ML20134C198

Compliance Evaluation Rept for Certification of Us Enrichment Corp Paducah Gaseous Diffusion Plant Paducah,Ky

ML20134C198
Person / Time
Site:	07007001
Issue date:	08/31/1996
From:	NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To:
Shared Package
ML20134B936	List: ML20134B933 ML20134C198 ML20134C209
References
NUDOCS 9609260074
Download: ML20134C198 (159)

Text

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i COMPLIANCE EVALUATION REPORT FOR THE CERTIFICATION OF THE UNITED STATES ENRICHMENT CORPORATION  ;

1 PADUCAH GASEOUS DIFFUSION PLANT PADUCAH, KENTUCKY DOCKET 70-7001  !

AUGUST 1996 l

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1 U. S. Nuclear Regulatory Commission i Office of Nuclear Material Safety and Safeguards l Division of Fuel Cycle Safety and Safeguards Washington, DC 20555 l

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B ADOCK 07007001 PDR I

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" . A.E OF CONTENTS l

l Chapter 1 INTRODUCTION . . . . .. .... . . . . ... . . . ... 1 l l

1.1 Introduction . . . .... ..... ..... ....... ..... . 1 l 1.2 Background ....... . . .... ..... . .. ... .. . . ... 2 1.3 General Plant Description . . . .. ... . ... . . .... .... 3 1.4 History of GDP Operation . .. ... . ... ... .. ..... ..... 8 1.5 Site Operations Summary ........... .. ........ . . ... 8 1.6 Authorized Activities . . . . . . ..... ........ .. . . 10 1.7 Codes and Standards . . . . .. .... ... . .. .... .......... 11 l l

l Chapter 2 SITE CHARACTERISTICS . . . ... ... . . .. ... . . 12 j 2.1 Geography and Demography . . . . .. ... .. ... . . .. . 12 2.2 Nearby Industrial, Transportation, and Military Facilities .... . .. 12 2.3 Climatology and Meteorology . . . ....... ... . .. . 13 2.4 Hydrology .................. ... . ..... . . .... 13 2.5 Geology and Seismology ..... . ........ ... ...... 13 Chapter 3 ORGANIZATION AND ADMINISTRATION .. . .. . . . .. 15 3.1 Organization ...... . . . ... ... ... ......... 15 3.2 Safety Review Systems . ... .... . .. ..... ..... 22 3.3 Operations . . . . .. . ... .... . . . ... . . . 23 l 3.4 Training . ..... . . . . .. .. ... ... .. .. 25 ;

3.5 Procedures . . ....... . ............ . . 28 1 3.6 Human Factors .. . . . . . ...... . . .... 29 3.7 Audits and Assessments . . ... . ...... ..... . . . 30 3.8 Quality Assurance .... . ............ ...... ..... . . 30 3.9 Event Reporting and Investigations . ........ . . .......... 33 3.10 Record Management ... . ... ... .......... . . . 34 3.11 Maintenance . . ......... ..... ... . . . .. 35 3.12 Configuration Management . . .... ..... ....... .. .... 37 3.13 Management Controls . . . .......... ... .. .. ....... .. 38 Chapter 4 FACILITY AND PROCESS DESCRIPTION . ............... ... . 40 4.1 UF. Recr,ipt and Feed .. .... ................. ......... 40 4.2 UFe Enri::hment .. ..... .... . ........... ... .... 46 4.3 UF, Product Withdrawal ... .... . . ...... ......... 50 4.4 UFe Tailt Withdrawal . . . . . .. .. ...... .. . ... . . 54 4.5 UF. Cylinder Storage . ..... .... ...... ....... . ... . 55 4.6 Chemical Facilities . ... .... . .. ..... ... . .. 56 4.7 Laboratory . . . .. .. ...... 56 4.8 Utilities . ............ . .. ..... . ...... .. . ...... 57 Chapter 5 ACCIDENT ANALYSIS . . .. .... ... .. . .. ..... 60 5.1 Accident Evaluation Methodology . . .. .. . .. . 60 5.2 Potential Hazards of Credible Accident Scenarios . . ... . ... . . 62 i.

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t Chapter 6 TECHNICAL SAFETY REQUIREMENTS . . . . . . ................... 69 1

l Chapter 7 RADIATION SAFETY ...................................... 75 l 71 ALARA ............................................... 75 l

7.2 Responsibilities . . . . . . . . . . . . . . . . . . . . . . .................. 76 l 7.3 Occupational Radiation Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 76 l 7.4 Exposure Controls and Exposure Experience . . . . . . . . . . . . . . . . . . . . . . 77 7.5 Airborne Radioactivity in the Workplace and Ventilation Systems . . . . . . . 78 7.6 Control of Surface and Personnel Contamination . . . . . . . . . . . . . ..... 79 7.7 Respiratory Protection Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 80 7.8 Instrumentation, Calibration, and Maintenance Program . . . . . . . . . . . . . . 81 7.9 Radiation Work Permit System . . . . . . . . . . ................... 82 7.10 Exemptions from 10 CFR Part 20 . . . . . . . . ................... 82 Chapter 8 NUCLEAR CRITICALITY SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 84 8.1 NCS Administrative Requirements . . . . . . . ..................... 84 8.2 Criticality Accident Alarm System . . . . . . . ..................... 89 8.3 Technical Criteria . . . . . . .................... ........... 90 Chapter 9 ENVIRONMENTAL PROTECTION AND WASTE MANAGEMENT ........ 93 9.1 Effluents ................................ .. ......... 93 9.2 Environmental Monitoring . . . . . . . . . . . . . . . . . . . . . ......... ... 95 9.3 Waste Management ........................... .......... 96 Chapter 10 CHEMICAL SAFETY .................... ................ 99 Chapter 11 FIRE PROTECTIO N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 101 11.1 Fire Protection Equipment . ................... ........... 102 11.2 Building Construction . . . . . . . . . . . . . . . ....................102 11.3 Process Fire Safety . . . . .............. .................103 Chapter 12 EMERGENCY PREPAREDNESS ............ ................105 Chapter 13 SECURITY AND SAFEGUARDS . . . . . . . . . . . . . . . . . . . . . . . . . . . . 110 13.1 Material Control and Accounting . . . . . . . . . . .................110 13.2 Physical Security and Transportation Protection . . . . . . ..........110 13.3 Classified Information ................. .................113 Chapter 14 DECOMMISSIONING .... ........... ...................115 Chapter 15 COMPLIANCE PLAN . . . . . . . . . . ..........................117 Chapter 16 ENVIRONMENTAL REVIEW . . . . . . . . . . . . . . . . . . ......... ... 121 Chapter 17 AUTHORIZATIONS AND EXEMPTIONS . . . . . . . . . . . . . . . . . . . . . . . . 122 Chapter 18 TERM OF CERTIFICATE . .. .............................123 i

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Chapter 19 CO NCLUSIO NS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 124 Chapter 20 ' ACRONYMS AND ABBREVIATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . 125 Chapter 21 R E FE R E N C E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 128 Appendix A PUBLIC COMMENTS AND NRC STAFF RESPONSES . . . . . . .. . . . . . . . . A-1 i

Appendix B INTERAGENCY CONSULTATION RESULTS . . . . . . . . . . . . . . . . . . . . . . B-1 l

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l LIST OF FIGURES Figure 1 Regional Location of the Paducah Gaseous Diffusion Plant . . . ...... 5 Figure 2 Paducah Gaseous Diffusion Plant Layout . .. . . . ... ....... . 7 Figure 3 Organization Structure .......... ...... .. . ..... . . 16 Figure 4 Uranium Enrichment Process Overview . . ........ . .. . .... 41 l

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l Chapter 1 INTRODUCTION 1.1 Introduction This report documents the United States Nuclear Regulatory Commission (NRC) staff compliance evaluation of the U.S. Enrichment Corporation (USEC) certification application for the Paducah Gaseous Diffusion Plant (PGDP) located in Paducah, Kentucky. The Paducah facility enriches natural uranium to a maximum of 2.75 percent 23sU by the gaseous diffusion method. The application consists of the safety analvsis report (SAR), a quality assurance program, technical safety requirements (TSRs), emergency plan, environmental compliance status report, fundamental nuclear material control plan, l transportation protection plan, physical protection plan, security plan for protection of l classified matter, waste management program, decommissioning funding program, l environmentalinformation, and a Department of Energy (DOE) prepared and approved compliance Plan. USEC submitted its initial application for certification by letter dated April 18,1995. By letter dated May 5,1995, the NRC rejected USEC's application i because it did not meet the requirements of Title 10 of the Code of Federal Regulations, l Parts 76.31 and 76.35. The NRC concluded that the application did not contain adequate information for the NRC to independently determine that the public health and safety, worker safety, the environment, special nuclear material and classified matter would be l adequately protected. The NRC and USEC met numerous times to discuss the application

! contents. USEC submitted near-final sections for NRC advance review for acceptability.

By letter dated September 15,1995, USEC submitted the revised application, without the Compliance Plan; the Compliance Plan was submitted on November 6,1995.

The staff requested additional informahon on the application by letters dated October 4, October 6, October 13, October 16, October 18, October 25, November 16, December 1, December 11, December 18,1995, January 29, February 15, June 19, and July 24, 1996. USEC responded by letters dated October 25, November 8, November 9, November 15, November 16, November 17, November 22, December 1, December 8, December 12, December 13, December 14, December 15, and December 22,1995 and January 3, January 5, January 10, January 11, January 18, January 19, February 9, l

February 12, February 16, February 20, February 23, February 29, March 1, March 6, March 7, March 11, March 13, March 19, March 20, March 27, March 28, March 29, April 3, April 4, April 11, April 17, April 18, April 25, April 26, May 2, May 16, May 22, May 31, June 27, June 28, July 2, July 3, July 18, July 19, July 26, and August 1, 1996. USEC submitted Revision 2 to the application on January 19,1996, Revision 3 to l the application on May 31,1996, Revision 4 to the application on July 26,1996, l Revision 5 to the application on August 1,1996, and Revision 6 to the application on August 12,1996.

, The staff requested additional information on the Compliance Plan by letters dated l

December 15 and 18,1995 and February 20,1996. USEC responded by letters dated January 19, February 5, March 27, April 2, April 4, April 9, April 18, April 24, April 26, April 30, May 1, May 9, May 24, May 30, June 10, June 17, and July 19,1996. USEC submitted Revision 2 of the Compliance Plan on February 5,1996. Revision 3, without 3 issues, was submitted on July 12,1996;the remaining issues were submitted by letters 1

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l dated July 15 and July 18,1996. Revision 3A, consisting of four issues, was submitted i on August 1,1996. In addition to the correspondence, NRC, USEC, and DOE met many times to discuss application questions and had numerous conference calls.

The application and all nonproprietary, unclassified supporting information and l communications are available at the NRC Public Document Room (2120 L Street, N.W., I Washington DC 20555) and at the Local Public Document Room (Paducah Public Library, 555 Washington Street, Paducah, Kentucky 42003) under Docket 70-7001.

As part of the staff's consideration of the application, there was a public comment period on the application. Notice appeared in the Federal Reoister (60FR49026)on September 21,1995, allowing for a 45-day public comment period on the application. A second notice appeared in the Federal Reaister (60FR57253)on November 14,1995, l providing for a 45-day public comment period on the Compliance Plan. In addition a public meeting was held on December 5,1995, at the Paducah Information Age Park Resource Center in Paducah, Kentucky. The comments received at the meeting and during the i comment period are discussed in Appendix A. l

1.2 Background

The President signed H. R. 776, the " Energy Policy Act of 1992"(the Act), into law on October 24,1992. Among other things, the Act amended the Atomic Energy Act of 1954 to establish a new Government Corporation, the U.S. Enrichment Corporation, for the purpose of managing and operating the uranium enrichment enterprise owned and previously operated by the DOE. The Act provided that within 2 years after enactment of

• 1 the legislation, the NRC is to promulgate standards that will apply to the two operating j gaseous diffusion plants to protect the public health and safety from radiological hazards and to provide for the common defense and security.

The Act further directs the NRC to establish a certification process under which the two gaseous diffusion plants at Piketon, Ohio, and Paducah, Kentucky, to be operated by USEC, will be certified annually by the NRC for compliance with those standards. The Act also requires the NRC to report annually to Congress on the status of the gaseous diffusion plants.

On February 11,1994, the Commission published in the F_eJJ_ral Reaister for comment a proposed new Part 76 to Chapter I of Title 10 of the Code of Federal Regulations (CFR) which establishes requirements and procedures for the certification process. The final rule was published September 23,1994, and became effective on October 24,1994. Under a Memorandum of Understanding between NRC and the DOE, DOE agreed to continue i nuclear safety, safeguards, and security oversight of the gaseous diffusion plants until the NRC has certified that the facilities are in compliance with the standards (Part 76) or approved a plan for achieving compliance with the standards.

DOE remains responsible for decommissioning of the site and retains ownership of the facilities, USEC is leasing the facilities. The Environmental Protection Agency (EPA) jurisdiction cemains unchanged. The NRC consulted with EPA as part of the certification 2

process, as required by the Act. The NRC staff also met with the Occupational Safety and Health Administration (OSHA) which has regulatory responsibility for occupational safety at the facilities. Consultation with agencies is discussed in Appendix B.

The USEC leased the two gaseous diffusion plants from DOE beginning on July 1,1993.

The plants have operated continuously since the early 1950s.

The objective of this review is to determine whether the PGDP operations comply with the regulations contained in 10 CFR Part 76. The review considers the management organization and administrative programs provided to assure safe operation of the facility.

The review identified and evaluated those elements of plant operation, termed important to safety, which must function at the highest level of reliability.

1.3 General Plant Description The regulations in 10 CFR 76.35(a)(1) require USEC to include in the SAR the " activities and locations involving special nuclear material and the general plan for carrying out these activities." This information is provided in Chapters 1,2, and 3 of the SAR. The regulations in 10 CFR 76.35(a)(2) also require USEC to provide the "name, amount, and specifications (including the chemical and physical form and, where applicable, isotopic content) of the special nuclear material, source and byproduct material the Corporation l proposes to use, possess or produce, including any material held up in equipment from i previous operations." The possession limits for NRC regulated source material, byproduct material, and special nuclear material are listed in Table 1-3 of the SAR. The table specifies the maximum quantity of regulated material that may be possessed by PGDP at any given time.

PGDP can possess up to 655,000 metric tons of uranium as source material,10 curies I thorium as laboratory chemicals and calibration sources,10,000 metric tons of uranium enriched up to 2.75 weight per cent (w/o) as special nuclear material,500 kg uranium enriched up to 9.95 w/o in calibration sources, up to 10,000 grams uranium enriched greater than 10 w/o but less than 20 w/o that is held up in equipment or used as calibration sources, up to 1000 grams uranium enriched to greater than 20 w/o that is held up in equipment or used in calibration sources, 0.5 Ci plutonium as laboratory chemicals and calibration sources, up to 2 curies (Ci) of byproduct material,500 mci Am-Be neutron sources, and other byproduct material and plutonium that exist as contamination as a '

consequence of the historical feed of recycled uranium. Based on the possession limits, PGDP is considered to be a Category 111 facility for safeguards and security purposes.

1.3.1 Site Description The Paducah facility is located in the northwestern corner of Kentucky in western McCracken County. The plant is located within a reservation of approximately 3,423 acres, of which approximately 748 acres are within the controlled plant security fence. The area surrounding the f acility is predominantly rural, immediately adjacent to the site is the West Kentucky Wildlife Management Area (WKWMA). The WKWMA is leased to the Commonwealth of Kentucky and consists of about 2,080 acres of the reservation. The WKWMA draws thousands of visitors annually for recreational purposes 3

such as hunting and fishing, horseback riding, field trials, hiking, and bird watching.

Bordoring the reservation to the north and northeast is the TVA Shawnee Steam Plant on the hio River. The facility is located about 10 miles west of Paducah, Kentucky, and 3.6 miles south of the Ohio River. Figure 1 shows t regionallocation of the PGDP.

1.3.2 Plant Description The principal process and purpose of the PGDP is the production of enriched uranium for l reactors. The uranium fuel cycle starts with the mining and milling of uranium ores to produce yellow cake, and then the conversion of the yellow cake into uranium hexafluoride (UF ). The UF is then shipped to an enrichment facility where the concentration of fissionable 23sU is increased. This enriched UF is transported to other fuel cycle facilities I where it is processed and fabricated into fuel assemblies and then sent to the nuclear power reactor. The gaseous diffusion plant (GDP) is a type of enrichment facility.

The gaseous diffusion separation process depends on the separation effect arising from l molecular effusion (i.e., the flow of gas through small orifices). When a mixture of gas  ;

molecules is confined in a vessel, the average velocity of the lighter molecules is greater j than that of the heavier molecules. Therefore, the molecules of the lighter gas strike the vessel walls more frequently than the molecules of the heavier gas. If the walls of the container are porous with holes large enough to permit the escape of individual molecules, but sufficiently small so that bulk flow of the gas is prevented, then the lighter molecules escape more readily than the heavier ones. The gas consisting of the escaped molecules is then enriched with respect to the lighter component of the mixture. I 1

The primary purpose of the enrichment facilities at the PGDP is to produce uranium enriched in 23sU assay and to strip uranium partially depleted in 23sU content to an I economically feasible assay. The PGDP enrichment facility consists of about 1800 operating stages arranged in two parallel cascades. The cascade buildings are designated as C-331 (400 stages), C-333 (480 stages), C-335 (400 stages), C-337 (480 stages), and C-310 (60 stages). The surge and waste building, C-315, does not contain any operating stages. The degree of isotopic separation in an efficiently operating diffuser cascade is only about 0.2 percent per stage. Consequently, between 500 and 700 stages are required between the feed point and product withdrawal point to enrich uranium from normal feed at 0.71 w/o 23sU to product ranging from 0.95 to 2.0 w/o 23sU. These stages are called the enrichers. An additional 700 to 1100 stages are used to strip the 23sU isotope from normal feed to a tails withdrawal assay of 0.2 to 0.3 w/o 23sU. These stages are called the strippers. The plant can produce 11.3 million separative work units (SWUs) annually at a rated power consumption of 3040 megawatts.

The basic unit of the gaseous diffusion process is the gaseous diffusion diffuser (or converter). Compressed UFe feed gas is made to flow inside a porous membrane or barrier tube. Approximately one-half of the gas passes through the barrier into a region of lower pressure. This gas is enriched in the component of lower molecular weight (23sU) and is sent to the next diffuser. The gas that does not pass through the barrier is depleted with respect to 23sU and is sent to the previous diffuser. Upon leaving the diffusion chamber, the enriched and depleted streams have to be recompressed to the barrier high-side pressure to make up for frictionallosses. Because the degree of enrichment achieved in a 4

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. 1 single diffusion stage is very small, to achieve useful enrichment levels, the effect must be multiplied many times over by making use of a cascade of many stages in series. The exact number of stages required is determined by the enrichment needed.

The main components of a gaseous diffusion plant are: large cylindrical vessels called diffusers that contain the barrier; compressors used to compress the gas to the pressures needed for flow through the barrier tubes and from one stage to another; electric motors to drive the compressors; heat exchangers and cooling circuit for removing the heat of i compression; and piping for stage and interstage connections and control valves to adjust the gas flow. In addition to this process stage equipment, gaseous diffusion plants require  ;

auxiliary systems such as the UF, feed and withdrawal systems, an extensive electrical '

power distribution system, and cooling towers to dissipate the waste process heat. The following contains a brief description of the operations that occur in each of the main buildings. Figure 2 shows the plant layout.

C-360 Toll Transfer and Samolino Buildina This building has systems in place for cylinder receipt, unloading, inspection, weighing,  !

cold pressure checking, sampling, and shipping. There are 4 autoclaves for sampling and transfer of UFe to customer owned cylinders.

C-333-A C-337-A Feed Vaoorization Facilities These buildings contain the cascade feed facilities. The feed cylinders are placed in -

autoclaves and heated to convert solidified UFe to a pressurized vapor which can be controlled by valves and flow measuring devices to maintain distribution of UFe gas through heated piping to appropriate points in the enrichment cascade. C-333-A contains 4 feed stations and C-337-A contains 5 feed stations.

C-331. C-333. C-335 C-337 Process Buildinas These buildings contain the heart of the enrichment process. C-331 and C-335 each t contain 4 units and are known as the "00" buildings. There are 10 cells per unit and 10 stages in a cell. C-333 and C-337 each contain 6 units and are known as the "000" buildings. There are 10 cells per unit and 8 stages in a cell. Each stage contains a mntor, compressor, converter, control valve, coolant system, and associated instrumentatioi .

C-333 and C-337 also contain freezer / sublimer units that are used for inventory control.

These units allow the excess UF, inventory to be rapidly removed from the cascade by freezing it in storage vessels and then returning it to the cascade by sublimination when required.

C310. C-310-A Purae and Product Buildina This building contains the equipment for product withdrawal. C-310-A contains the liquefication process. C-310 contains the cylinder filling operation. The product UF,is transferred from the gaseous to the liquid state, loaded into cylinders, and solidified in preparation for shipment.

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C-315 Surae and Waste Building This building serves 3 functions: surge systems, process tails withdrawal system, and a dry air plant (which is numbered C-620). UF, gas is compressed and condensed as liquid UF, and drained as hot liquid at above atmospheric pressure. The cylinders are weighed and moved outside to cool; after cooling the cylinders are moved to a storage yard.

C-400 Chemical Ooeration Buildino (Cleanino Buildino)

This building houses the decontamination and uranium recovery equipment.

C-300 Central Control This building houses the central control room. All PGDP communications, enrichment operations, and electrical distribution system operations are monitored and/or controlled, either directly or indirectly, from this room.

C-745-A throuch C-745-U Cvlinder Storace Yards These areas are used for interim and long-term storage of UF, cylinders.

1.4 History of GDP Operation Construction began at the PGDP in 1951. The plant began operating in 1952 and was fully operational by 1955, supplying enriched uranium for commercial reactors and military defense reactors. Construction at the plant included more than 100 permanent buildings with over 7 million ft 2of gross floor area. Additional construction included an extensive road and railroad network; expansive utility systems for electrical power distribution; process and sanitary water supply, treatment, and distribution systems; storm drainage system; sewage treatment facilities; and a dry air supply system.

1.5 Site Operations Summary The GDPs have over 100 years of collective operating experience (including Oak Ridge and Portsmouth). During that time, no incidents at any of the GDPs have caused death or serious injuries to plant personnel from exposure to radioactive materials or radiation nor have there been any incidents that have resulted in off site releases of radiation or radioactive materials that could cause committed doses in excess of established limits.

(USEC,1996) in the early 1970s, the GDPs initiated the Cascade improvement Plan and the Cascade Uprating Program to increase efficiency and capacity; the programs were completed in the early 1980s. The design of these upgrades incorporated appropriate safety improvements based, in part, upon the GDP operating experience gained since the 1950s. The GDPs produced a Final Safety Analysis Report (FSAR) for each site between 1980 and 1985. These documents identified the major safety events (major hazards, initiators, and sequences) and established an envelope for safe operation, as defined in DOE orders. The FSARs analyzed the unique risks associated with the operation of the GDPs, examined the impact of these hazards during accident and process upset conditions, and evaluated the risk to both on- and off-site personnel. According to DOE, the FSAR 8

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l understanding of the overall operation and associated hazards, and senior staff with first- i hand experience and understanding of the theory and operations. In 1989, DOE initiated an upgrade program for the FSARs to correct weaknesses and errors and to formalize some l of the analysis. This effort continues and is expected to be complete in February 1997.  ;

The upgrade program is discussed in greater detail in Chapter 15 of this CER.

l The PGDP has experienced several incidents during its operating history, some of which i have resulted in release of radioactive material. Some of the more significant events are l briefly discussed below. Additional information concerning these events is available in Chapter 4 of the SAR.

In November 1956, there was a major fire in the C-310 Purge and Product Building. The fire originated at a withdrawal compressor position where a seal failure resulted in the escape of process gas and a small quantity of fluorine. The fluorine reacted vigorously with the lube oil on the compressor surfaces. The heat from the fire caused the lube oil i

supply line to rupture, causing the fire to grow sufficiently to ignite the roof. The roof and I side walls above the cell floor collapsed causing extensive damage to the structure and equipment. PGDP has now installed a sprinkler system to minimize the possibility of reoccurrence.

In November 1960, PGDP was using a temporary vaporizer to heat the UFe. An overfilled cylinder was placed in the vaporizer and heated. The cylinder ruptured releasing 6,800lb of UFe . Steam chests for side feeding was discontinued; autoclaves are now required for heating cylinders.

In December 1962, an exothermic reaction in C-337 resulted in the release of about 3,400lb of UFe. The incident occurred after Cell 3 had been taken off-stream for treatment. An explosion in Stage 3 separated the converter head from the shell, moving the Stage 3 compressor across the aisle and against Cell 5. The Stage 4 compressor was moved several feet off its base. Extensive damage was done to the building. The incident was caused by a spontaneous reaction involving the treatment gases, residual UFe, materials of stage construction, and R-114. Cell treatment procedure changes were implemented to prevent reoccurrence.

In January 1978, equipment damage occurred in C-315 as the result of a UF,/ hot metal reaction. The direct cause of the incident was the failure of an automatic recycle control valve in the UF, gas compression loop to operate properly during a large inventory shift in the cascade. The gas compressor loop consisting of four centrifugal pumps, two gas j coolers, and associated piping and valves had to be replaced, along with building siding. 1 About 30lb of UFe was released. The PGDP made some design changes intended to prevent a recurrence of the event, j in December 1993,'PGDP experienced a coupling failure in C-337. Equipment damage was caused by excessive vibration of a compressor drive train. Although the estimated release was less than 2 lb of UFe, the incident did result in damage to the compressor and converter and associated piping, three adjacent converters; the surging from the air inleakage led to the deblading of four other compressors.

i 9

. _ - - _ ~ . - . . - . - - _ . . - ._ _.

In December 1994, PGDP experienced a release due to a pipe rupture. During cold-feeding of a damaged tails cylinder, the plant experienced a freeze-out of UF,in a feed line. During subsequent reheating of the UF, freeze-out, the feed line ruptured releasing approximately 10 pounds of UFe. Cold-feeding will not be an authorized activity under the NRC for the initial certification. PGDP can request authorization, with supporting analysis, in the future.

1.6 Authorized Activities i

The PGDP authorized activities for each regulated material are listed in Table 1-4 of the SAR. The authorized operations are discussed in more detailin Chapter 3 of the SAR. The activities to be conducted at the PGDP are:

1. Heating UF, cylinders and feeding contents into the diffusion process.  ;

2. Enrichment of natural uranium up to 2.75 percent enrichment by weight 23sU.

3. Receipt, storage, inspection, and acceptance sampling of cylinders containing uranium enriched up to 2.75 percent by weight 23sU.

4. Filling, assay, storage, and shipment of cylinders containing uranium enriched up to 2.75 percent by weight asU. 2

5. Cleaning and inspection of cylinders used for the storage and transport of source or special nuclear material.

6. Storage of process wastes containing uranium, transuranic elements, and other contaminants and decay products.

7. Process, characterize, package, ship, or store low-level radioactive and mixed wastes (storage of mixed wastes is limited to 90 days).

8. Radiation protection, process control and environmental sample collection, analysis, instrument calibration and operation checks.

9. Maintenance, repair, and replacement of process equipment. l I

10. Process Control Laboratory analysis and testing. I

11. Transfer between cylinders.

12. Calibration and use of portable health physics and fixed laboratory equipment.

13. Nondestructive testing and analyses of product and process streams.

14. Storage of special nuclear material and byproduct material remaining in equipment and f acilities from previous operations.

10

i .

i l

15. Calibration of neutron measuring instrumentation.

16. Use of internal sources in density meters.

17. Use of americium in smoke detectors.

l The activities listed above are those which the NRC has reviewed and will certify that they meet regulatory requirements. If additional activities are planned, USEC will need to perform a safety analysis and propose TSRs as necessary prior to conducting the activity.

I 1.7 Codes and Standards Appendix A to Chapter 1 of the SAR contains a list of the various industry codes and standards and NRC regulatory guidance documents that have been referenced in the PGDP application and in the responses to questions on the application. In many cases, the extent of the commitment needs to be clarified. Compliance Plan issue 45 commits USEC to review the commitments and to compile a listing of the specific sections of codes, standards, and regulatory guides to which USEC is committed. The results of this review will be transmitted to the NRC within 90 days after certification. This will be complete prior to the NRC assuming jurisdiction. The staff concludes that the plan and schedule are acceptable.

l 11 1

Chapter 2 SITE CHARACTERISTICS 2.1 Geography and Demography The PGDP is located at 37 06' 49" north latitude and 88 47' 39" west longitude in McCracken County, Kentucky. The site is in a generally rural area and was previously part of the Kentucky Ordinance Works. The largest cities within a 50-mile radius are Paducah, Kentucky, located approximately 16 air km (10 miles) to the east, and Cape Girardeau, Missouri, located approximately 64 air km (40 miles) to the west. The city of Metropolis, Illinois is situated approximately 5 miles north and east of the plant. There are several small communities located within a 5-mile radius of the plant boundary. Two unincorporated communities, Grahamville and Heath are located approximately 2 miles east of the plant. Portions of 28 counties are included within a 50-mile radius: 11 of which are in Kentucky,4 in Missouri,10 in Illinois, and 3 in Tennessee. The population within the 80-km (50-mile) radius is about 300,500 persons (MMES,1993). Paducah has a population (1990) of 27,256.

Highway US 60 is the closest major highway, and intersects the plant access road 3 miles south of the plant. Interstate Highway 1-24 is east of the plant and intersects US 60 near the city limits of Paducah.

2.2 Nearby Industrial, Transportation, and Military Facilities There are a number of industries located within 7 miles of the plant; many are located across the Ohio River in Illinois. The largest of these industries is Allied Signal, incorporated; an NRC licensee. Allied is licensed to produce uranium hexafluoride, in fact much of the feed material for the Paducah plant comes from Allied. Because of the distance and prevailing wind direction, a release of uranium hexafluoride at Allied would be unlikely to affect PGDP operations. None of the other industries would impact PGDP operations. DOE also conducts activities on-site and adjacent to the site, including activities in leased or USEC controlled areas. These activities are self-regulated by DOE.

These activities include, but are not limited to, groundwater contamination abatement and control, viaste storage activities associated with environmental contamination remediation, waste disposal pilot projects, and limited decontamination and decommissioning activities to comply with the Resource Conservation and Recovery Act (RCRA) requirements. These activities should not adversely impact PGDP operations. However, this situation will require special attention and coordination, after certification, to ensure that DOE activities do no negatively impact the safety of USEC operations regulated by NRC. DOE and USEC signed a resolution to address how shared site issues are to be resolved. There are no military facilities within 5 miles of PGDP.

Roadways within the fenced area consist of approximately 23 miles of paved surface.

Several paved roads branch to the periphery of the plant. The plant access road extends from the main entrance to Kentucky Highway 1154 which in turn connects with US 60.

The rail system at the site consists of approximately 17 miles . The rail spur enters the site west of building C-720 and branches to several areas inside the fence. Many of the buildings have direct rail service.

12

Barkley Regional Airport is located approximately 4 miles from the Paducah facility. Water transportation is available on the Ohio, Tennessee, and Cumberland rivers located within 15 miles of the site. The Mississippi River flows approximately 30 miles to the west.

2.3 Climatology and Meteorology The region in which the PGDP is located has a humid-continental climate. Summers are generally dry; precipitation occurs primarily in the spring and fall. Wintecs are characterized by moderately cold days; the average temperature in January, the coldest month, is about 1.7 C (35 F). Summers are warm and humid with average temperatures in July of 26 C (79 F). Annual precipitation is about 120 cm (47 in). The prevailing wind direction is from the south-southwest.

, Prevailing winds are from the southern quadrant at a 95th percentile velocity o' about l 1 m/s. Historically, the maximum windspeeds for the area range from 80 to 90 miles per hour. Northerly winds occur in the winter and southerly winds occur during the semmer.

An average of six tornadoes are sighted in Kentucky on an annual basis. Historicahy. there l

have been no reported tornadoes at the plant site. Additional information is available in ,

l Chapter 2 of the SAR.

l 2.4 Hydrology I

l The PGDP site is located in the western part of the Ohio River Basin. Two small Ohio River

! tributaries, the Big Bayou Cree.k on the west and Little Bayou Creek on the east, carry l surface drainage from the plant site. These two streams join north of the site and -

discharge into the Ohio River. Big Bayou Creek and its tributaries drain an area of 18.6 square miles. The area of the Little Bayou Creek basin is 8.5 square miles.

The historical flood of record for this area occurred in January 1937 with a high water mark of 347.0 feet above mean sea level which is 28 feet below the lowest elevation of the PGDP.

2.5 Geology and Seismology The facility is located at the northern end of the Mississippi Embayment. The Mississippi Embayment consists of unconsolidated sediments of Cretaceous to Quaternary age that dip gently downward to the south.

