ML20154G708

Safety Evaluation Supporting Amend 8 to License TR-2

ML20154G708
Person / Time
Site:	Waltz Mill
Issue date:	09/30/1998
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20154G692	List: ML20154G687 ML20154G703 ML20154G708 ML20154G727 ML20154G743
References
NUDOCS 9810130283
Download: ML20154G708 (15)

Text

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pa aer p k UNITED STATES g j NUCLEAR REGULATORY COMMISSION e $ WASHINGTON D.C. 20066 0001

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l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 8 TO FACILITY LICENSE NO. TR-2 CBS CORPORATION ACTING THROUGH ITS WESTINGHOUSE ELECTRIC COMPANY DIVISION

1.0 INTRODUCTION

By letter dated July 31,1997, as supplemented on March 20, and July 10,1998, the Westinghouse Electric Corporation, which subsequently changed its name to "CBS Corporation acting through its Westinghouse Electric Company Division" and is now considered the licensee (see Amendment No. 7 to the TR-2 license), submitted a request for authorization to dismantle and dispose of component parts of the Westinghouse Test Reactor (WTR) and to terminate Facility Licen, No. TR-2. The licensee requests that, after the reactor vessel, the remaining internals and the biological shield are removed from the facility, all residual radioactive material be transferred to the site's material license (SNM-770), and that the TR-2 license be terminated.

2.0 EVALUATION 2.1 Introduction The staff has reviewed the licensee's application to dismantle and dispose of component parts of the WTR and to terminate the facility " Possession Only" (TR-2) license.

The application included a Decommissioning Plan (DP), response to Nuclear Regulatory Commission (NRC) questions end an Environmental Report (March 20,1998). In 1963, the NRC issued a Possession Only amendment to License TR-2. Reactor fuel has now been shipped offsite.

The staff's review considered the DP's discussion of: (a) management responsibilities and commitments to continue following applicable regulations, regulatory guides, standards, and personnel protection plans, including procedures; (b) use of appropriate equipment and instrumentation, radiation survey methods, training, personnel dosimetry, radioactive waste disposal; and (c) the final radiological survey of the facility.

Termination of the facility license requires NRC to conclude that the reactor vessel and its internals, and the biological shield have been removed and that the remaining radioactive 9810130283 980930 PDR ADOCK 05000022 P PDR

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l material has been transferred to an existing Materials License (SNM-770). The acceptance I criteria to be used by the NRC staff are stated in several guidance documents; for example, NUREG-1537 (Chapter 17) and in Regulatory Guide 1.86, " Termination of Operating Licenses for Nuclear Reactors."

l 2.2 Site and Facility Description l

The Waltz Mill site is located approximately 30 miles southeast of Pittsburgh in Westmoreland County, Pennsylvania (Fig.1). The site is located about 3 miles west of the town of New Stanton between the towns of Madison and Yukon (Fig. 2). The WTR facility is located in the northwest portion of the Waltz Mill site (Fig. 3).

l The WTR retains NRC License Number TR-2 (Possession Only). The WTR license includes the reactor structure, reactor systems, the reactor containment building, the rabb!t pump room, the sub-pile room, the polar crane, and the WTR portion of the transfer canal.

The WTR was a low pressure, low temperature, water-cooled 60 MW reactor housed in a cylindrical vapor containment structure (Fig. 4). Since a permanent shutdown in 1962, all I

fuel and some of the reactor internal contents have been removed from the reactor vessel and from the Waltz Mill site. The reactor vessel hra been drained of all water and the vessel head is secured on the vessel.

The licensee has developed a detailed DP (Ref.1) to address the activities required to terminate the TR-2 license. The activities required for license termination are removal of (1) the remaining reactor vessel internal contents; (2) the reactor vessel; and (3) the l biological shield. Following these decommissionirig activities, Westinghouse is committed l to request traasfer of the remaining residual radioactive material and WTR facilities to their SNM-770 license.

Thu Westinghouse Waltz Mill WTR Decomrnissioning Plan was prepared using Draft Regulatory Guide DG-1005, " Standard Forrr.at and Content for Decommissioning Plans for Nuclear Reactors" and the applicable regulatory requirements associated with 10 CFR 50.82(b). A fuel element failure occurred April 3,1960. A new fuel element had failed and melted, blocking the coolant channel. Fission products accumulated in th? head tank and produced high radiation levels in some plant areas. The reactor was manually scrammed. The plant was evacuated and secured. Surveys and decontamination commenced immediately. A full description of the incident and recovary action are in the Report on WTR FuelElement Failure, April 3,1960 (Ref. 3). Over the next 6 months, repairs were made to the reactor. The contamination from this event was contained and subsequent decay and further decontamination aspects have been covered by the sue

" Characterization Report."

