ML20217A982

Safety Evaluation Supporting Amend 15 to License DPR-4

ML20217A982
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Site:	Saxton File:GPU Nuclear icon.png
Issue date:	04/20/1998
From:	NRC (Affiliation Not Assigned)
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ML20217A969	List: ML20217A964 ML20217A976 ML20217A982
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NUDOCS 9804220381
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2066M001

4 9 . . . . . ,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO.15 TO ,

l AMENDED FACILITY LICENSE NO. DPR-4 GPU NUCLEAR. INC.

SAXTON NUCLEAR EXPERIMENTAL CORPORATION DOCKET NO. 50-146

1.0 INTRODUCTION

By letter dated November 25,1996, as supplemented on May 30, June 4 and 16, August 21 and September 16,1997, and February 3 and 9,1998, and March 31,1998, GPU Nuclear, Inc. and the Saxton Nuclear Experimental Corporation (SNEC) (the licensees) submitted a request for amendment to Amended Facility License No. DPR-4 for the SNEC Facility (SNEF or f acility). In its evaluation of the licensees' request, the NRC staff also referred to the SNEF Decommissioning Environmental Report submitted on April 17,1996; the response from the licensees to NRC questions about the environmental report dated July 18,1996, and March 3 and 31,1998; and the SNEC Facility Updated Safety Analysis Report, Revision 0, submitted on October 25,1936; Revision 1, submitted on August 21, 1997; and Revision 2, submitted on February 3,1998. As part of the review of the amendment request, the staff also reviewed the Saxton Nuclear Experimental Corporation Facility Decommissioning Quality Assurance Plan submitted by letter dated November 8, 1996, as supplemented on May 30,1997, and February 3 and 9,1998.

The requested changes to the license and Technical Specifications (TSs) would (1) accommodate decommissioning activities at the SNEF, (2) establish specific TS controls over decommissioning activities, (3) establish limiting conditions for performing decommissioning activities, (4) extend exclusion area controls to include the Decommissioning Support Facility (DSF), (5) establish requirements for a Radiological Environmental Monitoring Program (REMP) and an OffSite Dose Calculation Menual (ODCM),

and (6) establish requirements for technical and independent safety reviews. In addition, the amendment authorizes other administrative and editorial changes to the TSs as-sociated with the changes described above.

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2.0 BACKGROUND

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The SNEF is located in Pennsylvania, about 160 kilometers (100 miles) east of Pittsburgh and 145 kilometers (90 miles) west of Harrisburg in the Allegheny Mountains,1.2 j

kilometers (three fourths of a mile) north of the Borough of Saxton in Liberty Township, i

Bedford County. The site is on the north side of Pennsylvania Route 913,27 kilometers

! (17 miles) south of U.S. Route 22, and about 24 kilometers (15 miles) north of the j Breezewood Interchange on the Pennsylvania Turnpike.

1 Built in 1960-62 under a license to SNEC, the facility was licensed as a demonstration power reactor at several different power levels during its history. The maximum licensed i

power level was 35 megawatts of thermal energy (MWT), although the prevailing licensed power level was 23.5 MWT. From 1962 to 1972, the facility was primarily used for research and training and for tcsting mixed oxide fuels. The reactor was also used to simulate fuel-cladding failures, which resulted in some plutonium, americium, and fission-product contamination remaining at the facility. In 1972, the facility was shut down and placed in a condition equivalent to what is now defined by the NRC as "SAFSTOR"(safe storage), and its operating license was changed to possession-only status.

l In 1972, all fuel was removed from the SNEF containment vessel (CV) and was returned to the U.S. Atomic Energy Commission at its Savannah River Plant in South Carolina. As a result, the licensees are no longer responsible for the facility's spent fuel. In addition, the control rod blades and the superheated steam test loop were shipped off the site. Af ter the

' fuel was removed, equipment, most tanks, and piping external to the CV were also i removed. Buildings and structures that supported reactor operations were partially  ;

decontaminated in 1972-74. '

i in preparation for demolition, final decontamination of reactor support structures and buildings was done in 1987-89. This process included decontamination of the control and auxiliary building, radioactive waste disposal facility, yard pipe tunnel, and filled drum l l storage bunker, as well as removal of the refueling water storage tank. After the NRC accepted the final release survey, these buildings were demolished in 1992.

The Saxton Soil Remediation Project was completed in November 1994. This was a comprehensive project involving soil monitoring and sampling. Soil that was located within the site perimeter and found to be contaminated with radioactive material was excavated, packaged, and shipped to a licensed radioactive waste disposal facility.

i in preparation for release of the site for unrestricted use, the licensees now propose to decontaminate and dismantle the SNEF CV; the concrete shield wsillocated around the northwest and northeast quadrants of the CV; the tunnel sections that are immediately l

adjacent to the outer circumference of the CV; and remaining portions of the septic system, welts, and associated underground piping. These structures contain or are suspected to contain residual radioactive material.

j The SNEF is possessed under a 10 CFR Part 50 license and its associated TSs. SNEC and GPU Nuclear are joint licensees of the SNEF and are subsidiaries of the same parent company, General Public Utilities, Inc. NRC approved a license amendment request (Amendment No.13) on May 10,1996, to designate GPU Nuclear as a co-licensee, i

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.- i 3-On February 16,1996, SNEC sebmitted a decommissioning plan for the SNEF. On April 17, i 1996, SNEC submitted the. Facility Decommissioning Environmental Report, and on May 17, 1996, SNEC submitted Technical Specifications Change Request (TSCR) No. 58. The NRC initiated a review of the submitted documents and sent a request for additional information l (RAl) to the licensees on June 6,1996; the licensees responded on July 18,1996.

On July 29,1996, the NRC published a final rulemaking (effective August 28,1996) that substantially changed the decommissioning process. The changes to the regulations eliminated the requirement to submit a decommissioning plan and the requirement that the i

' NRC review and approve the decommissioning plan before the start of decommissioning activities. The revised regulations resulted in the decommissioning plan being considered the post shutdown decommissioning activities report (PSDAR). Because of this revision to the regulations, NRC staff halted its review of the SNEF decommissioning plan.

The amendments to the regulations were based on the fact that the degree of regulatory oversight required for a nuclear power reactor during its decommissioning stage is much less than that required for the facility during its operating stage. The amendments to the regulations establish a level of NRC oversight commensurate with the level of safety concerns expected during decommissioning activities.

The purpose of the PSDAR is to give the NRC staff sufficient information to assure the proper NRC oversight of any significant decommissioning activitir;, to require the licensee i

to examine its plans for the funding of decommissioning activities, and to require the licensee to examine its plans for decommissioning to assure that the activities will not result in any environmental impacts that have not been previously considered. The NRC staff is I

not required to review and approve the PSDAR. Instead, the staff examines the PSDAR and l

1 makes a determination about whether or not it provides the information required by the  !

regulations. The NRC staff has completed its examination and has determined that the l l SNEF PSDAR in combination with the SNEF Environmental Report meets the requirements '

! of 10 CFR 50.82(a)(4)(i) regarding the licensees' planned decommissioning of the SNEF.

On October 25,1996, the licensees submitted the SNEC Facility Updated Safety Analysis Report, P.evision O. On November 25,1996, the licensees submitted an application for i license amendment (TSCR No. 59), which included proposed decommissioning TSs and l license changes that are the subject of this safety evaluation. The licensees withdrew l TSCR No. 58.

