ML20216D082

Environ Assessment by Us NRC Related to Request to Authorize Facility Decommissioning

ML20216D082
Person / Time
Site:	Saxton File:GPU Nuclear icon.png
Issue date:	03/31/1998
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20216D066	List: ML20216D061 ML20216D082 ML20216D120
References
NUDOCS 9804150171
Download: ML20216D082 (48)

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l Environmental Assessment by the U.S. Nuclear Regulatory Commission j Related to the Request to Authorize j Facility Decommissioning {

Saxton Nuclear Experimental Facility - Saxton Nuclear Experimental Corporation and GPU Nuclear, Inc.

Docket No. 50-146 March 1998 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Reactor Program Management 1

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t l CONTENTS Fig u re s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv Ta ble s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv i Acronyms .. . ..... ............................................. ....... ..v 1 Introduction . . . . . . ...... ... ......... ..... ..... .... ..........1-1 1.1 Propose d Action . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 -2 1.2 Need for Proposed Action . . . . . . . . ... ... ... . ... .. .. .. ... 1-3 1.3 Decommissioning Altematives . . . . . ... .......... .................1-3 l 1.3.1 DECON . . . . . . . . . . ......... ..... .. . .... ...... .... 1-3 )

1.3.2 SAF STO R . . . . . . . . . . . . . . . . . . . . . . ....... .... .. ... ... .. 1-3 1.3.2.1 Further Deferral of Dismantlement . . . . . . . . . . ......... . . . . . . . . 1 -4 1.3.3 ENTOMB .. ..... ............. ............ ............... 1-5 1.3.4 No Action . . . . . . . . . . . . . . . . .. .. .... . . . . .......... 1-5 1.3.5 Alternative Use of Resources . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 1.3.6 Chosen Decommissioning Altemative .... . . ...... . . . . . . . . . . . 1 -6 2 SNEF Site Description .. . . . . . .. ... . . ... . . .. .. . 2-1 2.1 Geography and Demography . .... . . . .. ... . . . . . . . 2- 1 2.2 Climate .. . .. .. .. ... . ... . ...... . ... . . . 2-1 2.3 Hydrology .. .. ...... . .. . . . .. .. ... . .. 2-2 2.4 Geology . .. . .. ... ... ..... .. .... ... .... . ..... .... 2-3 2.5 Other Environmental Features .. ... .. .. ..... ...... .. . . 2-4 3 SNEF Decommissioning Plans . . . . . . . . . . . .... ......... . ........ 3-1  !

3.1 Preliminary Activities . . . ...... .............. ........ . ..... . 3-1 3.2 Plant Radioactive MaterialInventory . . . ..... . ...... . . . . . . . . . . . 3-2 3.2.1 Activation Levels . .. ...... ... ... . ..... ........... . . . . 3-2 3.2.2 Area Dose Rates and Contamination Levels . . . . . . . . . . . . .... . . 3-2 l 3.3 Decommissioning Plans . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-10 l

. . 3-10 3.3.1 Preparation Phase . . . . ...... ..... .. .. ..... ..... ..

3.3.2 Operations Phase . . . . . . . . ... ... . ........ . .. .... . . . . 3-11 3.3.3 Survey Phase . . . . . . . ... ... . .. . .. ............ ..... ... 3-13 3.3.4 Site Restoration Phase . . . ... .. . .........................3-14 t

I 4 Environmental Effects of Proposed Action . . . . . ..... . . .... ... . . . . . . . 4- 1 4.1 Effects on Human Activitie s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 )

4.2 Effects on Terrain, Vegetation, and Wildlife . . . . . . . . . . . . . .. . . . . ...... .... 4-1 l 4.3 Effects on Adjacent Waters and Aquatic Life . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.4 Effects of Released Radioactive Material ............... ............ . 4-1 4.5 Effects on Groundwater . . . . . .. . ...... .. . ....... . ... . 4-5 4.5.1 Groundwater Flooding of Underground Structures . . . . ....... . 4-5 4.5.2 Groundwater Monitoring Program . .. . ... . . .... .. .. 4-6 4.5.3 Groundwater Monitoring Results . . .. . . . .. .. . . .. . 4-7 ii E

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4.6 Occupational Radiation Exposure ..... . ............. ............. . 4-7 l 4.7 Effects of Released Chemical and Sanitary Wastes . . . . . . . . . . . . . . . . . . . . . . . 4-7 '

4.8 Effects of Nonradiological Material . . . . . . . . . ............................4-9 j 4.9 Socioeconomic Effects . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-9 1 i

l - 5 Post ulated Accidents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 i j 5.1 Material-Handling Accident--Dropped Resin Vessel . . . . . . . . . . . . . . . . . . . . . . . .. 5-1 i 5.2 Fire Accident-Combustible Waste Stored in the Yard . . . . . . . . . . . . . . . . . . . . . 5-2 i 5.3 Rupture of Vacuum Filter Bag . . ........................... .......... 5-2  !

5.4 Segmentation of Components or Structures Without Local Engineering Controls . . 5-2 5.5 Oxyacetylene Explosion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-3 5.6 Explosion of Liquid Propane Gas (LPG) Leaked From a Front-End Loader . . . . . . . 5-3 i 5.7 Failure of Liquid Waste Storage Vessel ...................... .......... 5-3 5.8 in Situ Decontamination of Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-4 5.9 Loss of Support Systems . . . . . .. .................................5-4 5.10 Extema! Events . . . . . . . .. .............. ....... ..... . .. . . . . . 5-5 i 5.11 Offsite Radiological Events . . . . . . . . . . . . . . . . . . . . . ..................5-5 5.12 Containment Vessel Breach . . . ................... ................ 5-5 l 5.13 Summary of Postulated Accident impacts ............................. . 5-6

! 6 Agencies and Persons Consulted ano tiources Used . . . . . . . . . . . . . . . ....... . 6-1 1

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7 Conclusion . .. ... .. .. .... .... ....... ... ................... . 7-1 l

. 8 References . .. . . ... . . ... . .... . .............. 8-1 l l

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o,i Figures Fig ure 3.1 Saxton Site Plan . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-3 Figure 3.2 SNEF Containment Vessel, Sectional View (Looking West) . . . . . . . . . . . . . . . 3-4 Figure 3.3 SNEF Containment Vessel, Sectional View (Looking North) . . . . . . . . . . . . . . . . 3-5 Figure 3.4 SNEF Containment Vessel, Plan View (812 ft level) . . ............ ......36 )

Figure 3.5 SNEF Containment Vessel, Plan View (795 G Level) . . . . . . . . . . . . . . . . . . . . 3-7 j Figure 3.6 SNEF Containment Vessel, Plan View (781 ft Level) . . . . . . . . . . . . . . . . . . . . . 3-8 Figure 3.7 SNEF Containment Vessel, Plan View (765 ft level) . . . . . . . . . . . . . . . . . . . . . 3-9 Figure 3.8 Time Lines for SNEF Decommissioning . . . ............ . . . . . . . . . . . . 3- 10 '

Figure 4.1 SNEF Monitoring Well Locations. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-8 Tables Table 1.1 Occupational Dose Comparison Between Decommissioning Altematives . . . . . . 1-4

. Table 3.1 Calculated Activation Products Within the Reactor Pressure Vessel . . . . .. . . 3-2 l

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>k Aeronyms AEC U.S. Atomic Energy Commission ALARA As Low as is Reasonably Achievable CFR Code of Federal Regulations CV Containment Vessel DECON Decontamination DSB Decommissioning Support Building DSF Decommissioning Support Facility DOT U.S. Department of Transportation ENTOMB Entombment EPA- U.S. Environmental Protection Agency GEIS Generic Environmental Impact Statement (NUREG-0586)

GPU General Public Utilities Corporation HEPA High-Efficiency Particulato Air (filter)

INEEL ldaho National Engineering and Environmental Laboratory LPG Liquid Propane Gas MSL Mean Sea Level' MWT Megawatt Thermal Energy NPDES National Pollutant Discharge Elimination System NRC U.S. Nuclear Regulatory Commission ODCM Offsite Dose Calculation Manua!

PaDEP Pennsylvania Department of Environmental Protection PAG Protective Action Guide PaGS Pennsylvania Geological Survey PENELEC Pennsylvania Electric Company PSDAR Post-Shutdown Decommissioning ActivF.ies Report PWR Pressurized Water Reactor RWDF Radioactive Weste Disposal Facility SAFSTOR Safe Storage SNEC Saxton Nuclear Experimental Corporation SNEF Saxton Nuclear Experimental Facility TMi-2 Three Mile Island, Unit 2 USAR Updated Safety Analysis Report y h

V q'I 1 Introduction The Saxton Nuclear Experimental Corporation (SNEC) facility (SNEF) is located in south central Pennsylvania, near the Borough of Saxton in Bedford County, with SNEC as the original licensee. Built in 1960-42, the facility was licensed as a demonstration power reactor at several different power levels during its history.' The maximum licensed power level was 35 megawatts of thermal energy (MWT) although the prevalent licensed power level was 23.5 MWT. For most l of its operational life (1962-72), the facility functioned at full power and was used for research and training. The reactor was used for testing mixed oxide fuels (Ref.1), where experiments with fuel cladding that was intentionally " failed" resulted in some of the plutonium, americium, and fission product contamination remaining at the facility. The facility was placed in a condition equivalent to the status later defined by the U.S. Nuclear Regulatory Commission (NRC) as

'SAFSTOR" (safe storage) after it was shut down in 1972, and the operating license was converted to possession only status (Ref. 2)..

The SNEF is maintained under a Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part .50) license and its associated technical specifications. The license is due to expire on February 11,2000. In 1980, SNEC formed on agreement with GPU Nuclear to use GPU Nuclear and its resources to maintain, repair, modify, or dismantle buildings and structures at the SNEF on an as-needed basis. Both SNEC and GPU Nuclear are subsidiaries of the same parent company, General Public Utilities Corporation (GPU). Although SNEC remains the owner of the facility, a license amendment request was submitted to designate GPU Nuclear as a co-license holder. This request was approved by the NRC in Amendment No.13, dated May 10,1996.

Thus, GPU Nuclear now has corporate responsibility to comply with all applicable regulations, the provisions of the license, and facility-specific technical specifications. ,

Reactor support structures and/or buildings were decontaminated and demolished in 1987-89.

The licensees now propose to decontaminate and dismantle the facility's remaining structures-namely, the containment vessel (CV), which contains almost all of the remaining

. radioactive material on the site, the concrete shield wall located around the northwest and northeast quadrants of the CV, the tunnel sections that are immediately adjacent to the outer circumference of the CV, and remaining portions of the septic system, weirs, and their associated underground piping. The licensees propose to undertake these operations in preparation for releasing the site for unrestricted use. On February 16,1996, the licensees submitted a decommissioning plan in accordance with the regulations in 10 CFR 50.82 in effect at that time (Ref. 3). Based on the requirements of 10 CFR 5153 (Ref. 4), they submitted a facility decommissioning environmental report on April 17,1996 (Ref. 5). In response to requests for additiona! information from NRC, the licensees responded to July 18,1996, with additional information on the decommissioning plan and environmental report, and on March 3 and 31,1998, with additional information on the environmental report.

On July 29,1996, the NRC amended its regulations (effective August 28,1996) on decommissioning and termination of operating licenses tuouclear power reactors [61 FR 39278 i

(1996)]. The amendments c!arified ambiguities in the existing rule and codified procedures intended to reduce the regulatory burden, provide greater flexibility, and allow for greater public participation in the decommissioning process. The new rule allows greater flexibility regarding the type of actions that can be undertaken without NRC approval. The requirement for a decommissioning plan was removed from the regulations. Information on a licensee's plans for facility decommissioning would be provided to the NRC in the form of a post-shutdown decommissioning activities report (PSDAR)(10 CFR 50.82(a)(4)). Information required from the 1-1 1

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V latnopuction licensee in a PSDAR consists of (1) a description of the planned decomtr issioning activities along with a schedule for their accomplishment, (2) an estimate of expected costs, and (3) a discussion that provides the reasons for concluding that the environmental impacts associated with site-specific decommissioning activities will be bounded by appropnate previously issued environmental impact statements. Based on the new requirements, a PSDAR must be submitted to the NRC and be made available to the public for comment. Further, decommissioning plans for power reactors that were either submitted for approval or approved before the effective date of this rule will be considered as PSDAR submittals as per the . l provisions of 10 CFR 50.82.