Therefore, based on the above and on the Waltz Mill Facility " Characterization Report" l (Ref. 2), the staff concurs with the conclusion in the plan that no significant events in the operating history occurred which would inhibit the decommissioning of the reactor.

Furthermore, the staff finds this section of the plan adequate.

2

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l 2.3 Decommissioning Plan (DP) Background and WTR Management During Decommissioning l Decommissioning, as defined in the DP, will be accomplished by removal and disposal of I

the remaining reactor vessel internal contents, the reactor vessel, and the biological shield.

The remaining WTR facility components and the remaining residual components and the remaining residual radioactivity will be transferred to the Westinghouse SNM-770 Materials License. There are no radiologicallimits applicable to the transfer of structures, materials, l and equipment to the SNM-770 license, other than the radioactive materials possession l limits specified in the SNM-770 license. Prior to the transfer, the SNM-770 license will be amended as necessary to include the remaining WTR associated radioactive material l

inventory. Future use of these structures, materials, and equipment will be appropriately maintained in accordance with the SNM-770 license conditions and site procedures j controlling occupational and public exposure.

In addition to removing the reactor vessel internal contents, the reactor vessel, and the binlogical shield, decontamination and dismantlement activities may be performed on other structures and equipment located within the WTR containment building. These other activities are not required for WTR decommissioning; however, they are addressed herein as optional activities that may be undertaken under the authority of the TR-2 DP, prior to transfer of remaining residual radioactivity and WTR facilities to the SNM-770 license. The approved acceptance criteria associated with the retired facilities in the SNM-770 Remediation Plan will also be used for these other areas.

Precedent for transferring the residual radioactivity to the SNM-770 license has already been established by Amendment Nos. 3,4, and 6 to the TR-2 license. These amendments transferred previous WTR facilities to the SNM-770 license (Truck Lock Building, Facilities Operations Building, and WTR Basins).

The DP outlines the organizational structure (Fig. 5) by which dismantling and decommissioning will be managed and implemented. The direct responsibility for operational oversight of activities conducted under the TR-2 license and Waltz Mill Site Radiation Protection Program rests with the Waltz Mill Site Manager (current title is Manager, Resources and Support Operations) who reports directly to the Division General Manager (current title is NSD General Manager). The Waltz Mill Site Manager will continue to have overali responsibility for the facility and the functional groups for (a) operations; (b) engineering; (c) industrial hygiene; (d) safety; (e) security; (f) environmantal compliance; (g) f acilities support; and (h) radiation protection.

Reporting to the Waltz Mill Site Manager is the Radiation Protection Manager (current title is Industria! Hygiene, Safety and Environmental Compliance Manager) to whom the Radiation Safety Officer (RSO) reports. The RSO is responsible for the establishment and guidance of programs in radiation protection. The RSO also evaluates potential and/or actual radiation exposures, establishes appropriate control measures, approves written procedures, and ensures compliance with pertinent policies and regulations. Under RSO's l direction, health physics personnel administer the established site policy, collect samples, j perform analyses, take measurements, maintain records, and generally assist in performing l 6

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the technical aspects of the radiation protection program. The health physics staff reports directly to the RSO. The RSO will be supported by adequate staff, facilities and equipment and will hold a position with the organizational structure providlng direct access to senior management.

The Remediation Team Program Manager reports to the Waltz Mill Site Manager. The Remediation Team Program Manager will coordinate the elements of the functional groups of the Waltz Mill decommissioning organization, Reme . tion Team, and decommissioning contractors, as it applies to decommissioning activities. The Remediation Team reports to the Remediation Team Program Manager.

The existing Radiation Safety Committee required under the SNM-770 license will monitor decommissioning operations to ensure they are being performed safely and according to Federal, state, and local regulatory requirements (NRC, Environmental Protection Agency, Pennsylvania Department of Environmental Protection, Department of Transportation, etc.).

Members of this committee are appointed by the Division General Manager. The Radiation Safety Committee will review major decommissioning activities dealmg with radioactive material and radiological controls. In addition, the Radiation Safety Committee will review and approve changes to the DP that do not require prior NRC approval (10 CFR 50.59).