The facility rests on essentially flat-lying unconsolidated sediments more than 300 ft thick, I

which line unconformable on Paleozoic limestone that dips gently to the north. Although there is extensive faulting of the paleozoic formations north of the site, there is only one hypothetical fault mapped within 5 miles. The Paducah West Geologic Quadrangle Map shows a hypothetical fault starting about 3 miles east of the site near Chiles and trending northeast under the Ohio River. Eight other faults are shown on the Paducah West, the  !

Lovelaceville, the La Center, and the Bandana Geologic Quadrangle Maps. Six of these eight faults are based on boring data and are hypothetical. The two remaining faults are i

shown as small surface faults on the Paducah ht Geologic Quadrangle are apparently l pre-Pleistocene in age.

! 13 i

l l.

1 The thick unconsolidated cover of sediments that underlie the facility are essentially undeformed; there is no evidence of zones of alteration, irregular weathering, or structural weaknesses aside frorn the normal weathering of sediments at the ground surface and presumable at uncomformable stratigraphic contacts.

The potential for seismically induced liquefaction of unconsolidated sediments at the site represents the most significant engineering geology consideration. Based on recent study, a zone of potentially liquefiable sediments has been identified.

Several earthquakes are known to have occurred in the New Madrid seismic zone and would have caused impacts in the Paducah area. The most significant occurred in 1811 and 1812 with an estimated Richter magnitude of as high as 8.7. According to SAR 92.6, the evaluation basis earthquake for the PGDP was determined to have an estimated peak ground acceleration of 0.18g.

The ability of the facility to withstand a seismic event is one of the topics of the SAR upgrade project. Further discussion of the seismic aspects is contained in Chapter 15 of this CER.

14

t Chapter 3 ;NIZATION AND ADMINISTRATION The regu!z O CFR 76.35(a)(7) require that the SAR contain "A description of the  ;

managemer, wls and oversight program to ensure that activities directly relevant to '

nuclear safety and safeguards and security are conducted in an appropriately controlled i manner that ensures protection of employee and public health and safety and protection of ,

the national security interests." Chapter 6 of the SAR describes the organization and  ;

management controls utilized by USEC to meet this requirement.

The characteristics of an organization which can safely handle special nuclear and source materialinclude a clear assignment of responsibilities, including: (1) responsibility for the various components of safety; (2) an effective safety review system; (3) a training program for proper operation and proper conduct of the safety components;(4) a clear and accurate procedures; (5) a means of knowing compliance with the rules and procedures with a  !

means of assuring necessary compliance; and (6) an investigation process to enable j understanding and promote fixing of significant institutional problems. The following sections briefly describe the organization and management controls in place at USEC.

3.1 Organization USEC is a wholly-owned Government corporation which was established by the Atomic Energy Act, as amended. Members of the USEC Board of Directors were appointed by the President and confirmed by the U.S. Senate. Corporate offices are currently located in Bethesda, Maryland. USEC has hired a contractor, Lockheed Martin Utility Services, Inc.,

(LMUS) to operate the plant. USEC retains responsibility for the safe operation of the facility. USEC approves the management structure and key positions, assignment of individuals to key positions, and qualifications, responsibilities and authorities for key positions.

3.1.1 Safety Responsibilities [

USEC and LMUS have established an organization that has independent chains for the safety functions. The organization is discussed in r;6.1 of the SAR. By TSR 3.3, USEC is j required to use qualified individuals in facility positions, and to meet the responsibilities and I qualification requirements described in the SAR for the key staff positions. TSR 3.2.1 requires USEC to establish and define the lines of authority, responsibility, and ,

communication. The TSR also provides for the safety functions having organizational j freedom to ensure independence from operations. Figure 3 shows the organization i structure.

The Executive Vice President, Operations has overall responsibility for safe operations of the facility. This position has the authority to direct the General Manager to place the facility in a safe condition. Any operation directed to be shut down by the Executive Vice ,

President cannot be restarted without his/her concurrence. This position is appointed by I the USEC Board of Directors.

1 The Vice President, Production has overall responsibility for all activities within the USEC '

production organization, including the functions of operations, maintenance, plant support, 15

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engineering, transportation, materials handling and storage, and industrial, radiological, and l nuclear safety. The individual in this position has shutdown authority and must concur on l

the restart decision. The Vice President, Production is appointed by the USEC Board of l Directors.

t

! The USEC Nuclear Regulatory Assurance and Policy Manager is responsible for the

! management of USEC nuclear regulatory assurance functions, which includes the quality l assurance policy. This position is responsible for the USEC Ouality Assurance Plan and for j determining the status, adequacy, and effectiveness of the Quality Assurance Plan. This position is independent from production and reports to the Executive Vice President, Operations.

The USEC Safety and Health Assurance and Policy Manager is responsible for plant fire l and police services, nuclear material control and accountability, security, emergency

! management activities, and for the development and management of the related corporate

! policies. This position is independent from production and reports to the Executive Vice President, Operations.

The USEC Environmental Assurance and Policy Manager is responsible for the development and management of corporate environmental and waste management policies and for these

! activities at the plant. This position is independent from production and reports to the Executive Vice President, Operations.

The positions discussed above are all located at the headquarters office; the positions i discussed below are located at the PGDP site. i The Safety, Safeguards and Quality Manager is a USEC employee but is located at the site.

He/she reports to the Executive Vice President, Operations. This position has the responsibility for the oversight of plant operations to ensure that the health and safety of the public and workers are adequately protected, to ensure compliance with safety, safeguards, and quality requirements, and to ensure implementation of USEC policies, l procedures, and management expectations. This position manages the Safety, Safeguards l and Quality Office and directs plant quality assurance functions involving audits and oversight of plant operations as well as a nuclear safety assurance function. He/she is also responsible for nuclear material control and accountability. This position has shutdown authority.  ;

I The Safety, Safeguards and Quality Manager has an extensive role in the organization.

This position serves as the onsite presence for USEC and will be responsible for monitoring activities for USEC. This position is also responsible for the quality assurance (OA) program and the nuclear material control and accountability program. The QA program is in the early stages of development and will require a great deal of involvement. USEC needs to ensure that the support staffing is adequate to implement the necessary programs. This is an area in which the NRC staff has interest and will be carefully monitored through the observation and inspection program.

J 17

l l

The General Manager is responsible for the safe operation of the plant, for compliance with I all applicable NRC regulatory requirements, and for adherence to applicable USEC policies.  ;

i He/she is responsible for production, training and procedures, site and facilities support, l engineering, transportation, materials handling and storago, occupational, environmental, and nuclear safety. The General Manager also has responsibility for the primary day-to-day i l interface with the NRC for issues of adequate safety / safeguards and regulatory l compliance. The General Manager has shut down and stop work authority for all portions j of the plant. This position also has startup authority, concurrence must be obtained for l those operations shut down by the Executive Vice President, Operations; Vice President, l Production; or the Safety, Safeguards and Quality Manager. The General Manager is a l LMUS employee and reports to the Vice President, Production. This position is appointed by the President of LMUS with concurrence by the Vice President, Production and the Executive Vice President, Operations.

The Enrichment Plant Manager is responsible for the day-to-day production activities at the site including operations, maintenance, work control, and production support. This position has the same shutdown / restart authority as the General Manager, except that the concurrence of the General Manager must also be obtained. This position is appointed by the General Manager with concurrence by the President, LMUS and the Vice President, ,

Production.

l The Operations Manager is responsible for the safe operation of the enrichment cascade, plant utilities, chemical services, and feed and withdrawal facilities. This position may act for the General Plant Manager and the Enrichment Plant Manager. The Operations Manager has shutdown authority for that part of the operations for which he/she has responsibility.

l l The Maintenance Manager is responsible for providing safe and reliable performance of preventive, predictive, and corrective maintenance on production facilities and equipment.

l This position reports to the Enrichment Plant Manager.

I The Production Support Manager is responsible for the technical functions in direct support l

of production. This includes health physics program, laboratory operations, quality control, i and waste management program. He/she has the authority to stop work or shutdown l operations in areas for which he/she has responsibility. This position reports to the Enrichment Plant Manager.

l l

The Environmental Safety and Health Manager is responsible for establishing and ,

implementing the environmental monitoring program, environmental protection programs, and the industrial and chemical safety programs. This includes activities associated with environmental compliance, occupational safety and health, industrial safety, chemical safety, and industrial hygiene. This position reports to the General Manager and has stop work and shut down authority for activities that could cause environmental, safety or health concerns.

18

I l

! The Engineering Manager is responsible for engineering activities in support of operations, l including design, fabrication, and construction of plant modifications or additions; the configuration management program; systems and reliability engineering; and nuclear l safety. This individual reports to the General Manager.

l The Site and Facilities Support Manager reports to the General Manager and is responsible l for plant fire and police services, security, non-production related facility maintenance, and

( the shared site program.

The Training and Procedures Manager is responsible for preparation, presentation, and recording of employee orientations and for technical and qualification training programs' development and implementation. He/she is also responsible for the development and i implementation of the procedures management program.

The Plant Shift Superintendent Manager coordinates the activities of the plant shift

superintendents. He/she provides technical and administrative support for the plant shift superintendents. He/she also has responsibility for emergency management.

The Plant Shift Superintendent represents the General Manager, and has the authority and responsibility to make decisions as necessary to ensure safe operations, including stopping work and placing the plant in a safe condition. He/she can also authorize restart after l shutdown for non-routine reasons, however, approval of the Enrichment Plant Manager is necessary for operations shutdown by upper management. This position is responsible for j accumulation and dissemination of information regarding plant activities. The Plant Shift l Superintendent serves as incident commander during plant emergencies and is responsible j for making event notifications. He/she has the authority to act for the General Manager '

and Enrichment Plant Manager, in their absence, regarding operational matters. l l

TSR 3.1.1 assigns corporate responsibility for overall GDP safety to the Executive Vice l President, Operations. TSR 3.1.2 assigns responsibility for the overall safe plant i operations to the General Manager. The Plant Shift Superintendent responsibilities are assigned by TSR 3.1.3. The Plant Shift Superintendent is responsible for the operational aspects of the plant and for the central control room command function. TSR 3.1.4 assigns the Division Managers responsibility for operations conducted within their facilities.

These TSRs assigning responsibility are similar to the Westinghouse Standard Technical Specifications (WSTS). The staff concludes that the organization structure and assignment of responsibilities are consistent with good industry practice, meet the requirements of 10 CFR Part 76 and are, therefore, acceptable. l l

3.1.2 Technical Qualifications l l

The regulations in 10 CFR 76.35(a)(3) states that the SAR must include: "The qualifications requirements, including training and experience, of the Corporation's l management organization and key individuals responsible for safety in accordance with the l regulations in this chapter." Section 6.1.1 of the SAR describes the minimum l

qualifications needed for the key positions. It is the responsibility of USEC to ensure that i

19.

individuals in these positions meet the qualification requirements. TSR 3.3 requires facility positions to be filled by individuals whose experience / training qualify them for the position.

The minimum qualifications for key positions are described in the following paragraphs.

Executive, Vice President Ooerations The Executive Vice President, Operations shall have a bachelors degree or equivalent technical experience,10 years of management experience, and 6 years of nuclear experience.

Vice President Production The Vice President, Production shall have a bachelors degree in engineering or the physical sciences or have equivalent technical experience. He/she shall also have 6 years of technical nuclear experience and 6 years management experience (may be concurrent).

Nuclear Reaulatory Assurance and Policv Manaaer The Nuclear Regulatory Assurance and Policy Manager shall have a bachelors degree or equivalent technical experience and 4 years of nuclear experience.

Safety and Health Assurance and Policy Manaaer The Safety and Health Assurance and Policy Manager shall have a bachelors degree or equivalent technical experience and 4 years of nuclear experience.

Environmental Assurance and Policv Manaaer The Environmental Assurance and Policy Manager shall have a bachelors degree or equivalent experience and 4 years of environmental management experience.

Safetv, Safeauards and Quality Manaaer The Safety, Safeguards and Quality Manager shall have a technical degree. He/she shall l also have 15 years of nuclear experience with 3 years of management experience in quality assurance, nuclear safety oversight, engineering and technical support, or regulatory affairs. Either this position or a quality assurance manager that reports to this position ,

must have a minimum of one year experience in quality assurance.

l l

General Manaaer The General Manager shall have a bachelors degree in engineering or the physical sciences or equivalent technical experience. He/she shall have 6 years of nuclear experience and 6 years of management experience (may be concurrent).

Enrichment Plant Manaaer The Enrichment Plant Manager shall have a bachelors degree in engineering or the physical sciences or equivalent technical experience. He/she shall have 6 years of nuclear experience and 6 years of management experience.

Ooerations Manaaer The Operations Manager shall have a bachelors degree in engineering or the physical sciences or equivalent technical experience. He/she shall have 4 years of nuclear experience with at least 6 months in a gaseous diffusion plant.

20

Maintenance Manaaer The Maintenance Manager shall have a bachelors degree in engineering or the physical sciences or equivalent technical experience. He/she shall have 4 years of nuclear experience with 6 months in a gaseous diffusion plant.

Production Suocort Manaaer l The Production Support Manager shall have a bachelors degree in engineering or the physical sciences or equivalent technical experience. He/she shall have 4 years of nuclear experience and 6 months in a gaseous diffusion plant.

Environmental Safety and Health Manaaer The Environmental Safety and Health Manager shall have a bachelors degree in engineering or safety disciplines, the physical sciences or environmental sciences, or equivalent j technical experience. He/she shall have 4 years of nuclear experience with at least 6 months at a gaseous diffusion plant.

Enaineerina Manaaer The Engineering Manager shall have a bachelors degree in engineering or the physical sciences and 4 years of nuclear experience. He/she shall also have at least 6 months experience in a gaseous diffusion plant.

Site and Facilities Sucoort Manaaer l The Site and Facilities Support Manager shall have a bachelors degree or equivalent  ;

technical experience and 4 years of nuclear experience with at least six months at a l gaseous diffusion plant. i l Trainina and Procedures Manaaer The Training and Procedures Manager shall have a bachelors degree or equivalent technical l experience. He/she shall also have 4 years of experience in nuclear facilities.

l i Plant Shift Suoerintendent Manaaer The Plant Shift Superintendent Manager shall have a bachelors degree or equivalent technical experience and 4 years nuclear experience. He/she shall have at least 6 months experience at a gaseous diffusion plant.

Plant Shift Suoerintendent The Plant Shift Superintendent shall have a bachelors degree in engineering or the physical sciences or equivalent technical experience and 4 years experience at a gaseous diffusion plant, or a high school diploma plus 12 years experience at a gaseous diffusion plant.

Operations and Maintenance Suoervisors First-line operations supervisors shall have a high school diploma and 3 years of plant operations experience with at least 6 months in a gaseous diffusion plant. First-line maintenance supervisors shall have a high school diploma and 3 years of plant maintenance experience with at least 3 months in a gaseous diffusion plant.

a

21

The minimum technical qualifications provided in the application are sufficiently detailed to I enable the staff to determine that the established technical qualifications are consistent with good industry practice, meet the requirements of 10 CFR Part 76 and are, therefore, acceptable.

By letter dated June 26,1996, USEC informed the NRC that a new Site and Facilities Support Manager had been hired. Although the individual does have over 17 years of commercial nuclear experience, he will not have the required 6 months of GDP experience by the time the NRC issues the Director's Decision on certification, in the interim, USEC will provide an advisor with 3 years of GDP experience to support the individual until he  :

meets the experience requirements. This compensatory measure is acceptable to the staff. I 3.2 Safety Review Systems The regulations in 10 CFR 76.68(a) require that plant changes must be approved by a safety review committee. USEC has established a safety committee to assist in the oversight function required by 10 CFR 76.35(a)(7) and to meet the requirement in 76.68(a). The safety committee, which is the Plant Operations Review Committee (PORC),

is described in SAR section 6.2 and in TSR 3.10. PORC functions in an advisory role and supports the General Manager. USEC has also established an ALARA subcommittee and may from time to time establish other subcommittees to provide assistance in conducting the reviews and assessments required by the PORC. The following paragraphs describe the PORC and the subcommittees.

3.2.1 PORC The PORC activities are described in TSR 3.10. The TSR establishes the membership, qualifications, meeting frequency and quorum, functions, responsibilities, and required records. The PORC performs multi-discipline reviews of plant activities to ensure that the day-to-day and proposed activities are conducted in a safe manner. The General Manager approves the procedure implementing the PORC activities. The Safety, Safeguards, and Quality Manager is responsible for auditing and oversight of PORC activities.

The PORC responsibilities include reviews of: (1) all proposed procedures and procedure changes as required by TSR 3.9.2; (2) all proposed changes to the SAR; (3) all proposed changes to the plans submitted with the application; (4) all proposed changes to the TSRs, the TSR basis statements, the Certificate of Compliance, or the Compliance Plan; (5) all proposed changes to the plant or the plant's operations that require a written safety analysis in accordance with 10 CFR 76.68;(6) all nuclear criticality safety evaluations and approvals; (7) all proposed requests for enforcement discretion; and (8) NRC-required event reports.

PORC membership is multi-disciplinary and shallinclude experience in the fo: lowing functional areas: cascade and chemical operations, engineering, maintenance, nuclear safety, nuclear criticality safety engineering, radiological safety, quality assurance, safeguards, and chemical, industrial, and environmental safety. The members have a bachelors degree in engineering or the physical sciences or equivalent technical experience; and 4 years of nuclear experience with at least 6 months experience at a gaseous diffusion 22

plant. The member representing nuclear criticality safety engineering must have the qualifications of a Senior Criticality Safety Engineer. A quorum consists of the chair and six members. The quorum must include members with technical competence in operations, engineering, nuclear criticality safety engineering, radiological safety, and quality assurance. The PORC meets at least once per calendar month. Written records shall be maintained..

The current PORC charter has not been reviewed to ensure consistency with TSR 3.10 and some of the procedures for implementation are not in place. These items are addressed in Compliance Plan issue 20, however, they will be completed prior to NRC assuming jurisdiction.

3.2.2 ALARA The ALARA committee is a subcommittee of the PORC and is discussed in 95.3.1.2 of the SAR. The committee has no approval, stop, or start work authorities;its authority is limited to review and recommendations. The ALARA Committee functions to: (1) communicate management's commitment to the ALARA Program; (2) monitor the implementation of the ALARA program and serve as advisor to plant management for maintaining occupational dose and environmental dose in accordance with ALARA principles; and (3) review, for the purpose of occupational dose and environmental dose reduction, proposed designs, practice, selected suggestions, and selected project schedules. Membership is designated by the PORC chair and represents various functional disciplines of the plant. The Radiation Protection Manager currently serves as the chair.

The' committee meets at least semiannually. Minutes and reports of the ALARA Committee are maintained as part of the PORC records package.

3.2.3 Subcommittees l

The PORC chair may establish subcommittees to provide assistance in conducting reviews and assessments, however, PORC retains the overall responsibility for the required reviews. The PORC chair approves the subcommittee procedures, membership, and member qualifications.

The commitments for a safety committee have been reviewed and the staff concludes that the commitments are consistent with good industry practice, meet the requirements of 10 CFR Part 76 and are, therefore, acceptable. TSR 3.10 requiring USEC to have a PORC is acceptable.

3.3 Operations Operations is one of the topics required by 10 CFR 76.87(c)to be included in the TSRs.

Operations is discussed in SAR 66.5. TSR 3.19 requires USEC to establish, implement, and maintain the operations program described in the SAR and lists several elements that must be addressed by the program. The TSR on the operations program is acceptable to the staff.

i 23 l

The work force for the facility is divided into a day shift and four rotating shifts which ,

provide continuous coverage of plant operations. The gaseous diffusion process operates continuously. The day shift works primarily Monday through Friday from 7:00 a.m. to 3:30 p.m. The day shift provides administrative support, activities such as design and fabrication, procedure development, classroom training, planning, and preventive maintenance. Most of the plant staff works on the day shift. The rotating shift organization has the prime responsibility for continued plant operation, exchange of information, and response to abnormal and unusual conditions to ensure safe operation of the facility. Typical activities include providing oversight and direction for all plant operations, monitor systems and equipment for proper performance, conduct routine back shift maintenance and emergency equipment repair, prepare equipment for day shift repair / preventive maintenance functions, and respond to emergency situations. TSR 3.2.2 establishes appropriate minimum staffing levels for the plant. Overal! staffing levels for the shifts are not fixed but are based on the expected or planned activities. The average shift staffing on back shifts is approximately 90. TSR 3.2.2 also establishes overtime guidelines for staff who perform safety functions. Currently, USEC is not able to comply with overtime guidelines that limit overtime to no more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period, or no more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7 day period. Compliance Plan issue 42 addresses the "

overtime. USEC commits to providing sufficient staffing and submit a revised TSR to reflect these guidelines by March 31,1997. In the interim, the TSR will aliow 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> in a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period and no more than 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> in any 7 day period. This situation is l acceptable since it is for a limited timeframe. The action, justification, and schedule are acceptable to the staff.  ?

Each shift organization is composed of a Plant Shift Superintendent (PSS); an assistant  ;

PSS; a cascade coordinator; a shift engineer, first line managers for the cascade buildings, power operations, utility operations, chemical operations, and maintenance; safety and health representatives who perform health physics and safety functions: Security Shift commander; Fire Department Shift Commander; and operators, maintenance mechanics, security police officers, and firefighters.

Operational activities are controlled by the PSS whose normal watch station is in the C-300 Central Control Facility (CCF). The CCF is the hub of the plant operational activity. ,

The overall UF, enrichment process is monitored at this location. Key plant operations can be performed remotely from here, key alarm systems are monitored, and plant communications systems as well as offsite communication capabilities are located at the CCF. Typical operational activities that are monitored and controlled from the CCF include determining and establishing optimal plant power level, executing or altering the i maintenance work plan, and maintaining necessary manpower level to support plant operations. l l

Plant operations, shift routines, and operator responsibilities are governed by conduct of  :

operations procedures. These procedures cover such items as shift routine, alarm l response guidelines, logs and rounds, required reading program, control of operator aides, I I

daily instruction and long term orders, etc. A key element of the shift routines is operator rounds; typically rounds consist of verifying operating parameters, c6mponent alignment, and surveying for abnormal conditions. Management provides day-to-day guidance on plant operations via operational memos and daily instructions; these guidance documents 24 1

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do not replace procedures. The Problem Reporting Program provides the mechanism for employees to make problem reports to management regarding abnormal events or conditions that may have potential to negatively impact safety, health, or security.

TSR 3.23 addresses worker protection for UF, hazards. USEC is required to establish, implement, and maintain worker protection measures to minimize the risk and mitigate the j consequences of releases of UF, reaction products with moist air, and other associated process chemicals. This TSR is acceptable to the staff.

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l The description of operations is adequate. The TSR on overtime guidelines when revised as described in the Compliance Plan will be consistent with industry practice and the TSR on minimum staffing provides adequate site coverage; the TSRs are, therefore, acc1ptable, l

l Compliance Plan issue 23 addresses aspects of the operations program that are currently not in place. Areas of noncompliance are in procedures and training (also covered by Issues 24 and 27). USEC will continue to use procedures and the training that is currently

)

1 utilized under DOE until the new procedures and the training program are in place. This is  ;

acceptable to the staff. )

3.4 Training The regulations in 10 CFR 76.35(a)(5) requires USEC to submit a " training program that meets the requirements of 9 76.95". According to 10 CFR 76.95 a training program must be " established, implemented, and maintained for individuals relied upon to operate, maintain, or modify the GDPs in a safe manner. The training program shall be based on a -

systems approach to training."

USEC describes its training program in 96.6 of the SAR. By TSR 3.4, USEC is required to establish, implement, and maintain the program as described in the SAR. The training program at PGDP consists of a number of training elements, some of which utilize the systems approach to training and some that do not. The following sections briefly describe the training program in place at PGDP.

3.4.1 Organization and Administration The training organization consists of a centralized staff which report directly to the training i manager. The staff consists of technical trainers, administrative personnel, and mid-level managers who are directly responsible for assisting with the training program in the following functional areas: general employee training; operations and maintenance technical training; radiological protection training; environmental, safety, and health training; subcontractor training; and training instructor /developec qualification. Training staff are also assigned to interface with functional line managers to coordinate training for functional areas such as cascade operations, general plant suprort, health and safety, and r,hemical, utility and power.

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USEC has established general employment policies and procedures to cover selection of personnel. Minimum qualification standards for key positions are documented and  :

maintained. For positions requiring SAT based training, the requirements are defined in  !

Training Development and Administrative Guides. l The training staff is responsible for Training Instructor / Developer Qualification. Training is  ;

provided to designated subject matter experts and the technical trainers who develop  :

and/or conduct qualification training. Training is provided in the use of satin conducting analysis, design, development, implementation, and evaluation of plant training programs.

The program includes initial training and periodic re-evaluation of skills and knowledge in l both material development and/or instructional competency. Instructors / developers are required to spend a minimum amount of time working in the area they instruct. Instructors l are expected to stay up-to-date on procedure / policy changes, modifications to TSRs, and i lessons learned.

Training consists of initial and continuing training. The initial training is intended to instill j an understanding of the fundamentals, basic principles, systems, procedures, and j emergency responses involved for a particular job. This training includes the applicable process safety training such as technical requirements training and nuclear criticality safety l training. The TSR training is designed to provide sufficient understanding of the safety limits and limiting safety system settings, limiting conditions for operation, surveillance requirements, design features, and administrative controls necessary for safe operation of the plant. Continuing training is provided to maintain and improve job-related knowledge ,

and skills. This would include lessons-learned, procedure and equipment modifications and i refresher training in such topics as emergency response, criticality safety, health physics, l TSRs, general plant rules, etc.

Employees are trained commensurata with their duties. All new employees, subcontractors, and visitors who require unescorted plant access receive general employee training consisting of radiological safety, nuclear criticality safety, hazard communication, emergency preparedness, and other general topics. This training is required biennially.

l Training attendance records, examinations, employee qualification record, and program records are maintained to document each employee's training. The training program records contain the job analysis data and history of the module, j 3.4.2 Systems Approach to Training The systems approach to training (SAT)is utilized at PGDP for those personnel who operate, maintain, or modify Q items or the structures, systems or components necessary  ;

to meet the double contingency principle. Training for personnel who operate items '

identified on the AO list differ from training for Q activities in the rigor and formality for the  ;

individual elements of the SAT. The SAT approach is discussed in the following sections.

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3.4.2.1 Systematic Analysis A needs/ job analysis is conducted using job incumbent / supervisor written surveys or the table-top method with subject matter experts. This is used to identify tasks affecting worker or public safety, safeguards, or protection of the environment as identified in the application. A facility-specific task list is developed and analyzed for the following l positions: cascade operators, chemical operators, electricians, instrument and electronic l mechanics, maintenance mechanics, system engineers, cascade coordinators, plant shift superintendents, health physics technicians, laboratory technicians, waste management operators, nuclear criticality safety engineers and specialist, and on-site transportation of UFe. Each task is rated on degree of difficulty, importance, and frequency and from this analysis tasks are selected for training. The tasks selected for training are matrixed to the associated procedures and training materials. These matrixes and associated training materials are reviewed every 3 years and updated as necessary by changes in procedures, facility systems / equipment, or job scope.

3.4.2.2 Learning Objectives Learning objectives for a specific task are established and incorporated in the training module developed for the task. Learning objectives state the knowledge, skills, and abilities the trainee must demonstrate to successfully complete the training. The objectives are updated as necessary based on changes in procedures and equipment.

3.4.2.3 Training Design and implementation Classroom lesson plans, on-the-job training (OJT) guides or other instructional materials are developed. These materials provide the guidance and structure to ensure consistent delivery of information from trainer to trainer and class to class. The classroom lessons are l used mair.ly to provide cognitive learning on fundamentals, theory, basic operating and maintenance principles, individual systems, system inter-relations, safety requirements, and overviews of the GDP processes. Instructional materials such as video, computer-based training, and self-study are sometimes used.

OJT is a method of providing in-field training and evaluation. The training is conducted in the actual work environment and demonstrates actual performance. This method is only implemented if managemen determines that manpower and operational conditions will not be impacted by the training "ctivities.

The lesson plans, OJT, and other materials receive technical reviews by designated subject matter experts, instructional reviews by training personnel, and final approval for use of line management. Training materials are approved by responsible line management and the l training staff before issuance.

3.4.2.4 Evaluation of Trainee Mastery Trainee progress is evaluated by technical trainers and line managem'ent through several different methods such as written examinations, oral examinations, and practical tests to 27 l

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l ensure mastery of the job performance requirements and learning objectives. Training courses consist of blocks for which evaluation can be performed for each individual block  !

or for several blocks together. Remediation is provided as appropriate. l 3.4.2.5 Evaluation of Program l

Training effectiveness is evaluated through student and supervisor feedback. Students evaluate the course and instructor at the end of the training; and a post evaluation is conducted.3-6 months after completion of the training. First-line supervisors provide l

feedback on the student's actual performance. The program is also subject to assessment  !

by line and training management through the Self Assessment Program and through the Quality Assurance Audit Program. The feedback obtained is utilized to refine or revise the training programs.

The PGDP SAT program addresses the necessary elements of a good program and meets the requirements of the regulations. The effectiveness of the program will be determined I by how wellit is implemented. The staff has reviewed the training program and concludes I that it is consistent with industry practices and meets the intent of the regulations and is, l therefore, acceptable.

1 Although the program as described in the SAR is acceptable, the systems approach to  !

training is still in the development stages and not fully implemented. This program is covered by Compliance Plan issue 24. The current schedule has the initial round of i training to be completed by June 30,1997. The second round will be completed by December 31,1997. The second round is for those positions that were not identified in

• the early stages as needing a SAT based program and include systems engineers, cascade coordinators, and nuclear criticality safety engineers / specialists. As part of the justification for continued operation , USEC will continue to use the training program in place under DOE. This will be acceptable until the full SAT based program is complete.

3.5 Procedures Although a procedures program is not specifically required by the regulations, it is  !

considered an essential part of the management controls and oversight program required by 10 CFR 76.35(a)(7) and by NOA-1. USEC is committed to the use of approved and controlled written procedures to conduct nuclear safety, safeguards, and security activities  !

for the protection of the public, plant employees, and the environment. Procedures prescribe the essential actions or steps needed to safely and consistently perform safety related activities. The procedure program is described in 66.11 of the SAR. TSR 3.9 addresses the procedure program. The program is briefly described in the following  !

paragraphs.

USEC uses a four level procedure hierarchy. Level 1 consists of policy statements issued by USEC and apply to all GDP personnel. Level 2 is standard practice procedures that i apply to both sites or to more than one organization. Level 3 procedures are issued at the organizationallevel and apply to more than one group. Level 4 procedures are those l issued and applied within a group or subfunction. Section 6.11.4.1 and Appendix A to 96.11 of the SAR describes the minimum activities that shall be covered by written l l

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procedures. Topics covered are administrative procedures; system procedures that address startup, operation, and shutdown; abnormal operation / alarm response; maintenance procedures that address system repair, calibration, inspection, and testing; emergency response; and any task that is described in, or implements a commitment that is described in the SAR, TSRs, and plans submitted with the application.

In Appendix B to 96.11 of the SAR, USEC has listed the specific subsections of ANS 3.2-1994 that will be utilized in the procedure program. Procedures are developed or modified l

through a . formal process. Procedure development, control, and use is a process that consists of nine basic elements: identification, development, verification, review and comment resolution, appr' oval, validation, issuance, change control, and periodic review.

PORC review is required for procedures required by TSR 3.9 and for intent changes to those procedures. All procedures are periodically reviewed to ensure continued accuracy l and usefulness. Emergency, Operating, Alarm Response and procedures dealing with l highly hazardous chemicals are reviewed on a 1-year cycle. All procedures designated as

In-Hand, which involve liquid UFe handling activities, off-normal procedures, and Nuclear Material Control and Accountability procedures are reviewed on a 3-year cycle; all other i procedures are on a 5-year review cycle.

The procedure TSR 3.9 requires that written procedures shall be prepared, reviewed, approved implemented, and maintained. The TSR covers the review and approval of l procedures and allows for temporary changes. The procedure program as described in 56.11 and the TSR have been reviewed by the staff, the program is consistent with good industry practice and is, therefore, acceptable.

USEC is currently in the process of a Procedure Upgrade Program; not all of the procedures needed to implement programs in the application have been prepared or upgraded and approved. This issue is addressed in Compliance Plan Issue 27. The JCO presented in the Compliance Plan is acceptable. The site will continue to use existing procedures until new or revised procedures are complete. Procedures that contain action statements and operating limits from the TSRs will be in place prior to NRC taking jurisdiction. Levels 2,3, and 4 procedures that are related to AQ-NCS items will be completed by December 31, 1996;those related to O items by March 31,1997; and those implementing AQ and NS items will be completed by December 31,1997. This approach and schedule is acceptable  ;

to the staff.

3.6 Human Factors i Human factors is not specifically addressed in the regulations. However, USEC has proposed a human factors program. PGDP incorporates human factors considerations in engineering design work associated with new equipment and facility modifications; preparation, validation, and use of procedures; and in development of training and qualifications of personnel who operate, maintain, or modify structures, systems, and components relied upon for safety. Human factors is considered in problem reporting and investigation. Human actions required by the TSRs to prevent or mitigate accidents are systematically evaluated for human factor considerations on a 3-year cycle, including l 29

accessibility, visibility, ergonomic capability, suitability of the environment for the required activity, and interferences. This program will result in human factors considerations for those actions important to safety. The staff concludes that the program is acceptable.

