ML20202C624

Responds to NRC 980116 RAI Re Decommissioning Plan & Request for Termination of Triga Mark I & Mark F non-power Reactor Licenses R-38 & R-67

ML20202C624
Person / Time
Site:	General Atomics
Issue date:	01/28/1999
From:	Asmussen K GENERAL ATOMICS (FORMERLY GA TECHNOLOGIES, INC./GENER
To:	Alexander Adams NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
38-67-3032, TAC-M98498, TAC-M98499, NUDOCS 9902010128
Download: ML20202C624 (12)

Text

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g EENERAL ATORNCE q: January 28,1999

' 's 38/67-3032 Via Express Delivery Service Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 ,

A'ITN: Mr. Alexander Adams, Jr.

Subject:

Docket Nos. 50 89 and 50-163; License Nos. R-38 and R 67 Respectively:

Submittal of Response to Request for Additional Information (TAC Nos. M98498 and M98499) A

Reference:

Adams, Alexander, Jr., Letter to Dr. Keith E. Asmussen, REQUEST FOR ADDITIONAL INFORMATION (TAC Nos. M98498 and M98499), dated January 16,1998 I

DearMr. Adams:

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% This letter is in reply to your request for additional information (Ref.) regarding General Atomics' (GA's) submittal of a decommissioning plan and request for termination of its TRIGA*

Mark I and Mark F non-power reactor licenses R-38 and R-67. The enclosure to this letter includes a restatement of each request / question immediately followed by GA's corresponding response. GA has made corresponding appropriate revisions to its TRIGA* Mark I and Mark F research reactors decommissioning plan and is preparing to submit the revised plan, under separate cover, for your continued review and approval.

GA appreciates your assistance in considering its request and is hopeful for an expeditious approval. If you should have any questions regarding this submittal, or require any additional information, please do not hesitate to contact me at (619) 455-2823.

Very truly yours, f.

Keith E. Asmussen, Director Licensing, Safety and Nuclear Compliance

Enclosure:

GA's responses to NRC's request forinformation (2 copies)

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REQUEST FOR ADDITIONAL INFORMATION

. GENERAL ATOMICS TRIGA MARK 1 AND MARK F RESEARCH REACTORS DOCKET NOS. 50-89 AND 50-163 January 28,1999 Request:

1. Section 1.2. On Figures 1-3 and 1-4 please show the boundaries of the Mark I, Mark F and State of California licenses.

Response

1. The referenced Figures have been modified to clearly illustrate the scope of the Mark I and Mark F decommissioning. (

Reference:

Figures 1-3,1-4, and 1-5 on Pages 1-4,1-5 and 1-6)

Only facility areas and equipment that were essential to past operations and are needed to ,

support possession of the two reactors an: included and are termed the TRIGA* Reactors t Decommissioning Scope (TRDS). Other parts of the building and yard fall under the NRC and State site wide material licenses and the Site-Wide Decommissioning Plan. The TRDS is not i in conflict with other documents relating to the TRIGA' Reactors Facility.

Request:

2. Section 1.2.1. Because it is not licensed by the NRC, the NRC cannot approve decommissioning of the remaining components of the Mark III reactor. Please describe what steps,if any, have been taken with the State of California to decommission the Mark III and terminate that license.

Response

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2. References to decommissioning of the former Mark III reactor have been deleted.

Decommissioning of this area of the TRIGA* Reactor Facility will be conducted under the ,

i Site-Wide Decommissioning Plan. Approvals for this plan were given in NRC License SNM- l 696, Amendment #45, dated April 29, 1998; and CA State License 0145-37, Amendment

1. 127, dated August 26,1997.

Planning for the decommissioning of the former Mark III area has been completed but actual remediation is not scheduled until later in 1999. A DECON option similar to that planned for l the Mark I and Mark F has been selected. The plan calls for the building to remain through regulatory release.