In accordem Wh the provisions of the new rule and 10 CFR 50.71, the licensees also submi".sd the SNEC Facility Updated Safety Analysis Report (USAR), Revision 0, on October 25, 1996, Revision 1 on August 21,1997, and revision 2 on February 3,1998. Because of restrictions in the current license and technical specifications against the performance of decommissioning activities, SNEC Technical Specification Change Request 59 was submitted on November 25,1996, as supplemented on May 30, June 4 and 16, August 21, and September 16, 1997, and February 3 and 9,1998. The requested changes to the license and technical specifications would allow decommissioning of the SNEF by (1) accommodating .

decommissioning activities at the SNEF, (2) catablishing specific technical specification controls over decommissioning activities, (3) establishing limiting conditions for performing decommissioning activities, (4) extending exclusion area controls to include the SNEF decommissioning support facility (DSF), (5) establishing requirements for a Radiological Environmental Monitoring Program and an Off-Site Dose Calculation Manual (ODCM), and (6) establishing requirements for Technical and Independent Safety Reviews.

The SNEF was initially licensed before enactment of the National Environmental PoCcy Act and before a formal environmental impact assessment or statement was required. Hence, in accordance with current requirements in 10 CFR 51.53(d), SNEC developed and submitted the l

SNEC Facility Decommissioning Environmental Report to address actual or potential environmental impacts resulting from the decommissioning of the SNEF, including decontamination, dismantlement, and site restoration activities leading to license termination.

1.1 Proposed Action

. The licensees have proposed a DECON (i.e., decontamination) altemative as their preferred option (Ref.1). Specifically, they propose immediate dismantlement of the SNEF in order to (1) remove all remaining components from the site and all structures down to 91 cm (3 ft) below ground level, (2) stabilize radiological conditions at the site, and (3) establish a basis for requesting that the site be released for unrestricted use. At present, the licensees intend to have  ;

low-level radioactive waste generated at the SNEF site shipped to the Waste Management '

L Facility in Barnwell, South Carolina, or to Envirocare in Utah.

l As currently planned, all ute decommissioning activities will be completed in 1999, the NRC license will be terminated in 2000, and nonradiological site restoration will be finished in 2001.

As defined by NRC regulation 2 in 10 CFR 50.2, decommissioning means to remove a facility or site safely from service and reduce residual radioactivity to a level that permits release of the property for unrestricted use and termination of the license. Based on this definition, non- .

radiological site restoration after the license is terminated is not part of decommissioning; thus, i this assessment does not address restoration activities beyond what is necessary to allow unrestricted use of the site.

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lentooucnos 1.2 Need for Proposed Action

, The proposed action is necessary because of SNEC's 1972 decision to cease operations i

permanently at the SNEF. As specified in 10 CFR 50.82, any licensee may apply to the NRC for j authority to surrender a license voluntarily and to decommission the affected facility (Ref. 3).

Further,10 CFR 51.53(d) stipulates that each applicant for a license amendment to authorize decommissioning of a production or utilization facility shall submit with its application an environmental report that reflects any new information or significant environmental change associated with the proposed decommissioning activities (Ref. 4).

r .1.3 Decommissioning Alternatives Decommissioning is intended to take a nuclear facility out of service and reduce residual l radioactive contamination at the site to levels that will permit termination of the facility license.  !

Altemative methods to accomplish decommissioning are evaluated in NUREG-0586, " Final Generic Environmental impact Statement on Decommissioning of Nuclear Facilities,"(GEIS) l l

(Ref. 6), and include the following: DECON, ENTOMB (entombment), or SAFSTOR. Licensees l are required to evaluate the various altematives and to propose and justify the one they prefer. I The chosen altemative must be implemented in accordance with applicable regulations. These methods and the "No Action' attemative are addressed below. The information provided herein l summarizes information in the GEIS relative to the licensees' proposal.

I 1.3.1 DECON  !

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! The DECON decommissioning attemative consists of the onsite decontamination and/or removal l and transport of all contaminated or activated equipment, components, systems, and structures

! from the facility to an authorized site. This attemative is generally intended to reduce radioactive '

l contamination at the facility to a level that permits unrestricted use shortly after the cessation of i facility operation. The GEIS (Ref. 6) found DECON to be an acceptable decommissioning l attemative and, in general, the preferred approach. The licensees have proposed the DECON decommissioning attemative for all remaining structures and systems at the SNEF.

The licensees state that, in this case, the major advantages of immediate dismantlement are that this approach will most quickly remove contaminated and activated components from the SNEC site (i.e., from below ground level), which in tum will stabilize radiological conditions at the site and allow its release for unrestricted use. The licensees also contend that immediate dismantlement will allow them to make use of GPU's remaining expertise relative to the SNEF L and Three Mile Island, Unit 2 (TMI-2), to plan and implement dismantlement activities (Ref.1). In I addition, sites for the disposal of low-level radioactive waste generated in Pennsylvania are currently available at the Waste Management Facility in Bamwell, South Carolina, and at  ;

Envirocare in Utah under existing contracts maintained by the licensees. Thus, the waste can currently be sent directly to burial, thereby reducing the cost of decommissioning (Ref.1).

i 1.3.2 SAFSTOR The SAFSTOR attemative involves putting a facility in a safe condition and maintaining it in that J state until it can be decontaminated and dismantled. During SAFSTOR, a facility is left intact with fuel removed from the reactor vessel and with radioactive fluids drained from systems and disposed of in a safe manner. The SAFSTOR period would generally be sufficient to allow for substantial radioactive decay to take place, thus reducing both the level of radioactivity and the 1-3

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k lunoovenos quantity of contaminated and activated material that must eventually be dealt with. The GEIS (Ref. 6) recognizes SAFSTOR as tan acceptable decommissioning attemative.

1.3.2.1 Further Deferral of Dismantlement 1

in effect, the SNEF has been in a SAFSTOR condition since it was shut down in 1972. Complete dismantlement has already been deferred for 25 years. The option of deferring dismantlement for 30 more years has been analyzed by the licensees (Ref.1). Deferring dismantlement of the facility's remaining structures for another 30 years would include the benefit of further radioactive decay, thereby reducing the potential for personnel exposures.

Table 1.1 provides the licensees' comparison of radiation doses for the various altematives. The SNEF has already been shut down for more than 25 years, long enough to permit the decay of virtually all of the shorter-lived radionuclides. Had facility decontamination and dismantlement i taken place shortly after shutdown, occupational radiation doses could have been 50-100 times l those currently projected. By comparison, a delay of 30 more years might reduce potential radiation doses by about half their current values. Thus, most personnel exposure savings to be gained from deferring dismantlement have already been realized. i Deferral for 30 more years also has several disadvantages. First, the experience base currently available would be lost. GPU, the licensees' parent company, currently employs individuals who worked at the SNEF while it was operational. Their knowledge of the plant from that era has proven invaluable (Ref.1). In addition, GPU Nuclear has recently remediated and demolished the controt, auxiliary, and radioactive waste buildings and structures at the SNEF and placed TMI-2 in post-defueling monitored storage. The skills of people who worked on these projects are directly applicable to the work yet to be done at the SNEF, and these same people will no longer be available in 30 years (Ref.1).

Moreover, high groundwater conditions at the SNEF site could lead to loss of containment, which could cause either an unmonitored release path or groundwater flooding of the lower elevations of the CV (Ref.1). The reactor vessel and other associated contaminated systems extend below Table 1.1 Occupational Dose Comparison Between Decommissioning Alternatives Integrated Doses - Person-Sievert (Person-Rem)

Iad SAFSTOR DECON 30-Year Deferral immediate Asbestos Remediation 0.023 (2.3) 0.049 (4.9) l System Dismantlement 0.081 (8.1) 0.177 (17.7)

Reactor Vessel and Steam Generator Removal 0.029 (2.9) 0.063 (6.3) l Structure Decontamination and Dismantlement 0.002 (0.2) 0.0035 (0.35)

Waste Management 0.011 (1.1) 0.025 (2.5)

Total 0.146 (14.6) 0.318 (31.8) 1-4 I

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t lutnopocnos ground level, and groundwater flooding would create an extremely difficult dismantlement scenario by increasing the quantity of radioactive waste and the overall cost of the dismantlement process. Since the steel liner is covered by concrete on the inside (i.e., below grade level), degradation of the liner could go undetected. In addition, the high moisture content of the atmosphere inside the facility would hasten the degradation of CV systems and structural components that will be needed to support dismantlement (e.g., the polar crane and related equipment). These conditions could make decommissioning less safe and more complicated for workers as the components continue to deteriorate. As a result, dose-savings due to radioactive decay could readily be offset by an increase in other risk factors.

The continuing need to carry out maintenance and comply with surveillance requirements for the SNEF provides an added disadvantage. This situation would include an escalating effort over the next 30 years to manage a deteriorating facility that will never again be used and could pose increasing radiological risk to both the local public and the environment.

Finally, the cost of radioactive waste disposal is likely to increase markedly over a 30-year period (Ref.1).. At present, disposal costs are rising at a much higher rate than inflation. Sites licensed for the disposal of low-level radioactive waste generated in Pennsylvania are currently available in South Carolina and Utah, and wastes can be sent directly to burial. Future waste disposal choices are less certain, introducing the possibility of long-term radioactive warte storage at the SNEF site. This is undesirable because the site is located on a flood plain. Clearly, the SNEF site is unsuited, and was never intended, for the long-term storage of radioactive waste.

For all these reasons, the licensees have decided not to opt for a deferral of 30 more years before beginning the dismantlement and decontamination of the SNEF.

1.3.3 ENTOMB The ENTOMB alternative involves encasing radioactive systems and contaminated material in a structurally long-lived substance (e.g., concrete). An entombed structure must be appropriately maintained, and surveillance must be carried out until the radioactive material decays to a level that permits unrestricted use of the property. The GEIS (Ref. 6) did not consider the ENTOMB

option as viable for decommissioning reactors because of uncertainties related to cost and i .

because the reactor site would be indefinitely dedicated to the storage of radioactive material.

1.3.4 No Action The "No Action" altemative implies that a licensee would simply abandon a facility. Where there l

is no need for long-term, onsite storage of fuel, the objective of decommissioning is to restore the site of a radioactively contaminated facility to unrestricted use. To ensure that the potential for negative impacts on public health and safety is kept within acceptable bounds, some action is required-even if that action is as minimal as conducting a final radiation survey to verify I acceptably low radioactivity levels. Therefore, as concluded in Section 2.4.1 of the GEIS (Ref.

- 6), the "No-Action" altemative is not viable for decommissioning. NRC regulations do not permit a licensee simply to abandon a contaminated facility after operations cease.

1.3.5 Alternative Use of Resources The licensees state that sufficient funds are either currently available or will be provided by the owners of the SNEF to complete dismantlement and decommissioning of the facility in a safe and satisfactory manner. Most of the expense will be covered by a decomrrassioning trust that 1-5

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lefnopvenom was established for the specific purpose of fully funding the cost of decommissioning the site for unrestricted use. The proposed action would remove reactor-related radioactive material from the site down to unrestricted release levels and transport it to safe, monitored custodianship at a licensed disposal facility. Consolidating low-level radioactive materials in this way is generally regarded as a benefit to the public and to the environment. Based on these considerations, DECON represents the optimum use of available resources.

1.3.8 Chosen Decommissioning Alternative Based on the above-discussed analysis, the licensees chose the DECON altemative for decommissioning the SNEF. Immediate dismantlement will place the SNEF in a stable, secure, and decontaminated condition within the shortest time practical (Ref.1). 1 Radiological conditions at the facility are now at a level that allows workers to remove components from the facility safely, without posing significant threats to the safety and health of either workers or local residents. In addition, the technology already exists to decommission the ,

site safely and efficiently. The integrated dose determination for conducting an immediate l dismantlement is credible and consistent with current industry norms. Differences in the calculated integrated doses presented in Table 1.1 are relatively small. Compared with the i advantages of immediately removing the site as a potential source of environmental radioactive j l contamination, the benefit gained from further deferral of decommissioning would be extremely j limited (Ref.1). 1 The staff finds that the process used by the licensees in choosing the DECON decommissioning altemative is consistent with the method set forth in the GEIS (Ref. 6) and that the proposed  ;

altemative meets the requirements specified in 10 CFR 50.82. The staff concludes there is l I

reasonable assurance that immediate decommissioning will not pose a significant radiological risk to the health and safety of the public or to the environment. ,

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l 2 SNEF Site Description 2b Geography and Demography The SNEF site consists of 0.465 fenced hectares (1.15 acres) on a parcel of approximately 60 hectares (150 acres) owned by the Pennsylvania Electric Company (PENELEC). The site is located about 160 km (100 mi) east of Pittsburgh and 145 km (90 mi) west of Harrisburg in the

! Allegheny Mountains,1.2 km (0.75 mi) north of the Borough of Saxton in Liberty Township, l Bedford County, Pennsylvania. It is on the north side of Pennsylvania Route 913,27 km (17 mi) south of U.S. Route 22, and about 24 km (15 mi) north of the Breezewood Interchange of the Pennsylvania Tumpike.

i The SNEF was built adjacent to the Saxton Steam Electric Generating Station of PENELEC, a  !

l subsidiary of GPU. This coal-fired station operated from 1923 to 1974 and was demolished i between 1975 and 1977. Much of the property consists of gently sloping open grassland that was restored to its natural condition after the Saxton Steam Electric Generating Station was demolished. A cemetery adjoins the eastem boundary of the proper 1y, and undeveloped woodlands and residential areas are located along the northem, southem, and westem boundaries. The area surrounding the site is generally rural, forested, and mountainous.