3.7 Audits and Assessments An audit and assessment program is not specifically required by the regulations but is considered part of the management controls and oversight program required by 10 CFR 76.35(a)(7) and the quality assurance program required by 10 CFR 76.35(d). PGDP has established a system of audits and assessments that is designed to ensure that the health, safety and environmental programs are adequate and effectively implemented. The Audit and Assessment Program is described in SAR Section 6.8. TSR 3.5 requires USEC to implement the program described in the Quality Assurance Program (OAP) and the SAR.

)

The program is designed to ensure comprehensive program oversight every 3 years.

Audits are conducted by the Safety, Safeguards and Quality organization as part of the OA Program. The Audit Program is conducted in accordance with procedures and checklists by qualified auditors. Audits are used to verify the effectiveness of health, safety, and environmental programs and their implementation. Audits are also used to determine the effectiveness of the Assessment Program. Audit results are documented and reported as specified in plant procedures. The Audit Program is conducted in accordance with $2.18 and Appendix A of the Quality Assurance Program.

The Assessment Program at PGDP consists of program assessments, divisional self assessments, management self assessments, and problem reporting. Personnel from the area being assessed may perform the assessment if they do not have direct responsibility for the specific area being assessed. Observations are resolved by the responsible division management. The Assessment Program is used as a tool to determine the effectiveness of the programs, compliance, and to verify corrective actions. Problem reports are issued on items of noncompliance or nonconforming conditions. In addition to the assessments, all plant employees are responsible for writing problem reports on safety, operating, and noncompliance items. Problem reports are screened and corrective actions are taken as appropriate.

The staff has reviewed the PGDP Audit and Assessment Program and concludes that it is consistent with good industry practice and is, therefore, acceptable. )

i 3.8 Quality Assurance '

The regulations in 10 CFR 576.35(d) and 576.93 require USEC to submit a quality l assurance (OA) program that satisfies "each of the applicable requirements of ASME NOA-1-1989" or " acceptable alternatives to the applicable requirements." The regulations require USEC to " execute the crite.;iin a graded approach to an extent that is commensurate with the importance to safety." USEC submitted the " Quality Assurance Program" (OAP) with the application. The OAP establishes the minimum requirements for those items, activities, and services within the scope of the OAP. USEC has committed in the OAP to meet the Basic Requirements and Supplementary Requirements of ASME NOA-1-1989 or has committed to alternatives acceptable to the NRC.

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3.8.1 QA Organization

\

The Safety, Safeguards, and Quality Manager is responsible for ensuring implementation of l the OAP activities at the site. This manager directs PGDP quality assurance functions involving audits and oversight of plant operations. The OA organization is divided into three groups: Independent Assessment, Quality Systems and Nuclear Safety Assurance, independent Assessment performs audits and assessments of activities that affect safety, quality, and the environment. Independent Assessment also ensures the effectiveness of corrective. actions. Quality Systems maintains the QAP, prepares procedures, reviews other procedures for inclusion of quality requirements, and participates in operational readiness reviews. Nuclear Safety Assurance performs engineering assessments, provides nuclear safety oversight, and performs selected operational experience reviews to determine affect on plant safety. In addition, there are other positions in the organization, particularly the Engineering Manager, that have OA responsibilities. Personnel responsible for QA activities are acceptably independent from operations personnel. )

3.8.2 Quality Assurance Program USEC has established a OA program that has three categories: O, AQ, and NS. Systems, structures, and components (SSCs) are categorized as Q, AQ, or NS by Engineering. The requirements of the main body of the OAP apply to the Q items and activities. Appendix A  !

of the OAP defines the extent to which the OAP applies to AO items and activities.

Appendix A, Section 1 describes the OA program for Nuclear Criticality Safety (AO-NCS) items and activities required to meet the double contingency principle. Appendix A, Section 2 describes the OA program for other AO items and activities. Appendix A, Section 3 describes the OA program for AO structures. The formal OA Program is not applied to NS (non-safety) items. SAR s3.15 lists the systems and boundaries for the O and AQ items, except for AQ-NCS items. AQ-NCS SSC boundary definitions are documented in a manual for each facility. In accordance with TSR 3.22, the system l boundary documents shall identify utilities required by the SSC to perform its safety function.

The requirements of the OAP apply to activities affecting the ability of O and AQ SSCs to perform their intended function. These activities include designing, purchasing, fabricating, handling, receiving, shipping, storing, cleaning, erecting, installing, inspecting, testing, operating, maintaining, repairing, and modifying. The QAP is implemented through policies, procedures, instructions, specifications, drawings, procurement documents, contractual documents, and other documents. These documents provide measures that ensure that QA activities are planned and accomplished to meet the goals and objectives of the OAP.

Procedures that implement the requirements of the OAP are reviewed by the affected organizations and authorized by the responsible manager.

The OAP requires the establishment of an indoctrination and training system to provide

~

confidence that proficiency is achieved and maintained by PGDP personnel in the performance of their quality affecting activities. The QAP acceptably describes the qualification of personnel who verify the quality of items and work activities at PGDP.

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Quality is verified through surveillance, inspection, testing, checking, and auditing of items and work activities. The OAP requires that quality of items and work activities be verified by qualified personnel who are not responsible for performing the actual work. Quality verification activities are performed in accordance with procedures, instructions, and/or checklists by personnel who have been qualified in accordance with applicable codes, standards, or company training programs. Training and qualification of other personnelis addressed in Section 3.4 of this CER.

The OAP establishes a comprehensive audit system to ensure that the OAP requirements and related supporting procedures are effectively and properly implemented during operations. Audits willinclude an objective evaluation of OA practices, procedures, and instructions; work areas, activities, processes, and items; the effectiveness of implementation of the OAP; and conformance with policy directives.

The OAP requires documentation of audit results and review by the management personnel who have responsibility in the area audited to determine and take appropriate corrective action as required. Reaudits or surveillances are performed to determine that deficiencies have been effectively corrected and that the corrective action precludes repetitive occurrences. Audit findings are provided to the General Manager, the USEC Nuclear Regulatory Assurance and Policy Manager, and the Executive Vice President, Operations.

On the basis of its review, the staff concludes that the OAP for PGDP is structured in accordance with 10 CFR 676.93 and ASME NOA-1-1989. The OAP forms the foundation for the overall quality assurance program and describes how the requirements of 10 CFR 976.93 and ASME NOA-1-1989 are satisfied. The OAP and its implementing procedures control quality-related activities involving O and AO items to satisfy the requirements of 10 CFR s76.93. The staff concludes that the OAP is consistent with good industry practice, meets the requirements of 10 CFR s76.93 and the basic and supplemental requirements contained in ASME NOA-1-1989, and is therefore acceptable.

The OAP is not fully implemented at the site and the program is covered by Compliance Plan issue 29. Procedures and training for the conduct of audits, procurement, handling, and storage activities described in the OAP have not been fully developed or upgraded. In addition, noncompliances in other areas such as configuration management, maintenance, and records management also impact the OA implementation. Procedures and training on the audit aspects will be complete by December 31,1996 as will aspects for AQ-NCS items and services. Procedures and training for aspects related to O items will be in place by March 31,1997, and for other AO items and services will be complete by December 31,1997. The O and AO systems have been identified, however, the boundaries and support systems have not been identified in all cases. This will be complete before NRC assumes jurisdiction for O and AQ-NCS items and by October 1,1997, for other AO items. Until USEC reaches full compliance, they will utilize procedures and performance criteria in place under DOE. The key aspects will be in place I by the time NRC assumes jurisdiction. The plan, schedule, and justification are acceptable to the staff.

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3.9 Event Reporting and Investigations The regulations in 10 CFR 76.120 and other applicable sections referenced in 10 CFR i 76.60 identify the reporting requirements for the GDPs. The PGDP Event Reporting and l Investigation Program is described in 96.9 of the SAR. In addition to the requirements for l oral notifications and written reports, USEC is required to determine root causes, adequate I corrective actions, and lessons leamed.

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USEC com.mitted to an Event Reporting and Investigation Program which is part of its management controls and oversight program. The program includes identification and categorization of events, NRC notifications, and an analysis to determine root causes, corrective actions, and lessons learned.

The program appears adequate to ensure that abnormal conditions are identified, reported, and investigated. Plant personnel are required to report to their line manager or Plant Shift Superintendent (PSS) abnormal events or conditions that may affect plant safety. The PSS is responsible for categorizing and reporting events to the NRC. Table 6.9-1 of the SAR is i a list of event reporting requirements that USEC is committed to. It includes reporting criteria and the reporting timeframe. The table contains items such as criticality reporting in accordance with NRC Bulletin 91-01 and safety system actuation reporting; although i j these are not required by the regulations, USEC has agreed to report them. Reports of l classified information are done in accordance with the Security Plan for the Protection of  ;

Classified Matter. Compliance Plan Issue 44 deals with operational trips and alarms that l are set to avoid an actual actuation of a safety system. PGDP plans to review system designs where operational trips and alarms coincide with the setpoints for safety system actuation based on the same monitored parameter and the same equipment actuated. The results of this review will be submitted for NRC review and approval by December 31, 1997. In the interim, the trips and alarms will be treated and reported as an actuation of a safety system. This is acceptable to the staff.

An investigation will be conducted for each reportable event in accordance with written procedures. The Nuclear Regulatory Affairs Manager will form the investigation team and ensure the selection of qualified investigator (s). The investigation includes, at a minimum, analyzing available information, interviewing involved individuals, identifying root causes, and developing corrective actions. Documentation related to the event will be retained in accordance with record retention requirements described in s6.10 of the SAR.

Responsible management will develop and approve corrective actions to address root causes. The PORC will review and concur with the corrective action plan. The schedule for implementing corrective actions is defined in each corrective action plan. These actions are entered into a commitment management database to ensure proper implementation and  ;

closure. Management will verify the completion of the corrective action before its closure i in the database.

Lessons learned information from observations, events, problem reports, and expenences from GDPs and other related industries are reviewed. The informatio'n is communicated I within the plant and between the two GDPs. Management reviews the information and  ;

determines if further action needs to be taken.  !

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S The staff has reviewed the PGDP Event Reporting and Investigation Program and concludes that it is consistent with good industry practice and is, therefore, acceptable. Although the procedures to implement this program have been completed, the associated training has not been conducted. Training will be completed prior to the time NRC takes jurisdiction over the PGDP. The procedures will become effective on the date NRC assumes l jurisdiction. Until NRC takes over jurisdiction, PGDP will continue to make its reports to l DOE following established DOE criteria. Compliance Plan Issue 25 is acceptable to the l staff.

3.10 Record Management A records management program is not specifically required by the regulations, however it is considered part of the management controls and oversight program required by 10 CFR 76.35(a)(7)and a necessary part of the OAP. The PGDP Records Management and Document Control (RMDC) Programs are described in SAR 66.10. TSR 3.24 addresses the record retention program.

USEC committed to centralized RMDC Programs. The records must be legible, readily retrievable, protected from damage, and retained for a certain period of time. The RMDC programs are implemented through procedures. The procedures address the criteria for the handling, verification, identification, authentication, indexing and filing, retention and disposition, correction, protection, storage, transmittal, retrievability, distribution, and assessment of records and controlled documents. These programs are administered by the Manager of Documents and Records.

The RMDC Programs appear to be adequate. The programs ensure that records and documents important to safety and safeguards and security will be controlled, maintained, and distributed according to written procedures. The programs also contain measures for special conditions, such as handling contaminated records and classified information.

The Records Management Program contains requirements for access control to ensure that only authorized personnel have the access to records. The record storage areas are fire rated and can protect and preserve records from loss, theft, tampering, unauthorized access, damage, and deterioration.

The use of computer codes and computerized data in the RMDC Programs is controlled and maintained according to a written manual and procedures. It addresses both access security and virus prevention. Precautions are taken to ensure validity of computer codes and data.

The QA Program ensures that USEC's suppliers establish a comparable Records Management Program. It also ensures that changes and corrections to records and documents are reviewed and approved by authorized personnel and are distributed and used at the locations where the prescribed activity is performed.

The RMDC Programs have been reviewed and it is concluded that thb programs are consistent with good industry practice and are, therefore, acceptable.

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4 As described in Compliance Plan issue 26, USEC has not completed the development and implementation of administrative and technical procedures for the centralized RMDC programs. The RMDC procedures will be updated along with other procedures in the Nuclear Safety Procedures Upgrade Project. Pre-existing records and documents will be turned over and incorporated into the centralized records management system via a turnover schedule that is based on their importance to safety. The records and documents needed to demonstrate that the facility can be operated and controlled within the operational envelopes specified by the application for NRC certification are either retrievable.or are being recreated. Legacy records and documents, which w not readily retrievable at the time of certification, will be retrieved, updated, or re-created when they are needed for the evaluation of proposed plant changes pursuant to 10 CFR 76.68. All records and documents will be incorporated into the centralized records management system by June 30,1998. The schedule is acceptable to the staff.

3.11 Maintenance The regulations in 10 CFR 76.87 require the TSRs to address maintenance. The description of the PGDP Maintenance Program is contained in 96.4 of the SAR. By TSR 3.15 USEC is required to establish, implement, and maintain the program.

The maintenance unit consists of five groups; the group managers report to the Maintenance Manager who in tum reports to the Enrichment Plant Manager. The five groups are Mechanical, Electrical, Instrument and Controls, General Production Service Shops, and Calibrations and Electronics.

1 The maintenance and surveillance program consists of a mix of corrective maintenance,  ;

preventive maintenance, and instrument calibration. Routine maintenance work is '

identified, prioritized, planned, scheduled, executed, and closed out in accordance with a work control program. The Maintenance Program is conducted in a graded approach commensurate with safety as described in the Quality Assurance Plan.

Corrective maintenance are those actions intended to check, troubleshoot, and repair equipment that has degraded or failed. Corrective maintenance is initiated by issuance of a work order that identifies and prioritizes the need for maintenance. A job package is assembled that identifies necessary materials and schedules, identifies required support services, and provides for post maintenance testing as required. Work is performed after a pre-job briefing; completed maintenance is documented.

Preventive maintenance are tasks performed on a periodic basis to prevent failures, j facilitate performance, and maintain life of equipment. Maintenance procedures identify I and specify the schedule for preventive maintenance. The bases for tasks and their periods of performance are developed through a review of manufacturer recommendations, available industry standards, and historical operating information. Engineering (the design authority) approves and controls the documentation of preventive maintenance bases.

l Preventive maintenance is scheduled according to plant conditions arid the specified task i

frequencies. TSRs specify actions to be taken if scheduled maintenance activities are not completed within the required period for equipment relied upon for safety. For items not 35

governed by TSR requirements, the Operations Manager and system engineer evaluate the effect of extending the scheduled period. This includes a determination of any compensatory actions and a rescheduling of the preventative maintenance. Permanent changes to preventive maintenance tasks and schedules requires design authority approval.

Work control procedures apply to both corrective and preventive maintenance. Corrective maintenance is initiated by any manager identifying the need for maintenance by issuing a work order. The Work Control Supervisor reviews the work order to determine the need for a work. package. Procedures in the work control program i.dentify equipment that requires a work package before maintenance can be performed.

Work control procedures define the content of a work package. The minimum content for a safety system work package consists of the work order, a planning checklist, written work instructions, and a safety system data sheet. The package may also contain equipment specific procedures, quality control inspection requirements, nuclear criticality safety analysis, and health physics requirements. The package is developed using a planning sheet that lists the criteria for items and systems requiring System Engineering, Industrial Safety, Security and Fire Services, Health Physics, Nuclear Safety, and Waste Management approval. Engineering review and approvalis always required for modification to plant equipment. The appropriate Maintenance Group Manger reviews and approves work packages prior to starting work. Authorization for release of equipment is required from designated operations managers before removing equipment from service.

The staff concludes that the Maintenance Program meets the requirements of 10 CFR 76, is consistent with industry practice, and is, therefore, acceptable.

PGDP Compliance Plan, Issue 22, describes the aspects of the Maintenance Program that are not currently being implemented. The maintenance history and trend analysis program will not be in place until September 30,1997. The work control process needs to be upgraded for O and AO items; this will be complete by December 31,1996 for AQ-NCS items, by April 30,1997 for O items, and June 30,1998 for other AO items. Revision of the measuring and test equipment and process equipment calibration programs; completion dates are the end of 1996 for AQ-NCS items, by March 31,1997, for O items, and ,

June 30,1998, for other AO items. Procedures and training need to be upgraded to meet l new requirements in the application. Procedures and training related to AQ-NCS items will be finished by December 31,1996, by March 31,1997 for O items, and for other AO items June 30,1998. TSR related items will be complete prior to NRC assuming jurisdiction. Vendor's manuals used for maintenance activities will be entered into the document control and records management system by March 31,1997. Interim measures that will be implemented until the entire program is in place are described in the justification for continued operation. Work control procedures developed under DOE will continue to be followed. A minimum work package composition has been described.

System engineers have been assigned responsibility for specific safety systems. The system engineers will provide technical support for maintenance activities including work package development, determination of post-maintenance testing requirements, 36

l 3'

observance of surveillance testing, and review of procedures. The staff finds the justifications for continued operation and the schedules for achieving compliance to be acceptable.

3.12 Configuration Management The regulations in 10 CFR 78.68 require USEC to " maintain records of changes in the plant and of changes in the programs, plans, policies, procedures and operations described in the approved application, and copies of the safety analyses on which the changes were based." This is accomplished via the Configuration Management Program which is described in 66.3 of the SAR.

P The Configuration Management Program is used to control changes and maintain the plant configuration to ensure accurate, current design documentation that matches the plant's physical configuration while complying with applicable requirements. Nuclear Regulatory Affairs controls documentation and coordinates implementation of approved changes to the TSRs and conditions of the certificate of compliance, controls documentation of changes to TSR Basis statements, maintains records of changes and change control documentation,  :

and maintains the program plans and documentation of changes. They also maintain a l flowdown listing of procedures and training derived from TSRs and the program i descriptions and plans in order to ensure that the overall configuration of conditions as described in the application are maintained. Control of changes to the physical plant and control of the physical plant configuration are managed by the Design Authority.

The Configuration Management Program includes identification of the structures, systems,

• equipment, components, and design features credited for safety and safeguards, organizational descriptions of duties and responsibilities; and administrative controls, procedures, and policies, to implement and document activities that maintain the plant baseline configuration.

Engineering has established the Change Control Board which provides' a technical review of all proposed changes. The review determines the necessity of the change and the requirement for review and approval by the PORC. If the change is not a substitution, the proposal proceeds through the design modification process. An engineering group is assigned to process the change request. A design / project team is assigned that may include personnel from Engineering, Safety Analysis, Safety, Nuclear Criticality Safety, Operations, and Maintenance as appropriate. A safety evaluation is performed by Safety Analysis. Modifications are evaluated for any required changes to procedures, trs;ning, testing, or regulatory documentation.  !

l The staff has reviewed the Configuration Management Program and concludes that it is consistent with industry practice and meets the intent of 10 CFR 76.68, and is therefore acceptable.

Although the program once fully implemented will be acceptable, there are aspects of the program that are not currently in place. These areas of noncompliance are described in the PGDP Compliance Plan, issue 21. Elements not fully implemented include the flowdown of commitments into procedures and training, O and AO components have not been 37

identified, development of the baseline documentation that establishes the design requirements has not been completed for all O and AQ items, development and implementation of improved records management and document control program is not complete, development / upgrade of engineering procedures associated with the change control process, the assessment program to evaluate the effective implementation of the configuration management program, and the training program for the new programs and l procedures. The completion dates for those items related to the O and AQ-NCS lists is by l December 31,1996, those associated with other AO systems is in October 1997, items I associated. with TSRS will be complete by the time the NRC assumes jurisdiction, engineering procedures will be complete by March 31,1997, and the training and assessments aspects will not be compete until the end of 1997. Interim measures that will be taken until the prograrn is fully in place are described in the Compliance Plan. All requests for changes, other than like substitution, will obtain engineering services and will be reviewed by the Change Control Board. Design requirements will be developed / recovered on a case-by-case basis to support any change request review.

Listings of all changes to plant and equipment safety systems shall be maintained for each calendar year and will be available for review. Change control procedures currently in l place under DOE will be used until the development of Configuration Management Program procedures is complete. The staff finds the justifications for continued operation and the j schedules for achieving compliance to be acceptable.

3.13 Management Controls The regulations in 10 CFR 76.35(a)(7) require the SAR to contain a " description of the j management controls and oversight program to ensure that activities directly relevant to nuclear safety and safeguards and security are conducted in an appropriately controlled  :

manner that ensures protection of employee and public health and safety and protection of l national security interests". USEC and LMUS have established management systems with associated policies, administrative procedures, and management controls to ensure protection of the health and safety of workers and the public, protection of the environment, and for the common defense and security. Management Systems and Programs are described in Chapters 5 and 6 of the SAR, the TSRs, and in the Program Plans.

Preceding sections discussed many of the programs that come under the consideration of management controls. Primary among these are an organizational structure that has clear assignment of responsibilities and independent reporting chains for the safety functions, PORC, OA, a configuration management program, an audit and assessment program, and an investigation and reporting process. The PORC provides the necessary review for management to make informed decisions. The Audit and Assessment Program provide assurance that programs are being implemented in accordance with regulations and procedures. A QA Program is in place to promote safe, reliable, and efficient plant operation. PGDP investigates incidents to determine root cause and lessons teamed.

Items from the lessons learned are integrated into the Procedures and Training Programs as appropriate. PGDP has a commitment tracking and corrective action management system that prioritizes plant actions consistent with their safety and safegua~rds significance.

These items along with a Procedures and Training Program, a Maintenance Program, the Configuration Management Program, and other programs will provide the necessary tools 38

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

6 0

i for USEC to operate in a safe, reliable fashion once in place. It is apparent from both the staff observation reports and the DOE inspection reports, that operations at PGDP have experienced numerous problems in the area of management controls. Many aspects, such as PORC, engineering evaluations, and QA are still in the developmental stages. i Additionally, the Corrective Actions Program and root cause determinations have -

experienced implementation problems. Procedure adherence has also been a problem at the plant. These activities will be closely followed by the NRC staff to ensure that the programs are being effectively implemented. .

l Some management control aspects are not currently in place and are addressed in l Compliance Plan issue 19. Items from the OAP, TSRs, and other requirements have not l been flowed down to policies and procedures. Additionally organizational responsibilities l

and authorities have not been flowed down to position descriptions. An initial flowdown l will be completed by the time the NRC takes jurisdiction of the PGDP. The final flowdown will not be complete'until the Procedure Upgrade Program is complete on December 31, i 1997. This approach is acceptable to staff; the key aspects will be in place before NRC assumes jurisdiction. In the interim, USEC will continue to utilize the tools currently in place, including procedures, training programs, tracking systems, etc. The justification for continued operation and the schedule for completion are acceptable to the staff. ,

I a

l l

39

Chapter 4 FACILITY AND PROCESS DESCRIPTION The regulations in 10 CFR 76.35(a)(8) require USEC to provide a " description of the principal structures, systems, and components of the plant " Chapter 3 of the SAR provides the facility and process description.

The PGDP is designed to separate a feed stream containing the naturally occurring proportions of uranium isotopes into a product stream enriched in the 23sU isotope and a tails strearn depleted in the 23sU isotope. The chemical form of the working material of the plant, UFe , does not require chemical transformations at any stage of the process. In the three primary steps of the process, UFe is volatilized from a feed cylinder, passed through the diffusion process, and condensed into product or tails cylinders. A process overview of the PGDP uranium enrichment operations is presented in Figure 4. The description of the process presented in this chapter follows the flow of UF, from its receipt through processing to the disposition of product and tails material. The descriptions are drawn from the SAR. Additional detail and building and equipment drawings are provided in the SAR. At the end of each section is a listing of the O and AO systems (except for AQ-NCS). Both the O and AO system boundaries are further described in SAR 93.15.

4.1 UFe Receipt and Feed Feed material is received at PGDP in cylinders that contain 10 to 14 tons of UFe in the solid state. Cylinders containing UFe may arrive at PGDP by rail or truck. Normal assay feed materialis received at the C-360 Toll Receiving and Transfer Facility or the cylinder unloading facility in C-400. The feed cylinders are unloaded, inspected, and weighed.

• Some of the cylinders are liquified in an autoclave for sampling to ensure conformance with feed material specifications and for uranium accountability. Each cylinder is cold pressure checked prior to heating; the pressure is an indication of purity.

For sampling the UF, cylinder is positioned in one of four 96-inch diameter autoclaves in C 360 with the cylinder valve at the 12 o' clock position. The cylinder is connected to a manifold by means of a copper tubing pigtail. After pressure testing all connections, the UF cylinder valve is opened, the emergency cylinder valve closer is attached, the autoclave shellis closed, and a hydraulically operated rotating ring seals the movable shell of the autoclave to the fixed head.

A thermovent bleeds atmospheric air from the shell. Autoclave temperature is controlled by utilizing a cascade control scheme within a feedback loop controller. The steam piessure is limited to a maximum of 8 psig which corresponds to a saturated steam temperature of 235 F. During heating, the cylinder is valved to a rupture disc on the manifold header to mitigate the possibility of cylinder rupture. The rupture disc will relieve cylinder contents to surge drums at 100 psig (the lowest MAWP of cylinders heated). A pressure transmitter monitors the cylinder pressure and closes the steam supply valves if the cylinder pressure exceeds 90 psia.

When the heating cycle is completed, the cylinder valve is closed, an~d steam is evacuated from the autoclave. The pigtail is purged and evacuated, the autoclave is opened, the pigtailis disconnected and the cylinder is rotated to the 3 or 9 o' clock position to permit 40

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i 1

41

liquid sampling. The pigtail is reconnected and pressure tested and the cylinder valve is ,

partially opened before closing the autoclave shell. The steam block valve may be opened l to maintain temperature during sampling.

During sampling, the inlet valve to the pipette in the sample cabinet is opened and the l

pipette is allowed to fill. The pipette sample evacuation valve is opened and the entire sampling system is flushed with liquid UFe to the evacuation drums. The evacuation valve is closed and the pipette is valved to an evacuation source to ensure all UF,is removed from the pipette and sampling manifold. All system valves are closed and the pipette inlet valve is opened to fill the pipette with liquid UFe. After closing the inlet valve, the sample is transferred to a sample container. Following sampling, the pigtail is purged, evacuated, and disconnected from the UFe cylinder. The UFe cylinder is then rotated to return the cylinder valve to the 12 o' clock position. l After sampling, the cylinders are moved by crane and cylinder cart to a storage yard where the cylinders are allowed to resolidify prior to transport to the feed facilities in C-333-A and C-337-A. The dual hoist / trolley and control systems on the east bridge crane is being replaced with a single polar hoist / trolley and control system. The crane will be replaced by September 30,1996. Compliance Plan Issue 4 addressing the cranes is acceptable to the staff.

There are five feed stations located in C-337-A and four feed stations located in C-333-A.

These stations allow simultaneous feeding of material with different assay levels. Each feed station has two 72-inch diameter containment type autoclaves, a feed metering device, a control valve for the upstream pressure, evacuation dad purge lines, an l evacuation jet for heeling cylinders, and associated headers and valves. I Each autoclave in the feed facilities is a cylindrical pressure vessel approximately 22 ft long with an internal diameter of 6 ft which is mounted with its long axis horizontal. The autoclave shell moves approximately 15 ft from the fully closed to the fully opened position. The moveable shell of each autoclave is secured to the fixed head with a j hydraulically operated flanged closure, which is sealed with a gasket.

The autoclaves in the feed vaporization facilities are designed to provide steam heating of 2.5,10,and 14-ton cylinders of UFe to a maximum operating temperature of 235 F.

Autoclaves are designed to meet the seismic criteria of a horizontal acceleration of 0.18g and a vertical acceleration of 0.12g and to fulfill the requirements of Section Vill, Division 1, of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. The autoclaves are containment-type autoclaves with a MAWP of 200 psig at 250 F. Autoclaves are designed to minimize standing water inventory to contain the products of a worst case UFe release. The feed vaporization autoclaves have a pressure relief valve that will relieve any pressure within the autoclave in excess of the MAWP of the autoclave to atmosphere through a vent line, which extends above the roof.

Each autoclave has a cylinder valve closer that consists of an air motor outside the autoclave head and a shaft to connect to the cylinder valve. When a' cylinder is placed in the autoclave, the shaft is connected to the cylinder valve stem when the valve is opened.

42

I The valve can then be closed from outside the autoclave by pressing the air motor start l button. The air motor is not reversible, so the valve must be opened manually by an l l operator.

l l

Service lines penetrate the autoclave head and are closed by redundant isolation valves i that close automatically to prevent the escape of reaction products in the event of a UFe release inside the autoclave. The service lines that penetrate the head include the steam supply line, conductivity cell steam sample lines, UF, feed line, condensate drain line, and thermal vent. The autoclave relief line connects to the rupture disk and relief valve. A cylinder valve closer shaft containing packing and flanges to prevent leakage, a pressure tap, and temperature probes also penetrate the stationary head. ,

Each autoclave is designed to withstand the reaction pressure developed by the exothermic l reaction of UFe with the steam vapor and the condensate present. The maximum pressure generated by the reaction can be controlled by limiting the amount of water present. To ensure that the maximum allowable water inventory of the autoclave is never exceeded, i redundant cor..fer sate level probes are mounted at the same elevation in the condensate l drain slightly b'; low the autoclave floor. Any accumulation of condensate in the drain line reaching the level of these probes automatically shuts the steam supply isolation valves and thermovent line and prevents the addition of more water. l l

In the event of a release, reaction products of UFe and water are detected by a 1 l conductivity cell and/or the pressure instrumentation, either of which effect autoclave containment by closure of isolation valves at the penetrations. The conductivity cell which constantly measures the conductivity of a sample of the autoclave atmosphere,is intended 1 to detect small UF, releases that may not result in a pressure increase detectable by the high pressure isolation system. The cell is sensitive to the HF component of the UF, reaction with water. When triggered, this cell initiates a signal that closes all isolation I valves.

For large releases the use of a pressure sensor provides a more rapid isolation than that obtained from a conductivity cell. For those releases having a large and rapidly rising pressure the autoclave isolation system is initiated by the pressure sensor.

l As a protection against over pressure, each autoclave is provided with a rupture disc backed by a 3-by 4-in. pressure relief valve. The relief valve is set to relieve in accordance i with ASME Code, Section Vlli, and is designed to handle a 40% UFe,60% HF mixture l resulting from a UF, release rate of 1,786 lbs of UF e/ min. Pressure sufficient to open the

' relief valve will vent the autoclave to the atmosphere. This relief system would function i only if there were an excessive amount of water in the autoclave at the time of a UF.

release within the autoclave.

Autoclave temperature is maintained constant by an automated controller with a maximum l l

of 235 F. A timed start-up system provides the warning mechanism to protect against a plugged valve or pigtail. If the cylinder pressure fails to reach 24 psia within the allowed ,

time frame, the steam supply isolation valves are closed and alarms are sounded. Heating j of feed cylinders is started approximately three hours prior to the estimated time that the cylinders will be needed for feed. i 43 l

9 l

l The UFe feed flow system maintains the desired flow rate to the cascade. Alarm conditions available at each feed station are: (1) indication that the feed control valve is fully open, signaling that a loss of flow is imminent (2) indication of low UFe header pressure (3) indication of high UFe header pressure and (4) indication of high total UFe flow to the cascade.

Since material with different assay levels can be fed simultaneously from different feed stations, a feed header crossover system buffer is provided. The piping upstream of feed jets that connect feed lines carrying UFe of different assays is double blocked and maintained at a vacuum. This pressure is monitored by a pressure indicating switch set to alarm in the event of a change in pressure between the block valves. The piping downstream of feed jets that connect feed lines carrying UFe of different assays is also double blocked and is maintained at a positive pressure to prevent assay mixing. This pressure is monitored by a pressure indicating switch set to alarm in the event of a decrease in pressure between the block valves.

Cylinder evacuation, or " heeling," is accomplished by connecting the nearly empty cylinder to the throat of a venturi jet station and routing the UFe from a full cylinder through the venturi, which provides a suction on the nearly empty cylinder. This will remove the residual UFe from the cylinder. Heeling can also be accomplished by connecting the nearly empty cylinder to other low pressure sources (e.g., cascade surge drums, purge and evacuation pumps, etc.). When the cylinder is acceptably heeled, the steam supply and the cylinder valve are closed. The autoclave pressure is reduced to atmospheric by jetting the autoclave through a steam venturi, and the autoclave is opened. After pigtail purging and evacuation, the empty cylinder is removed from the autoclave with the bridge crane and transferred to the storage yard.

The overhead bridge cranes handle only cylinders that are empty or contain solid UFe.

Each crane hoist has a shoe brake that is spring actuated in the event of a power loss. A geared up/down limit switch is connected to the cable drum to prevent exceeding limits for lowering or hoisting the load. When activated, it will stop the motor and activate the shoe brake. Each of these cranes uses an H-frame-type sling to lif,t the cylinders with its single hook.