Physical and operational separation are sufficient to assure that this D&D activity will have no effect on the Mark I and Mark F reactors. Physically, the Mark III reactor room and poo.!

structure are more than 25 feet from the Mark I and Mark F reactors and in a separate part of the facility. The Mnk III and associated lab rooms were added to the building complex some time after construction and initial operation of the Mark I and Mark F reactors. Connections to common systems such as the pool water heat removal and treatment system (already valved out) or electrical alarm systems have been or will be isolated. Remediation and waste packaging and removal operations will not intrude into the TRDS in the building.

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I Request:

3. Sections? 1.2.1 and 2.3. It appears from your decommissioning plan (DP) that you propose to completely demolish the reactor building. Your termination survey and NRC's confirmatory survey would be of the site. NRC normally approves DPs that propose complete demolition of structures when removing the activated and contaminated material threatens the structural integrity of the building and could have an impact on non-radiological safety. liowever, section 2.2.2 of your DP shows that the radioactive portions of the facility are confm' ed to the reactorinternals and biological shield. Also section 3.3 appears to show a decommissioning process where the radioactive materialis removed without jeopardizing the structural integrity of the reactor building. Please justify demolition of the reactor building before the confirmatory survey is requested from NRC or revise your DP to show the conduct of the tennination and confirmatory surveys on the decontaminated and dismantled building. Please revise other portions of your DP as needed.

Response

3. Further review of the levels of contamination in the TRDS and feasibility of maintaining building integrity while removing activated and contaminated materials has led to a change in the DECON approach presented in the plan. GA proposes that the building and yard areas comprising the TRDS can be remediated and demonstrated to be below appropriate release levels without dismantlement of the building. The plan has been revised throughout to indicate this change.

Request:

4. Section 1.2.2, What part of the total cost shown in the table is related to the decommissioning of the Mark I and Mark F reactors?

Response

4. The costs pmsented now represent the estimated cost of decommissioning the TRDS including  ;

only the Mark I and Mark F reactors. (

Reference:

page 1-11)

Request:

5. Section 1.2.2. It is mentioned that removal and shipment of fuel are not considered i decommissioning tasks. liowever, possession, on-site handling, and transportation must comply with NRC regulations and may require NRC authorizations. Explain how these fimetions are to be managed whenever fuel is on site before and during actual dismantlement and decontamination of the TRIGA* reactor facilities.

Response

5. Decommissioning of the TRDS is planned separately for the Mark I and the Mark F.

Decommissioning of the Mark I mactor does not affect the ability to meet all spent fuel material siorage and handling requirements following isolation of services such as alarms and the pool water treatment system which are cornmon to the FSC and the Mark I reactor. Thus, the Mark I reactor decommissioning can be completed and all Mark I equipment would te removed or remediated.

The Mark F decommissioning will be accomplished in two phases separated by the mmoval of spent fuel material stored in the Fuel Storage Canal (FSC) associated with the Mark F reactor pool. Although the Mark F reactor and the FSC share the pool structure, all of the 2

i equipment associated with the Mark F reactor itself (e.g. core grid plate, irradiation experiment housing, reactor controls, etc.) can be safely removed without affecting the Ltructures and equipment supporting safe fuel storage and handling. This will be accomplished by removing equipment from the reactor pool with minimum clearance to the pool and Door and removal of equipment without passing across the FSC. Crane interlocks  ;

and physical barriers (posts and chains or equivalent) will be utilized to ensure compliance  :

with this requirement. Further, equipment and structures not associated with the FSC (that is, everything but the entire Mark F pool, fuel maintenance and protection systems, and fuel  ;

handling systems) would be removed or remediated. By retaining all equipment and systems j necessary, all required fuel related functions (e.g. all Technical Specification requirements for fuel possession, on-site handling, and transportation) will be met in the same manner as they are now until appropriate approvals are obtained and the fuel is removed. Phase 2 of the Mark F decommissioning would then proceed by completing the rernoval or remediation of FSC related equipment and structures and the Mark F pool. (

Reference:

Section 1.2.1, page 1-10)

Request:

6. Section 1.2.3. Please provide infonnation that shows that General Atomics Technologies Corporation has been meeting the terms and conditions of their sinking fund discussed in your submittal of May 20,1996 (Reference 10.5). If the sinking fund does not contain sufficient funds to carry out decommissioning, where will actual funding come from for decommissioning activities?