The staff considers the discussions and commitments in the DP appropriate and acceptable, based in part on the fact that any residual radioactive material will be transferred to SNM-770.

2.4 Current Radiological Status of the Facility A radiological characterization survey of the facility was made (Ref. 2). Exposure rate measurements were obtained for the operating floor sub-pile room, rabbit pump room, test loop shield cubicles, test loop dump tank pits, primary coolant pipe tunnel, truck lock platform, beam port platform, reactor head stand platform, reactor head platform, and the general background for " Clean Areas." Exposure rates and smears from all areas are given in the " Characterization Report" (Ref. 2). The staff has reviewed all of the dose rates identified and concludes that the report is complete; therefore, the staff considers the licensee's estimates of the radiological conditions to be reasonable, and the planning for additional radiation measurements to be acceptable.

2.5 Decommissioning Alternative The DP proposes to decontaminate down to the levels required by the NRC for termination i of the license (the Decon alternative with some transfer of material to SNM-770). The staff concurs with the licensee's plan on this method of decommissioning.

2.6 Decommissioning Organization and Responsibility I The DP identifies the overall organizational structure by which the licensee will manage the facility dismantlement leading to decommissioning. An organizational chart (Fig. 5) identifies the roles of personnel and identifies key positions for both implementation and oversight of the project. The staff judges that the licensee is adequately aware that 8

l Figure 5 WTR Decommissioning Project Responsibility Matrix ESBU Division Rebulatory Quality Systems General AiTairs Assessment Manager 5

1 Radiation Safety ',

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Committee s i

e Radiation l Safety Waltz Mill Site ~8 A(Tairs Omcer Mauger Project Director (NOTE) , , , , , , , ,

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Industnal Hygiene.

Safety and Remediation Tearr QuahD FaciH6es Program Environmental Assurance Manager Manager Compliance Manager Project Technical Lead I

i Morrison OTS Duratek Security Safety Radiation Knudsen Health

& & Safety Operations Physics Access Environmental O mcer Mana6er Mana6er Control Lead (NOTE)

NOTE: The WM Radiation Safety Omcer reports to the Industrial Hygiene, Safety and Environmental Compliance Manager and is also the Secretary of the Radiation Safety Committee.

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l experienced and highly competent staff is required to dismantle a reactor while maintaining  !

due regard to protecting the public, environment, and workers from significant radiological risk. Accordingly, the staff considers the licensee's plan acceptable in this respect.

2.7 Regulations, Regulatory Guides, and Standards The licensee acknowledges responsibility for both its and the contractor's compliance with all applicable regulations. The State of Pennsylvania Health Standards and other applicable l standards will be followed. The staff judges the licensees's awareness of its responsibilities to be acceptable.

2.8 Training and Qualifications The DP discusses the need for training (Section 2.6) the dismantlement staff in certain functions, and outlines both the functions and the key individuals who will do the training.  !

l Individuals (employees, contractors, and visitors) who require access to the work areas or a radiologically restricted area will receive training commensurate with the potential hazards to which they may be exposed.

Radiation protection training will be provided to personnel who will be performing remediation work in radiological areas or handling radioactive materials. The training w;ll ensure that decommissioning project personnel have sufficient knowledge to perform work l activities in accordance with the requirements of the radiation protection program and accomplish as low as is reasonably achievable (ALARA) goals and objectives. The principal objective of the training program is to ensure that personnel understand the responsibilities and the required techniques for safe handling of radioactive materials and for minimizing exposure to radiation.

Specifically, radiation worker training (RWT) will be required for decommissioning project personnel working in restricted areas and will be commensurate with the duties and i

! responsibilities being performed. Personnel completing RWT will be required to pass a written examination on the material presented. Completion of this training will qualify an individua! for unescorted access to radiologically controlled areas. RWT will include the following topics:

. Fundamentals of radiation,

• Biological effects of radiation,

. External radiation exposure limit and controls,

. Internal radiation exposure limits and controls,

. Contamination limits and controls,

. Management and control of radioactive waste, including waste minimization practices,

. Response to emergencies, and

. Worker rights and responsibilities.

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in addition to a presentation of the topics identified above, participants in RWT will be required to participate in the following demonstrations: '

• The proper procedures for donning and removing a complete set of protective clothing (excluding respiratory protection equipment);

• The ability to read and interpret self-reading and/or electronic dosimeters; The proper procedures for entering and exiting a contaminated area, including use of proper frisking techniques; and An understanding of the use of Radiation Work Permit (RWP) by working within the requirements of a given RWP.