UFe detection heads are installed above the autoclave head ring, the heated housing at the autoclave head, above the jet station piping, in the piping trench, and along the west wall in C-337-A. These systems consist of one detector each and will detect leakage from the autoclave seal, the heated housing piping, the jet station piping, or the piping trench. If a i leak is detected, an alarm is sounded locally at each autoclave and on the UF, detector

! alarm panel.

Autoclaves are connected to the steam supply, air supply, UFe jet exhaust system, and cascade feed station by piping which is enclosed in heated housings or steam-traced to prevent freeze-out of UFe. Feed lines are equipped with 100 psig rupture discs which l relieve to surge drums to prevent cylinder over pressure.

44 l l

1

l 1e

! All autoclave piping to the outside containment valve is installed according to national ,

( standards and engineering piping specifications with a pressure rating of 200 psig  ;

I minimum. Piping downstream of the containment valves is installed with pressure rating equivalent to its intended service.

UF, valves installed in the feed facilities are designed and tested to engineering specifications and have an externally buffered bellows stem seal backed up by a packing  ;

gland secondary stem seal. Pneumatically operated valves have air to open-spring to l close-type. actuators. The actuators on containment isolation valves are required to always "

fail to the safe position. Actuators are equipped with limit switches that indicate the open or closed position of the valve on a local panel. Valve bodies are of steel or monel construction, corresponding to pipe size material specifications.

Isolation valves are suitable for UF, gas or liquid service and a temperature range from 70oF to 300 F. Pressure ratings are 200 psig for all valves that are used as primary containment valves. Valves located downstream from these valves are installed with a l pressure rating sufficient for intended service. t l

l The autoclave manualisolation system contains three manual push buttons, one at the Operations Monitoring Room door, one at the crane bay exit near the local cylinder yard, ,

and one in the ACR. When one of these buttons is pressed, each facility autoclave is placed into containment. The buttons are used upon confirmed UF, outleakage to mitigate l

the release. This item is addressed in the Compliance Plan.

l l There are several TSRs associated with these operations. These are briefly described in i Chapter 6 of this document.

1 The O systems associated with the sampling and transfer operations include: autoclave j water inventory control system, autoclave high pressure isolation system, autoclave l pressure relief system, autoclave steam pressure control system, UF, release detection system in the laboratory area (zone 1), UF, release detection system in the heated i

housings (zones 5-8), UF, release detection system located in the transfer room ( portions l of zone 4), scale cart movement prevention system, criticality accident alarm system, l overhead bridge crane, cylinder lifting fixtures, scale carts, hydraulic lifts, UFe cylinders and pigtails, and the liquid UF transfer piping and valves. The AO systems associated with sampling and transfer include: autoclave opening prevention system, low cylinder pressure system, high cylinder pressure system, cylinder pressure relief system, UF, release detection system at the autoclave head, cylinder interlock system, and sample cylinders.

The O systems associated with the feed facilities include: autoclave water inventory control system, autoclave high pressure isolation system, autoclave pressure relief system, ,

j autoclave steam pressure control system, UFe release detection system, criticality accident l alarm system, overhead bridge cranes, UF, cylinder lifting fixtures, UFe cylinders, and the l l UF pigtails. The AO systems associated with the feed facilities include: the autoclave i opening prevention system, low cylinder pressure system, high cylinder pressure system, cylinder pressure relief system, and the UF, release detection system at the autoclave head.

45 l

i

. - - - - - - ~ - _ - -- - -. .. - - .- - ..

l 4.2 UF, Enrichment The PGDP enrichment facility consists of about 1800 operating stages arranged in two parallel cascades. The cascade buildings are C-331 (400 stages), C-333 (480 stages),

C-335 (400 stages), C-337 (480 stages), and C-310 (60 stages). -The degree of isotopic separation is only about 0.2 % per stage, therefore, 500 to 700 stages are required between the feed point and product withdrawal point to enrich uranium from normal feed to a product ranging from 0.95 to 2.75 w/o 23sU. These stages are called the enricher. It takes 700.to 1100 stages to strip the 23sU isotope from normal feed to a tails withdrawal assay of 0.2 to 0.3 w/o 23sU; these stages are called the stripper.

A unit consists of ten cells. There are four units in C-331 and C-335 and six units in C-333 and C-337. A cell contains eight stages in C-333 and C-337 and ten stages in C-331 and C-335. Each stage contains a motor, compressor, converter, control valve, coolant system, and associated instrumentation. . Each stage is connected to the next l upper and lower stage throughout the cascade. The piping to a unit is arranged so that l any unit or cell can be bypassed, without shutting down any other cells.

The stage converter is designed to support the barrier materialin contact with the process l gas stream, separate the enriched fraction of the process gas stream from the portion depleted in the lighter isotope, and remove the heat of compression of the process gas in the converter by the use of cooling coils. The process gas enters the barrier region of the converter via a common inlet nozzle, where a portion of the process gas diffuses through the barrier tube wall and passes through the process gas cooler to the next compressor.

The stream that diffuses through the barrier tube wall and is slightly enriched in the 23sU isotope is called the A-stream. The gas that does not diffuse through the barrier tube wall l sleeves is slightly depleted in the 23sU isotope (the B-stream). The A-stream is routed to l the next higher stage compressor in the cascade, while the B-stream is routed to next lower stage compressor.

In order to efficiently operate the cascade, the stages near the feed point must have higher  ;

total flow rates than those near the withdrawal points. The flow taper required for l efficient use of power is achieved by sequencing different equipment sizes and by varying  ;

the pressures across each size equipment. At the PGDP facility, 3 sizes of equipment is used in the cascade. Buildings C-331 and C-335 contain "00" equipment, C-333 and C-337 contain "000" equipment, and C-310 contains 2X-sized equipment.

The largest equipment in use in the PGDP cascade is designated as 000-sized equipment.

The first size reduction in the cascade sends the process gas into 00-sized equipment, and a final reduction near the top of the cascade takes the process gas into 2X-sized converters. Both "000" and "00" stages use axial flow compressors to move the process gas through the system, while the smallest equipment in the purge cascade uses centrifugal compressors.

46

1 i

The enrichment cascade is arranged in two parallel cascades. The two cascades are connected at tie-in points. Tie lines between buildings are used to adjust the length of the cascade as required, and provide a means to move process gas from building to building without the need for withdrawal and feed facilities in each building. The tie lines are housed in elevated heated housings.

At those points in the cascade where there is a change in equipment size and a transition in pressure, the gas pressure can be increased b/ passing it through a booster station.

These stations are also used at the product and tails withdrawal points and at the junction with the purge cascade. The compressors are either axial or centrifugal, depending on the flow rate and pressure ratio required. In general, axial compressors are used to boost the l "A" (enriched) stream, while centrifugal compressors are used to boost the "B" (depleted) l stream. An axial compressor is used to boost the bottom overlap to the upper cascade, while centrifugal compressors are used on the top overlap.

t Surge drums and freezer / sublimer units provide surge volume for fluctuations in inventory

( and power. Freezer / sublimer (F/S) units are located in certain cells in the process buildings and can be operated independently or in conjunction to remove excess UFe inventory, to conduct rapid power drops, to stop UFe outleakage, or to adjust the power load.

The F/S is cooled with R-114 creating a low pressure vessel into which UF, flows from the cascade B-line and solidifies on the heat exchanger tubes. A control valve on the inlet line l is used to control the rate at which UF,is added to the vessel commensurate with the rate l at which cascade power load is being decreased. The freeze rate is normally controlled

! between 60 and 100 lbs/ min.

l To remove UF, from the F/S and return it to the cascade, the freezing process is reversed.

l The vessel is heated with R-114 and the UFe changes phase from solid to gas j (sublimation). F/S pressure is limited to less than the triple point by the cascade pressure to j preclude UF liquefaction during the sublimation process. There are a' total of 52 i freezer /sublimers yielding a total weight of UFe that can be stored in these I

freezer /sublimers equivalent to approximately 640 MW of cascade power load. The freezer / sublimer vessel and UFe piping are nickel-plated steel.

l A number of functional trips have been incorporated into the freezer / sublimer syst9=a. A l high UFe pressure alarm / trip is activated when the UFe pressure reaches 18 psia. 's ne

! R-114 and RCW temperature trips operate when either the R-114 or RCW temperature decreases to 82 F and places the freezer / sublimer in the hot standby mode and sounds an alarm. A differential pressure trip will place the freezer / sublimer system in the hot standby mode should the R-114 decrease to within 2.5 psi of the RCW pressure, indicating a tube rupture within the R-114 condenser /reboiler. The high weight trip places the freezer / sublimer in the cold standby mode when the unit reaches approximately 9,000lb of UFe/R-114 in the 10-MW vessel or approximately 18,000 lb in the 20-MW unit.  ;

Each cascade building also has surge drums, which are used for evacuation, purging and cell treatment activities. The surge drums are large cylindrical vessels made of SA-28SC l steel mounted with their long axis horizontally. They are normally operated so that their

pressure remains below atmospheric pressure. Each bank of drums is enclosed within an i 47 m

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insulated housing similar to the cell and bypass line housings. These housings are maintained at elevated temperatures to prevent condensation of the process gas.

Process piping, with its associated flanges, valves, and expansion joints moves the UFe  ;

from compressor to converter, cell to cell, and ouilding to building. Process piping varies in  !

size from 3 in. to 54 in. in diameter. UF, piping is nickel-plated internally to reduce corrosion. Process piping is heated either by routing the piping through heated enclosures i or by steam tracing and insulation. The cascade piping requires expansion joints to allow  ;

for some pipe misalignment and thermal expansion. The joints are of the monel  ;

bellows-type.

i The four main process buildings are grouped in two pairs, C-331 and C-333, and C-335 i and C-337. The purpose of the process buildings is to house the equipment and much of l the support systems necessary for the isotopic separation of uranium. All process buildings are connected by tie-line piping to form a cascade of the desired throughput i capacity.

l The ground floor houses the equipment required to provide auxiliary services to the cell floor process equipment. The cell floor (or second floor) houses the stages necessary for the isotopic separation of uranium.

l l Cells, booster stations, and UF, piping are enclosed in a housing of ribbed galvanized sheet I

metal and transite sheets attached to steel framing. The purpose of the housing is to retain heat so that the UF,is maintained in the gaseous state as required by the process.

The framing and housing enclosing cells and booster stations are independent of the building framing and are supported by the cell floor.

Each cell is served by a coolant condenser. A steel cylindrical lube oil storage tank is housed approximately at the center of each unit above the cell BP housing in the 00  ;

buildings and housed in rooms on the building roof in the 000 buildings. >

Each cascade unit is provided with a lubricating system to maintain a continuous supply of oil to the compressor and motor bearings. This same system supplies oil to operate the hydraulic stage control valves in the "000" buildings. The process recirculating oil system is designed to maintain a continuous and adequate supply of lubricating oil at a controlled temperature and free of abrasive matter to the bearings of axial compressors, centrifugal compressors, and electric motors of the cascade process equipment. The unit tube oil systems also serve the booster pumps and motors as well as hydraulically operated stage control valves in the OOO-size equipment.

Individual units within the cascade areas are equipped with completely independent recirculating lube oil systems, which include storage capacity, recirculating pumps, oil strainers, oil cooler, and a gravity supply tank. This equipment, coupled with the necessary supply, drain and vent piping, and the instrumentation for manual and automatic control, comprises the basic components of the system. .

The cell floor area is served by permanent overhead bridge cranes. These cranes are used to handle heavy process equipment during maintenance or modification.

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l Two types of UF, detection systems are used to monitor selected equipment in the cascade buildings: the ionization chamber type and the turbidity type. lonization type detectors are installed at the following locations: cell housings, bypass housings, "B" seals, UF, condensers, product withdrawal pumps, UF, drain stations, and autoclave areas where UF, could conceivably be released. The sensing element is an ionization chamber in which air is made conductive by the use of an alpha emitter. The sensitivity of the ionization chamber is controlled by adjusting the size of the opening over the alpha emitter.

l Turbidity type indicators use a photoelectric cell which actuates an alarm when UFe is I

detected and are used in the C-310 and C-315 ductwork. An air sample is continuously I moved through a sample chamber which becomes opaque due to the presence of UFe . An electronic detection unit senses this condition and alarms at a preset point. This system is used only to detect UFe in the ductwork from any source.

The purge cascade is located in building C-310. The functions of the purge cascade include isotopic enrichment of 23sU and separation of lighter molecular weight contaminant gases from the UF, for safe venting to the atmosphere. There are 10 cells with six stages each in C-310.

The top cells in the PGDP cascade are called the purge cascade. The purge cascade is located in the C-310 building with a portion of the product withdrawal equipment located in the C-310-A building. The purge cascade removes light molecular weight gases, allows the withdrawal of product from the cascade, and removes intermediate molecular weight gases from the cascade system. Below the product withdrawal point and a few cells above it, the purge cascade provides the same isotopic enrichment function as other enrichment '

cascade equipment. Above the product withdrawal point, the function is to separate light molecular gases from the residual UFe which remains above product withdrawal point and vent them via the tops purge stack to the atmosphere. Both functions are accomplished by using equipment similar to that in the enrichment cascade to effect separation using gaseous diffusion.

The ten cells in the purge cascade are normally operated as a separate cascade tied into the top of C-335. Nitrogen or lights separation from the residual UF,is accomplished primarily in the top four cells while the lower four cells are used primarily for isotopic separation. The remaining middle cells provide for a transition as the UF, concentration rapidly drops.

Lights containing only traces of UF, will normally be withdrawn from the cascade at the top cell in the purge cascade which is normally cell 2. From the top of cell 2 the lights pass to the tops booster station and then to the purge rate control system. A portion of the lights is recycled to maintain the lights front while the remainder is passed through one of two parallel banks of Alumina traps then on to the jet system and exhausted to the atmosphere by venting through a stack outside the C-310 building. The vent stack contains various sampling systems to track stack emissions of fluorides and selected radionuclides.

Control facilities at PGDP consist of local control panels for each cell in the enrichment i cascade, an area control room (ACR) for each process building, and a plant CCF located in l 49

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the C-300 Central Control Building. The ACR instrumentation permits control room operators to monitor the operation of process equipment with the ability to control and/or shut down some equipment. The CCF instrumentation monitors the process buildings' status and also permits operators to control the operation of some equipment.

The CCF's function is to monitor, coordinate, and/or control critical plant process, power distribution, utilities, communications, plant alarm systems, and emergency operations. All CAAS actuations and all UF, release detection systems actuations alarm in the CCF. The Cascade coordinator directs adjustments in er.ch building in order to maintain the electrical power load.

Each process building has an ACR located approximately at the building's center on the ground floor. The purpose of each ACR is to permit operators to monitor process equipment, make changes in operation, and take corrective action to mitigate abnormal operating conditions. Most operating cells can be shut down and isolated from the ACR.  ;

The C-310 cells, however, must be shut down at the local panels. j l

Each cell in the enrichment cascade has a local operating area on the ground floor l consisting of a cell operating panel, motor breaker panel, valve control center, buffer control panel, and a freezer / sublimer panel for those cells containing freezer /sublimers. The cell panels are essentially the same in all buildings except for arrangement of instrumentation on the panels. )

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There are several TSRs related to the enrichment operations. These are briefly discussed in l

Chapter 6 of this document. l l

The Q systems associated with the enrichment operations include: the UF release l detection systems (cell housings, cell exhaust ducts, bypass housings, and over B-boosters), R 114 coolant over-pressure control system, freezer / sublimer high-high UF, l

weight trip system, intermediate gas removal high temperature control system (although on I the Q list, this piece of equipment is nor used), and the criticality accident alarm system.

The AQ systems associated with the enrichment operations include: the process gas l coolers, low pressure datum system, high pressure datum system, surge drum pressure / room temperature instrumentation, space recorders, cascade piping and equipment, and cell remote manual shutdown system.

4.3 UF, Product Withdrawal l The product withdrawal systems for PGDP are housed in C-310 and C-310-A. There are l two complete withdrawal systems that permit simultaneous withdrawal of two product streams with different 23sU concentrations. One system is identified as the top product withdrawal system, while the other system is designated the side product withdrawal system. Either system can be used to withdraw material of any assay up to the plant limit of 2.75 w/o 23sU. When the C-315 Tails Withdrawal Facility is unavailable, tails withdrawal may be performed at C-310. Each facility has local operating controls and the ACR in 50

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C-310 provides additional instrumentation. Control of the assay level is accomplished by monitoring of the withdrawal stream with mass spectrometers and/or by analyzing samples j collected during the filling process.

l UF, from the diffusion cascade is compressed using Nometex pumps to a pressure of approximately 30 psia and then cooled to about 160 F to condense the gas. The liquefied i UF, flows by gravity into 21/2 or 10-ton cylinders. The liquefaction and accumulation is conducted in C-310-A. Each cylinder being filled is mounted on a scale that monitors the l cylinder weight. When the predetermined cylinder weight limit is almost reached, an l l audible alarm on the scale is sounded to alert the operator, and the valve in the UF, drain l line to the cylinder automatically closes. The filled cylinder is disconnected and moved l l outside the building for cooling and solidification of the UF, product. Noncondensible l l contaminant gases remaining in the cylinder are removed by connecting the cylinder to the '

" burp" station at C-310 and evacuating it. Gases evacuated from the cylinder are passed  !

through sodium fluoride traps to remove any UF, before discharging the remaining gases to I the atmosphere through the C-310 stack.

The Normetex pump uses a fixed and a moving spiral vane. The moving vane has an eccentric motion causing " pockets" between the vanes to open and close to compress the gas and move it from the inlet to the discharge of the pump. This design permits compression of UFe without the use of dynamic seats between the process and atmosphere.

A spacer column connects the fixed vane to the pump body, and the moving vane is attached to the base of the spacer column by two metal bellows that separate the UF.  ;

from atmosphere and serve as an inlet to the pump. The space between the two bellows '

is buffered and monitored. An inlet control valve regulates the amount of flow through the pump by controlling the suction pressure. The pump discharge is through a check valve at the pump center. The flow is routed to one of three UF condensers.

UF, enters the tube side of the condenser at approximately 30 psia, where it is cooled and condensed to a liquid by R-114 coolant passing through the shell side. The R-114 is cooled with RCW in a coolant condenser similar to the process building coolant condensers. A portion of the UFe flow and that part of the noncondensible gases not entrapped in the liquid UF, are returned to the cascade via the condenser vent. These gases are normally rourned near the stage of the cascade from which they are withdrawn.

A vent control valve regulates the condensing pressure and the flow of gases back to the cascade. The condenser pressure is controlled and measured in the C-310 control room.

Two UFe liquid accumulators serve the withdrawal system. The product accumulator is a 21,000lb capacity nickel-lined tank used in the top product system. The side accumulator is monet-lined steel with a 4,300-lb capacity. The accumulators provide surge volume by

" floating" on the drain line. A vent line with a control valve is provided to permit the return of noncondensibles to the cascade and to control pressure, f There are two cylinder filling stations or withdrawal positions in norrrial use in C-310. Each has a cylinder cradle arrangement mounted on a cart, which is moved on a floor track l

j system. A cylinder to be filled is placed in the cradle and the cart is moved into position 51

l on a scale at the filling station. A removable pipe or " pigtail" connects the filling station to the cylinder valve. Withdrawal pigtails are wrapped with electrical heat tracing covered I with insulating tape to reduce the likelihood of freeze-out. Each cylinder filling station has l an exhaust hood connected to a common exhaust duct, HEPA filter, and fan, which is then exhausted to atmosphere. This prevents the accumulation of any residual gases that might arise from withdrawal operations.

While the liquid UF,is being drained from a condenser or accumulator into the cylinder, the weight of the material withdrawn can be read from the scale to determine when the weight is approaching the cylinder target weight. The cylinder target weight is set at or below the cylinder fill limit. When this target weight has almost been reached, an alarm on the scale sounds to alert the operator and a valve in the UF, drain line automatically closes to prevent overfilling of the cylinder. Additional filling of the cylinder to the target weight is then performed using a local manual override. If the target weight or the cylinder filllimit l should be exceeded, the excess material may be removed by valving the cylinder to the evacuation header at the withdrawal position. Cylinder filllimits are based on a maximum UFe temperature of 250oF, a certified minimum internal volume for each cylinder, and a 5% safety factor.

After being filled, cylinders are moved to the cooldown yard, where they remeM for a minimum of five days for 10- and 14-ton cylinders and for a minimum of three days for 21/2-ton cylinders. Cylinder saddles are used to position cylinders for storage.

Enriched product material is, withdrawn into 10-ton cylinders at C-310, which after cool-down are moved to the C-400 building where they are loaded into overpacks installed '

on railcars or flatbed truck trailers. The cylinders are then shipped to the Portsmouth facility. The emptied Paducah product cylinders, with a small amount of heel material, are returned to PGDP. The cylinders are received at C-400 where they are unloaded and moved back to the C-310 product withdrawal facility for refilling with Paducah product.

The product withdrawal area has one 20-ton overhead bridge crane that is used to move liquid-filled cylinders from the scale cart to the storage area. The crane is a double-block )

crane equipped with a special lifting beam specifically designed for lifting liquid UF, 1 cylinders. The crane hoist has two d.c. rectified shoe brakes. One is used as a holding I brake and the other is on a timer that acts within one second as an emergency brake l should the holding brake fail. These brakes are spring actuated in the event of a power l loss. A geared up/down iimit switch is connected to the cable drum for use while lowering or hoisting the load. When activated, it will stop the motor and activate the shoe brakes.

This switch will reset automatically once the motor is reversed. Two paddle-type limit switches prevent a collision between the lifting beam and the upper crane structure. Each has a weight that hangs on a wire from the crane trolley. If the lifting beam comes in contact with one of these weights, the tension in the wire is released and the crane hoist motor is de-energized. Once activated, the paddle-type limit switches require a manual reset to resume crane operation.

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[ To prevent the crane from carrying a load into the piping associated with the C-310-A burp station, a zone control has been installed. Also, proximity sensors and mechanical rail stops are located at the end of the bridge and trolley rails to prevent the crane from running off the end of its tracks.

Cylinder movement from the scale cart to the cylinder cool-down area is minimized and i restricted to the lowest practicable height above ground.

After the liquid UFe product has solidified and cooled to ambient temperature the cylinder is moved to the burp station. The cylinder is then connected via a pigtail to a manifold which contains two exhaust lines. One is for cylinder pigtail purge gases, which are exhausted to i

the cascade, and the other is connected to a trapping system. Any R-114 or other noncondensible gases that might remain in the cooled cylinder are removed by pulling a vacuum on the cylinder with an air jet through the trapping system. The pressure in the cylinder is reduced to 8 psia or less to remove these noncondensible gases. A cooling water spray may be applied to the cylinder skin. This aids in cooling the cylinder during the burping operation which minimizes the amount of UF, removed. The gases evacuated from the cylinder are passed through a sodium fluoride (NaF) trap to absorb UFe that might  ;

be pulled from the cylinder. The exhaust from these traps is then released to the atmosphere.

l Activation of the UFe release detector head above any drain position pigtail will cause the dual drain line block valves to close, the cylinder valve closer to close the UFe cylinder valve, and the evacuation valves to open to evacuate the manifold and pigtail. Activation of any of the ceiling detector heads will result in closure of the redundant drain line block valves at all of the drain stations to isolate the drain header.

The C-360 Toll Transfer and Sampling Facility Movides the systems for transferring the  !

product into customer cylinders and for shipping the cylinder. For transfer of material, the ouent cylinder is heated.

The cylinder is positioned in one of four 96-in diameter autoclaves in C-360 with the cylinder valve at the 12 o' clock position. The cylinder is connected to a manifold by means l of copper tubing pigtail. After pressure testing all connections, the UFe cylinder valve is l opened, the emergency cylinder valve closer is attached, the autoclave shell is closed, and l a hydraulically operated rotating ring seals the movable shell of the autoclave to the fixed head.

Steam is admitted into the autoclave and a vent bleeds atmospheric air from the shell.

Autoclave temperature is controlled by utilizing a cascade control scheme within a feedback loop controller. The steam pressure is limited to a maximum of 8 psig which corresponds to a saturated steam temperature of 2350F. During initial heating, the cylinder is valved to a rupture disc on the manifold header to mitigate the possibility of cylinder rupture. The rupture disc will relieve cylinder contents to surge drums at 100 psig (the lowest MAWP of cylinders heated). A pressure transmitter monitors the cylinder pressure and closes the steam supply valves if the cylinder pressure exceeds 90 psia.

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After heating, the steam is shut off, the cylinder valve is closed, and steam is evacuated '

from the autoclave. The autoclave is then opened, the pigtail is disconnected, and the cylinder is rotated to position the valve in the 6 o' clock position. The transfer pigtail is j connected to the cylinder and the transfer line which leads to a drain manifold located on  :

the lower level. An empty receiving cylinder is attached to the drain manifold by another pigtail. The cold pressure is verified to be less than 5 psia prior to filling. The empty  ;

cylinder is cradled on a cart that rests on a scale at the drain station. 1 After the parent cylinder valve is reopened, the autoclave is closed and steam flow is ,

resumed. A flow of liquid UFefrom the parent cylinder to the receiving cylinder is ]

established by opening the drain manifold valve and the receiving cylmder valve. Transfer 1 can be made from only one autoclave at a time. During liquid transfer from an autoclave, l the drain line block valves for the remaining autoclaves are closed.

The manifold valve at the drain station is closed automatically at a predetermined setting on the scale. The scale setting is always below the final cylinder filllimit to prevent I overfilling. An operator can then change the scale setting or use the jog button to fill the cylinder to the proper weight. Cylinders inadvertently filled beyond the desired weight are  ;

evacuated to the building evacuation drums until the desired weight is reached.

Cylinders are normally filled at a temperature around 180oF while reheating is limited to a temperature of 235 F. After block valves are closed, the pigtail is purged and ,

disconnected. The receiving cylinder is moved to the storage area and allowed to cool a 1 minimum of three days for a 21/2-ton cylinder or five days for a 10-ton cylinder to ensure I solidification of the UF, prior to shipment to the customer. .

The TSRs related to the product withdrawal operations are discussed in Chapter 6.

The O systems associated with product withdrawalinclude: the UFe release detection  ;

systems (at the Normetex pump, the withdrawal station low voltage systems, and condenser and accumulator heated housings), scale cart movement prevention system, criticality accident alarm system, normetex pump high discharge pressure system,  !

overhead bridge crane, UFe cylinder lifting fixtures, scale carts, the UFe cylinders and pigtails, and the condenser, accumulator and liquid UF, process piping and valves.

4.4 UFe Tails Withdrawal l l

The process for tails withdrawalis similar to that for product withdrawal. The tails <

withdrawal facility is located in Building C-315. The two primary purposes of tails withdrawal facility are to provide a cascade surge volume and to compress and condense the UFe to permit withdrawal. The surge volume is provided by two 28,000lb Hortonspheres and is used to control cascade inventory fluctuations.

These two functions are accomplished by routing the B-stream from the bottom of the cascade in the C-331 building to a Hortonsphere as well as using the suction of the Normetex pumps or high-speed compressors. One or more of the three Normetex pumps operate in parallel and discharge to the UFe condensing system. Two high speed compressors in C-315 that are maintained in standby can be used in lieu of the Normetex 54

Q pumps. The flow through the Normetex pumps or high speed compressors controls the pressure in the on-stream Hortonsphere. The process gas is retumed from the on-stream Hortonsphere via a control valve to the bottom of the cascade in C-331 building as the A-stream. Since the on-stream Hortonsphere floats on the discharge of the low-speed compressors, it provides a surge volume for changes in tails withdrawal as well as flows to and from the cascade.

The UF, from either the Normetex pumps or high-speed compressors is piped at approximately 30 psia to a condenser, where the gas is condensed by cooling to a temperature of about 160oF. Three UF, condensers are used in the tails withdrawal facility. Two 10-ton nickel-lined steel accumulators are located down stream from the condensers in the tails withdrawal system. Each accumulator can be used for short-term storage of the liquefied tails material while a cylinder is valved off or being changed.

Normally, only a small amount of UF, is maintained in the accumulators, which float on the line ready for immediate use.

The liquid UF, flows by gravity into a 10- or 14-ton tails storage cylinder at one of the four cylinder filling positions. During filling, the cylinders rest on cradles on rail-mounted carts positioned on scales at each station. The scales provide a weight readout and an adjustable audible alarm to alert the operator when the fill limit is approached. A valve in the UF, drain line then automatically closes to prevent overfilling of the cylinder. Before moving a cylinder from the filling station, an accountability weight is established so that, in the event of an overfill, evacuation of the excess UF can be performed with the cylinder in the drain position. After a cylinder has been filled with tails material,it is carefully transported outside by the use of the air-operated scale cart. It is then lifted by a double-block 20-ton crane and carefully transported to the temporary storage area for tails cylinders where it remains until its contents have cooled and solidified. The cylinder is then moved to a long-term storage yard.

The tails withdrawal TSRs are discussed in Chapter 6.

The O systems associated with tails withdrawalinclude: the UFe release detection systems (at the Normetex pump, the withdrawal station low voltage systems, high-speed centrifugal pumps, and condenser and accumulator heated housings), scale cart movement prevention system, criticality accident alarm system, normetex pump high discharge pressure system, overhead bridge crane, UFe cylinder lifting fixtures, scale carts, the UFe cylinders and pigtails, and the condenser, accumulator and liquid UF, process piping and valves.

4.5 UFe Cylinder Storage Operation of PGDP requires the storage of large inventories of UFe in cylinders. This occurs in the form of interim storage facilities for feed and product stockpile and long-term storage for the cylinders containing depleted UFe tails.

! Fourteen active UFe cylinder storage yards are provided at the site for interim and long-term storage. These storage yards, designated as C-745-A through Q, provide i storage for nearly 52,000 cylinders. However, only yards A, B, E, H, J, and Q are leased 55 l

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l by USEC. Four more cylinder storage yards are being constructed, C-745-R through U, l with a capacity for an additional 15,000 cylinders. In addition to these yards, space is l

provided at C-310, C-315, C-333-A, C-337-A, and C-360 to serve as a staging srea for cylinders scheduled for feeding, for empty feed and withdrawal cylinders,and as a cooldown area for cylinders which have been filled with tails or product. There is also a cylinder staging area at C-400 for shipping and receiving product cylinders. Tails cylinders that are in long-term storage are double stacked. Aisles are provided for access to visually inspect the cylinders and valves.

Because of their weight, the cylinders require the use of large specialized handling equipment and a firm foundation for both transport and storage. Design considaations for l new and recently developed cylinder storage areas include 3-in. of dense graded aggregate covered by 5-in. of compacted limestone covering. The yard lighting is designed to provide an illumination level of 0.5 footcandle. Storm drainage is accomplished by a sub-surface drainage system and by sloping the storage pad to existing drainage to minimize corrosive exposure of cylinders.

The cylinders are placed on supports to prevent them from rolling and allow them to be l stacked. The saddles also help in corrosion prevention by keeping the cylinders off the ,

ground and as dry as possible. Steel reinforced concrete cylinder saddles are used in l newer cylinder yards for cylinder support.

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4.6 Chemical Facilities The C-400 Chemical Operations facility and C-409 houses decontamination, uranium l

recovery, cylinder cleaning and testing, and pipe cleaning equipment. Decontamination I facilities include Branson and Lewis degreasers, eight cleaning and rinsing tanks, alkali tank, acid tanks, spray booth, and dry honers. C-400 also contains the drum crusher, receiving booth, and hand tables. Uranium recovery operations consist of stainless steel tanks that receive all uranium bearing solutions and uranium precipitation system. Other facilities, including processes for F2 , Cl2, CIF3 , H 2SO4, and HNO3, are located within the fenced portion of the site.

The only O system associated with the chemical f acilities is the criticality accident alarm system. The AO systems are the chlorine system, CLF3 system, and the F2 system.

4.7 Laboratory l

l The laboratory is housed in C-710, with one iaboratory housed in C-409. The laboratory supplies analytical services to all functions in the plant. Analytical services and include sample analysis for process control, environmental monitoring, material specification verification, personnel health and safety monitoring, and troubleshooting.

l The only O system for the lab is the criticality alarm system. There are no AQ systems.

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4.8 Utilities The PGDP requires a number of auxiliary systems that are essential for plant operations.

The major utilities include electric, water, steam, compressed air, and nitrogen.

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Dry. compressed air is one of the utilities required at PGDP for plant operations. Pneumatic instruments, controllers, and valves are used for control of the many plant processes. Air is also used in the enrichment process to operate purge and evacuation systems. In addition, air is used.for plant support production equipment operations, steam plant operations, laboratory facilities, test facilities, maintenance functions, process seals, and other miscellaneous uses. It is supplied to PGDP from the air stations through a distribution system.

Nitrogen is utilized in the process buildings for purging process equipment and for process seat feed. While nitrogen is distributed throughout the plant at 50 psig, the pressure within the process buildings is reduced to correspond to the specific usage. The loss of nitrogen flow to the process buildings would result in an automatic transfer of instrument air to the building nitrogen header. This transfer could be accomplished without a interruption of gas flow to the compressor shaft seals. Therefore, normal operation of the seal system would continue with no expected loss of material containment.