Response

6. This is addressed in a separate letter because the response involves proprietary infonnation.

Request:

7. Section 1.2.4. Please provide a copy of the Quality Assurance Program Document prepared for decommissioning of the reactors.

Response

7. A copy of the Quality Assurance Program Document (QAPD) for the TRIGA* Reactor Facility i Decommissioning is attached.

Request:

l l 8. Table 2-1. Does the information in this table consider the activation of reactor structure and components such as the reactor tank, concrete and soil surrounding the reacter? For example, Europium-152 is a common isotope from the irradiation of concrete, but according to your table, is not an expected radionuclide. Do your calculations show that soil was activated by operation of the reactor? If so, what volume of soils is activated?

l Response:

8. Table 2-1 (page 2-6) has been revised to address all expected radionuclides for release of the building structures as well as soil. Calculations indicate that no activated soil has resulted from opration of the reactors. Decommissioning experience of the University of Texas TRIGA > reactor also supports this conclusion (

Reference:

"The University of Texas at Austin Decommissioning Plan for the U of T TRIGA Reactor, College of Engineering, Taylor Hall, 3

Room 131," License No. R-92, Docket No. 50-192). Sampling is planned during final surveys to confirm this conclusion.

Request:

9. Sections 2.2.3 and 4. Please discuss the difference between criteria to release the site and structures for unrestricted use at the end of decommissioning activities and the criteria to dispose or remove potentially contaminated materials from the site during the decommissioning process.

a. Please clearly discuss criteria for release of the site, structures and components for unrestricted use at the time of license termination. Address both contamination and radiation fields. Please see the new Subpart E of 10 CFR Part 10 " Radiological Criteria for License Termination." Because, as discussed in 10 CFR 20.1401(b)(3), you have submitted your DP before August 20,1998, the release criteria applicable to your application are the appropriate sections of " Cleanup Criteria" for reactor licenses in the " Action Plan to ensure Timely Cleanup of Site Decommissioning Management Plan Sites" dated April 16,1992 (57 FR 13389). Please note that some guidance in this document is for materials licensees and may not be applicable to your situation.

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b. Please discuss your plans for disposal of radioactive material created by decommissioning activities. This question refers to material that will not be released for unrestricted use by l termination of the reactor licenses.

Response

9. A Section 2.8 has been added which clearly addresses the criteria for unrestricted release of the site, structures, and components at the time of license termination. Briefly, the criteria given in Section 2.8 are those found in Regulatory Guide 1.86 and a dose rate due to residual radioactivity of not more than 5 microR/hr above natural background measured at 1 meter from the surface. A Section 3.1.3 has also been added which discusses appropriate controls -

including survey methods and detection capabilities - of materials and/or items removed during deconunissioning. Briefly, the criteria is that no discemible facility related radioactivity will be released. Clarifications have also been made to Sections 4.1,4.2,4.3, and Appendix B.

i Radioactive waste material generated during Mark I and Mark F decommissioning activities will be sent to the Nevada Test Site (NTS) waste treatment, storage, and disposal facility operated by DOE for wastes which meet the NTS Waste Acceptance Criteria (WAC).

Permission from DOE to utilize this facility is based on a specific agreement between GA and l DOE covering site decommissioning funding and waste disposition responsibilities. Waste l

I which cannot meet the NTS WAC such as radioactive contaminated lead or cadmium (mixed waste) will be sent to other authorized waste facilities such as Envirocare in Utah or treated to i render the waste non-hazardous prior to disposal.