There is a requirement in the SNEF license that GPU Nuclear shall not dismantle or dispose of the facility or the property occupied by the facility without prior approval of the Commission. The TSs state that the licensees are prohibited from taking any action that results in alteration of the CV, removal of major radioactive components, or dismantling of components. These requirements can only be changed by amending the facility license.

TSCR No. 59 requested removal of these requirements and proposed other changes to the license and TSs to support decommissioning activities.

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After approval of the license amendment, the licensees could begin to perform major decommissioning activities without specific NRC approval using the process described in

' 10 CFR 50.59. The licensees are required to maintain records of 10 CFR 50.59 reviews and to report on these reviews to the NRC. The licensees' process for performing these i

reviews, the reviews themselves, and major activities conducted at the site will be carefully l evaluated during NRC inspections.

3.0 EVALUATION 3.1 Introduction The licensees have proposed extensive modifications to the SNEC TSs to provide for planned decontamination and dismantlement activities. The existing TSs are appropriate for i

a facility that was maintained in SAFSTOR with routine activities limited to security, '

surveillance, and maintenance, and specific activities (such as removal of asbestos l- insulation and old electrical systems) that were approved by NRC on a case-by case basis.

The proposed chang 2s to the TSs are intended to accomplish the following:

o Permit decommissioning activities at the SNEF:

I o Establish control over decommissioning activities; e

Establish limiting conditions for performing decommissioning activities; e

Extend exclusion area controls to include the SNEF's DSF which consists of the decommissioning support building, the material handling bay, and the personnel access facility; '

Establish requirements for a REMP, an ODCM, and a Process Control Program (PCP) for radioactive waste; e Establish requirements for technical and independent safety reviews; and e

Accommodate desirable administrative and editorial changes.

The degree of regulatory oversight required for a nuclear power reactor facility during its decommissioning stage is less than that required during its operating stage. During the operating stage, fuelin the reactor core becomes highly radioactive and contains the l overwhelming majority of the radioactive material on the site. Safe control of the nuclear fission process involves the use and operation of many complex systems and the establishment and maintenance of proper operating procedures with appropriately trained staff to ensure that the reactor is properly operated and maintained, and that radiation exposures to the public and operating personnel are maintained within regulatory limits and

( are as low as is reasonably achievable (ALARA).

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During a nuclear power reactor's decommissioning stage, the nuclear fission process has stopped and the fuel assemblies are permanently removed from the reactor core. Once the  !

fuel assemblies are removed from the site, the potentialimpact on public health and safety of activities performed by the licensee during decommissioning is far less significant. {

Contaminated and activated areas of the facility must still b6 controlled to limit the spread of radioactive inatorial and to maintain personnel radiation exposures at ALARA levels, j

Radiological hazards presented by credible events during decommissioning activities at the SNEF are much less severe than those presented when the plant was operational. The I reduction is a direct result of the absence of irradiated fuel and of the ladioactive decay that has occurred in residual raaterials during the 25 years since the cessation of reactor i operations. This 25-year period was sufficient to pern..t the decay of virtually all shorter- I lived radionuclides. Consequently, limiting conditions for operation (LCOs) or surveillance '

l requirements that are normally found in the TSs for an operating plant or a permanently

! defueled plant in which spent fuelis stored in the fuel pool, are not required. The only comparable limiting conditions retained in the proposed TSs for the SNEF are those involving the release of airborne radioactive materials via the monitored ventilation system, l which includes high-efficiency particulate air (HEPA) tilters.

l l The remaining requirements are appropridely specified in the Administrative Controls I l

section, the contents of which were found to be generally consistent with the i

administrative controls section of the improved standard technical specifications (STSs) for  !

power reactors. As in the STSs, the proposed TSs now include a set of definitions for  ;

terms applicable to all associated decommissioning activities. The definitions establish a common framework of understanding and help avoid ambiguities associated with the  !

implementation of the TSs.

3.2 Accident Analvses i

l The licensees for the SNEF reviewed the decommission!ng activit!ss described in the PSDAR  !

l and the postulated licensing-basis accidents identified in the updated safety analysis report l

(USAR). The licensees then analyzed potential accidents identified by this review. After i comparing their calculated doses with the recommendations of the U.S. Environmental l

Protection Agency (EPA) for emergency planning purposes (Ref.1), they concluded that no l

l potential accidents could result in doses larger than a small fraction of the EPA protective l action guide (PAG) of 1000 mrem whole body dose for an individual located at the site boundary. The site boundary as defined in the proposed TS and used in the licensees' l l

accident analysis is the same as the exclusien area boundary used in 10 CFR Part 100. The l definition of exclusion area in the TSs is a historical term that refers to an area of the SNEF 1

! in which considerations of security and access restrictions exist.

l The NRC staff has evaluated the significance of accidents at the SNEF by using site criteria contained in 10 CFR Part 100 (25-rom whole-body dose to an individual located on the exclusion area boundary). In 10 CFR 100.11, footnote 2, the Commission emphasized that the use of 25-rem as a site guide was not intended to imply that this level of exposure constituted en acceptable limit for emergency doses to the public under accident conditions.

Rather, the 25 rem dose was selected as a reference value for use la evaluating reactor sites for accidents with an exceedingly low probably of occurrence and, therefore, a low l

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l 6-l 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 end 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 6 rem for whole-body doses, l

in addition, the staff compared accident consequences with annual dose criteria specified in 10 CFR Part 20 for normal reactor operation (100 mrom whole-body dose to individual members of the public) (10 CFR 20.1301).

3.2.1 Material Handlina Accident-Drooned Resin Vessel The worst case material handling accident considered involved dropping a steel l

domineralizer vessel that containe 3ll of the used resins remaining at the site during removal l of the vessel from the CV. Using conservative atmospheric dispersion parameters and dose calculation methodologies, the licensees calculated offsite doses as a result of this accident, at less than 1.5 mrem whole-body to an individual standing at the site boundary for the duration of the event. This calculated dose is a small fraction of the annual dose criterion  !

of 100 mrem to the whole body cited in 10 CFR 20.1301 for normal reactor operation and represents an even smaller fraction of the 6-rem value that is considered well within the 25-rem reference value for a whole-body dose cited in 10 CFR 100.11(a).

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 the CV. However, the amount of radioactive material that could be released from internal surfaces of these vessels is less than the radioactive inventory estimated for the domineralizer vessel containing used resins. (For example, the surface contamination in the reactor vessel was estimated to be 11.8 Cl, compared with 17 Ciin 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 be 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.

3.2.2 Accident Fire-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 CV 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 0.3 mrom. This calculated dose is a small fraction of the annual dose criterion of 100 mrom to the whole-body cited in 10 CFR 20.1301 for normal reactor operation, and

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represents an even smaller fraction of the 6-rem value that is considered well within the 25 rem reference value for a whole-body dose cited in 10 CFR 100.11(a). Thus, the staff concludes that a fire at the SNEF would pose no serious radiological risk to the general I public.

3.2.3 Ruoture of Vacuum Filter-Bao '

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 a release resulting from a rupture of a vacuum filter-bag at less than 0.09 mrem. This calculated dose is a small fraction of the annual dose criterion of 100 mrem to the whole body cited in 10 CFR 20.1301 for normal reactor operation, and represents an even smaller fraction of the 6-rem value that is considered well within the 25-rom referer":e value for a whole-body dose cited in 10 CFR Part 100.11(a). Thus, the stalf cdnclubes that a vacuum filter-bag rupture accident at the SNEF would pose no serious radiological risk to the general public.

