ML20115K051

Forwards Responses to 960606 Request for Addl Info Re Questions Resulting from Review of SNEC Facility Decommissioning Plan & Environ Rept

ML20115K051
Person / Time
Site:	Saxton File:GPU Nuclear icon.png
Issue date:	07/18/1996
From:	Kuehn G GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20115K054	List: ML20115K051 ML20115K082 ML20115K114 ML20115K141 ML20115K145 ML20115K165 ML20116A044 ML20116A058 ML20116A073 ML20116A081 ML20116A110 ML20116A119 ML20116A132
References
C301-96-2038, NUDOCS 9607250207
Download: ML20115K051 (46)

Text

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GPU Nuclear Corporation U GM Route 441 South P.O. Box 480 Midd!ctown. Pennsylvania 17057-0480 (717)944-7621 Writer's Direct Dial Number.

1 17) 948-8720 July 18, 1996 C301-96-2038 U. S. Nuclear Regulatory Commission Attn: Document Control Desk V?-^ington, D.C. 20555 Ow..tlemen:

Subject:

Saxton Nuclear Experimental Corporation Operating License No. DPR-4 Docket No. 50-146 Response to the Request for AdditionalInformation Regarding the SNEC Facility Decommissioning Plan and Environmental Report Enclosed are responses to the June 6,1996 Request for Additional Information regarding questions resulting from the review of the SNEC facility Decommissioning Plan and Envirc.anental Report.

The responses to questions are complete to the extent possible. In those instances where responses are incomplete, the provided response addresses the specific information required and commits to provide a full response following completion of a required action or obtaining the infonnation. Requests for specific documents were handled in a similar manner. Consideration j was given to maintaining the level of docketed information from the GPU Nuclear facilities consistent and therefore, GPU Nuclear has chosen not to submit specific documents. The documents handled in that manner are identified in the response to the respective request.

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Sincerely, l

9607250207 960718 6 ADOCK 0500

g. i

. A. Kuehn Jr.

%DR Vice President SNEC ,

WGH Attachments .

h cc: Administrator, Region I /

NRC Project Manager NRR g NRC Project Scientist, Region I g & 'l \

/Y c' i lh v.fi)<7 ; I 9/[ ' I fy {

C301-96-2038 Attachment 1 Page 1 of 44 1 Page 2-2, Section 2.1.2 - There is a mention of high ground water. Please discuss any safety implications of this condition as it might relate to de dismantlement, decontamination, and eventual site restoration of the Saxton Nuclear Experimental Facility (SNEF).

Response: The "high ground water" condition referred to in section 2.1.2 of the Decommissioning Plan (DP) does not impose any safety implications related to dismantlement, decontamination, or site restoration of the Saxton Nuclear Experimental facility. The reference to the ground water conditions as they exist at the site is made to point out the risk (s) involved with not moving forward with decommissioning of the Containment Vessel (CV). As stated in section 2.1.2 of the DP, failure or breach of the CV liner could lead to intrusion of uncontrolled quantities of ground water into the I structure. This would greatly complicate the decontamination and dismantlement (D&D) l effort and increase the cost. Prompt removal of the CV while the liner is intact will I remove this threat. l The current ground water conditions have existed since the formation of Raystown Lake following construction of the dam. Those same conditions were present during the j previous major decommissioning phases in 1972 through 1974 and again from 1986 through 1992. During these periods, major D&D activities were conducted while  ;

managing the ground water situation using a variety of conventional industrial techniques such as mechanical diversion, collection in sumps and pumping, and isolation. Discharges of ground water were made to the nearby Raystown branch of the Juniata River as required to keep ground water from interfering with D&D activities.

It is anticipated that similar methods will be employed to accomplish the D&D of the service tunnel which surrounds the CV and the dismantlement of the CV following free release. .

2. Page 2-2, Section 2.1.2 - The Saxton Nuclear Experimental Corporation (SNEC) has studied and submitted to the NRC in the past information on the degradation of the SNEF steel shell. Please summarize the results of the most recent information you possess on degradation of the SNEF steel shell.

Response: The most recent assessment of corrosion of the Saxton Nuclear Experimental Corporation CV shell was made in 1995. It concluded that a worst case scenario would result in through wall penetration of the steel liner by the year 2015 and would most likely occur in an area of ground contact. It addressed active protective measures and stated they were unnecessary to ensure containment shell integrity below grade if decommissioning is accomplished by the year 2000, but may be needed if decommissionirg is significantly delayed. Although interior surface corrosion of above grade welds on the steel liner was observed, ultrasonic testing (UT) of these areas confirmed they are not at as great a risk as the below grade surfaces.

C301-96-2038 Attachment 1 Page 2 of 44

3. Page 2-3, Section 2.1.2, Second paragraph - there is mention of concern about the flood plain. Please discuss any safety implications of this condition as it might relate to the dismantlement, decontamination, and eventual site restoration of the SNEF.

Response: According to records kept by the U.S. Geologic Survey (USGS) from 1889 to present, the maximum observed flood level at the site occupied by the SNEC facility was 809.5 feet above mean sea level (msl) recorded in March of 1936.

In a study performed in 1969, the Army Corps of Engineers (ACOE) concluded that the 225 year period of flood recurrence level was at 812.0 feet above msl. The Federal Emergency Management Agency (FEMA) and the ACOE were both contacted in March ,

1996 to verify the validity of the projections. Both agencies indicated that no additional l l

studies have been performed which would update this data nor was there any data to contradict these projections.

l The flood of record and 225 year recurrence projection are referenced in the 1972 Saxton Decommissioning Plan and Safety Analysis Report, previously submitted. l The predominant grade elevation for the decommissioning support structures is j approximately 812.5 feet above msl. The finished slab elevation of the Decommissioning l Suppon Building will be approximately 813.5 feet above msl. This places the main  !

1 structure and that portion subject to the effects of flooding above both the 100 year and 225 year flood recurrence elevation. Therefore the structures will not be affected by any i reasonable flooding event at the site.

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Any flooding which could affect the structures would occur with sufficient notice to  !

permit implementation of steps to minimize any impact to the public. )

The installation of support services for the Containment Vessel (CV) will use existing penetrations all of which are above the referenced flood levels. Additionally these l

installations will be engineered so as to ensure a seal can be made to prevent water l intrusion into the CV. l l

4. Page 2-5, Table 2.1 Please give more details about the information on which this table is based. Discuss major assumptions, for example, which radionuclides would lead to the majority of the predicted doses, and how much of the doses would be caused by external exposures and how much by internal body burdens.

J Response: The occupational dose estimate is based on external exposures only.

Engineered controls and/or personal respiratonj protection would be designed to minimize internal body burdens (ingestion or inhalation caused doses). The occupational doses j were based on the results of radiological surveys taken by SNEC facility radiological i

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C301-96-2038 Attachment 1 Page 3 of 44 controls personnel from 1991 through 1995. These surveys provide both component

" contact" and "gcneral area" dose rates.

Dismantling activities would be sequenced such that equipment with the highest source of radiation exposure would be removed as early as possible, to reduce " general area" radiation fields.

The total occupational dose for each task (i.e., asbestos remediation, systems dismantlement, etc.) is the sum of the occupational doses received while performing each of the individual activities (i.e., work area set-up, pipe removal, pump removal, etc.)

performed under that task.

Individual activity doses were estimated by multiplying the estimated hours spent in the radiation field, by the dose rates indicated on the radiological surveys. Both " general area" and " contact" dose rates were accounted for in estimating occupational exposure.

The use of ALARA practices and a reduction in " general area" dose rates as the decommissioning work progressed were considered in estimating the occupational doses.

Waste Management Task doses were estimated as a ratio (percentage) of the exposure received by the decontamination and dismantling workers, and Health Physics technicians.

The ratio was identified by an analysis of occupational exposures recorded at a similar decommissioning project.

To account for the decay in the source term between the time of the radiological survey (1991 through 1995), and the year (s) of"Immediate Dismantlement"(1997-1998), the occupational dose estimate was reduced by 4.5% (based on the decay of Cs-137 for 2 years).

The Occupational Dose Estimate for the "30 Year Deferral" Decommissioning Altemative was developed based on the assumption that the occupational dose for the "Immediate" Decommissioning Alternative was reduced based on the decay of Cs-137 for an additional 30 years. The isotope Cs-137 was conservatively selected as the basis for the decay rate since it has a longer half-life than Co-60.

5. Page 2-6, Section 2.2 - The report mentions additional detailed engineering and planning here and in other places. Please give a projected schedule for NRC's receiving that information for review.

Response: The additional detailed engineering and planning mentioned on Page 2-6, Section 2.2 will be performed as preliminary activities are completed and permits advanced work to proceed. It was not the intention of GPU Nuclear to include this material in the decommissioning plan: the statement was meant to constme that continued planning and engineering work would go on as decommissioning tasks progressed.

Therefore, no projected schedule for NRC's receiving that information for review will be

C301-9A2038 l Attachment 1 Page 4 of 44 provided. The information will be available for NRC review during site activity inspections or prior to performance of activities ofinterest to NRC which are identified to GPU Nuclear.

6. Page 2-6, First full paragraph - Scabblers and CO2 blasters may not be familiar terms to most people and should be described in an appropriate place in the Decommissioning Plan.

Also, the last sentence states that use of water will be minimized. Does this imply that water will not be used for dust control purposes? Please clarify.

Response: The decommissioning plan is a technical document purposefully written for individuals with experience and a familiarity with the terminology of the nuclear power industry and activities associated with the decommissioning of a plant. It is expected that people not familiar with the terminology would make an effort to understand concepts or new material encountered. The two terms chosen as an example to be described are very simplistic. If an effort were persued to describe them and the numerous others like them, using a similar argument, it would result in a sizable expenditure of effort with little value added.

Concerning the use of water, it will be used for tasks where an advantage will result from its use. The sentence in question was meant to explain that water will not be used in excess thereby causing storage control and liquid waste disposal problems which would be associated with the use oflarge volumes ofwater.

7. Page 2-9, Number 4 - You refer to radiological characterization of portions of systems to ensure that release criteria have been met. How does this differ from radiological surveys to ensure that release criteria have been met?

Response: In the manner used in the referenced section, the terms " characterization" and

" survey" should be considered synonomous.

8. Page 2-10, Paragraph A.l. - You discuss work instructions. Are these generalinstmetions or developed specifically for each job?

Response: Both general and task specific work instructions will be developed and implemented as needed and appropriate for thejob. For example, a general work procedure may be developed on the use of a specific cutting tool, however the complicated removal of a large component which might require the use of that tool would have a task specific work instruction. Tasks will be reviewed for complexity and overall risk to determine the need for a specific instruction or procedure.

All radiological work affecting quality, health and safety of the public or project personnel, or regulatory requirements will be performed using written procedures. See section 7.3.4 of the DP.

C301-96-2033 Attachment 1 Page 5 of 44

9. Page 2-10, Paragraph A.2. - Please provide a discussion on your plans for draining contaminated water from piping and proposed methods for handling the liquid.

Response: As mentioned in section B. of page 2-10 of the DP, systems were drained as part of the plant shutdown process in 1972 and little residual water is expected. Indeed during characterization activities most liquid systems were breached and only minor amounts of condensed liquid were encountered, generally less than one gallon.

During dismantlement, any contaminated water will be collected as described in Section B and stored in suitable containers for eventual processing. Processing options include solidification, purification or shipment to a qualified off-site vendor for processing and disposal. Several hundred gallons of contaminated water is currently stored in the CV awaiting processing /. GPU Nuclear expects any contaminated water drained from piping

to be a small fraction of this volume.

i 10. Page 2-10, Paragraph A.4 - There is in general need for more detail on the radiological control program. An example is noted here. Please provide more information on contamination barriers, catch basins, and use of continuous airbome monitors, giving more details of the plans, including filters and other means of controlling radioactive materials or preventing releases from enclosures to occupied areas.

Response: The decommissioning plan is meant to present plans for the SNEC facility decommissioning in general terms. Details of activities such as radiological controls are contained in long established GPU Nuclear programs which will be implemented in support of the decommissioning of the SNEC facility. The example noted cannot be specifically addressed since consideration of radiological conditions, equipment installation accessibility, and the experience of personnel on equipment choice will vary from location to location. The Radiological Control Program and procedures provide both the means to satisfy regulatory requirements and suflicient freedom to choose the best method for controlling radioactive materials or preventing releases from enclosures to occupied areas.

The program manuals and procedures are available for review during site inspections.

I 1. Page 2-10. Paragraph A.4. - A number of places were noted where qualifying language is used unnecessarily, thus detracting from the value of the commitment. An example is noted here. The Decommissioning Plan should contain a firm commitment on the utilization of the listed controls with a "will be employed" rather than a "may be employed" Another example appears on page 2-15 under " Concrete Shield Wall" The last sentence begins with "If applicable, . " Should not contaminated rubble always be processed as radwaste?

Response: The use of" qualifying language" provides GPU Nuclear the option to use various controls including some which may not currently exist and those which will be developed through actual work practice.