Request:

10.Section 2.3.1, Table 2-4. The items in this list are normally considered facility components that are removable only after the decommissioning amendment is approved. For the NRC licensed reactors please commit to not removing these components until the decommissioning amendment is issued or provide justification for removing these components under the l possession only amendment. If removal of these components will occur under the l decommissioning amendment, please revise the DP accordingly.

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Response: ,

10.GA will not remove items normally identined as facility components that are removable under a decommissioning plan until the plan is approved. Table 2-4 has been deleted to reflect this commitment. In addition, the text has been revised to indicate how items may be removed under 10CFR50.59 where allowed (

Reference:

Section 2.3.1.1.4, page 2-10) (Note that Table 2.4 of the revised Decommissioning Plan was Table 2.5 in Revision 0.)

Request:

i i 1. Section 2.3. Under 10 CFR 50.82(b), an amendment to existing reactor licenses will be issued by NRC to authorize dismantlement and decommissioning. These amended licenses will include relevant Technical Speci6 cations (TSs) and other features, including applicable ,

Physical Protection and Emergency Plans. Please submit proposed documents for our review ~

and evaluation. If you wish to use the existing documents please state this. You may want to consider developing these documents to address relevant issues both while any fuel is on site, and after fuel removal.

Response

11.While fuelis still present in the FSC, GA has elected to utilize the existing TSs and Physical Protection and Emergency Plans. The commitments made in these documents can be met during Phase 1 of the decommissioning project because of the independent or separable nature of the various structures, equipment, and services. GA will reevaluate the TSs and upon removal of the fuel and submit appropriate changes for NRC approval if warranted.

Request:

1 12.Section 2.3. Use of a localized fme mist of water is mentioned. How will the contaminated water be disposed of?

Response

12.Section 2.3 was revised to state that contaminated water will be minimized where practicable through collection and recirculation. When contaminated water is generated, it will be processed at GA's NWPF or utilizing an equivalent system through settling, filtration, ion exchange,or other appropriate means to remove contamination as authorized by GA's Broad Scope Radioactive Materials License (No. 0145-37) issued by the State of California and meet release criteria for the sanitary sewer. This is consistent with approved current GA practice.

Water that cannot be processed to meet requirements in such a manner will be solidified (stabilized) with portland cement and dispositioned as low-level waste.

Request:

13. Section 3.1.2. The DP refers to approved procedures. Please explain the approval process and its relationship to the reactor TSs.

Response

13. Procedures have been utilized to implement the requirements of the TSs. These procedures are prepared or reviewed by the PIC and reviewed by the CRSC utilizing a formal system de6ned in the administrative procedure governing operations under the reactor possession only license. These procedures are for the specine implementation of the TSs requirements.

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Procedures for the decommissioning of the reactors unuer an approved Decommissionmg I Plan are prepared to control the operations under this plan. These procedures am prepared and approved in accordance with the QAPD and are reviewed by the CRSC to ensure that there is no impact on the ability to meet TS requirements under reactor possession only conditions with the fuel located in the FSC.

Section 3.1.2 addresses corporate Health Physics (HP) procedures which are controlled in ,

accordance with an HP administrative procedure. The HP procedures supplement the reactor TSs and also address topics not covered by the TSs such as work performed by HP Department personnel. These procedures are reviewed and approved by the HP Manager and the Director of LSNC.

Request: l l

14.Section 3.1.3. Please discuss a dose estimate to members of the public from l decommissioning activities.

Response

14. Site perimeter access controls will restrict members of the public from entering the area surrounding the facility in which decommissioning activities will be taking place.

Accordingly, the dose to members of the public as a result of decommissioning activities is estimated to be negligible. This is consistent with the estimated dose to members of the public l

as a result of decommissioning the reference research reactor as given in the " Final Generic Envimnmental Impact Statement on decommissioning of nuclear facilities," i.e., NUREG-0586. More specifically,in NUREG-0586 the dose to the public due to decommissioning and truck transport of radioactive waste from decontamination and decommissioning "at the reference research reactor" (which like General Atomics' facility, is a TRIGA reactor) "is estimated to be negligible (less than 0.1 person-rem)."

l Request:

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15. Table 3-2. Please revise this table to clearly show the collective dose from decommissioning the NRC licensed reactors separate from the State of California licensed material.

