ML20217N613

Trip Rept of 940524 Meeting W/Ge Re New Powder Facility.List of Attendees & Handouts Encl

ML20217N613
Person / Time
Site:	07001113
Issue date:	08/26/1994
From:	Troup G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	Dan Collins NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML20217N558	List: IA-97-246, Forwards One Liners & Weekly Repts Re Activities Occurring within Division of NMSS IR 05000731/2008001 ML20217N553 ML20217N593 ML20217N613 ML20217N684 ML20217N704 ML20217N728 ML20217P042 ML20217P099
References
FOIA-97-246 NUDOCS 9708260270
Download: ML20217N613 (14)

Text

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e UNITED STATis

/# 5 880h NUCLEAR REGULATORY COMMISSION I YT TA.G GIA AUG 2 61994

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l MEMORANDUM T0: Douglas M. Collins, Chief Nuclear Materials Safety and Safeguards Branch i

Division of Radiation Safety and Safeguards THRU: Edward J. McAlpine, Chief Radiation Safety Projects Section .

Nuclear Materials Safety and Safeguards Branch l FROM: Gerald L. Troup Sr. Fuel Facility Project Inspector l Radiation Safety Projects Section

SUBJECT:

TRIP REPORT / MEETING

SUMMARY

- GENERAL ELECTRIC NEP NEW POWDER FACILITY (DOCKET 70-1113)

On August 24, 1994 a meeting was held at NMSS with representatives of General Electric to discuss the New Powder Facility. General Electric had proposed this meeting to provide information to the NRC very early in the project I concerning the scope of the project and the preliminary schedule so that NRC management was informed. The list of attendees is in Attachment I and a summary of the discussions, including the GF hand out, are in Attachment 2.

During the course of the meeting, certain other topics were discussed as they related to possible impacts with the licensing of the Neu Powder Facilit Ten Eyck stated that the new 10 CFR Part 70 would be issued in February,y. Ms.

1995.

Thus, it will be effective in von year, which will be approximately the same time GE will be submitting the 11censo amendment request to operate the facility. GE will have to plan for the new regulations as they move towards the license amendment request.

Mr. Burnett stated that the Secretary of Energy has advised Chairman Selin that the new Waste Vitrification Facility to be constructed at Hanford will be subject to licensing by the NRC. The schedule for the licensing of this facility will fall at the same time as the GE license amendment and license renewal. The work load for the DOE facility may impact available resources in HMSS to handle the GE licensing activities.

Mr. Pierson also pointed out that the licensing manager for GE is also responsible for B&W NNFD. The NNFD license renewal is scheduled for completion in 1995 and will require most of his time. Consideration is being given to assigning a new project manager for one of the projects, probably GE. This will result in a learning period for the new project manager.

I have had discussions with GE personnel about h2ving a similar meeting with Region 11 management. Mr. Kipp and Mr. Vaughan both expressed interest in having such a meeting, espe:ially after the contracts get formalized and the preliminary project schedule is developed. Because of other factors, this 9700260270 970019 PDR FOIA NICHOLS97-246 PDR g ; , O') 0 P l/

4 meeting will probably be scheduled for October.

During the Fuel Cycle conference call on August 23, the pre-operational inspection requirements for this type of facility was discussed. Siemens is also installing a Direct Conversion facility and is slightly ahead of GE.

Siemens has already submitted the intent to construct letter and is doing site preparation and facility (UF, cylinder storage pad) relocation. I have talked to Mr. Hooker, RIV WCF0, about the ins'ection a program and his thoughts on the {

1 subject. As more details become availa)le, an inspection outline will be developed based on discussions between Ril, RIV WCF0 and NMSS/FCOB, and expanded as more details become available.

Some of the material contained in Attachment 2 was identified as " business sensitive" which should be treated as " proprietary." Therefore, Attachment 2 has been marked as containing 10 CFR 2.790 information and should be treated accordingly, 1

i Attachments:

1. List of Attendees

2. Meeting Summary t

I A'I'TACHMENT 1 LIST OF ATTENDEES August 24, 1994 General Electric

(

C. Kipp, General Manager, Nuclear Energy Production i i

C. Vaughan, Project Manager. Licensing and Safety, New Powder i Facility l

M. Chilton, Project Manager Operations, New Powder Facility NRC- Puel Cycle Safety and Safecuards R. Burnett, Director E. Ten Eyck, Deputy Director HEC- Puel Cycle Licensina Branch R. Pierson C. Emeigh E. Flack R. Wilson M. Klasky E. Keegan NRC- Puel Ovele Ooerations Branch J. Roth C. Harmon, III NRC- Reaion II G. Troup

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ATTACHMENT 2 Meeting Summary August 24, 1994 Proiect Scope GE NEP is going to build a new facility for the production of UO2 powder using the ' direct conversion" (" dry") process.

Negotiations are still in progress with BNFL and FDFC (Framatome) for licensing of existing technology. GE is also negotiating with several " equity partners" to share the cost. GE will have at least 51% of any partnership so there is no question about l

' foreign ownership" or about the decommissioning funding arrangements. All of these negotiations are still in progress, l l although decisions are expected by September 1.

The facility will have c nominal capacity of 1,000 MTU/yr using 3 process lines, although the facility will be designed with space to install a foorth line in the future. The intention of GE is that once he " direct conversion" lines are operating and producing qualified product, the existing ADU ("weta) lines will be shutdown and decommissioned.

(

GE was very direct in that this project is intended to reduce the environmental risk from the liquid waste from ADU and the required waste lagoons. Liquid waste will still be generated from the Solvent Extraction system in URU and from UCON. However, the adry process" will reduce the production of wastes and permit the current inventory of stored waste to be worked off.

(Record Notet. Although not specifically discussed at this meeting, GE has been able to reduce the Nitrate Waste inventory by sending the waste to a local paper plant to afeed" the waste digester. The URLS project has been cleaning up the sludge in the existing Nitrate lagoons so that the nitrate liquida can be processed through a Bio-Denitrification (BDN) facility. A pilot facility has been tested and a larger facility is being planned.

The in the next CaFstage will be to develop a process system for the wastes lagoons.)

2 Facility Descrintion The ft.

new facility will be a 3 story building about 135 ft. x 155 square base projection, located on the North side of the existing the partnership FMO/FMOX building. The exact location is dependent of requirements.

One proposal is de place the new facility on the North side of the existing roadway (which will be next to the new Shipping Container Refurbishment Facility / Warehouse). Powder would be transferred in hulk containers "~. rough a special (" moderation controlled") chute to the existing press feed area.

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The by product f rom the " dry process" is HF, which will be collected in water to produce about 50% hydrofluoric acid. With a uranium content of s 5 ppm U, this product can be sold for commercial use. The facility design includes a loadirg facility for the product. GE has had preliminary discussions uith commercial suppliers cencerning the disposition of the acid.

The existing ADU process system uses " thermally disassociated" ammonia as the source of hydrogen gas for the calciner feed. The new facility will use hydrogen gas directly. This will be supplied from a gas facility to be installed and maintained by a contractor. The hydrogen facility and the gas supply system will require special consideration in the Hazards Evaluation process required by the new 10 CFR 70. Disassociated ammonia will continue furnaces. to be the source of hydrogen gas for the sintering I

i Enrichment One ADU process line (line #3) is qualified to process Uranium up to 5% enrichment. URU is also qualified for 5%. The Gadolinia Shop is only qualified for 4%.

The new lines will be desi ned and qualified for a maximum enrichment of 6%. This enr chment is presently not used anywhere but long term projections dictate that the new facility should be designed for the maximum poesible enrichment. Considerable work (and many facility modifications) will be required to make FMO ,

i acceptable for 6% enrichment. (CQMMENT: UF6 can be shipped in Model 30 B cylinders up to 5% enrichment. 6% enrichment can be shipped in Model 8-A cylinders, which are 8 inches in diameter and have a capacity of 115 kg. UF6 'I  ;

The intent is that blenders will be 1 tonne or 3-tonne and transfer containers will be 500 ):g. Powder " check hoppers" will I

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be 130 kg. hoppers cooled with n:.trogen gas. At 6% enrichment, moderation control will be VERY important.

Schedule The present (initial) schedule for the project ist Sept 1, 94 Sept- Oct

Technical decision (pick BNFL or FBFC process)

Contracts signed / implemented Nov, 94  : " Request to Construct" letter to NRC; GE is currently taking soil samples, doing radiological testing, etc. ; expect " build at risk" response Feb, 95  : begin construction Jan, 96  : License Amendment request Mar, 97  : begin Hot Start-Up and Qualification Sept, 97  : Fully operational GE has established this cchedule HUI indicated that they wish to L

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move this up at least 6 months and maybe as much as a year, meaning mid '96 for start up and late '96 for fully operational.

GE also stated that they e :pect an over-lap of about 6 months between the new facility and the existing ADU lines. Once the new i f acility is qualified to produce a satisf actory product, the shutdown and deccmmissioning of ADU will begin.

GE (Vaughan and Chilton) said that once a project schedule is  !

developed, it will be provided to Region II so we can start to j plan our inspection activities for the project. Vaughan also stated that any questions about the status of the project should '

he addressed to him.

(COMMENT! Siemens is also installing ' dry conversion" with the intention of shutting down their ADU lines. We will have to anticipate that GE will attempt to move up their schedule to beat

! Siemens to being "first" in installing the new product line, l

since they are competitors in the domestic BWR fuel line.]

i l Licensing The GE license has been in " timely renewal" since 1969. The current schedule has the license renewal for February, 1997.

There appears to be a dioconnect between the license renewal (which is based on the ADU process) and the new facility license

, f or the " dry process" .

GE is proposing that the new facility be de-coupled from the license renewal so that the facility ic not a " hostage" of renewal while FCLB/NMSS says they cannot review the renewal while the new facility is also being reviewed under a separate license amendment application.

It appears that the two license situations will be treated separately unless the new facility is licensed before renewal.

Then, a revision to the renewal application will be required, which could result in a delay in license renewal. Timing is essential.

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Nuclear Market: Craig Kipp l

Powder Facility: Ylike Chilton

License Issues: Charlie Vaughan 4

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!  ? Wilmington Site

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Media Release For more infonnadon i:ontacu FredJackson. 910*675 5201 GE announces investment in Wilmington facility WILMINGTON, N.C., October 13,1994 - GE announced plans to make a

. major facility and equipment investment at its commercial nuclear ibel plant located near Wilmington, N. C. The project involves new fuel processing equipment which will be installed in an addition to be built on the present chemical processing building. Construedon of the new facilitywill l@ in 1995 with start up in 1997.

"'Ihis muld-million dollar investment here is a c.trong indicator of GE's commitment to Wilmington, N.C.." said GE spokesman FredJackson. "Since the new processing facility replaces existing equipment, this announcement does not mean newjobs for the Wilmington GE plant. However, in the long run, this project represents an important investment in long term global corn petitiveness for GE in Wilmington.

"GE's electric power generadon businesses operate successfully throughout the world, and haw a posidve impact on the U.S. trade bcnce. The new Wilmington fuel conversion plant reflects this global nature," Jackson said.

D.e new facility will be constructed through a new corporadonjointly owned by GE andJapan Nucles: Fuel UNF). JNF isjointF/ held by CE, Hitachi and Toshiba. GE wi11 operate the plant, serving its own production needs as weu as those ofJNF which supplies theJapanese market.

A French company, Franco-Belge de Fabricadon de Combustibles (FBFC),

will provide the process technology.

GE employs some 2,400 people at its 1,600 acre site in Cutie Hayne. The plant serves the electric power generadon indusuy with nuclear fuel and components. Aircraft engine parts also are produced at the plant.

Earlier this year Wilmington was made the headquarters of GE % clear Energy's fuel business when GE moved its fuel technology and commercial organizations from SanJose, Calif., to Wilmington.

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. lo- /UL GE Wilmin;rton, NC NEP Dry Conversion Process P.O. Box 780 M/C K89

. Wilmington. NC 28402 (910) 675-5656 OUR RFJ: NRC/CMV/8002

, FILE #: 2.08.3 December 15,1994 l

Mr. E. D. Flack Licensing Branch, NMSS U. S. Nuclear Regulatory Commission Mail Stop T 8. D 14 Washington, D. C. 20555 0001

Reference:

NRC License ENM 1097, Docket 701115

Subject:

Request to Begin Construction Replacement of Uranium Conversion Facility and Equipment With reference to activities authorized by NRC license SNM 1097, GE's Nuclear Energy Production facility in Wilmington, NC, hereby applies for authorir.ation to construct the facilities necessary to replace the current UFe conversion equipment. This willinclude the construction of rso HF recovery system to be operated in conjunction with the new conversion equipment. GE will also be requesting permission to operate the facilities and equipment consistent with the current license and information to be supplied later which discusses the details of the facilities. General details of the replacement facilities and equipment are attached to this letter.