3.2.4 Seamentation of Comoonents or Structures Without Local Enoineerino Controls 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 a release from the segmentation of contaminated components or structures without or during the loss of local engineering controls at less than 1.5 mrem.

This calculated dose is a small fraction of the annual dose criterion of 100 mrem to the whole-body cited in 10 CFR 20.1301 for normal reactor operation, and represents an even smaller fraction of the 6-rom value that is considered well within the 25-rem reference value for a whole-body dose cited in 10 CFR 100.11(a). Thus, the staff concludes that a segmentation accident at the SNEF would pose no serious radiological risk to the general public.

3.2.5 Oxvacetvleno Exolosion Oxyacetylene torches may be used to segment piping for reactor cooling systems and other piping systems within the CV. 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 0.05 mrem. This calculated dose is a small fraction of the annual dose criterion of 100 mrom to the whole-body cited in 10 CFR 20.1301 for normal reactor operation and represents an even smaller fraction of the 6-rem value that is considered well within the 25 rem reference value for a whole-body dose cited in 10 CFR 100.11(a). Thus, the staff concludes that an accidental oxyacetylene explosion at the SNEF would pose no serious radiological risk to the general public.

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3.2.6  !

Exolosion of Llauld Prooane 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 0.4 mrem. This calculated dose is a small fraction of the annual dose criterion of 100 mrom to the whole-I body cited in 10 CFR 20.1301 for normal reactor operation, and represents an even smaller I

fraction of the 6-rem value that is considered well within the 25-rem reference value for a l

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.

3.2.7 Failure of Llauld Waste Stornos Vessel Failure l

This an'alysis was based on the assumption that a tank containing 500 gallons 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 'iO4 of the radioactive materialin the tank would become airborne, an assumption based on data published in DOE HOBK-3010-94 (Ref. 2). The referenced handbook contains an analysis of experimental measurement of airborne release fractions under simulated vessel leakage 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 pressurized at or below 50 psig. Under these circumstances, and 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 5 x 108 mrom.

The licensees reasoned that the low volume and remote location of liquid radioactive waste i

' 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.

Because tritium does not bind with soil, th6 licensees performed an additional calculation that assumes the failure of a 3800 L (1000 gal) tank of radioactive liquid waste that contains levels of tritium greater than that found in water in the CV sump. An analysis of 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 10 8 mSv (7 x 104 mrem).

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 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). 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 airborne or a liquid pathway.

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9-3.2.8 in Situ Decontamination of Svstems 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 t

systems or tanks to reduce radiation dose rates before dismantlement or general area j

decontamination. Because the use of chemicals for decontamination will be limited and the i

' amount of contaminated chemicals produced would be small, the staff concludes that potential radiological releases from accidents involving in situ decontamination of systems will be small and are bounded by the dropped resin vessel and explosion events analyzed by the licensees.

! 3.2.9 Loss of Suonort Systems Electric power, cooling water, and compressed-air systems provide support to decommissioning activities. Loss of these systems could potentially 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 l

during decommissioning activities. A inss 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 airborne contamination.

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

Regulations established by the Occupational Safety and Health Administration require that crana holsting 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 l would be bounded by the material-handling accident analysis.

A loss of compressed air or cooling water being used to support decommissioning activities t

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.

Thus, the staff concludes that loss of support systems during any anticipated decommissioning activities would not cause a radiological accident that could adversely affect the radiological health and safety of the public.

3.2.10 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 external events on the SNEF would not adversely affect the radiological health and safety of the public.

3.2.11 Offsite Radioloalcal Events l Offsite radiological events related to decommissioning activities are limited to those associated with the shipment of radioactive material. Radioactive materials will be shipped in accordance with applicable regulatory requirements. The Radioactive Waste Management  !

Program and the Operational Quality Assurance Plan ensure compliance with these '

requirements. Historically, many accident-free shipments have been made in accordance with these requirements. The staff concludes that compliance with shipping requirements ensures that neither the probability of an occurrence nor the consequences of an offsite event would significantly affect the radiological health and safety of the public.

3.2.12 Containment Vessel Breach it is possible that the steelliner 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 stop further groundwater intrusion, it would effectively eliminate l the likelihood of groundwater contamination from such a long-term breach. An above-grade I 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 l through the monitored HEPA filtration unit. The licensees have committed to include i 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.

3.2.13 Conclusion 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 l individuallocated at the site boundary during a postulated material-handling accident would l be less than 1.5 mrem to the whole-borty. The results of other onsite accidents are equal l to or below this value. The limiting accident case represents a very small fraction of the 6-rem dose considered to be well within the reference value of 25-rem to the whole bor'y for an individuallocated 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 100 mrem 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 the duration of these events is also well below the EPA PAG of 1000 mrom whole-body dose. The staff finds the range of accidents analyzed by the licensees to be appropriate, 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. Thus, 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.

3.3 Pronosed Chanoen to the Technical Soecifications The licensee has proposed changes to the TSs to permit decontamination and dismantlement of the SNEF. In Sections 3.3.1 through 3.3.7, these changes are discussed.

3.3.1 Channes to Permit Decommissionino Activities at tha.EEE The current TSs restrict activities at the SNEF and do not allow decommissioning activities.

3.3.1.1 Chanoes to Permitted Activities Actions permitted at the SNEF, as specified by Section A.3 of the current TSs, are not broad enough to accommodate decommissioning activities. These activities focus at present on possession of licensed material, radiological characterization of the facility, and specific actions otherwise approved by the NRC on a case-by-case basis. The proposed TS 2.0 would allow active decommissioning and dismantlement of the facility.

" Decommissioning Activities" as a defined term is proposed to be referenced in TS 2.0,

" Principal Activities." (Section numbers were changed throughout the TSs because of reformatting.) " Decommissioning Activities" are defined in proposed TS 1.0.3 as follows:

The term DECOMMISSIONING ACTIVITIES describes all of those activities needed to decommission the SNEC Facility and return the site to unrestricted use. Examples of these activities include: PRODUCTION ACTIVITIES needed to conduct decommissioning such as physical dismantlement; radioactive waste preparation, treatment, packaging and shipment; radiation protection activities, construction and installation of support systems, structures and components, and final status survey.

Because proposed TS 2.0 describes the activities to be conducted at the SNEF during decommissioning, the proposed changes in principal activities are acceptable to the staff.

Section B.1 of the current TSs contains a statement of GPU Nuclear's responsibility for maintaining the CV and performing characterization activities. Proposed TS 3.1 specifies that GPU Nuclear has the responsibility for safely performing decommissioning activities.

Unos of authority, responsibility, and communication are procedurally defined and established organizationally for important management staff positions. The position titles responsible for conducting all activities safely and in accordance with the regulations are j 4

stated in the proposed TSs. The positions in the organization responsible for carrying out decommissioning activities, assuring implementation of license conditions, and for radiation safety at the SNEF are given in the proposed TSs. All radiological controls personnel have stop work authority in the proposed TSs in matters relating to or impacting on radiation safety. These proposed revisions to the TSs are consistent with, and take into account changes to, the principal activities at the SNEF and, thus, are acceptable to the staff.

