, a statement such as "No activities related to radioactive materials will be performed unless the ,

monitors specified in Table 2.1 are operable".

Technical Specification 2.2 has been changed to require the identified instruments be E operable when licensed : ,tivities related to radioactive materials are being performed. The USNRC suggested statement has been incluN into the basis.

3

2. Provide analysis to demonstrate that the Tntium ofonitor is no longer required after the heasy water is removedfrom the reactorfacility.

An analysis has been prepared which demonstrates that the tritium monitor is no longer needed after the heavy water has been removed from the facility. The analysis is attached as Appendix 1.

3. In Table 2.1, there are set pointsfor the tritiurr, the movingfilterparticubte, and the filter bank monitors that are based on an annual calibration. Additional specificiy us needed. Therefore, provide a value or a multiple of backgroundfor these monitors, or provide the specific calibration procedure to establish these set points.

Technical Specification 3.2 for the operating reactor has been changed to provide set points for the Tritium monitor, moving filter particulate raonitor or the filter bank monitor. The set point for the Tritiam monitor is 10 mR/hr due to the inherent limited sensitivity of the monitor and the potential interference from aaturally occurring radon.

4. TS 2.3 specifies applicability of "as long as heavy wateris stored. " Containment integrity should be required as long as this barrier to radioactive materials is required. Provide calculational veryication thatfor allpotential activities after removal ofthe heasy water (including decommissioning activities) containment integrity is not required.

Alternatively, provide a revision to the TS to require containment integrity during activities with radioactsve materials.

Once the heavy water is removed from the site, a major source of radioactive material (the tritium) will be removed. The spent fuel has already been removed from the site. NUREG-1537, 1

Part 2, " Guidelines for Preparing and Reviewing Applications for Licensing Non-Power Reactors", identifies the maic purpose of the containment is to mitigate accident consequences for the operating reactor. The dynamics associated with decommissioning are greatly less than those associated with an operating reactor. local temporary containments can be used to contain any potential radioactive material that may be generated during decommissioning operations.

Precedence has been set using these temporary containments during decommissioning research reactors at UCLA, UC Berkeley, and University of Kansas. These temporary containments effectively minimize the spread of radioactive material througho :t the facility. They heva also been used during numerous other decommissioning projects at USNRC licensed facilities. There are many other facilities that have greater amounts of radioactive material that do not require a containment building. An analysis of the potential release of the GTRR Source Term indicates that if all were released, the combined fractions only equals 3 % of the allowed release. See Appendix 2.

5. Table 3.1, footnote (1) limits applicability to conditions that could cause a "high radiation area. " This monitorprovides indicationfor radioactive conditions below this threshold and the applicability should be changed to so indicate. Further, it should be consistent with the requirementfor the instrument in TS 2.2.

Footnote (1) has been changed to apply to any licensed activity which could cause a radiation area. This is consistent with Technical Specification 2.2.

6. Table 3.1, footnote (2) limits applicability to conditions that could cause " airborne release to the containment " Provide a reference to the methodology that will be used to determine these conditions (e.g., in basis provide veripcation that afacility procedure will be established and used).

Footnote (2) and basis have been changed to provide a better definition of when the appropriate monitors will require surveillance and maintenance. The daily checks must be peformed when licensed activities that could cause an airborne release to the containment of greater than twelve DAC-hours per week as defined in 10 CFR Part 20.1002.

7. TS 5.6.a indicates "and fax. " Fax service is not at the number specified and is not required by USNRC guidance. Eliminate the referral tofar orprovidejustification.

Reference to sending a fax report has been eliminated.

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Georgia Tech Research Reactor (GTRR) Appendix 1 Neely Nuclear Research Center Georgia Institute of Technology Analysis to Demonstrate That the Tdtium Monitor is Not Required After the Removal of the Heavy Water From the Facility Question 2 from NRC's December 12,1997 letter asks us to provide mathematical calculations showing that the tritium moni:or is not needed afer the removal of the heavy water.

The stack flow rate from the containment building is approximately 31000 standard it' per minute. .

(31000 ft ' ) * (60 min) * (24 hr) * (365 d) , 28300 ml min

• hr

• d

• yr I ft '

This equates to the removal of 4.6 E14 ml of air from the containment building per year.

Currently there are approximately 2468 gMons of heavy water in the storage tank. The concentratian of tritium in the heavy water is approximately 140 Ci per ml.