As part of this licensing action, GE will be requesting permission to re-institute the commercial sale of HF acid, which is a byproduct of the replacement conversion process, and will propose modifications to the licensed criticality safety program to place more reliance on the parameter of moderation control.

The project requires the constmetion of two new ctmetures. The first will be connected to the existing FMO X nora building face and will contain the senices and UFe conversion equipment, including ceramic treatment, to prepare nuclear ceramic grade UO2 p wder for pressing. This will be a three-story structure similar in height to the current FMO/FMO X structure with a footprint of nominally 20,000 square feet. The second facility is for the recovery and shipment of HF acid. This will be a high bay metal building with a footprint of nominally 1,500 square feet and g p 2'""&

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I will be located north and slightly west of process equipment stmeture. There will also be some rearrangement of the roadways and storage in the e.rea. The area where construction te.kes place is mainly covered by asphalt; however, some new ground will be utilized. Generallayouts are included in the attachment to this letter.

The project will result in significant improvements in our safety and environmental management programs. Atmospheric discharge pathways will be reduced, liquid discharges and associated treatment byproducts will virtually be eliminated, worker exposures will remain very low and the project is consistent with and augments the long term closure and decommissioning plans for the site. Based on the characteristics of the project, no Environmental Impact Statement, Negative Declaration or Environmental Impact Assessment is required under the exclusion provisions of 10 CFR 51.22.

The current schedule rquires constrinction to begin mid-February 1995, a detailed license submittal January 1996 and operations involving special nuclear material to begin during March 1997. GE will not process any special nuclear material until the NRC has completed their review and approved operations.

Your concurrence to proceed with construction is requested no later than mid-January 1995 as the contractors will be mobilizing during January.

GE recognizes that signi5 cant additional information must be submitted in advance of approving operations. As in the past, GE continues to be committed to working closely with the NRC to establish appropriate scope, guidance and schedules. GE shall also provide the additionalinformation required to facilitate the licensing process.

Pursuant to 10 CFR 170.31,1.A, " full cost" charges will apply to this review and a specific fee for processing this amendment is not required.

GE personnel would be pleased to discuss this matter further with yn and your staff as deemed necessary. This project is being coordinated within the regulatory agencies by the undersigned. Jim Klapproth continues as the regulatory interface for the site operations; he can be reached at (910) 675 5608. I can be reached at (910) 675 5656.

Very truly yours, GENERAL ELECTRIC C iPANY

<Y $1f a.

Charles M. Vaughan roject Manager Licensing and EHS for Dry Conversion Licensing Chronological File SNM 10J7

Mr. E.D. Flack, Licensing Branch, NMSS, USNRC 1 December 15,1994 ATTACHMENT REPLACEMENT OF URANIUM CONVERSION FACILITY AND EOUIPMENT Puroose The current UF6to UO conversion 2 lines are 25 years old and have operated safely and effectively over the years. The technology, whiO capable of producing the required product, is not state of-the-art, chenucal intensive, complex and costly to maintain effectively. The process equipment is also nearing the end of its useful life without significant investment.

This project will replace the current ADU Process conversion operations and

, equipment with new equipment and process that provides a much more efficient and environmentally friendly operation. The process is referred to as dry conversion which is a much more direct approach to the conversion of UF6to UO powder 2 for the fabrication of LWR fuel elements. Because the process is more direct, there are no chemicals added, wastes are reduced, operations are simplified, effluents are virtually eliminated and process equipment, including the safety systems, are much less complex, easier to understand and to maintain.-

This project is th'e culmination of the work GE performed at Wilmington with the GECO direct dry conversion lines and our global evaluation of conversion technology.

. LOCATION The location of the new structures are identified in Figure 1. The conversion portion of the facility is connected to the current FMO-X building on the north wall beginning north of the current UO2 powder queuing and pellet pressing area. The HF Recovery Cell, a cooling tower, overhead pipe bridge

Q Mr. E.D. Flack, Licensing Branch, NMSS, USNRC 2 December 15,1994 and a small UF6 cylinder queuing area are slightly removed to the north and west from the FMO-X buildirg and the new UF6 to UO2 conversion facility.

BUILDING CONSTRUCTION The three story, nominally 20,000 square foot building footprint housing the conversion activity and the support equipment rooms will be constructed as a compartmentalized building ising a combination of cast in-place concrete and concrete block. The design basis for wind is 120 MPH and for seismic, Zone I from the N. C. State Building Code will be followed. The compartmentalized structure is important from a nuclear safety and fire protection standpoint. The height is slightly taller than the current FMO-X building (s 14 feet higher). The roof design is such that leakage of water is highly unlikely and it slopes slightly to the west to aid in the evacuation of ,

rain water. The building will be supported by an auger cast concrete piling foundation (constructed in accordance with proper hydrogeological preservation techniques) and will include footings for a fourth conversion line, should that be necessary in the future. The exterior finish over the concrete structure will be similar to that currently used on the FMO-X building.

The HF recovery, storage and load-out building is two stories in height with a nominal footprint of 1500 square feet. The building will be metal sided, steel framed and will generally match the construction of the other buildings in the area. The facility is designed to minimize any release of HF to the environment during operation, storage or load-out. The building connects to the main process building via a pipe bridge.

The overhead pipe rack is steel construction similar to those used throughout the site.

The cooling tcwer cellis approximately 1,000 square feet and the same type construction as similar units currently used on site.

The construction will take place on the north side of FMO-X. The majority of this area is currently used as pan of a driveway and/or storage of various

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Mr. E.D. Flack, Licensing Branch, NMSS, USNRC 3

~ December 15,'1994 items and materials associated with the current fuel manufacturing operations.

Therefore, the new construction does not claim any appreciable area tl.at was .

not already being used as a part of current activities. Very minimal natural vegetation (weeds and scrub) will be removed as a result of providing an area

)

for this construction.

The present general drainage area flow pattern will be maintained and possibly enhanced by the construction.

All construction will be in accordance with all State and local building, fire ,

and safety codes.  !

' Construction traffic will be routed through a security gate on the NW corner of the controlled access area. A new staging area will be c'onstructed at this point which is a safe distance from all nuclear processing operations. This area will be used to stage the construction workers and other facility workers because its location is the most appropriate area for their assemblage.

.- BUILDING USE The new conversion building will be used for the conversion of UF6 to UO 2 powder, the powder processing steps of blending, panicle formulation and preparation as press feed, powder packing, and to house the necessary electrical, auxiliary process and HVAC equipment to support operadons.

Throughput is nominally 1000 MTU per year with the capability to expand to a nominal capacity of 1500 MTU (consistent with current capacity). The building will tie directly to the FMO/FMO-X building with appropriate powder transfer corridors, and to the adjacent Warehouse for final preparation of powder shipments.

The new HF Recovery building will be used to recover the HF produced in the conversion reaction as 50% (typical ) HF acid. Minimal storage capacity -

'and a tank truck loading area will also be part of this building.

Mr. E.D. Flack, Licensing Ilranch, NMSS, USNRC 4 December 15,1994 RADIOLOGICA L SAFETY The majority of the operations planned for the new manufacturing area are very similar to current operations and our experience with full-scale operation of the GECO conversion process. Additionally, we have reviewed the operational performance of other direct conversion techniques and find that the individual as well as the collective radiological dose will be equal to or lower than current operations. Some design enhancements may further reduce the very low radiological exposure levels to our workers. Releases to the environment should be no more than current operations and since some sources are eliminated, we are expecting a decrease. UF6 releases outside the facility are very unlikely due to a unique vaporization equipment and enclosure design.

Our plan is to apply our current radiological control program to these operations with no changes. Therefore, with respect to radiological safety, we are not currently proposing modifications to our license nor asking for exemptions to the regulations as a condition of operating this new facility.

In summary, no significant radiological safety problems are expected.

NUCLEAR CRITICALITY SAFETY Nuclear criticality safety for the project represents the only area where a licensed programmatic change will take place. The facility and process design have been structured such as to allow the safe handling of large ,

batches of UO2 powder containing low moderation. As a result, the l operations rely on the control of the parameter moderation to prohibit a criticality accident. While not new in the nuclear industry,it does represent a change in this facility's operating basis. GE will be modifying internal

• l criticality safety guidelines and the facility license to provide for this option and to clearly identify the commitments for operating safely in this manner.

The facility is designed to eliminate significant sources of moderation and to provide effective controls in places where materials that have moderating effects are required.

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_ Mr. E.D. Flack, Licensing Branch, NMSS, USNRC 5

December 15,1994 '

GE plans to utilize the basic safety system design as is used ai the Framatome Fabrication Combustible (FBFC) Pierrelatte and Romans plants.

Both facilities have significant operational history to draw from and have been thoroughly reviewed by the French CEA. GE will be conducting independent HAZOPs for the facilities and operations and willindependently verify all neutronics calculations with our approved codes and methods.

- Details of the changes in the license and the demonstrations of safety will be provided at the time the amendment information to support operation of the facility is submitted (currently scheduled in January 1996).

ENVIRONMENTAL

. Pre-Construction Phase: The area to be occupied by the. dry conversion building on the north side of FMO X was evaluated to determine if there were any pre-existing con d~ itions of environmental concern which should be addressed before construction.

'Ihe first evaluation utilized information on the F-Series monitoring wells (Refemace General Electric Environmental Report, NEDO-3166, date May .

1989) which surround the FMO - FMO-X complex, F-7, F-9 & F-10, being-downgradient from the proposed construction' location. All these wells show norma' background results with an average pH of 6.34 (high 7.14, low 5.29),

fluoride values averaging <l.0 ppm (1.0 is minimum detectable), a nitrate average of 0.40 ppm (0.02 is minimum detectable). These valves confirm the

- absence of uranium or the inorganics characteristic of the process operated in this general region of the plant.-

The FX Series of monitoring wells (on the perimeter of FMO_- FMO-X am

- usert to monitor the movement of uranium and associated inorganics that potentially could migrate from under FMO - FMO-X. This series of wells was added several years ago after some leaks were identified in the Chemical processing area of the plant.

Monitoring wells FX-7AU and FX-7AL (screened in the upper and lower

, layers of the surficial aquifer respectively) are positioned on the southwest

Mr. E.D. Flack, Licensing Branch, NMSS, USNRC 6 December 15,1994 corner of the constmetion area and are down grad}ent of any potential sources originating from the FMO - FMO X operations.

During our evaluation earlier this year, FX-7AU showed 0.46 mg/l nitrate (as N)(1992 average 0.97,1993 average 0.16), <0.05 mg/l fluoride (1992 average 0.09,1993 average 0.13) and <0.02 mg/l uranium (1992 average

<0.02,1993 average <0.02). Likewise, FX-7AL showed <0.03 mg/l nitrate (as N)(1992 average 0.03,1993 average 0.75),0.12 mg/l fluoride (1992 average 0.14,1993 average 0.13) and <0.02 mg/l uranium (1992 average

<0.02,1993 average <0.02). We also evaluated a host of otherinorganics in March 1994, all of which were in normal natural bounds for our location.

These values also support the conclusion that there is nothing in the ground waterin the region of the proposed new construction that needs consideration in the construction or operational phase.

Some 75 feet west of the construction site, monitoring well FX-3B has shown napthalene ranging from 4310 ppb to 5750 ppb with some occasional traces of benzene and ethylbenzene. The values except napthalene are below reporting levels. We believe that these organics have originated from the treated wood pilings used in the construction of FMO - FMO-X. Several years of monitoring indicates that this condition should have no adverse impact on the site or the construction of the new facilities.

The soilin the area to be occupied by the building has been evaluated. The area was divided into nominally 30' x 30' grids (25 samples from a 150' x 150' area). Grass and/or asphalt were removed from each location and then a 3" diameter auger was used to remove a 61/4" soil core. These samples were analyzed for uranium and also split to form two composite samples each from the eastern half and western half of the construction area. The uranium value average 0.92 ppm (3.28 high,0.06 low] (see Table 1). TCLP values were below the regulatory level (Table 1 - 40 CFR 261) and Gross alpha and beta values were all well below any level of significance (see Table 2).