Because of a reorganization within GPU Nuclear, the position of Vice President, Nuclear Safety and Technical Services, has been eliminated and responsibility as related to the SNEF has been transferred to the Vice President Engineering. This results in proposed changes to

, TSs 3.1.2,3.5.5.1, and 3.5.5.4 in which the title Vice President, Nuclear Safety and l

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r Technical Services, is replaced with Vice President Engineering. Because the level of management attention within GPU Nuclear related to the SNEF has not decreased, and the

' independence of oversight provided by the Radiation Safety Officer and Radiation Safety Committee has not changed, this change is acceptable to the staff.

As described in Section B.1.a.2 of the current TSs, the responsibilities of the Program l Director of the SNEF include administration of all facility functions, direction of ali i

decontamination and characterization activities, and assurance that the license requirements and TSs are implemented. Proposed TS 3.1.3 would stipulate that the Program Director's

! responsibilities include all decommissioning activities and compliance with license requirements and TSs. The staff finds that the oroposed TS modifications are consistent with the revisions in principal activities at the SNEF and more aptly describe the Program Director's responsibilities. For these reasons, the staff considers that this proposed modification to the TSs is acceptable.

I Section B.1.c.2 of the current TSs requires the presence of the radiadon safety officer (RSO) or a qualified designee whenever activities are in progress within the CV. Proposed TS 3.2.2 would require the onsite presence of the RSO or group radiological controls supervisor (GRCS) whenever CV entry, production activities, maintenance, characterization, and/or radioactive waste management activities are in progress in any radiologically controlled area. The proposed revision was necessary because waste management activities will take place outside the CV during facility decommissioning. The staff finds that the proposed revision accurately specifies the conditions and times requiring the onsite presence of the RSO or GRCS due to expanded principal activities at the SNEF. Therefore, the staff considers that this proposed modification to the TSs is acceptable.

Section B.1 of the current TSs also references the SNEC Organization Chart, Figure 2, which the licensees have proposed be omitted from the TSs. Proposed TS 3.1 states that organizational requirements shall be documented in the SNEF USAR. The USAR states that the organization is depicted on Figures 2.3-1 and 2.3-2 of the PSDAR. The licensees state I'

' that the organization charts in the PSDAR contain auditable organizationalinformation that is qualitatively superior to that provided in the current TSs.

Section B.1.a.5 of the current TSs contains a description of the roles and responsibilities of the SNFF Site Supervisor. Proposed TS 3.1.4 would omit specific reporting responsibilities and would focus on providing onsite management and continuous oversight of production activities. Managerial relationships are identified in organizational charts, functional descriptions of departmental responsibilities and relationships, and job descriptions.

Section B.1.s.4 of the current TSs defines the supervisory responsibilities of the GRCS, and notes that the GRCS reports to the RSO. Proposed TS 3.1.6 would omit this reporting relationship because it has been noted elsewhere (e.g., in organizational charts, functional descriptions of departmental responsibilities and relationships, and job descriptions).

The " Administrative Controls" section in the proposed TSs described above capture the essential aspects of the organizational structure; thus, the staff finds that these changes to the organizational charts and position requirements are acceptable.

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Section B.1.d.1 of the current TSs addresses personnel selection and training requirements.

, These requirements are in the proposed Section 3.3.1 of the TSs. Proposed TS 3.4, l l

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" Training," would stipulate that GPU Nuclear maintain training programs for those personnel performing decommissioning work at the SNEF. Proposed TS 3.4.1 would require job specific training or knowledge for accomplishing each task or project goal dictated by i the work scope for dismantlement or decommissioning. Proposed TS 3.4.2 would permit a l competency demonstration in lieu of training for the performance of specialized tasks, techniques, and equipment operations. These requirements are new and reflect recognition I of the need for additional training to perform decommissioning activities. The staff finds l that these proposed additions are appropriate and acceptable, j Section B.1.d.2 of the current TSs requires briefings for personnel conducting permitted activities, including maintenance and characterization activities. Proposed TS 3.3.2 would l

l specify briefings for personnel performing decommissioning or associated activities on the applicable site conditions and applicable requirements for the assigned task. The staff finds '

that this modification was necessary to provide an accurate reflection of increases in the  :

principal activities to be conducted at the SNEF. Thus, the staff finds this that change to the TSs is acceptable.

l 3.3.1.2 Administrative Chanaes Due to Chanaes in Permitted Activities  !

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Section A.3 of the current TSs, which describes principal activities permitted at the SNEF, was revised in proposed TS 2.0 to eliminate the description of Pennsylvania Electric l Company (PEC) personnel on the adjacent PEC property. Section A.2.d of the current TSs l t

is proposed for elimination. This section requires PEC personnel to report any problems that '

i they have observed at the SNEF. Because of the increase in the licensees' staff on the site, l- PEC personnel are no longer needed to report on the condition of the SNEF. The NRC staff finds that this administrative change is acceptable.

Sections B.3.b.1 and B.3.c.1 of the current TSs define those activities that should be I covered by written procedures. Proposed TSs 3.6.1.2.1 and 3.6.1.3.1, respectively, would l include decommissioning activities to be consistent with the need to identify principal activities and extend controls to the DSF. Because these changes are commensurate with '

the activities at the SNEF due to decommissioning, the staff concludes that this is an acceptable administrative change.

Section B.3.b.4 of the current TSs requires the establishment of written procedures for i activities that "could impact CV integrity and/or could result in a measurable release to the environment." CV integrity is limited to those features of the CV liner required to serve as  !

both a contamination barrier and an intrusion barrier. Since the objective of this requirement is to avoid the release of measurable quantities of radioactive material to the l environment, and since the CV is now only one potential source of radioactive l contamination within the SNEF site boundary, there is no need to identify CV integrity per se in this requirement. Therefore, proposed TS 3.6.1.2.4 reads: " Activities that could result in a measurable release to the environment." The staff concludes that this is an acceptable i administrative change.

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Section B.3.c.2 of the current TSs requires that radiation levels and airborne activity surveys be obtained before beginning workin the controlled area of the CV. Proposed TS 9.6.1.3.1 would require that all decommissioning activities and maintenance work under Health Phys!ca control be consistent with the requirements of 10 CFR Part 20. The licensees reasoned that to demonstrate compliance with 10 CFR Part 20, radiological conditions must be assessed before any work begins in a controlled area and, therefore, no need exists to include this detailin the TSs. The staff finds that the revised requirement effectively encompasses the previous requirement and concludes that this is an acceptable administrative change.

Section B.5.e of the current TSs defines record retention requirements pertaining to design changes and maintenance of the SNEF. The licensees reasoned that during decommissioning, it is appropriate to maintain " records of reviews performed for changes made to procedures or equipment pursuant to 10 CFR 50.59." This wording is used in proposed TS 3.9.6, replacing the original text. The proposed revision should casure that records of changes that result from decommissioning activities will be app:cpriately maintained. The staff concludes that this is an acceptable administrative change.

Section B.6.b.3 of the current TSs specifies maintenance and design changes to be included in the annual report to the NRC. Proposed TS 3.8.2.2 would specify that decommissioning changes are also included in this report. This modification to the TSs w;ll provide for a complete report of changes at the site during the year. The staff concludes that this is an acceptable administrative change.

3.3.2 Chanaes to Establish Soecific Controtr_ Over Decommissionina Activities With the transition to full decommissioning activities at the SNEF, additional TSs are needed and existing TSs need to be modified to appropriately control the expansion of activities.