Steam is primarily utilized in the process buildings for heating pipe enclosures to maintain the desired operating temperatures to prevent freeze-out of UFe. Remote bulb thermostats transmit impulses to air relays which regulate the air loading on the steam control valves to achieve and maintain the set temperature. The condensate system functions as a method of removing the water formed by the cooling of the steam via steam traps. Steam traps are used to retain steam in the piping system until it has condensed and given up its latent heat. At that point, the condensate and air are discharged by the trap.

During the compression of UFe, a large amount of the shaft horsepower is converted to thermal energy. The process coolant system removes the excess heat to maintain a desirable UF operating temperature. This requires the coolant system to remove approximately 95% of the energy supplied to the process system.

Each converter has a gas cooler installed in its shell. The UF stream which has diffused through the barrier tubes of the converter then passes through the fins on the finned tubes of the heat exchanger. This facilitates the transfer of heat from the process gas through the fins and tube walls into the R-114 coolant flowing inside the cooler tubes. The j vaporized R-114 progresses to a water-cooled condenser, transferring its heat and l changing back to the liquid phase. The condenser is at a higher elevation than the cooler which makes it possible to maintain a liquid head in the condenser outlet line sufficient to promote gravitational flow around the R-114 loop.

The R-114 condensers are cooled by recirculating cooling water (RCW) which flows to one of several large conventional-type cooling towers. It is cooled by releasing the UF, heat of compression to the atmosphere. The cooled water collects in a basin'at the base of the  ;

cooling tower. Large pumps in a pump house next to the cooling tower serve to pump the  ;

RCW through its loop to the process building condensers and back to the cooling towers.

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Each "000" cell has two coolant systems, while "00" cells have one coolant system serving all 10 stages. Each of these systems contains about 20,000 to 25,000lb of coolant. The coolant system pressure is maintained higher than the pressure of both the UFe system and the RCW system. Therefore,in case of leakage, the coolant leaks into the UF, or the RCW, thus preventing water from leaking into the coolant system and then potentially into the process gas.

The process system electrical equipment consists of three major elements: transformers, switchgear., and static capacitors. The transformer transforms the voltage to the proper value for use by the process motors. Switchgear distributes the power and provides on-off switching to the motors. Power factor correction is provided by the static capacitors.

Auxiliary and backup power systems provide continuing reliability for the caccade.

The staff is awere that the descriptions contained in Chapter 3 of the SAR may not match the "as found" condition of the facility in all instances; this aspect is being remedied as part of the SAR upgrade effort. The regulations also provide a mechanism for dealing with "as found" conditions. Chapter 3 of the SAR does provide an adequate description of the facility and process. Therefore, the staff concludes that the facility and process descriptions contained in Chapter 3 of the SAR meet the requirements of the regulations, and are, therefore, acceptable.

There are three Compliance Plan Issues that are associated with the autoclave.c, cascade cell trip, and UFe leak detector sensitivity testing; these are discussed below.

Issue 3 concerns upgrades to the autoclaves. These upgrades include:

(1) Installation of manual push buttons for the feed facility autoclave containment system.

These will be used to place the autoclaves in containment upon confirmation of a UFe release. The three manual push buttons will be installed by September 30,1996. The associated TSR will be submitted by October 31,1996; (2) The pressure monitoring instrumentation serving the autoclave safety system for C-333-A, C-337-A, and C-360 autoclaves will be replaced by October 31,1997. The existing instrumentation will be used until replaced; (3) The UFe detection system for zones 1 and 4 in C-360 will be modified so that upon detection of a release, multiple valves will be closed on the transfer and/or sampling piping.

Currently only one containment isolation valve is closed. Until modified, sampling will not be allowed during transfer operations. This modification will be complete by August 31, 1997; (4) Performance criteria and test plan for determining plant air isolation check valve operability on loss of plant air and for allowing the bottled nitrogen system to operate the emergency valve closings for C-310, C-315, and C-360 have not been developed. This will be complete by December 31,1996; 58

[ . 1 (5) Installation of low instrument air pressure switch to initiate containment upon loss of instrument air in C-360. The switch will be installed by August 31,1997; (6) Modification of the controls for the autoclaves in C-360 to prevent them from being inadvertently opened when a autoclave high pressure isolation system containment signal l is present. This is being administratively controlled until the modification is complete  !

(October 31,1997), i l

l (7) Obtaining a code interpretation from the ASME code committee concerning the C-360 parent daughter transfer system and the need for pressure relief on the transfer manifold. l The ASME code interpretation will be submitted to the NRC by December 31,1996; and (8) Provide the capability to perform a pressure decay test separately for the inner and outer containment valves for all autoclave and provide assurance that backpressure does not mask leaks during the test. This will be complete by December 31,1996; additionally a new TSR to reflect the configuration will be submitted to the NRC by December 31, 1996. The upgrades listed in issue 3, the schedules for the upgrades, and the justifications for continued operation are acceptable to the staff.

Issue 48 requires PGDP to reconfirm specific vaiues for battery performance and air pressure requirements for air circuit breakers necessary to ensure operability of the l cascade cell trip function. The values will be verified by independent review by August 30, 1996. If new limiting values are determined a new TSR, TSR basis statement and/or revised SAR section will be submitted within 30 days of verification.

Issue 49 deals with the sensitivity of the UF leak detector testing. USEC has not .

developed a testing method that establishes the correlation between the detectability of  !

test smoke and the detectability of UF and its reaction products. Plant experience does -

indicate that the detectors do have capability to detect UFe . All detectors will be field tested by July 31,1998.

The plans of actions and schedules and the justifications for continued operations for these issues are acceptable to the staff.

59

Chapter 5 ACCIDENT ANALYSIS 4 The regulations in 10 CFR 76.35(a)(4) require the SAR to include an " assessment of l accidents based on the requirements of 676.85." A " reasonable spectrum of postulated accidents which include internal and external events and natural phenomena" is to be )

I considered in the accident analyses. Chapter 4 of the SAR contains the PGDP accident analysis. By TSR 3.20, USEC is required to make changes to the accident analysis in accordance with the plant design change control process described in SAR 66.3. The SAR  !

is based, in part, on the 1985 Final Safety Analysis Report (FSAR) and approved safety l evaluations performed by the plant during the intervening time period. The SAR does not fully incorporate all of the information that is being generated by the current DOE site-wide safety analysis upgrade effort. A number of areas and supporting safety analyses are expected to be updated concerning the descriptions of hazards, plant SSCs (structures, systems, and components) and human activities relied on for safety. The upgrade is scheduled for completion by DOE in Februay 1997 in accordance with Compliance Plan issue 2.

The potential consequences of postulated accidents include personal injury, health effects from acute exposure to toxic chemicals, non-stochastic effects from acute radiation exposure, and risk of latent cancer because of exposure to radioactive material. The purpose of accident analysis is to (1) investigate the nature and consequences of accidents in order to determine the impact of the facility on the health and safety of the workers and the general public, and (2) identify limiting conditions for operation (LCO) to assure the safe storage. handling and processing of radioactive and other hazardous materials at PGDP.

5.1 Accident Evaluation Methodology The staff review included the accident scenario initiators, related release mechanisms, and the potential consequences of each scenario described in Chapter 4 of the SAR with regard  ;

to the hazards associated with (1) fire, (2) toxic / radioactive material release, (3) explosion, (4) radiation exposure, and (5) nuclear criticality. The staff considered the facility design and the description of equipment and operations as presented in the SAR and in USEC's responses to staff questions. In addition, the operating history of the three GDPs was also reviewed. As a result, it was determined that the major hazard at a GDP was an accidental release of a large amount of UF or an accidental criticality. Section 1.4 of this CER summarizes the past UF, release accidents at the GDPs. Although there have been no criticality events at any of the three GDPs, criticality safety was evcluated.

5.1.1 UFe Releases Uranium hexafluoride chemically reacts with the moisture in the air to form uranyl fluoride (UO,F2 ) and hydrogen fluoride (HF). The acute toxic chemical effects of exposure to l uranium (U) and HF have previously been evaluated by the NRC in NUREG-1391 (NRC, 1991) and by Pacific Northwest Laboratory for the NRC in PNL-10065 (PNL,1994). The NRC analysis concluded that the chemical effects of exposure to uranium in soluble form exceeded the acute radiological effects and that the threshold for renal injury (nephrotoxicity) appears to lie very near 3 micrograms (#g) U per gram kidney tissue which 60.

f o results from a single inhalation intake, by a standard person, of about 30 milligrams (mg) U. NRC staff evaluation of the data provided in the Portsmouth application SAR, Chapter 4, Appendix A, Figure A.1-1, indicates that the thresholds for renalinjury are airborne concentrations of 15 mg-U/m2 for exposure times of one hour or more and 25 mg- i U/m2 for exposure times of 30 minutes or less.

Regarding the radiological hazards of soluble uranium, the equivalent radiological doses for a 30 milligram intake via the inhalation pathway are about 50 millirem for natural uranium, 110 millirem for 2.75 percent enriched and 350 millirem for 10 percent enriched uranium.

For the ingestion pathway, the corresponding radiological doses are substantial!y lower.

Therefore, the primary concern for exposure to soluble uranium is from the chemical rather than radiological effects.

According to NUREG-1391 (NRC,1991), exposure to HF at a concentration of 25 mg/m for 30 minutes has been identified as the level with no significant effects, either short-term or long-term. However, significant irritation of lung tissue could occur at this concentration I (PNL,1994). The American Industrial Hygiene Association (AlHA,1988) has defined l Emergency Response Planning Guidelines (ERPG) as (a) ERPG-1 (4.1 mg-HF/m8 ) at which nearly allindividuals could be exposed for one hour without experiencing other than mild, l transient health effects or objectionable odor; (b) ERPG-2 (16.4 mg-HF/m3 ) at which nearly all individuals could be exposed for one hour without developing irreversible or other serious health effects which could impair an individuals ability to take protective action; and (c) ERPG-3 (41 mg-HF/m ) at which nearly all individuals could be exposed for one hour without developing life-threatening health effects.

1 The odor threshold for human beings to HF is 0.0333 to 0.1333 mg-HF/m3 . The l equivalent stoichiometric half-hour intake of soluble uranium based on the assumption that the UF -H2 O reaction goes to completion, would range from 0.06 to 0.24 mg. Similarly, I the stoichiometric amount of uranium in a UFe release resulting in an airborne concentration  !

of 50 mg-U/m 3(equivalent to a half hour intake of 30 mg-U) corresponds to an airborne HF  ;

concentration of 17 mg-HF/m*. It should be noted that for relatively long exposures (on the order of tens of minutes) resulting from UFe releases, health effects from uranium intakes would be limiting. However, for short exposures (a few minutes or less), the health effects from HF would be more limiting. For most of the large UF release scenarios, it is assumed that the exposure time is limited to 30-minutes or the release duration, whichever is shorter, for a member of the public located off-site. For workers located on-site, exposure times for most UFe release scenarios would likely be shorter than 30 minutes or the release duration, as a direct result of early detection (alarms, sight, smell, etc.), emergency response, and procedures implementing the "See-and-Flee" policy and the comprehensive Worker Protection TSR (TSR 3.23).

A description of the dispersion modeling used by USEC/ DOE in evaluating the impact of UF, releases at the PGDP is presented in SAR 64.7. Since there are no residences located within one kilometer of any of the postulated release points, a distance of 1 kilometer was conservatively assumed for reviewing the impact to an off-site resident. For large UF.

releases inside the process buildings, the dispersion modeling established UO2F2 and HF concentrations in the cell floor and operating floor of a process building and release rates I l

! from the building to the environment.

61

in November 1995 and March 1996, the staff obtained draft versions of HGSYSTEM/UFe ,

a computer atmospheric dispersion model, and used them to perform independent modeling calculations for severallarge UFe release scenarios, including large releases of liquid UFe (14 ton cylinder valve failure) and gaseous UF,(seismic event). These calculations were performed to determine the appropriateness of the Justification for Continued Operation present3d in the Compliance Plan's SAR upgrade issue, and to confirm the results of the dispersion analyses described in the SAR. Although the formats of the consequences differed (the SAR presented plume overlays of health effects whereas the staff's analyses generated.results in terms of uranium airborne concentrations), the results of the NRC staff's analysis compared reasonably well with the results contained in the SAR. The SAR upgrade willinclude dispersion analyses using a revised version of HGSYSTEM/UFe. More detailed confirmatory analyses will be subsequently performed by the NRC staff following submittal of the upgraded SAR.

5.1.2 Criticality All criticality pathways and scenarios cannot be identified with a high degree of certainty.

Therefore, the effects of a representative postulated criticality event were evaluated in the SAR to characterize the hazard associated with facility operations. The scenario is based on the accumulation of 5% enriched uranium solution in a 55-gallon waste drum. The consequence of this scenario is similar to, or bounds, all other criticality scenarios at PGDP. The analysis of the event used the KENO V.a and Scale Version 4.1 to determine the critical height and neutron multiplication factor, respectively. The SAR validation calculations were for systems with 5 w/o or less of 23sU. The staff independently validated some calculations for systems with 2.75 w/o or less of 23sU.

SAR 94.4 describes several criticality accident scenarios for selected plant areas and includes an updated version of the accident analysis contained in the 1985 FSAR and analysis previously approved by DOE for the " limited scope" plant assay increase from 2.0 to 2.75 w/o U235 enrichment. Additional criticality analyses, described in Appendix A to Chapter 4 of the SAR (Higher Assay Upgrading Project (HAUP) Safety System Analysis KY-792, Revision 1.a), have been developed to support a plantwide assay increase.

The consequence analysis presented in Table 4.4-1 and Figure 4.4-1 of the SAR, indicate that the effects of an accidental criticality due to all credible events involving 5 w/o U 235 are limited to the site. No attempt was made to estimate the probabilities of accident sequences other than to determine that the criticality safety criteria should provide assurance that the probabilities of the identified sequences are very small.

Processes for which double contingency criteria cannot be met were identified and the proposed criticality safety controls were evaluated by the NRC staff for inclusion in the TSRs.

5.2 Potential Hazards of Credible Accident Scenarios

~

Staff review of the PGDP potential UFeaccident scenarios presented in Chapter 4 of the SAR included evaluation of equipment and procedures against potentia! failure modes. The 62

review identified and confirmed design features and operating changes which could detect and prevent accidents and response actions which could mitigate the consequences. As a result of this review, the adequacy of several proposed TSRs and application of OA to the safety related Structures, Systems, and Components (SSCs) was determined.

The description of process design, equipment, instruments and controls, and administrative controls and safety programs presented in the PGDP SAR provided a basis for reviewing the adequacy of potential accident scenarios. The scenarios presented in the PGDP SAR are for plant segments defined by process function. The plant segments are Feed, Enrichment Cascade, Product Withdrawal, Tails Withdrawal, UFe Sampling and Transfer, UF Cylinder Storage, Chemical Facilities, Waste Management and Laboratory Facilities.

5.2.1 UF, Releases The SAR identifies seven worst case bounding scenarios that were analyzed to develop a representative range of postulated UF, releases for the PGDP regardless of power level.

Two bounding UFe release scenarios involving an outdoor rupture of a 14 ton liquid UF.

cylinder and an evaluation basis earthquake were independently evaluated by the staff in terms of impacts to off-site population. Source terms were obtained from SAR Table 4.7-13.

l Assuming reasonably conservative HGSYSTEM/UFe input data for a 14-ton (28,000 pound) liquid UFe cylinder valve failure scenario (opening at the bottom of the cylinder with a i diameter of 1 inch, Pasquill-Gifford atmospheric stability class D (PG-D) and a wind speed l of about 5 m/s), a 1/2 hour exposure to plume centerline airborne concentrations at 1

• kilometer from the release point resulted in a standard-person uranium intake of roughly 60 mg. For PG-B with a 3 m/s wind speed, an intake reduction of about an order of magnitude was predicted by HGSYSTEM/UFe at the same distance. I For a seismically induced UF release of about 64,000 pounds, a very conservatively

)

estimated 1/2 hour plume centerline intake (point source,20 meter stack horizontal release, no building wake effects, PG-F, and 1.5 m/s wind speed) at 1 kilometer was calculated to be roughly 100 mg U. At 1 kilometer, the ground level uranium concentration was roughly half of that in the plume centerline. Assumption of an area source (several release points) would significantly reduce these projected intakes, since very little plume overlap was observed for PG-F at 1 kilometer even for widely separated j release points in a process building.

It should be noted that at short distances (up to several hundred meters) for the cylinder valve failure and seismic scenarios, calculations showed that severe renalinjury or death from exposure to uranium or HF cannot be ruled out. It should also be noted that application of the straight line Gaussian dispersion methodology provided in Regulatory Guide 1.145 (NRC,1983a) for the seismic scenario, raised intakes at 1 kilometer by almost an order of magnitude.

~

Although it is clear that the primary chemical hazard present at the PGDP is UFe, a structured hazard audit to identify materials, equipment, and energy sources which could pose a radiological / chemical threat to the health and safety of the worker or public has not 63

yet been conducted. Completion of the hazard audit is addressed in issue 2 of the PGDP Compliance Plan. However, the consequences of the two developed scenarios are expected to bound the other potential chemical hazards. Areas and buildings where the significant quantities of hazardous material exist, were generally determined based on the information contained in the PGDP SAR. This information is summarized below for the major plant facilities where a significant hazard exists to a member of the public located off-site, a worker located on-site or to the environment.

Enrichment Cascade Systems (C-331. C-333. C-337. C-335 and C-310 uo to the Normetex Pumo Suction) Large quantities of UF, are processed in the diffusion cascade, located primarily in Buildings C-333, C-337, C-331 and C-335. Operating at the maximum power of 3040 MW, Buildings C-333 and C-337 would contain in process, roughly 710,000 and 630,000 pounds of gaseous UF , respectively, within 480 of the largest "000" stages and within process piping and equipment. Buildings C-331 and C-335 would l contain roughly 175,000 and 185,000 pounds of gaseous UFe, respectively, in 480 of the '

relatively smaller "00" stages, and in process piping and equipment. Building C-310 would contain only about 2,500 pounds of gaseous UF,in 60 of the smallest stages. It should be noted that at CUP power, even though most of the Process Gas Hi-Side (compressor exhaust) pressures of the large stages exceed atmospheric pressure, most of the cascade is operated at sub-atmospheric pressures and therefore most process gas confinement failures even at maximum power would initially result in ambient air or seat buffer gases leaking into the cascade system as opposed to process gas leaking out, in addition, USEC is committed to applying NOA-1 in a graded manner for controlling the quality of piping and equipment containing large quantities of UF, and maintaining the process gas pressure (Hi-side compressor exhaust pressure) below 25 psia which was the cascade test pressure during the cascade equipment improvement and upgrade project. This pressure limitation also ensures that the pressure of a cell operating near 25 psia would drop to, or below, atmospheric pressure soon after the cell is tripped, terminating the initial driving force for any release. The cascade system is described in Chapter 4 of this CER.

The Cascade segment of the plant includes piping and equipment (compressors, converters, valves, heat exchangers, etc.). The staff reviewed the seven credible UF, i release accident scenarios pertaining to the Cascade segment of the plant as presented in

{

Chapter 4 of the SAR. These scenarios and their bounding UF release quantities inside the cascade building (in gaseous form unless specified) are: (1) cell overheat and rupture -

5 to 10 lbs; (2) fatigue failure - 1,000 lbs; (3) heavy equipment drop - 4,000 lbs; (4) vehicular impact - 6,000 lbs; (5) valve closure failure - 10,000lbs; (6) coupling failure -

17,500 lbs; and (7) seismic event - 64,000 lbs.

The primary controls relied upon to prevent or mitigate significant potential accident scenarios in the Cascade segment of the plant, and to ensure that the bounding source term is not exceeded, have been identified in the TSRs and SAR Chapters 3 and 4 and include: (1) cascade pressure limitations and protection; (2) freezer / sublimer high weight trip system; (3) freezer / sublimer UF, and R-114 over pressure protection; (4) R-114 coolant overpressure protection system; (5) UFe release detection system; (6) fire protection system; (7) cascade cell trip function; (8) heavy equipment handling c:ontrols; (9) prohibition on applying direct heat to UF, solidified plugs; (10) UF, piping and equipment 64

[

l l

l l

l (greater than 2 inch in diameter) quality control and confinement requirements; and (11) administrative controls specifically identified in Section 2 of the TSRs and contained in )

programs and procedures required by Section 3 of the TSRs.  !

The staff has reviewed the cascade accident scenario descriptions and the associated controls presented in the submitted PGDP TSRs, including the TSR Basis Sections, and PGDP SAR Section 3.15. Based on the information provided in the USEC application, the staff has determined that the scenarios described appear to constitute a reasonable 1 spectrum of postulated accidents and that the safety controls for preventing significant UF, I releases are adequate. In addition, the staff has reviewed and found acceptable, the safety l rationale contained in the justifications for continued operatio,in the Compliance Plan l issues dealing with the cascade segment of the plant.

UF, Feed (C-333A. C-337A). Withdrawal (C-310. C-315). Transfer and Samolino (C-360)

Facilities In the Feed, Transfer (Toll Enrichment) and Sampling facilities, the major pieces of equipment identified as potentially hazardous because of the presence of liquid UFe and j thermal energy, are the feed, transfer and sampling autoclaves. There are 22 autoclaves at l PGDP. Cylinders containing as much as 14 tons of UF, are heated by steam in these l autoclaves to liquify the solid UFe. In addition to being able to provide confinement, the j autoclaves have several appropriate safety systems to prevent or mitigate accidental  !

releases of UFe. The autoclaves and their associated safety systems and operations are described in Chapter 4 of this CER.

The Feed segment of the plant (C-333A and C-337A) includes steam heated feed autoclaves and associated valves, piping, and controls. The significant potential accident scenarios inside autoclaves include failure of cylinder valves and pigtails, hydraulic rupture due to overheating a UFe cylinder or heating an overfilled UFe cylinder, and cylinder rupture caused by an explosive hydrocarbon-UFe reaction. The significant potential accident scenarios outside autoclaves in the feed facilities include mechanical failure of a cylinder (eg. cylinder drop), and failure of UF, cascade feed piping.

The Toll Enrichment and Sampling segment of the plant (C-360) includes steam heated sampling and transfer autoclaves and associated valves, piping, and controls. The significant potential accident scenarios inside autoclaves include failure of cylinder valves I and pigtails, hydraulic rupture due to overheating of a UF, cylinder or heating of an overfilled UF, cylinder, and rupture due to hydrocarbon-UF, reaction. The significant l potential accider t scenarios outside autoclaves include mechanical failure of a liquid UFe cylinder (eg. cylinder drop), failure of cylinder valves and pigtails, and failure of the UFe transfer and mmpling piping.

In the Product (C 310) and Tails (C-315) withdrawal facilities, the major hazard is UFe which exists above atmospheric pressure mostly as liquid in cy;inders (up to 14 tons), and process equipment and piping. Compressors withdraw process gas (tails and product) from the cascade raising the pressure above 1 atmosphere. The gaseous UFe is cooled and liquified before being withdrawn into product and tails cylinders. Product and Tails withdrawal operations are described in Chapter 4 of this CER.

l I

65

1 I

The Product and Tails Withdrawal segments of the plant include UF, withdrawal compressors, condensers, accumulators, and associated valves, piping, cylinders and controls. The significant potential accidents scenarios in the Product and Tails Withdrawal l

segments include mechanical failure of UF, cylinders, failure of cylinder valves and pigtails, I hydrocarbon-UF reaction, and failure of UF, confinement (compressor to withdrawal  !

manifold). l i

The staff reviewed credible UFe release accident scenarios pertaining to the UFe Feed, I Withdrawal, Transfer / Sampling and Cylinder Handling operations at the plant as presented  ;

in the SAR. The worst-case accident associated with these facilities involves the rupture I of a cylinder filled with 28,000lbs of liquid UF .e The specific scenarios identified by USEC and their bounding UF, release quantities (in gaseous form unless specified) are: (1) a 14-ton UF, cylinder rupture - 28,000 lbs liquid; (2) UF, cylinder pigtail f ailure - 660 lbs liquid; and (3) explosion caused by liquid UFe/ hydrocarbon oil reaction during UFe cylinder heating

- 28,000 lbs liquid. .

l The primary controls to prevent or mitigate significant potential accident scenarios at the Feed segment and the Toll Enrichment and Sampling segment of the plant that have been identified by USEC in TSRs and SAR Chapters 3 and 4 as being relied upon for safety are:

(1) autoclave shell pressure limitation; (2) autoclave high pressure containment system; (3) autoclave high pressure relief system;(4) autoclave high steam pressure steam shutoff system; (5) UFe (smoke) detection system; (6) autoclave high condensate level steam shutoff system; (7) UF, cylinder temperature, filling, weight, and heating limitations; (8)

UF cylinder handling (rolling, tilting, disconnection, valve protection and lifting j requirements); (9) UF, freezeout heating limitations; (10) UFe cylinder, UFe cylinder lifting crane and fixtures, UFe cylinder hydraulic lift, and pigtail and UF cylinder scale cart quality control requirements; (11) UFe piping and equipment quality control and confinement j requirements; (12) autoclave confinement, quality control and testing requirements; and l (13) administrative controls specifically identified in Section 2 of the TSRs and contained in programs and procedures required by Section 3 of the TSRs.

UFe withdrawal and UF, cylinder handling accident scenarios and their bounding UF, release quantities (in gaseous form unless specified) identified by USEC for the Product and Tails Withdrawal segment are: (1) 14-ton UFe cylinder rupture - 28,000 lbs liquid; (2) UF, cylinder pigtail failure - 660 lbs liquid; (3) explosion caused by liquid UFe/ hydrocarbon oil I reaction during withdrawal- 28,000lbs liquid; (4) withdrawal compressor (Normetex Pump) f ailure - 250 lbs; (5) withdrawal compressor (Normetex Pump) seal f ailure - 10 lbs; (6) condenser / accumulator / withdrawal manifold piping f ailure - 1,000 lbs liquid; and (7) buffered expansion joint single bellows wall failure - small.

The primary controls to prevent or mitigate significant potential accident scenarios at the Product and Tails Withdrawal segments of the plant that have been identified by USEC in TSRs and SAR Chapters 3 and 4 as being relied upon for safety are: (1) UFe withdrawal (Normetex) pump discharge pressure limit; (2) UF condenser coolant pressure limit; (3) UFe withdrawal (Normetex) pump high discharge pressure shutdown system; (4) coolant (R-114) high pressure relief system; (5) UF (smoke) detection and isolation system; (6) UF, (smoke) detection and alarm system; (7) fire protection sprinkler system; (8) UFe cylinder scale cart movement prevention system; (9) UF, freezeout heating limitations; (10) UFe 66 8

0

cylinder, pigtail, UFe cylinder lifting crane and fixtures, and UF, cylinder scale cart quality control requirements; (11) UFe cylinder fill weight limitations; (12) UF, cylinder handling (disconnection, valve protection and lifting) requirements; (13) UFe accumulator and l condenser minimum wall thickness requirements; (14) UF, cylinder, pigtail, accumulator, condenser, UFe piping and equipment quality control and confinement requirements; and (15) administrative controls specifically identified in Section 2 of the TSRs and contained in programs and procedures required by Section 3 of the TSRs.

According to USEC, other than introducing normal industrial hazards, no other facility introduces significant additional risk to on-site worker health and safety from potential accidents resulting in unconfinement of radioactive material. Minor occupational hazard is introduced in the decontamination building (C-400), waste management facilities and i

technical services facilities. The potential consequences include personal injury, health l effects from acute exposure to toxic chemicals, non-stochastic effects from acute radiation exposure, and risk of latent cancer because of exposure to radioactive material.

The staff has reviewed the non-cascade accident scenario descriptions and the associated controls presented in the submitted PGDP TSRs, including the TSR Basis Sections, and PGDP SAR Section 3.15. Based on the information provided in the USEC application, the staff has determined that the scenarios described appear to constitute a reasonable spectrum of postulated accidents and that the safety controls for preventing significant UFe releases are adequate. The staff has also reviewed and found acceptable the safety rationale contained the justifications for continued operation in the Compliance Plan issues dealing with the non-cascade segments of the plant.

5.2.2 Criticality in the Enrichment Cascade Facilities the following plant areas were evaluated for criticality accident scenarios as presented in Chapter 4 of the SAR: (1) cascade equipment; (2) surge drums; (3) freezer /sublimers; (4) uranium hexafluoride/ refrigerant 114 separator system; (5) cylinder burping / sodium fluoride traps; (6) seal exhaust / wet air pump system; (7) handling and storage of contaminated waste; (8) HEPA filtered vacuum cleaners; (9) sample and calibration buggies; and (10) tops purge trapping system.

General criticality control over the cascade equipment is provided by moderation control, supplemented by process control (temperature and pressure) for cascade components.

Material handled outside of the cascade system is controlled by favorable spacing and volume controls; characterization by duplicate, independent laboratory analysis; or, the j results of a survey confirmed by a single laboratory sample analysis. Material removed from spacing controls (such as the sample and calibration buggies) are handled and stored with batch limits. Contaminated waste is characterized using duplicate, independent i laboratory samples for waste streams, and by establishing always safe mass limits for l waste drums.  !

l In the Chemical Facilities (C-400 and C-409) the following areas were evaluated for l

criticality accident scenarios as presented in Chapter 4 of the SAR: (1) C-400 spray booth; j (2) C-400 uranium recovery unit; (3) C-409 uranium recovery system; (4) C-400 cylinder wash; and (5) field decontamination.

l.

! 67 r

• 1

. _ . . _ ._ _ __ _ _ _ _ _ _ . _ . - _~ _ _ _ . . _ _ . _ _ -_ .

1 4

i l

Chemical Facility criticality control is provided by the use of geometrically favorable tanks, or taking two representative independent samples prior to solution transfer; safe slabs or  ;

continuous monitoring by safety instrumentation in sump pits; administrative controls over mass of uranium contained in equipment to be decontaminated; double-key interlocks over j the uranium recovery systems, and limits on enrichment, concentration and use of i

, favorable geometry containers in C-400; a combination of concentration / density limits and use of favorable geometry containers in the C-409 system; and, spacing for small safe i geometry containers.

j

In the UF, Handling Facilities the following areas were evaluated for criticality accident scenarios as presented in Chapter 4 of the SAR: (1) feed autoclaves; (2) C-360 condensate drain tanks; (3) C-360 scale /evapurator pits; (4) product cylinders; (5) i accumulators; (6) C-310 scale pits; and (7) Normetex Pumps.

Criticality control for the UF, handling facilities is provided by water inventory, operational and configuration control for the autoclaves; favorable geometry or instrumented pits; j cylinder integrity; and moderation and mass control for Normetex pumps.

In the Laboratory the following areas were evaluated for criticality accident scenarios as presented in Chapter 4 of the SAR: (1) liquid uranium salvage; (2) solid uranium salvage; j (3) sampling laboratory; and (4) uranium analysis laboratory. l l

l Criticality control in the laboratory is provided by concentration control over liquid uranium i j salvage and mass control in cold traps; spacing control and batch limits over solid uranium salvage and laboratory uranium samples. l I The staff has reviewed the USEC criticality accident analysis and safety controls and determined that, in combination with the Compliance Plan, they provide adequate assurance of safety and ate acceptable. 1 I

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68 I

Chapter 6 TECHNICAL SAFETY REQUIREMENTS l The regulations in 76.35(e) require the application to contain " Technical Safety l Requirements in accordance with 976.87." The regulations further require a basis statement for the requirement to be part of the application, but not part of the TSR The L TSRs were to consider the information from the safety analysis report and contain l appropriate references to established procedures and/or equipment to address the 14 j topics listed in 76.87(c). The TSRs are to include safety limits, limiting control settings (LCS), limiting conditions for operation (LCO), design features, surveillances, and

! administrative controls as appropriate. The TSRs are contained in volume 4 of the I application. TSR 3.6 requires USEC to controll the TSR Basis statements in accordance with the plant change control process described in 96.3 of the SAR. The TSRs consist of facility / equipment specific TSRs and administrative TSRs. Section 1 contains TSRs related to use and application and includes definitions, time intervals for surveillances, intent of terms, and applicability statements. Section 2 contains the facility-specific TSRs, including  !

TSRs on the autoclaves, UF, detection systems, criticality accident alarm systems, cylinder i handling, cylinder filling, cylinder heating, fire protection system, and other process-related I equipment. Section 3 contains the TSRs related to administrative controls, including responsibility assignment, the organization, staffing, PORC, procedures, training, criticality safety, and commitments to the radiation protection, fire protection, chemical safety, environmental protection, radioactive waste management, and maintenance programs, as well as the other topics required by the regulations.

