ML20210P138
| ML20210P138 | |
| Person / Time | |
|---|---|
| Site: | Neely Research Reactor |
| Issue date: | 02/08/1999 |
| From: | Hertel N Neely Research Reactor, ATLANTA, GA |
| To: | Mendonca M NRC (Affiliation Not Assigned) |
| Shared Package | |
| ML20210P143 | List: |
| References | |
| TAC-MA2362, TAC-MA2363, NUDOCS 9908120084 | |
| Download: ML20210P138 (12) | |
Text
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Georgia mBGffdnaQ@ Neely Nuclear Research Center echn@.gy e00 t'eet>o or>ve Atlanta, Georgia 30332-0425 a (404) 894-3600 FAX: (404) 894-9325 February 8,1999 United States Nuclear Regulatory Commission i1555 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
Subject:
Response to USNRC Request for Additional Information (TAC No. MA2363)
References:
- 1. USNRC Letter dated December 28,1998,
Subject:
Request for AdditionalInformation (TAC No. MA2362)
- 2. USNRC Docket No. 50-160
Dear Mr. Mendonca:
Enclosed is our response addressing your request for additional information.
Please contact me at 404-894-3601 if there are any questions regarding our response. I Si e rel ,
Q '
l Nolan E. Hertel l
Director, Neely Nuclear Research Center
Enclosure:
as stated Cc: Dr. G. Wayne Clough, President, Georgia Institute of Technology Members of the Technical Safety Review Committee
. 1' Notary: --
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u A Unit of the University System of Georgia An Equal Education and Employment Opportunity Institution 9908120084 990200 PDR ADOCK 05000160 P PDR -
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- RESPONSES TO GEORGIA INSTITUTE OF TECIINOLOGY (GT)
RESEARCII REACTOR DECOMMISSIONING PROJECT REQUEST FOR ADDITIONAL INFORMATION (RAl) l
GENERAL COMMENT
S Only QC aspects of the QA plan are presented. We would like copies of the QA plans for GT and NES and the interface between the two.
Response: It is the intent that, when a DC is selected, a QA Program will be prepared by the DC, subject to review and approval by Georgia Tech, that will encompass the entire D & D organi:ation.
Since many alpha readings are presented in the documents, an explanation of the source of this contamination is needed.
Response: Ofthe 892 direct alpha readings taken, 8 (or <1%) were above the MDA for the survey instrument. Radon and its daughters are suspected due to the results of the Reactor Building air sampling.
SPECIFIC COMMENTS- Characterization Report Book i
- 1. Please explain the last sentence on page 8. We assume the survey is not intended to release the NNRC for unrestricted use.
Response: The characteri:ation surveys do not release sites for unrestricted use. The sentence was included in the Characteri:ation Reportfor clarification purposes.
- 2. A-4,6,8,10 Please provide some explanatory text on these calibration sheets so that we can better understand their significance.
Response: An instrument's Response Range, or Control Chart, is established with a known source at the start of the project to establish the acceptable range. Instruments are then checked daily with the same source to verify that their responsesfall within this range. Ifnot, the instrument would be removedfrom use until the deviation was resolved and acceptable responses were demonstrated.
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- 3. B-1: Please explain the discrepancies between the on-site and independent analysis of samples. Discuss the effect of this discrepancy (if any) on the final survey.
Response: The discrepancies were primarily due to conversion, and some transcription, errors. A corrected version ofAttachment B-1 is enclosedforyour review.
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- 4. 16: Please discuss why removable contamination need not be removed.
Response: It was not the intent to imply that removable contamination need not be removed. The specifications "fixedphis removable contamination" were added to clarify that direct readings should be taken before the measured area is swiped. The specifications on this page will be changed to match the exact wording of NRC Reg.
Guide 1.86.
!5.17: Please confirm that the numbers presented in the paragraph on airborne radioactivity limits are the maximum acceptable concentrations.