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C301-96-2038 Attachment I ;

Page 6 of 44 l l

Concerning the second example of the concrete shield wall, the reference was meant to provide assurance that contaminated rubble would not be disposed ofin a local landfill and that GPU Nuclear would strive not to include non contaminated rubble as rad waste. ,

This would eliminate unnecessary shipments and a need to process the specific material as j rad waste. These decisions will be made as the project proceeds.  ;

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12. Page 2-12, Number 4. - Discuss removal procedures for " inaccessible piping" including methods and radiological status and impact. How will the integrity of structures disturbed to remove piping be ensured?

l Response: The " inaccessible piping" refers to that piping whose exterior surface is not readily accessible. The piping in this catagory consists of floor drains embedded in concrete and those sections of process piping which pass through concrete walls, floors, and ceilings. All of this piping is considered internally contaminated and will be treated as such until proven otherwise by survey / sample. The radiological status of all piping is described in the Characterization Report.

Inaccessible piping will be removed either as the concrete is removed around it or by core boring.

As required by OSHA Regulation 29 CFR 1926.850 Subpart T-Demolition, an engineering survey must be performed prior to any demolition operations to determine the possibility of collapse of any portion of the structure. Such surveys wil be performed and measures implemented as required to ensure structural integrity. Such measures would include; floor jacks, shoring, floor loading restrictions, or complete structural removal.

Much of the process piping passes through wall penetrations and its removal will have no impact on the structural integrity.

13. Page 2 There is mention of piping meeting the release criteria. Discuss the methods to be used to relate radiation survey results or measurements to the release criteria. Give

& tails of the basis for setting thresholds on go/no go limits.

Response: Most of the piping in the SNEC CV was internally contaminated during plant operation. All piping samples removed during characterization efforts have shown some level ofinternal contamination. Because of the presence of the reactor vessel (RV) and the high levels of contamination found there, piping within the reactor cavity is either activated or contaminated. Therefore, most if not all of this piping and components will be disposed of as radioactive waste. However, there is a possibility that some potentially clean piping may be found in other parts of the site during dismantlement.

When radiation above background is detected on accessible internal portions of pipiag sections, remaining piping or tubing sections from the same system will be assumed to be above release criteria. In addition, when an in-line trap, drain or catch point retains

C301-96-2038 Attachment i Page 7 of 44 be internally contaminated above release criteria. contamination The presence of sufTicient gamma emitters in a piping system would allow d sensitive to gamma emitting radiation to be used to identify relevant radionucli sections of small bore piping or tubing. Ifidentified, these pipe sections would b disposed of as radioactive waste.

The internal surfaces oflarge diameter piping can be accessed with many stan large bore piping as acca permits. detector probes attached to ex In all cases so far encountered, radiation is internally contaminated above release criteria. measurements alo Small diameter piping and tubing may require other measurement or t wipe test through the internals of a tube or small bore pipe can detect the pr loose surface contamination.

present to be above release criteria.If detected, it is likely that there is suflicient contam Under this condition, the pipe or tube would be disposal as radioactive waste. assumed to be, contaminated above release When survey techniques are inconclusive or cannot establish that a piping section retains only background levels ofradionuclides the pipe / tubing section will be assumed to be contaminated above release criteria.

Survey techniques for release of potentially clean piping include calibrated dete extended into piping runs in a controlled fashion.

Detectors typically used for direct measurement of total surface activity within embedded piping include cylindrical g proportional detectors with appropriate diameters. In addition, Geiger-Mueller de and other detector types and sizes for specific situations may also be used.

in the past has developed or assisted in the developme entation for selected situations. That included the uranium / fuel debris hold-up measurem TMI-2 to assess the remaining core inventory scattered throughout plant s accident.

The previous D&D effort involving the reactor support buildings at the SNE Other field use equipment includes portable HpGe an sensitivities.

on final survey activities at the Fort St. Vrain facility on Service Co. of Colorado. ,

and the challenges involved withInthe survey of piping.This work has k addition, GPU Nuclear will be supporting the SNEC facility decommissioning. ma 14.

Page 2 Do you anticipate that you will be able to remove all contaminated a activated components that do not meet the release criteria before demolition of t containment vessel (CV) or will some components and areas be unaccessible u

C301-96-2038 Attachment 1 Page 8 of 44 is demolished? If some areas must wait until demolition to be surveyed, please discuss your procedures for accomplishing this.

Response: Yes, it is anticipated that all contaminated and activated components that do not meet the release criteria will be removed before demolition of the CV, l

15. Page 2 How will you ensure that unacceptable amounts of radioactive material will not be present next to and under the CV structure len on site?

Response: The Final Survey Plan, which will be submitted for NRC approval, will contain detail on how the applicable release criteria will be met and what surveys will be required.

Deep core bores have been performed on the north side of the CV and have not indicated the presence oflicensed radioactive material above current release limits. Other areas next to the CV are inaccessible at this time due to the tunnel structure which surrounds the CV.

) In addition to the core bore results, ground water monitoring near the CV has never detected any gamma emitting isotopes attributed to licensed operations.

W As part of the Final Survey, additional samples near the CV and outside the pipe tunnel will be required to verify the site meets free release criteria. Once the CV interior is l decontaminated, core bores could be used to sample the area immediately under the CV.

Based on the deep core sample results, the ground water monitoring and the absence of ground water intrusion into the CV, we have confidence in the below grade integrity of the CV liner and have no reason to suspect contamination of the area next to or under the CV.

! 16. Page 2 Please provide estimated dose rates expected to be experienced by personnel during removal of the reactor vessel, pressurizer, and steam generator and during the packaging activities for each vessel. How were these determined? What are the predominant isotopes? Also, specify expected dose rates around each vessel after it has been packaged for transport to the disposal facility. Discuss ALARA considerations 4 related to the removal of these vessels. Please discuss in detail how the reactor vessel (and steam generator and pressurizer, if lifled out the CV roof) will be lifted from the CV.

Provide information such as crane type and set up to be used and safety precautions to prevent dropping of components.

Response: Dose rates and predominant isotopes for the components listed are given in the characterization report which is furnished. Reductions in dose rates for these components

during removal cannot be projected at this time as the specifics associated with component removal are not yet in place. To be conservative, the as found unshielded dose rates were used for the person-rem estimates. Dose rates for the pressurizer and steam generator are sufliciently low so as to already meet the normal conditions of transport with the exception of a small hot spot on the steam generator. The reactor vessel will require the

C301-96-2038 Attachment 1 Page 9 of 44 application of some shielding but the sequence has not been decided nor are the specifics of the shielding known at this time.

ALARA measures to minimize exposure during removal of these vessels will be implemented in accordance with the Radiation Protection Plan, such measures would include: the use of temporary shielding, mock up and pre-job training to minimize the time spent on the task, the use of experienced personnel, the incorporation oflessons learned from previous similar jobs, the use of cameras to govide remote viewing, the use oflong handled tools, restrictions on the number of personnel near the work site, etc.

Detailed planning related to removal of the RV and the type of crane needed to accomplish the task is being developed.

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17. Page 2 Please provide more details on methods and procedures to be used in the removal and replacement of the section of the CV to permit removal of the reactor vessel.

What special precautions will be employed to ensure that these actions will be accomplished safely? Will the use of a covering on the new CV opening affect releases assumed for evaluated accident scenarios?

Response: Details related to the removal and replacement of the CV section to permit the reactor vessel removal have not yet been developed. Use of a process very similar to that employed at the Pathfinder decommissioning project is envisioned. It incorporated cutting a hole in the CV dome to provide access for a crane to remove the reactor vessel. The removed section was replaced and sealed.

Proper industrial safety and lifting and rigging practices will be employed to ensure safety is not compromised. Radiological safety will be ensured by monitoring to prevent any unmonitored releases and by implementing contamination control measures to prevent the spread of radioactive material. The use of a replacement cover will not impact releases evaluated as part of the accident scenarios analyzed in section 3.4 of the DP.

The use of a covering on the CV opening created to remove the reactor vessel will not affect releases assumed in the accident analyses. The covering will permit the CV to provide the same degree of environmental boundary that existed prior to installation of the opening.

18. Page 2-23, First paragraph - Explain " fixed or contained within a plastic barrier?"

i Response: The term " fixed" is used to indicate that any loose surface contamination will be immobilized to prevent it from spreading. This may entail the use of a " fixative" agent such as paint, strip coat, foam, encapsulant, etc., hence the term " fixed" In lieu of or in combination with these measures a barrier made of plastic or a plastic like substance may also be used as a cover to prevent the spread of contamination. j 1

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C30196-2038 Attachment 1 Page 10 of 44

19. Page 2-23, Section 2.2.1.4.4 - In referring to 10 CFR Part 71, please provide evidence j that the quality assurance requirements of Part 71.101 have been approved.

Response: A GPU Nuclear program will be developed / revised and implemented to I conduct 10 CFR 71 activities prior to the performing the affected activities.

20. Page 2-25, Section 2.2.1.4.5 - How confident are you that the roadway stmetures, l

j bridges, etc., along the proposed transponation route (both in Pennsylvania and South Carolina) will accommodate the weight and dimensions of the reactor vessel? Has the proposed route been defmed and evaluated? If so, please identify all potential problem areas and describe anticipated resolution requirements, including the need for coordination with the duly constituted entities or authorities holding jurisdiction over the structures.

. Response: Packages of greater weight have traveled over much of the route expected to be used. Preliminary discussions with several experienced vendors indicate that the route

chosen will accommodate the transport of the large components without difliculty.

Details concerning the route and packaging specifics for the reactor vessel have not been developed. A qualified vendor, experienced in similar projects will be employed to assist with the packaging and shipment of the large components. A detailed transportation plan 4 will be developed to ensure the route is suitable, the transportation is performed safely and the interested parties along the route are informed about the plan and the status of the shipment.

4 It is expected that the NRC Transportation Branch will be required to approve the shipment of the reactor vessel. The details relative to the transportation of the reactor 1

vessel will be covered fully in submittals made to obtain specific approval for shipment. It is not assumed that approval of the DP would abrogate the approvals required for specific aspects of the decommissioning including such shipments.

21. Page 2-28, Table 2.2 Please give the bases for and assumptions used in developing this table.

Respons: The Occupational Exposure for the "Immediate" Dismantlement Alternative was developed based on the same assumptions used in developing Table 2.1-1, 3 "Immediate" Decommissioning Alternatives. The table values are the same. See response to question 4.

22. Page 2-30, Section 2.3 - Please provide a description of the qualifications, experience, and i

responsibilities ofindividuals in positions important to the management of the decommissioning activities. Also, please describe the administrative controls to be used

to ensure adequate health and safety protection for work to be performed by contractors, and plans for ensuring that the contractors are adequately qualified and experienced cn the subject of radiation safety.

C301-96-2038 Attachment 1 Page 11 of 44 Response: These documents are available for inspection and may also be furnished upon request.

Personnel holding key management positions have a strong background in their specific fields. These areas of expertise include complex program management, supervision and oversight, planning and scheduling, radiological : antrols practices, radioactive waste packaging, and decontamination.

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Individuals assigned to the SNEC facility decommissioning project have a long history of J successful involvement with the TMI-2 clean-up and recovery program. For example, the l

GPU Nuclear Vice President, Nuclear Services who is responsible for the overall SNEC facility decommissioning was the Vice President in charge of TMI-2 during defueling and transition to monitored storage. The Program Director, SNEC facility, who is responsible for decommissioning the SNEC facility in a safe and efEcient manner, was the TMI Site Director responsible for defueling and transition to monitored storage of TMI-2.

The SNEC Facility Site Supervisor, who is responsible for the management and oversight of all production activities, was the Manager of Waste Management for TMI-2 during the clean-up and recovery. The Saxton Radiation Safety OfDeer, responsible for the conduct and oversight of all radiation protection activities, was the Radiological Controls Field ,

Operations Manager for TMI-2 during the transition to monitored storage. Members of the technical support team also have extensive TMI-2, SNEC facility, DOE, and commercial decommissioning and operational experience.

Regarding the occupational and radiological safety requirements for contractors, their personnel assigned at our sites are held to the same high standards of performance as our own personnel. Contractors' programs must comply with GPU Nuclear standards at a minimum.

Finally, the training programs used to train / qualify all radiation workers are accredited by INPO.

23. Page 2-36, Section 2.4 - Please provide more details on the proposed training for staff and contractor personnel involved in the decommissioning activities, including qualifications of the personnel expected to perform the training. Please provide a copy ofyour training program.

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Response: GPU Nuclear is an INPO accreaned nuclear training institution. As such, task I and job analyses are perforrned to develop new or to apply existing training programs. l All personnel involved with training associated with the SNEC project are currently full time instmetors in the GPU Nuclear Training Department. Their qualifications and the programr they teach are quite voluminous and fill a small library. These materials are available for inspection or examples for a more limited scope request could be provided.

As an example, the procedure which governs instructor qualification and a sample lesson plan are provided.