The population density of the area is low, with small concentrations in the valleys and along main L

. highways. The population of the three surrounding counties has declined between 1980 and 1990. As recorded during the 1960 census, the estimated populaiion of the Borough of Saxton during construction of the SNEF was 975. Thirty years later, the population was recorded at 838, )

a decline of 16.3 percent. l l ..

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The nearest population center is the city of Altoona (pop. 51,881,1990 census), which is about l 32 km (20 mi) north-northwest of the SNEF site. The nearest incorporated towns other than the l l Borough of Saxton are all located to the east--Coalmont Borough, about 4 km (2.5 mi) away I from the site; Dudley Borough, about 5.5 km (3.4 mi) away, and Broad Top about 8.5 km (5.3 mi) j away.

The nearby Raystown Branch of the Juniata River is widely used for recreation by local residents, primarily for boating and fishing. The vast majority of recreational activities along the l river, however, are centered downstream of the site on Raystown Lake. The lake is the focus of ,

one of the better recreational areas in this part of Pennsylvania, and this area has been j intensively developed by the Federal Govemment. Over 475,000 visitors per year enjoy the boating, fishing, camping, hunting, picnicking, and other recreational attractions of the region.

2.2 Climate The climate of south-central Pennsylvania is best described as a study in contrasts. During the l late spring, summer, and early fr.1, the region is dominated by air masses that originate in the southeastem United States. Warm and humid conditions are normal during this time of year, )

along with air mass thunderstorms and precipitation associated with cold fronts. These frontal boundaries are more active (weather-wise) during the spring and autumn, when the polar jet stream is over the region.  ;

1 The winter season is cold and often overcast, and winter air masses are generally cold and dry.

Winds associated with these air masses are generally from the west-northwest, originating in 2-1

% j SNEF SITS DescnirTiosi central Canada and moving down into the region behind active cold fronts and . low-pressure systems that move north along the Atlantic seaboard. The area experiences a high percentage of cloud cover, due in part to its close proximity to the Great Lakes. As the cold, polar air passes over the relatively warm lakes, condensation forms and produces lake-effect snows, which are common along the shores of the lakes. Drying occurs as the distance from the lakes increases, and a constant cloud cover dominates westem Pennsylvania. In addition, the region has steep-sided valleys, with mountain winds during the day. The winds cause an increase in cloud cover as daytime heating causes rising air currents and subsequent condensation (clouds).

Precipitation in the region is mainly due to air mass thunderstorms, the passage of cold fronts from the west, and low-pressure storms thai move along the Appalachian Mountains through the St. Lawrence Valley. These storms generally produce copious amounts of rain from a northeasterly direction. Annual amounts can range from 75 to 100 cm (30 to 40 in). One-quarter of winter precipitation occurs in the form of snow. Major fall and winter coastal storms that l- produce large amounts of precipitation in the eastem half of the state have minimal effect on the l- SNEF site.

i l Winds in the Saxton region are influenced by topographic features. The facility lies in a large l

valley framed by Terrace and Saxton Mountains to the east and by Tussey Mountain and the Allegrippis Ridge to the west. The mountains and valley are generally situated southwest to northeast. With the large-scale wind flow out of the west, " wind channeling" occurs at the lower levels and gives rise to a small-scale southwesterly flow up the valley. The varying topographic l regime also causes valley-slope circulation pattems. During the daytime, beginning in midmorning and continuing until around sunset, the wind crosses the vallt.y and blows up the sides of the mountain as daytime heating near the surface creates unstable, rising air and an j increase in cloud cover. Beginning around midnight and continuing until shortly after sunrise, the l wind tends to blow down the mountain slope as the land surface along the slopes cools more rapidly than at the base of the valley. This coc.or, denser, stable air sinks toward the valley and moves down the canyon. Wind speeds are generally light at the SNEF site (below 4.5 m/s or 10 l mph), primarily due to the width of the valley in which it is located.

L 2.3 Hydrology l

l The site (as well as a portion of the PENELEC area and surrounding uncontrolled lands)is located in the 100-year flood plain of the Raystown Branch of the Juniata River. A srnall stream i known as Shoup's Run flows west and transects PENELEC property to the south of ti SNEF L and empties into the Raystown Branch of the Juniata River. Normal elevation of the rit near l the facility is about 242 m (794 ft) above mean sea level (MSL); the site and adjacent property lie about 5 m (17 ft) above river level.

The primary body of water in the vicinity of the facility is the Raystown Branch of the Juniata River, which meanders along its water course in an overall flow direction to the northeast and generally borders the northem and westem edges of the property. The watershed extending upstream from the Borough of Saxton is about 1958 km 2 (756 mi2 ). About 55 km (34 mi) downstream from the site, the Raystown Branch of the Juniata River is dammed, impounding the river to form Raystown Lake. At normal pool level, the lake is 43 km (27 mi) long and has an area of 3360 hectares (8300 acres).

Underlying the site are three distinct subsurface zones that have different water-bearing and transmitting properties. The site is immediately underlain by a filllayer that consists of fly ash, cinders, and/or sitt and sand-size sediment. This fill layer is underlain by a layer of boulders in a I

2-2 ,

I

e,

]

%,f 1

SNEF SITE Desemenom silty clay matrix. ' Bedrock lies beneath this boulder layer. Field permeability tests were conducted by the licensees' consultant in selected bore holes and laboratory mechanical '

analyses were performed on construction fill material to obtain relative indications of the ability of the various subsurface zones to transport water (Ref. 3).

The red silty sand fill material was well graded, containing about 45 percent passing a No. 200 l sieve. The well-graded nature of the fill suggests a very low permeability, probably ranging between 104cm/s and 10 cm/s. The ash fill material, however, is believed to have substantially greater permeability than the red silty sand fill. Actual permeability values for the ash fill are l unavailable: friable particles may heve been altered by the mechanical analysis technique.

l The licensees reported that, in general, the construction fill and boulder laryers were less permeable than the bedrock. Tests indicated that the boulder layer acted as a barrier or i confining layer to the flow of groundwater between the construction fill and the bedrock, l: essentially isolating the shallow groundwater from the deeper, bedrock groundwater. The permeability of the bedrock varied with depth. Test results indicate rock permeability ranging from moderate values (about 1.06 x 10-8 cm/s) to negligible values (no flow recorded in the test sections). . The highest permeability was at the boulder layer-bedrock interface-probably a function of the weathered, fractured nature of the top of the bedrock. Based on test borings, the L licensees suspect that other zones of comparatively high permeability may be present in the L bedrock.

Groundwater is measured at depths of about 1-1.5 m (3-5 ft) below the surface in the immediate site vicinity. Groundwater level observations in test borings also indicate a groundwater gradient of 3-4.5 m (10-15 ft) over e distance of 180-240 m (600-800 ft) from the site to the river.

Groundwater movement within the bedrock beneath the site is predominantly controlled by fractures in the bedrock. Groundwater also moves within the spaces (bedding planes) between the individual rock layers of the bedrock. The direction of groundwater flow is influenced by the

_ i orientation of these fractures and bedding planes.

A hydrogeologic investigation conducted in 1992 identified the specific orientations of the two dominant fracture pattems and of the bedding planes. One fracture pattem trended northeast-southwest and dipped (tilted) moderately to the northwest. The second fracture j pattern trended northwest-southeast and dipped steeply toward the southwest. The bedding

planes trended northeast-southwest and dipped moderately toward the southeast.

! 2.4 Geology The SNEC site lies ir, the Appalachian Highlands in the Ridge and Valley physiographic province. ,

This province compnses attemate successions of narrow ridges and broad or narrow valleys trending generally toward the northeast. This is a region of attemating hard and soft sedimentary

. rocks that have been severely folded by lateral compression into a series of anticlines and synclines. The ridge is of Tuscarora quartzite, and small amounts of Pleistocene gravel and recent alluvium are found along the river. Most of the area is underlain by strata from the Upper Devonian age. Although coal is mined in the general area of the site, no coal has been . reported l

to lie beneath the site, nor has the site been undermined. The ridges immediately to the northwest of the site rise to a height of 400 m (1300 ft), and ridges to the southeast rise to 460 m (1500 ft), The elevation of the site is about 247 m (811 ft) above MSL.  !

Soil Descriotion: Split-spoon samples collected during the hydrogeological investigation (Ref. 7) and samples from hand-dug pits indicate that the surficial soil in the vicinity of the CV contains 2-3 l

J

.)

ff SNEF Srrn DesenteT,oss two types of construction backfill: (1) well-graded reddish, sitty, fine-to<,oarse sand with some fine to medium gravel and (2) a mixture of well-graded ash and cinders. Both of these fill materials were deposited during station construction. The depth of the fill generally ranges from 1-2 m (3-6 ft), altho' ugh the fill may be deeper at those locations where construction excavation took place.

Underlying the fill materials is a layer of boulders. This layer is generally 1.2-1.8 m (4-4 ft) thick and separates the fill material from the top of the bedrock. The material making up the boulder  !

matrix is a silty clay. The sitt and clay appear to be localized in the boulder layer and have not been found in the fractured bedrock below (Ref. 7).

S.edrock Geoloov: The Pennsylvania Geological Survey (PaGS) has identified the bedrock underlying the facility as " marine beds" of the upper Devonian age. The PaGS designated this bedrock as the "Foreknobs Formation," but the unit has also been called a lower member of the

" Catskill Formation." The bedrock is composed of multiple layers of red and green siltstone and sandstone (also identified as gray to olive brown shales, graywackes, and sandstones). Depth to bedrock at the site is generally about 2.4-3.7 m (8-12 ft) below the surface (Ref. 8).

In 1981, during the licensee's hydrogeologic investigation, many bedrock outcrops were examined throughout the region. These outcrops substantiate the premise that the plant site is located on the westem limb of a major syncline that strikes generally northeast, dipping to the east. Some minor intemal folding is present within various bedding members, though the overall dip of the major structure is to the east. An understanding of the bedrock orientations and fracture patterns of these Devonian rocks is important for ascertaining the flow directions of groundwater within the bedrock (see Section 2.3).

2.5 Other Environmental Features Historical: The SNEF site and adjoining PENELEC property do not contain any known historical or archaeological areas. The project site was disturbed during construction of the SNEF and a nearby coal-fueled power plant, which preceded the SNEF. j Endancered Soecies: There are no known endangered or threatened plant or animal species on either the SNEF site or the adjacent PENELEC property.

i 1

{

l 2-4

b i - 3 SNEF Decommissioning Plans 3.1 Preliminary Activities All fuel was removed from the SNEF CV in 1972 and was retumed to the U.S. Atomic Energy Commission (AEC) at its Savannah River Plant in South Carolina (Ref.1). As a result, the licensees are no longer responsib!e for the facility's spent fuel. In addition, the control rod blades l and superheated steam test loop have been shipped off site. After fuel removal, all equipment, tanks, and piping extemal to the CV were removed. The buildings and structures that supported reactor operations were partially decontaminated during the 1972-74 time frame (Ref.1).

In 1986, after required approvals had been obtained, about 800,000 L (210,000 gal) of water that had seeped into the lower levels of the reactor support structures / buildings were discharged into the Raystown Branch of the Juniata river (Ref.1).. The water was very slightly contaminated by radioactive material leached from contaminated surfaces within these structures. i t-

! Final decontamination or removal of reactor support structures and buildings was performed in 1987,1988, and 1989, in preparation for demolition (Ref.1). Included in the decontamination were the Control and Auxiliary Building, Radioactive Waste Disposal Facility (RWDF), Yard Pipe Tunnel, and Filled Drum Storage Bunker. In addition, the Refueling Water Storage Tank was remcved. Upon acceptance of the final release survey by the NRC, these structures were j vemolished in 1992.

The Saxton Soil Remediation Project was completed in November 1994 (Ref. 9). Soil within the site perimeter found to be contaminated with radioactive material was excavated, packaged, and shipped to a licensed radioactive waste disposal facility. This program successfully reduced ,

radioactive contamination levels.