Response: The section on airborne radioactivity limits will be changed to read:
" Concentrations of airborne radioactivity in radiologically and non-radiologically controlled areas shall comply with the standards established in 10CFR20, Appendix B, j Table 1 or Table 2, as appropriate. "
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- 6. 17: Please provide additional details pertaining to the paragraph concerning Retention Tanks ;
radioactivity release limits.
Response: The section on retention tank limits will be changed to read: " Concentrations in liquid to be releasedfrom the site shall comply with the standards established in 10CFR20, Appendix B, Table 2 or Table 3, as appropriate. "
- 7. D-1,2: Please discuss the meaning of the negative numbers.
See Response to Question 8 below.
- 8. D's: What is the meaning of"a measured activity below the MDA?"
Response to Questions 7 and 8: The established daily ambient background subtracted from the measured value may result in a negative number. Daily MDAs are calculated using count times, background values, efficiencies, and detector geometries.
NUREG/CR-5849 recommends that actual data be presented and usedfor calculational purposes. This avoids the skewing ofdistribution residts and inaccurate conclusions that might resultfrom substituting MDA valuesfor the actual data.
- 9. 25: Where are the results of the asbestos and lead in paint survey? Were these materials found contaminated with radioactive materials?
Response: Summary tables of the asbestos and lead analysis results are included in Attachments K and L These results and locations may then be compared to the rad survey data to determine their contamination status.
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SPECIFIC COMMENTS - Decommissionian Plan l
- 1. 9: Please clarify the last sentence in the second paragraph concerning other licensed activities on the site. Is this to be a discussion of the Co-60 irradiator and the interface between the Co-60 storage pool and the reactor?
Response: "Other licensed activities " refers to radioactive material usage as defined in the Georgia Institute of Technology's Broad Byproduct Materials License No. GA117-1
, (SNM) issued by the State of Georgia. This license includes the Co-60 irradiation l sources stored in the water pool and used in hot cell irradiations.
Neely Nuclear Research Center (NNRC) houses a wide variety of radioactive materials l used in academic research. Theformer Georgia Tech Research Reactor was housed in a l distinct separate containment building within the NNRCfacility. This decommissioning l plan encompasses the containmentfacility plus several specified reactor support areas l (tank farm, waste treatment room) outside the containment. While reactor l decommissioning is taking place, academic radiation research will continue at the NNRC
. under the Georgia broad license in separate and distinct laboratories that have no connection with the reactor containmentfacility or NRC Tech Spec allowed activities.
No radioactive research will be conducted in the specified (NRC controlled) l decommissioning areas.
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- 2. 25: What is meant by reactor " faces" In 2.2.2?
Response: The "reactorfaces" refers to the eight exterior surfaces of the octagonal reactor shield wall.
- 3. 31: Please explain the meaning of section 2.3.2.20 (i.e., remaining U-235).
Response: The text of 2.3.2.20 will be changed to read " Small quantities ofremaining U-235 (contained in core monitoring " fission chambers" and experimentfails) will be packaged and shipped to an appropriate licensed disposalfacility. " This will clarify' that there are only gram quantities of U-235 remaining, notfuel elements.
- 4. 32: Is the outdoor characterization that which has already been performed or a new characterization?
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- . Response
- The title "Characteri:ation" L misnomer. Decommissioning activities have l the potential to impact surrounding areas, especially with the movement, packaging, and shipping ofcontaminated materials. Although strict controls will be in place to prevent or control this, routine " verification" (or " confirmation ") surveys and sampling will be performed concurremly with decommissioning activities. To avoid confusion and to be consistent with the terminology " Facility Yard" used in the Characteri:ation Report, the title ofthis section will be changed to " Yard Surveys ".
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I 5. 33: Please describe the purpose for installation of "new retention tanks." Are the "new L retention tanks" 'a part of this decommissioning plan and, therefore, a part of our review of the decommissioning plan?
Response: Mention ofnew retention tanks was put in the Decommissioning Report so it is clear that new tanks will be installed in place of the old tanks, after release of the facilityfor unrestricted use, and the room's structure is not to be disturbed, ifpossible.