Response

l 15. Table 3-2 (page 3-11) has been revised to include only those activities associated with the i scope of the Decommissioning Plan. I Request-l

16. Section 3.3. Please discuss the potential consequences to the public and the workers, assuming a poolleak while fuel is still on the site, perhaps initiated by nearby dismantlement activities, or even by a seismic event.

Response

16.The following information was added to Section 3.3.

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Consequences of a pool leak are low because the pool water is continuously treated and contains negligible radioactivity. The main function of the pool is to provide shielding for workers positioned near or over the FSC during any required handling, inventory related, or l

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training operations. The water is not required for fuel cooling. Any failure to meet shielding requirements would result in worker restrictions on approach to the FSC until the requirement cbuld be met. The other potential consequence would be due to flowing water carrying loose contamination to a new location within the facility, outside the facility, or underneath to the soil. Since loose contamination is minimal, the risk of the spread of contamination is low. GA is experienced and prepared to contain and decontaminate any spread of contamination. There is no potential for airborne contamination from such an event.

Request:

17. Figure 4-3. Please discuss the use of the sampling system depicted in this figure.

Response

1 17.The Decommissioning Plan has been revised to clarify the sampling approach to be utilized l and the figure has been deleted.

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Request:

18.Section 6. Please discuss continued compliance with Parts 50 and 73 for your physical security plan.

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Response

18. Section 6 was revised to add the following information.  ;

1 Compliance with Parts 50 and 73 of the physical security plan is assured because all of the I elements of the plan an: maintained until the fuel is removed off site. Decommissioning  !

activities will be conducted with the fuelin the FSC and all physical security and surveillance l in place. Once isolation of common services has been implemented, the security of the Mark F reactor and control rooms can be maintained without impact on decommissioning activities in the Mark I reactor room. Workers will be limited in number and appropriately trained before entry to the site. Oversight will be provided during any entry into the Mark F room for i

purposes of activities, e.g. reactor equipment removal, not associated with the FSC or fuel.

Request: l 19.Section 9. Your definition of unreviewed safety question is different from that in 10 CFR 50.59. Amendments to the regulations in August 1996 extended the application of 10 CFR 50.59 to NPRs whose licenses no longer authorize operation of the reactor. Please amend this section of your DP to be in accordance with the regulations.

Response

l 19.The text has been revised to be consistent with the regulations.

Request:

20. Appendix A. With the change in license status of the Mark I to possession-only, do you plan to conduct additional characterization measurements? If so, please describe the results of those measurements, or if not yet conducted, describe the planned measurements.

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Response: ,

20. Sufficient characterization combined with information from decommissioning of similar reactors formed the basis for the approach outlined in the Decommissioning Plan. Additional measurements will be made, as appropriate, during initiation and/or execution of each major activity planned in remediating the reactor.

Request:

21. Appendix B, Section 2.2, page B-8. Please give best estimates for the schedule for fuel shipment off-site, and for submitting the request to NRC for license terminations.

Response

21. DOE is contractually committed to take the spent fuel at INEEL. (A letter documenting the planned shipping date has been requested from DOE.) The Decommissioning Plan has been modified to describe the arrangements and shipment " schedule" in Section 2.3.2. The shipping schedule and the time required to complete Phase 2 of the Mark F plan, in turn, determine the date for the license termination request. Our best estimate for fuel shipment:

July 2003, submitting request for license termination to NRC: January 2004, based on receiving DOE permission to ship fuel on May 1, 2003. Any slippage from this date will result in day-to-day slippage of the July 2003 and January 2004 dates. j Request:

22. Appendix B, Section 3.1.2. Are any of the hazardous materials discussed in this section  :

activated or contaminated? Please discuss types and quantities of any mixed wastes that result from decommissioning.