The licensees have proposed changes to the TSs to add these controls.

3.3.2.1 Fire Protection Proposed TS 3.7 would establish new requirements for fire protection, As part of its review of the fire protection program, the staff referred to Revision O of the Emergency Response Procedure and Emergency Plan (ERPEP),6575 ADM 4500.06, dated April 8,1997. The licensees have determined that a significant fire at the SNEF is highly unlikely. To provide reasonable assurance that the risk of a fire-induced radiological hazard to the public, the environment, and facility personnel is minimized, tne licensees have implemented a fire protection plan as part of the ERPEP. The fire protection plan, Exhibit 4 of the ERPEP, describes the administrative controls on transient t.ombustibles, hot work, personnel training, interface with the local fi e department, maintenance of fire protection equipment, and the annual fire protection inspections of the SNEF conducted by the Three Mile Island Nuclear Station fire protection engineer. Section 4.3 of the ERPEP describes the notification and evacuation procedure to be followed by facility personnel upon the discovery of smoke or fire. Proposed TS 3.7 prescribes that procedures will be established for the prevention of fires, the response of plant personnel to fires, the fire suppression equipment available onsite, and the training requirements for personnel performing fire watch dutm,.

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Fire suppression of incipient fires is provided by onsite personnel who are trained annually in the use of portable fire extinguishers. For fires beyond tha incipient stage, suppression is provided by the Saxton Fire Department, which is trained periodically by Three Mile Island Nuclear Station fire protection personnel. The fire department is assisted by the onsite radiological control technicians for fires in radiologically controlled areas.

The licensees' fire protection program, as described in the ERPEP and in proposed TS 3.7 provides for defense in depth in areas important to facility safety, to achieve the following objectives: prevent fires from occurring; rapidly detect, contt 8, and extinguish fires that do occur; and ensure that the risk of fire-induced radiological he tards to the public, the environment, and facility personnel is uinimized. Therefore, the staff finds that the proposed fire protection program is acceptable for decommissioning activities at the SNEF.

3.3.2.2 Insnection Raouirements The need to periodically verify the physical and radiological status of the facility during a monitored condition equivalent to SAFSTOR led to the stipulation of repirements for the content and frequency of facility inspections in the TSs. While the SNEF was unmanned or when specific limited tasks were performed, periodic inspections assured that conditions at the facility were essentially static or that changes to the status were acceptable. With the move to change the status of the facility to active decommissioning, the licensees have proposed modification of inspection requirements.

Activities to verify plant conditions discussed in Section B.4.a.2 of the current TSs will be performed in accordance with the requirements of applicable procedures. As noted in proposed TS 3.5.3.1, inspection requirements pertaining to verification that access points to the exclusion area are secured and that intrusion alarms are operable will continue to be performed. Each procedure will stipulate the frequency and manner in which inspection results will be documented. This is acceptable to the staff, f

Quarterly inspections and radiation monitoring activities as discussed in Sections B.4.a I through B.4.a.1 of the current TSs were performed to gain a profile of conditions at the site when the site was unmanned or when occasional activities were performed. The licensees propose to eliminate these TSs. Because the SNEF will have staff on site on a regular basis, inspections similar to those proposed for elimination would be conducted on an as- needed basis instead of quarterly. Proposed TSs 3.6.1.2 and 3.6.1.3, which pertain to requirements for procedures, would specify procedural requirements applicable to facility inspections and radiological surveys. Because this change is intended to provide a more accurate assessment of site conditinns than would be available through the current inspection program, this change is acceptable to the staff.

Since quarterly reports have been eliminated, the need to review " quarterly inspection reports" as required in Section B.2.a.2 of the current TSs no longer exists. Therefore, proposed TSs 3.5.2, which describes the responsibilities of independent safety reviewers, does not refer to " quarterly inspection reports." For the same reason, the text addressing

" quarterly inspection reports" in Section B.S.c of the current TSs, " Records," has been i

r omitted from proposed TS 3.9. Inspection results will continue to be evaluated and documented on forms associated with the appropriate procedure. Program evaluation results will be documented in accordance with proposed TSs 3.9.9 and 3.9.10. The staff concludes that these changes are commensurate with the increase in activities at the SNEF and are, therefore, acceptable.

To maintain control of changing plant conditions resulting from the full time presence of decommissioning personnel, the licensees concluded that quarterly observations were inadequate and replaced the quarterly requirements in the TSs. The staff concludes that replacement of the above-described TS requirements is acceptable.

3.3.2.3 Establishment of Soecific Controls Section B.6.a.2 of the current TSs, which requires that events affecting CV integrity be reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, has been revised in proposed TS 3.8.1 to require that events or incidents creating the potential for uncontrolled release of radioactive materials be reported in a manner consistent with 10 CFR 50.72 or 10 CFR 50.73. Because this change is consistent with the regulations, the staff concludes that this administrative change is acceptable.

l 3.3.3 Chanaes to Establish Limitina Conditions for Performina Decommissionina Activities l Decommissioning activities could result in effluent releases to the air. The licensees have '

installed a new ventilation system and have proposed TSs that establish limiting conditions for performing decommissioning activities.

Since the original, permanent plant ventilation systems are no longer functional, a temporary ventilation system (Figure 1) has been installed at the SNEF to perform the following functions:

o Provide for filtration and quantification of airborne radioactive particulate materialin effluents, e Provide for worker comfort by reducing CV temperature extremes, e Reduce the potential for confined-space restrictions by providing sufficient air volume changes, e Reduce CV interior radon concentrations that build up from naturally occurring sources and accumulate in the CV, and j e Provide sufficient face velocity at the CV/DSF (decommissioning support facility) opening to meet requirements for maintaining the air release pathway for the CV via the monitored ventilation system exhaust.

The ventilation system consists of one exhaust fan drawing air from the upper and lower portion of the CV. The exhaust fan is a 6500-cfm (184 m' per min) centrifugal unit fitted with pre-filters and HEPA filters for the removal of airborne particulates in the exhaust air.

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The exhaust fan and filtration units are located outside the CV on the north side and are ducted to the CV using the existing 17-in. (43 cm) CV ventilation penetration. The flow path is from the decommissioning support building walllouvers (or roll-up doors), through i

the decommissioning support building, 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 832 ft to 811.5 ft (253 m to 247 m). A 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 17 in. (43 cm) 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 through a short stack. The stack height and arrangement were i

selected on the basis of industrial safety considerations and to prevent the intrusion of

! debris. The stack height is not relevant to radioactive rolesse criteria for this situation. To j

indicate and monitor radioactive releases, a radiation monitor with isokinetic sampling has .

been installed downstream of the HEPA filter unit. The filtration unit v as designed and constructed in accordance with ANSI N509 and tested per ANSI N510.

The capacity of the system is sufficient to provide a face velocity at the planned CV/ material handling bay opening that ensures airflow into the CV and provides adequate i

! turnover of the CV air volume-allin accordance with industry standards. The face velocity of approximately 45 fpm (23 cm per sec) and CV air change rate of approximately 3 CV volumes per hour meet these goals. This flow arrangement provides for ventilation of the DSF and CV from low to high-contamination areas and sufficient face velocity at the l

planned CV/ material-handling bay opening to meet CV integrity goals (i.e., to prevent the inadvertent and unfiltered release of radioactive contamination or airborne radioactive material).