More recently, the licensees performed a site characterization study to determine the radioactive contamination and activation levels at the SNEF (Ref. 26). The results of the study and subsequent radiological surveys indicate that the following systems are contaminated or activated:

l

• Main coolant system; j l

• Pressure relief system charging and volume control system; l-

• Component cooling system;

• River water system; .

• Steam /feedwater system; l

• Cocling, heating, and ventilating system;  !

• Purificstion system;

• Rosin sluice system;

• Shutdown cooling system;

• Safety injection system;

• Sample system,

-* Storage well cooling system;

• Vents and drains system;

• Waste liquid system; and

• Septic tanks.

3-1

[ 1

- o, SNEF Deconsessommo Pt.Aus 3.2 Plant Radioactive Materialinventory 3.2.1 Activation Levels Under a contract with the licensees, TLG Services of Bridgewater, Connecticut, performed an analysis of the residual neutron activation products remaining in the SNEF's reactor pressure vessel, intemais, and surrounding lead, structural steel, and concrete (Ref.1). The analysis indicated that approximately 53.7 TBq (1452 Ci) of neutron activation products, mainly iron-55 (1.4 TBq or 37 Ci), cobalt-60 (22 TBq or 595 Ci) and nickel-63 (30 TBq or 811 Ci), would be present at 24 years after shutdown (until July 1996) in the reactor vessel wall and clad, intomats, insulation can, and support can (Ref.1). These results do not include activity from intomal and extemal surface contamination, which is normally a small percentage of the total activity for l

these components. Table 3.1 presents a breakdown of activation products within the SNEF reactor vessel and intemals.

Neutron activation analysis of the ENEF structure surrounding the reactor vessel indicates that a significant amount of concrete must be removed. Preliminary estimates indicate that -

approximately 310 m'(11,000 ft') will require removal. Concrete is removed because of both contamination and activation. The estimated total activity associated with the concrete is about 18,500 MBq (0.5 Ci).

3.2.2 Area Dose Rates and Contamination Levels l SNEC obtained site-specific radiological and environmental data in 1995, during implementation

! of the SNEC Site Characterization Plan (Ref.10). Typical results are presented in this section.

A site plan is provided in Figure 3.1, and Figures 3.2 to 3.7 depict radiation fields and l

1 l

Table 3.1 Calculated Activation Products Within the Reactor Pressure Vessel (July 1,1996)

Activity Location Megabecquerels Curies .

Core Baffle 23,800,000 643 l

Thermal Shield 4,920,000 133 Reactor Vessel Clad and wall . 330,000 9 Reactor Insulation Can and Support Can 150,000 4 l Lower' Core Guide Blocks 6.770,000 183 Lower Core Plate 8,400,000 227 i Lower Support Shroud Tubes and Tie Rods 370,000 10 Lower Core Barrel 2,150,000 58 Upper Core Plate 6,100,000 165 Upper Core Barrel 740.000 _ 20 j Total 53,730,000 1452  !

3-2 l i

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Figure 3.1 Saxton Site Plan l

l 3-3 l

1

. i

. 1 i

b. .') .

l SNEF Dscomissaommo PLANS l l

3 i j ,

l

.I l Average Contaminsbon Levels and General n

Area Exposure Rates q "] l

a. _ Poiar Orane L ]

1. ~

$6 s  !

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'f <1000 come100 eq cm

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suup Figure 3.2 SNEF Containment Vessel, Sectional View (Looking West) 3-4

t)

SNEF Dae==== ammo PLAus

\

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Average Contamriation Levels and General AREA 7 2 Nea Essa Rates  !

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j Figure 3.3 SNEF Containment Vessel, Sectional View (Looking North) 3-5

.)

SNEF DscoMascuno PLAss Average Contamination Levels and General Area Exposure Rates

<1.000 dom /100 se em

(-

Q- ~

-1.000 dom /100 so em ce Steel Platform 9 %O i ,

e' d 7 ....., I Naten $81 l 7

! l

<0.2 mRen GA  !

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58 2 Floor a . i Fuei Bridge

~ Walls SB 6 & SB 7 l

% <1.000 com/100 sa cm Relocated to Here l S s.7 v\p Ss-s l Floor Elev 818 d' ., l

,) Shiele Blocks Open dI S3-3 j

Hatch _m _-

g Ss 4 SB 5

<0 2 mR/n G A

<1000 cpm /100 sc em Floor [

m h Upper Dome (inside)

-1.000 dpm/100 sq cm W

Polar Crane u

j .l.300 dpm/100 sq cm avg Emergenc Personnel Aer Lock 1,700 opm/100 so cm '

Figure 3.4 SNEF Containment Vessel, Plan View (812 ft level) 3-6

5 SNEF Denaamammanamo Pt.Aus Average Contamination Levels and l

General Area Exposure Rates p -3.500 com/100 se em (

, OE h;,(;,.9 g e "-  :..,* ,

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Floor m ,s.sr.... s egyA: %c)d. r,$1.jy}d,a,

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I Figure 3.5 SNEF Containment Vessel, Plan View (795 ft Level) 3-7

e SNEF Deconnesseomino PLAms Average Contamination Levels and General Area Exposure Rates qy N.TNiEjyig;jJ.ji; l ..... A,t.t,

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? hi k!j,l) r' 200.000 dpm/

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+ .,'*.  ? Storage Well Floors j f%:; ::![', ..~" 0:!; ;L '

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Figure 3.6 SNEF Containment Vessel, Plan View (781 ft Level) 3-8

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e.

SNEF Deconsessiosmo PLAus Average Contamination Levels and General Area Exposure Rates

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Figure 3.7 SNEF Containment Vessel, Plan View (765 ft level) l 1

3-9

1 e

l SNEF DscoMhessioso PLAms 1

contamination levels for the CV. General area dose rates varied from near background to )

approximately 2 mSv (200 mrem) per hour. The highest dose rates measured were i approximately 80 mSv (8000 mrem) per hour in contact with the boric acid demineralizer and l 50 mSv (5000 mrem) per hour near the reactor vessel. Removable alpha surface contamination I measured on smears ranged up to 1400 disintegrations per minute per 100 cm2 (Area 6). I l 3.3 Decommissioning Plans l

The licensees have selected the DECON attemative for decommissioning the SNEF. Immediate l l dismantlement places the SNEF in a stable, secure, and decontamirated condition as soon as yactical (Ref.1). A tentative schedule for decommissioning actions at the SNEF is provided in Figure 3.8 (Ref. 24).

1 ACTIVITY YEAR i

1997 1998 1999 2000 2001 01l02 03 l 04 01 02 ! 03 04 01 02 03 04 01 02 03 04 01 02 l Preparation Phase Licentees' Sta[.1 (Thru 01 1998)

Nor pov1 N (finished 01 1998)

Operations Phase l Ooerations Activities  ;

Prenare Iaron Cnmnnnnnt Pomnval l l l taran Comoonent Pemoval 1 l Dermittioninq nneratinne Survey Phase l Submit ticente Termination Plan . l l

Final Survev  !

NRC Confirmatory Survey t1cenee Tarm,natinn - l l j Site Restoration Phase final Restoration = l l

l Figure 3.8 Time Lines for SNEF Decommissioning l

3.3.1 Preparation Phase Based on the results of the 1995 site characterization study, the licensees have performed conceptual engineering and planning to determine the most advantageous approach for decommissioning activated and contaminated systems and structures. They have committed to incorporate the following considerations into all conceptual and detailed engineering and planning:

l l 3-10

=

SNEF n========a plans

• Regulatory guidance,

+

Maintenance of occupational radiation exposure at levels that are as low as is reasonably achievable (ALARA),

• Management of low-level radioactive waste,

• Industrial safety,

• - Environmentalimpacts,

. Costs, and

• Schedule considerations.

In addition, the licensees have committed to use field-proven and state-of-the-art dismantlement techniques and to perform all SNEF decommissioning activities under a quality assurance program that meets all applicable NRC requirements (Ref.1). The NRC staff is reviewing the quality assurance program as part of the licensees' license amendment application.

3.3.2 Operations Phase The licensees have stated that all decommissioning activities will be conducted in accordance with the SNEF's Radiation Protection Program. The objective of this approach is to continue ensuring the radiological safety of the workers, visitors, and the general public during the decommissioning phase, Through its inspection function, the NRC will periodically inspect and monitor the effectiveness of the Radiation Protection Program, in addition to providing oversight of decontamination and dismantlement activities.

The scope of work for completing DECON decommissioning at the SNEF includes the removal of activated or contaminated systems and structures. Removal (or dismantlement and removal) will be performed for those systems and structures that do not meet the site release criteria (Ref.12) discussed in Section 3.3.3. Dismantlement and removal of pipes and metals will be performed using shears, portable band saws, diamond wire saws, abrasive wheel cutters, milling machines, and other suitable tools. Scabblers and CO 2blasters are being considered for removing fixed -!

contamination from concrete. The licensees will select the best attematives as part of their l detailed engineering and planning efforts. The use of water will be minimized due to the cost and time required to dispose of contaminated water (Ref.1).  ;

l The major tasks associated with the SNEF decommissioning of contaminated systems are l described below i

(1) A DSF adjoining the CV was built to segregate and package waste. Construction of the DSF has been completed under an NRC-approved license amendment.

(2) Removal and packaging of the majority of the asbestos insulation from affected systems j 1

and components have been completed under an NRC-approved license amendment.

(3) All piping and other portions of systems that are not at or below the release criteria will be l dismantled and removed, including all piping 8 cm (3 in) in diameter and smaller that cannot be readily characterized. The licensees intend to ship this small-bore piping to a 3-11 I I

SNEF neum .--a ptans licensed vendor for further decontamination, radiological surveys, segregation, and ultimate disposal.'

(4) The reactor pressure vessel will be removed as follows (Ref. 24):

• A temporary reactor vessel lay down facility will be installed within the site boundary. This facility will provide weather protection for the reactor vessel and will permit decommissioning personnel to conduct packaging operations.

• - Piping and instrumentation lines attached to the reactor vessel will be cut.

Interferences such as steel plates and concrete shield blocks located directly above the reactor vessel will be removed to permit a vertical lift of the reactor -

vessel.

• An opening in the steel CV dome of sufficient diameter will be made above the reactor vessel to permit its removal in one piece.

• The reactor vessel will be separated from the reactor vessel support assembly and placed on temporary support beams on the 247 m (812 ft) level.

• The vessel will be grouted and openings created by cutting operations will be sealed to preclude the release of surface radioactive contamination.

• Appropriate radiological contamination and airbome control measures will be implemented to prevent the spread of such material before the reactor vessel is removed from the CV.

• Using a crane, the reactor vessel, including the attached " insulation can," will be lifted from the temporary support beams, from the CV and placed onto the lay down area and shipping container.

• The crane will lift the removed section of the CV dome back into place, and the  :

weather-tight integrity of the opening will be restored. l (5) The steam generator and pressurizer will be removed as follows (Ref. 24): )

l

• All process piping attachments to the vessels will be cut and attached i instrumentation will be removed. The components will be grouted and openings l created by cutting the attached piping will be sealed to prevent release of contamination to surrounding areas during handling.

• Removal of these vessels will be performed by using temporary supports, cutting l l the four hanger rods, and then moving the vessels horizontally to be lifted up  !

l through the removable hatch. l l

• Removal of the steam generator and pressurizer vessels from the CV will be through a separate opening cut in the CV dome similar to the reactor pressure vessel opening.

(6) The CV at the SNEF is a circular steel structure that is approximately 33.2 m (10E ft) tall and 15 m (50 ft)in diameter, with approximately half of the structure extending below 3-12

f c

q

~

SNEF Dscoheessionimo Pl.ANS

. Ma The CV is subdivided into a reactor compartment with storage well, primary I I

,imrv;tment, auxiliary compartment, and operating floor. These areas are separated by ecocrete walls, floors, and ceilings. In addition, the below-grade portion of the CV is lined with concrete. The major tasks associated with dismantlement of the CV are as follows:

• Radiological characterization has been performed to determine the depth of penetration of activation and contamination into the concrete structures and to verify the extent of contamination on the remaining surfaces of the building.

l

• Systems and components will be removed from the CV.

• Contaminated concrete and other intemal building structures will be removed (e.g., support steel, grating, etc.).

l l - The CV shell will be decontaminated, as needed.

• A final site release survey will be performed to ensure that all contamination above the release criteria has been removed.

l l - An NRC confirmatory survey will be performed. l

- Remaining portions of the intemal CV structure, exclusive of the concrete, will be removed (the remaining concrete will be demolished, as necessary, to j accommodate backfill operations).

• The CV will be removed to a depth of 0.9 m (3 ft) below grade, and the void will be backfilled.