Based on the data GE has determined that there is no pre-existing condition that would negatively influence construction or operation of the facility. All records associated with these evaluations are available for inspection.

Mr. E.D. Flack, Licensing Branch, NMSS, USNRC 7 December 15,1994 Construction Phase: All construction will be accomplished utilizing standard construction techniques. The area has been thoroughly evaluated for soil characteristics, hydrological considerations and contaminants. No unusual conditions were identified, nor were any radiological contaminants located. The auger cast concrete pilings are to be installed utilizing approved techniques. The monitoring wells on the north side of FMO X should not be affected by the work or the facility, however, we have performed a hydrogeological evaluation of the area and,if required, know where replacement wells should be located. We anticipate no problems.

All local and State codes will be complied with during the constmetion. GE and the Architect Engineering (AE) firm will secure all appropriate Federal, State and Local permits for the construction and installation operations.

The construction and modification to the facilities and operations is consistent with the General Electric Environmental Report, NEDO - 31664, dated May l 1989.

i .

i Operational Phase: The project will have a positive environmentalimpact with respect to the facility's currently approved operations and with respect to the entire fuel cycle operation. My positive impacts are as follows:

(1) Eliminates the use of nominally 2 million gallons of 20% ammonium hydroxide and in excess of 100 thousand gallons of 50% (nominal) nitric acid consumed annually from the process to convert UF 6 to UO2 and process the associated scrap.

(2) Recovers the fluoride associated with UF6 as a useable resource,30 to 50 % HF acid at s 3 PPM U).

(3) Provides for the direct recycle of clean dry scrap without any chemical treatment or processing. This eliminates the nitric acid and ammonium hydroxide used in this operation, and in some cases the added burden of solvent extraction treatment. This short process cycle also has a favorable influence on reducing inventory levels.

-Mr. E.D. Flack, Licensing Branch, NMSS, USNRC 8-December 15,1994 (4) The C. F 2byproduct associated with the treatment of fluorides in the procers liquid waste streams is reduced by approximately 830 metric tons per year. , ' proximately 70' metric tons per year of C. F2 will continue to be

_ generate J by the treatment of dilute HF (10%) produces by scrubbing and maintenance operations.

! (5) The large waste treatment, storage and disch:.rge system required for the ADU process will r.o longer be required. This eliminates approximately 80,000 gallons / day aqueous discharges including a reduction of ammonia in the process effluent. It also affords GE the opportunity to modify the operations of these facilities and reduce decommissioning challenges associated with them. '

(6) The uranium is well contained in the process equipment and the facility resulting in low worker exposures and low atmospheric discharge.

Utilizing large containers minimizes make/ break connections that generate airborne and reduces handling so that extemal radiation is kept to a minimum.

(7) Due to the nature and simplicity of the process, along with bulk handling, the amount of radioactive and chemical wastes generated as a result of operations will be significantly reduced.

(8) Designing a new facility affords the opportunity to implement the lessons learned related to effective safety and operating systemsc This will be evident throughout the facility, its operations and the documentation of the safety and operational basis.

(9) The facility will be new, of current technological design, baselined using techniques ofintegnted safety analysis, and fully conflguration controlled from final design, through constructi on, and into its operational life.

Mr. E.D. Flack, Licensing Branch, NMSS, USNRC 9 December 15,1994 MATERIAL CONTROL AND ACCOUNTING The material control and accounting plan for this facility will be based on the existing principles and commitments currently in our approved license. All of the required updates will be made to our material control, measurements, measurement controls, accounting, statistical performance evaluations and inventory procedures. There are no adverse impacts in this area and, in fact, the design and operation of the facility with the large containers, simplified recycle and reduced waste streams should result in better overall MC&A performance.

All facility specific information in the license will be updated as a 10 CFR 70.32 (c) change, and all site retained information will also be updated to l reflect the new operations when they become operational. The DIQ update l for the IAEA will be completed, as required, a minimum of 70 days before operation of the facility using SNM.

INDUSTRIAL S AFETY During the construction phase, the AE/ Contractor is responsible for the safety and health of workers on the job. GE is requiring the contracting party to have a documented safety program and a designated safety coordinator on site. Our EHS professionals will maintain cognizance of the construction work.

Industrial safety features are being designed into the facility. The AE/ Contractor has the responsibihty to meet or exceed all codes and requirements. GE is maintaining cognizance through our review and oversight proces .

State building codes, NFPA 801 and NFPA 90A, are being used as fire protection standards. Additionally, special consideration is being aoolied to exclude moderation in the event of a fire and at the same time controu.cd extinguish the fire before any critical damage takes place.

Mr. E.D. Flack, Licensing Branch, NMSS, USNRC 10 December 15,1994 Chemical safety for the process is being included in design. The OSHA process safety management (PSM) requirements will apply and be  !

incorporated into the HF operations.

l The control room is being designed so as to be inhabitable in an emergency I for a period of time necessary to bring the operation to a safe shutdown condition.

When operational, the facility will be operated in accordance with federal and state safety requirements and will be covered under the same oversight program as applies to the existing site.

EMERGENCY PREPAREDNESS .

The construction activities and their impact on current operations are being l integrated into our current plans. Proper consideration is being given to evacuation, staging and accountability. The overall site emergency plan covers the spectrum of potential emergencies during the constniction phase of the project.

One element of the stan-up phase of the project is emergency preparedness readiness. To verify that all the systems and procedures are adequate and workable, a special drill to test and verify the key requirements will be scheduled early in the initial operations (first three months of operation). All facility specific information in the Radiological Contingency and Emergency Plan will be updated to reflect the new operations when it becomes operational (10 CFR 70.32 (i)).

PHYSICAL SECURITY The construction and operational phases will be protected in accord with our currently approved security plan. We will be making operational the NW access security gate to the controlled access area (CAA). There will be some re-routing of traffic during the construction phase. After construction, we will be opening a new entry way on the east side of the CAA adjacent to the current ingress / egress point. All facility specific changes will be updated in the Physical Security Plan pursuant to 10 CFR 70.32 (e).

6 Mr. E.D. Flack, Licensing Branch, NMSS, USNRC H December 15,1994 DECOMMISSIONING The facility is being designed with ease of decommissioning as a requirement.

A number of key considerations include minimizing the amount of contaminated equipment and materials, surface finish, materials of construction and methods ofinstallation. As a result, GE does not expect any adverse impact on the task of decommissioning this facility.

l This facility adds no new challenges to the overall task of decommissioning.

It does reduce the longer term challenges because it brings to a halt the ADU process of converting UF6 to UO2 much earlier than anticipated in previous decommissioning studies for the site and, therefore, reduces the burden and complexity that the chemical processes would have had on the longer term.

This is a positive benefit to the task of decommissioning.

The decommissioning files and records associated with the facility will be updated so that in the future all required information is available.

o t

Mr. E. D. Flack, Licensing Branch, NMSS, USNRC Decemher 15,1994 TALLE1 DRY PROCESS ARLA (30'X 30' GRIDS) i i

SolL SAMPLES FOR URANIUM l 346009 346008 346007 346006 346005 0.63 1.05 1.37 1.53 0.32 25DP 24DP 23DP 22DP 2tDP 346004 346003 346002 346001 346000 1.46 1.00 0.25 1.46 1.05 20DP 190P 18DP 17DP 16DP 345999 345998 345997 345996 345995 1.56 0.48 0.42 0.70 0J7 IfDP 14DP 13DP 12DP 11DP 345994 345993 345992 345991 345990 0.83 0.63 0.55 0.88 3.28 10DP 9DP SDP 7DP 6DP 345989 345988 345987 345986 345915 0.06 0.84 0.44 0,76 0.93 5DP 4DP 3DP 2DP IDP KEY: Sample ppm U Grid #

Grid Average: 0.92 Grid Std. Devt 0.643 i

- Mr. L . Flack, Licensing Branch, NMSS, USNRC '

December 15,1994 TABLE 2 DRYPROCESSAREA SOIL TCLP COMPOSITESAMPLES TEST DESCRIPTION East # 1 East # 2 West # 1 West # 2 Detection Units Limits Arsesdc(total)TCLP <0.1 < 0.1 <0.1 < 0.1 0.1 mg4 iter liarium (total)TCLP 0.56 0.56 0.27 0.25 0.05 mgniter Cadmium (total)TCLP 0.08 <0.05 < 0.05 < 0.05 0.05 mg4 iter Chromium (total)TCLP < 0.05 <0.05 < 0.05 < 0.05 0.05 mg4 iter Lead (total)TCLP 0.07 0.06 0.07 0.05 0.05 mgditer Mercury (total)TCLP <0.04 < 0.04 < 0.04 < 0.04 0.04 maniter Selenium (total)TCLP < 0.1 < 0.1 < 0.1 < 0.1 0.1 mg4 iter Silver (total)TCLP < 0.05 < 0.05 <0.05 <0.05 0.05 mediter Gross Alpha 1.0 +/- 0.4 0.8 +/- 0.4 0.5 +/- 0.4 <03 03 pCi/ gram Gross Beta 1.5 +/- 0.2 1.6 +/ .2 0.6 +/- 0.2 < 0.2 0.2 _ pCi/ gram

-_ W .:

April 7. 1995 General Electric Company ATTN: Mr. C. P. Kipp, General Manager- l GE Nuclear Energy Production P._0. Box 780 Wilmington, NC 28402 i

SUNECT: NOTICE OF VIOLATION (NRC INSPECTION REPORT NO. 70 1113/95-02) l_ ' Gentlemen:

This refers to the inspection conducted by E. D. Testa of this officef on March 6 8, 1995.

The inspection included a review of activities authorized for your Wilmington facility. At the conclusion of the inspection, the findings were enclosed discussed with those members of your staff-identified in the report.

Areas examined'during the inspection are identified in th$ report. Within these areas, the inspection consisted of selective examinations of procedures and representative records,-interviews with personnel, and observation of activities in progress. The purpose of the inspection was to determine whether activities authorized by the license were conducted safely and in accordance with NRC requirements. -

Based on the results of this inspection, certain of your activities appeared to be in violation of NRC requirements, as specified in the enclosed Notice of Violation (Notice). The enclosed Inspection Report also identifies activities

_.that violated NRC requirements that will not be subject to enforcement -action because your efforts in identifying and correcting the violation meet the criteria specified in Section VII.B. of the Enforcement Policy.

You are required to respond to this letter and should follow-the instructions speci_fied in the enclosed Notice-when preparing your response. In your response, you should document the specific actions taken and any additional actions you plan to prevent recurrence. . After reviewing your response to this Notice, including your_ proposed corrective actions and the results-of future inspections, the NRC will determine whether further NRC enforcement action is necessary to ensure compliance with NRC regulatory requirements.

t In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter and its enclosures-will be placed in the NRC-Public Document Room.

'The responses directed by this letter and the enclosed Notice are not subject to-the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of-1980,-Pub. L. No.96-511.

wd _

d GE 2

Should you have any questions concerning this letter, please contact us.

Sincerely, original signed by D. R. McGuire Douglas M. Collins, Chief Nuclear Materials Safety and Safeguards Branch Division of Radiation Safety and Safeguards Docket No. 70-1113 License No. SNM-1097

Enclosures:

1. Notice of Violation

2. NRC Inspection Report cc w/encls:

J. F. Klapproth, Manager , ,

Fuels and Facility Licensing  !

General-Electric Company 1 P. O. Box 780, Mail Code J26 i Wilmington, NC 28402  !

Dayne H. Brown, Director Division of Radiation Protection N. C. Department of Environment, Health & Natural Resources P. O. Box 2'/687 Raleigh, NC 27611-7687 Distribution w/encls:

E. McAlpine, Ril G. Troup R. Bellamy, RI G. Shear, RIII C. Cain, RIV F. Wenslawski, RIV PUBLIC Distribution w/o encis:

License Fee Management Branch

• 8ND TO PUBttC DCOUMENT ROOM) YES NO OFFICE Ril:ORSS Rit:ORSS SiONATURE NAME EDieste EJMcAlpine DATE 0417195 04 f 7195 04 # 195 04 # 195 04 / 195 04 1 19E coPYP 6ES) NO dEh NO YES NO YES 3 YE3 NO YES NO OFFICIAL RtCORD COPY 00CLMENT NAME: G:\0R55\RSPS\GE502RPI.EDT

10TICE OF VIOLATION General Electric Company Wilmington,- NC Docket No - 70 1113 License No. SNM-1097 During an NRC inspection conducted on March 6 8, 1995, a violation of NRC requirements was identified.

In accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2, Appendix C, the violation is listed below:

Safety Condition No. 5-1 of Special Nuclear Material (SNM) License Number 1097 (SNM-1097), requires that licensed material be used in accordance with statements, representations, and conditions contained in Part I of the Application dated October 23, 1987, and supplements and letters thereto.

Part 1, Section 2.2.1.3 of the licensee's Application for License No.

SNM-1097 specifies in part, that a) the Criticality Safety Function will establish a criticality safety control program including criteria, i procedures and training b) analysis and approval of proposed changes in process safety. conditions and processing equipment involving criticality Part 1, Section 2.7.3 requires in part that the changed activity will  !

not be initiated until the nuclear safety analysis demonstrating safety of the activity has been completed, a preoperational inspection has been conducted to verify that the installation is in accordance with the nuclear safety analysis, and appropriate procedures and/or instructions are in place. The results of these analysis are documented in nuclear safety reviews and maintained for the period of time they-remain applicable and in accordance with the records retention requirements of this license.

The Criticality Safety Analysis of FM0X Combustible' Trash and Waste Handling performed on or about April 6,1981 specified that the yellow polyethylene bags used for trash would.either separate from their holder or the bag would fail if a safe batch mass were placed inside.

Contrary to the above, on March 6.- 1995 the bags would not pull from the holder nor would they fail if a safe batch limit as authorized in NEDE-24697 Appendix A was placed inside the bag in the holder.

This is a Severity Level IV violation (Supplement VI).

Pursuant to the provisions of 10'CFR 2.201, General Electric Company is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: -Document Control Desk, Washington, D.C. 20555

'with copies to the Regional Administrator, Region 11, and the Chief, Radiation Safety Projects Section, Region ll, within 30 days.of the date of the letter tranmitting this Notice of Violation (Notice). This reply should be clearly Enclosure 1 pge. av w urv

2 marked violation: as a " Reply to a Notice of Violation" and should include for each disputing the (1)violation the reason for the violation, or, if contested, the basis for (2) the corrective steps that have been taken and the results achieved, (3) the corrective steps that will be taken to avoid further violations, and (4) the date when full compliance will be achieved. Your i response may reference or include previous docketed correspondence, if the ('

correspondence adequately addresses the required response, if an adequate reply is not received within the time specified in this Notice, an order or Demand for Information may be issued as to why the license should not be modified, should notsuspended, be taken. Where or revoked, or why such other action as may be proper good caust to extending the response time, is shown, consideration will be given ,

Dated at Atlanta, Georgia this ry td day of April 1995 l

l

~

- '= 7'mO C80g*> UNITEo STATES o

• ' ,s - NUCLEAR REGULATORY COMMISSION I V'

-9 REGION ll

?-

.j 101 M ARIETTA STREET. N.W., SUITE 2900 ATLANTA. GEORGIA 303234190

.( Ox]

_ Report No : .

70-1113/95 i j

t. icensee: General Electric Company Nuclear Energy Production P. O. Box 780 Wilmington, NC 28402 Docket-No.: 70-1113 License No.:. SNM-1097 Facility Name:

! Nuclear Fuel and Components Manufacturing Plant inspection Conducted: March 6-8, 1995 l

Inspector: *

\ Y 1...aau T. D. Testa, ' P.E. , Senior \ fuel Faci Mty

,s

' -' 7 9 3 Project inspector- Date Signed Approved by:

Yi 'd Cll.

E. J. ficAlpine, Chief \

4!7!93 Date Signed Radiation Safety Projects Section Nuclear Materials Safety and Safeguards Branch Division of Radiation Safety and Safeguards

SUMMARY

Scope:

This routine, unannounced inspecticn involved onsite review of the licensee's nuclear criticality _ safety program. In addition previously identified inspection findings were re, viewed. follow-up actions related to Results:

-Within the scope of the inspection, two violations-(one cited and one noncited) (VIO_95-02-01 and NCV 95-02-03 Paragraph 3 and 5) were identified.

_ ane new item for review during-future inspections was identified concerning the licensee's Paragraph 3). use of multiple-document control stamps (IFI 95-02-02 Enclosure 2 EW

?/l'

2 The licensee's corrective actions, taken in response to various previously identified violations and follow-up items, appeared adequate. The violations and follow-up items that were closed as a result of this inspection are identified in Paragraph 3 (a through e).

I It was noted positions.

supervisory that the licensee had made some changes in their management and '

Housekeeping in the facility was very good.

l 4

4 44

1 REPORT DETAILS

1. Persons Contacted

• D, Barbour, Supervisor, Radiation Protection J. Bradbur

• D Brown, y, Regulatory Team

• D.

• R.

Dowker, Team Leader, Operations Support-Chemical PLTeam Lea Foleck, Sr. Specialist, Licensing Engineering

• T. Hauser, Manager, NQA/EHS

• J. Huffer, Criticality Safety

• B. Kaiser, Manager, Fuel Fabrication

• R.

i

• J. Kennan, Program Manager, Compliance Auditing

! *A. Klapproth, Manager Fuel Facility Licensing i

Mabry, Nuclear Safety Engineer, Nuclear Safety Engineering

• 0. McCaughey, Manager, Fuel Quality

• R.

• L. Patterson, Team Leader, Fuel Fabrication Production Paulson, Program Manager, Criticality Engineering

• J. Pierce, Senior Engineer

• S. Selby, Team Leader, 00 Production Team

*W.

i Sependa, Manager, Chemical Product Line

• M. Shea, Criticality Safety

• G. Smith, Leader, FM0 Maintenance .

• D. Snell, Manager (Actin

• H. Strickler,-Manager, g), Nuclear Energy Production

• C, Tarrer, Manufacturing Engineer-Environmental Protection & Industrial- Safety

• K. Theriault, Team Leader, URV

• T.

Winslow, and Manager, Emergency Preparedness, Security, Material Control Accountability Other personnel.

office licensee employees contacted included engineers, technicians, and

• Denotes those present at the exit interview conducted March 8, 1995

2. Operations Reviews (88020)

Safety Condition No S-1 of Special Nuclear Material (SNM) License Number 1097 (SNM-1097), requires that licensed material be used in Part I of the Application dated Octoberaccordance with statements, represe 23, 1987, and supplements and letters thereto, The inspector toured the manufacturing facility and observed heavy yellow trash bags in use in many areas. These btgs were approximately 30 gallon in size and secured in a holder stand with a locking ring clamp arraignmeat. The bags are used for combustible and noncombustible trash throughout the facility. The inspector observed that the bags were held in position in a very secure manner moderate force. and would not come free from the holder using The inspector requested that the licensee test the bajs to ensure that they would either come free or the bags 1

2 would separate (tear) with less than the safe batch limit as authorized in NEDE 24697, Appendix A. This is approximately 25 Kg at 4.00 percent enrichment. The licensee tested the holder and bags to about 35 Kg and found that the bags and holders did not ensure that a geometrically safe container could be assured. The i inspector reviewed the Criticality Safety Analysis of FM0X Combustible and Noncombustible Trash and Waste Handling performed April-7, 1981 in which criticality safety was based on a combination of safe batch (administrative controls) and geometry controls. The licensee believes that these strong yellow bags have been in use for three or more years. The use of these strong

' yellow trash bags in this type of holder does not provide assurance that the safe geometry control can be assured in all credible events. The licensee immediately removed the bags from

- the holders and placed them in free form on the floor pending a reevaluation of the holder and bags. A temporary operating

instruction (TOI) was issued covering the use of the bags and holders. The licensee was informed that the use of these bags in l the holders without a criticality safety analysis was a violation ,

l of Part 1, Section 2.2.1.3 of the Application for License No. SNM 1097 which specifies in part, that a) the Criticality Safety Function will establish a criticality safety' control program including criteria, procedures and training b) analysis and approval of proposed changes in process conditions and processing equipment involving criticality safety. Section 2.7.3 requires in part that the changed activity will not be initiated until the nuclear safety analysis demonstrating safety of the activity has been completed, a preoperational inspection has been conducted to verify that the installation is in accordance with the nuclear safety analysis, and appropriate procedures and/or instructions are in place. The results of these analysis are documented in nuclear safety reviews and maintained for the period of time they remain applicable and in accordance with the records retention requirements of this license.

Failure to maintain geometry control when the heavy yellow polyethylene trash bags were installed in the trash bag holder.

1113/95 02-01). (Violation (VIO) 70-The inspector also toured the facility control rooms and found several examples of facility drawings with different configuration stamps. The licensee uses several stamps to designate controlled drawings.

Revisions associated with configuration control are in active evolutior, and the licensee has an evolving configuration control program The inspector informed the licensee that the resolution of the use of multiple configuration stamps on drawings would be trackad as an inspector follow-up ltem.

Resolution of the use of multiple configuration stamps on controlled drawings in the control rooms (Inspector follow-up Item (IFI) 70-1113/95-02-02).

Within the scope of the inspection one violation and no deviations were observed.

s __

3 3.

. Status on Previously Identified Inspector items (92701, 88015, 88020) a.-

(Closed) IFl 70 1113/94-011-01: Review the licensee's plans and actions related to the acid leak Unusual incident Report (URI).

The inspector reviewed the acid line material selection and 1

{

physical of relocation the newly andlines.

relocated selectively observed the physical location  !

The licensee completed their walkdown of the system on February 27,1995. The inspector informed licensee representatives that the item would be closed.

b.

(Closed) IFl 70-1113/94-009-01: Review corrective actions for the l-l- oxidative furnace in the Uranium Recovery Unit. This item

! involved the corrective actions taken by the licensee following a l jam of boats in the oxidative furnace which occurred on June 13, 1994 and was reported by the licensee and recorded in the NRC database as event number 27386. The inspector ruiewed the licensee's corrective actions which included: redesign of the furnace muffle with a slightly lower profile and corrugated top and bottom, modification of the procedure relative to how to remove furnace boat jams, inspection of the muffle on a six-month

!. frequency, updete of the -Technical Report, and revision of the nuclear criticality safety analysis and associated nuclear safety requirements.

The inspector toured the area with the licensee, inspected the visable corrective actions and found the corrective actions to be adequate. The inspector informed licensee representatives that the item would be closed. '

c. (Closed)'IFl 70-1113/94-003-02:

outside nuclear criticality safetyReview audit. corrective actions on This item involved the level of documentation available to support the use of the Oak Ridge composition for concrete in the nuclear criticality analysis calculations for the separations between slab tank. The licensee contracted for in situ measurement of the hydrogen content of the concrete separations. Multiple measurements were performed which yielded a slightly lower hydrogen content than the Oak Ridge concrete composition. With the slab tanks in alternatin positions (slab tanks between every other concrete wall)g, the licensee's calculation produced a K,n of 0.84175 (assuming optimum moderation and taking credit for the rebar in the concrete walls).

This met 'the license requirement of K,n .of less than 0.90 unoer normal conditions. The inspectar informed licensee representatives that the item would be closed,

d. (Closed) VIO 70-1113/93-012-03: Rotary Pellet Press 4B installed without a hole being drilled in the sump cover plate as required by the nuclear criticality safety -analysis. The inspector visually observed and physically verified that the 4B press has holes drilled in the-sump cover as required by the nuclear U criticality safety analysis. The inspector also reviewed the licensee's January 27, 1995 Document titled Revalidation of the Criticality Safety Basis for the Ceramic Area--Closure of

1 4

' Ril-93-12-02 (03). The inspector informed licensee representatives that the item would be closed.

i

e. (Closed) IFI 70-1113/93-010 01: Follow-up on Administrative mechanism to keep safety-technical reports current. The inspector reviewed the licensee's Procedure 10-10 titled Configuration Management Program - Fuel Manufacturing, Revision 0, dated

, June 10, 1994. The procedure describes the configuration management program which is designed to ensure that controlled documentation is updated. The document also establishes the responsibilities and methods to be used for updating the controlled documentation. The inspector selectively reviewed several of the recent change requests and the revisions and had no further questions. .The inspector infcrmed licensee representatives that the item would be closed.

4. Organization Changes (88005)

a. The new Manager Chemical Product Line assumed the position on January 30, 1995. There are no license requirements for this position. The inspector noted that the individual has extensive experience in progressively more responsible nuclear project management.

b. The license requirements (Part I, Chapter 2, Section 2.2 Organizational Responsibility and Authority and Section 2.5 Personnel Education'and Experience Requirements) for the recently appointed Program Manager Criticality Safety Engineering were reviewed. The individual arrived onsite November 1994 and assumed the' program manager position on January 30, 1995. Indoctrination Report No. 1 dated January 3, 1995 was reviewed and the inspector noted that the individual has made satisfactory progress on the outlined ' orientation schedule'. The educational, training and experience of this individual meet the license requirements.