When activities that could generate a measurable release of airborne radioactive material are planned inside the CV and/or decommissioning support building, the ventilation system is required to be in operation. Administrative procedures require that airflow be managed to f ensure that it is routed from low- to high-contamination areas and through the ventilation I system.

i The CV dome will only be opened to permit removal of the reactor vessel, steam generator,

and pressurizer. At such times, airflow will be verified in the proper direction and other )

potential airborne generating activities will be suspended in the CV and DSF. The performance of the ventilation system will not be degraded by the method used to reseal i

the CV openings after components have been removed, i Alarms associated with the ventilation system provide indication locally and at the GPU Energy Dispatch Facility, which is manned 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day. Administrative controls are provided to ensure that proper notification and actions are taken in the event of an alarm.

During activities having the potential to produce airborne contam; nation, proposed TS 2.1.1 would require that the filtered and monitored CV/ decommissioning support building ventilation system be operated in such a manner that the release pathway is through the ventilation system exhaust. The licensees have committed to suspend activities with the I

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potential to produce airborne contamination if ventilation is lost. Proposed TS 2.1.2 would require that when the CV/ decommissioning support building ventilation system is in operation, the ventilation exhaust monitoring instrumentation will be operated

' ' nultaneously. The ventilation system will be shut down if the exhaust monitoring

. ;trumentation is inoperable.

The licensees have proposed TSs 3.5.3.1 c and d, which require monitor channel checks, source checks, channel tests and channel calibrations of the ventilation system offluent particulate monitor, and a check of the HEPA filter to verify efficiencies. These surveillances are performed in accordance with the requirements of the ODCM, which was found acceptable by the staff.

The staff concludes that the design of the ventilation system is acceptable for the conduct of decommissioning activities. The staff also concludes that the proposed TSs restrictions

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on the conduct of decommissioning activities and the operation of the CV/ decommissioning support building ventilation system are acceptable.

3.3.4 Chanoes to Extend Exclusion Area Controls to include the Decommissionino Suonort Facility (DSF)

The licensees have proposed changes to the TSs to extend exclusion area controls to the DSF because radioactive material will be handled in the DSF.

Section A.2.a of the current TSs defines the exclusion area as "that portion of the Saxton Nuclear Experimental Corporation property enclosed within the fence containing the containment vessel." The text of TS A.2.a refers the reader to Figure 1 of the TSs, which is the facility layout. The figure contains a footnote allowing the exclusion area to be as small as the CV or as large as the outer security fence. The term " exclusion area" as used in the TSs differs from the definition used in 10 CFR Part 100. The definition in the TSs is related to access control and security. Proposed TS 1.1.2 would describe the exclusion area as consisting of "that portion of the SNEC facility property enclosed within a fence and building boundaries as posted." The footnote in the proposed TSs is similar to the one in the current TSs. This change better defines the exclusion area and describes how it will be recognized. The proposed change permits the exclusion area to be defined on an as-appropriate basis and allows the exclusion area to be dynamic, thereby allowing variations that are consistent with the requirements of decommissioning activities that are in progress.

For this reason, this change to the TSs is acceptable to the staff.

Section A.2.b of the current TSs reads: "Except for authorized entry the following access points shall be maintained locked." The access points were the gate to the exclusion area fence and the CV access door. Proposed TS 1.1.3.1 states: "Except for authorized entry,  ;

access points to the exclusion area will be secured." This revision would use the terms  !

" access points" and " direct access to the exclusion area" simultaneously to accommodate additional flexibility to add or remove points of entry to the exclusion area. Using the word

" secured" allows the access points to be fixed in a closed position by means other than a

" lock." This change is acceptable to the staff because it allows the licensees additional flexibility to conduct decommissioning activities while maintaining the security of the exclusion area. I

,, l 19-Section B.4.a.2.a of the current TSs requires " verification that the locks at all entrances to the Containment Vessel exclusion area fence are locked," whereas the proposed TS 3.6.3.1.a specifies " verification that exclusion . area access points are secured at the completion of each authorized entry," and requires verifying the status of exclusion area access points.

Section A.2.c of the current TSs requires an intrusion alarm on the CV. The proposed TS 1.1.3.2 requires that the CV and DSF be equipped with an intrusion alarm that is activated whenever the site is not manned. There is Oso a requirement in proposed TS 1.1.3.3 to keep the CV and DSF secured when the site is not manned.

Section B.4.a.2.b of the current TSs requires quarterly " verification of the operability of the CV intrusion alarm." Proposed TS 3.5.3.1.b would specify that " verification of the operability of the exclusion area intrusion alarms shall be performed quarterly."

The staff concludes that the combination of secured access, intrusion alarms, administrative controls, and the nearly daily activity at the site acceptably ensure the security of the exclusion area during decommissioning.

3.3.5 Chanoes to Establish Raouirements for a Radioloalcal Environmental Monitoring Prooram. an Offsite Dose Calculation Manual. and a Process Control Prooram Changes establishing descriptions and specific requirements for a REMP, an ODCM, and a PCP are incorporated into the proposed TSs. The licensees have proposed to revise the TSs to include wording that is consistent with Appendix l to 10 CFR Part 50, Generic Letter 89-01, and the revised 10 CFR Part 20. These changes were made to identify programs for monitoring releases of radioactive material and/or estimating the typos, amounts, and radionuclide concentrations of radioactive waste produced during decommissioning.

The licensees have proposed TS definitions related to the control of licensed radioactive material, the radioactive effluent monitoring program, the radiological environmental monitoring program, radioactive waste processing, and the site boundary location. The staff finds that the proposed definitions are consistent with 10 CFR Part 20 and the guidance contained in NUREG-0472, NUREG 1301, and Generic Letter 89-01. Therefore, the staff finds that the proposed definitions are acceptable.

The ODCM is the supporting document of proposed TSs 3.6.2.1 and 3.6.2.2 and l implements the radiological effluent controls program and the REMP. Proposed TS 3.6.3 identifies the content of the ODCM and the methodology for making changes to the manual.

The ODCM contains the controls, bases, and surveillance requirements for liquid and gaseous radiological effluents. In addition, the ODCM describes the methodology and parameters to be used in calculating offsite doses from radioactive liquid and gaseous effluents. The ODCM also describes the methodology used for calculating the liquid and gaseous effluent monitoring instrumentation alarm and trip set points to meet the design objectives in Appendix 1 to 10 CFR Part 50. The ODCM also defines the requirements for the SNEF REMP and contains a list of the specific sample locations used in the REMP.

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The ODCM will be maintained at the site for use as a reference guide and training document for accepted methodologies and calculations. Changes in the calculation methods or parameters will be incorporated into the ODCM to ensure the ODCM represents the current methodology in all applicable areas. Changes to the ODCM will be implemented in accordance with the TSs.

The ODCM uses methodologies and guidelines that reflect the shutdown status of the facility and are generally consistent with the guidance in NUREG-0133, NUREG-1301, Regulatory Guide 1.109 (Revision 1), and the Radiological Assessment Branch Technical Position (Revision 1). Therefore, the staff concludes that the SNEC ODCM, Revision 0, is acceptable and that proposed TSs 3.6.2.1, 3.6.2.2, and 3.6.3 are appropriate and acceptable.

Proposed TS 2.1.3 would require an analysis to verify that release criteria have been satisfied before batches of liquid waste process affluent are released. All effluent release calculations would comply with the ODCM and, therefore, this TS is acceptable to the staff.