J C301-96-2038 Attachment 1 )

Page 12 of 44 l I

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24. Page 3-6, Section 3.1.2 - Please provide a Copy of the Saxton Site Characterization Plan (6575 PLN-4520.06) with updates, and a copy of the Saxton Characterization Report, Reference 11. I Response: Copies of the documents requested have been included with this submittal. l

25. Page 3-9, Section 3.1.2.2 - You refer to using RESRAD to determine release criteria. Is this to meet proposed future release criteria? If so, please discuss how you will meet current release criteria.

Response: No, the discussion in section 3.1.2.2 refers to contaminated concrete and the  :

pathways analysis used to determine allowable residual activity that would produce less than 15 mrem to the population if allowed to remain in concrete buried at the SNEC facility site atler dismantlement. This was done in order to estimate the amount of concrete that might be generated, not to definitely determine allowable residual. It has always been the intent GPU Nuclear to comply with all regulations / guidelines currently in effect e.g., Reg. Guide 1.86 and/or NUREG-0586 " Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities", NUREG-1500 " Working Draft Regulatory Guide on Release Criteria for Decommissioning", NUREG/CR-5512, Volume 1, Residual Radioactive Contamination from Decommissioning: Technical Basis for Translating Contamination Levels to Annual total Dose Equivalent", NUREG/CR-5849,

" Manual for Conducting Radiological Surveys in Support of License Termination" etc., or any future requirements necessitating pathways analysis.

26. Page 3-9, Section 3.1.2.2 - You discuss possible adjustment of the volume of activated concrete that will be removed. When will this adjusted volume be available?

Response: Off-site sample analysis confirmed on-site concrete radionuclide analysis for significant gamma emitters such as Cs-137, Co-60 and Eu-152. These analysis results are presented in the SNEC Site Characterization Report supplied for your information. A small change in the estimated concrete volume can be obtained by more specifically applying core bore results to affected regions in and around the reactor vessel (~16%

reduction). This is not a significant reduction and comes with no more assurance of being correct than the original estimate which conservativly assumes the complete removal of the 5' interior concrete wall south of the reactor vessel. Therefore, no adjustment will be applied at this time. Additional samples will be taken during the dismantlement phase when major components have been removed allowing access to more of the concrete surface area. At that time, a meaningful adjustment in concrete waste volume can be made.

27. Page 3-10, Section 3.1.2.3. - Explain "10 CFR Part 61 type analysis," in relation to the data presented.

C30196-2038 Attachment 1 i Page 13 of 44 i i

Response: "10 CFR Past 61 type analysis" is a common industry term used to refer to the analysis of samples of radioactive waste streams or sources by various methods which are used to satisfy the requirements of 10 CFR Part 61.55 " Waste classification" Land disposal facilities require licensees to provide information to demonstrate compliance with I this section of the regulation. This analysis identifies typical reactor produced radioactive isotopes as well as so called "hard to detect nuclides" l

28. Page 3-10, Section 3.1.2.4 - Have you verified that the disposal site you intend to use will accept the TRU waste discussed in this section?

Response: This issue is not a concern as long as the SNEC facility ships waste materials within current regulatory requirements for packaging and transportation. Waste shipped to the Barnwell, S.C. burial site and other waste handling facilities will be packaged and shipped to meet appropriate criteria, but in all cases the TRU concentrations will be within

required concentrations when averaged over the volume of material within individual waste packages. l

29. Page 3-11, Table at the top of the page - Because septic tank B contains radioactive material, discuss the potential for concentrations in sludge in the tank, and for seepage into the surrounding soil. Also give data for soil and flora samples in the eilluent pathways.

Response: Please refer to Appendix D, Grid G-7 of the furnished SNEC Site

, Characterization Report (SCR). Data concerning the septic tank is provided there. The

potential of seepage from the tank will be examined when the tank is removed and radiological surveys of the surrounding ground can be made. If contaminated soil is identified, it will be addressed. Data is being compiled on the soil and flora sample results for the effluent pathways and will be provided as soon as it is available.

30. Page 3-11, Section 3.1.2.5 - The activation discussed was apparently calculated. Please compare with relevant confirmation measurements.

4 Response: Please refer again to the SCR which contains measurement information and summarizes conclusions from the activation analysis report prepared by TLG services.

Calculated data has shown to be comparable to measurement data. See SCR, section 4.3 and TLG Activation Analysis Report.

31. Page 3-12, Section 3.1.2.6, Last paragraph - Give explicit EPA criteria, and compare with NRC criteria for the same conditions. Please provide a copy of Reference 12, Saxton Site Remediation Report.

Response: Explicit EPA criteria for hazard material analyses is based on 40 CFR 261.21-24 Subpart C requirements for characterizing material as hazardous. This EPA regulation presents acceptable testing procedures for the four Subpart C characteristics (i.e.

ignitability, corrosivity, toxicity, and reactivity). Results of these tests are found in 4

C301-96-2038 Attaclunent 1 Page 14 of 44 Appendix 3 of Ref.12, "Saxton Soil Remediation Report" which has been included as an attachment. While NRC regulations do not provide prescriptive testing procedures for determining whether a waste is radioactive or chemically hazardous, they do require generators to manage waste to protect worker and public health and safety. Per 10 CFR 20,61, and DOT regulations the soil was characterized to meet shipment manifesting, physical form, radioactive classification for A, B, or C criteria and waste container labeling. Laboratory results of soil radioactivity analyses are found in Appendix 4 of Ref.

12.

32. Page 3-13, Paragraph 1 - Examination of Figure 3.1-11 indicates that the well in which tritium has been periodically detected (GEO-5) is the most easterly of all SNEF monitoring wells. A comparison of Figure 3.1-11 with figure 1, Saxton Site Plan, in the Final Release Survey of Reactor Support Buildings, dated April 1990, indicates this well is approximately 25 feet west-southwest of the site of the former Rad Waste Disposal Facility. Inasmuch as GEO-5 is somewhat closer to this structure than to the Rad Waste Disposal System tunnel, what is the basis for assuming that the tunnel is the source of the tritium? Explain why the existing wells are sufficient to (a) define the extent of the contaminated zone, (b) determine whether substantially higher concentrations of tritium or other radionuclides exist within the contaminated zone, and (c) identify the most probable lource of the tritium and any other radionuclides detected.

Reggqs: The source of the tritium (as detected in the GEO-5 well) is thought to emanate from the Radioactive Waste Disposal System which included the RWDF (Radioactive Waste Disposal Facility) and segments of the Pipe Tunnel which tied into this facility. The DP and the environmental report incorrectly stated only the tunnel. With slight variances in the ground water flow it is difficult to determine with any certainty which location is the source of tritium, although the RWDF provides a higher degree of suspect. However, the tritium concentrations tend to fluctuate, due to seasonal variations, from less than detectable to a maximum of 760 pCi/L. This maximum concentration is stillless than 4%

of the EPA drinking water limit of 20,000 pCi/L and would have no impact on site remediation.

(a.) Per the SNEC Decommissioning Environmental Roport and its mentioned Ref 10 (GEO Engineering Phase I Report of Findings), Figure No. 3, the depth of the GEO-5 well is approximately 15 feet. Table 1 of this report also indicates the groundwater level to be approximately 4-5 feet below grade which supports the design depth of the well. In addtion Figure No. 1 of the GEO report indicates the groundwater flow to be toward the west.

This results in the GEO-5 well to be directly located hydraulically downgradient of the RWDF and representative to sufficiently monitor the contamination zone.

(b.) Given the proximity of the GEO-5 well and sampling results for this location from the past 2 years it is estima*-d that peak concentrations will be

1 l

l C301-96-2038 l Attachment 1 I Page 15 of 44 adequately identified. Attached is a graph indicating the seasonal trends for i tritium from this location. See Attachment 2, reference 2.

(c.) As previously mentioned the most probable source of tritium is from the area where the RWDF was located. With the exception of positve tritium occurring periodically, there has been no detection of positive activity from other nuclides.

33. Page 3-16, Table 3.1 This table contains some operating events that occurred at the SNEF. Did any of these events or any other event that occurred during operation of the SNEF result in contamination outside the SNEF fence? If so, what action was taken and what are your plans to ensure that these areas are considered during decommissioning?

Resnonse: Yes. Unknown events lead to the contamination of small plots of soil sometime prior to the resumption of decommissioning activities in 1986. These areas

< were extensively characterized and remediated as part of the 1994 Saxton Soil Remediation Project. Refer to the report of that project, furnished per your request in Question 31. Neither the aerial surveys nor soil characterization surveys have detected any additional contaminated areas beyond the SNEC facility fence. There is no infonnation which would lead us to conclude that any exist.

.I

34. Page 3-17, Section 3.2 - RADIATION PROTECTION - The potential for inhalation of radioactive materials in particulate form will probably represent the major radiological health hazard for workers who will perfonn the dismantlement and decontamination tasks. Therefore, a comprehensive description of the respiratory protection program should be provided in the Decommissioning Plan that includes the types of respiratory protection available for use, examples of how each type is expected to be utilized, and the guidelines that will be employed to determine the appropriate type of protection to be required. Please address the type of respiratory protection anticipated during dismantlement activities that might be expected to generate airborne activity, such as scabbling, CO2 blasting, or removal of embedded contaminated piping.

Resoonse: Currently and during all future decommissioning phases, the SNEC Facility a

Site Respiratory Protection Program is in effect. The SNEC Facility Site Respiratory Protection Program is encompassed within GPU Nuclear's Respiratory Protection Program which is comprised of procedures based upon the technical guidance of numerous source documents. These sources are:

3 29 CFR 1919.134 (OSHA standard),

- 30 CFR 11 (NIOSH),

ANSI Z88.2,1992 Standards, 10 CFR 20 Sections 1701- 1704, Appendix A and B',

NRC Regulatory Guide 8.15 " Acceptable Programs for Raspiratory Protection " and a

C301-96-2038 Attachment 1 Page 16 of 44 NUREG-0041 " Manual of Respiratory Protection Against Airborne Radioactive Materials -

Over the years, the GPU ~ Nuclear Respiratory Protection Program has been inspected by the NRC and has been found in compliance with all NRC standards. The GPU Nuclear Respiratory Protection Program is fully implemented at the SNEC facility and is available for review.

The types of respirators available for use at the SNEC facility are:

Full face, negative pressure, air purifying respirators, Full face, positive pressure, air purifying respiratirs, Full face, continuous flow, airline respirators, Hood / helmet, continuous flow, airline respirators, Full face, pressure demand, self contained breathing aparatus and Half face, negative pressure, air purifying respirators.

35. Page 3-37, Section 3.2,1 - Please provide a copy of the SNEC Radiation Protection Plan, 6575-PLN-4542.01. Who would make a determination that an alternative procedure is

" equivalent" to this Plan and what criteria would be used to determine that an alternative is " equivalent"? Please provide a copy of the Saxton facility ALARA procedure.

Resoonse: Due to consideration given to maintaining the level of docketed information from the GPU Nuclear facilities consistent, a copy of the SNEC Radiation Protection Plan will not be provided as requested. It is available at the site for review.

The review anu approval process for procedures and plans at the SNEC facility is implemented via procedure 6575-ADM-4500.07 "SNEC Procedure Development, Change Requests and Safety Reviews" The term equivalent as used in this section refers to a plan or document which would supersede the current plan. The same levels of review and approval that implemented the current plan would be required to approve any substantive changes and would also be required for any document or procedure which

supersedes this plan. The SNEC Facility ALARA procedure is being developed and will be available at the site upon completion ofits review and approval. The procedure is scheduled for September 1996 implementation.

36. Page 3-37, Section 3.2.1, Radiological Controls Program - With regard to the need for assessing internal radiation exposure for those workers that are to be assigned to any activities on the dismantlement and decontamination of the Saxton facility that may present a radioactivity inhalation or ingestion hazard, please discuss your plans for conducting wholebody counts prior to, during and following these activities.

i C301-96-2038 a

Attachment 1 l Page 17 of 44 l ,

l Response: The requirements for conducting bioassay and assessing internal radiation i exposure are established in SNEC procedure 6575-ADM-4500.31. The procedure

{ requires bioassay:

[ -

baseline or prior to initial RWP entries,

} -

annually or after the baseline,-

termination of employment, j -

prior to going to another licensed facility other than one operated by GPU j j Nuclear, '

{

following the assignment of 10 mrem internal whole body dose (CEDE) or j- greater in any one day, following the assignment of 50 mrem internal whole body dose (CEDE) or j greater in any seven day period, following an unexpected skin contamination on the neck and face or positive

. nasal smears,

{: -

anytime the inhalation of radioactive material is suspected and i -

supplemental as directed by the RSO.

e l l The normal bioassay method is by whole body counting (WBC) which is performed at the TMI site. Based on the internal radioactivity exposures involved and their decay emitters, I other bioassay methods may be utilized. All bioassay methods and internal assessments 4

will be performed in accordance with existing TMI procedures.

I j 37. Page 3-44, Section 3.2.1 - Please provide a copy of the Off-site Dose Calculation Manual.