3.3.3 Survey Phase The licensees have committed to decontaminate the SNEF in a manner that complies with the l radiological criteria promulgated after submittal of the environmental report-that is, as specified l in 10 CFR Part 20, Subpart E, Radiological Criteria for License Termination (Ref.12). The purpose of the final radiation survey is to demonstrate that the SNEF meets the final site release l criteria in accordance with the above-cited regulations. The radiological criteria for unrestricted use in 10 CFR 20.1402 are that a site will be considered acceptable for unrestricted use if the residual radioactivity that is distinguishable from background radiation results in a Total Effective Dose Equivalent to an average member of the critical group that does not exceed 25 mrem (0.25 mSv) per year, including that from groundwater sources of drinking water, and the residual ,

i radioactivity has been reduced to levels that are ALARA. Determination of the levels which are i l ALARA must take into account consideration of any detriments, such as deaths from l l transportation accidents, expected to potentially result from decontamination and waste disposal. t l The critical group as defined in 10 CFR 20.1003 means the group of individuals reasonably l l expected to receive the greatest exposure to residual radioactivity for any applicable set of .

circumstances (Ref.12). To demonstrate compliance with these criteria, the licensees must i determine the contributions to dose from direct exposure and from ingestion and inhalation for  !

credible and likely exposure scenarios. The licensees will obtain the following radiological data  ;

1 to support their calculations:

• Surface activity levels (removable and fixed),

• Deposited activity in volumetrically contaminated material, 3-13

, I l

SNEF Deconnessommo Pt.aus

• Direct exposure rates, a Radionuclide concentrations in soil and ground / surface water, and

+

Total site inventory of residual radioactive material.

The cata must be accurate and represent actual site conditions, while accounting for the considerable variability of natural and enhanced background radiation. The final radiation survey l will be carried out under the control of a survey plan that will incorporate sound statistica!

l methods of data collection and analysis. The final radiation survey plan, which has not yet been L submitted, will specify the following:

l

• Equipment, methods, and procedures used to determine residual radioactivity and l

j extemal dose rates; Methodology for data review and analysis, including the application of appropriate statistical and pathway modeling; a Physicalcontrolof samples;

• Training and qualification of survey personnel;

• Data quality objectives; and

• Quality control / verification.

The final survey plan and related implementing procedures will follow the guidance in the l applicable standards at the time of the final survey. At present, that guidance is contained in NUREG/CR-5849 (Ref.13). Guidance needed to implement the newly published release criteria in 10 CFR Part 20, Subpart E, will be used when it becomes available. Note that this guidance is under development and is available in preliminary form as draft NUREG-1505, NUREG-1506, and NUREG-1507 (Refs.14,- 15, and 16). The licensees have also committed to comply with all applicable regulations of the State of Pennsylvania and localjurisdictions.

i 3.3.4 Site Mestoration Phase The NRC will conduct an independent verification to confirm that decontamination activities on or near the SNEC site have removed all radioactive material attributable to SNEF operations to levels specified by the site release criteria. Site restoration activities will begin only after the NRC has confirmed that decontamination has been satisfactorily accomplished and has terminated the license. These activities are scheduled to be completed by late 2000. As the j SNEF site is wholly contained within PENELEC property, the licensees anticipate that the ownership of the property on which it is located will revert back to PENELEC after the NRC l license is terminated. The licensees have not committed to the exact method of site l restoration-that is, whether the site will be retumed to preconstruction status or converted to other industrial uses. It should be emphasized, however, that nonradiological site restoration activities are not regulated by the NRC.

3-14 I

1

, j 1

l 4 Environmental Effects of Proposed Action 4.1 Effects on Human Activities The licensees estimate that no more than about 40 personnel will support decommissioning activities at the SNEF site at any given time. Due to the small size of the work force, the impact of decommissioning on local housing, services (e.g., schools), and transportation should be slight.

The SNEF is located next to an electrical substation. The substation and electrical power transmission lines in the vicinity of the site will be unaffected by decommissioning activities.

l 4.2 Effects on Terrain, Vegetation, and Wildlife l

No known endangered or threatened plant or animal species have been identified on or make use of the SNEF site. That portion of the 0.465-hectare (1.15-acre) site not occupied by facility structures is composed primarily of open grassland that does not provide good habitat for wildlife. Further, no endangered or threatened species have been identified on the adjacent PENELEC property, which is essentially composed of open grassland with scrub vegetation and  !

trees growing along property boundaries. Areas that have remained undisturbed since the coal-fired station was demolished have generally reverted to open fields or woodlands and thus provide a better habitat for wildlife. l l

The proposed decommissioning activities will take place on previously developed areas of the site or on adjacent open areas of the PENELEC property. These activities include construction of temporary support facilities such as office trailers, the DSF needed for segregating and packaging of waste, and the borrowing of fill material needed to backfill the CV void. Those areas of the site left in their natural state will not be disturbed by these activities. Hence, there will be no effect on the existing terrain or vegetation in previously undeveloped areas of the site.

During the removal and demolition of structures at the SNEF, waterfowl and other wildlife that sometirnes make use of adjacent areas will be disturbed and/or displaced by demolition activities. Demolition activities in the area will last a very short period of time, however, and will be limited to as small an area as necessary, thereby limiting their potential impact on the environment.

4.3 Effects on Adjacent Waters and Aquatic Life Proposed decommissioning activities at the SNEF are not expected to have any adverse effects on adjacent surface waters or associated aquatic life. All or most of the decommissioning work l will be done in previously developed areas of the site. Since this work will involve minor construction activities to remove and demolish remaining structures, a comprehensive soil l erosion and sedimentation control plan will be implemented to limit soil disturbance and potential

! siltation of the river. The content and implementation of the control plan are designed to meet l the requirements of Pennsylvania Code 102.4.

4.4 Effects of Released Radioactive Material As part of routine decommissioning operations, limited quantities of radioactive material may be released to the environment in liquid and airbome effluents. An effluent control program has 4-1

}

EuvinosusuTAL Errects or PnorosED Action been implemented to ensure that the release of radioactive material to the environment is minimal and does not exceed established limits. Federal effluent limits are set at low levels to protect the haalth and safety of the public. The licensees have committed to conduct operations l l

in a manner that holds radioactive effluents to small percentages of these Federallimits.  !

Releases during decommissioning the SNEF will be a small fraction of the limits in 10 CFR l Part 20.

The ODCM supports the SNEF's technical specifications and implements its radiological esiluent controls. The ODCM contains the controls, bases, and surveillance requirements for liquid and l gaseous radiological effluents. This document also describes the methodology used to calculate l instrumentation alarm and trip set points for monitoring liquid and gaseous effluents. The ODCM follows the methodology and models suggested by NUREG-0133, NUREG -1301, Regulatory Guide 1.109, Revision 1, and the Radiological Assessment Branch Technical Position, Revision 1, for calculating offsite doses caused by plant effluent releases. The licensees have applied simplifying assumptions in this manual, where applicable, to provide a more workable document for meeting radiological effluent control requirements.

l Airborne Radioactive Effluents i After termination of operations at the SNEF in 1972 and prior to dismantlement of all radiological waste systems, radioactive gas was retained long enough for decay of the principal radionuclides to take place, then it was released to the atmosphere. Hence, radioactive gases that require processing are no longer present.

The licensees have installed a new ventilation system because the original, permanent plant I ventilation systems are no longer functional. This system performs the following functions:

• Provide for filtration and quantification of airbome radioactive particulate material in effluents,

• Provide for worker comfort by reducing CV temperature extremes,

• Reduce the potential for confined-space restrictions by providing sufficient air volume changes,

• Reduce CV interior radon concentrations that build up from naturally occurring sources and accumulate in the CV, and

!

• Provide sufficient face velocity at the CV/DSF opening to meet requirements for l maintaining the air release pathway for the CV via the monitored ventilation system

' ]

exhaust.  !

l The ventilation system consists of one exhaust fan drawing air from the upper and lower portion I of the CV. The exhaust fan is a 184-M' per min (6500-CFM) centrifugal unit fitted with pre-filters and high-efficiency particulate air (HEPA) filters for the removal of airbome particulates in the exhaust air The exhaust fan and filtration units are located outside the CV on the north side and are ducted to the CV using the existing 43-cm (17-in) CV ventilation penetration. The flow path through the DSF is from the Decommissioning Support Building (DSB) wall louvers (or roll-up doors), through the DSB, through the material handling bay, through the planned CV/ material handling bay opening, and across the CV operating floor to be exhausted through exhaust registers attached to a plenum that runs from elevation 253 m to 247 m (832 ft to 811.5 ft). A 4-2 l

l

EnvmoshumTAL ErrscTs or Pnorosso Action i

duct connection is provided inside the CV on the inlet plenum to allow connection of a flexible duct hose for local ventilation needs. The plenum connects to the existing 43-cm (17-in) CV ventilation penetration that is provided with an isolation damper and is connected to the filtration unit. Air flows from the filtration unit to the fan and is exhausted via a short stack. To indicate and monitor radioactive releases, a radiation monitor with iwkinetic sampling has been installed ,

downstream of the HEPA filter unit. The filtration unit was designed, constructed and tested in accordance with applicable American National Standards institute standards.

The CV atmosphere will also be monitored by portable air samplers, if necessary, if decommissioning activities require additional control of airbome radiological contamination, portable HEPA filtration units, including those built into vacuum cleaners, will be used. Effluent-monitoring instrumentation will be used to identify discharges of airbome effluent as required and to demonstrate compliance with the SNEF ODCM limits as promulgated by applicable regulations.

The licensees calculated that the dose to the public from decommissior@g activities other than transportation of radioactive waste will be about 0.14 person-mSv (1/ person-mrem). This includes about 0.1325 person-mSv (13.25 person-mrem) from the rar pathway and about 0.005 person-mSv (0.5 person mrom) from the liquid pathway from the .elease of groundwater in the pipe tunnel discussed in Section 4.5. The GEIS (for the reference test reactor) estimated that the dose to the public from decommissioning activities other than transportation of radioactive waste will be negligible (less than 1 person-mSv (100 person-mrom)). The reference test reactor is used for comparison because its operating power level (60 MWT) and usage (research) is a  ;

better match to the SNEF than the reference pressurized water reactor (PWR) (power level of 3500 MWT used for power production). Also, decommissioning activities at the SNEF in the  ;

1970s are specifically discussed in NUREG/CR-1756, " Technology, Safety and Costs of  !

Decommissioning Reference Nuclear Research and Test Reactors" (Ref. 25).  !

Effluents from concrete removal activities, pipe segmentation, and structural segmentation and j CV liner decontamination activities were considered by the licensees with the first two activities ]

contributing about 97 percent of the population dose. Removal of the reactor vessel, steam j generator and pressurizer will not contribute significantly to effluent releases because they will ,

be handled as complete components. The licensees used information from the site  !

characterization report (Ref. 26) to determine the radioactive material available for release and  !

applied conservative assumptions from NUREG/CR-1756 (Ref. 25) for performance of the ventilation system. The licensees used methodology from the ODCM and Regulatory Guide 1.109 (Ref. 27) for dose calculations. Conservative assumptions for atmospheric dispersion factors for a distance of 16 km (10 mi) from the SNEF were used in the calculations. The release of particulates beyond this distance will be insignificant. However, the licensees performed a very conservative bounding calculation for a large population to a distance of 80 km l (50 mi) from the SNEF which resulted in a maximum additional population dose of 0.3 person-mSv (30 person-mrem).

Liquid Radioactive Effluents The generation of radioactive liquid wastes may occur during the decontamination and dismantlement of the systems and structures at the SNEF. However, the licensees have  !

indicated a desire to keep the use of water to a minimum during decommissionti,4 Because of this, with the exception of slightly contaminated groundwater in the pipe tunnel discussed in ,

Section 4.5, no specific sources of liquid wastes have been identified at this time. If liquid waste is produced by activities at the SNEF, the resulting liquid waste stream will be processed using )

4-3

i r

W.

l EnvetosMsNTAL ErrscTs or Pnoposso Action l techniques that meet ALARA goals and the requirements of the regulations. Plant equipment L

that had been used to process liquid radioactive waste was removed during earlier demolition j activities. Liquid radioactive wastes generated during decommissioning will be processed using l

temporary systems supplied by GPU Nuclear or by experienced vendors and contractors, as '

appropriate. Temporary waste treatment systems (filtration units or domineralizers) will be connected to tanks for storage of processed water prior to discharge. These systems may include temporary ventilation with filtration for airbome radiological contamination control. Once it has been verified that the stored processed water meets the discharge limits specified in the ODCM and allowed by 10 CFR Part 20, the water will be released. Effluent-monitoring instruments will be used to identify discharges of liquid effluent and to demonstrate compliance i with ODCM limits and applicable NRC regulations.