The new retention tanks are notpart ofthe decommissioningprocess.
- 6. ~38: Please explain the techniques to be used by the TSRC to audit the D & D projects. (Note that audits by TSRC are not mentioned in Section 7.2.5.2).
Response: The final QA Plan will detail the auditing organization, methods, and i proceduresfor the D & D project. This will be available when the DC is selected and finalprojectplans andprocedures are developed. However,for clarity, some editing will beperformedin section 7.2.5. These changes are asfollows:
- 1) Delete the word " Contractor's"from 7.2.5
- 2) 7.2.5: . Delete the sentence "The Executive Engineer or NNRC Director designee willperform an independent audit as an over-check ofthe DC"
- 3) ' New 7.2.5.1 "The Executive Engineer or NNRC Director designee will perform an independent audit as an over-check of the DC. The RSO will be responsiblefor the routine review and audits ofitems 7.2.5.2 through 7.2.5.6. "
- 4) Insert new 7.2.5.1 (above) and renumberpresent items 7.2.5.1-7.2.5.6 as 7.2.5.2
-7.2. 5. 7.
- 5) New 7.2.5. 7 - Change wording to "The TSRC will be responsiblefor the review and approval of documentation, to be developed by the DC, that will be submitted to the NRC. " l
- 6) Addnew 7.2.5.8 " Additionally, the responsibilities ofthe TSRC will conform to the specifications in Section 5.2 of the G.I.T. 's Decommissioning Tech j Specs. "
~7. 42: The order of the review and approval of the ALARA plan is unusual and is different than the plan for procedures, etc. (Is the order reversed?)
Response: The approval path for the ALARA Plan is more extensive than other procedures and is intentional. This involvement of the Executive Engineer review and approval was not previously indicated in section2.4.4. The following will be added to section 2.4.4. "The Executive Engineer will review and approve the ALARA Plan. The Executive Engineer will perform independent audits of the DC. Additional review, approval, and/or audit responsibilities may be assigned by the Director of the NNRC as !
theprojectprogresses. "
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- 8. 43: Please confirm that the Health Physics Program and the GTRR Site Health and Safety Plan are the same plan.
Response: The GTRR Site Health and Safety Plan is one component of the Health Physics Program.
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- 9. 43: Will an instrument calibration be donejust prior to the final survey?
Response: Although there are no provisions that require instrument calibrations just prior to the final survey, calibrations on instruments used in the final survey will be
. performed as needed. Calibrations will be performed with the radionuclides ofconcern ,
or appropriate correction factors developedfor the different radionuclides of concern.
Section 4.0 discusses the instrumentation methodsfor the Final Survey.
10.~45: Please' verify that the TSRC will not only review but also approve the Environmental
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Program Plan. l 1
Response: This sentence will be corrected to read "All plans and procedures will be reviewed and approved by the TSRC", to be consistent with 2.4.3.
- 11. 45: Please discuss the relationship between the existing and the new onsite emergency plan to be developed by the DC.
Response: On December 5,1998, the NRC exempted the Georgia Tech Research Reactor facilityfrom having an Emergency Plan. An on-site Emergency Plan that is site-specific
- as well as project-specific must be prepared by the DC to accountfor the responsibilities ofall individuals involved in the project.
- 12. 48: Please explain the meaning of" low average radiation exposures" to the workers. Should not Table 3.2 contain an administrative limit for an annual or project dose?
l Response: The words "with low average radiation exposure limits to the workers" is non-specific and will be changed to read "while keeping exposures within the objectives ofALARA. "
l Table 3.2 will be modified to add an annual administrative exposure level of1,000 mrem. 1 i
Additional changes in 3.1.2.5: The use of the word " limits" in Table 3.2 and the text of section' 3.1.2.5 is misleading. These are goals of the ALARA program, but will be \
allowed to be exceeded under certain conditions and within certain constraints. These conditions and constraints will be clearly detailed in the project-specific ALARA plan to 1 be prepared by the DC. Therefore, the word limits will be replaced by the word levels in the last sentence of3.1.2.5 (pp. 47), the paragraph preceding Table 3.2 (pg 48), the title for Table 3.2, and within Table 3.2 (pp. 48).