Response

The Off-site Dose Calculation Manual is in draft and awaiting specifics on the i design of the CV ventilation system. A copy will be provided when the approved document'is issued.

l

! 38. Page 3-47, Section 3.3.2.1 - Please provide more information on design of the temporary l filtration systems referenced in this section, and on the HEPA filtration system to be j provided for the Decommissioning Support Building. Also, discuss methods to be employed for testing all HEPA filters. Please provide diagrams and descriptions of operation of the complete decommissioning ventilation system including radiation

monitors, alarm or automatic action setpoints and automatic or manual actions.

. Besponse: Details are not yet available concerning the CV ventilation system or the l associated monitors. Design work began on July 16,1996. The information on the j

! decommissioning ventilation system as delineated in the question will be furnished as soon l as it is available.

i i A separate ventilation system for the Decommissioning Support Building (DSB) is not j planned at this time but will be supported from the CV system. Temporary, portable j HEPA ventilation units may be employed in the DSB as needed. These are industry standard units.

4

l C301-96-2038 Attachment 1 ;

Page 18 of 44

39. Page 3-47, Section 3.3.2.1 - This section discusses the use of continuous air monitors (CAM)if necessary. How will the decision be made to use a CAM instead of a portable air sampler? What actions will be taken if a CAM alarms? Will the CAMS be able to turn off the ventilation system if conditions warrant?

Response: Local continuous air monitors (CAMS) are routinely operated in the vacinity of selected radiological work activities as directed by radiological controls department personnel. CAM placement will be decided based upon the work scope and the specific ,

radiological conditions at the work site. All radiation workers are trained on the proper response to a CAM alarm. If a CAM alarms, workers are directed to immediately stop work, place the job in a safe condition, leave the area and notify Radiological Controls, In addition to local area monitors, a radiological airborne monitor will be installed in the l exhaust train of the CV ventilation unit. This effluent monitor will be interlocked with the  !

CV ventilation system to secure it should conditions warrant.

l l

40. Page 3-49, Section 3.3.2.3, Last paragraph - How and when will changes to the 1 Decommissioning Plan be submitted?

Response: Please refer to section 1.1.3 of the Decommissioning Plan. Changes will be submitted to the NRC as the need arises based on site conditions. Internal copies of the Plan are on " controlled distribution" and changes are accomplished in accordance with written procedures governing review, approval and distribution. l

41. Page 3-50, Section 3.3.2.3, (1), (2), and (3)- Will external shielding be required to allow these large components to be shipped intact? Discuss protection of the public from these '

radiation sources.

Response: GPU Nuclear intends to meet the normal conditions of transport as defined in 10 CFR 71 and 49 CFR 173. The maximum doses from these packages are no different than for any other radioactive waste package which meets the normal conditions of transport. In order to meet the normal conditions of transport, external shielding will likely be required on the reactor vessel. A small area of the steam generator may require shielding depending upon the type of transport used (open or closed). The pressurizer will not require any shielding to meet the normal conditions of transport.

42. Page 3-50, Section 3.3.2.3 (1) and (2) - Please specify how the vessels will be sealed prior to shipment. Also, please describe the purpose and characteristics of the proposed concrete / grout.

Response: Specifications have not been developed at this time for details such as sealing the vessels to meet the required packaging specification e.g. " strong tight container", type B package etc. These specifications will be fumished or made available for review when complete.

C301-96-2038 Attachment 1 Page 19 of 44 The use of grout or similar fill material may be needed to comply with disposal l requirements at the burial facility. Please refer to the response to question 44. .

t

43. Page 3-50 to 53, Section 3.3.2.3 - Offsite vendor, offsite processor, offsite volume reduction facility etc. are mentioned seve;al times. Please explain in more detail who and where these offsite facilities may be, something about their qualifications and experience, including their authorization to possess by-product radioactive material, and the planned subsequent disposition of the materials and components.

Response: Decisions concerning the ofTsite vendors who will suppon the f decommissioning activities at the SNEC facility will be determined later in the project. j Vendors will be considered based on their experience and capability to successfully {

complete tasks similar to those to be performed at the SNEC facility site. Vendors will be chosen in accordance with the requirements of Section 7.3.9 of the DP and GPU Nuclear  ;

procedures. I 1

44. Page 3-51, Section 3.3.2.3 (3)- What criteria will be used to decide on the need for filling the pressurizer with grout?

Response: As described in section 3.3.2.3 (3) of the DP, volume reduction options may be available for use in disposing of the pressurizer. If so, grout would not be used. If direct burial is chosen at the Barnwell Facility, their license requires that at least 85% of the package void space be filled. In this case it is likely that grout inject!on would be used.

Credit is not taken for use of the grout in any other way.

45 Page 3-51, Section 3.3.2.3 (4)- This section indicates that some components may be  !

decontaminated on-site. What criteria will determine if a component is decontaminated  !

on-site or sent off-site? What methods of decontamination will be employed on-site and where will these activities take place?

Response: Whether a component would be decontaminated on or off site will be driven j principally by decisions based on economics and practicality. It will not be practical to j perform extensive decontamination of components at the site due to the limited j infrastmeture. Therefore, the capability will be limited. Generally, component j contamination levels and component arrangement and make-up are such that  ;

decontamination to free release levels is not practical and will not be pursued. The extent of decontamination intended is that consisting of minor decontamination of tools, equipment and hardware used in support of decommissioning and not intended for disposal. Such decontamination would use routine methods commonly employed at nuclear facilities such as water / soap, CO2 blasting, hand wiping; etc. When practical such activities will take place in the CV. In some cases a temporary vendor supplied facility such as those commonly used during outages at nuclear power plants may be used.

These facilities are usually a self contained decontamination trailer type facility which employ the above techniques.

C301-96-2038 Attachment 1 Page 20 of 44

46. Page 3-54, Section 3.3.2.5 - You discuss using information from the characterization report to determine the field processes that will be used to remove radioactive material.

When will these determinations be made and how will NRC be informed of these decisions?

Response: The removal of contaminated materials and components is an iterative process which involves the interpratation of data and the application of that information to field decisions on a continuous basis. In some cases decisions regarding removal processes or techniques will be made in a generic sense in that the most efficient method or the only practical method for removal will be employed. The radioactivity levels reported in the characterization report will be used to guide the selection of these processes in order to minimize exposure, ensure safe work practices and to be eflicient. )

It is not the intent of GPU Nuclear to inform the NRC of these routine, daily decisions l whether to employ for instance a power band saw or a reciprocating saw to cut a pipe. I These are the types of decisions we are refenng to.

I

47. Page 3-55, Section 3.3.2.6 - Here it is stated that large components will be moved and prepared for shipment as soon as practicable, but on pages 3-51 and 3-52 there are l comments that suggest these large components could be used as shipping containers for smaller components and debris. Please explain in more detail what is planned.

l Response: The words "as soon as practicable" are used to indicate as soon as they were ready to be shipped but not before so. This is not meant to indicate that no other work l has taken place which may have generated radioactive waste for shipment and disposal, l indeed, there is considerable waste ready to be packaged and shipped now. If possible it makes good economic sense and is a wise use of burial space to consolidate such waste as described when possible. This is a common industry practice and would be employed on this project whenever practical.

48. Page 3-61, Third paragraph - Please provide a description of the vendor supplied stations noted here including their capabilities, methods of operation, decontamination solutions expected to be used, the need for contaminated solution storage tanks, etc.

Response: Details on these temporary facilities are not yet available, however, responses to several other questions concerning decontamination capabilities will give some idea of the type of stations intended. Such facilities are used routinely at other sites and are frequently used at GPU Nuclear facilities.

49. Page 3-63, Section 3.3.4 - Please provide additional detail on the new opening that will be made in the CV for access from the decommissioning support building. How large is the opening? Will there be a way to seal this opening when not in use?

C30196-2038 Attachment 1 Page 21 of 44 i

Response: This opening will be slightly smaller than the Material Handling Bay (MHB) which encloses the intended opening. The MHB is 16 feet wide. The MHB will be 4

attached to the CV and will have a door (s) to isolate the MHB and hence provide closure on the CV. It is planned to shut and lock these doors when access is not required.

1 l 50. Page 3-65, Table 3.3 For the waste classification of components listed on this table to be determined at a later time, when will this determination be made? What is your estimate of these components at this time?

Response: The re-evaluation will take place between now and the end of 1996. The items you refer to are the: 1) Shut Down Cooling Heat Exchanger,2) Regenerative Heat i Exchanger, 3) Boric Acid Demineralizer Vessel,4) Storage Well Demineralizer Vessel and the 5) Purification Demineralizer Vessel.

l The TBD listing (Table 3.3-2) for these items has to do with waste classification. When using the available waste stream information from current sample data, these items initially appear to be greater than Class C waste materials. This was determined by comparing exposure rate measurements with calculated mR/h/Ci values provided by a radiation shielding computer code model. In all cases, background radiation levels were not removed from suney data prior to comparing to the computer model output. This elevates the estimated curie loading for several of the items listed above. When background is stripped from the measurement results, several of the above items will be Class C waste. We believe that only item 3 and possibly item 2 from the above listing will i remain as greater than Class C waste after background subtraction. I Item 2 has a hot spot located in one area that is significantly higher than the rest of the unit, forcing the average contact exposure rate higher and resulting in a greater than Class C waste classification. Specific suneys and compartmentalized modeling will produce a more realistic result and lower the curie loading for this unit to allow a Class C assignment.

Item 3 will be initially treated the same way as item 2. However, a more extensive approach may be required, such as extracting a sample of the internals from this unit or from some attached piping.. In order to accomplish this more aggressive component interrogation / treatment, better access and localized shielding may need to be provided.

The details of the re-evaluation of this component have not been decided, but should be available within the time frame listed above in paragraph one.

Other TBD listings on this table refer to estimates of the curies present on hardware located throughout the SNEC containment vessel (Surface Contaminated Objects - SCO).

Some of these total curie TBD listings have been addressed in the SCR. Those TBD curie listings not provide in the SCR are for items located in the containment vessel. The total curie estimate for all these items is surface area dependent and although contamination levels per unit area are known, surface areas of several of these items have not been calculated. In all cases these items represent much less than 1% of the curie loading for the entire facility. These items will be assessed prior to packaging for shipment when they

C30196-2038 Attachment 1 Page 22 of 44 can be separated form resident background radia' ion levels and proper physical measurements can be taken. This is the best approach considering ALARA personnel dose mandates.

51. Page 3-68, Section 3.4, ACCIDENT ANALYSES - Was the accidental release of the contents of a radioactive liquid waste storage tank analyzed? What is the maximum quantity and concentration of contaminated liquid expected to exist on site during decommissioning activities? Please provide a bounding accident analysis for the failure of a liquid waste vessel.

Resoonse: For a bounding estimate of the contents of a radioactive liquid waste storage tank, it is assumed that the floors and walls of the spent fuel pool (Area 6) were completely decontaminated using 500 gallons of water. The surface areas and mean smearable activity for these areas were take from the SNEC Site Characterization Report, Section 4.1.5 and Table 4-44 respectively. A smear efliciency of 10% was also assumed so the activity available for removal by decontamination was 10 times the smearable activity found. The total calculated activity is shown below:

AVERAGE SURFACE SURFACE TOTAL AREA CONTAM ACTIVITY 2

(ft') (uCi/cm ) (uCi)

FLOOR 217 270000 24519 WALLS 4025 47000 79166 TOTAL 103685 Assuming this activity is contained in 500 gallons of water, the resulting nuclide concentrations using the Area 6 distribution are as follows:

C301-96-2038 Attachment 1 Page 23 of 44 CONCENTRATION ISOTOPE (uCi/cc)

H-3 0.00E+ 00 FE-5 5 8.3 5 E-05 CO-60 2.3? E-02 NI-59 0.00E+ 00 NI-63 4.33 E-0 4 SR-90 7.64 E-0 5 N B-94 0.00E+ 00 A G-108 M 0.00 E+ 00 SB-125 0.00E+ 00 CS-134 0.00E+ 00 CS 137 3.01 E-02 EU 152 0.00E+ 00 EU 154 0.00E+ 00 EU-155 0.00E+ 00 PU 238 1.06 E-0 5 PU-239 2.76 E-05 AM 241 4.3 5 E-0 5 PU 241 9.3 5 E-0 5 CM 242 3.53 E-07 CM 243 9.99 E-07 TOTA LS 5.48 E-02 The tank is assumed to develop a leak and all of the liquid is released. It is assumed that SE-5 of the activity in the tank goes airborne. This is a highly conservative assumption, as DOE-HDBK-3010-94 lists this as the bounding release fraction for a tank pressurized up to 50 psig. A tank used to store this type ofliquid would be at atmospheric pressure so the release fraction should be substantially less than this value.

An atmospheric dispersion factor (X/Q) of 4.14X10-3 sec/m' is used to calculate the airbome activity concentration at the site bounday (200 meters). This conservative value is calculated for a 1 m/s wind speed and a G stability category. Off-site doses are calculated using the parameters and methodology ofEPA 400. The whole body dose to an individual standing at the site boundary for the duration of the release is calculated to be less than 5X10 mrem.