Response

22.The generation of some mixed waste is anticipated during the decommissioning process based ,

on the knowledge that lead and cadmium have been located in close proximity to the reactor as '

part of the equipment utilized in performing experiments. We estimate 45 cubic feet of radiologically contaminated lead and S cubic feet of radiologically contaminated / activated cadmium. Appendix B, section 4.2.2 of the plan has been revised to specifically address the points raised in this question.

Request:

23. Appendix B, Section 4.1.2, page B-19. Please discuss further your bases for estimating zero potential exposures to the public after license termination. What release criteria are assumed? .

Should NRC consider this to be GA commitment?

Response

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23. The text has been revised to indicate that the potential exposures to the public after license i

termination are negligible. Remediation will be conducted such that the surface contamination criteria given in Regulatory Guide 1.86 are met and the exposure rate does not exceed 5 micro-R/ hour above background at one meter above the surface.

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Request: ,

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24. Section 4.1.3, page B-20. Please discuss the basis for the number of truck trips for hazardous and radiological waste. What total quantities of these materials are assumed to be involved? Ilow many trips will be required to transfer all irradiated TRIG A fuel off the site?

Response

24. As stated in Section 4.1.3 (page B-23), it is expected that the total number of truck trips necessary to ship all TRDS radiological and hazardous waste will not exceed twelve. The total quantity of radiological waste (including mixed waste) is estimated at less than 4000 cubic feet. The total quantity of hazardous waste is estimated at less than 50 cubic feet. Fewer than eight truck trips are estimated to be sufficient to remove the fuel from the site.

1 Request: j

25. Section 4.2.3, page B-20. Please give the basis and best estimates of quantities of uncontaminated construction debris to be trucked off the site.

Response

l 25.The selection of the decontamination option rather than dismantlement has resulted m a substantial reduction in the amount of clean waste associated with the project. A total of less than 20,000 cubic feet of clean waste has been estimated. This is based on an evaluation of the amount ofinterior material and equipment which must be removed to access all activated and contaminated areas requiring remediation or confirmatory surveys. l Request:

j 26.Section 4.8, page B-22. Please discuss human health effects on decommissioning workers. I Please give best estimate (and explanation) ofinternal doses to the public.

Response

26. The total dose estimated for decommissioning workers is 20 person-rem for the entire project evolution. This estimate will be achieved by utilizing ALARA practices including planning of work activities, utilization of engineered safeguards, and minimization of exposure times. The decommissioning will be conducted under a Work Authorization system using written procedures to ensure proper planning, training, and evaluation of potential risks. It should be noted that a total dose of 20 person-rem is consistent with 18.6 person-rem given in Table 7.3-3 " Summary of radiation safety analyses for decommissioning the reference research reactor" of the " Final Generic Environmental Impact Statement on decommissioning of nuclear facilities"(NUREG-0586).

The internal dose to members of the public as a result of decommissioning activities described in GA's TRIGA facility decommissioning plan will be negligible. The dominant intemal exposure pathway for members of the public is inhalation. The dose to the public is estimated to be negligible because access to the area surrounding the facility is restricted and because decontamination activities with potential for airborne activity will be conducted utilizing engineered safeguards such as HEPA equipped enclosures. Further, continued operation of the facility HEPA system provides additional protection for all decontamination activities ,

conducted within the building. Thus, potential airborne radioactive will be negligible; and, l therefore, the potential internal dose to the public is also negligible. '

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The estimate of negligible internal dose to members of the public can also be obtained from the  !

estimate, given for the aference research reactor in the " Final Generie Environmental Impact '

Statement on decommissioning : nuclear. facilities" (NUREG-0586). In Section 7.3.1 of ,

NUREG-0586, the dose to the public as a result of' decommissioning operations at the '

refennce research reactor - including truck transportation of radioactive waste - is " estimated -l to be negligible (less than 0.1 person-rem)." This estimate of less than 0.1 person-rem i includes both internal (from inhalation and ingestion) and external exposure doses. j i

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