Proposed TS 3.6.2.1 contains requirements for the radioactive effluent controls program and proposed TS 3.6.2.2 contains requirements for the REMP. The proposed TSs will establish a programmatic controls program which will present the methodology for controlling and monitoring releases of radioactive materialin gaseous and liquid effluents, limit the dose to a member of the public to the values in Appendix l to 10 CFR Part 50, and l conduct an environmental monitoring program.

l The licensees have proposed TS 3.6.2.1 for the control of liquid radioactive effluents to unrestricted areas, which limits the concentration of radioactive material to "10 times the concentrations specified in 10 CFR 20 Part 20.1001-20.2402, Appendix B, Table 2 Column 2."

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The proposed TS for the control of radioactive liquid effluents establishes limitations on the concentrations of radioactive material released in liquid effluents to the Juniata River conforming to ten times the effluent concentration values of 10 CFR 20.1001-20.2402, Appendix B, Table 2, Column 2. The basic requirements for TSs concerning radioactive effluents from nuclear power reactors are stated in 10 CFR 50.36a. These requirements l

' Indicate that compliance with effluent TSs (which have incorporated the requirements of Appendix l to 10 CFR Part 50 and are implemented by the ODCM) will keep average annual releases of radioactive materialin liquid effluents and the resultant dose at small percentages of the dose limits for individual members of the public specified in 10 CFR 20.1301. These 10 CFR 50.36a requirements further indicate that operational flexibility is allowed, compatible with considerations of health and safety, which may temporarily result l

' in increases higher than such small percentages but still within the dose limits specified in 10 CFR 20.1301. It is further indicated in 10 CFR 50.36a that, when using operational flexibility, best efforts shall be executed to keep levels of radioactive materials as is low as is reasonably achievable (ALARA) within the numerical limits stated in Appendix l to 10 CFR Part 50.

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' Historically, nuclear power plant limits for liquid effluent concentrations were based on the values specified in the "old" 10 CFR 20.106, which referenced Appendix B, Table 11 maximum permissible concentrations. These referenced concentrations are specific values that relate to an annual dose to an individual member of the public of 500 mrem. As stated in the introduction to Appendix B of the "new" 10 CFR Part 20, the liquid effluent concentration values contained in Appendix B, Table 2, Column 2, are based on an annual dose of 50 mrom. The use of a concentration limit equal to ten times the effluent concentration values will allow the same degree of operational flexibility that was allowed by the previous limits; incorporation of the requiremen'.s of Appendix 1 to 10 CFR Part 50 and 40 CFR Part 190 dose limits into the TSs and the ODCM will assure compliance with 10 CFR Part 20.

The staff finds that the proposed TSs are consistent with the requirements of 10 CFR 20.1301,10 CFR 20.1302,40 CFR Part 190,10 CFR 50.36a, and Appendix I to 10 CFR Part 50. Additionally, the proposed TSs are consistent with the guidance in NUREG-0472, NUREG-1301, Generic Letter 89-01, and Regulatory Guide 4.8 for environmental TSs for nuclear power plants.

Therefore, the staff has determined that the proposed changes are acceptable.

In conjunction with the proposed TSs, the licensee has submitted Revision 0 of its PCP.

Proposed TS 3.6.2.3 describes the PCP and tells how changes are made to it. The PCP contains the methodology that will be used to ensure that solid radioactive wastes will be processed and packaged in such a way as to assure compliance with 10 CFR Parts 20,61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

i The staff has determined that the licensees' PCP contains the methodologies, guidance, and  !

references to plant and vendor procedures that are generally consistent with the guidance in NUREG-0133, NUREG-1301, and GL 89-01. Therefore, the staff finds that the licensees' I PCP (Revision 0) and proposed TS 3.6.2.3 are acceptable. l l

3.3.6 Chanaes to Establish Reautrements for Technical and indeoendent Safety Reviews '

and a Qualitv Assurance Proaram Section B.2 of the current TSs contains requirements for a Radiation Safety Committee and an audit function provided by GPU Nuclear. These requirements are carried into proposed TSs 3.5.4 and 3.5.5 The major changes for the Radiation Safety Committee requirements are an expansion to provide independent overview and assessment of decommissioning activities. The frequency of meetings is increased from once to at least three times per year to account for the increase in activity at the site. Section B.2.b of the current TSs discusses audits. Proposed TS 3.5.4 discusses audits with minor wording changes that do not affect the meaning of the TSs. The staff concludes that the Radiation Safety Committee and the audit functiori will continua in an acceptable manner commensurate with the increase in deccmmiss!oning activities.

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22-The licensees proposed adding sections to the TSs to identify specifications for the l responsibility, scope, and qualifications of personnel performing technical and independent safety reviews of f acility and procedure changes associated with the decommissioning of '

the SNEF. The proposed text describing the review requirements are contained in proposed TSs 3.5.1 and 3.5.2.

The review requirements incorporated in the proposed new sections of the TSs are necessary to satisfy requirements for evaluating changes and to determine the need for NRC review and approval. Proposed TS 3.5.2.5.1 specifically addresses review of those changes i performed without prior approval from the NRC under the provisions of 10 CFR 50.59. l These two new TSs establish explicit requirements for the licenseos' management to ensure .

that all proposed changes to facility systems and procedures are given full review by I independent personnel and that these changes are approved at an appropriate management  !

level. Both technical and safety reviews are required by the proposed TSs. The staff l concludes that these TSs are acceptable and that the required reviews will minimize the potential for personnel errors.

I The licensees have proposed a new TS,3.6.2.4, " Quality Assurance Program," which I states that quality assurance (QA) program requirements shall be established in a l decommissioning OA plan and stipulates how revisions to the plan will be made. During j NRC's review of the decommissioning plan, SNEC committed to developing and l Implementing during the decommissioning phase an approved QA plan that meets the i requirements of 10 CFR 71.101, and to do this before performing 10 CFR Part 71 activities, i By a letter dated November 8,1996, the licensees submitted the Saxton Nuclear Experimental Corporation Facility Decommissioning Quality Assurance Plan (OAP),

Revision O. This plan was supplemented on May 30,1997 and February 3 and 9,1998, to become Revision 1. The OAP reflects the permanently shut down, defueled status of the SNEF and commits to continue compliance with the NRC approved "GPU Nuclear Operational Quality Assurance Plan," 1000-LPN-7200.01, for the packaging and transportation of radioactive waste.

Criterion i of the " General Design Criteria for Nuclear Power Plants" specifies criteria for l quality standards. It specifies that activities performed on structures, systems, and l

components important to safety have quality standards that are commensurate with the importance of the safety functions to be performed. Further,10 CFR Part 71, " Packaging and Transportation of Radioactive Material," discusses procedures and quality standards to be applied to the packaging and transportation of radioactive material. Subpart H of 10 CFR l Part 71, " Quality Assurance," and its 10 CFR 71.101(b), " Establishment of Program,"

I states that a licensee shall apply each of the applicable criteria in a graded approach (to an extent that is consistent with its importance to safety).

As noted above, the regulations allow for quality requirements that reflect the importance of safety functions to be performed. At the SNEF, the operational phase of the plant has been over for many years. The principal remaining work activities at the site are the dismantlement and shipment of the disassembled CV, equipment, and certain systems.