The results are summarized below:

C301-96-2038 l Attachment 1 Page 24 of 44 ACODENT DESCRPTICN RUPTURE OF UQUID PADWASTE STORAGE TANK ACTIVITY AVAllAE1.EFCRRELEASE 0.104 0 VENTILATION FILTF% TION EFFICENCY 0 DURATlON CF RELEASE 2 FES ACDVITY ACTIVITY RELEASE DCF TOTAL PRESENT REl. EASED FMTE (REMAE/ DOSE ISOTOPE (uC) (ud) (uGAec) uCAc) (REM)

H-3 0.00E+ 00 0.00E+ 00 0.00E+ 00 7.70E+ 01 0.00E+ 00 F555 1.58E+ 02 7.90503 1.10E4)6 3.20E+ 03 2.91511 COL 60 4.53E+ 04 2.26E+ 00 3.14E44 3.20E+ 05 8.33509 NI-59 0.00E+ 00 0.00E+ 00 0.00E+ 00 2.60E+ 05 0.00E+ 00 NI-63 8.20E+ 02 4.10502 5.69506 7.50E+ 03 3.54510 SR90 1.45E+ 02 7.23503 1.00E46 1.60E+ 06 1.33508 NB94 0.00E+ 00 0.00E+ 00 0.00E+ 00 5.00E+ 05 0.00E+ 00 AG-108M 0.00E+ 00 0.00E+ 00 0.00E+ 00 3.40E+ 05 0.00E+ 00 SS125 0.00E+ 00 0.00E+ 00 0.00E+ 00 1.90E+ 04 0.00E+ 00 CS 134 0.00E+ 00 0.00E+ 00 0.00E+ 00 5.60E+ 04 0.00E+ 00 CS 137 5.70E+ 04 2.85E+ 00 3.96504 3.80E+ 04 1.24507 E1J 152 0.00E+ 00 0.00E+ 00 0.00E+ 00 2.70E+ 05 0.00E+ 00 E11154 0.00E+ 00 0.00E+ 00 0.00E+ 00 3.40E+ 05 0.00E+ 00 EU-155 0.00E+ 00 0.00E+ 00 0.00E+ 00 5.00E+ 04 0.00E+ 00 PU-238 2.00E+ 01 1.00503 1.39E47 4.70E+ 08 5.41 E47 FU.239 5.23E+ 01 2.61503 3.63507 5.20E+ 08 1.56E46 AM-241 8.23E+ 01 4.12503 5.72E4)7 5.30E+08 2.51506 PU-241 1.77E+ 02 8.84E43 1.23E46 0.90E+06 1.01507 CM-242 6.67501 3.34E435 4.63E49 2.10E+ 07 8.06510 CM-243 1.89E+ 00 9.46E-05 1.31508 3.70E+ 08 4.02508 TOTALS 1.04E+ 05 5.18E+ 00 7.20504 4.90E46 No liquid pathway evaluation was made, since the low volumes ofliquid radwaste and their distance from the river would preclude direct entry into the river. Any entry into the river would be through the groundwater system. Any dose from this pathway would be insignificant since virtually all of the activity in the water would be bound up in the soil, and the release rate to the river via groundwater would be very slow.

52. Page 3-68, Section 3.4, ACCIDENT ANALYSES - Please provide the bases or information sources for the radionuclide mixtures assumed for the source terms for each of the accidents summarized in this section.

Response: The bases for the nuclide mixtures in the accident analysis source terms are samples collected and published in the Saxton Characterization Report, Reference 11 of the Decommissioning Plan. The specific Saxton characterization data used for each analysis are as follows:

Material Handling Accident - Dropped Demin Vessel SX866950118

a C301-96-2038 Attachment 1

. Page 25 of 44 :

1 '

4 1

Material Handling Accident - Dropped Steam Generator SX835950071

!- SX825950111

, SX865950053 j Fire - Combustible Waste in the Yard AREA 6 Mix

Vacuum Filter Bag Rupture AREA 6 Mix l 1

Pipe Segmentation w/o Eng Controls SX865950053

{ Oxyacetylene Explosion SX865950053  ;

i

' ~ LPG Explosion AREA 6 Mix i SX861950135

, 53. Page 3-68, Section 3.4, ACCIDENT ANALYSES - Please analyze another accident l

{ scenario in which it is assumed that the reactor vessel is dropped from the maximum lift l

height outside the CV during the removal sequence, or describe your rationale for not j 4 considering this accident.

t

Response

As stated on page 3-69 of the decommissioning plan, dropping of the reactor

{ pressure vessel was not analyzed since it would be highly unlikely to mpture during a materials j handling accident due to the nature ofits construction. In addition, the surface activity in the l vessel, as stated in the Decommissioning Plan, is 11.8 Ci compared to 17 Ci in the

, demineralized vessel that has been analyzed for the materials handling accident. The J

demineralizer vessel contains more loose activity than the pressure vessel. The radionuclide ]

l distribution assumed for the demineralizer vessel drop is also richer in transuranics (the major j dose contributor) than the distribution in the pressure vessel. All other assumptions used in

{ the materials handling accident would be the same. As a result, the greater activity available

! for release from the demineralizer vessel clearly shows that dropping this vessel provides the j' most bounding dose estimate for postulated materials handling accident scenarios.

i l 54. Page 3-68, Section 3.4, ACCIDENT ANALYSES - Please address the potential health j impact for on-site personnel of each of the accidents analyzed in this section.

3 l Besponse: The estimated maximum occupational Committed Effective Dose Equivalent l (CEDE) and Committed Dose Equivalent (CDE) to the bone surface for each accident

scenario is summarized below. All doses are in Rem.

t i a I 1

I i 1 r I

l C301-96 2038 Attaciunent 1 l Page 26 of 44 4 l

Accident Type CEDE CDE Drgpped Demineralizer Vessel . ,,,_0.,4_, ,,_,7. 2 ,,_ ,

Fire in Combustible Waste 0.08 1.2_ l Oxyacetylene Explosion 0.7 0.04..

Pipe Segmentation _. 0 02_ .. 0 3.. . j i yacuum Fiter Rupture 0.004 _

. ,_0.07_ l 1 LPG Explosion 0.004 0.07 1

j The calculations were performed using conservative assumptions and dose conversion factors from the following references:

Saxton Nuclear Experimental Corporation Decommissioning Plan i

i NUREG/CR-0130, " Technology, Safety, and Costs of Decommissioning a Reference Pressurized Water Reactor Power Station", Volumes 1 and 2. l 1

Federal Guidance Report No. I1, US EPA 1988

, The calculations are documented in GPU Nuclear Calculation RAF 6612-96-014.

1 i 55. Page 3-68, Section 3.4.1, Introduction - Was the second sentence intended to imply that j the calculated whole body doses presented for each accident described in Section 3.4 are total effective dose equivalent (TEDE) values? Please clarify. Also, please provide a

, description of the inhalation dose calculations for each accident scenario analyzed.

Response

The calculated whole body doses presented in each accident analysis re total

! effective dose equivalent (TEDE) values using the inhalation dose conversion vaiues prosided i

by EPA 400. To determine the inhalation dose, the activity is assumed to be released over a two hour period to determine a release rate (uCi/sec). The release rate is then multiplied by

the atmospheric dispersion coefficient described in on page 3-70 of the Decommissioning plan 3

(sec/m ) to determine the concentration at the site boundary. The EPA 400 dose conversion 3

factors (mrem /hr/uCi/m ) are then used to calculate the offsite dose at the site boundary for the two hour release. In actuality, the offsite dose delivered is independent of the duration of the

! release since the total activity released is function of the accident rather than its duration.

I Releasing the activity over a one hout period would reduce the exposure time by one half but would double the dose rate. The total dose would be the same for any release duration.

l 56. Page 3-70, Section 3.4.1.1 - With respect to plans for filling the resin vessel with grout

[ !

prior to lifling, will the grout and resin be mixed or is the purpose of the grout just to fill the void?

Response: As descibed in the response to Question 42, the primary purpose of the grout is to fill the void space to comply with disposal criteria. This would also result in

stabilization of the waste form (note that no credit is taken for this benefit in the accident analysis).

i

C301-96-2038 Anachment 1 Page 27 of 44

57. Page 3-71, Section 3.4.1.2 - The release fraction of 1.5 X 10 4 used in the accident scenario for a combustible waste fire was based on the combustion of a relatively small volume of waste (one cubic meter) contaminated with uranium. It is expected that the larger volume of waste assumed in the accident scenario would generate higher temperatures, resulting in a higher release fraction. Also, the use of uranium to simulate the behavior of most of the radionuclides assumed in this scenario is questionable.

Comparisons with the methodology described in DOE-HDBK-3010-94 also suggests that the release fractions calculated by you are lower than might be expected. Please provide a more representative basis for the release fraction assumed for this accident scenario. Also, please provide the bases for the radionuclide inventory assumed for this analysis and explain why it should be considered representative of the wastes described.

Response: The release fraction specified in the analysis ofcombustible waste fire in Section 3.4.1.2 of the Decommissioning Plan was chosen because it was the release fraction specified in NUREG /CR-0130 for this of event. The use ofuranium to simulate the behavior of nuclides in this mix is appropriate since more than 96% of the offsite dose from the event is the result of transuranics released from the waste. It is not clear that a fire involving an amount of 3

material greater than 1 m would generate significantly higher amounts of heat since the density and composition of the fuel is the same. The experimental data in DOE-HDBK-3010-94 show release fractions ranging from 3E-5 to SE-4 with median and average values of 8E-5 and IE-4 respectively. In fact, in only one data point was significantly higher than the value used in the Saxton analysis. Based on the information provided in NUREG /CR-0130 and DOE-HDBK-3010-94, a release fraction of 1.5E-4 appears to be reasonable. Even using the bounding value of SE-4 identified in the DOE document would only increase the offsite dose in the analysis to 0.85 mrem. This value is still well below the bounding analysis dose of 1.5 mrem that is specified in Section 3.4.1 of the Decommissioning Plan.

The basis for the nuclide inventory is the isotopic distribution given for Area 6 in the SNEC Site Characterization Report. This is the most highly contaminated area of the plant and it is believed that the surface contamination in this area provides a good approximation of the distribution that will be present in combustible waste. As discussed in Section 3.4.1.2 of the Decommissioning Plan, the activity in the Sealand Van is assumed to be 99.8% of the Type A LSA limit for this type of container and the assumed mix.

58. Please analyze a fire in the CV or decommissioning support building. Please describe the fire protection plan to be in place for decommissioning.

Response: The type of materials that would be involved in a fire would be the same whether the fire occurred in the yard area or in the CV or decommissioning support building (DSB).

The difference is that a fire in the support building or CV would have the advantage of HEPA filtration prior to release to the environment. As a result,2000 times more activity would need to be involved in the fire in the CV or DSB to produce the same dose as the fire in the yard that has been analyzed in the Decommissioning Plan. Since the activity in the yard fire was

C301-96-2038 Attachment 1 Page 28 of 44 assumed to be 1.79 Ci, this would equate to over 3,500 Ci a fire in one of thes Such a fire is not a credible event because of the level of contamin material, limited availability to fire because of storage and because the fire is existing analysis in Section 3.4.1.2 of the Decommissioning Plan.

Conditions within the CV are such that there is almost no combustible materia There is a separete power supply for both the crane and for lighting and outlets. B circuits can be deenergized from outside the CV.

During an inspection of the CV by a GPU Nuclear Fire Protection Engineer and members of the local fire company, agreed that the fire risk was insignificant due to the low combustibility loading and any fire would be contained by the CV without a challenge. If the CV fire loadin conditions change radically, the effect on the risk will be revisited.

The DSB will be provided with a fire detection system and conditions within maint such that combustible matericls will be stored in drums or LSA boxes essentially unavailable to fuel a fire.

An increased fire potential may exist should there be the need to expose combustible materials for the period required to repackage t ,

however, the volume will be maintained small enough so as not to challen of the available portable fire extinguishment equipmment.

The Fire Protection Program is contained within the Emergency Response Proced the SNEC facility 6575-ADM-4500.06.

The procedure is being revised and updated to address the increased number of temporary support buildings and work activities necessary to decommission the facility. Section 4.6 of the procedure addresses the response to smoke or fire by identifying required notifications and Exhibit 1,1.0 provide a description of agreements, training and material support provi;ed the local fire and ambulance company to deal with a fire emergency at the SNEC racility.

59.

Page 3-73, Sect;on 3.4.1.4, Surface contamination - (a) Justify the method used to estimate the surface contamination for segmenting a pipe; (b) explain the f the equation for total activity generated near the bottom of the page; (c) explain th of the 208 microcuiles per gram; (d) show which radionuclides contribute the of the dose; and (e) this dose is approximately the same as that computed for the

" maximum credible event." Please explain the stated conclusion that the event of 3.4.1.1 bounds all other potential events.

?