140 Ci ,

ml

'3785

, gallon ml

• M8 #

This equates to 1.3 E9 pCi of tritium in the heavy water. For this calculation, it is assumed that 99%

of the heavy water (and tritium activity) will be removed, leaving 1.3 E7 pCi of tritium. If all of the remaining tritium were to evaporate, and be released up the stack, the average annual concentration at the stack release point would be:

1.3 E7 p Ci

= 2.8 E 8 pCi/ml 4.6 E14 ml A XIQ value of 4.1 E-4 sec/v'(Technical Specificationsfor the Georgia Tech Research Reactor, Georgia Institute of Technology) can be used to take into account the dilution from the stack to the site boundary.

X( Ci/ml]= 5#" * ' '

• *' *

• l *i" m' ml min ft ' 60 sec This gives an average concentration of 1.7 E-10 Ci/ml at the site boundary, well below the effluent concentration of 1 E-7 pCi/ml from Appendix B of 10CFR20. In fact, even before the removal of the heavy water, if all the tritium were released, the average concentraticn at the site boundary (1.7 E-8 pCi/ml) would not exceed the effluent concentration.

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Georgia Tech Research Reactor (GTRR) Appendix 2 Neely Nuclear Research Center Georgia Institute of Techrology Analysis to Demonstrate That Containment Integrity is Not Required After the Removal of the Heavy Wa'ter From the Facility Question 4 from NRC's December 12,1997 letter asks for mathematical calculations showing that containment integrity is not required after the removal of the heavy water.

The vast majority of radioactive material in the containment building is contained in the teactor biological shield and components. A SCALE code calculation of the activation products was performed by Dwayne Blaylock (Activation Products in the Biological Shield of the Georgia Tech Research Reactor, M.S. Thesis, Crorgia Institute of Technology,1997). His calculations found that the major isotopes were C-14, Co-60, and Zn-65. In May of 1997, the estimated activities were 2.2 -

E6 pCi,5.5 E6 Ci, and 4.4 E6 pCi, respectively.

The C-14 is primarily contained in graphite. In the worst-case scenario, if the graphite wne to catch fire and all of the C-14 was liberated, then the avera'ge annual concentration at the stack release point would be:

2.2 E6 p Ci

= 4.8 E- 9 pCi/ml 4.6 E14 ml 3

The X/Q dilution factor of 4.1 E-4 sec/m reduces this value to:

8 8 y, j 4.1 E-4 sec ,4.8 E-9 Ci,31000 fl . 0283 m ,1 min m

8 ml min fl 8 60 sec 2.9 E-11 pCi/ml, well below the effluent concentratian limit of3 E-9 pCi/ml.

The Co-60 and Zn-65 are activation prodwts in aluminum, stainless steel, and other metals. 'Ihis material is not likely to be released in gaseou or particle form. However, if one assumes that 1% of the Co 60 and Zn-65 were released, the average annual concentration at the stack release point would be:

E E Co-60: = 1.2 E-10 Ci/ml Zn-65: = 9.6 E-11 pCi/ml 4.6 E14 ml 4.6 E14 ml Taking into account the X/Q dilution factor, the average worst-case concentration at the r , boundary for Co-60 is approximately:

8 X[pCi/ ml} = 4.1 E-4 sec

• 1.2 E-10 pCi

• 31000 fl * .0283 m'

• I min m

8 mi mm ft, 60 sec

7.2 E-13 pCi/ml (below the effluent concentration of 5 E-11 pCi/ml).

For Zn-65, taking into account the X/Q dilution factor gives an average concentration of 8

y, 4.1 E-4 sec ,9.6 E-Il Ci,31000ft . 0283 m' , I min m' W mm R' @ wc 5.8 E 13 pCi/ml (below the effluent concentration of 4 E-10 pCi/ml).

A sum of the fractions for each of the major isotopes (H-3, C-14, Co-60, Zn-65) yields:

1.7 E- 10 2.9 E-11 7.2 E -13 5.8 E-13

= 0.03 lE-7 3E 9 S E-11 4 E-10 i

Since the sum of the fractions is 1, even in the worst-case scenario for all releases of mdioactive materials in the containment building, only a small fraction of the effluent concentrations would be reached. In fact, even before the removal of the heavy water, the sum af the fractions yields:

1.7 E - 8 2.9 E -Il 7.2 E -13 5.8 E-p = 0.20 lE-7 3 E-9 SE-Il 4 E-10 Thus, even before the removal of the heavy watcr, the worst-case scenario would not result in a -

release to the environment in excess of 10CFR20 effluent concentration limits.

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