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- 13. 49: The approval path for RWP in 3.1.3 is not consistent with the approval path in 2.4.1 and 1
2.4.2. Pleasejustify your reasons for this charge.
. Response: The sentence in 3.1.3 will be corrected to read "The RSO and the Director of
' the NNRC will review and approve all RWPs. " The addition ofthe underlined words as l
well as the deletion of "or their designees" will make this statement consistent with 2.4.1 '
and 2.4.2.' The final QA plan will indicate individuals responsible for review and approval, as well as their alternates (or designees),'should the primary signatory be unavailable.
- 14. 50: Please describe the monitoring capabilities installed in the portable exhaust systems used with CCE.
)
Response: Air exhaustedfrom portable exhaust systems will be radiologically monitored either locally or by the building exhaust monitoring system. The monitoring location will be determined by adherence to the " Airborne Radioactivity Control Procedure" (section
. 3.1.2.2.3) and by specific work procedures that will be generatedfor each task. These procedures will establish controls based on an evaluation of the potentialfor airborne contaminationfor each task.
- 15. 57: Please describe the " designated licensed disposal site" for mixed waste.
1 Response: The designated disposal site depends on the availability and cost of a site at
= the time ofdecommissioning and can change during the course of a decommissioning.
Therefore, a specific disposal site is not named at this time. However, to ensure that disposal sites meetproper criteria, the word " licensed" will be added to the disposal site description.
- 16. 58: Since the Radiological Accident Analysis is discussed in the DC's emergency plan, please provide a copy to the NRC.
Response: Delete present "Other potential accidents" paragraph, and replace with the following underlinedtext.
" Worst case" analyses were performedfor accidents that would contribute dose to a decommissioning worker or an individual at the site boundary. The dose to a decommissioning worker assumes that dewatered sludgefrom the tank with the highest concentrations is contained in a 55-gallon drum. This drum is dropped during transport and the nearby worker inhales apercentage ofthe dispersed activity. Using conservative assumptions, the CEDE to the worker would be 0.0285 the ALI, or 143 mrem. See Appendix Bfor calculations.
The dose at the site boundary assumes afire in the graphite of the biological shield and components.' This analysis was performed by Georgia Tech in response to NRC comments on the P.O.L request. The sum of thefractions of the effluent concentrations
' at the site boundaryfor the major iso: opes (H-3, C-14, Co-60, Zn-65) yielded a value of 0.03. For details of this analysis, refer to Georgia Institute of Technology Letter to the 6 Grdoc b
l USNRC, dated 2N/98, Appendix 2, " Analysis to demonstrate that containment integrity is not required after the removal ofthe heasy waterfrom thefacility" In summary, the scenarios listed above pose no major ha:ards to the public or decommissioning personnel under the proposed decommissioning activities. If an accident did occur the Emergency Plan, that will be developed by the DC, would be implemented.
l l 17. 67: Are the QA and Radiological Control plans different documents?
1 Response: Section 7.1 discusses Quality Assurance Provisions. The two references to a l Radiological Control Program will be deletedfrom this paragraph.
Additional changes: The terms " Radiological Control Program" and " Health Physics
- Program" were both used to indicate the same program. To avoid confusion, the term
" Radiological Control Program" will be the only one used in the D-Plan. The four places where the term " Health Physics Program" was used (pages ,1 and 43) will be changed to say " Radiological Control Program ".
l 18,67: Does the QA plan conform to ANS 15.8 or some similar document?
l Response: A statement will be added to the Decommissioning Plan requiring the QA l Plan to confbrm to ANS 15.8.