Response

SX865950053. The surface contamination on the segmented pipe was based on This sample was chosen because it had the highest transuranic content of th piping analyzed during characterization. This particular sample was analy surface contamination in an acid bath, then analyzing the contamination remove it provides a good representation of the surface contamination on this type ofp

. The results of the analysis showed the specific activity in the materid ramoved was convert this contamination to a unit of surface area, it was conservatively assumed that surface activity in the pipe was distributed in the first i

1/16"(0.16 cm) of the pipe surface. This i

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

l C301-96-2038 Attachment 1 Page 29 of 44 is considered a conservative assumption since it is likely that the fdm ofcontamination is much thinner than 0.16 cm. Using this assumption ofcontamination depth, the mass of a 1 cm2 area >

of pipe 0.16 cm thick was determined to be 1.27 g. The measured activity from 1 g of the surface material was 208 uCi, so the activity in 1.27 g was calculated to be 264 uCi. As a result, the surface contamination on the pipe was calculated to be 264 uCi/cm2. As can be seen from this calculation, the assumption of the contamination in the first 0.16 cm in the pipe is conservative compared to a more probable thinner contamination distribution because the thicker assumption prc'.uces more weight (hence more activity) per unit area of the pipe.

The total activity generated is given in the Decommissioning Plan by the following formula:

1 Total Activity Generated = (Surface Contamination Level)(Kerf Width)(Pi X Length of Pipe ID)

j. Having found the contamination per unit surface area of the pipe, the activity released during l the segmentation is the product of the contamination per unit surface area of the pipe and the j

surface area involved in the pipe cut. The kerf width (width of the cut) and the circumference  !

l' of the pipe (PI x pipe diameter) provides this area. Activity released from pipe that is i vaporized below the surface contamination is ignored as it is negligible compared to the  ;

amount released from the smface contamination. The SNEC Facility SiteCharacterization l Report demonstrates that activation of this piping is not significant.

1 The dose by isotope for this event and the maximum credible event are given below- i l

DROPPED PIPE CUT DEMIN i DOSE DOSE : 1

ISOTOPE (FEM) (REM)

H-3 0.00E+ 00 0.00E+ 00 FE-55 2.90509 1.42509 COL 60 1.97E48 5.76509 NI-59 5.10508 2.77508

! NI-63 1.18507 7.45508 l SR90 1.63507 9.63507 l N594 1.42E-09 0.00E+ 00 l AG 108M 0.00E+ 00 0.00E+ 00 SS125 0.00E+ 00 0.00E+ 00 CS 134 0.00E+ 00 0.00 E+ 00 CS-137 5.43E49 1.20E-07 EU-152 0.00E+ 00 0.00E+ 00 EU-154 3.89509 2.31508 l EU-155 1.46510 8.79510 '

PU 238 1.84504 1.69504 l PU-239 4.47504 5.38504 l

AM-241 7.07504 8.09504 PU-241 9.75505 1.44 E-04 CM-242 0.00E+ 00 0.00E+ 00 'I CM-243 0.00E+ 00 0.00E+ 00 i

TOTALS 1.44503 1.46503 l

In both analyses, over 99% of the dose is delivered by transuranic isotopes It is true that both scenanos produce essentially the same offsite dose. However, the dropped demin vessel

_ _ , - _ _ _ , _ _ . u . . _ - - - .

l i

l C301-96-2038 Attachment 1 Page 30 of 44 produces a slightly lager dose, so it was picked as the maximum event. In either case, both accidents are bounded by the less than 1.5 mrcm value stated in the introduction to Section 3.4 of the Decommissioning Plan.

60. Page 3-77, Section 3.4.1.6 - You discuss filter banks being ruptured during this accident.

It appears that the number of filters used comes from reference 14. How many filters will be in use at Saxton and what is the results of the analysis using Saxton information?

Response: The number of filter banks ruptured in this analysis was taken from Reference 14 of the Decommissioning Plan because it provides a highly conservative dose estimate for the accident scenario. While the exact number of filter banks that will be installed in the decommissioning ventilation system is not known at this time, it will certainly be far less than

50. Since the activity released to the emironment is the result of the loaded filters mpturing, the use of 50 filter banks in the analysis provides a bounding estimate for this accident scenario.

61. Page 3-78, Section 3.4.1.8 - Do you plan to install any emergency lighting in the CV to allow for exit in the case of power loss?

Resoonse: Yes, emergency lighting will be installed.

62. Page 3-80, Section 3.4.1.9 B. - As decommissioning activities progress, what is the CV response to flooding of the SNEC site as related to structural response and buoyancy?

l Response: Questions relative to the stmetural response and buoyancy concerns with the l

Saxton Nuclear facility CV were first addressed in the "Saxton decommissioning Plan and Safety Analysis Report" dated April,1972 and previously supplied to the NRC. That l report demonstrates that the stmetural integrity cnd negative buoyancy of the CV in its present configuration is assured up to a projected flood level of 826.7 feet above mean sea level (amsl). This flood level occurs at the Saxton facility site with a projected frequency )

of approximately once every 3,500 yers. The maximum historical observed flood level of record at the site was 809.5 feet amsl. The assurance of negative buoyancy under these conditions does not include the effects of soil adhesion in preventing upheaval and hence is very conservative.

According to Army Corps of Engineers (ACOE) and U. S. Geologic Survey (USGS) data, the 100 year flood recurrence level at the Saxton facility site is below the predominant site grade of 812 feet amsl. Therefore under both normal and realistic flood conditions, the structural integrity and negative buoyancy of the CV is assured.

During the facility decommissioning process, components and structural materials will be l removed from the CV. In its present condition, the CV weight including all components l and structural materials is 3,249 tons. At a postulated 100 year flood recurrence level of i 811 feet amsl, the buoyant force acting on the CV is 2,583 tons. This leaves a margin of 666 tons of negative buoyancy at the 100 year flood r "currence level.

C30196-2038 Attachment 1 Page 31 of 44 l

Applying a safety factor of 1.1 as prescribed by NUREG 0800, " Standard Review Plan for Review of Safety Analysis Reports for Nuclear Power Plants", results in a weight of 2,841 tons needed to preclude floatation (1.1 x 2,583 = 2,841). This results in a negative i buoyant force of 408 tons at the 100 year flood recurrence level (3,249 - 2,841 = 408 tons).

l The total weight of all equipment associated with the CV is approximately 175 tons, l therefore all equipment may be removed from the CV while still maintaining negative buoyancy at the 100 year flood recurrence level. This leaves an additional mass of 233 tons of material which could be removed from the CV at the 100 year flood recurrence l level without risking floatation (408 tons of reserve buoyansy - 175 tons of equipment =

233 tons). I If additional material beyond the 408 tons is required to be removed, either replacement weight will be added or additional analysis will be performed to ' refine the estimate.

1 I

Structural integrity of the CV will not be compromised so as to risk failure until the restoration phase, at which time unconditional free release and license termination will .

I have taken place.

63. Page 4-1, Section 4.1.1 - With respect to the use of the RESRAD code in general, please describe the assumptions employed la the various calculations, the pathways assumed, the limiting pathways, and projected annual dose rates and/or body burdens ofinternal radionuclides. Where experimental data are required, discuss the source of those data, and the range ofvalidity and applicability to the Saxton facility and site.

Describe in detail your analyses that ensure that the proposed release criteria are met, including projections for the next 1000 years, as applicable, following unrestricted release.

Response: In developing radionuclide concentration guidelines in soil using RESRAD, all significant pathways for the critical population have been and will be considered. These pathways are:

1. Direct exposure from the contaminated soil;

2. Internal dose from inhalation of airborne nuclides, excluding radon;

3. Internal dose from ingestion of:

- plant foods grown in contaminated soil and irrigated with contaminated water,

- meat and milk from livestock fed with contaminated fodder and water,

- drinking water from a contaminated well or pond,

- fish from a contaminated pond, and

- contaminated soil.

I i

1 C301-96-2038 l Attachment i Page 32 of 44 l Various assumptions have and will be used for SNEC compliance modeling. The most significant data inputs as determined by past sampling assumes:

1. Cs-137 is the predominant nuclide

2. Site affected area (contaminated zone) is approximately 1.1 acres,

3. Thickness of contaminated zone (Soil depth of 6 inches),

i 4. Include all pathways (residential scenario),

5,1000 yr exposure period

, 6. Nuclide concentrations scaled to meet NRC dose acceptance criteria.

I

7. Valid RESRAD default parameters - See attached list used in example calculation.

The limiting pathway is direct exposure from Cs-137 remaining in the soil or buried debris.

Soil concentrations for Cs-137 will be reduced to meet NRC cleanup criteria. Other nuclide concentrations will be limited based on scaling factors with Cs-137 and the respective doses added to meet release criteria.

64. Page 4-2, Section 4.1.1 - Please clarify the last sentence of this section.

l Resnonse: At present, the proposed release criteria as developed by the " Enhanced Participatory Rule Making Process" has a more restrictive release criteria for drinking water than the proposed 15 mrem /yr from all other pathways. GPU Nuclear is committing to meet that more restrictive U. S. EPA standard of 4 mrem /yr.

65. Page 4-2, Section 4.1.2 - The release criteria for surface contamination meet applicable NRC guidance. However, please address specific release criteria for materials and components containing neutron induced radioactivities, and the bulk free flowing materials already noted.

Response: Specific release criteria for neutron induced materials has not been promulgated by the NRC and is still under review based on the new release criteria to be implemented in the revision to 10 CFR 20. The NRC has not yet finalized guidance on acceptable methods of demonstrating compliance with the revised criteria. Three NUREGs are under development which will provide such guidance: NUREG-1505 "A Nonparametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys", NUREG-1506 " Measurement Method for Radiologica! Surveys in Support of New Decommissioning Criteria", and NUREG-1507 " Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions" It is the intent of GPU Nuclear to comply with the guidance promulgated when it is available.

Release criteria for " bulk free flowing material" has been promulgated by the NRC in IE i Bulletin No. 80-10, IE Circular No. 81-07, and Information Notice No. 85-92. GPU Nuclear is committed to meet this criteria.

b l

I C301-96-2038

! Attachment 1 Page 33 of 44

l. 66. Page 4-3, Section 4.2, FINAL SURVEY METHODOLOGY - Natural and enhanced j backgmund radiation is mentioned. Exactly what is meant by enhanced?

i-Response: " Enhanced natural background" refers to naturally occurring radionuclides

!- ' which have been concentrated in the environment by some external process. This has 3

posed challenges at some sites in determining true " background" due to the presence of j such things as phosphate fertilizers, fly ash, oil and natural gas facilities, etc. Such effects

! on background must be taken into account, however, these variations are not the l responsibility of the licensee.

1

{ 67. Section 5.0 - The remaining cost of decommissioning the SNEF is $22.2 million (1995

dollars). Does this cost include activities to be performed after license termination such as

} demolition of the CV and site restoration? You indicate that trust fund collections are continuing at the annual rate of $2.5 million and that any shortfall prior to full collection j will be made up from GPU operating funds. For how many mor'e years will trust fund

[ collections be made? Assuming four more years of collection leaves a shortfall of $4.1 j million. However, the regulations require that licensees have the full amount of

decommissioning funds in external trusts at some point in the decommissioning process

! and cannot rely on internal funding. Please indicate how you will meet the regulations in

10 CFR 50,75 and 50.82 for decommissioning funding. Please note that earnings on the money in the trust fund can be used to compensate for the shortfall and that only funding for activities prior to license termination need to be considered.

B_esponse: There is no shortfall of funds for decommissioning. The quoted remaining j cost of $22.2 million for SNEC facility decommissioning includes funds for activities that  !

l are not classified as " decommissioning" by NRC regulations in 10 CFR 50.75. The $22.2 million dollar estimate includes $3.8 million for contingency and 0.77 million dollars for l site restoration. Therefore, only $17.63 million are required for decommissioning activities performed prior to license termination. Currently there is over 8 million dollars in the tmst fund. Additional contributions of $2.5 million per year lfo- the next 4 years are projected to be made to cover the activities necessary for license ta mination. l Therefore, the decommissioning trust fund will contain over $18 million (1995 dollars) to l cover expected expenses of 17.63 million dollars for those activities necessary for license termination.

7 1 Met-Ed is currently contributing $1.3 million per year , Penelec is contributing 1 million i dollars per year and Jersey Central Power and Light is contributing 0.2 million dollars per year. Thus, a minimum of $2.5 million dollars per year is being collected from rates for decommissioning . However, these funds can only be spent in accordance with the percentage ownership each company has of the facility (44% ownership for JCPL,32%

ownership for Met Ed and 24% ownership for Penelec). Nonetheless, there will be a total of $2.5 million of available funds contributed each year to the decommissioning fund because any shortage in contribution from any individual owner company will be made up out ofoperating funds.

C301-96-2038 Attachment 1 Page 34 of 44

68. Please briefly describe your industrial safety program for SNEF decommissioning.

Response: The Occupational Safety Program which is in effect for SNEC facility decommissioning activities is an extension of the program in use at Three Mile Island.

The program is documented through the GPU Nuclear Policy and Procedure hianual and the GPU Nuclear Safety & Health hianual.

The Three Mile Island Industrial Hygienist is assigned the collateral duty of oversight of the safety and industrial hygeine activities at the facility. In this capacity, he facilitates implementation of the company's program through facility inspections and by providing guidance to on-site supervision.