Some of these processes will probably generate a product that will require further processing for disposal as solid radioactive waste as discussed below.

! With the exception of release of groundwater discussed in Section 4.5, no other significant liquid l releases have been identified by the licensees. Atmospheric releases of radioactive material to the environment is the major source of radiation to the public from routine decommissioning activities, it is expected that any public dose from liquid releases will be less than doses from l the air pathway discussed above.

Based on its review of the licensees' dose assessment methodology, parameter definition, and assumptions used in the evaluation, the staff agrees with the licensees' determination of public dose from routine decommissioning activities that result from releases from the SNEF to the air and liquid pathways (shipment of radioactive waste is discussed below). The staff has determined that this dose will be bounded by the values in the GEIS (for the reference test reactor) and that dose to the public from decommissioning activities that result in releases to the air and liquid pathways will be negligible (less than 0.001 person-Sv (0.1 person-rem)).

Transoortation and Disoosal of Solid Radioactive Waste The licensees have developed a formal process control program that specifies the formulet, sampling requirements, analyses, tests, and determinations necessary to verify that sol %fied radioactive wastes will be processed and packaged in a manner that complies witF ( .; 10 CFR Parts 20,61, and 71; (2) burial ground requirements; and (3) Department of Tonsportation (DOT) regulations goveming the disposal and shipment of solid radioactim wastes. In addition, the licensees have committed to comply with applicable requirements of the Pennsylvania Department of Environmental Protection (PaDEP) under the Natic.al Pollutant Discharge Elimination System (NPDES).

Transportation of radioactive waste can cause both occuNtional and offsite radiation exposures.

Occupation exposure is included in the projections discussed in Section 4.6. The estimated cumulative radiation exposure to the public is the swn of the small individual radiation exposures that are assumed to occur when members of the public are in the vicinity of a low-level radioactive waste shipment for brief periods of time. Dose limits on shipments must comply with DOT regulations specified in 49 CFR Part 173, Subpart 1. Types of cargo, packaging, and limits on inventories shipped will be bounded by DOT regulations for both truck and rail shipments.

The licensees estimate that 580 m* (20,500 ft') of low-level radioactive waste will be shipped from the SNEF site to low-level radioactive waste burial sites in South Carolina or Utah during decommissioning of the CV. In addition, activities already completed at the site have resulted in 4-4

i I-EuvinochENTAL ErrscTs or Pnorosso Action the shipment of waste off site. The licensees estimate that the total waste shipped off site for all decommissioning activities will not exceed 2700 m3 (95,000 ft8). By comparison, the GEIS 3

(Ref. 6) estimated that 4930 m (174,000 ft') of low-level radioactive waste would be shipped by truck from a decommissioned reference test reactor to a licensed disposal facility. The quantity of radioactive waste generated by decommissioning activities at the SNEF is bounded by the j GEIS .

The licensees have performed a conservative calculation that shows about 0.3 acres of land will be used for waste disposal. This is bounded by the estimate of 0.5 acres of land required for disposal of waste from the reference test reactor in the GEIS.

The estimate of the radiation exposure to the public in the GEIS associated with shipments of radioactive waste from the reference test reactor is 0.022 person-Sv (2.2 person-rom). The licensees calculate that the exposure to the public from shipments from the SNEF are approximately 0.0086 person-Sv (0.86 person-rem). The licensees have assumed that 100 shipments are needed and that each shipment contains radioactive material that results in the maximum exposure rates allowed by DOT regulations. This is a very conservative assumption because many of the waste shipments from the SNEF will consist of concrete with a very low dose rate per shipment. The licensees used assumptions and dose values from NUREG/CR-1756 (Ref. 25) in the calculation of radiation exposure to the public from waste shipments.

These assumptions and dose values are applicable to shipments from the SNEF. The staff has determined that the projected cumulative radiation exposure to the public from the transportation of radioactive waste is clearly bounded by the GEIS.

4.5 Effects on Groundwater I

4.5.1 Groundwater Flooding of Underground Structures  !

The service tunnel that surrounds the SNEF CV contains significant quantities of groundwater.

The tunnel is a below-grade concrete structure located as shown in Figures 3-1,3-2, and 3-3.

The tunnel ceiling is at grade, approximately 247 m (811 ft,6 in) above MSL, whereas the floor is approximately 245 m (805 ft) above MSL--or about 1.8 m (6 ft) below grade. Groundwater levels vary at the site, depending on season and weather, but they generally average about 246 ,

m (807 ft) above MSL (Ref 7). Hydraulic pressure forces groundwater into the tunnel through the construction joint between the tunnel floor and the CV shell. Water levels in the tunnel have been observed to f!uctuate considerably with groundwater changes. During periods of severe drought, the tunnel has been dry; at other times, the water level has reached the ceiling. i Contamination on the tunnel's interior concrete surfaces, principally by cesium-137, has leached j into the water and led to low-level concentrations of radionuclides.

In 1986, a comparable situation existed in the other below-grade structures that have since been demolished at the site. Approximately 795,000 L (210,000 gal) of groundwater ,

contaminated with very low concentrations of radionuclides were removed and discharged to the Raystown Branch of the Juniata River (Ref.17). A similar process would likely be used to remove groundwater from the service tunnel. A bounding calculation has been performed to determine the maximum possible dose to a member of the public if this water were to be discharged under worst-case conditions (Ref.18). Using the maximum batch release flow rate ,

and historic minimum river flow, the maximum organ dose would be approximately 7 x 104 mSv l 4

(7 x 10-8 mrem), whereas the maximum whole-body dose would be about 4.5 x 10 mSv (4.5 x 10-3 mrem). These levels are significantly below applicable limits.

4-5 i

9 EnvutosumuTAL ErrscTs or Paorosso Action The licensees also performed a calculation of population dose from this release. It was assumed that 100,000 fishermen consume fish and water from the Juniata River or Raystown Lake. The

~

licensees calculated a population dose of less than 0.005 person mSv (0.5 person-mrom). This dose is included in the total population dose discussed in Section 4.4. All releases from the SNEF would comply with the SNEF ODCM and with applicable regulations.

The primary mechanism for a release of radionuclides to the groundwater would be an accidental spill of radiologically contaminated water. Temporary systems used for processing water will be designed to reduce the probability of spills to the ground. Written procedures and work instructions will be used to reduce the potential for spills and to mitigate potential spillage in a timely manner.

If a spill of radiologically contaminated water does occur, groundwater at the facility should not be adversely affected. Fission and activation products in the water (primarily cesium-137, cobalt-60, and small quantities of transuranics) will be adsort>ed onto the soil as the water percolates through the ground. Numerous studies focusing on retention of these radionuclides by soil (Ref.19) show that they are typically retained in the first 10-30 cm (4-12 inches) of soil.

As a result, they are not immediately available for transport into the groundwater. Should such a spill occur at the SNEF, the affected soil would be sampled and analyzed for radionuclide content. Soil containing quantities of these nuclides in excess of applicable release criteria (Ref.11) would be excavated and transported off site to a licensed disposal facility. As a result, these types of radionuclides would not find their way into the local groundwater.

The only radionuclide that could reach the groundwater from such a spill would be tritium, which is not retained by the soil. Current concentrations of tritium in the water at the SNEF are relatively low when compared with applicable release limits. The highest concentrations are found in the CV sump and range from 11 to 22 kBq/cc (3 x 10d to 6 x 10d pCi/cc). Since there is no source of tritium production at the site, the concentration of tritium will not increase with time.

In fact, if decontamination activities create radiologically contaminated water, the concentration of tritium in liquid wastes will decrease through dilution. The low concentrations of tritium in water, coupled with the finite nature of such a spill, will not appreciably affect tritium concentrations in groundwater at the site.

4.5.2 Groundwater Monitoring Program A hydrogeological investigation was conducted in 1992 to determine the direction of groundwater flow in the shallow aquifer of the SNEF site (Ref. 20). Eight overburden (shallow) groundwater monitoring wells were installed for this purpose. Groundwater elevation contour maps indicate that groundwater within the overburden soil flows west, toward the Raystown Branch of the Juniata River (Ref. 5). Additional information was gathered during the 1992 investigation for installing deeper, bedrock monitoring wells for reliably monitoring the CV with a minimal number of wells.

l Two bedrock groundwater monitoring wells were installed in 1994. These wells, designated l MW-1 and MW-2, were drilled to a depth of about 16.7 m (55 ft) and are located, respectively,  !

approximately 6 m (20 ft) northwest and 9.1 m (30 ft) west of the CV. A third well(GEO-9) was I drilled near MW-1 to a depth of 15.2 m (50 ft). A vertical piezometer was installed in this well in 1994 for the sole purpose of measuring bedrock groundwater elevation. The bedrock groundwater monitoring well locations were chosen with the intent of intercepting any groundwater contamination that might result from CV leakage. These locations were based on an inferred direction of flow of the bedrock groundwater, which in tum was based primarily on a j l

4-6  !

l l

ENVIRONMENTAL EFFECTS oF PaorosED AcnoN geological investigation of two dominant bedrock fracture patterns and bedding planes visible in three outcrops in the general vicinity of the site. The wells were drilled at an angle of approximately 25 degrees from vertical to increase the probability of intersecting each of the 1 dominant bedrock fracture pattems. '

Using the wells described in this section, a groundwater monitoring program has been conducted at the SNEF site since 1994. The location of each of these wellis depicted in Figure 4.1.

4.6.3 Groundwater Monitoring Results Contamination resulting from SNEF operations has been detected in only one of the shallow monitoring wells. The tritium concentration in well GEO-5 has ranged from less than detectable (7.4 Bq/L or 200 pCi/L) to a maximum level of 28 Bq/L (760 pCi/L). This concentration is less than 0.08 percent of the applicable concentration of tritium in water specified by 10 CFR Part 20 (i.e. , 10-2 pCi/ml) for assessing and controlling the d se to the public due to effluents, and less than four percent of the U.S. Environmental Protection Agency (EPA) limit for drinking water (i.e.,

20,000 pCi/L). Well GEO-5 was drilled to a depth of 4.5 m (15 feet), is located about 67 m (220 feet) east-northeast of the CV, and is the most distant monitoring well up gradient from the CV. The RWDF, a suspected source of the tritium contamination, was decontaminated and dismantled in 1987-89. Well GEO-5 is about 7.6 m (25 feet) west-southwest of and down gradient (the direction of groundwater flow) from the RWDF site. There are no other monitoring wells in that vicinity. Thus, the information available to confirm the RWDF as a source of tritium contamination is limited.

From a public health standpoint, tritium concentrations observed in the groundwater to date are insignificant. Currently, there are no known public uses of shallow groundwater in the vicinity of the SNEF. For license termination ,the licensees must meet the release criteria for the site which includes radionuclides in groundwater. This will be addressed before the final survey is conducted and the license is terminated.  ;

4.6 Occupational Radiation Exposure The licensees have determined the projected occupational radiation exposure for performing l decommissioning activities at the SNEF. The projected radiation exposure estimate for  !

decommissioning is 0.32 person-b (31.8 person-rem). The licensees

• past estimates for l occupation exposure for work that P 3 been performed have been reasonablely accurate. The l' GEIS states that occupation exposure for decommissioning the reference test reactor is 3.44 person-Sv (344 person-rem). The staff concludes that the occupational radiation exposure from l decommissioning the SNEF is bounded by the GEIS.

4.7 Effects of Released Chemical and Sanital/ Wastes l

During decommissioning, water from an existing groundwater well located on the adjacent PENELEC property will be the source for sanitary water. The use of groundwater for sanitary and drinking water is regulated by the PaDEP. If the groundwater well is used as a drinking water source, it may be necessary to provide nonradiological water treatment.

The licensees have indicated a desiro to keep the use of water to a minimum during decommissioning. No chemical radiological decontamination is planned, so the use of hazardous chemicals is not anticipated during the decommissioning process. Liquid discharges 4-7

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EnvmosuustAL ErrscTs or Pnorosso Action from the facility are regulated by the NPDES permitting system administered by the PaDEP. All liquid waste streams will be sampled, tested, and processed as necessary before discharge to ensure that effluents are in compliance with applicable PaDEP - NPDES permit limits.

Accordingly, the quality of the nearby water resources is not likely to be affected Holding tanks will be used during decommissioning for the collection of sanitary waste. These tanks shall be closely monitored and pumped out by a PaDEP-licensed contractor for offsite disposal at a licensed facility.

- 4.8 Effects of Nonradiological Material Asbestos Surveys for asbestos were conducted in the CV during May 1995. Bulk insulation samples were taken of various components, piping systems, and vessels throughout the containment building.

The quantity of asbestos requiring removal was estimated at about 32 cubic meters. The majority of asbestos has now been removed.