Within the scope of the inspection, no violations or deviations were identified.

5. Nuclear Criticality Safety (88015 and 88020) 2 Facility Changes and Modifications The licensee reported a loss of control under procedures developed pursuant to NRC Bulletin 91-01 on January 31, 1995 and the NRC-recorded the incident as event number 28310. The incident involved discovery that the active engineered controls associated

-with termination of inputs to the V-225 tank (favorable geometry) upon' detection of high concentration. The automatic shutoff of steam to the head-end concentrator remained effective. The licensee's nuclear safety analysis had established the requirement for two-controls for the tank -- favorable geometry and concentration control (280 gram uranium per liter). The

5 licensee's investigation team could not determine the root cause of the incident and the team could not verify that the concentration control had ever been implemented. The team determined that there had been no enforcement to ensure all the necessary paperwork was complete and accounted for, no correction to the NSR/R was made, and no in-plant physical inspection to verify that actual active engineered controls were in place and functional had been accomplished.

Tne licensee established both short and long term corrective actions. The short term actions which were taken prior to restart of the process were: 1) comparing the active engineered controls list with NSR/Rs that applied to URV to verify that the requirements were met, 2) physically decoupling V-225 from the nitrate waste process and V-103, 3) changing the nuclear safety requirement liter, for V-225 and V-244 from 280 to 350 grams uranium per

4) adding a double block and bleed valve between V-27.5 ud the head end concentrator and interlock to close at a density of 1.37 grams per cubic centimeter, and 5) modify the distributed control system to support the above changes. The longer term corrective actions were: 1) evaluation of all completed NSR/R changes between May 28, 1991 ar.d December 31,'1994 to ensure adequate documentation, 2) review of current configuration management system to assure adequacy, 3) conduct roundtable training sessions with all Chemical Product Line personnel on the incident, and 4) review the event with Chemical Product Line staff.

The incpector reviewed the corrective actions and changes that had been made to the head-end concentrator portion of the URV process as a result of a loss of concentration control with the licensee.

The changes made by the licensee involved a double block and bleed valve combination being installed between V-225 and the head-end concentrator in conjunction with other piping modifications. The inspector noted that the licensee had completed the change in accordance with internal procedures associated with change control and license requirements. The inspector walked down the changes, discussed them with the cognizant engineer and found them to be as specified on the as-built drawings. The inspector concluded that the facility modification would effectively control the concentration in the V-225 tank.

The inspector informed the licensee that the failure to maintain controls required by the safety analysis was a violation of regulatory requirements. This violation, however, will not be subject to enforcement action because the licensee's efforts in taking immediate action to correct the violation met the criteria specified in Section Vll.B of the Enforcement Policy (NCV 70-1113/95-002-03).

O

b. Criticality Calibrations and Monitoring System 1 The inspector reviewed the licensee', W iementation of a criticality accident monitoring system required by 10 CFR 70.24.

The licensee's program involves: 1) daily checking of the control terminal and panel indicator lights for indication of problems,

2) monthly source response checks 0 all detectors equip an internal check source and testini of the alarm horn, ped and with

3) yearly responn checks l with an external gamma sour (ce.includin, portable backup detectors) '

The inspector roviewed the licensee's procedure, NSI No. 0 4.0, Rev 33, dated December 8, 1994, and data covering the past two years. The inspector noted that the most recent annual check had been performed during the period of December 5 7, 1994 and that the previous check had been performed during the period of December 7 8, 1993. The inspector also noted that the licensee was taking prompt action when in' ations of malfunction were observed. On January 4, 1995 .ag a routine test, the licensee i

got indication that the detet...,r at the " Elephant Gun", DAM 22 had an alarm and light that was not working. A work order was immediately processed (WO No. 95 nl549 00) and the system was checked. The system was founo to te operating satisfactorily and the licensee concluded that the indication of trouble were false i

1 .

and due to the test individual pressing the wrong test keys.

I

! Within the scope of the inspecU on, no violations or deviatic.ns were identified.

6. Exit Meeting (99701)

The inspector met with licensee representatives indicated in Paragraph 1 at the conclusion of the inspection on March 8,1995, Ihe inspector summarized the scope and findings of the inspection. Although

! proprietary documents and processes were reviewed during the inspection, the proprietary nature of these documents is not reflected in this report. Dissenting commeMs were not received from the licensee.

Typ.g item Number Status Description and Reference IFI 70 1113/94 011-01 Closed Review the 1.icensee's Plans and actions related to the scid leak Unusual incident Report (Paragraph 3).

IFl 70 1113/94-009-01 Closed Review Corrective Actions for URV 0xidation Furnace (EN27386)

(Paragraph 3).

IF! 70 1113/94 003-02 Closed Review Corrective Actions on 04 side Nuclear Criticality Safety Au'Jit (Paragraph 3).

7 11ng item Number Statys Qpscription and Reference (continued)

V10 70 1113/93 012 03 Closed Rotary Pellet Press 40 installed without a hole being drilled in sump cover plate as required by the nuclear criticality safety analysis (Paragraph 3).

IFl 70 1113/93 010 01 Closed Follow up on Administrative mechanism to keep safety analysis reports current (Paragraph 3).

VIO 70 1113/95 002 01 Open Failure to maintain geometry control when the heavy yellow polyethylene trash bags were installed in the trash bag holder (Paragraph 2).

IFl 70-1113/95 002 02 Open Resolution of the use of multiple configuration stamps on controlled drawings in control rooms (Paragraph 2).

NCV 70 1113/95 002 03 Closed Failure to maintain control specified in the nuclear criticality safety analysis (Paragraph 5).

l

I ol REGION 11 4178 audit et>0af I

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-June 2, 1995 General Electric Company ATTN: Mr. C. P. Kipp, General Manager i

GE Nuclear Energy Production P. O. Box 780 Wilmington, NC 28402

SUBJECT:

NRC INSPECTION REPORT NO. 70 1113/95 04 Gentlemen:

This refers to the inspection conducted by W. B. Gloersen of this office on May 1-5, 1995. The inspection included a review of activities authorized for your Nuclear Energy Production facility. At the conclusion of the inspection, i the findings were discussed with those members of your staff identified in the  !

report.

Areas examined during the inspection are identified in the4 report. Within these areas, the inspection consisted of selective examinations of procedures and representative records, interviews with personnel, and observation of activities in progress. The purpose of the inspection was to determine

' whether activities authorized by the license were conducted safely and in accordance with NRC requirements.

Within the scope of the inspection, violations or deviations were not identified.

In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this_ letter _ and its enclosure will be-placed in the-NRC Public Document Room.

Should you-have any questions concerning this letter, p1 ease contact us.

Sincerely, original signed by D. M. Collins Douglas M. Collins, Chief Nuclear Materials Safety and Safeguards Branch._

Division of Radiation Safety and Safeguards Docket No._70-1113' License No. SNM-1097

Enclosure:

HRC Inspection Report cc w/ enc 1: (See page 2)  :

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GE 2 cc w/ encl:

, r. Klapproth, Manager fuels and facility Licensing General Electric Company P. O. Cox 780, Hail Code J26 Wilmington, NC 28402 Dayne H. Brown, Director Division of Radiation Protection N. C. Department of Environment, Health & Natural Resources I P. O. Box 27687 {

Raleigh, NC 27611 7687 Djitribution w/ enc 1:

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Report No.: 70 1113/95 04 Licensee: General Electric Company Nuclear Energy Production P. O. Box 780 Wilmington, NC 28402 1 Docket No.: 70 1113 License No.: SNM 1097 Facility Name: General I.lectric Company Nuclear Energy Production Inspection Conducted:

May15,1)95 Inspector:/[

% B. Gloersed

@# b/// N papSigned Approved by: wf

. McAlpine, Ch ef 6!C/fa' D6te Signed Radiation Safety Projects Section Nuclear Materials Safety and Safeguards Branch Division of Radiation Safety and Safeguards

SUMMARY

Scope:

This routine, unannounced inspection was conducted in the areas of radioactive waste management, including radioactive liquid and airborne effluents, environmental protection, previously identified inspection findings, and licensee response to NRC Bulletin 94 02.

Results:

The licensee's organization to implement the required environmental protection and radioactive waste management programs had remained stable and had not changed significantly since the last inspection. The licensee's remediation and monitoring program for the residual contaminated soil and groundwater beneath the FM0/FM0X Building com) lex continues. The groundwater chemistry conditions have remained stable tarough the monitoring periods.

The licensee had implemented an effective program to monitor and control liquid and gaseous radioactive effluents and to maintain the activity released in those effluents to concentrations which were within the limits specified in 10 CFR 20 for release of radioactive material to unrestricted areas.

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REPORT DETAILS

1. Persons Contacted Licensee Employees i

• D. Barbour, Radiation Protection Coordinator

• J. Bradberry, Project Manager, Regulatory Team

• D. Brown Team Leader, Environmental Processes Team - Chemical Product Line (CPL) .

• T Crawford, Senior Engineer, Environmental Protection  !

• D. Dowker, Team Leader, Operations Support - CPL N. Gutermuth, Industrial Safety Specialist D. Hassler, Supervisor, Heating, Ventilation and Air Conditioning (HVAC)

• B. Kaiser, Manager, Fuel Fabrication Product Line

• R. Keenan, Program Manager, Compliance and Auditing

• A Habry, Engineer, Nuclear Safety

• R. McGowan, Team Leader, Uranium Recovery Lagoon Sludge / Waste Treatment

• S. Hurray, Manager, Radiation Safety

• R. Pace, Manager, CHEHET Lab .

S. O' Conner, Manufacturing Engineer

• H. Shaver, Engineer, Nuclear Safety

• G. Smith, Team Leader, FHO Haintenance

• H. Strickler, Manager, Environmental Protection and Industrial Safety

• F. Walker, Manager, Shipping and Traffic

• T. Webb, Specialist. Environmental Protection

• P. Winslow, Manager, Emergency Preparedness, Security, Material Control and Accountability Other licensee employees contacted included engineers, technicians, and administrative personnel.

• Attended exit interview on May 5, 1995.

2. Status of Previously-Identified Inspection findings and Followup Items (92701,92702)

a. (Closed) Inspector Followup Item Compare analytical results for gross alpha (IFI) and 70-1113/94-06-01:

gross beta analyses of a final process lagoon sample and two effluent stack samples.

As part of the NRC confirmatory measurements program, the inspector requested the licensee to provide to the NRC a liquid sample from the discharge of the final process lagoons and two effluent stack samples (one from the chemical area recirculation unit No. S49 (CRCK) and one from the calciner discharge No. 2 (CALC 2) stack) so that a comparison could be made between the NRC and licensee results. The purpose of these comparative measurements was to verify the licensee's capability to measure quantities of radionuclides accurately in both the liquid and gaseous waste streams. The licensee utilized the services of a

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j j 2 l contract laboratory to perform the liquid effluent sample analysis. Analytical services for the NRC were provided by the DOE's Radiological Environmental Sciences Laboratory (RESL) in Idaho falls, Idaho. Attachment 1. Table 1 provides a comparison of the licensee's results to the NRC's results for the samples i noted above. Attachment 2 provides the criteria for assessing the j agreement or disagreement between the analytical results.

I 1

4 As indicated in Attachment 1. Table 1, the licensee's results for i the liquid effluent sample and one of the stack samples were

!- within-the comparison ratio limits as specified in Attachment 2.  ;

j 4 However the licensee's results for the chemical area  !

recirculationunitNo.549  !

outside of the comparison ra(CRCK) stackparticular tio limit. This sample was biased sample may- low and '

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' have had significant loading on the filter paper due to the  ;

chemical pro)erties of the air stream being sampled. Particulate  !

loading on t10 filter pa)er would result in absorption of alpha i p particle emission from tie sample surface. The inspector i discussed with the licensee the sample result agreements and

! disagreement and it was decided to collect and split two >

additional stack samples, including one from the CRCK stack. The  ;

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samples were to be analyzed for gross alpha and gross beta  !

radioactivity by both the licensee and its contractor laboratory '

and mailed to the NRC for comparative analysis.