69. Please briefly describe any emergency planning you will have in place for SNEF decommissioning.

Response: Emergency plans and procedures will be in place to provide response to the following types of events or incidents: unauthorized entry or attempted entry into the CV, attempted or actual break-in of the SNEC facility, fire, flooding, tornado, contaminated injury, and the inadvertant release of radioactive material.

70. Please provide a dose estimate for members of the public during SNEF decommissioning activities. Please consider someone at the site boundary and at the nearest residence to the SNEF.

Response: As discussed in the SNEC DP (Sections 3.3.2.1 and 3.3.2.2) and the SNEC Environmental Report ( Section 5.4), systems will be in place to reduce as necessary the amount of radioactive materials released to the environment. Releases of airborne particulates will be minimized by using a HEPA filtered ventilation system. Since all radioactive gases have been decayed and released, processing of gaseous wastes will not be necessary. Generation ofliquid wastes will be minimized to keep water processing costs low and to keep liquid discharges to a minimum. It is expected that intrusive groundwater will be the largest component of discharged water. Liquid wastes and intmsive groundwater will be processed as necessary using temporary systems (e.g.

filtration units or demineralizers) prior to discharge. Systems to monitor airborne and liquid emuents also will be in place. The emuent monitoring instmmentation will be used to demonstrate that discharges are low and comply with the limits defined in the Saxton Offsite Dose Calculation Manual (ODCM).

The systems that will be used for reducing and monitoring airborne and liquid emuents have not been purchased. It follows that the specific capabilities of these systems are not known. Therefore, the amount of radioactive material that will be released to the environment during decommissioning activities can not be estimated at this time.

Additionally, estimating emuents at this time would be discult because decommissioning l

i l

C301-96-2038 Attachment 1 Page 35 of 44 I

activities are expected to be dynamic. However, effluents are expected to be low and as I previously mentioned discharges will comply with ODCM limits. Demonstrating compliance with ODCM limits will ensure that doses to the public are low and consistent with the guidelines in USNRC 10 CFR 50, Appendix I. Maintaining doses within the i guidelines of USNRC 10 CFR 50, Appendix I demonstrates that releases of radioactive l effluents are being maintained "as low as reasonably achievable" (ALARA).

Tlie ODCM uses the methodology and models suggested by USNRC NUREG-0133 and USNRC Regulatory Guide 1.109, Revision 1, for calculation of offsite doses due liquid and airborne effluents. Maximum whole body and organ doses from liquid effluents will be calculated for ingestion of fish and drinking water. Other pathways contribute negligibly to the doses from liquid effluents. - For airborne effluents, the maximum organ

. dose will be calculated for 1) inhalation,2) consumption of milk, 3) consumption of green leafy vegetation,4) consumption of meat and 5) standing on the ground. Conservative assumptions, dilution factors, X/Qs and D/Qs will be used to calculate doses. The .

calculated doses, therefore, will be higher than actual doses received by people living near SNEF.

Although all doses from SNEC decommissioning activities can not be estimated at this time, doses to the public are expected to be lower than any dose listed in Section 3.4 of-the SNEC DP (Accident Analyses). A bounding dose calculation was performed for 1 releasing the groundwater in the service tunnel to the Juniata River (Ref. 29 of the SNEC ,

Environmental Report). Using worst case conditions, the doses were found to be a small fraction of the 10 CFR 50 Appendix I limits.

SNEC Decommissioning Environmental Report Ouestions  ;

71. Page 5-1, Section 5.1 - Please discuss the potential impact on the local transportation ,

conditions ofincreased commuter traffic and movement of materials in and out of the SNEF as a result of decommissioning activities.  ;

Response: Other than the additional staffing, transportation related issues may include waste shipments and the arrival of heavy operating equipment. However, most of the heavy equipment needed for the decommissioning project has been used already at the Saxton facility. And no local traffic related issues occurred. Within the waste shipment category, only the disposal of the reactor vessel, steam generator and pressurizer present transportation challenges. All other activities will follow standard procedures in effect throughout earlier dismantlement projects.

Disposal of the steam generator, pressurizer and reactor vessel is discussed in sections 2.2.1.3,2.2.1.4 and 3.3.3.4 of reference 8. The disposal of these large components will likely require three oversized shipments to an appropriate rail terminal. All routes will be inspected for adequacy and will be identified on the shipping permit. As with all

~ . . _ _ .

i l

C301-%-2038 .

Attachment 1 l Page 36 of 44 oversized shipments, a trip plan will be filed with local officials. Other waste shipments )

will utilize normal freight vehicles.

No adverse impact to local transportation is expected from other shipments. Fewer than 100 total waste shipments are anticipated over the 21/2 year decommissioning schedule.

For comparison,165 waste shipments were completed within a three month schedule during the Saxton Soil Remediation Project (reference _4). The soil shipments were completed without incident or complaint. Additionally, radwaste shipments have been made from the Saxton facility since 1962.

Because much of the anticipated transportation activities occurred without incident during prior Saxton projects, no adverse impact is expected to the local trame conditions.

Trame increases from the increased stamng also should have minimal effect. The proposed stafling will be comparable to stamng levels during operation of the facility.

72. Page 5-2, Section 5.3 - Please make specific comparisons between local and/or State of Pennsylvania requirements and the SNEC Soil Erosion and Sedimentation Control Plan, Resnonse: Under the Pennsylvania Code of Regulations relating to erosion control (25 PA Code Chapter 102), requirements for an erosion and sedimentation control plan are listed (Section 102.5). They are as follows:

~

Pennsylvania Environmental Regulations Section 102.5.

1 Erosion and sedimentation control plan. '

(a) The erosion and sedimentation control plan shall be prepared by a person trained and experienced in erosion and sedimentation control methods and techniques.

(b) The erosion and sedimentation control plan shall be designed to prevent accelerated erosion and sedimentation and shall consider all factors which contribute to erosion and sedimentation, including, but not limited to, the i following:

(1) The topographic features of the project area.

(2) The types, depth, slope and areal extent of the soils.

(3) The proposed alteration to the area.

(4) The amount of runoff from the project area and the upstream watershed area.

(5) The staging of earthmoving activities.

(6) Temporary control measures and facilities for use during earth '

moving.

(7) Permanent control measures and facilities forlong term protection.

(8) A maintenance program for the control facilities including disposal of materials removed from the control facilities or project area.

I C301-96-2038 Attachment 1 Page 37 of 44 Guidelines for compliance with the above requirements are described in the April,1990 Erosion and Sediment Pollution Control Program Manual and the November,1996 Erosion and Sedimentation Control Plan Develooment Checklist and Worksheets. Both of these documents were developed by Pennsylvania's Bureau of Soil and Water Conservation, Division of Soil Resources and Erosion Control.

The regulations _first require the preparer of the plan to be trained and experienced in l erosion and sedimentation control methods and techniques. The preparers of the SNEC

! Soil Erosion and Sedimentation Control Plan hold relevant degrees; one having a Master's i Degree in Environmental Pollution Control and the other having a Bachelor's Degree in  !

i Biology which includes courses in Ecology. The main contributor to the plan has had I i intermittent experience in erosion and sedimentation control practices over the past 16 i

years at the Three Mile Island site located in Middletown, Pa.

1-

The specific requirements, as lised under Section (b) above, h' ave been incorporated into j the site plan as necessary. Please see the enclosed site plan for a more detailed 4

comparison. i j 73. Page 5-2, Section 5.4 - Please describe the SNEC radiological effluents, both liquid and l' airborne, and their controls in more detail, to provide sufficient bases for evaluation of l effectiveness. Please include information about the assumed parameters, scenarios, and j methods used to estimate projected doses to the public. Provide comparisons between ]

j projected doses and regulatoiy limits.

{ Resoonse: See response to Question 70.

I 74. Page 5-4, Section 5.5, First Paragraph - Please provide more detail on plans for processing j contaminated water expected to be generated during decontamination of the SNEF,

! including that currently contained in the CV pipe tunnel. What is the range of radionuclide

. concentrations that has been measured in this water? What is the planned pathway for-water that is to be released to the environment? In view of the high groundwater level as f discussed on page 5-5, has consideration been given to use of a dewatering system in order to lower the groundwater level and thus minimize or eliminate infiltration during i decontamination of the pipe tunnel?

Response: Details on the plans for processing contaminated water are not available at this time. Contaminated water processing of will be accomplished in accordance with all applicable regulations and the quality assurance requirements of the decommissioning plan. The range of radionuclide concentrations in the CV pipe tunnel are given in the characterization report, table 4-31. Water released to the environment would be released via the existing outfall to the Raystown Branch of the Juniata River. Previous, exhaustive attempts to lower the ground water level via dewatering systems to support work in the

l C301-96-2038 Attachment 1 Page 38 of 44 pipe tunnel were unsuccessful. The close proximity of Raystown Lake has influenced ground water level at the SNEC facility site.

75. Page 5-5, Top Paragraph - What volume of tritium contaminated water is currently in the CV sump? Where are the other principal sources of tritium and what are the volumes of each? What are the planned release pathways for this water?

Response: The CV sump currently contains approximately 116 gallons of water. This volume varies as condensation from the CV is collected in the sump.

The other principal sources of tritium are given in the SNEC Facility Site Characterization Report which is included with this submittal. From a volume perspective the principal source of tritium would be the concrete in the facility, however the concentration is very low. The entire facility is estimated to contain 4.15 curies of tritium.

Water containing low levels of tritium may be released to the Juniata River as a normal liquid effluent in accordance with the off-site dose calculation manual. This would include releases ofintrusive ground water which may become slightly contaminated. As mentioned throughout the decommissioning plan we plan on minimizing the volume of water used during the decommissioning process, small volumes of more contaminated water may be processed for oft-site disposal at a licensed disposal facility after solidification or absorption.

76. Page 5-5, Second Paragraph - Please provide your best estimate of contamination levels on the inner surfaces of the pipe tunnel. Also, please provide a copy of Reference 29.

Response: The " pipe tunnel" is filled with near surface ground water at this time as it was during characterization and so extensive measurements of surface contamination were not possible. However, surveys ofloose surface contamination have been made in support of personnel entries which indicate surfaces are generally less than 5,000 dpm/100cm2 . In addition to these surveys, core bore samples were taken at several locations in the tunnel.

These samples were analyzed and the results are given in the characterization report.

The " pipe tunnel" in question was part of the same tunnel which was decontaminated and demolished in 1992. Surveys from that section prior to decontamination and from the still present " steam pipe" tunnel section indicate that loose surface contamination is less than 2

5,000 dpm/100cm , while the fixed or total contamination ranges from < minimum 2

detectable activity (MDA) to 230,000 dpm/100 cm with an average of approximately 2

20,000 dpm/100cm ,

77. Page 5-6, Section 5.7 - The validity of the assessment of the radiation exposure of members of the public, which appears to be solely based on estimates contains on the NRCs Generic Environmental Impact Statement (GEIS) and a comparison of the volume of waste assumed to be shipped from the reference test reactor and the volume expected

C301-96-2038 Attachment 1 Page 39 of 44 to be generated by SNEF. In fact, the second paragraph of section 3.1 of the GEIS states that site specific assessments will be required for the environmental report submitted with the application for license modification prior to decommissioning a specific facility. Please provide an independent assessment of the integrated radiation exposure of members of the public or demonstrate that each of the important parameters in the dose calculation for the SNEF decommissioning is bounded by the parameters assumed by Battelle Pacific Northwest Laboratories for the reference test reactor dose calculations used in the GEIS.

Response: The estimated radwaste volume (580 m3 ), stated in section 5.7 of the Environmental Report, was taken from Table 3.3-2 of the proposed Saxton Decommissioning Plan. This volume figure is only an estimate of what remains onsite as part of the decommissioning process. Processed waste still needs to be estimated and therefore will add to the 580 cubic meter estimate but not more than 10% (58 cubic meters). The reason the 580 cubic meter estimate is so low when compared to the NUREG-0586 estimate (4930 cubic meters)is that radioactive materials from past site work have not been factored into the overall estimate. When considering actual radwaste shipped as a result of preliminary decon (from 1972-1974), reactor support building '

dismantlement and demolition (1986-1992), and radioactive soil disposal (1994) the overall radwaste volumes become more representative of the parameters calculated by Battelle Pacific Northwest Laboratories (BPNL)(reference NUREG/CR-1756, Volume 2 of 2, Table 1.2-8). To date, approximately 1835 m3 (243 m3 of demolition debris and 1592 m3 of soil) have been shipped off-site. The total radwaste volume generation for the entire project, including past work, is estimated to be in the area of 2400-2700 m3 , l This estimate is approximately 55% of the NUREG-0586 bounding conditions and the BPNL parameters. The dose to the public is negligible (less than 0.1 person-rem), per Table 7.3-4 of NUREG-0586, for a test reactor in a 30 year safestor condition.