Both the EPA and the Occupational Safety and Health Administration have established regulations that apply to the removal of asbestos-containing material. These regulations include requirements related to notification, recordkeeping, handling, air emissions limits, and disposal.

All activities involving asbestos at the facility were conducted ir, accordance with Federal and State regulations. All asbestos found in the facility was assumed to be radiologically contaminated, which required its disposal as mixed low-level radioactive wyte.

Hazardous Waste j The generation, storage, transportation, and disposal of hazardous waste are regulated by the PaDEP under Pennsylvania's Solid Waste Management Act (35 P.S. 6018.101 et Seq.).

Decommissioning may be expected to generate very small amounts of hazardous waste.

Decontamination and dismantlement activities primarily use nonhazardous chemicals or mechanical processes. Potential sources of hazardous waste include lead-based paint that was used to cover much of the painted surfaces of the facility and mercury-containing instruments i l and switches. Other minor sources of hazardous waste may be encountered during l

decommissioning; however, the amount of waste generated is expected to be less than the limit l for a small quantity generator as specified by Pennsylvania hazardous waste regulations.

The staff concludes that the potential hazard due to nonradiological material will not effect offsite activities or conditions. The staff is satisfied that the licensees have exhibited an acceptable i level of appreciation for the hazardous nature of this material and are planning decommissioning

! activities accordingly.

4,9 Socioeconomic Effects Socioeconomic impacts associated with the SNEF stem mainly from the shutdown of the facility in 1972, which resulted in the loss of certain jobs and income to the community.

Decommissioning of the SNEF should provide a small short-term increase in income for the community.

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P 5 Postulated Aooidents The licensees for the SNEF considered the decommissioning activities described in the PSDAR and the postulated licensing-basis accidents identified in the USAR. The licensees then -

analyzed potential accidents that were identified by this review. After comparing their calculated

, doses with the recommendations of the EPA for emergency planning purposes (Ref. 28), they l concluded that no potential accidents could result in doses larger than a small fraction of the

l. EPA protective action guide (PAG) of 10 mSv (1000-mrom) whole-body dose for an individual i located at the site boundary. In addition, for the purpose of comparison with the GEIS, the

! licensees analyzed a severe transportation accident and determined the lung dose from the material-handling accident. The list of accidents in the GEIS is not an exact match with the accidents considered for the SNEF. However, the staff has determined that the accidents analyzed by the licensees are appropriate to the SNEF.

The NRC staff has evaluated the significance of accidents at the SNEF by using site criteria contained in 10 CFR Part 100 (250 mSv or 25-rem whole-body dose to an individual located un the exclusion area boundary). In 10 CFR 100.11, footnote two, the Commission emphasized that l the use of 250 mSv (25 rem) as a site guide was not intended to imply that this level of exposure l- constituted an acceptable limit for emergency doses to the public under accident conditions.

I Rather, the 250-mSv (25-ram) dose was selected as a reference value for use in evaluating reactor sites for accidents with an exceedingly low probability of occurrence and, therefore, a low risk of radiation exposure for members of the general public (10 CFR 100.11, footnote 2). The NRC has stated in the " Standard Review Plan" (NUREG-0800) that plant siting and dose-mitigating engineered safety features systems are acceptable if the whole-body dose from accidents is well within the exposure guideline values of 10 CFR Part 100. The NRC defined "well within" as 25 percent of the 10 CFR Part 100 values, or 60 mSv (6 rem) for whole-body doses, in addition, the staff compared accident consequences with annual dose criteria specified in 10 CFR Part 20 for normal reactor operation (1 mSv (100 mrem) whole-body dose to individual members of the public)(10 CFR 20.1301). Summaries of these evaluations are provided in the following subsections.

8,1 Material-Handling Acciden6-Dropped Rosin Vessel The worst-case material-handling accident considered involved dropping a steel domineralizer vessel that contains all of the used resins remaining at the site during removal of the vessel from containment. The licensees calculated offsite whole-body dose as a result of this accident at less than 15 pSv (1.5 mrom) to an individual standing at the site boundary for the duration of the event. This calculated whole-body dose is a small fraction of the annual dose criterion of 1 mSv (100 mrem) to the whole-body cited in 10 CFR 20.1301 for normal reactor operation, and  :

represents an even smaller fraction of the 60-mSv (6-rem) value that is considered well within l the 250-mSv (25-rem) reference value for a whole-body dose cited in 10 CFR 100.11(a).

Be':ause this is the worse-case accident for the SNEF, the licensees also calculated the dose to I the lung to allow a comparison with the GEIS because accident doses in the GEIS are given as lung doses. The dose to the lung from this accident is calculated to be less than 0.003 mSv (0.3 mrom). The dose to the lung for the worse-case accident in the GEIS is 0.16 mSv (16 mrem).

The staff concludes that the dose to the lung for a member of the public from the worst-case accident for the decommissioning of the SNEF is bounded by the GEIS. I I

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J PosuuTuoAccioants The licensees considered the risks and consequences of an accidental drop of a large component (e.g., the reactor vessel, steam generator, or pressurizer) during removal from containment. However, the amount of radioactive material that could be released from intemal surfaces of these vessels is less than the radioactive inventory estimated for the used resins contained by the domineralizer vessel. (For example, the surface contamination in the pressure vessel was estimated to be 0.437 TBq (11.8 Ci), compared with 0.63 TBq (17 Ci) in the domineralizer vessel.) Further, in the event of vessel rupture as a result of an accidental drop, radioactive resin beads would be much more mobile than would the surface contamination within the reactor vessel, steam generator, or pressurizer.

The staff concludes that dropping the domineralizer vessel represents the worst-case, or bounding, dose estimate for postulated material-handling accident scenarios. The staff further concludes that the resin-drop accident at the SNEF would pose no serious radiological risk to the general public.

5.2 Fire Accident-Combustible Waste Stored in the Yard Combustible waste materials stored in the yard area were identified as the most serious fire hazard for the SNEF. Radioactive material released during a fire in this area would not be limited by the confinement building or the HEPA filtration system. Using conservative atmospheric dispersion parameters and dose calculation methodologies, the licensees calculated the whole-body dose to an individual standing at the site boundary for the duration of the release at less than 3 pSv (0.3 mrem). This calculated dose is a small fraction of the annual dose criterion of 1 mSv (100 mrem) to the whole-body cited in 10 CFR 20.1301 for normal reactor operation, and represents an even smaller fraction of the 60-mSv (6-rem) value that is considered well within the 250-mSv (25-rem) reference value for a whole-body dose cited in 10 CFR 100.11(a). Thus, the staff concludes that a fire accident at the SNEC facility would pose no serious radiological risk to the general public. ,

8.3 Rupture of Vacuum Filter Bag Using conservative atmospheric dispersion parameters and dose calculation methodologies, the i licensees calculated the whole-body dose to an individual standing at the site boundary for the duration of a release resulting from a vacuum filter bag rupture at less than 0.9 pSv (0.09 mrom).

This calculated dose is a small fraction of the annual dose criterion of 1 mSv (100 mrem) to the whole-body cited in 10 CFR 20.1301 for normal reactor operation and represents an even smaller fraction of the 60-mSv (6-rem) value that is considered well within the 250-mSv (25-rem) .

reference value for a whole-body dose cited in 10 CFR Part 100.11(a). Thus, the staff concludes j that a vacuum filter bag rupture accident at the SNEF would pose no serious radiological risk to {

the general public. j 5.4 Segmentation of Components or Structures Without Local Engineering Controls Using conservative atmospheric dispersion parameters and dose calculation methodologies, the l licensees calculated the whole-body dose to an individual standing at the site boundary for the i duration of a release from the segmentation of contaminated components or structures without

or during the loss of local engineering controls at less than 15 pSv (1.5 mrem). This calculated dose is a small fraction of the annual dose criterion of 1 mSv (100 mrem) to the whole-body cited in 10 CFR 20.1301 for normal reactor operation, and represents an even smaller fraction of the 60-mSv (6-rem) value that is considered well within the 250-mSv (25-rem) reference value for a 5-2

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)

PosTuuTso AcciosuTs i I

whole- body dose cited in 10 CFR Part 100.11(a). Thus, the staff concludes that a segmentation accident at the SNEF would pose no serious radiological risk to the general public. )

8.8 Oxyacetylene Explosion Oxyacetylene torches may be used to segment piping for reactor cooling systems and other piping systems within the containment building. Violent explosions can occur when acetylene and oxygen are incorrectly mixed. Using conservative atmospheric dispersion parameters and dose calculation methodologies, the licensees calculated the whole-body dose to an individual standing at the site boundary for the duration of the release from such an explosion at less than 0.5 pSv (0.05 mrom). This calculated dose is a small fraction of the annual dose criterion of 1 mSv (100 mrom) to the whole-tody cited in 10 CFR 20.1301 for normal reactor operation, and represents an even smaller fraction of ths 60-mSv (6-rem) value that is considered well within the 250-mSv (25-rem) reference value for a whole-body dose cited in 10 CFR 100.11(a). Thus, the staff concludes that an accidental oxya atylene explosion at the SNEF would pose no serious radiological risk to the general public.

8.8 Explosion of Liquid Propane Gas (LPG) Leaked From a Front-End Loader Using conservative atmospheric dispersion parameters and dose calculation methodologies, the licensees calculated the whole-body dose to an individual standing at the site boundary for the duration of the release from such an explosion at less than 4 pSv (0.4 mrem). This calculated dose is a small fraction of the annual dose criterion of 1 mSv (100 mrem) to the whole-body cited in 10 CFR 20.1301 for normal reactor operation, and represents an even smaller fraction of the 60-mSv (6-rem) value that is considered well within the 250-mSv (25-rem) reference value for a whole- body dose cited in 10 CFR 100.11(a). Thus, the staff concludes that this type of accidental explosion of LPG at the SNEF would pose no serious radiological risk to the general public.

8.7 Failure of Liquid Waste Storage Vessel This analysis was based on the assumption that a tank containing 1900 L (500 gal) of radioactive liquid waste at atmospheric pressure developed a leak and that all of the liquid was released.

The analysis further assumed that a release fraction equivalent to 5 x 104 of the radioactive material in the tank would become airbome, an assumption based on data published in DOE-HDBK-3010-94 (Ref. 29). The referenced handbook contains an analysis of experimental measurement of airbome release fractions under simulated vesselleakage below the level of the liquid contents and under a range of pressures. The staff considers this to be a conservative assumption since the handbook lists this as the bounding release fraction for a tank pmurized at or below 50 psig. Under these circumstances, and using conservative atmospheric 4.asion parameters and dose calculation methodologies, the licensees calculated the whole-body dose to an individual standing at the site boundary for the duration v the release was calculated at 4 4 less than 5 x 10 mSv (5 x 10 mrem).

The licensees reasoned that the low volume and remote location of liquid radioactive waste potentially available for release would preclude direct entry into the nearest river (i.e., the Raystown Branch of the Juniata River). Any entry into the river would have to be through the groundwater system, and the potential dose from this pathway would be insignificant because virtually all radioactive material released would be bound up in the soil. Thus, the release rate to the river via groundwater would be very slow.

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'e PosTuuTso AoccamTs Because tritium does not bind with soil, the licensees performed an additional calculation that assumes the failure of a 3800 L (1000 gal) tank of radioactive liquid waste that contains levels of I tritium greater than that found in water in the CV sump (See Section 4.5.1). An analysis of I radiation doses was performed which considered the use of groundwater for crop irrigation and as a source of drinking water. The maximum dose to a member of the public from these pathways is 7 x 104 mSv (7 x 104mrem).

These calculated doses are a small fraction of the annual dose criterion of 1 mSv (100 mrom) to the whole-body cited in 10 CFR 20.1301 for normal reacto! ' operation, and represents an even smaller fraction of the 60-mSv (6-rom) value that is considered well within the 250-mSv (25-rem) reference value for a whole-body dose cited in 10 CFR Part 100.11(a). The staff concludes that an accident involving failure of a liquid waste storage vessel at the SNEF would pose no serious radiological risk to the general public by either an airbome or a liquid pathway. ,

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5.8 in Sifu Decontamination of Systems '

Large-scale chemical decontamination of systems is not anticipated as part of the decommissioning of the SNEF; however, limited application of chemicals may be used on systems or tanks to reduce radiation dose rates before dismantlement or general decontamination of the area. Because the use of chemicals for decontamination will be limited and the amount of contaminated chemicals produced small, the staff concludes that potential radiological releases from accidents involving in situ decontamination of systems will be small j and are bounded by the dropped resin vessel and explosion events analyzed by the licensees.

5.9 Loss of Support Systems i Electric power, cooling water, and compressed-air systems provide support to decommissioning i activities. Loss of these systems could affect work activities in many plant areas, and could also affect the systems themselves.