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- 19. 59: Why is the TSRC not mentioned in 7.2.5.27 Response: See responsefor item #7, page 42for the D-Plan above.
i EDITORIAL, COMMENTS - Characterization Report Book 1
- 1. 24: Last line of first paragraph: Section 8.3 should be Section 8.3.3 correct?
l Response: Section 8.3 will be changed to Section 8.3.3.
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- 2. 15: Please provide a copy of Reference 3 (QA plan for site characterization). I l Response: Enclosed is a copy ofNES Document 82A9085. l l
- 3. D-2: Page is missing.
Response: Enclosed is a copy ofPage D-2.
- 4. 21: Where is Appendix A? Should this be Attachment A?
Response: Reference to " Appendix A " will be changed to " Attachment A. "
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- 5. 29: The last sentence ends abruptly and does not continue on next page.
1 Response: Sentence will be changed to say, "All of the stored materials and equipment l
} in the walk-in l l
dpm/100 cm'. food had elevated direct py activity with a maximum reading
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- 6. 48: Is reference I for the University of Washington Research Reactor correct?
Response: "(Reference 1) " is incorrect and will be deletedfrom this sentence.
- 7. J-8,9: Please editorially correct these pages.
Response: 11 appears that these were poor copies. Clearer copies will be provided with a more prominent page title or description.
EDITORI AL COMM ENTS - Decommissioning Plan
- 1. 20: Please provide drawings that are clearer.
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Response: Clearer drawings are attached and will be provided in the final Decommissioning Plan. )
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- 2. '24: Please verify our assumption that the May 1998 version of the Radiological l Characterization Report is the correct version (or January 1998?). j Response: The May 1998 version of the Radiological Characteri:ation Report is the correct version. The date of the latest Characteri:ation Report will be changed to Afay 1998.
- 3. 25,29: Because of the poor quality of Figure 1.7, it is not possible to locate the Bismuth shield blocks, lead cover plate, etc.
Response: A clearer drawing indicating locations of reactor components is attached and will be provided in thefinal Decommissioning Plan.
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APPENDIX B ,
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- 1. From Attachment M of Site Charactedcab0n, a gamma spevbuscopy analysis of the waste tanks i yielded the highest value at LIQ-6, " Suspect Waste Tank Sludge." The sample weighed 765g. at 550 ml geometry for an effectNe density of 1.39 g/ml.
- 2. The default density of soil is 1.6 g/cc and water is 1.0 g/ce, therefore the H2O and soil components of the sample can be represented by the followng equation:
1.6(X) +1.0(1-X) = 1.39 Where: X = the fraction of soil and (1-X) = the fracton of water Solving: X= .65 and (1-X) = .35 3 l
In the proposed scenario, the sludge is dried and placed into a 55-gallon (7.5 ft*) drum geometry. !
Due to this volume reduction, a scaling factor of at least 1/.65 (or 1.54) should be applied to correct for the increased specific actMty in the dried media. For conservatism, a factor of 2 shall be used.
- 3. From the gamma spectrum for Liq 4, the actuty in a 55-gallon (7.5 ft ) drum is calculated:
Cs-137 = (5.55 pCilg)(2 SF)(1.6 g/cc)(28317cc/ft )(7.5ft )(1E-6 uCi/pCi) = 3.77 uCi Co-60 = (125 pCi/g)(2 SF)(1.6 g/cc)(28317cc/ft')(7.5ft')(1E-6 uCi/pCi) =84.95 uCi {
- 4. In the " Dropped Drum" scenario, an occupational worker inhales 1% (another conservatNe estimate) of the actMty in the drum. The inhaled value is compared to the Stochastic All' to yield a Committed EffectNe Dose Equivalent (CEDE):
(3.77 uCi Cs-137)(.01) + (84.95 uCi Co-60)(.01) = .0285 ALI 2E2 uCi ALI 3E1 uCi All CEDE = .0285 ALI (5 rem /All) = 0.143 rem or 143 mrem
' 10CFR20, Appendix B, Table 1, Column 2.
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