The staff reviewed the proposed TSRs against the application, the 1985 FSAR, the current i Operational Safety Requirements (OSRs), safety controls incorporated into existing l procedures, nuclear criticality safety requirements for ongoing activities, and current plant safety practices. The staff determined that the TSRsincorporate the safety requirements of the current OSRs and other safety controls utilized by the DOE to ensure safe GDP operations. Several OSR requirements were also clarified during transition to TSRs. The proposed TSR safety limits, limiting control settings, and limiting conditions for operations either incorporated similar OSR limits or were more conservative based upon new information. TSRs defining administrative or programmatic requirements were consistent with other similar NRC requirements or current DOE mandates as a part of the Regulatory i Oversight Agreement. The administrative control and programmatic TSRs are discussed in I other sections of the CER; a summary of the facility specific TSRs is as follows.

Section 1 of the TSRs is the use and application section. Appropriate definitions are I provided by TSR 1.2.1 through TSR 1.2.17. TSR 1.3 defines acceptable time intervals for surveillance (i.e., biennial surveillances could be conducted up to 2 years 6 months). 1 Acronyms and intent of terms (shall, should, may) are covered by TSRs 1.4 and 1.5 respectively. Applicability statements for safety limits (1.6.1), operating limits (1.6.2),

surveillance requirements (1.6.3), and conditions outside TSRs (1.6.4) are provided. The wording for these are similar to the WSTS and are acceptable.

Section 2.1 contains the TSRs for the Toll Transfer and Sampling facility (C-360) and Section 2.2 contains the TSRs for the feed facilities (C-333-A and C-337-A). USEC has established operational modes (TSR 2.1.1 and TSR 2.2.1) that are appropriate for the operations that occur in the feed and sampling and transfer facilities. Safety limits for the 1

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autoclave shell pressure have been set at 110 % of the MAWP (TSR 2.1.2.1 and TSR 2.2.2.1). A safety limit for the UF cylinder temperature (TSR 2.1.2.2 and TSR 2.2.2.2) l has been set to ensure proper ullage for particular cylinder types in use at the facility.

These safety limits are appropriate and provide adequate margin of safety.

Many of the TSRs for these facilities relate to the operation of the autoclaves. The autoclave high pressure isolation systems cause the autoclaves to go into containment I

mode and sound an alarm if the intemal pressure reaches 15 psig. A pressure of 15 psig is I

assumed to indicate that a UFe release has occurred within the autoclave. Placing the autoclave into containment minimizes potential external releases of UFe., The TSRs (2.1.3.1 and 2.2.3.1) state that the actuation pressure shall not exceed 15 psig and that the autoclave containment shall be operable. The TSRs set appropriate potential abnormal conditions, required actions, and surveillances to ensure operability of the systems. One of the surveillances is to conduct a quarterly autoclave pressure decay or leak rate test at a pressure of at least 90 psig.

The autoclave pressure relief system (TSR 2.1.3.2 and 2.2.3.2) shall be operable and actuation pressure set at a level to prevent the rupture of the autoclave (157.5 psig or 105

% of the MAWP). This system prevents the rupture of the autoclave and the subsequent l l

uncontrolled release of UF, by permitting small controlled releases. Appropriate potential l abnormal conditions, actions, and surveillances have been established. The autoclave steam pressure control system provides an indirect means of controlling cylinder l temperature below the safety limits such that a loss of ullage and cylinder overpressurization could not occur. TSRs 2.1.3.3 and 2.2.3.3 establish the LCS on pressure and require the system to be operable. Appropriate potential abnormal conditions, actions, and surveillances have been established in the TSRs. The autoclave water inventory control system shall also be operable to prevent overpressurization cf the autoclave or the possibility of a criticality upon a UF, release. This system limits the amount of water in the autoclave; thereby minimizing the possibility of water mixing with fissile uranium in amounts sufficient to cause a criticality. Limiting the quantity of water is also the most effective method of limiting the maximum pressure generated from a large release. Appropriate potential abnormal conditions, actions, and surveillances are established in TSRs 2.1.4.3 and 2.2.4.2.

The UF release detection systems are required to be operable. These systems alarm upon detection of UF ; some of the systems also cause automatic actions to mitigate the release. The detection system known as the laboratory area (zone 1) isolates the sampling manifold from the parent cylinder by closing the sampling valves to end the release. The system identified as the transfer room (portion of zone 4) automatically isolates the drain manifold from the parent cylinder by closing the block valves and automatically isolates the cylinder by closing the cylinder valve. The detection system in the autoclave heated housings, piping trench, jet station, and along the west wallin C-337-A sound an alarm to alert operating personnel to initiate corrective / mitigative action. TSRs 2.1.4.1,2.1.4.2a, 2.1.4.2b, and 2.2.4.1 establish appropriate potential abnormal conditions, actions, and surveillances to ensure operability.

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l The TSRs for the product and tails withdrawal operations (C-310 and C-315) are contained l in 62.3. USEC has established appropriate operational modes (TSR 2.3.1) for the activities

that occur in these facilities. Safety limits for the Normetex pump discharge pressure (TSR l 2.3.2.1) and the UF condenser coolant pressure (TSR 2.3.2.2)have been established to l ensure the integrity of the expansion joint bellows and UFe condensers respectively. TSR l 2.3.3.1 requires the Normetex pump high discharge pressure system to be operable and l sets an appropriate actuation pressure of 42 psia for the LCS. TSR 2.3.3.2 requires the R-114 coolant overpressure system to be operable and sets appropriate rupture disk actuation pressures at 105 % of the MAWP. Appropriate potential abnormal conditions, ,

l actions, and surveillances have been established for both TSRs.

Several TSRs have been established for the operability of the UF, release detection systems and associated isolation systems. TSR 2.3.4.1 is for the system referred to as the withdrawallow voltage system. This system, upon detection of a UF, release, automatically isolates the withdrawal position to limit the release quantity. The system referred to as the withdrawal station ceiling low voltage system (TSR 2.3.4.2) l automatically isolates all withdrawal positions. The detection system at the Normetex pumps (TSR 2.3.4.3) automatically shuts Cun the pump and closes the discharge valve to decrease the system pressure and mitigate the release. There are also UF, release l l detection systems that do not result in automatic actions; they only alarm. These systems ,

l are located in the heated housings for the condenser and accumulator (TSR 2.3.4.4) and l

, the C-315 high speed centrifugal pumps (TSR 2.3.4.5). All of the TSRs have appropriate l l potential abnormal conditions, actions, and surveillances established to ensure operability i

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of the system.

TSR 2.3.4.6 establishes assay limits for the product withdrawal (2.75 w/o 23sU) and the tails withdrawal (1.0 w/o 23sU) areas that keep operations within the bounds of the criticality analysis.

Cylinder filling and handling TSRs have been established for the feed, withdrawal, and toll transfer and sampling facilities. TSRs 2.1.4.4 and 2.3.4.14 require the cylinder scale cart movement prevention system to be operable. This system prevents the movement of the cart while the cylinder is attached to a pressurized pigtail. Appropriate potential abnormal j conditions, actions, and surveillances have been set. TSRs 2.1.4.12 and 2.3.4.17 state j that the accountability weight of UF drained into a receiving cylinder shall not exceed the l standard filllimit. Normal heating of a cylinder that contains more than the standard fill i limit is prohibited by TSRs 2.1.4.6 and 2.2.4.4. Heating of an overfilled cylinder can result

! in rupture of the cylinder. The scales used for cylinder weight verification are required to be operable by TSRs 2.1.4.21,2.2.4.12, and 2.3.4.23; the surveillances include calibration and functional tests to ensure operability. TSRs 2.1.4.7 and 2.2.4.5 prohibits I certain cylinders from being heated. Certain serial number ranges of type T, O, and OM cylinders lack a certified volume and therefore cannot be heated to liquification. Cold feeding is not currently an authorized activity.

Prior to heating, the cylinder cold pressure shall be :s 10 psia in accordance with TSRs

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2.1.4.9 and 2.2.4.7. The cylinder cold pressure check provides a mechanism to detect the presence of excessive amounts of gaseous impurities in the cylinder that could potentially cause overpressurization of the cylinder during heating. Cylinder valve clarity shall be 71 i

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demonstrated prior to initial heating of a cylinder (2.1.4.19 and 2.2.4.10) and prior to heating a cylinder following heating interruptions (2.1.4.20 and 2.2.4.11). An open pathway assures that cylinder heating will not cause an unrelieved pressure buildup. TSRs l 2.1.4.8,2.2.4.6, and 2.3.4.16 prohibit heating and/or filling a damaged cylinder; a visual inspection is conducted to determine damage to the cylinder.

l Cylinder filling is covered by TSRs 2.1.4.10,2.1.4.11, and 2.3.4.15. TSRs 2.1.4.10 and l 2.3.4.151imit the amount of unknown material that can be in a cylinder to be filled. This

! prohibition.is to provide moderation control for nuclear criticality safety by limiting the j amount of moderating material in the cylinder. The unknown material could also cause an unwanted vigorous reaction with the UFe. TSR 2.1.4.11 timits the number of autoclaves that can be valved to the transfer station to one autoclave (in the transfer mode). The autoclaves in C-360 share.a common transfer header; this action prevents UF releases into other autoclaves. Cylinder handling TSRs require the cylinder pigtails to be disconnected from the UF manifold during rolling or tilting (TSR 2.1.4.13)to prevent

) damage to the pigtail and require the cylinder valve to be closed prior to disconnecting the cylinder from the pigtail to prevent UF, release from an open source (TSRs 2.1. 4.14, 2.2.4.8, and 2.3.4.18). Valve protectors are required to be installed prior to movement of the cylinder per TSRs 2.1.4.15 and 2.3.4.19. The cylinder valve is one of the most likely failure points on a cylinder while the cylinder is being moved. The valve protector provides l additional protection against valve damage. Liquid filled cylinders can only be lifted with overhead cranes that meet specified design criteria (TSRs 2.1.4.16 and 2.3.4.20).

Movement of one UF, cylinder over another cylinder when either contains liquid UF is

prohibited by TSRs 2.1.4.17 and 2.3.4.21. This prohibition preserves the accident

! analysis assumptions. All of these TSRs have appropriate potential abnormal conditions, l actions, and surveillances.

Section 2.4 contains the TSRs for the enrichment cascade. USEC has established appropriate operational modes in TSR 2.4.1. Appropriate safety limits have been established for the freezer / sublimer UFe weight limit (2.4.2.1) and the coolant (R 114) overpressure protection systems (2.4.2.2). Upon activation of the weight trip, the freezer / sublimer is placed in a standby mode, precluding addition of more UFe. The freezer / sublimer weight limit system prevents overfilling with UFe. Overfilling with UF, could cause a stress rupture of the R-114 tubes. Overpressurization and rupture of the R-114 into the UF, system could result in the subsequent loss of UFe containment. These two systems have appropriate LCSs, LCOs, potential abnormal conditions, actions, and surveillances established in TSRs 2.4.3.1 and 2.4.3.4. TSRs 2.4.3.2 and 2.4.3.3 require the freezer / sublimer UF, and R-114 vent line manual block valves to be sealed in the open position to assure the availability of the pressure relief system. TSR 2.4.3.5 states that the intermediate gas removal high temperature control system shall not be operated; this system has not been operated and there are no plans for operation. Appropriate potential abnormal conditions, actions, and surveillances have beer. established.

USEC has also established TSR 2.4.4.11 to limit the cascade high side pressure to s; 25 psia to prevent rupture of the cascade components and subsequent UF, release. The safety limit is set at 40 psia by TSR 2.4.2.3 Appropriate potential abnormal conditions, actions, and surveillances have been established. TSR 2.4.4.12 requires the DC control power and air pressure for cell trip of UFe stage motors to be operable. Several of the 72.

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accident scenarios require operating personnel to respond to process conditions and alarms by de-energizing the process motors (cell trip) in order to bring the cell below atmospheric pressure and thus mitigate or terminate any UF, releases. Appropriate potential abnormal conditions, actions, and surveillances have been set for the cell trip function.

The UF release detection system is required to be operable prior to steady state operations above atmospheric pressure (TSR 2.4.4.1). The TSR requires a set minimum number of detecter heads for operability at the cell housing roof, cell exhaust duct, cell bypass, unit bypan, and B booster pump. Appropriate potential abnormal conditions, actions, and surveillances have been established. In TSR 2.4.4.3, USEC has established appropriate assay limits for cascade equipment. The assay limits keep operations within the bounds of the analysis. Measurement of the product stream provides reasonable assurance that l overall plant assays are within limits. USEC has also established a TSR for dealing with cascade deposits of uranium. The TSR requires moderation control be maintained when deposits of UO2F2 exceeds a safe mass. Moderation controlis the single contingency for ensuring subcriticality in the enrichment cascade. As long as modaration control is maintained, criticality is not possible. TSR 2.4.4.4 establishes appropriate potential abnormal conditions, actions, and surveillances. TSR 2.4.4.13 prohibits movement of large process equipment over other process equipment without communication with the area control room. The TSR allows for necessary movement of equipment in a controlled l manner and is acceptable.

l Equiprnent removal for all f acilities is covered by TSRs 2.5.4.1,2.5.4.2,2.5.4.3, and 2.5.4.4. These TSRs require a presurvey and post survey of equipment to be removed to determine if uranium deposits exist which may require special handling, require opening to be covered with fireproof covers, and control the time allowed for decontamination.

Appropriate potential abnormal conditions, actions, and surveillances have been established. TSRs 2.1.4.18,2.2.4.9,2.3.4.22, and 2.4.4.14 appropriately prohibit the use of direct heat on a UF, plug until flow clarity has been assured. Application of direct heat to the middle portion of a plug can cause local liquefaction of UFe resulting in large hydraulic forces in the pipe, creating the potential for a UF release due to a pipe rupture. I USEC has also established design features for UF, cylinder slings and lifting fixtures (2.1.5.1,2.2.5.1, and 2.3.5.1), cranes (2.1.5.2, 2.2.5.2, and 2.3.5.2), UF cylinders (2.1.5.3, 2.2.5.3, and 2.3.5.3), cylinder pigtails (2.1.5.4, 2.2.5.4, and 2.3.5.4), scale carts (2.1.5.5 and 2.3.5.5), hydraulic lifts (2.1.5.6), interlock switch for hydraulic lifts l

(2.1.5.7), rail stops on hydraulic lifts (2.1.5.8), and UF, condenser and accumulator i minimum wall thickness (2.3.5.6). All of the design features and surveillance requirements are appropriate.

Section 3 of the TSRs contains the administrative controls. TSR 3.17 commits USEC to the packaging and transportation quality assurance program that is described in the NRC-approved version of UEO-1041," Radioactive Material Packaging and Transportation Quality Assurance Program." The staff approved the packaging and transportation quality i assurance program by letter dated March 21,1996. TSR 3.21 states that USEC is not dependent upon outside agencies to provide the level of safety described in the TSR and i

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1 that USEC controls the facilities, structures, systems, and components that are relied upon in the TSRs. All of the other safety topics that 76.87(c) required to be addressed by the TSRs have been addressed by UUEC and are discussed in other chapters of this CER.

The staff has reviewed the TSRs for the Paducah facility in conjunction with the application and concludes that they establish the necessary controls, provide the necessary program commitments for the facility, and meet the requirements of 10 CFR Part 76 and are, therefore, acceptable. The SAR upgrade project may result in the identification of new l safety systems and the need for additional TSRs;it could also result in the elimination of current safety systems and a recommendation to eliminate some of the TSRs. Once the SAR upgrade project is complete, the TSRs will need to be reevaluated based on any new information. ,

While USEC will be allowed to make changes to the TSR basis statements, the staff believes that USEC should not make changes to the TSRs themselves without prior staff i approval. Additionally, in order to impose the TSRs on USEC, the staff recommends the '

following condition:

The United States Enrichment Corporation shall conduct its operations in accordance with the Technical Safety Requirements that are contained in l Volume 4, Revision 5 of the Application dated August 1,1996, as modified j l by Revision 6 of the Application dated August 12,1996. Changes to the Technical Safety Requirements shall require NRC approval prior to implementation.

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I Chapter 7 RADIATION SAFETY l

The regulations in 10 CFR 76.60(d) require USEC to comply with the applicable provisions I of 10 CFR Part 20. In accordance with 10 CFR 20.1101, USEC is required to develop, l document, and implement a radiation protection program commensurate with the scope l and extent of activities to ensure compliance with the provisions contained in Part 20. The PGDP radiation protection program is described in SAR 65.3. TSR 3.13 requires USEC to

( establish, implement, and maintain the program described in the SAR. The TSR also l

requires the following elements to be addressed in the program: HP technicians training and qualifications, personnel exposure control and measurement, contamination control, radic active material control, radiological protection instruments and equipment, and records and reports. The PGDP radiation protection program involves the entire range of facility operations which could affect worker safety pertaining to radioactive materialin normal operations or during accident conditions. The radiation protection program is briefly described below.

l 7.1 ALARA l

USEC is committed to maintaining radiation exposures in accordance with the ALARA l principle. The responsibility for establishing the ALARA policy rests with the Executive Vice President, Operations. The General Manager has the overall responsibility and l authority for the ALARA program and the Radiation Protection (RP) Manager is responsible

! for implementing the ALARA program. The staff finds USEC's proposed ALARA policy to be in accordance with 10 CFR Part 20 and its associated responsible functional positions to be adequate.

USEC is committed to establishing an ALARA Committee. The ALARA committee's authority is limited to reviews and recommendations only. Included in the Committee's charter for the purpose of maintaining occupational doses ALARA are; (1) annually evaluating the implementation of the ALARA program; (2) establishing the annual radiation l exposure goals and monitoring radiation exposure and airborne activity trends; (3) advising the PGDP management and the PORC for the purposes of maintaining occupational doses ALARA (4) reviewing proposed design changes with a projected collective dose greater than 1 person-rem; (5) reviewing practices resulting in a collective dose greater than 1 person-rem; and (6) reviewing revisions to ALARA procedures. Being a subcommittee of l

the Plant Operational Review Committee (PORC), the ALARA Committee chair (RP Manager) reports to the PORC chair.

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Membership of the ALARA Committee includes representatives from RP, Laboratory, Waste Management, Operations, Maintenance, Environmental, Safety and Health, Site and Facilities Support, Emergency Management, Training and Procedures, Nuclear Safety, Industrial Safety and Health, and the Oil, Chemical, Atomic Workers Bargaining Unit. A quorum consists of the chair or vice-chair (Production Support Manager) and five members.

The ALARA committee meets at least semiannually and as directed by the chair. Meeting minutes are provided to each ALARA Committee member and the PORC.

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I 7.2 Responsibilities  !

l The RP Manager directs the RP organization and provides technical oversight of all I radiological protection procedures. The RP Manager, as well as the HP support personnel (HP technicians and their supervisors), have the authority to stop work to maintain the integrity of the RP program. The RP Manager has direct access to the General Manager and the Enrichment Plant Manager. Some of the RP Manager's duties include training of l personnelin the use of radiological program support equipment, determining the need for issuing and closing out radiation work permits, and conducting the radiological  !

occupational monitoring program. Some of the RP support personnel duties include dosimetry, bioassay, instrumentation and calibration functions, and the guiding of workers in the radiological aspects of the job.  :

The minimum qualifications of the RP Manager is a bachelors degree in engineering, health physics, radiation protection, or the physical sciences or equivalent technical experience, i and four years experience in radiation protection, including six months at a uranium  ;

processing facility. HP technicians and their supervisors are required to have technical ,

qualifications pertinent to their assigned duties; training includes initial, on-the-job and I continuing training. Table 5.3-1 of the SAR provides the HP technician course curriculum. l l

7.3 Occupational Radiation Protection 1 l

PGDP has defined the following radiological areas for protection of workers from the chemical toxicity of uranium and from radiation.  ;

i Radiological Material Areas (RMAs): Includes areas or rooms containing radioactive l material more than 10 times Appendix C to 10 CFR 20.1001-20.2401. For 23sU, this I threshold corresponds to 1000 Ci or 30 grams, for 234 U this threshold corresponds to 0.01 pCi and for natural uranium this threshold corresponds to 55 mg. l i

Contamination Control Zones (CCZs): Includes areas generally with removable surface contamination levels up to those provided in SAR Table 5.3-6. Certain discrete areas may exceed these levels. For uranium, the level would be 1,000 dpm/100 cm 2, l

Contamination Areas (CAs): Includes areas with removable surface contamination levels between 1 and 100 times those provided in SAR Table 5.3-6 averaged over 1 m2 . For uranium this range is between 1,000 and 100,000 dpm/100 cm2 averaged over 1 m2 ,

High Contamination Areas (HCAs): Includes areas with removable surface contamination levels above 100 times those provided in SAR Table 5.3-6. For uranium the levels would be above 100,000dpm/100 cm2averaged over 1 m 2, Fixed Contamination Areas: Includes areas with removable surface uranium contamination levels less than 1,000 dpm/100 cm2 but total (removable + fixed) contamination levels greater than 1,000 dpm/100 cm 2. Also includes areas with direct radiation levels greater than 50prem/hr at 1 meter from the surface.

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Soil Contamination Areas: Includes areas with volumetric uranium concentration levels greater than 30 pCi/g.

Airborne Radioactivity Areas (ARAs): Includes areas with potential to exceed airborne soluble uranium concentration levels greater than 50pg/m .

Radiation Areas (ras): Includes areas with direct radiation levels between 5 and 100 mrem /hr at 30 cm from the surface.

High Radiation Areas (HRAs): includes areas with direct radiation levels greater than 100 mrem /hr at 30 cm from the surface.

USEC will post caution signs for RMAs, ARAs, ras, and HRAs in accordance with the posting requirements of 10 CFR Part 20.

7.4 Exposure Controls and Exposure Experience External radiation monitoring dosimeters (NVLAP-accredited thermoluminescent dosimeters, i or TLDs) are issued to monitor external exposures and exchanged at least quarterly (i2 l weeks). Other dosimeters such as finger rings and direct-reading dosimeters are issued where the standard TLDs cannot provide the desired information or are not practical.

USEC states that internal exposures for the PGDP workers will be evaluated by bioassay i procedures to determine intakes based on the biokinetic model in ICRP 30 and that air l sampling will only be used to trigger special bioassay sampling. USEC is committed to I performing bicassay for all personnel who have a potential to exceed doses of 100 mrem per year or 1 mg soluble uranium per week. The staff finds this criteria for performing bioassay and the action levels proposed in SAR Table 5.3-4 adequate to ensure compliance with the 10 CFR Part 20 dose limits. l l

Since the primary internal radiological dose hazard at the PGDP is uranium in soluble form (Class D) and enrichment is less than 5 percent, meeting the 10 mg weekly intake limit over the entire year provides reasonable assurance that the Annual Limit on Intake (ALI) will not be exceeded. Inhalation of 10 mg of depleted uranium with an Activity Median Aerodynamic Diameter (AMAD) of 1.Opm would result in a committed effective dose equivalent (CEDE) of less than 10 mrem.

USEC proposes to collect monthly bioassay (urine) samples for routinely determining uranium intakes. The lowest action level proposed by USEC is 5 g U/l. Using ICRP Report 54, " Individual Monitoring for intakes of Radionuclides by Workers: Design and Interpretation," which utilizes the current ICRP metabolic models, the NRC staff estimates that a lower limit of detection of 5 g/l for uranium in a 24-hour urine sample (1.4 liter) collected during the tenth, twentieth and thirtieth day following a single intake, corresponds to intakes of about a 1,2 and 4 mg of soluble uranium, respectively. These are below the 10 mg weekly intake limit. It should be noted that based on the metabolic l model for uranium presented in ICRP-26 and ICRP-30, which provides data for an AMAD of 1.Opm, the NRC staff calculated an approximate 30 percent increase in uranium deposition in the kioney for a 0.2 pm AMAD particle size. During normal PGDP operations, the NRC l 77

l staff expects any uncontainment of uranium to result in particulates with AMADs greater than 0.3 m. In addition, an evaluation of 31 workers accidentally exposed to natural I uranium in 1986 indicates that the ICRP guidance may overestimate the amount of uranium  !

present in urine at 7 days, since bioassay data indicates more rapid excretion of Class D uranium than originally believed (Fisher et al,1990). This implies that the potential intake l which might go undetected undar the proposed bioassay progr, n could be somewhat higher than estimated using the ICRP model. Nevertheless, the NRC staff concludes that it  !

is unlikely that these potential nonconservatisms would practically account for a large l enough factor so as to result in the PGDP not being able to detect weekly intakes involving l more than 10 mg of soluble uranium.

l l As stated by USEC in Table 5.3-9 of the SAR, during the years 1992,1993 and 1994, I l there were no individual external exposures greater than 500 mrem and the largest internal l radiological dose was 38 mrem CEDE. The highest total effective dose equivalent (TEDE) for those years was less than 350 mrem. It should be noted that 10 CFR Part 20 requires radiation monitoring if there exists a potential for exceeding 10 percent of the annual occupational dose limits which includes a 5,000 mrem TEDE limit. The collective radiological doses for 1992,1993 and 1994 were 7.1,7.6 and 6.8 person-rem.

7.5 Airborne Radioactivity in the Workplace and Ventilation Systems

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i USEC uses continuous air monitors, portable and fixed air-sampling equipment and lapel l samplers in accessible areas where the airborne concentrations could exceed 10 percent of l the derived air concentration (DAC) averaged over 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Low-volume air samplers are located throughout the PGDP in Buildings C-310, C-315, C-331, C-333, C-333A, C-335, -

C-337, C-337A, C-360, C-400, C-410, C-710, C-720 and C-757. Routine low-volume air sample media are changed at least once a week. However, the sample filter nay be exchanged sooner if needed.

USEC uses portable HEPA (high efficiency particulate air) filter systems for general radiological protection purposes and when an unprotected worker may potentially exceed a 1 0.8 DAC-hr exposure. As part of Compliance Plan issue 41, the HEPA filter systems required to control worker exposure will be determined and modified appropriately by July 1997. It is possible that some HEPA filter systems currently being used may not be required for worker protection. However, the above mentioned criterion of 0.8 DAC-hr is acceptable to the NRC staff, since it corresponds to an intake of 1.3 mg or less of soluble uranium. The actions, justification, and schedule are acceptable to the staff.

For HEPA filter units that have differential pressure gages, USEC is committed to checking the differential pressure prior to use or daily when used continuously. The maximum allowable differential pressure proposed by USEC is 5 inches water'.

HEPA filter systems used for personnel protection are tested periodically in accordance with Section 5 - Visual Inspection, Section 8 - Airflow Capacity and Distribution Tests, Section 10 - HEPA Filter Bank In-Place Test, Section 11 - Adsorber Bank in-Place Test and Section 15 - Laboratory Testing of Adsorbent of ASME N510-1989 entitled " Testing of Nuclear Air Treatment Systems" with two exceptions. The adsorber bank in-place test is performed with an alternative, but equivalent, Halide gas, and the laboratory testing of 78

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adsorbent will test for HF loading instead of radiciodine. Partial conformance with ASME N510 as described by USEC is acceptable to the NRC staff since Section 7 - Mounting Frame Pressure Leak Test of ASME N510 is termed " Optional" and the other remaining

, Sections i.e. Section 6 - Duct and Housing Leak and Structural Capability Tests, Section 9 l - Air-Aerosol Mixing Uniformity Test, Section 12 - Duct Damper Bypass Test, Section 13 -

l System Bypass Test and Section 14 - Air Heater Performance Test, of ASME N510 are either not applicable or do not provide a significant additional margin of safety for the types of operations conducted at the PGDP. The two exceptions mentioned above are also acceptable to the NRC staff since the proposed deviations cause the requirements to apply directly to the PGDP operations.

USEC is committed to maintaining the average air velocity above 100 lfpm through openings of hoods used to sample uranium or containing readily dispersible uranium. This is consistent with good radiological safety practice.

Ventilation equipment is typically designed such that normal air flow or leakage flows are generally from areas of lesser contamination to areas of higher potential contamination.

l However,in the PGDP process buildings, general air flow is from the Cell Floor which has a ,

l comparatively higher potential for contamination and UF, releases to the ground (operating)  !

l floor. However, the combination of (1) use of continuous air samplers, (2) UFe release  ;

detectors in cells operating above atmospheric pressure being able to detect UF, releases )

as small as a few pounds, (3) periodic contamination surveys, (4) low numbers of workers i present in the process buildings, (5) most of the cascade operating below 1 atmosphere, )

l (6) worker protection requirements as part of the TSRs, and (7)large mixing volumes of l

the cell and ground floors, contribute to lessen the hazard of this unfavorable ventilation scheme. Considering all of these factors, and operational history of the PGDP, the NRC I staff does not find this unfavorable design of the process building ventilation system to result in a significant hazard to workers present on the operating floor.

l 7.6 Control of Surface and Personnel Contamination

! In radiological areas, worker exposure to surface contamination is minimized by proper use l of surveys, posting, protective clothing and equipment. USEC is committed to limiting skin j l or personal clothing contamination at egress from radiological areas to no more than 1,000 l 1 dpm/100 cm2alpha, and 5,000 dpm/100 cm 2beta / gamma These levels are typically the detection limits of the instrumentation used for personal contamination monitoring at the PGDP. USEC's action levels for surface contamination on laundered protective clothing are 1,000 dpm/100 cm2alpha, and 20,000 dpm/100 cm beta 2

/ gamma. It should be noted that most of the radioactive material that could contaminate protective clothing at the PGDP will be in soluble form and therefore will be readily removed by laundering.

USEC is committed to providing routine contamination survey monitoring in I

lunchrooms /breakrooms adjacent to and within CCZs and permanent boundary control stations on a daily basis; feed / withdrawal stations, contaminated maintenance areas, change rooms and UF, sample handling laboratories on a monthly basis; CCZs on a quarterly basis and the remaining uranium processing areas on an armual basis. Based on the results of the surveys, Radiological Areas will be reposted or decontaminated in accordance with the ALARA principle, as necessary. Even though the survey frequencies 79

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l proposed by USEC are generally lower than those recommended in Reg Guide 8.24 entitled l

" Health Physics Surveys during Enriched Uranium-235 Processing and Fuel Fabrication" I dated October 1979, the staff finds it acceptable for the following reasons: strict I requirements related to the containment of UFe, use of continuous air samplers throughout the PGDP, UF, release detectors in uranium processing areas being able to detect UF.

releases as small as a few pounds, low numbers of workers present in the process buildings, most of the cascade operating below 1 atmosphere, and large mixing volumes in most of the uranium processing areas.

USEC is required by regulation to survey incoming and outgoing shipments of radioactive material. In addition, USEC is committed to restrict release of materials and equipment for unrestricted use if removable surface contamination levels exceed those presented in SAR Table 5.3-6. These levels are in accord with NRC guidance related to unrestricted release of equipment and material.

The NRC staff finds USECs program to control surface and personnel contamination acceptable. The basis for accepting USEC's proposed surface contamination criteria is as follows. Most of the contamination present at the PGDP is in the form of natural, slightly enriched and depleted uranium. For removable surface contamination (23sU) of 1,000 dpm/100 cm 2averaged over an entire facility, using a resuspension factor of 5x10-5 per meter (IAEA,1970), the NRC staff calculated a weekly intake (40-hour exposure) via inhalation of less than 0.4 mg of uranium. For fixed uranium surface contamination of 5,000 dpm/100 cm 2(5 percent enrichment), and assuming an infinite planar source and 100 percent occupancy, the NRC staff calculated an annual deep dose equivalent of less than 1 mrem.

7.7 Respiratory Protection Program USEC utilizes respiratory protection at the PGDP.10 CFR Part 20, Subpart H, provides the requirements for an acceptable respiratory protection program. As stated,in 10 CFR Part 20 and SAR Section 5.3, respiratory protection is only to be relied on when process or other engineering controls are impracticable. The respiratory protection program at the PGDP is administered by the Industrial Hygiene group.

Respirators are used (1) upon entering ARAs, (2) during breach of containment systems, (3) when removable surface contamination exceeds 100 times the values listed in SAR i Table 5.3-6 (e.g.100,000 dpm/100 cm 2for uranium), (4) when work on contaminated '

surfaces has the potential to generate airborne radioactivity and (5) when a worker can potentially be exposed to 0.8 DAC-hours during a work shift. For situations where certain physical limitations, such as heat stress, may contradict safety or not be in accordance with ALARA, the RP Manager could authorize non-use of respirators. These criteria and the condition for non-use of respiratory protection are acceptable to the NRC staff and consistent with the requirements of 10 CFR Part 20. The main reasons for this acceptability are as follows: workers in ARAs without respiratory protection have the potential to intake more than 2.4 mg of soluble uranium; for facilities with an average removable surface contamination (23sU) of 100,000 dpm/100 cm2 , application of a 80

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resuspension factor of 5x10-5 per meter (IAEA,1970) would result in a weekly intake (40-hour exposure) via inhalation of less than 40 mg of uranium; and 0.8 DAC-hours during a work shift corresponds to an intake of less than 1.3 mg of soluble uranium.

7.8 Instrumentation, Calibration, and Maintenance Program Monitoring instruments under the RP program at the PGDP are used to (1) count air and smear samples, (2) measure direct dose rates (mR/hr), (3) continuously monitor the air (alarming continuous air monitors) and (4) detect skin and clothing contamination. Typical detection limits for these detectors (except for the direct dose rate meters) are presented in SAR Table 5.3-8.

l 10 CFR 20.1501(b) requires assurance that instruments and equipment used for j quantitative radiation measurements (dose rate and effluent monitoring) are calibrated l periodically for the radiation measured. The National Council on Radiation Protection (NCRP) and Measurement states that the required frequency of calibration ranges from once every few weeks to annually depending on the amount of use an instrument receives, the environmental condition it is used under, and the historical experience of each l instrument type (NCRP,1991).