3 l

For tracking purposes, this IFl will be closed. However, a new

If! pertaining to the comparison of the two additional stack '

i samples will be tracked (IFl: 70-1113/95-04-01). ,

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b. (Closed) Unresolved item (URI) 70-1113/94-06-02: 10 CFR 70.59 j semiannual report did not include total quantities of i

Technicium 99 (Tc-99) released in liquid effluents,  ;

! Table 5.1 of the License Application requires the licensee to

, i collect-composited samples from the discharge of the final process i lagoons semiannually and determine the concentration of Tc-99.

l The inspector reviewed Tc-99 analytical results from 1990 to 1994 of composited samples from the discharge of the final process 3 lagoons. The Tc-99 concentrations ranged from approximately

! 14 picocuries per liter I the effluent concentratio(pC1/1) to 200inpC1/1.

n limit (ECL) 10 CFR When comparedB, 20, Appendix to ,

[ Table 2, Column 1, the percent ECL ranged from approximately

. 0.02 percent to 0.33 percent. Guidance pertaining to the -

reporting significance-of a specific radionuclide was provided in-l Regulatory Guide 4.16.- Monitoring and Reporting Radioactivity in Releases of Radioactive-Materials in Liquid and Gaseous Effluents >

From Nuclear Fuel Processing and Fabrication Plants and Uranium

• i Hexafluoride Production-Plants, Revision (Rev.) 1, dated Dec. ember 1985. Regulatory Guide 4.16 states that a liquid effluent release is significant if the concentration of a radionuclide averaged- '

over a calendar quarter is equal to or greater than ten percent of ,

1, -the appropriate ECL specified in 10 CFR 20, Appendix B, Table 2 -

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3 Column 1. Although Regulatory Guide 4.16 suggests that liquid effluents should be sampled quarterly in order to establish that the radioactivity in the effluent is insignificant, the licensee collects composited samples from the discharge of the final process lagoons semiannually. This sampling method should be representative since the composite is of daily samples of effluent collected over a 26 week period. This item is considered closed.

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3. Environmental Protection (88045) l Safety Condition 5-1 of License No. $NM-1097 renewal requires that licensed material be used in accordance with the statements, i representations, and conditions contained in Part I of the Application  !

and supplements thereto. Chapter 5 of Part I specifies the requirements for the Environmental Protection Program.

Table 5.1 of Chapter 5 of the License Application outlines sam)1ing  !

frequency, parameters analyred, action levels, Minimum Detecta)1e  ;

Concentrations (MDCs), and sampling-types collected. Figures 5.3, 5.4, ,

5.5, 5.6, and 5.7 of the same chapter contained maps showing the  !

, locations of sampling sites throughout the area.

The inspector reviewed the licensee's environmental protection program with respect to management controls, quality control, and program  !

l implementation.- The )rogram provided representative measurements of radioactivity in the .11ghest potential exposure pathways and verification of the accuracy of the effluent monitoring program of ,

environmental exposure pathways. Accumulation of radioactivity in the environment can thereby be measured; trends assessed, to determine l

whether the radioactivity resulted from plant operations; projections  !

made of potential dose to off-site populations based on the cumulative j measurements of-any plant-originated radioactivity; and detection-of  !

unanticipated pathways for the transport of radionuclides through the  ;

environment. -The program was designed to detect the effects, if any, of  :

.- plant operation on environmental radiation levels by monitoring i radiation oathways in the area surrounding the plant site, it also verified t1at the measurable concentrations of radioactive materials and levels of radiation were not higher than expected on the basis of the '

effluent measurements. ,

l- a. Organization  ;

Chapter 2 of Part I of the License Application describes the licensee's general: organization and administrative requirements.

Specifically, Section 2.2.1.6 specifies the responsibilities of the environmental protection function.

The Environmental Protection and Industrial Safety Department was t

responsible for the environmental protection functions. The inspector reviewed the licensee's organization, staffing levels

and lines of authority as they related to the department to verify i

! that the licensee had the organizational structure-to control and

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4 monitor environmental radiation exposures and releases of radioactive material. The inspector noted that the Environmental Protection and Industrial Safety Department had not changed significantly since the last inspection and was composed of a manager, a senior environmental protection engineer, a nuclear safety engineer, an environmental protection specialist, and an industrial safety specialist. From interviews and discussions with staff personnel, the inspector determined that the staff was knowledgeable, experienced, and competent.

The inspector concluded that the licensee's Environmental Protection and Industrial _ Safety Department satisfied the stated requirements and noted that its staff was capable to implement effectively the program.

No violations or deviations were identified,

b. Augmented Groundwater Monitoring Program During March 1992, the licensee initiated an augmented groundwater monitoring program as part of a remedial action plan for contamiaation found beneath the fuels Manufacturing Operations (FH0/fM0X Building complex. Additional remedial actions included the remova)l of contaminated soils in source area, the sealing of the construction joint in the slab tank room, and the installation of a horizontal collection system beneath the slab tank room. The majority of the uranium contamination was removed by excavation.

The shallow, horizontal groundwater collection system was installed to mitigate migration of the residual contamination from beneath the building. Wells were also installed along the perimeter of the FM0/fM0X Building to monitor the upper and lower sand layers of the surficial aquifer and the principal aquifer.

The program included routine collection of water samples from the wells and the shallow groundwater collection system, and analysis

, of those samples for uranium, fluoride, and nitrate concentrations. These three parameters were the primary indicators of groundwater contamination associated with the material released from the FH0X slab tank room and were referred to as the critical parameters.

The inspector reviewed the 1994 Annual Assessment of the Groundwater Hydrology and inorganic Chemistry in the Vicinity of the' General Electric Company FH0/fM0X Building Area Report which was submitted to the State of North Carolina. The groundwater sampling, analysis, and assessments were-performed by the Research Triangle Institute (RTI). The inspector reviewed the data--in the report to assess the effectiveness of the groundwater remediation actions beneath the FM0X slab tank room, it was noted in the report that nitrate, fluoride, and uranium concentrations were decreasing as a function of time in the horizontal collection system. The greatest rate of decrease in concentration occurred when ground water-remediation system was initiated. As expected,

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the rate of decrease continued to diminish. This was an indication that the horizontal collection system was effective in removing contamination from its area of influence beneath the FM0X Building.

In addition, historical and recent grounJwater monitoring results l

for the three critical parameters indicated that the lower layer of the surficial aquifer and the principal aquifer had not been l contaminated by the inorganic constituents associated with the FMOX slab tank source. However, upper surficial groundwater contamination by these inorganic constituents was det.cted by the monitoring system surrounding the FH0/FM0X Buildirg. The base of  !

this shallow aquifer was approximately ten feet below the land  :

surface. The primary area of contaminatior, was localized on the west side of the FMOX Building, situated near the slab tank room.

l The two monitoring wells at this location were both characterized by concentrations of nitrate and fluoride that exceeded the State of North Carolina groundwater standards and consistently contained uranium just above the licensce's minimum detection limit of 0.02 eg/ liter. Isotopic results indicated that the uranium vas l most likely related to fuel production processes since it was not i

characterized as naturally occurring. The presence of these three critical inorganic parameters in the groundwater suggests a localized impact to the shallow groundwater quality on the exterior of the building. The steadily increasing concentration of nitrate in the downgradient well (from the source) and the subsequent periodic detections in uranium in this well in 1994, suggests that the shallow contaminant plume was following the groundwater flow and moving slowly toward the northwest in this localized area outside the FMOX Butiding. The licensee was continuing to monitor the upper surficial aquifer wells on a quarterly frequency.

The inspector concluded that the licensee's groundwater

, remediation program was effective in reducing the contaminant concentrations and that the monitoring program was adequate to track the migration of the contaminants and was being etfectively implemented.

No violktions or deviations were identified.

c. Program implementation and Results The inspector accompanied a licensee representative and observed the collection of the five ambient air particulate samples. The samples were collected in accordance with procedural guidance.

The ambient air sampling equipment was operational and the flow meters had been calibrated within the specified frequency.

The inspector re.iewed selected sampling results from vegetation, water, sediment, and soil samples collected quarterly in 1994 as part of the licensee's routine environmental monitoring prograa.

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6 Most of the sampling locations around the facility were below the appropriate action level specified in Table 5.1 of the -

Application. The inspector did note that, in 1994, a few soil sample locations had elevated gross alpha radioactivity concentrations that exceeded the licensee's action level specified in Table 5.1. In those cases, the licensee issued an Environmental Action level (EAL) investigation. The inspector reviewed selected EAL investigations that documented elevated soll sample results that occurred in 1994. The EAL investigations were adequately documented and the licensee had either taken appropriate corrective action or was still assessing the data to i determine the appropriate corrective action to take.

Based on the above reviews, it was concluded that the licensee had effectively implemented the required environmental monitoring >

program.

No violations or deviations were identified. I

d. Non-Routine Environmental Sampling Program The inspector reviewed the soil sample and gro ndwater monitoring well resulu collected in the vicinity of where the new Dry Conversion Process (DCP) facility was being constructed. The DCP facility was being constructed to replace the. current UF, conversion equipment. The area to be occupied by the DCP Building on the north side of the FMOX Building was evaluated to determine if there were any pre-existing conditions of environmental concern which should be addressed before construction. '

The licensee utilized both historical and current groundwater monitoring data from monitoring wells placed around the FMO-FHO-X complex. The FX series of wells was added several years ago after some leaks were identified in the chemical processing area of the i

, plant to monitor the movement of uranium and associated inorganics ,

that could migrate from under FM0-FM0X complex.- In summary, the '

licensee did not identify any contamination from uranium or other inorganic material that would require reporting or remediation in

, the vicinity of the new construction site.

1 In addition, the licensee evaluated the soil in the area to be occupied by the DCP facility by collecting soil samples and performing direct radiation soil surface scans. The area was

. divided into 30 feet x 30 feet grids. One soil sample was ,

collected and one direct radiation measurement was made in each grid. Direct radiation readings were essentially at normal -

background levels (0.05 mR/hr). Total uranium concentrations in the soil ranged from less than one pCi/gm to approximately

, eight pCi/gm. In summary, there was no indication of soil -

contamination in the construction area.

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The inspector noted that the licensee submitted a letter dated December 15, 1994, to the NRC Office of Nuclear Materials Safety and Safeguards (0NHSS) regarding the proposed construction of the DCP facility. That letter included a description of the i construction project, with details, including, purpose, location, building construction and use, radiological safety, nuclear l criticality safety, environmental safety, industrial safety.

l material control and cccountability, emergency preparedness, physical security, and decommissioning. The ONHSS, in a letter dated December 21, 1994, determined that construction of the DCP buildings would not significantly affect the quality of the environment and had no objection to the licensee constructing the butidings. In addition, the NRC indicated that the licensee ,

should maintain the results from the pre-operational soil sampling I in the decommissioning file required by 10 CFR "0.25(g).

No violations or deviations were identified. l

e. Calcium Fluoride Relocation Project The inspector discussed with the licensee the )roposed Calcium  ;

Fluoride (Caf,) Relocation Project that would unvolve the relocation of approximately 1.4 million cubic feet of uranium bearing Caf, to an above ground covered storage facility, in

' addition, the inspector toured the two in-ground storage locations where the Caf, had been buried. The following provides a brief

' history of the calcium fluoride generation, storage, and proposed removal of the material.

Calcium fluoride is a byproduct resulting from the treatment of ammonium fluoride liquid generated by the ammonium diuranate (ADU) conversion process. This waste stream contained relatively low concentrations of uranium. Caf, sludges were placed in two, in-ground storage locations from the time of conversien operation i

commencement in 1968 to the time of Waste Treatment facility (WTF) operation in 1972. These in-ground storage locations were referred to as the central storage area and the northwest storage area. Since 1972, Caf, sludge containing uranium had been accumulated in the Caf, lagoons at the WTF. When the fluoride waste treatment portion of the uranium recovery unit was started in 1985, the uranium concentration in the Caf sludge was reduced to the point that it was possible to dispose ,of most of the Caf, sludge as a non-uranium bearing material, as long as the conditions specified in Section 1.8.5 of the License Application were met. During the period since 1985, most of the Caf, sludge had been disposed of at the GSX-Laidlaw facility (a Resource Conservation Recovery Act (RCRA) hazardous waste burial facility) in Pinewood, South Carolina. For a brief period, the licensee was able .to ship some of the Caf, sludge to a facility in Pittsburgh, Pennsylvania where it was used as a fluxing agent in the steel industry.