Individuals whose work assignments require the use of respiratory protection devices will receive respiratory protection training in the devices and techniques that they will be required to use. The training program will comply with the requirements of 10 CFR Part 20 Subpart H, Regulatory Guide 8.15, NUREG-0041, and 29 CFR 1910.134. Records of training will be maintained which include trainee's name, date of training, type of training, test results, authorization for protective equipment use, and instructor's name.

The staff considers the DP's discussion of personnel training to be acceptable.

2.9 Radiation Protection Program The DP discusses the licensee's radiation protection program. It is clear that both industrial and radiological health and safety is considered by the licensee to be top priority for the entire dismantling and decommissioning project. The DP designates direct responsibilities and oversight functions of key positions in the decommissioning organization, and a!so commits that the DP will develop written radiation work permits for all relevant tasks that willinclude specific ALARA objectives. The DP discusses the use of self-contained systems and filters to prevent the release of airborne radioactive particles either to the rest of the reactor building or to the unrestricted environment. The staff considers that the licensee's plan places adequate emphasis on control of radiation exposures, and is acceptable.

2.10 Industrial Safety and Hygiene Program The licensee's plan discusses various specific plans to control and limit potential nonradiological risks and hazards. The staff considers these steps acceptable.

2.11 Cost Estimate and Funding The DP presents a cost estimate to complete the tasks, and cites the Financial Assurance Mechanism in place, which was reviewed by the staff and found to be acceptable. The staff finds the cost estimate and funding commitment acceptable.

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1 2.12 Dismantling and Decontamination Tasks and Schedules l

The DP presents task analysis, the schedule for completion, and radiation dose estimates for specific tasks. The DP discusses segmenting and removing radioactive components and materials, contamination control, respiratory protection, local shielding, and radwaste disposition. The radioactive dose for individual decommissioning tasks is projected in Section 3 of the DP.

The staff considers Section 3 of the plan acceptable, and based on some independent audits, the dose estimates appear to be reasonable.

l 2.13 Safeguards and Physical Security l As the fuel has been shipped offsite, the physical security task has been reduced to ensuring that excess control of the facility and the radwaste will prevent inadvertent i exposure to members of the public. The DP addresses this (Section 8) issue acceptably. l 2.14 Radiological Accident Analysis Because the fuel has been shipped offsite, the accidents discussed in the facility safety analysis report are not credible. Potential accidents (Section 3.4) have been reduced to ,

those involving inadvertent overexposures to the residual radioactive materials.

The following postulated accident scenarios have been analyzed considering the radionuclide levels and isotopic composition of components to be processed, and the anticipated decommissioning activities:

1. Dropping of contaminated concrete block / rubble;

2. Fire / explosion;

3. Canal sediment criticality and handling; and

4. Rupture of a "gh efficiency particulate air (HEPA) vacuum bag.

The components with the highest radionuclide levels were used in the accident analysis.

Therefore, accidents that were analyzed bound the radiological consequences from other accident scenarios. In evaluating the postulated accidents, conservative assumptions were made by Westinghouse and the NRC staff when data or knowledge to support more realistic analyses were lacking.

The following are the major assumptions used in the following accident analyses by Westinghouse and by the NRC staff's audit calculation:

1. No filtration of radiological effluent is assumed for the accident analyses;

2. A worst case atmospheric dispersion factor of 3.53 E-02 sec/m (R.G.1.145);

3. Dose receptor is located 100 meters from the radioactivity release point;

4. The wind speed is 1 mile per hour;

5. The atmospheric stability condition is extremely stable (Pasquill G);

6. All releases to the environment are assumed to be ground level releases; and 12

._ . .. . _. , _ . _ _ _. ~._._ _ _ _ _. _ _ _ ..__.,___ _ _ _

7. A standard breathing rate of 3.333 E-04 m /sec.

The results of the Westinghouse' accident calculations as confirmed by the NRC staff are as follows:

l

! 1. Dropping of contaminated concrete rubble / block-20 to 30 mrem total effective

! dose equivalent (TEDE);

2. Fire / explosion-small (less than 10 mrem TEDE);

l 3. Canal sediment criticality and handling-this is an optional area as defined in DP l Section 2.7. All care must be taken to ensure subcriticality of the canal l sediment as this has the potential for the highest TEDE: and

4. Rupture of a HEPA vacuum bag-small (less than 10 mrem TEDE).

V i

Because of the training of all involved personnel, the quantity of radioactive material on site, and the radiation protective program, the probability of the identified postulated

accidents is expected by the NRC staff to be very small.