78. Page 7-4, Section 7 4 - The environmental report discusses both the aerial surveys and

" comprehensive soil monitoring and sampling work." Can you compare the results of these two methods, and are they consistent? Please give specific values for Cs-137 deduced from the aerial surveys. Please give details of the analyses that project doses to occupants ,

of the SNEC site, pre-remediation, now, and in the future.

Response: Comparison of Cs-137 soil activities between the EG&G aerial survey (1989) ,

and the onsite soil sampling work (1993) are not consistent. The purpose of the aerial -

survey was to measure Cs-137 concentrations in surrounding areas (outside the Saxton restr:cted area) to determine if there was wind blown contamination emanating from the site property. The measurements made as part of the " comprehensive soil monitoring and sampling work" were performed onsite to characterize the radiological constituents of the soil. The results oflatter study showed measurements significantly higher than background and are documented in the soil remediation report. However, the aerial survey did compare favorably to other studies performed offsite, where background concentrations of Cs-137 for areas surrounding Saxton were determined. In 1988 EG&G made in-situ measurements at off-site locations which compared favorably to the

C301-96-2038 Attachment 1 Page 40 of 44 1989 aerial survey. Results of these measurements are listed in Table 1 of the "In Situ Survey General Public Utilies Facility and Surrounding Area" and in Table 2 of the "An Aerial Radiological Survey of the Saxton Nuclear Experimental Corporation Facility and Surrounding Area." Both documents are provided.

As a result of the 1993 soil characterization sampling and subsequent disposal of onsite soil (1994) the post remediation Cs-137 concentration average is <l.0 pci/g (<3 mrem /yr).

Results of this work is documented in the soil remediation report.

79. Page 7-5, Last paragraph - Please provide a copy of References 7,9,10, and 11 and describe the rationale used in positioning the two bedrock wells. How well has the direction of groundwater flow been established in the bedrock aquifers? Has any radioactive contamination been detected in either of these wells that could be attributed to SNEC activities?

Response: The positioning of the two bedrock wells (MW-1 and MW-2) was based on the recommendation of GEO Engineering of Dover, New Jersey. In August of 1992, GEO Engineering was contracted to investigate the extent of the overburden groundwater along with the depth to the bedrock surface and the orientations of the bedrock groundwater flow pathways. To determine the flow pathways in the bedrock, three nearby bedrock outcrops were investigated. All three outcrops were similar in fracture pattern and bedding plane orientations, indicating the direction of bedrock groundwater movement for the general area, including the SNEC facility. GEO Engineering reports, dated November 18,1992 and June 7,1994, discuss their findings, recommendations and subsequent installation of the groundwater monitoring system. 'Please refer to References 10 and 11 cited in the SNEC Facility Decommissioning Environmental Report.

Collection and analysis from the bedrock monitoring wells began in July 1994 and since that time no radioactive contamination has been detected from these wells. This was previously documented in the Decommissioning Environmental Renort and in the SNEC Facility Decommissioning Plan on page 3-13. Analyses of the overburden groundwater wells hydraulically downgradient of the containment vessel (GEO-3, GEO-6, GEO-7 and GEO-8) also have not detected any radioactivity. Additionally, three wells (GEO-1, GEO-4 and GEO-5) serve as background sampling points for monitoring the containment vessel (CV), since these wells are located hydraulically upgradient of the CV. Monitoring Station GEO-5 is the only point that has shown positive tritium activity intermittently which possibly is attributed to the demolition of former reactor support buildings (e.g.,

Rad Waste Disposal Facility).

80. Page 7-6, First Paragraph - Please provide a description of the gas displacement sampler and how it is used to monitor significant fractures and bedding planes. Is there a means of isolating these zones in boreholes MW-1 and MW-2? Please describe the #1 Morie Filter Pack material in the bottom 25 feet of each borehole as depicted in Figure 7.5-1. Also, is

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C301-96-2038 Attachment 1 Page 41 of 44 I

there any use being made of groundwater from the overburden zone above bedrock in the I vicinity of the SNEF7 Response: The gas displacement sampling system was retrofitted in all the overburden i monitoring wells and initially installed in the bedrock monitoring wells during the spring of 1994. This system allows dedicated sampling to prevent the potential for cross contamination between wells and will achieve minimal agitation of subsurface waters as is the case in a bailer-type collection system. Water is obtained from the well by injecting compressed gas (air) into a one-inch diameter schedule 80 PVC riser pipe and thus displacing the water sample up to the surface via a discharge line. There is a check valve and a 10p pore size sintered polyethylene filter at the lower end of the riser. A high pressure regulator is used in conjunction with the compressed gas cylinder for adjustment of the sample flow rate. The gas displacement samplers (Geomons manufactured by Aguifer Systems, Inc. of Bloomfield, New Jersey) have water-tight well head fittings and i

oversize risers to maximize the capture of water for slow recharging wells.

As part of the analysis performed by the contracted hydrogeologic consultants (GEO Engineering), it was determined that bedrock monitoring wells should be installed at an angle in order to maximize the interception of fractures and bedding planes. The boreholes (MW-1 and MW-2) were drilled into bedrock at an angle of approximately 25 from the vertical to accomplish this. By sealing the annular space with grout and bentonite pellets above a depth of approximately 30 feet, fractures and bedding planes in the areas which would intercept potential outleakage from the containment vessel are monitored. Those areas are isolated in the boreholes and this is what is referred to as significant fractures and bedding planes (for monitoring purposes). . Construction details can be seen on Figure 7.5-1 of the Decommissioning Environmental Report.

The #1 Morie Filter Pack materialin the bottom 25 feet of each bedrock borehole consists of a silica quartz sand which serves as a filter medium for removing sediment. Morie Filter Pack material is sterilized before packaging for sale. Morie #1 connotates the grain size which is a sand fine as opposed to a coarser gravel filter material.

The only potential use of groundwater from the overburden zone in the vicinity of the SNEC Facility (i.e., encompassing those areas which the groundwater could potentially impact) is a pumped well located within the property boundary. It is located in the southern area of the property, approximately 75 meters from the containment vessel.

This well is believed to have been installed in the 1920 - 1930 time frame with some recent upgrades to the pumping / storage tank system. Actual construction details of the well could not be located. It is a non-potable sanitary water source, solely used by company personnel for personal hygiene and washing vehicles and other equipment. The Pa.

Department of Environmental Protection has verbally approved the well for this use and signs are posted at all distribution points preventing the consumption of this water.

Nevertheless, routine analyses of this well water have indicated that radioactive contamination is not present.

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! 81. Page 7-6, Second Paragraph - Is the detection of tritium in GE0-5 noted in this paragraph j the only incidence in which radioactive material has been detected in the overburden i i monitoring wells? Describe the analytical methods used to detec.t and measure the i concentration of radioactive material in water samples collected in the overburden as well

! as the bedrock, including the sensitivity or minimum detection limit of the instrumentation i used.

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Response

The intermittent detection of tritium in Groundwater Monitoring Station GEO-5 is the only incidence in which radioactive material has been detected. All other 3

overburden and bedrock monitoring wells have shown no positive activity for tritium nor '

! any plant related isotope.

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Ground water samples are analyzed for tritium by filtering the sample, mixing with a i scintillation fluid then counting in a liquid scintillation counter, An appropriate count time j- is used to reach a required sensitivity of 200 pCi/L. Samples are also placed in counting j containers for gamma analysis using High Purity Germanium detectors. Required j sensitivity for Co-60, Cs-134, and Cs-137 is 15 pCi/L.

l The laboratory uses approved analy,ical procedures, NIST traceable standards and i sources, and complies with the guidance recommended in Regulation Guide 4.15 for the  !

analysis of all samples. I 1

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82. Page 7-6, Third Paragraph - You state that soil sampling is conducted on an as needed .

j. basis. Please give some examples of when this sampling would be conducted.

} Response: Since the soil remediation work was completed in 1994 there has not been a j systematic need to conduct soil sampling as part of the quarterly environmental j surveillance. Future decommissioning work (such as excavations of sanitary waste

! systems, CV demolition and identified areas still requiring soil remediation) may impact site drainage and, therefore will be evaluated to determine if routine soil sampling is i required. At this time only biased soil sampling will be performed in order to assess

} compliance with NRC site release limits.

] 83.- Pages 7-7,7-8, Section 7.6 - In accordance with discussions during the site visit on May

[ 9,1996, please submit the SNEC plan for the final radiological survey, including methods to provide and ensure consistency and compliance with release criteria.

j Response: GPU Nuclear has requested Saxton Nuclear Experimental Corporation site free release criteria in the DP that is consistent with the proposed revision to 10 CFR 20 concerning license termination residual radioactivity limits. The NRC has not yet finalized

guidance on acceptable methods of demonstrating compliance with the revised criteria.

] Three NUREGs are under development which will provide such guidance: NUREG-1505

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C301-96-2038 l Attachment 1 Page 43 of 44 "A Nonparametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys", NUREG-1506 " Measurement Method for Radiological Surveys in Support of New Decommissioning Criteria", and NUREG-1507 " Minimum l

Detectable Concentrations with Typical Radiation Survey Instmments for Various l Contaminants and Field Conditions" It is our intent to comply with the guidance promulgated when it is available.

Final termination survey activities will not take place until the year 1999-2000 time frame.

It is expected that the proposed release criteria developed as part of the enhanced participatory rulemaking process and the compliance guidance will be approved before then. Ifit is not, we intend to use that guidance which is available (at this time, l NUREG/CR-5849 " Manual for Conducting Radiological Surveys in Support of License Termination" and NUREG/CR-5512 " Residual Radioactive Contamination From Decommissioning").

Per our discussions with your staff, it is our intention to submit a separate "Saxton Final Survey Plan" for NRC approval as other licensees have done as part of the decommissioning process. We have prepared an approved final survey plan previously as part of the release of the Saxton facility outbuilding demolition and have provided input on the final survey plan for the Fort St. Vrain decommissioning project. In addition the company participated in the enhanced participatory rulemaking process and we me very familiar with the requirements.

The final survey plan will incorporate the use of data quality objectives as called for in the draft guidance to demonstrate compliance with the release criteria. The plan will incorporate the following items:

A detailed description of the types, extent, and locations of the measurements and samples that will be obtained.

A description of the equipment and techniques that will be used for measuring, sampling, and analyzing the data.

A description of the methods for interpreting and evaluating the data quality.

A list of quality control requirements for ensuring data quality.

Detailed implementing procedures will be in place to carry out the final survey plan requirements.

Instmmentation will be selected which will be capable of measuring levels sufliciently below the release or action guideline values. These instruments will be calibrated using standards and sources which are traceable to the National Institute of Standards and Technology (NIST). These calibrations and operability checks will be made using sources I

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Page 44 of 44 which are representative of the radionuclide mix or mixes encountered at the site. All instrument calibration and maintenance will be conducted in accordance with industry j recognized practices and standards and approved procedures.

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! All aspects of the survey will be documented in accordance with the plan requirements and

! approved procedures. The final survey report will be presented in a format which will l stand alone and not require the use of other supporting data or documents to conclude j that the applicable release criteria has been met.

The quality assurance program de. ailed in Section 7.0 of the Decommissioning Plan will

. be implemented during all phases of the final survey to ensure the validity of the results.

Given the changes which are pending relative to the regulations on termination release -

j criteria and the compliance guidance, we feel it is prudent to wait to incorporate those i aspects in the Final Survey Plan rather than submit a document which would be outdated and inadequate prior to the start of such survey activities.

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i C301-96-2038 Attachment 2 Page 1 of 1 Requested reference documents:

1. 1994 Saxton Soil Remediation Project Report- supports question 31 response.

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2. Graph of Tritium Activity for Well GEO-5 from 7/94 - 4/96- supports question 32 response.

3. Residual Radioactivity Program Test Problem Default Data- supports question 63  !

response.

4. SNEC Soil Erosion and Sedimentation Control Plan- supports question 72 response.

5. Aerial Radiological Survey of SNEC Facility and Surroundings,- supports question 78 response.

6. In Situ Survey General Public Utilities Facility and Surrounding Area - supports question 78 response.

7. Preliminary Hydrogeological Investigation Saxton Nuclear Experimental Station - i supports question 79 response..

8. Geologic, Chemical, Radiometric and Geotechnical Studies of Samples from Eleven Drill Holes in Superficial Materials- Volume I, and Report on Drilling and Radiometric Analysis of Samples Collected at Sites of Spent Resin and Liquid Waste Tanks- Volume 2 -

support question 79 response.

9. Phase I Report of Findings - Groundwater Investigation at the Saxton Site -; supports question 79 response.

10. Summary of Field Work Saxton Nuclear Experimental Station - supports question 79 response.

I 1. Training Instructor Qualification and Lesson Plan Samples - support question 23 response.

12. Saxton Site Characterization Plan - supports question 24 response.

13. SNEC Facility Site Characterization Report - supports question 24 response.

14. TLG Activation Analysis Report - supports question 30 response.

15. Maximum Offsite Dose from Release of Saxton Pipe Tunnel Water - supports question 76 response.

j ATTACHMENT 2 REFERENCE 1-SUPPORTS QUESTION 31 RESPONSE I

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