NRC licensees must make routine survey measurements with reasonable accuracy and reliability. Reliability is a function of the detector systems, instrument usage, manufacturing quality, and the user's calibration and maintenance programs. USEC has proposed to calibrate instruments based on specifications derived from the applicable vendor manuals or other nationally recognized guidance such as NCRP 112 (NCRP,1991).

In addition, USEC is committed to using calibration sources which are within 5% of the stated value and have documented traceability links to NIST with the exception of large area uranium slab sources, which are certified to 10% of the stated value. USEC is also committed to calibrating air flow measurement devices for air samplers on an annual basis to within 20% of the standard. Currently some of the sources used for setting setpoints and source checking hand and foot monitors are only accurate to 10%; this is addressed by Compliance Plan Issue 11. USEC has requisitioned new sources, all affected instruments will be recalibrated using the new standards within 12 months of receiving the new standards. The hand and foot monitors are calibrated using sources and equipment accurate to 5%. The use of less accurate sources to source check and set alarm setpoints should not reduce the protection of the worker. The plan of action and schedule l l and the justification for coatinued operations are acceptable to the staff. l The NRC staff has reviewed USEC's commitments related to the leak testing of sealed sources and found them to be consistent with general NRC guidance.

USEC has proposed to perform source or operability checks of portable RP instruments on a daily basis or prior to use if not used on a daily basis. Operability checks of instruments, such as the hand and foot monitors, that have intrinsic functional test features are l performed automatically on a continuous basis. In addition to continuous testing, these instruments are also source checked once a week.

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7.9 Radiation Work Permit System USEC is committed to using Radiological Work Permits (RWPs) for work activities in CAs, HCAs, ARAs, ras and HRAs. RWPs implement radiological and uranium's toxicity safety requirements. The RP Manager designates RP personnel who are authorized to approve, issue, update, revise, and close RWPs. In SAR Section 5.3.1.5, USEC states that RP assesses the radiological conditions applicable to a task request. Since in the same SAR I section, USEC also states that RWPs are based on the toxicity of uranium, it may be inferred that RP would also assess the uranium toxicity conditions applicable to the task request. Based on its assessment, RP includes information concerning protective clothing  ;

and radiological hold points, and approves and issues the RWP. RP closes the RWP upon  !

work completion, RWP expiration (maximum effective time period - 1 year), or changes to l the radiological or uranium's toxicological conditions of the job. Prior to closure, RP l evaluates the area to ensure it has retumed to an acceptable condition.

l 7.10 Exemptions from 10 CFR Part 20 i 1

USEC has requested an exemption from 10 CFR Part 20 requirements related to labeling containers.10 CFR 20.1904 requires each container of radioactive material be labeled such that the radionuclide(s) including their estimated quantities, radiation levels, enrichment, and forms are identified. USEC states that it is impractical to label every container located in ras. As a compensatory measure, USEC has proposed to place one caution sign in the area stating that every container may contain radioactive material. In i addition, USEC is committed to surveying all containers removed from contaminated or potentially contaminated areas to ensure that contamination is not spread around the plant i site. USEC has also requested an exemption from labeling UFe cylinders per 10 CFR 20.1904,since these are readily identifiable. As a compensatory measure, USEC has proposed to have UF, cylinders constantly attended by qualifitd Radiological Workers during movement. The staff finds the on-site radiological safen impacts that could result from this exemption to the requirements of 10 CFR 20.1904 to 'ae minimal and recommends that this exemption be granted.

The staff concludes that the radiation protection program is consistent with good industry practice, meets the requirements of 10 CFR Part 20 and 76, and is, therefore acceptable.

1 An item of noncompliance covered by Issue 10 is posting of radioactive material areas. l Building work areas will be reposted to reflect " Restricted Area" and " Contamination <

Control Zone" prior to NRC assuming jurisdiction. Areas which contain unlabeled, but potentially radioactive, material containers will be posted by the end of the year. There are also some leased areas which have not been completely characterized regarding radioactive material present. The characterization and any necessary reposting will be complete by November 30,1997. Compliance Plan issue 9 addresses plant procedures necessary for the annual radiation protection program review; these procedures will be in place by October 1,1996. Compliance Plan Issue 41 addresses the use of portable and installed HEPA filters. Portable and installed HEPA filters are used routinely at PGDP. However, those that are required to control worker exposure have not been identified. These will be identified and testing procedures will be reviewed to allow for testing in accordance with 82

i I SAR 65.3.2.10 by June 10,1997. In addition, procedures for conductin0 daily differential l pressure checks on portable units equipped with differential pressure gauges will be revised by August 30,1996. The justifications for continued operation, plans of action and schedules for these issues are acceptable to the staff.

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O Chapter 8 NUCLEAR CRITICALITY SAFETY The regulations in 10 CFR 76.87(c)(3) require the TSRs to address criticality prevention. In addition,10 CFR 76.89 requires USEC to maintain and operate a criticality monitoring and audible alarm system. The nuclear criticality safety program is also part of the management controls and oversight necessary to protect the public health and safety required by 10 CFR 76.35(a)(7). The nuclear criticality safety (NCS) program is described l in 65.2 of the SAR and required by TSR 3.11.

TSR 3.11 establishes the foundation for the NCS program. USEC is required to establish, l implement, and maintain the program as described in the SAR. The TSR further requires )

the NCS program to address the following elements: adherence with ANSI /ANS standards, NCS responsibilities, process evaluation and approval, design philosophy and review, criticality accident alarm system coverage, procedure requirements, posting and labeling requirements, change control, operation surveillance and assessment, and technical aspects. The TSR requires all operations involving uranium enriched to 1.0 w/o or higher and 15 g or more of 23sU to be based on a documented NCS evaluation and to be performed in accordance with an NCS approval. The TSR sets the minimum margin of subcriticality of 0.02 in k,,, and a k,,, of s; 0.9634 (including the bias, uncertainty, and the margin of subcriticality) for all criticality calculations. The TSR further requires the double contingency principle to be used as the basis for design and operation of processes using fissionable materials; for those instances where double contingency is not met, TSRs shall be established, implemented, and maintained to prevent criticality from occurring. The ,

staff concludes that this TSR sets an acceptable foundation for the NCS program. l 8.1 NCS Administrative Requirements The administrative aspects of the NCS program are discussed in the following subsections.

PGDP has committed to ANSI /ANS 8.1-1983, ANSI /ANS 8.7-1975, and ANSI /ANS 8.19- 1 1984.

8.1.1 NCS Organization and Responsibilities The General Manager has overall responsibility for NCS, approving the implementation of nuclear criticality safety approvals (NCSAs). Responsibility and authority for the implementation and oversight of the NCS program is delegated to all levels of management. Organization managers are responsible for ensuring that operations are identified and evaluated for NCS prior to beginning the activity and for ensuring implementation of the requirements contained in the NCSAs. First-line managers are responsible for ensuring personnel are aware of NCS requirements and that they receive the necessary training.

The NCS section is independent from groups performing operations which require NCS evaluation. The NCS section reports to the manager of Nuclear Safety. The manager of NCS is responsible for the administration of the program, including reviewing the overall effectiveness of the NCS program, ensuring that staff are trained and qualified in accordance with procedures, and ensuring that NCS evaluations and NCSAs are prepared and technically reviewed by qualified NCS engineers.

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Qualified NCS and senior NCS engineers are responsible for: 1) providing NCSAs for fissile material operations: 2) performing facility walk-throughs of facilities that handle fissile i material and advising appropriate supervision of any NCS concerns; 3) participating in investigations of incidents involving NCS and in the determination of recommendations for eliminating such incidents; 4) assisting in plant emergency preparedness planning; 5) providing support to the PORC; and 6) participating in the review of procedures which  ;

involve fissile material operations to verify NCSA commitments are properly reflected.

Additionally, the senior NCS engineer is responsible for verifying sufficient information is documented in nuclear criticality safety evaluations (NCSEs) to allow independent analysis, verifying credible process upsets related to criticality safety are properly identified and evaluated, verifying compliance with the double contingency principle, checking for accuracy, and verifying applicability of the calculational methods.

The minimum requirements for a WCS engineer are: 1) a baccalaureate in engineering,  ;

mathematics, or related science; 2) familiarization with the facility by completing a minimum 1 year in the NCS section; 3) complete KENO V.a training course; 4) perform a minimum of four evaluations under the direction of a senior NCS engineer; 5) perform walk- ,

through inspections under the guidance of a qualified NCS engineer; 6) receive NCS l surveillance team training; 7) attend a nationally recognized criticality safety course; and 8) 1 year of organized training in the physics of nuclear criticality (for those without a physics or nuclear engineering background). The minimum requirements for a senior NCS engineer are: 1) completion of the requirements for a NCS engineer; 2) perform a minimum of four technical reviews of nuclear criticality safety evaluations under the supervision of a qualified senior NCS engineer or the NCS section manager; 3) completion of one year as a qualified NCS engineer and two years PGDP site-specific experience; and 4) be approved by the NCS section manager.

8.1.2 Process Evaluation and Approval Every operation involving 15 g or more of 22sU and uranium enriched to 1 w/o or higher 23sU is evaluated for NCS prior to initiation. The operation and related NCS requirements are documented in an NCSA; the evaluation is documented in a NCSE. The evaluation and approval process is governed by written procedures.

The organization responsible for performing an operation initiates the NCS process by completing a NCSA request form or Part A Request for Criticality Safety Evaluation. Part A documents the operating organizations request for NCS evaluation and the description of the operation. The form is approved and signed by the manager of the operating group.

The form is then submitted to the NCS section for analysis.

An NCSE is prepared to document the analyses performed as specified in the NCSE procedure. Techniques such as an NCS parameter checklist, What-If analysis, or Hazard  ;

and Operability Study are used to identify and document potential upset conditions  ;

presenting NCS concerns. An analysis for each identified process upset condition is then performed to demonstrate double contingency.

l The NCS evaluation process involves: (1) a review of the proposed operation and l procedures, (2) discussions with the subject matter experts to determine the credible 85

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process ucsets which need to be considered, (3) development of the controls necessary to meet the double contingency principle, and (4) identification of the assumptions and l equipment (i.e., physical controls) needed to ensure criticality safety.

Once the NCSE is complete, a technical review of the evaluation is performed and I documented by a senior NCS engineer. I Part B of the NCSA is prepared based on the results of the NCSE. Part B documents the conditions.of approval of the operation. These conditions of approvalinclude the process conditions that must be maintained to meet the double contingency principle or preserve the documented basis for criticality safety and restrict the modes of operation to those that l' have been analyzed in the NCSE. The requirements to be included in operating procedures and postings are identified. If needed, development of TSRs would also be included. ,

l The NCSA approval process first involves the acceptance of the NCSE and NCSA by the technical reviewer. A 10 CFR 76.68 review will be performed to determine if prior NRC approvalis necessary. The NCSE and NCSA are reviewed by the PORC and, if acceptable, approved by the General Manager. The PORC reviews the NCSE and NCSA to verify technical accuracy, ensure all credible process upsets have been identified and ensure consistency with other NCSAs and other potentially conflicting requirements or regulations.

Once approved, the NCS controls, limits, evaluation assumptions, and safety items are verified to be fully implemented in the field. The verification documentation is maintained as a quality record along with the NCS evaluation. The manager of the operations group signs Part B. The NCSA is then issued as a controlled document.

First-line management is responsible for implementing the conditions delineated in the NCSAs through the use of such tools as training, operating procedures, postings, and labels. First-line management ensures postings and labels are prepared and verifies that they are properly installed as required by the NCSA. The operating procedures are prepared or modified by first-line management to incorporate the NCSA requirements.

First-line management is responsible for ensuring the employees understand both the procedures and NCSA requirements before the work begins.

The NCS program applies the double contingency principle by implementing controls either on two different parameters or by implementing two controls on one parameter. Controls include passive barriers such as structures, vessels, and piping; active engineered features such as valves, themocouples, and flow meters; and administrative controls that require human actions to be taken in accordance with approved procedures. When two controls are implemented for one parameter, the violation or failure scenarios of the controls shall be independent. Application of this principle ensures that no single credible event can result in an accidental criticality or that the occurrence of events necessary to result in a criticality is not credible. USEC has committed to the double contingency principle in TSR 3.11 and has committed to ANSl/ANS 8.1.

There are three operations at the PGDP which do not meet the double contingency principle. These are product cylinder operations, operation of the enrichment cascade, and removal of large cascade equipment (e.g., compressors, convertors, G-17 valve, etc).

These operations have been evaluated to be safe and are described in the accident 86

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I analysis. There are TSRs to ensure controls are in place for these operations. Sections 2.4 and 2.5 of the TSRs list controls for operation of the enrichment cascade and for removal and maintenance of enrichment cascade equipment, respectively. Section 2.3 of the TSRs provides the controls associated with ensuring moderation control for the product cylinders.

All other operations shall comply with the double contingency principle. In the event future i operations are found to not comply with the double contingency principle, USEC is required I to develop a TSR to cover the single contingency. The TSR requires NRC approval prior to implementation.

Emergencies arising from unforeseen circumstances can present the need for immediate l action. If NCS expertise or guidance is needed immediately to avert the potential for a l criticality accident, direction will be provided verbally or in writing. Such direction can include a stop work order or other appropriate instructions. A NCSA or other form of l documentation will then be prepared to justify the actions taken once the emergency l condition has been stabilized. This documentation shall be prepared within 49 hours5.671296e-4 days <br />0.0136 hours <br />8.101852e-5 weeks <br />1.86445e-5 months <br /> l following the stabilization of the emergency condition.  !

l 8.1.3 Design Philosophy and Review l

l Designs of new fissile material equipment and processes have to be approved by the NCS section prior to implementation. Designs willinclude the use of favorable geometry or engineered controls on mass, moderation, volume, concentration, interaction, or neutron absorption, as the preferred approach over the use of administrative controls. The -

preferred design approach includes two goals. One is to design equipment with NCS independent of the amount of internal moderation or fissile concentrations, the degree of interspersed moderation between units, the thickness of reflectors, the fissile material density, and the fissile material chemical form. The other goal is to minimize the possibility of accumulating fissile materialin inaccessible locations and, where practical, to use favorable geometry for those inaccessible locations.

Fissile material equipment designs and modifications are reviewed to ensure that favorable geometry and engineered controls are used to advantage. Administrative limits and controls are implemented in NCSAs to satisfy the double contingency principle for those cases where the preferred design approach cannot be met.

8.1.4 Procedure Requirements Operations to which NCS pertains shall be governed by written procedures. These procedures contain the appropriate NCS controls for processing, storing, and handling of fissile material. The NCSA requirements which require employee actions shall be incorporated into the operating procedure. NCSA requirements are identified by placing a

" commitment stamp" in the left hand margin next to the appropriate procedure step.

l Identifying these requirements in this way ensures changes to these requirements are not

! made without review and approval by the NCS Section. The NCSA requirements are incorporated into the appropriate procedures as required by the NCS procedure.

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9 New and modified procedures are reviewed by the NCS Section. The NCS Section reviews the procedures to verify that the appropriate NCSA requirements have been incorporated and to verify that the proposed operation complies with NCS program requirements.

8.1.5 Posting and Labeling Requirements NCS limits and controls for areas, equipment, and containers are presented through the use of postings and labels as specified in approved NCSAs and procedures. Postings and labels are proposed, reviewed, and approved during the review and approval process. The limits and controls are posted on nuclear criticality safety requirements signs as required by the plant NCS procedure. The design of labels are prepared by the operating organization and concurred on by NCS. The approved NCSAs snecify the wording for the postings.

Limits and controls are printed in an appropriate size typeface and the postings and labels are placed in conspicuous locations.

8.1.6 Change Control Functional and physical characteristics of operations controlled for NCS are described in NCSAs and NCSEs. These components and features which are identified in the NCSAs and NCSEs are analyzed to determine the " boundary" of the system, encompassing those items that are essential to ensure operability. The boundaries are identified on system  !

drawings and the configuration is verified to be as-built. These components and features are documented in a manual for each facility. Each time a change to a facility is planned, this manualis reviewed by the individual planning the change to determine if the change affects SSCs relied on for safety. The design control manual specifies the organizations required to perform reviews of changes to SSCs. The required approvals are obtained  ;

before the change is implemented. The Configuration Control Board verifies the required reviews have been perforrned before approval. If an item is relied on for the criticality safety of an operation it will be identified through the work control process as an NCS SSC .

and NCS Section approvalis required before implementing the change. The NCS Section  !

reviews the NCSE for this specific operation and determines if the change affects the analysis performed and conclusions made in the NCSE. The change request will be approved by NCS only if the change does not impact NCS or once a revised NCSE has determined that the change is acceptable and meets NCS' program requirements. In this way modifications to controlled operations are evaluated and approved prior to implementation. The items which require configuration change control are identified as specified by the safety system boundary identification procedure. These components and features are then reviewed during surveillances, assessments and walk-throughs of the facilities to verify that unauthorized modifications have not been made. l 8.1.7 Operation Surveillance and Audits To ensure that the NCS program is properly established and implemented, USEC utilizes surveillances, assessments, audits, and walk-throughs. The NCS section performs internal surveillances of the plant's implementation of NCS limits and controls in accordance with I procedural requirements, need, and results of incident trending. The operating organizations also perform surveillances for fissile material operations on an annual basis.

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NCS personnel provide technical support. The surveillances include the inspection of facility modifications, operating procedures, compliance with NCSAs, postings, and waste generation and handling.

l NCS walk-throughs are performed to determine the adequacy of implementation of NCS l requirements and to verify that conditions have not been altered to adversely affect NCS.

! Additionally, audits are conducted in accordance with the QA program to determine the adequacy of the overall NCS program, including the adequacy of the NCSEs, NCSAs, internal surveillances, and implementation of requirements.

The results of the surveillances, audits, and walk-throughs are documented and reported to appropriate managers. Identified deficiencies are documented and corrected according to the problem reporting system.

8.2 Criticality Accident Alarm System A Criticality Accident Alarm System (CAAS)is provided to alert personnelif a criticality accident should occur. The system utilizes a distinctive audible signal to notify personnel in the affected area to initiate evacuation, thereby reducing personnel exposure to emitted radiation.

The CAAS provides detection and alarm coverage for postulated criticality events that would produce an absorbed dose in soft tissue of 20 rad of combined neutron and gamma radiation at an unshielded distance of 2 m from the reacting material within 1 minute as required by 76.89(b). The detection criteria are met by setting PGDP detectors at 10 milliroentgen per hour above the background radiation rate for the area (s) of coverage.

The CAAS detects gamma dose rate. The system uses clustered detectors with each cluster containing three scintillation detectors. Activation of any two of the three detectors in a cluster willinitiate evacuation alarms. The failure of any major component of the system will result in a notification that indicates the need for corrective maintenance. The location of detectors and setpoints is based on results of dose calculations and detector tests performed at critical experiment facilities. The need for CAAS coverage is considered during the NCS evaluation process.

TSRs are in place to ensure operability of the criticality accident detection coverage and the criticality accident alarm. A CAAS is required (by TSR) for areas, equipment, or processes which contain greater than 700 grams of 23sU at an enrichment 21.0 w/o 23sU.

The TSRs (2.1.4.5a, 2.1.4.5 b, 2.2.4.3a, 2.2.4.3b, 2.3.4.7a, 2.3.4.7b, 2.4.4.2a, 2.4.4.2b, 2.6.4.1a, and 2.6.4.1 b) establish appropriate potential abnormal conditions, actions, and surveillances for the CAAS. The "a" TSRs cover the detection function and the "b" TSRs cover the audibility function. Required actions to be taken if the system is inoperable include ceasing movement of fissile material. The staff concludes that these TSRs are acceptable.

The regulations require a criticality monitoring and audible alarm system in all areas of the facility. The regulations (76.89(a)) also allow USEC to request appro' val to exclude areas from the monitoring requirement. By letter dated April 19,1996, as revised by letter dated August 15,1996, USEC submitted a request to exclude areas from CAAS monitoring.

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This request is still under review by the staff. The staff will complete its review of the exclusion request during the transition period. Compliance Plan Issue 8 states that the compliance plan issue (8) will be modified to provide a plan and schedule for establishing adequate coverage if the NRC does not agree to the request to exclude certain areas from CAAS monitoring. In the interim, the staff understands that the areas subject to the request do not have CAAS coverage.

There are three Compliance Plan issues that concern aspects of the CAAS; issues 8,46, and 50. Compliance Plan Issue 8 also coveis the use of a portable CAAS in building C-710 in areas that are not adequately covered by the permanent system. The permanent CAAS in C-710 will be modified by June 30,1998. The portable system, building howlers, and radios will be used to compensate for the deficiencies discussed in the Compliance Plan until the system is fixed. This is acceptable to the staff.

Compliance Plan issue 46 concerns CAAS horn audibility. There are snme areas in C-331, C-335, and C-337 where the CAAS horns are not audible to personnel due to high ambient noise levels. PGDP is conducting a full sound survey of all buildings with CAAS horns.

Plant modifications will be made to ensure that the CAAS horns can be heard throughout the affected areas of the process buildings. These modifications will be complete by December 15,1998. As a compensatory measure for the inaudible CAAS alarms, the building howlers will be sounded within 10 seconds following actuation of any CAAS detection alarm cluster. The 10-second time delay for building howler actuation would not have a significant impact on evacuating the areas of the process buildings where the CAAS alarms cannot be heard. The justification for continued operation with the compensatory measure and the plan of action and schedule are acceptable to the staff. ~

i Compliance Plan issue 50 concerns criticality accident alarms for nearby buildings. The  !

evacuation area is determined based on a 12 rad exposure from a postulated criticality.

Several of the leased buildings which are located within the evacuation area do not have evacuation horns and/or light actuated by the clusters detecting the accident. The alarms from the adjacent alarmed buildings cannot be heard within most of the unalarmed buildings. Therefore, the requirement that a criticality accident be annunciated into all areas requiring evacuation is not met. As compensatory measures postings will be I installed to identify each unalarmed building within an evacuation path and personnel I entering the unalarmed buildings will be required to have a radio capable of receiving emergency information and will be required to monitor the radio continuously while inside.

Announcements would be made over the radio. The necessary modifications will be complete by December 15,1998. The staff finds the plan of action and schedule along with the justification for continued operation and compensatory measures to be acceptable.

8.3 Technical Criteria 8.3.1 Parameters The use of NCS parameters have been described in detail in SAR 55,2. Moderation, volume, interactions (spacing), geometry, mass, enrichment, density,' heterogeneity, concentration, and reflection are all parameters that are used for NCS control. The specific controlis documented in NCSEs and NCSAs. Pararneters are considered in the NCSEs.

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d Criteria have been established in SAR 95.2 to ensure appropriate control of the parameter.

An exception is the use of neutron absorbers. USEC has agreed that before a neutron absorber is used, for the purposes of complying with the double contingency principle, the details specifying the neutron absorber control program shall be submitted to the NRC for review.

8.3.2 Methods of Calculation NCS calculational methods have been derived and validated for the determination of criticality safety values associated with all process and storage operations performed at PGDP. To assist in these calculations, PGDP has validated computer codes, provided appropriate reference handbooks, and performs hand calculations in a prescribed manner.

For those cases where adequate references or experimental data are not available, NCS computational analyses are performed, which involve the calculation of k,,, to determine if the system will be subcritical under both normal and credible abnormal process conditions.

Computer codes that simulate the behavior of neutrons in a process system or that solve the Boltzmann transport equation are used.

Computer calculations of k ,, provide a method to relate analytical models of specific system configurations to experimental data derived from critical experiments. A critical experiment is defined as a system which is intentionally constructed to achieve a self-sustaining neutron chain reaction or criticality. Critical experiments which have specific,

, well-defined parametric values and are adequately documented are termed benchmark experiments. Computer codes are validated using experimental data from benchmark experiments which, ideally, have geometries and material compositions similar to the systems being modeled. The PGDP computer code validation establishes the upper limit for calculations as k,,,s 0.9634 in accordance with the validation report KY/S-221, Rev.1

" Validation of the Paducah Gaseous Diffusion Plant Nuclear Criticality Safety Code System for the ENDF/B-IV 27 Group Cross Sections," dated January 1996. Reference to the minimum margin of subcriticality of 0.02 in k,,, plus the associated uncertainty and bias has been retained as a requirement and thus is consistent with k ,,s0.9634 provided in the referenced validation report.

The computer codes and cross sections used in performing k,,, calculations are maintained as part of the configuration control program. Access to the code is controlled, confirmation tests are conducted and comparisons against an archived production version is performed.

Both the administrative aspects and the technical aspects of the NCS program have been reviewed by the staff. In addition, onsite reviews of each program element were conducted between November 1994 and April 1995 by knowledgeable NRC personnel.

The staff concludes that the program and TSR commitments are consistent with good industry practice, meet the requirements of 10 CFR Part 76, and are, therefore, acceptable.

Several aspects of the NCS program are not fully implemented and are addressed in Compliance Plan issues 5,6,8,46, and 50. Issues 8,46, and 50 are discussed in CER 98.2: issues 5 and 6 are discussed below.

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Compliance Plan issue 5 concerns nuclear criticality safety approval documents. There are operations at PGDP for which the NCSEs are incomplete, formal documentation is unavailable, or are not fully documented. There are also some administrative aspects of the NCS program that have not yet been proceduralized or documented. The procedural changes to resolve the administrative aspects will be complete by October 31,1996. The completion date for preparing and approving the NCSE and NCSA documentation for operations with 1 w/o or higher 2asU and 15 grams or more 23sU is December 2,1996. All operations with > 2 w/o rasU and all new operations either have, or will have prior to start up, properl.y documented NCSAs and NCSEs. In the interim, PGDP will continue to use the existing documents and controls. This is acceptable to the staff.

Compliance Plan issue 6 concerns nuclear criticality safety approval implementation. There are currently some inconsistencies between the specification in the NCSAs and the supporting implementation procedures and work-site postings. Field verification of the procedures and postings will be compiled and documented by December 2,1996. This is acceptable to the staff.

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Chapter 9 ENVIRONMENTAL PROTECTION AND WASTE MANAGEMENT The regulations in 10 CFR 76.60(d) requires that USEC comply with the applicable provisions of 10 CFR part 20. USEC describes its radiological environmental program in 65.1 of the SAR. By TSR 3.16, USEC is required to establish, implement, and maintain the program described in the SAR. The Radioactive Waste Management Program is required I by TSR 3.14. This chapter briefly discusses the USEC management of effluents and l waste. USEC follows the ALARA principle as it pertains to releases. It is USEC policy to l

prevent or. minimize the discharge of radioactive material to the environment. USEC has a pollution prevention program in place at the site.

9.1 Effluents The environmental program includes a system of process and administrative controls to prevent releases above regulatory limits and to maintain effluents ALARA. Emission points and the emission controls are discussed in the following paragraphs. TSR 3.8 requires USEC to control emissions as described in SAR 95.1.

PGDP has only two monitored emission points. One is the C-310 Purge Vent Stack. This stack is continuously sampled prior to venting. This system discharges the gases of lower molecular weight that have entered the cascade and those vented from the cylinder burp station. The gases first pass through chemical traps and then are vented to the stack.

Action levels have been established for discharges from this point with associated actions that range from reviewing the data to closing the vent / stack. The other monitored emission point is the C-335 UFe /R-114 Separation System. The continuous sampler is only operational when the system is operational. This system is used when relatively large amounts of R-114 coolant have entered the cascade and mixed with UFe. The gas mixture is passed through the system to recover and return UF, to the cascade. The discharge from the cold trap passes through chemical traps before the gases are vented to the roof or routed to the seal exhaust / wet air system. The emissions from this system are typically low. A summary of radionuclide emissions is provided in the Environmental Compliance Status Report.

The remaining airbome emission points are not monitored but are evaluated on a set frequency agreed to by EPA. There are several discharge points from the C-400 facility; the decontamination spray booth, uranium recovery unit, dryers, and cylinder cleaning.

The decontamination spray booth is equipped with mist eliminators to remove any droplets of contaminated liquid from the booth exhaust prior to venting. The filter exhaust air from the uranium recovery unit flows through an entrainment separator prior to discharging through a roof vent. Air from the dryers passes through lint traps prior to discharge.

Exhaust from the cylinder cleaning station is discharged outside the building without treatment.

Operations from the C-710 Laboratory which involve gaseous UFeoccur in closed systems with emissions limited by the use of cold or chemical traps.

Process buildings C-310, C-331, C-333, C-335, and C-337 each have a wet air exhaust system. The wet air evacuation system consists of vacuum pumps that remove air from 93

i diffusion equipment. The exhaust is passed through chemical traps, pump oil, and mist eliminators before venting. These buildings also have seal exhaust stations that collect and vent air that has leaked through the shaft seals of cascade compressors. The air passes through chemical traps, pump oil, and mist eliminators. These vents may be combined 3 with the wet air exhaust system.

l The C-409 facility exhaust from cylinder washing passes through a soda-ash solution i scrubber prior to discharge. Air from cylinder drying is exhausted outside the building with

no further. treatment. The filter exhaust air from the treatment of the aqueous solutions

! flow through an entrainment separator prior to discharge. Air from the receiving booth for

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contaminated equipment exhausts through a HEPA filter system.

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The consequences of radionuclides released to the atmosphere from PGDP is determined I by calculation of the committed effective dose equivalent (CEDE) to the maximally exposed

! person and to the entire population residing within 80 km (50 mile) of the plant. The j maximally exposed individual is located approximately 2,400 m north of the plant site. The

dose calculations are made using the CAP-88 package of computer codes. This is an

accepted methodology. The dose (1994) to the maximally exposed individual from both j routine and unplanned release was 0.016 mrem. The calculated CEDE was 0.011 person-rem. PGDP is required to submit an annual report to the Environmental Protection Agency that summarizes the airborne radionuclide emissions from the plant.

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Wastewater streams at PGDP can be classified into the following categories: recirculating l Water blowdown, nonprocess wastewater, process wastewater, sanitary wastewater, and e runoff. USEC leases 13 outfalls at the site,10 of these are monitored for radiological contaminants. The Environmental Compliance Status Report provides monitoring data for i these outfalls.

Nonprocess wastewater includes once-through cooling water from pumps, air conditioners, steam condensate, drinking water drains, safety shower drains, eyewash bath drains, and

, other uses of potable and nonpotable water. Nonprocess wastewaters are discharged 4

directly to the outfalls. Process wastewaters are treated for removal of contaminants and are sampled and analyzed prior to discharge. The C-400 and C-409 facilities generate the

! majority of and process all radiologically contaminated wastewaters. Prior to discharge or

. reuse, decontamination solutions are processed in one of the treatment facilities. Treated wastewaters from these facilities are discharged through Outfall 001.

i l Sanitary wastewaterincludes all grey-water and black-water discharges and some

wastewaters for cleaning activities, including wastewaters from the laundry and safety 2

equipment cleaning. Sanitary wastewater is treated at the C-615 Sewage Treatment Facility prior to discharge through Outfall 008.

Runoff includes runoff from building roof drains, parking lots, roads and facility grounds.

Runoff from buildings and terrain on the west side of the facility discharge through Outfall 008. Outfall 009 and 016 dischargos runoff from the southwest side of the plant. Outfall 013 collects and discharges runoff from the southeast corner of the plant including the C-745D and C-745G cylinder storage yards.

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o The recirculatory cooling water blowdown from the cooling towers is treated at the C-616 Wastewater Treatment Facility for phosphorus. The effluent discharges to the C-616F lagoon and then through Outfall 001.

Since there are no intakes of surface water for domestic or routine livestock watering purposes in Big or Little Bayou Creek and no water intakes on the Ohio River within 15 miles downstream of the plant, PGDP does not calculate a dose from liquid effluents.

The facility meets the dose limitations contained in the regulations. The PGDP effluent program meets the requirements of the regulations, and is, therefore, acceptable.

9.2 Environmental Monitoring PGDP has onsite and offsite permanent stations to collect ambient air samples. PGDP uses high-volume air samplers that use treated filter paper to collect particulate radionuclides and particulate gaseous fluorides. Samples are analyzed weekly for gross alpha and beta deposition. Action levels have been established that when exceeded trigger an investigation. The high-volume samplers have been in operation since August 1995.

USEC will use the 1996 data to compare the calculated effective dose equivalent from the high-volume ambient air samplers and the actual release data. This will be completed by June 30,1997 when PGDP submits its 1996 annual report to EPA. This schedule is acceptable to the staff. This issue is discussed in Compliance Plan Issue 38.

PGDP has an onsite meteorological tower located in the southern section of the site. It is equipped with instrument packages at the 10 m and 60 m levels to measure air j temperature, dew point, and wind speed and direction. There is also ground level instrumentation to measure solar radiation, barometric pressure, and precipitation.

Thermoluminescent dosimeters (TLDs) are located along the site perimeter fence, at the reservation boundary, at nearby residences and communities, and at two remote )

background locations distant from the site. The TLDs measure external gamma radiation j and are collected and read on a quarterly basis.