Offsite power is used to energize tools, cranes, lighting, and air-filtering equipment operated during decommissioning activities. A loss of power to plant ventilation and filtering systems could disrupt airflow paths and render the HEPA filters ineffective. In the event of loss of offsite power, the licensees will suspend any work activities with the potential for causing airbome contamination.

A loss of offsite power could result in loss of power to material-handling equipment.

Occupational Safety and Health Administration regulations require that crane-hoisting units be equipped with holding brakes. Although loss of power is not expected to result in crane or hoist failure, the radiological consequences of such an event would be bounded by the analysis for material-handling accidents.

A loss of compressed air or cooling water being used to support decommissioning activities will interrupt work, but it will not result in the release of radioactive material. Thus, a loss-of-cooling-water event or a loss-of-compressed-air event would not adversely affect the radiological health and safety of the public.

The staff concludes that loss of electric power during any anticipated decommissioning activities would not cause a radiological accident that could adversely affect the radiological health and safety of the public.

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r Posm.ATso AcciosuTs 5.10 External Events External events described in the USAR were reviewed to evaluate the potential radiological consequences from a natural or manmade event at the SNEF during the decommissioning phase. The staff concludes that the effect of extemal events on the SNEF would not adversely l affect the radiological health and safety of the public.

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5.11 Offsite Radiological Events Offsite radiological events related to decommissioning activities are limited to those associated

! with the shipment of radioactive material. Radioactive material will be shipped in accordance l with applicable regulatory requirements. The Radioactive Waste Management Program and the l Operational Quality Assurance Plan ensure compliance with these requirements. Historically, many thousands of accident-free shipments have been made in accordance with these requirements. The staff concludes that compliance with shipping requirements ensures that l

neither the probability of an occurrence nor the consequences of an offsite event would significantly affect the radiological health and safety of the public.

, However, to allow a comparison to the GEIS, the licensees calculated the maximum lung dose to a member of the public from a severe transportation accident. This is the accident in the GEIS  !

that has the largest maximum dose to a member of the public. j The licensees assumed that a truck carrying two large shipping containers is in an accident and .

is completely consumed by fire. This type of shipment is representative of shipments from the SNEF. The containers contain 99.8 percent of the Type A LSA limit for this type of container which would bound shipments from the SNEF. The radionuclide mixture in the containers is  !

representative of the loose surface distribution of radionuclides found in areas of the CV. The i licensees used assumptions similar to those in NUREG/CR-1750 (Ref. 25) for release fraction, l

location of the maximum exposed individual, and atmospheric conditions. The maximum j l calculated lung dose from this accident is 0.068 mSv (6.8 mrem). This is well within the dose in l

_ the GEIS of 0.16 mSv (16 mrem) and the staff concludes that this accident is bounded by the GEIS.

5.12 Containment Vessel Breach j lt is possible that the steel liner of the CV could be accidentally breached during

! decommissioning operations. A below-grade breach would result in groundwater intrusion.

l Although the in-leakage would be a nuisance, it could be readily eliminated by plugging. Not only would plugging halt further groundwater intrusion, it would effectively eliminate the likelihood of 1 i

groundwater contamination from such a long-term breach. An above-grade breach would likely involve a small cross-sectional area (relative to the large opening between the CV and DSF) which would be accommodated by the ventilation system. Airflow through a breach would be from the outside in, and exhausted air would pass through the monitored HEPA filtration unit.

The licensees have committed to include precautions in their facility procedures to reduce the probability that the CV will be challenged (either as a contamination barrier or as a barrier to groundwater intrusion). For these reasons, the staff concludes that penetration of the CV liner is a low-probability event that carries a minimal radiological consequence.

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POSTULATED AccesuTs 5.13 Summary of Postulated Aceidentimpacts These analyses demonstrate that credible accidents during the SNEF's decommissioning operations would not be expected to have a significant adverse radiological impact on the health and safety of the public or on the environment. The highest calculated dose to an individual located at the site boundary would be less than 15 pSv (1.5 mrom) to the whole-body during a postulated material-handling accident. The results of other onsite accidents are below this value.

The limiting accident case represents a very small fraction of the 60-mSv (6-rem) value that is considered well within the 250-mSv (25-rem) reference value to the whole-body for an individual located on the exclusion area boundary as stipulated in 10 CFR 100.11(a), Reactor Site Criteria, and in fact represents a small fraction of the annual dose limit of 1 mSv (100 mrom) as stipulated ;

in 10 CFR 20.1301, which is applicable for individual members of the public during normal operations. The highest calculated dose to an individual located at the site boundary for these i events is also well below the EPA PAG of 10 mSv (1000 mrom) whole-body dose. The staff finds the range of accidents analyzed by the licensees to be appropnate, the assumptions and methodologies used to be adequately conservative, and the radiological dose estimates to be well below accepted dose criteria for members of the public. Thub, the staff concludes that there is reasonable assurance that no credible accidents related to decommissioning activities at the SNEF would pose unacceptable radiological risks to the health and safety of the public.

This radiation dose to the maximum-exposed individual from accidental radionuclide releases during decommissioning of the reference test reactor is given in the GEIS as a 50-year

( committed dose equivalent of 0.16 mSv (16 mrem). This value bounds the radiation dose to the maximum-exposed individual 0.003 mSv (0.3 mrom) from accident radionuclide releases from the SNEF and 0.068 mSv (6.8 mrem) from the severe transportation accident. The staff concludes that doses from accidents from decommissioning the SNEF are bounded by the GEIS.

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l 8 Agencies and Persons Consulted and Sources Used l 4

I The NRC staff, with the assistance of technical personnel from the Idaho National Engineering i and Environmental Laboratory (INEEL), performed the review and evaluation of the licensees' ,

application for authorization to decommission, including the decommissioning environmental l report, the PSDAR, and the USAR. Principal contributors to this assessment include Alexander  !

Adams, Jr., of the NRC staff and William Gammill, Robert Carter, James Miller, and Allan Wylie of the INEEL 1 The staff consulted with the Pennsylvania State official regarding the environmental impact of  ;

the proposed action. The State official had no comments.  !

A public meeting was held on the proposed PSDAR for the Saxton facility on January 28,1997.

The meeting was held in the town of Saxton and included representatives from SNEC, the NRC, Bedford County, and members of the public living in the vicinity of the facility. The public's interests anci concerns are reflected in this assessment and in the associated safety evaluation.

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e l 7 Conclusion i

The staff reviewed and evaluated the SNEF Decommissioning Environmental Report prepared l by GPU Nuclear and submitted on April 17,1996, as supplemented on July 18,1996 and l

! March 3 and 31,1998. The staff also evaluated the licensees' plans for decommissioning the j facility as described in the PSDAR and their ability to manage radioactive waste generated  !

l during dismantlement. This evaluation included the Radiation Protection Program, which implements the regulatory requirements of 10 CFR Part 20 (Ref. 21) through plant procedures that are established to maintain radiation exposures ALARA.

In accordance with the proposed decommissioning plan described in the licensees' PSDAR, the  !

staff has determined that the environmental impacts, both radiological and nonradiological,  ;

associated with the decommissioning of the SNEF are bounded by the conditions evaluated in l the Generic Environmental Impact Statement (Ref. 6) and have been adequately evaluated in j the SNEF's decommissioning environmental report. The staff also finds that the proposed ,

l decommissioning of the SNEF complies with 10 CFR Part 50, Appendix 1, and 10 CFR Part 20.

! In many areas, the decommissioning of the plant will reduce the already small environmental effects associated with the plant. During the decommissioning process, the transportation of low-level radioactive waste will be higher than it was during the shutdown period or when the l- facility was operational. This increase, however, will not have a significant impact on either the i environment or the nearby population. The waste generated and shipped during i decontamination and dismantlement will be regulated in accordance with the applicable provisions of 10 CFR Parts 20,61, and 71 (Refs. 21,22, and 23) and with pertinent DOT regulations. l Based on a review of the licensees' proposed schedule for decommissioning, the staff finds that l this schedule is acceptable and that it is consistent with the level of effort required and the complexity of the undertaking. The staff further concludes that there are no significant

environmental impacts associated with the actions proposed and that these actions will not have

! a significant effect on the quality of the human environment. Pursuant to 10 CFR 51.31, an

environmental impact statement is not required.

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? l 8 References 1

1 1. Saxton Nuclear Experimental Corporation Facility Post-Shutdown Decommissioning Activities Report (issued as the Saxton Nuclear Experimental Corporation Facility Decommissioning Plan), GPU Nuclear, February 1996.

2. Amendment No. 8 to Amended Facility License No. DPR-4, Saxton Nuclear Experimental Corporation, U.S. Atomic Energy Commission, August 15,1972.

l 3.10 CFR Part 50," Domestic Licensing of Production and Utilization Facilities." U.S.

Government Printing Office, Washington, DC.

4. 10 CFR Part 51, " Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions." U.S. Govemment Printing Office, Washington, DC.

5. Saxton Nuclear Experimental Corporation Facility Decommissioning Environmental Report, GPU Nuclear, April 1996.

6. NUREG-0586, " Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities." U.S. Nuclear Regulatory Commission, Washington, DC, August 1988.

7. Preliminary Hydrogeological Investigation, Saxton Nuclear Experimental Station, Saxton, I PA, CrounckWater Technology, Inc.,1981.

8. Report on Drilling and Radiometric Analysis of Samples Collected at Sites of Spent Resin and Liquid Waste Tanks, SNEC Facility (Saxton, PA: Pennsylvania State University),

Westinghouse Electric Corp., General Public Utilities, January 16,1989.

9. Saxton Soil Remediation Project Report, GPU Nuclear, May 11,1995.

10. Saxton Site Characterization Plan, Procedure No. 6575-PLN-4520.06.

11. Saxton Radiation Protection Plan, Procedure No. 6575-PLN-4542.01.

12. 10 CFR Part 20, Subpart E, " Radiological Criteria for License Termination,' Federal Register (Volume 62, Number 139), July 21,1997.

13. NUREG/CR-5849, " Manual for Conducting Radiological Surveys in Support of License Termination," U.S. Nuclear Regulatory Commission, August 1994.

14. NUREG-1505, "A Non-Parametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys," U.S. Nuclear Regulatory Commission, August 1995.'

l

15. NUREG-1506, " Measurement Methods for Radiological Surveys in Support of Decommissioning Criteria," U.S. Nuclear Regulatory Commission, August 1995.  !

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16. NUREG-1507, " Minimum Detectable Concentrations with Typical Radiation Survey l Instruments for Various Contaminants and Field Conditions," U.S. Nuclear Regulatory )

Commission, August 1995. l l 8-1

o REFERENCES

17. "Saxton Nuclear Experimental Facility Final Release Survey of the Reactor Support Buildings," GPU Nuclear Corporation Report to U.S. Nuclear Regulatory Commission, April l 1990. l l

18. Radiological Analysis File Number 6612-96-007," Maximum Offsite Dose from Release of Saxton Pipe Tunnel Water," 8.A. Parfitt, March 22,1996.

19. Eisenbud, M., " Environmental Radioactivity," Third ed., Academic Press Inc.,1987. l

20. Phase i Report of Findings - Groundwater Investigation Saxton Nuclear Experimental Station, Saxton, PA, GEO Engineering,1992.

21. 10 CFR Part 20, " Standards for Protection Against Radiation." U.S. Govemment Printing Office, Washington, DC.

22. 10 CFR Part 61, " Licensing Requirements for Land Disposal of Radioactive Waste."

U.S. Govemment Printing Office, Washington, DC.

23. 10 CFR Part 71,

• Packaging and Transportation of Radioactive Material." U.S. Govemment Printing Office, Washington, DC.

24. Saxton Nuclear Experimental Corporation Facility Decommissioning Project, Presentation Slides, NRC meeting Decert.ber 4,1997.

25. NUREG/CR-1756, " Technology, Safety and Costs of Decommissioning Reference Nuclear Research and Test Reactors," U.S. Nuclear Regulatory Commission, Volumes 1 and 2, March 1982, cnd Addendum, July 1983.

26. Saxton Nuclear Experimental Corporation Site Characterization Report,1996.

27. Regulatory Guide 1.109 " Calculation of Annual Doses to Man from Routino Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix 1,* Rev 1, U.S. Nuclear Regulatory Commission,1977,

28. EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear incidents, U.S. Environmental Protection Agency,1991.

29 DOE-HDBK-3010-94, Airbome Release Fractions / Rates and Respirable Fractions for Nonreactor Nuclear Facilities, Volume I - Analysis of Experimental Data, U.S. Department of Energy, Washington, D.C., December 1994.

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