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8 As noted above, the Caf, sludge was presently stored in the following three locations at the licensee's facility: 1 (1) northwest storage treat (2) central storage area; and (3) lagoon storage area. The northwest storage area consisted of seven hallow trenches located on the high ground in the northwest quadrant of the facility. The central storage area consisted of two shallow basins and an L-shaped trench located at the site lagnon area. The inspector verified that access to the northwest and central storage areas was controlled by a lor.ked fence and that they were properly posted as a radioactive materials storage area. The lagoon storage area consisted of two fluoride waste lagooas located at the WTF. The licensee indicated that the fluoride lagoon storage area was near the end of its useful life.

Although Caf, sludge had not been routinely placed in the lagoon system since 1985, uranium bearing Caf, sludge is sent to the lagoons during start-up periods from the Ammonia Recovery Plant and during periods when the fluoride waste stream exceeds allowed limits f or Caf, recovery.

The licensee had installed ground water monitoring wells around these facilities. Groundwater samples were collected quarterly and analyzed for fluoride and total uranium. These results were routinely submitted to the North Carolina Department of Environmental Management. To date, the data indicated that there had been no migration of either the uranium or fluoride.

The licensee had planned removal and relocation of the Caf: sludge in a three phase program. Basically, the three phase program would sequentially excavate and empty the northwest storage area, central rtorage area, and the lagoon storage area in order to consolidate the Caf, in an enclosed above ground storage facility.

The licensee planned on storing the material until economic recovery of the uranium in the Caf, could be achieved. Following removal of the Caf, from the storage areas, the licensee planned to characterize the in-ground storage areas and take the necessary actions to reduce the residual contaminants to a level that may permit their release for unrestricted use. The inspector indicated that the licensee's progress in this Calcium fluoride Relocation ProjNt would be monitored during future inspections.

No violations or deviations were identified.

4. Radioactive Effluents (88035)

a. Liquid Effluents Safety Condition S-1 of Materials License No. SNH-1097, authorized the use of licensed materials in accordance with statements, representations, and conditions contained in Part I of the License Application and approved supplements thereto.

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l 9 i Section 5.1.2 of the Application described the treatment process, sampling and analytical controls for treated process liquid waste, i Liquid waste streams containing uranium were segregated as nitrate I waste, fluoride waste, and radioactive waste. Each of these {

separated waste streams were treated and processed through a  ;

quarantine tank system before they were released from fuel manufacturing operations. All processed wastes were collected, ,

treated and discharged from process lagoons to the Northeast Cape l Fear River. Treated process wastes (with the exception of nitrate wastes) were sampled and discharged at the outfalls of the two final process lagoons. A composite sam)le proportional to the discharge flow of the liquid waste disciarged to the river was required to be collected daily and chemically analyzed for uranium concentration. Weekly composite samples of those daily samples were required to be analyzed for gross alpha and gross beta activity.

Section 5.1.4.2 of the Application described the sampling and analytical controls for treated nitrate waste transferred to the Federal Paper Board Company's biological WTF. Each tank truck shipment of nitrate waste was required to be sampled and analyzed

. for uranium. Daily composites of the grab samples were also required to be analyzed for concentrations of uranium and weekly composites were required to be analyzed for gross alpha activity.

Table 5.1 of the Application established action levels for the results from each of the above required analyses except for gross alpha activity in the nitrate waste weekly composite sample.

The inspector reviewed selected procedures and verified that the action levels and provisions for collecting and analyzing liquid effluert samples were in accordance with the Application, in addition, the inspector reviewed selected weekly composited analytical results from January 1995 to March 1995 for discharged

, liquid wastes from the final process lagoons and noted that uranium, gross alpha, and gross beta concentrations in the samples were well below the specified action levels.

Based on the above review, it was concluded that the licensee had effective liquid waste release monitoring system and that liquid releases were effectively controlled and minimized.

No violations or deviations were identified.

b. Gaseous Effluents l Chapter 5 of Part I of the License Application described the licensee's Environmental Protection Program requirements.

Spe:ifically, Section 5.1.1 outlined airborne effluents and exhaust systems from the uranium processing areas. Each exhaust stack from uranium processing areas was required to be continuously sampled from a point between the final HEPA filter l

l l

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b, e

10 and the discharge to the atmosphere. Table 5.1 of the Application ,

specified the collection frequency, parameters of interest, action  ;

i levels, and detection limits. Depending on the particular stack,  ;

the filters in the stack samplers were required to be milected at least daily or weekly and measured for gross alpha and beta =

l activity. '

l (1) Records Review The inspector reviewed selected procedures and verified that l i

" the action levels and provisions for collecting and l analyzing gaseous effluent samples were in accordance with l 3

the License Application. In addition, the procedures 4

included provisions for summarizing the analytical results in weekly reports and notifying management if the action levels were exceeded.

l The inspector selectively reviewed weekly stack program reports covering the period from January 1995 - April 1995.

Based upon the records reviewed, the inspector noted that

• the specified action levels had been exceeded on only on i occasion during the period. January 1995 through April 1995.

{ On January 22, 1995, CHMNO 542 Unit No. I daily sample

! result was 1.lE-Il pCi/cc which exceeded the instantaneous licensee action level of 1.0E-ll pCi/cc. The licensee's

stack monitoring program flagged the action level result and stack location, and generated a request to perform an investigation and complete an Unusual Incident Report (UIR).

The HVAC Team Leader was the principle investigator who

' received the notice and initiated VIR No. 95-02. The apparent cause of this elevated stack result was due to the deterioration of the HEPA filters in exhaust L%it No. 541 due to acid fume intrusion from the scrubber unit and onto the HEPA fibers. Six out of 32 HEPA filters had

i. i deteriorated and thus allowed particles to by-) ass the i filtration system. The demisters in the scrub 3er apparently

were plugged which caused a change in flow patterns through the scrubber and decreased its efficiency allowing acid fumes to intrude and communicate with the filters. The licensee shut down exhaust Unit No. 541 and started stand-by Unit No, 546. The HEPAs were changed in Unit No. 541. In

, addition, the demisters in the Chemical scrubber were cleaneo.

In addition, the inspector reviewed quarterly and annual i concentrations and quantities of gaseous waste effluents released in 1994-and the first quarter of 1995. The 1994 average concentration was 4.30E-14 utt/cc. The total volume

of exhaust gas released for 1994 was 2.87E+15 cubic centimeters (cc). The total quantity of uranium released in 1994 was 124.6C9 1, while the total quantity of uranium

) released during the first quarter of 1995 was 47.5 uC1. For i the first quarter 1995, the average concentration was i

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6.90E-14 pCi/cc. The total. volume of exhaust gas released i was 6.90E+14 cc. From the data above, it was apparent that the Itcensee had not exceeded the License Application's quarterly limit of 1250 pCi/ quarter.

(2) Offsite Dose Estimates The inspector also reviewed the licensee's calculations and i

input parameters-to calculate offsite doses to the maximally exposed individual due to gaseous effluent releases using the US Environmental Protection Agency's COMPLY computer code. The COMPLY computer code was used for calculating the offsite total effective dose equivalent (TEDE) to the individual likely to receive the highest dose from licenseJ ,

operation to demonstrate compliance with 10 CFR 20.1301 and 20.1302 and with the National Emission Standards for Hazardous Air Pollutants (NESHAPS) in 40 CFR 61, Subpart 1.

Effluent releases via the licuid pathway were not part of the COMPLi code's input parameters. The total effective dose equivalent for 1994 was 0.2 millirem per year -

(mrem /yr), which was well below the NESHAPS limit.

(3) Collection of Samples The inspector accompanied a licensee technician to observe the collection of the routine weekly stack samples, as i

s)ecified in Table 5.1 of the Application. The inspector o) served the changeout of particulate filters at 30 effluent stacks. The inspector noted that the technician was knowledgeable about the location of all of the stacks to be j changed out, that the appropriate procedures were used, and that the individual was adequately knowledgeable of his duties.

, The inspector observed that the physical condition of the vent stack sampling stations was good, indicating that the <

licensee's maintenance )rogram was effective. Also, all units were cbserved to 1 ave been calibrated within the required time interval.

No violations or deviations were identified.

c, Solid Waste Safety Condition S-2 of Materials License No. SNM-1097, granted exemptions and special authorizations in Sections 1.8.1 through

1.8.14, Part I of the Application.

1 Section 1.8.5, Disposal of Industrial Waste Treatment Products,

, specified that, notwithstanding any requirements for State or j local government agency disposal permits, the licensee was authorized to dispose of industrial waste treatment products i

l l

l 12 i

without continuing NRC controls provided that either of the conditions in Sections 1.8.5.1 or 1.8.5.2 were met. For solid (

waste shipped for disposal at the RCRA hazardous waste burial facility in Pinewood, South Carolina, the tot.1 uranium concentration shall not exceed 250 pCi/gm, of which no more than 100 pCi/gm shall be soluble.

The inspector reviewed and observed the licensee's process for disposal of calcium sulfate (CaSO ) presscake at the GSX Hazardous Waste Landfill in Pinewood, South Carolina. In 1985, the NRC approved a license amendment request from the licensee to dispose l of industrial waste containing small quantities of low enriched uranium (LEU) at the GSX Hazardous Waste Landfill. This amendment was added to the special authorization section of SNM-1097. The licensee was authorized under the license condition to dispose of waste containing small quantities of LEU in accordance with Option 2 of the 1981 Branch Technical Position (BTP), " Disposal or Onsite Storage of Thorium or Uranium Waste from Past Operations,"

which basically limited the uranium concentration in the waste material to 250 pC1/gm (insoluble) and 100 pC1/gm (soluble).

~

The CaSO 4 was produced as a byproduct from the recovery of uranium from the onsite lagoon sludge. The Uranium Recovery Lagoon Sludge (URLS) project had been an on-going project to remediate the East and West lagoons. The licensee had determined that the average concentration of uranium in the sludge was 50 pCi/gm with a maximum concentration of 250 pC1/gm. During this ins)ection, the inspector observed that the licensee was processing tie West lagoon. The CaSO 4 sludge was disposed of as a sligl,Lly moist presscake and placed in three separate synthetic bags before being tied to a wooden pallet for placement in the hazardous waste site cell.

e The inspector reviewed selected leach /raffinate presscake sample results from December 1994 to March 1995. Concentrations of insoluble uranium in the presscake samples ranged from 28 pCi/gm to 132 pCi/gm. Concentrations of soluble uranium in the presscake samples were well below the 100 pCi/gm limit. The inspector reviewed URLS Process Control Procedure, F-4060, Second teach and Raffinate Filter Press, Rev. 3, dated July 27, 1993, and noted that the licensee collected three core samples from each presscake block. Each core sample was analyzed for total uranium using either a Kinetic Phosphorescence Analyzer (KPA) or a Scintrex Lazer Analyzer (UA-3). Before either analyzer was used, the licensee performed a quality control check using an in-house prepared standard to verify that the analyzer was operating within the control-limits specified by the manufacturer.

As part of the NRC confirmatory measurements program, the inspector requested the licensee to provide three presscake samples from Bag No. 1192 so that a comparison could be made

.. .I

". , i 13 between the NRC and-licensee results. The purpose of these comparative measurements was to verify the licensee's capability to measure quantities of uranium accurately in the presscake sample. As noted above, the licensee performed the total uranium measurements onsite. The inspector requested that the presscake samples be_ delivered to the NRC Region 11 office. The licensee agreed with the inspector's request and promptly mailed the samples to the NRC Region !! office on May 4, 1995. The inspector informed licensee management that the comparison of the sample analyses would be tracked as the same followup item noted in Paragraph 2.a-above (IFI 70-1113/35-04-01).

No violations or deviations were identified,

d. Instrume.itation i

j The licensee designated one out of three Tennelec Model LB 5100 i 1 alpha / beta gas proportional counters (TN-3) to count the l particulate filters collected from the vent stacks. The inspector

] reviewed the licensee's calibration records, for the TN-3 i alpha / beta proportional counter, dated Septeraber 28 1993.and l September 26, 1994. The calibration curves had been generated by a computer which identified the ' plateau" and o>erating voltage.

The licensee had seven alpha sources and three jeta sources traceable to National Institute of Science and Technology (NIST) sources. These sources had been decay-corrected and were available to calibrate the proportional counter. For the last calibration performed, the licensee used Th-230 for the alpha source and Sr-90 for the beta source. Typical TN-3 detector efficiency for gross alpha measurements was 37 percent. In addition, the licensee used a uranium source to perform the daily source checks. The inspector reviewed the daily source check data from January 1995 to April 1995, and noted no apparent problems.