2.15' Radioactive Materials and Waste Management l

The DP addresses the potential sources of solid, liquid, and gaseous radioactive wastes and l disposal. Solid wastes will consist primarily of reactor components mentioned previously-

! some demolished concrete and graphite. Items will be packaged and shipped to a licensed low-level radwaste disposal facility. The licensee is committed to follow all applicable l

regulations.

l Liquid waste may result from washing contaminated surfaces, or dust-suppression activities. All liquid disposition will either be in accordance with 10 CFR 20.2003, or by ,

solidification and treatment as low-level radioactive solid waste. Any gaseous radioactive l waste will probably be airborne particulates liberated during the demolition process. These particulates will be contained and trapped in HEPA filters that will also be disposed of with

.the other radioactive solid wastes. Other radioactive materials including waste material will

'be transferred to SNM-770.

The staff considers these aspects of the DP acceptable.

2.16 Technical Specifications The facility operating license was amended in 1963 to a " Possession Only" license. At l that time the facility licer'se contained a license condition which stated that the facility was l to be possessed as described in the " Final Shutdown Report for Westinghouse Testing L Reactor, WTR-172" dated January 25,1963. The licensee has provided a set of Technical Specifications (TS), in Appendix A to the WTR Decommissioning Plan submitted July 31, 1997, to be used during decommissioning and has agreed to an addition to these TS in the March 20,1998 letter. This addition has been included in the response to Question 8, and

!' will be considered as Section 4.2.3(5). The staff finds the Decommissioning TS as

! outlined herein acceptable, s

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2.17 Proposed Termination Radiation Survey Plan The WTR decommissioning activities will result in the removal of the reactor biological shield, vessel, and internal components. Also, decontamination and dismantlement activities of other structures and equipment associated with TR-2 may be performed under

, the provisions of the WTR Decommissioning Plan. After removal of the reactor vessel

! internal contents, the reactor vessel, and the biological shield, all remaining residual radioactivity and WTR facilities will be transferred to the SNM-770 license, where it will be addressed by the SNM-770 Remediation Plan. Upon completion, no materials covered by l

the 10 CFR Part 50 license will exist.

The method for determining that the WTR facility has met the decommissioning objectives and prerequisites for license termination will be an independent verification that the reactor vessel intemal contents, the reactor vessel, and the biological shield have been removed.

l This independent verification will be performed and documented.

The staff finds this aspect of the plan acceptable.

3.0 ENVIRONMENTAL CONSIDERATION

l The Commission has prepared an Environmental Assessment and Finding of No Significant impact (EA), which was published in the Federal Reaister on September 30,1998,63 FR 52308. On the basis of the EA and this safety evaluation, the Commission has determined that no environmentalimpact statement is required and that issuance of this amendment l approving decommissioning will have no significant adverse effect on the quality of the human environment.

4.0 CONCLUSION

Based on the NRC staff's review of the licensee's DP, it is concluded that the licensee is adequately cognizant of its continuing responsibilities to protect the health and safety of both workers and the public from undue radiologica! risk. The DP provides reasonable  ;

evidence that the licensee is prepared to dismantle the reactor and dispose of all significant reactor-related radioactivities in accordance with applicable regulations and applicable NRC guidance.' The staff, therefore, finds the licensee's plans to be acceptable.

The staff concludes that decommissioning operations can be conducted without undue risk to health and safety of the workers and the public and without causing any significant impact on the environment. The staff, therefore, finds the licensee's plan acceptaMu.

The staff has concluded, based on considerations discussed above that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed activities; (2) such activities will be conducted in compliance with the Commission's regulation; and (3) the issuance of this amendment will not be inimical to the common defense and security or the health and safety of the public.

14

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5.0 REFERENCES

1. Letter to U.S. Nuclear Regulatory Commission from A. Joseph Nardi, Westinghouse Electric Company, dated July 31,1997.

2. Westinghouse, Waltz Mill, " Characterization Report," Revision 0, February,1994.

3. Staff of the Westinghouse Test Reactor, Report on WTR FuelElement Failure, April 3, 1960, July 7,1960.

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4. Letter to T. Michaels from A. Joseph Nardi, March 20,1998 (Item 8).

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Principal Contributors: J. Miller, INEEL/LMITCO T. Michaels, NRC

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l Dated: Septenber 30, 1998 l

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