Thus, with respect to the regulations, and the current shutdown and defueled status of the site and CV, the staff anticipated that the QA program requirements for decommissioning would be scaled down (graded) to reflect quality standards that are commensurate with the

b importance of the remaining safety functions that are still necessary to ensure public health and safety, and occupational safety. Thus, grading of OA requirements is appropriato for those areas that cpplied during the operational phase. However, any grading or reduction of QA requirements would not be extended to the areas involving radiation, contamination, ,

packaging, transportation, or other current conditions appropriately under the control of the l OA program.

l The staff's review of the OAP found that it contained the necessary requirements to protect public health and safety, considering the onsite activities that remain to be done. On the basis of the staff's review of the OAP, the staff has concluded that the OAP, Revision 1,is acceptable in that it conforms to the quality standards appropriate for the circumstances at the SNEF, and is commensurate with the importance of the safety functions to be performed. The staff also concludes that proposed TS 3.6.2.4 is acceptable because it requires a OAP and acceptably controls changes to the OAP.

3.3.7 Other Adrninistrative and Editorial Chances to the Technical Soecifications Section B.3.a of the current TSs addresses procedural requirements of the OA program.

Proposed TS 3.6.1.1 would include general criteria for the applicability of procedures and I would eliminate a reference to the procedure control methodology. As a co-licensee, GPU Nuclear will support SNEC decommissioning activities within its procedure programs.

Section B.3.d of the current TSs, which referenced the GPU Nuclear procedure control methodology, would be revised. GPU Nuclear acted as a contractor to SNEC before the approval of the license transfer, when a differentiation between procedure methodologies for SNEC and GPU Nuclear was necessary. Elimination of the reference is appropriate because GPU Nuclear, as a co licensee, is now directly responsible for complying with TS requirements. This revision can be found in proposed TS 3.6.1.4.

1 Section B.6 of the current TSs contains requirements to submit reports to the Region I i Administrator. Because of a reorganization of responsibilities within NRC, all responsibility for the SNEF has been transferred to the Non Power Reactors and Decommissioning Project Directorate of the Office of Nuclear Reactor Regulation. The licensees have proposed changes to TS 3.8, " Reporting," to reflect this change. The licensees have also proposed l

updating references to " telegraph" with " facsimile" to reflect changes in information technology.

Section B.6.a of the current TSs requires that written reports of any occurrence of a potentially unsafe condition that could involve the tacility or the public be submitted to the NRC within 15 days. Proposed TS 3.8.1 would sperify a 30-day period. The original 15-day period remained in the TSs after industry reporting requirements were modified in 10 CFR 50.73. The proposed revision will make reporting requirements for events at the SNEC facility consistent with current regulations.

Section B.5 of the current TSs does not stipulate a retention period for records. Proposed TS 3.9 would specify the retention period as the duration of the license. This change would make the proposed TSs consistent with established guidelines for retention of records at nuclear plants.

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24-l A review of recordkeeping requirements determined that several new types of records would need to be retained to support activities associated with decommissioning. The licensee proposed that the following records need to be included in the TSs:

j- e records of all reportable events submitted to the Commission (proposed TS 3.9.1);

e records of principal decommissioning activities (proposed TS 3.9.2);

e records of training and qualification of members of the facility staff (proposed l TS 3.9.3);

i e 1 records of QA activities required by Section 18 of the SNEF Decommissioning  !

QAP, which are classified as permanent records by applicable regulations, codes and standards (proposed TS 3.9.7); and I

e records of reviews or audits required by TS 3.5.4 are proposed for addition to the I l

[

TSs. The proposed addition includes the retention of records of reviews previously not addressed and corrects the reference to the appropriate section of the TSs (proposed TS 3.9.8).

The administrative changes proposed by the licensees in the associated sections of the TSs do not represent a relaxation of requirements. The staff concludes that the proposed administrative changes to the TSs are appropriate for the status of the SNEF, have been adequately justified by the licensees, and are acceptable for implementation.

The licensees have proposed adding TS 1.0, " Definitions." This section contains definitions of terms specific to the proposed TSs to assist in the interpretation of the TSs. Significant definitions have been specifically discussed in the preceding evaluation of the licensees' proposed TSs. The staff has reviewed the remaining proposed TS definitions and has determined that they are acceptable for the status of the SNEF and help to clarify the TSs.

The TSs contain a number of proposed editorial changes made by the licensees to correct grammar, add clarity, and improve readability. The TSs have also been completely reorganized and renumbered by the licensees. The staff concludes that these changes are administrative in nature, and are acceptable.

3.4 Pronosed Chanaes to the Facilitv License The licensees have proposed changes to the SNEF license to reflect the change in status of the facility to allow decommissioning activities.

The licensees have proposed additions to paragraph 2.C of the license which contains specific portions of the regulations to which the license is subject. The licensees have proposed adding reference to 10 CFR 50.59 to this paragraph. This proposed change is consistent with changes to the regulations that became effective on August 28,1996, making the provisions of 10 CFR 50.59 apply to nuclear power reactors that had submitted

I c ..

s' their certification of permanent cessation of operations. Because the SNEF license had been modified to allow possession but not operation of the facility before the effective date of the rule, the certification of permanent cessation of operations is deemed to have been submitted. The change to the license is consistent which the regulations and is, therefore, acceptable to the staff.

l The licensees have proposed eliminating paragraph 2.C.(2) of the license which states:

i CPU Nuclear shall not dismantle or dispose of the facility or the property occupied by the facility without prior approval of the Commission.

The purpose of the license amendment is to permit dismaritling and disposal activities to be conducted. The process for license termination is clearly described in the regulations by I

) amendments that have been made to the regulations since this condition was placed in the SNEF license in 1972. Because this change to the license is consistent with the regulations

! and the proposed changes to the TSs, this deletion is acceptable to the staff.

l l

The licensees have proposed changes to paragraph 3 of the license to remove reference to the expiration of the SNEC corporate charter and to clarify the license condition by noting that the license will continue in effect until the Commission notifies the licensees that the i l

license is terminated. These changes are acceptable to the staff because they clarify the I license condition and are consistent with the regulations.

4.0 STATE CONSULTATION

l In accordance with the regulations of the Commission, the Pennsylvania State official was 3 notified of the proposed issuance of the amendment. The State official had no comments. '

5.0 ENVIRONMENTAL CONSIDERATION

A Notice of Issuance of an Environmental Assessment and Finding of No Significant impact relative to decommissioning was published in the Federa/ Register on April 15,1998 (63 FR 18459). Pursuant to 10 CFR 51.32, the Commission has determined that decommissioning the SNEF will have no significant impact on the environment.

6.0 CONCLUSION

The staff finds that the change in facility status to active decommissioning warrants the modifications to the license and TSs described in the licensees' application. The staff concludes that the changes to the license and TSs proposed by the licensees are acceptable for the change in activities on the site.

The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding j which was published in the Federal Register on March 12,1997 (62 FR 11494). 1

! l l

l l )

F 4' .

i.

f 26-l l On the basis of the considerations discussed above, the staff has concluded that (1) there i

is reasonable assurance that the health and safety of the public will not be endangered by site activities conducted in the proposed manner, (2) such activities will be conducted in compliance with the regulations of the Commission, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the haalth and safety of the

! public.

l

7.0 REFERENCES

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

2. DOE HDBK-3010-94, Airborne 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. l

! Principal Contributors: W. Gammill, INEEL R. Carter, INEEL A. Adams, Jr.

l S. Klementowicz l E. Ford E. Connell Date: April 20, 1998 l

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