The inspector noted that the calibrations had been performed

, within the pre ibed time period and were adequate. Typical sample countint times were as follows: (1) daily samples- 15 minutes; (2) weekly samples- 10 minutes;-(3) daily _ backgrounds- 10 minutes; and (4) daily source checks- five minutes.

The. inspector conclu'ded that the counters had been properly calibrated wit,:in the required time limits using NIST-traceable sources and that they had been stable over the last several years.

No violations or deviations were identified.

5. Records and Reports of Radioactive Effluents (88035) 10 CFR 70.59 requires the licensee to submit i. report to the NRC Region !! office, within 60 days after January 1 and July 1 of etch year, specifying the quantity of each the of the principal radionuclides released to unrestricted areas in liquid and gaseous effluents during the previous six months of operation, if the quantities of radioactive

s 14 materials released during the reporting periods are significantly above the licensee's design objectives previously reviewed as part of the licensing process, the report shall cover this specifically.

The inspector reviewed the Semiannual Effluent Release Reports for the period January 1, 1994 through December 31, 1994, and verified that they were submitted within the required time frame. Although the minimum reporting requirements of 10 CFR 70.59 were met, the inspector observed that the report format guidance specified in Regulatory Guide 4.16, '

Monitoring and Reporting Radioactivity in Releases of Radioactive Materials in Liquid and Gaseous Effluents From Nuclear Fuel Processing and Fabrication Plants and Uranium Hexafluoride Production Plants, Rev.1, December 1985, was not used. The quantities of radionuclides ,

l released to unrestricted areas presented in Attachment 1, Table 2 were derived from previous licensee effluent release reports.

The total gaseous radioactivity discharged during 1994, was 124.6 gCl of uranium, compared with 87.6 901 during 1993. Although there was an increase in total gaseous radioactivity discharged in 1994, during the i last five years there were no apparent statistically significant trends '

identified. Similarly, 86.1 mci of uranium wrs discharged to the Cape Fear River in 1994, compared with approximately 98.1 mci during 1993.

Although there was a decrease in total liquid radioactivity discharged in 1994, during the last five years there were no apparent statistically significant trends identified.

The inspector concluded that the licensee had implemented an effective program to monitor and control liquid and gaseous radioactive effluents.

The reporting of those effluents met the minimum regulatory requirements, however as noted above the reporting guidance specified in Regulatory Guide 4.16 was not used. The average concentrations of radioactive material released during 1994 in the liquid and gaseous effluents were well within the limits spei.iiled in the License Application and 10 CFR 20, Appendix B, Table 2.

No violations or deviations were identified.

6. Decommissioning Records (88035, 88045) 10 CFR 70.25(g) specifically requires licensees to maintain certain records important to the safe and effective decommissioning of the facility in an identified location until the license is terminated by the Commission. These records shall include drawings of structures and equipment in restricted areas where radioactive materials were used or stored, documentation identifying the location of inaccessible residual contamination, and detailed descriptions af unusual occurrences or spills of radioactive materials that can affect decommissioning.

10 CFR 70.25(g)(3) requires, except for areas containing only sealed sources (provided the sources have not leaked or no contamination remains after cleanup of any leak), that the licensee maintains a list to be updated every two years contained in a single document of

I 15 i

information pertaining to the safe and effective decommissioning of the i facility. 10 CFR 70.25(g)(3)(1)-(iv) specifies information to be j included in the list.

The inspector verified that the licensee had maintained a "Decomissioning File" in accordance with the requirements s)ecified in 4 10 CFR 70.25(g). In addition, the inspector reviewed with tie licensee i the onsite burial or abandonment of radioa.tive materials in location 4

around the GE facility. From discussions with the licensee and from direct observation by the inspector, it was noted that the licensee had two in-ground " storage" locations, comonly referred to as Northwest

Storage Area and Central Storage Area, in which CaF sludges containing

low concentrations of uranium were placed. The Caf sludges were placed

in these storage locations (or pits) from the time of conversion

operation starting up in late 1968 until the advent of the WTF in 1972.

. The licensee was in the process of developing a plan to relocate the )

Caf sludge for eventual recovery of the uranium (see Paragraph 3.e).

4 i

The decommissioning file was maintai'ted by the Manager, Radiation Safety  :

! and the inspector noted that, in general, the content and maintenance of 1 the., records satisfied the requirements of 10 CFR 70.25(g). The inspector also observed that the files were readily accessible and well

organized. The inspector noted that during the next update of the l

. decommissioning file in October 1995, if environmental monitoring

, program sampling results inside and outside of the controlled access i area (CAA) indicate a trend of elevated uranium concentrations, then the  !

i licensee should consider including these locations in the

decommissioning file. For example, the inspector noted that selected

soil sampling in the vicinity of sample location IA (approximately

100 feet east of the CP&L substation and 10 feet north of the perimeter j fence, just outside the CAA) and along the wet weather stream in the

vicinity of the-outfall from the final process-lagoons indicated-i elevated soil concentrations of uranium. These locations should be i considered for inclusion in the decommissioning file if an upward trend i

is identified. The licensee acknowledged the inspector's comments.

No apparcnt violations or deviations were' identified.-

i 7. Information Notices and NRC Bulletins (92701) 4 The-inspector determined that the following Info:mation Notice (IN) and j NRC Bulletin (NRCB) have been received by the licensee, reviewed for applicability, distributed to appropriate personnel, and that action, as appropriate, was taken or scheduled.

• IN 94-81: Accuracy of Bioassay and Environmental Sampling Results (November 25,1994)

• NRCB 94-02: Corrosion Problems in Certain Stainless Steel Packagings used to Transport Uranium Hexafluoride f (November 14,1994) b l

t 16 The NRC issued HRCB 94-02 to (1) notify licensees that some uranium hexafluoride transportation packagings manufactured with a phenolic foam in chlorides have exhibited pitting and corrosion problems; licensees that uranium hexafluoride transportation packagings(2) advise with high-chloride foam do not conform to the applicable NRC Certificate of Compilance (CoC); (3) remind itcensees that packagings that are not in accordance with the NRC CoC, or that are in an impaired physical condition, are not authorized for transport under the general license provisions of 10 CFR 71.12; and 4) require licensees to reply in writing regarding whether or not(they intend to discontinue use of the packagings with the high-chloride foam that are not in accordance with the NRC CoC.

Witnin 30 days of the date of NRCB 94-02 (November 14,1994),each licensee was required to submit a response indicating whether or not the licensee intends to discontinue use of the following packagings:

(1) Model No. GE-21PF-1 packagings that were fabricated by Nuclear Containars, Inc., af ter June 1,1986;- and (2) Model No. NCI-21PF-1 packagings that were fabricated by Nuclear Containers, Inc., during the period June 1, 1986, to November 30, 1991, ttit have NCI serial Hos. I through 486, and 487A and 488A. .

The inspector reviewed the licensee's response to the NRC in a letter dated December 13, 1994.

In that letter the licensee indicated that GE-

- NEP did not own or intend to use any of the identified packagings containing high-chloride foam for the transportation of UF.,

s)ecifically, certain GE-21PF-1 and NCI-21PF-1 packagings, as noted a>ove.

Occasionally, a defective 21PF-type overpack may be shipped to the itcensee's facility. The inspector verified that applicable procedures UF , Rev.19,were implemented dated March (Prod 80.05, Receiving Cylinders Containing 14,1995), to require that incoming overpacks be identified and any overpacks of the group tiescribed in NRCB 94-02 be segregated and returned to the shipper empty. The inspector had no further questions and it was noted that the licensee had met the response requirements specified in NRCB 94-02.

Within the areas inspected, no violations were identified.

B. Exit Int.srview The inspection scope and results were summarized on May 5,1995, with those persons indicated in Paragraph 1. The inspector described the areas inspected and discussed the inspection results, including likely informational content of the inspection report with regard ,o documents and/or processes reviewed during the inspection. Although proprietary

-documents and processes were occasionally reviewed during this inspection, the pro)rietary nature of these documents or processes has been deleted from tais report. Dissenting comments were not received from the licensee.

• l 4

17 1121 Item Number Status Description and Reference IFl 70 1113/94 06 01 Closed Compare analytical results for gross alpha and gross beta analyses of a final process lagoon sample and two stack samples collected.

URI 70 1113/94 06-02 Closed 10 CFR 70.59 semiannual report did not include total quantities of Tc 99 released in liquid effluents.

IFI 70 1113 '95 04 01 Open Compare analytical results for gross alpha and gross beta analyses for two stack samples, including one from the CRCK stack, and total uranium analyses for three CaSO. presscake samples from bag No. 1192 (Paragraphs 2.a and 4.c).

4 f

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', l

. l 4

Tabla 1: Comparison of NRC and General Electric Gross Alpha and Gross Beta Analytical Results for Gaseous and liquid Effluent Samples Sample Mat r.ig Analysis Concentration (uci/ml) Agree-Licensee ILRC +/- la htig Resolution anni CRCK' Gross a 8.25E-14 2.50+/-0.82E-13 0.33 3 No Gross G n.a.: 5.69+/-1.45E-14 ---

3 --

3 CALC 2 ' Gross a 3.69E-13 1.77+/-0.59E-13 2.08 3 Yes Gross B n.a 8 5.94+/-2.12E-14 ---

3 --

FPLE' Gross a 6.70E-08 5.00+/-2,00E-08 1.34 3 Yes Gross 8 6.50E-08 1.10+/-0.50E-07 0.59 2 Yes

' Chemical Area Recirculation Unit 549

'not analyzed

'Calciner Discharge No. 2

' final Process Lagoon Effluent Table 2: Effluent Summary for General Electric Company Nuclear Energy Production (1991-1994)

Quantity Peleased (microcuries)

E f fluent Nuclide 1191 1111 1191 1911 Gaseous U-234 6.53E401- 9.74E+01 7.31 E+01 1.04E+02 U-235 3.04E+00 3.87E+00 2.85E+00 4.06E+00 U-236 2.03E-01 3.63E-02 2.54E-02 3.69E-02 U-238 1.53E401 1.70E+01 1.16E+01 1.65E401 Lfquid U-234 4.80E+04 4.45E+04 8.19E+04 7.19E+04 U-235 2.24E+03 1.76E403 3.19E+03 2.81E+03 U-236 1.50E+02' l.47E+01 2.89E+0! 2.54E+01 U-238 1.13E+04 7.68E+03 1.30E+04 1.14 Th-234 N.D.' l.08E-01 N.D.' N.D.[+04 "None Detected Attachment 1

'. o 1 L.. l l  !

l i

l CRITERIA FOR COMPARISONS OF ANALYTICAL MEASUREMENTS This attachment provides criteria for the comparing results of capability tests and verification measurements. 1hese criteria are t'ased on empirical l relationships which combine prior experience in comparing radioactivity emission, and the accuracy needs of this program.

In these criteria, the judgement limits or " Comparison Ratio limits"' denoting l agreement or disagreement between licensee and NRC results are variable. This variability is a function of the ratio of the NRC's analytical value relative i

to its associated statis l program as " Resolution".}ical As the and analytical ratio of the NRC Jncertainty, value to referred its associated to in this l

{

i -

uncertainty increases, the range of the acceptable differences between the NRC l and iteensee values becomes more restrictive. Conversely, poorer agreement between NRC and licensee values must be considered acceptable as the resolution decreases.

. For comparison purposes, a ratio between the licensee's analytical value and the NRC's analytical value is computed for each radionuclide present in a given sample. The computed ratios are then evaluated for agreement based on

, the calculited resolution. The corresponding values for " Resolution" and the

" Comparison Ratio Limits" are listed in the Table below. Ratio values which

, are either above or below the " Comparison Ratio Limits" are considered to be l

in disagreement, while ratio values within or encompassed by the " Comparison

! Ratio Limits" are considered to be in agreement.

l TABLE NRC Confirmatory Measurements Acceptance Criteria Resolution vs. Comparison Ratio Limits Comparison Ratio Limits Resolution for Aareement

<4 0.4 - 2 5 4-7 0.5 - 2.0 8 - 15 0.6 - 1,66 l 16 - 50 0.75 - 1.33 l 51 - 200 0.80 - 1.25

> 200 0.85 - 1.18 l ' Comparison Ratio - Licensee value NRC Reference Value

' Resolution NRC Reference Value Associated Uncertainty

Attachment 2 l

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