Biological monitoring is conducted to assess the impact of plant operations on vegetation in the vicinity of the plant. Vegetation is sampled annually and analyzed for total uranium concentration and "Tc beta activity. Sample locations are selected annually and are based on meteorological data, land use census, and the availability of vegetation. Samples generally consist of food crops such as corn, squash, and tomatoes. PGDP has action levels that trigger an investigation if exceeded.

Soil and sediment samples are collected from the site on an annual basis. Sediment samples are analyzed for total uranium,2nPu,237Np, 2nU, 23oTh, and "Tc. Soil samples are analyzed for total uranium. Action levels are established that when exceeded trigger an investigation.

USEC collects water samples at the outfalls originating within the security area that discharge to waters of the United States. Continuous, flow proportional sampling is used in the continuously flowing KPDES outfalls. Grab samples are collected at the intermittent 95

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or rainfall outfalls. The sampling frequency is established by the KPDES permit. Samples l are analyzed for total uranium, # 5U, "Tc, 237Np, 23sPu, 2 Th, alpha, and beta as described in the SAR. USEC also collects water samples from the receiving streams around the plant.

1 The Environmental Monitoring Department is responsible for the collection, validation, and l reporting of all field and analytical data from the monitoring program. Data is reviewed to identify possible trends and compared against background locations. Currently some of the environmental data are not trended; this issue is addressed in Compliance Plan issue 37.

Current practices include reviews of environmental data to identify any unusual results that might indicate an increase in releases. This practice will continue until the necessary procedures are developed and implemented by December 31,1997. This is acceptable to the staff.

The monitoring program is acceptab!e to the staff.

The regulations in 10 CFR 76.35(g) requires USEC to submit a " compliance status report that includes the status of various State, local and Federal permits, licenses, approvals, and other entitlements, as described in !i 51.45(d) of this chapter. The report must include environmental and effluent monitoring data." As part of the application, USEC submitted an Environmental Compliance Report. The Environmental Compliance Report contained information on the environmental permits issued to the facility, including the principal permit limits, a summary of monitoring and emissions / effluent data for each permit, and a summary statement on the status of USEC compliance. USEC also provided a summary of the data from the environmental monitoring program. The report met the requirements of '

the regulation and is therefore acceptable.

The regulations in 10 CFR 76.35(c) require the application to contain "any relevant information concerning deviations from the published Environmental Impact Statement, ,

Environmental Assessments, or environmental permits under which the plants currently operate from which the Commission can prepare an environmental assessment related to the compliance plan." To meet this requirement USEC submitted a report called Supplemental Environmental Information. The information in this report and other information in the application was used in preparation of the environmental assessment that the staff prepared to support approval of the Compliance Plan. The report met the requirements of the regulation and is therefore acceptable.

9.3 Waste Management The regulations in 10 CFR 76.35(m) require "A description of the program, as appropriate, for processing, management, and disposal of mixed and radioactive wastes and depleted uranium generated by operations." To meet this requirement, USEC submitted two plans:

the Radioactive Waste Management Plan and Depleted Uranium Management Plan.

The Radioactive Waste Management Plan (RWMP) addresses the management of ~

radioactive and mixed wastes. USEC projects that about 60,000 ft of low-level radioactive waste is generated each year and about 860 ft2 of mixed waste is generated.

The RWMP describes the various wastestreams. Uranium is the primary radiological 96

I I contaminant in the waste. The wastes may also contain trace quantities of transuranics l (237Np and 23sPu) and fission products (Tc) from past processing of reactor return l material. Each waste, whether liquid or solid, is characterized as releasable or not by the use of radiological surveys, sampling and analysis and/or the radiological status of its area of generation. Wastes not suitable for release are classified in accordance with 10 CFR 61.55 prior to disposal.

Wastes are collected at the source in drums, tanks, or boxes appropriate to the type of l waste generated prior to treatment or transfer to a waste storage facility Mixed wastes I are transferred to a 90-day accumulation area equipped with secondary containment. '

After sampling, mixed wastes are transferred to DOE for storage until treatment / disposal i options are available. USEC does not store mixed waste beyond 90 days. USEC has designated waste storage areas for different low-level wastes. The C-757 building is used to inspect and sort waste for further processing or storage. The wastes are inspected for hazardous materials and are monitored for the presence of radioactive contamination using bulk or hand monitors. Wastes suitable for release to unrestricted areas are transferred to the nonradiological side of the facility. Radioactive wastes are containerized for shipment to storaga or treatment facilities. Liquid wastes are processed in C-400 and C-409 for the removal of uranium from the wastewaters. The processed wastewater is either reused or discharged through outfall 001.

The Depleted Uranium Management Plan describes USEC's program for the management and disposition of the depleted uranium produced at the facility. USEC has projected production levels of depleted uranium through the year 2005. Currently the cylinders containing the depleted uranium are stored onsite in cylinder storage yards. USEC conducts an initial inspection to document cylinder condition and then reinspects every 4 l

years to check for any indication of cylinder damage. Any damage discovered is evaluated for corrective actions which may involve more frequent inspection, cylinder repair, or emptying of the cylinder. )

l There are certain aspects of the Depleted Uranium Management Plan that are not currently l in place and are discussed in Compliance Plan issue 35. First, depleted uranium handling, moving, and stacking procedures are part of the procedure upgrade program and have not l yet been updated. USEC will continue to utilize existing procedures until the upgraded l procedures are in place at the end of 1996. The other item of noncompliance has to do with the initial and follow-up inspections. Currently USEC does not document the initial inspections and does not conduct follow-up inspections. USEC does conduct a visual inspection to detect damage and verify proper positioning in the storage yard; however, the inspections are not documented with a log sheet. PGDP will have the procedures to implement the baseline and periodic inspections by December 31,1996. This timeframe is acceptable. All USEC cylinders were either new or were cleaned and hydrostatically pressure tested prior to being filled with depleted uranium. Followup inspections on the first USEC cylinders will not be due until July 1997. USEC will conduct a baseline inspection for all cylinders currently in storage by December 31,1997. This schedule is l acceptable. -

l l There are several potential uses for depleted uranium that are being investigated by both l

USEC and DOE. However, for planning purposes, USEC is assuming that the ultimate 97

disposition strategy for the remaining inventory will consist of converting the solid UFe to U30, using the pyrohydrolis process in which the UFe will be converted into a gas and combined with steam. The UF, gas will react with the steam to form UO22F which is then converted to U30s in high temperature calciners. The U 30, would then be packaged and shipped to an authorized repository. This approach is acceptable to the staff.

The USEC Radioactive Waste Management Plan and the Depleted Uranium Management Plan are acceptable to the staff and meet the requirements of the regulations.

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4 Chapter 10 CHEMICAL SAFETY The regulations in 10 CFR 76.87 require the TSRs to address chemical safety. Chemical safety includes the chemical hazards derived from radioactive materials, and plant conditions related to the hazards of chemicals on or near the site that may directly or i

indirectly affect radiation risk.

TSH 3.18 requires USEC to establish, implement, and maintain a chemical safety program as described in the SAR. Section 5.6 of the SAR describes the Chemical Safety Program at PGDP.

The program integrates the environmental, safety, and health management systems. The chemical safety control strategy requires the identification and listing of chemicals in current use. The chemicals are categorized by potential chemical risk and nuclear safety significance, and are addressed by either the accident analysis in Chapter 4 of the SAR, the Process Safety Management (PSM) program, or the Industrial Health and Safety programs for chemical hazards. Existing plant programs are used to assure chemical safety through the incorporation of technical and administrative controls to manage risk. Those programs involve the development and use of operating procedures, site-wide safety procedures, operator training, maintenance, configuration management, emergency planning, incident investigation, audits and inspections, quality assurance, human factors, and detection and monitoring. Cross-references are made to the program requirements addressed in other sections of the TSRs.

The potential for a release nf hazardous chemicals and the impact on nuclear safety are

• analyzed in Chapter 4 of the SAR. The analysis considers both the radiological and toxicological effects of UFe. The analysis recognizes the possible risks of a severe chemical accident associated with UF, processing, and the chemical toxicity associated with the release of hydrofluoric acid and soluble forms of uranium that would result from the reaction of UF, and moisture in the air. The technical and administrative controls necessary to assure an acceptable level of safety for plant workers and members of the l

public from the processing of UFe are addressed in other sections of this CER.

l Plant conditions that may either directly or indirectly affect radiation safety are addressed  :

by one of two programs. A number of hazardous chemicals at PGDP are managed under l l the OSHA PSM program per 29 CFR 1910.119 to prevent an impact on nuclear safety, the l workers, and members of the general public. Those chemicals are chlorine, chlorine l trifluoride, and fluorine. Of these three chemicals, chlorine is above the threshold quantity l l and requires complete impiementation of the elements of the OSHA standard. Chlorine l l trifluoride and fluorine require administrative controls to restrict the respective inventory to levels below the threshold. The structures, systems and components (SSCs) associated l with those chemicals are considered AO under the QA program; the configuration control, l maintenance, and inspection are applied in a graded approach. The uses, locations, and

( quantities of those chemicals are described in the SAR. A Process Hazards Analysis (PHA) l is required by the OSHA PSM rule to be performed for each of the chemical systems handling those hazardous chemicals if the inventory is above the threshold limit.

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. l Two NRC site visits were conducted on November 14-18,1994, and January 9-13,1995, to assess the adequacy of the chemical safety program at PGDP, walkdown those hazardous chemical systems, and review the associated PHAs. Observations made during  ;

those site visits are documented in NRC Observation Report Nos. 70-7001/94003and 70-  !

7001/94004. The overall chemical safety program was in the process of being formalized at that time. The material condition of the hazardous chemical systems was found to be generally acceptable. The PHAs performed for the chlorine systems appeared to be thorough and rigorous.

The performance of a PHA provides an acceptable method of identifying potential hazards )

and recommendations for reducing the risks associated with the specified chemical systems. The OSHA PSM program at PGDP contains clear requirements that the PHA recommendations be evaluated by management and that the recommendations accepted by management be prioritized and performed according to an approved schedule.

Hazardous and toxic chemicals not covered by the above programs are managed by industrial hygiene and safety programs without regard to a threshold inventory barrier to enable the safo use of these chemicals. The chemicalidentification and inventory control feature involves three processes: (1) the baseline identification of chemical inventories  !

used onsite, (2) a formal Engineering Service Order program to address modifications to 1 existing systems in order to consider new or revised chemical applications onsite, and (3) contractor control to ensure that hazardous and toxic materia!s brought onsite are properly authorized and controlled. These proposed administrative controls provide an acceptable level of assurance that chemicals used onsite in less than bulk quantities will be identified, evaIuated and adequately controlled.

l The staff concludes that the chemical safety program meets the requirements in 10 CFR 76,is consistent with industry practice, and is, therefore, acceptable.

1 USEC does not currently have in place all the aspects of a mechanicalintegrity program for )

covered hazardous chemicals to assure adequate confinement measures for those hazards.

This is addressed in issue 40 in the Compliance Plan. The mechanicalintegrity program for maintenance and inspection PSM requirements will be implemented by May 26,1997, which is consistent with OSHA requirements. Until the program is complete, the facility will continue to use the program that existed under DOE regulatory authority. As an additional measure PGDP will use work packages when performing maintenance on chemical systems. The staff finds the justification for continued operation and the schedule for completion to be acceptable. Issue 43 of the Compliance Plan covers the use of hazardous chemicals by DOE and other entities located on the PGDP site. Under this item DOE and third party tenants will provide USEC information on their hazardous chemical use. These parties will be required to inform USEC prior to bringing hazardous chemicals onsite. The staff finds the schedule to be acceptable.

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! Chapter 11 FIRE PROTECTION

( The regulations in 10 CFR 76.35(a)(6) require the SAR to include a " description of equipment and facilities which will be used by the Corporation to protect health and minimize danger to life or property" such as " fire protection systems". The PGDP fire protection program is described in 55.4 of the SAR. TSR 3.12 requires the establishment, implementation, and maintenance of the fire protection program. The following paragraphs briefly describe the PGDP fire protection program.

l The program is implemented by Fire Services (FS), which is headed by a management l official who has been assigned the responsibility of the Authority Having Jurisdiction

! (AHJ). The AHJ is a qualified fire protection professional having a degree in engineering or .

! a technical curriculum and at least 6 years of applicable experience. The FS is staffed with  !

l Fire Protection Engineers and Fire Officers. The PORC provides oversight and review of l

the activities of the FS and all fire safety-related issues.

3 l The program includes performance of annual building surveys for the purpose of ensuring -

continued fire safety of the operations. This activity includes review of the fire safety features of the buildings, their occupancy, the hazards of the processes and storage in the

! buildings, and any other related issues. This is considered an important activity as the fire l safety status of a building may change with time because of changes in the facility, l processes, or management methods. A review of sample building surveys indicates a i l methodical effort applied to the task, and the surveys should be effective in timely l detection of many safety problems. [

! The program includes inspection, testing, and maintenance of the fire protection equipment, which includes fire water systems, sprinkler systems, and a fire alarm system.  !

The procedures and the testing and maintenance frequencies are based on applicable l

[ National Fire Protection Association (NFPA) codes. Major elements of the inspection l program and associated frequencies are described in the SAR.

PGDP's hot work permit system is designed to control cutting, welding, and other hot work conducted in a manner consistent with industry fire prevention practices. Line Managers, r who issue the permits, receive training on fire safety. FS is notified prior to the initial use -

l of a permit; field surveillance of work is conducted during routine inspections.

The program also includes training of the plant fire department personnel who provide 1 emergency response service and manual fire fighting capability. Training is based on l l national standard emergency response methodology and includes site specific issues.

Specific training activities include fire-fighting, hazardous material response, confined  ;

space rescue, emergency medical response, radiological emergencies, and rescue. The '

fire-fighting personnel participate in the drills and exercises on the emergency plan. The FS is also responsible for control of impairments of fire protection equipment, investigating fire l incidents, and providing input to plant design and modifications.

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• I 11.1 Fire Protection Equipme'nt

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The facility is protected by two fire water systems: a High Pressure Fire Water System (HPFWS) that supplies the sprinkler systems in most of the process buildings and a low pressure Sanitary and Fire Water System (SFWS), which supplies sprinkler systems in the ,

other buildings and other water needs. Both have independent water supplies. The HPFWS is supplied by a 300,000-gallonoverhead tank and surface reservoir having a ,

, capacity of 4.76 million gallons. Both systems have redundant electric and diesel-driven l fire pumps, The fire main systems are fitted with sectional and post-indicator valves and '

hydrants. PGDP is upgrading the fire protection system to enhance reliability. Two fire  ;

pumps associated with the HPFWS system are being replaced / refurbished; this task will be l completed by October 31,1996. PGDP is also reconfiguring the fire-water supply piping  !

to C-315. After reconfiguration, those area of C-315 in which O list systems are located i

will be supplied from the HPFWS system. The Compliance Plan item is scheduled to be l complete by March 31,1997. The plan and schedule in Compliance Plan issue 14 are l acceptable to the staff.

The fire alarm system is activated by water-flow sensors in the sprinkler systems and fire detectors and other sensors to detect plant upsets. The alarm is transmitted to an annunciator in the central plant control room, which is manned around-the-clock PGDP is upgrading the fire alarm system to enhance overall reliability of the system (Compliance Plan issue 13). These upgrades are scheduled to be complete prior to NRC assuming regulatory jurisdiction.

l The plant fire department is equipped with two pumpers, a ladder truck, and other

• l emergency response equipment. The installed fire suppression and fire fighting equipment j is adequate for the f acility.

l PGDP has TS As (2.3.4.8, 2.3.4.9, 2.3.4.10,2.3.4.11, and 2.3.4.12) that cover the fire protection system for buildings C-310 and C-315. These TSRs cover the operability of the distribution mains, water supply basin, the storage tanks, pumps, and the sprinkler system.

l The TSRs establish appropriate potential abnormal conditions, actions, and surveillances for the systems. TSRs (2.4.4.5,2.4.4.6,2.4.4.7,2.4.4.8, and 2.4.4.9) have also been established for the fire protection systems for C-331, C-333, C-335, and C-337. These TSRs cover the operability of the HPFWS distribution mains, sprinkler systems, water ,

supply basin, storage tanks, and HPFWS pumps. The TSRs establish appropriate potential l abnormal conditions, actions, and surveillances for the systems. PGDP has also l established TSRs to cover hot work operation when the fire systems are not operational (TSR 2.3.4.13 and 2.4.4.10). All of these TSRs are acceptable to the staff. l 11.2 Building Construction The process building construction comprises of structural steel frames, non-bearing wa!!s, concrete floors, and built-up roofs. The roofs contain layers of bituminous combustible material, so that the buildings do not comply with the requirements of Type I construction as classified by the National Fire Protection Association code NFPA 220, Types of Building 102 i

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! Construction. h1 wever, all of the process buildings are fitted with automatic sprinkler systems, which t ompensates for the departure from the Type I classification in that the probability of a fine within the building involving the roof is greatly minimized.

l The ventilation system for the process buildings is comprised of forced draft fans drawing l fresh air at the first floor air intakes and delivering into the second floor space. Air is withdrawn from the second floor space through the second floor deck by a duct system l and induced draft fans and returned to the atmosphere or partially recirculated within the l building. The ventilation system handles air at a large volumetric rate, which is necessary l for the particular application. While this is a fire protection concern, since high ventilation l rate may assist propagation of fire, especially since each floor of the process building is effectively one very large fire area without compartments, the sprinkler system reduces this risk to an acceptable level.

! Tunnels connect the switch houses, process buildings, anc central control facilities. The tunnels contain cables mainly for control and communications; some 440 V AC and 250 V DC power cables are also located in the tunnels. All cables are insulated to 600 volts except for communication cables that are located in separate low voltage trays. All cables have neoprene or PVC jackets which are considered flame retardant. The tunnels do not have automatic fire suppression systems or fire detectors. However, materials of construction are noncombustible, the trays have been maintained free of debris and/or combustibles, and transient combustible loadings are small. The only reasonable source of l ignition is electrical in origin.

The process buildings being very large, the maximum egress path lengths, prescribed by NFPA 101, Life Safety Code, from most parts of the buildings are exceeded. In TSR 3.23, PGDP is required to identify and mark emergency egress routes in process areas and maintain them free of obstruction.

11.3 Process Fire Safety 1

The predominant process-related fire hazard arises from the possibility of a breach in the i forced lubrication system for the compressors and other machinery and the resulting spill catching fire. The accident analysis section of the application recognizes the hazard.

i Apart from the lubricating oil, which is combustible, the average fire loading of the process l buildings is light. An aqueous film-forming foam (AFFF) is the ideal suppression agent for 1 an oil-spill fire. The facility fire department is capable of delivering AFFF from their l

pumpers. The buildings are fitted with automatic sprinkler systems, which also can help suppress and control propagation of an oil spill fire. The probability of an oil-spill fire ,

originating in the cell housing, or propagating into it from outside, is a concern that was j carefully considered. There is no automatic fire suppression system inside the cell housing.

However, the combustible loading inside the cell housing is light and UF, smoke detectors '

are installed in the housing and would alarm upon detecting smoke generated from a fire.

The plant Fire Services would thus be alerted of a fire in the cell housing and timely intervention with manned fire-fighting equipment would occur.

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4 In Compliance Plan issue 17, USEC has committed to complete a combustible loading analysis for the process buildings by June 30,1997. This schedule is acceptable to the staff.

11.4 Fire Hazard Analysis The facility has building surveys performed on the major buildings. These surveys provide building descriptions, changes since the test updating of the survey, and descriptions of the hazards therein. The documents provide a baseline for fire hazard evaluation. PGDP evaluates fire hazards for the major building annually and documents the findings in the building surveys.

11.5 Pre-Fire Plan The pre-fire plan for the facility provides action plans for credible emergencies, such as lube oil fires and UF releases; building access and exit points; control locations for utilities:

cautionary for protection against hazards from UFe and R-114 coolant release, potential oxygen-depleted areas, electrical hazards, etc.; and relevant building construction and occupancy details. Fire protection features, such as sprinkler systems and valves, hydrants, and hose houses are described and their locations identified on drawings. PGDP committed to update the pre-fire plans on an annual basis. However, PGDP has not updated the pre-fire plans recently; this issue is addressed in Compliance Plan issue 17.

The pre-fire plans will be updated by the end of 1996 which is acceptable to the staff.

The staff concludes that the PGDP fire protection program is acceptable and meets the requirements of 10 CFR Part 76.

An area of noncompliance addressed by issue 16 involves fire protection procedures. The fire protection procedures are part of the procedure upgrade program. Procedures specifically related to the application commitments have been upgraded and implemented, except the procedure for welding, burning, and hot work protection which will be implemented by the end of the year. Remaining procedures will be upgraded by June 1, 1997,in accordance with the procedure upgrade program. This is acceptable to the staff.

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1 Chapter 12 EMERGENCY PREPAREDNESS l The regulations in 10 CFR 76.35(t) require USEC to submit an " emergency plan that meets l the requirements of 976.91." Section 76.91 describes the type of information to be l

included in the Emergency Plan. USEC submitted an Emergency Plan for the PGDP with the application. The following paragraphs briefly describe the PGDP Emergency Plan.

l 12.1 Plant Description Chapter 1 of the Plan adequately describes the NRC-regulated activities and provides a l description of the uranium enrichment process. The Plan also specifies the substances associated with the enrichment process which could pose hazards if released to the environment and the locations where they are used or stored at the plant. The Plan l adequately describes the plant's main process buildings and the various other buildings located on the plant site, and the area near the plant, which is a rurallow-population area.

12.2 Types of Accidents  ;

Chapter 2 of the Plan adequately gives a brief description of each type of accident that could result in consequences beyond the DOE reservation boundary which could possibly require protective action recommendations to off-site agencies.4lt lists quantities and locations of radioactive and hazardous materials located at the plant.

12.3 Classification of accidents l Chapter 3 of the Plan describes the plant's system for classifying emergencies as alerts or

site area emergencies, and the action levels required by each. The Plan requires an alert be declared if an incident has led or could lead to a release to the environment of radioactive i

or other hazardous material bLt is not expected to require a response by off-site response organizations. The Plan requires a site area emergency be declared if an incident has led or could lead to a significant release to the environment of radioactive or other hazardous material and could require response by off-site organizations to protect persons off-site.

The definitions for alert and site area emergency are acceptable. Specific emergency l action levels for classifying and declaring an emergency are in an Emergency Plan implementing Procedure.

12.4 Detection of Accidents Chapter 2 of the Plan adequately identifies the means for detecting accidents or abnormal operating conditions in a timely manner. These include a criticality accident alarm system, UF, detecting equipment, a fire protection system consisting of automatic sprinkler systems and fire alarms, and various chemical detectors.

12.5 Mitigation of Consequences i

Chapter 5 and 6 of the Plan adequately describe the means for mitigating the consequences of an accident. The plant has systems and instrumentation available for i

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detecting abnormal operating conditions that could result in an emergency. Provisions and  ;

procedures are provided for evacuating personnel in the immediate incident area and controlling access to the surrounding accident vicinity. Mobile fire-fighting equipment is maintained onsite to support fire fighting, back up the fixed fire suppression systems, and ,

provide a hazmat response capability, and each major process building has an automatic )

sprinkler system. Provisions and procedures are provided for treating and transporting )

injured workers, and providing protective action recommendations to local off-site officials  !

in potentially affected off-site areas. The Public Address (PA) system is one of the ways l l

personnel on the DOE reservation are alerted that a threat or potential threat exists.

However, an item of non-compliance exists with the PA system in that it does not provide sufficient assurance that all on-site personnel can be notified of protective action recommendations because of dead spots. The PA system also has had system reliability problems. Additional PA speakers are required outside throughout the plant and in two j buildings. PGDP has plans to upgrade the PA system to extend coverage to those plant l l areas not currently covered. The existing PA system is augmented by the plant radio '

system and telephones to ensure notification of plant personnel, including those who may be in dead spots. The scheduled completion date for this upgrade, covered by Compliance Plan issue 31,is April 30,1997. The NRC staff feels this is adequate for upgrading the PA system because the plant has alternative ways of notifying personnel of an emergency.

Section 7.6 adequately describes the program for assuring emergency equipment is operable and properly stored and maintained.

12.6 Assessment of Releases Sections 5.2 and 6.4 of the Plan provide a description of the methods and equipment for assessing releases of radioactive or hazardous material. Assessment actions during an Alert include increased surveillance of plant instrumentation and visual observation of incident conditions, and monitoring event conditions for potential changes in the emergency classification level. Assessment actions during a Site Area Emergency include assessment of on and off-site exposures regularly to determine if and when on-site sheltering may be required. Additional activities can include performing continuing emergency assessments for mitigating events and on-site protective actions based on on-scene and field monitoring results, release information, and meteorological conditions for j radiological or hazardous material releases. Radiation detection equipment is used onsite for normal and emergency response use. If access is available to the monitors during a  :

release, water sampling is performed, and bioassay sampling is conducted to assess uranium uptake to individuals exposed or suspected of being exposed in the event of a UF, I release.

12.7 Responsibilities Chapter 4 of the Plan provides an adequate description of the responsibilities of plant personnel during an emergency. During an emergency the General Manager is authorized to declare an emergency and initiate the appropriate response. The Plant Shift Superintendent assumes a dual role as Crisis Manager and incident Commander until the Emergency Operations Center (EOC) is activated, at which time the General Manager or designee assun m the role of Crisis Manager. This is conducted by procedural checklists and, if possible, . ace-to-face briefings. The EOC is automatically activated for alerts and 106

i site area emergencies. Once the Crisis Manager responsibilities have been transferred from the PSS to the Plant Manager, the PSS maintains responsibility of incident Commander at the incident scene. Chapter 3 describes those responsible for notifying the NRC-Operations Center immediately after notifying the off-site authorities but no later than one hour after the declaration of an emergency. Chapter 7 of the Plan identifies the department responsible for maintaining and updating the Plan.

12.8 Notification and Coordination Chapter 3 of the Plan provides a commitment to notify off-site authorities of an emergency. Section 4.3 of the Plan describes provisions for requesting off-site assistance, and 95.6 and 65.7 describe medical transportation and treatment of contaminated workers. Chapter 6 of the Plan adequately describes the facilities and equipment at the plant for mitigating emergencies. Facilities include an emergency operations facility, an emergency operations center, a central control facility, a command post, a central alarm station, on-site medical facilities, and decontamination facilities. Equiprnent includes I communications equipment, emergency monitoring equipment, and a plant weather monitoring system. Part of the communications equipment is the Public Warning System I which consists of outdoor warning sirens within a two mile immediate Notification Area surrounding the plant, and Emergency Broadcast System announcements. However an item of non-compliance exists with the Public Warning System (Compliance Plan issue 30).

The sirens do not provide total coverage within the two mile immediate notification area.

Plans are in place to upgrade the existing sirens and provide full coverage in the immediate notification area. The sirens are scheduled to be installed and operational by March 15, 1997. Messages through the Emergency Broadcast System are available and functioning and are an alternate method for providing public warning. The NRC staff feels the alternate method for providing public warning is adequate and the scheduled completion date for upgrading the systems is acceptable.

12.9 information To Be Communicated Section 3.3 of the Plan describes the information to be communicated to off-site response organizations and the NRC during emergency notifications. The information communicated includes plant status conditions, radiological / hazardous materials release data, recommendations for protective actions for off-site response organizations, and other ,

applicable emergency information. I 12.10 Training Section 7.2 of the Plan describes the type of training that the general plant personnel, emergency response organization and support personnel, other DOE reservation personnel, and off-site emergency support organizations receive. Training records are retained to document readiness assurance.

12.11 Procedures for safe shutdown and recovery .

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Chapter 9 of the Plan describes the means of restoring the facility to a safe condition after l an accident. Recovery and restoration activities are conducted to maintain exposures as

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low as reasonably achievable. A recovery organization is established and managed by a Recovery Manager. This manager has overall responsibility for recovery activities which include checking safety equipment involved in the emergency and restoring it to normal conditions. In addition, TSR 3.7 requires USEC to establish, implement, and maintain procedures to prescribe plant response to earthquakes, tornado /high winds, and flooding / intense precipitation.

12.12 Exercises Section 7.3 of the Plan adequately describes the provisions in place to conduct biennial exercises of the emergency plan. Off-site response organizations and the NRC are invited to observe or participate in these exercises. In addition to the biennial exercises the plant conducts drills to test individual elements of the emergency plan. A drill and exercise working group is responsible for scenario development, scheduling, and identifying participants and evaluators. Formal critiques are conducted, deficiencies are identified, and corrective actions assigned and tracked through completion or implementation. Remaining critiques are reviewed by appropriate personnel for validity and to determine an appropriate method for corrective actions. Communications checks with off-site response organizations are conducted quarterly, and telephone numbers are checked and updated.

12.13 Hazardous Chemicals Chapter 10 of the Plan states that the plant complies with the Emergency Planning and Community Right to Know Act. Plant procedures have been developed for hazardous materials releases that are not classified as emergencies to ensure that requirements of SARA Title lit are met. Material Safety Data Sheets are maintained at various areas of the plant.

12.14 Comment From Offsite Response Organizations USEC submitted the letters they received from off-site response organizations commenting on the Emergency Plan. No significant problems with the plan were identified. One organization indicated that they could not transport contaminated individuals on their helicopter but otherwise would be able to provide transportation if requested. The plant indicated that they had other ambulatory services available, including their own, to transport contaminated individuals and that this was not a concern. ,

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Chapter 7 of the Plan indicates any changes to the Plan are communicated to the l appropriate off-site response organizations. Letters of agreement with off-site response  !

organizations are reviewed and upd1ted every four years or more frequently if needed. I I

12.15 Changes to Emergency Plan The regulations allow USEC to make changes to the Emergency Plan if the changes do not  !

decrease the effectiveness of the Plan and the changes are provided to the NRC and j

affected off-site response organizations within six months of the change. In 97.1 of the J l

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response organizations within 6 months of the change.

! The staff concludes that the Emergency Plan meets the requirements in 10 CFR 76, and is therefore acceptable.

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O Chapter 13 SECURITY AND SAFEGUARDS 13.1 Material Control and Accounting The regulations in 10 CFR 76, Subpart E requires USEC to meet specifi' equirements within Parts 70 and 74 for material control and accounting for Category Category ll, and Category 111 special nuclear material (SNM). The PGDP possest. ion limits for special nuclear material of low strategic significance (SNM-LSS) are such that only safeguards requirements for Category til SNM-LSS apply to USEC activities at this plant. Specifically, USEC must comply with the applicable requirements of 10 CFR is 70.51,74.11,74.13, 74.15, 74.17, 73.33, 74.81, and 74.82.

The NRC recognizes that USEC may opt to engage in production or nonproduction activities that involve other than Category lli material. In that event they must apply for and be certified by the NRC as meeting the applicable safeguards regulations in accordance with the category of material that USEC seeks to either possess and use, or possess only.

USEC must implement an NRC approved Fundamental Nuclear Material Control (FNMC)

Plan pursuant to 10 CFR 74.33(b)(2), achieve the general performance objectives of 10 CFR 74.33(a), maintain the system capabilities required by 10 CFR 74.33(c), and establish records which comply with the record keeping requirements of 10 CFR 74.33(d)(1).

Guidance for preparation of a FNMC Plan is provided in Regulatory Guide 5.67," Material Control and Accounting for Uranium Enrichment Facilities Authorized to Produce SNM of Low Strategic Significance" end in NUREG/CR-5734," Recommendations to the NRC on Acceptable Standard Format and Content for the FNMC Plan Required for Low-Enriched Uranium Enrichment Facilitier."

As part of the application, USEC submitted the " Fundamental Nuclear Material Control Plan" for the Paducah facility; because of its nature, this plan is not publicly available. The plan describes how the PGDP facility will meet applicable NRC material control and accounting requirements. The staff concludes that the FNMC Plan for the PGDP satisfies the performance objectives and system capabilities required by the regulations, therefore the Plan is deemed acceptable.

Relative to its contained commitments, the FNMC Plan is acceptable, however, there are several aspects that are not yet in place and are addressed in the Appendix to the Compliance Plan. This portion of the Compliance Plan is not publicly available. The Appendix contains 5 issues related to material control and accounting (A.1, A.2, A.3, A.4, and A.7). All items, with the exception of one subitem in A.1, will be completed on or before December 31,1996;the subitem in A.1 will be completed by December 31,1997.

The staff finds the plans and schedules to be acceptable and the justifications for continued operation contain adequate measures for continued operations until the facility is in full compliance 13.2 Physical Security and Transportation Protection The regulations in 10 CFR Part 76, Subpart E requires USEC to meet specific requirements within Parts 70,73 and 74 for physicel protection of Category I, Category 11, and Category 110

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111 SNM. Specifically, USEC must comply with the applicable requirements of 10 CFR I 73.67,73.71, and 73.74. I The NRC recognizes that USEC may opt to engage in production or nonproduction activities that involve other than Category lli material. In that event they must apply for and be certified by.the NRC as meeting the applicable safeguard regulations in accordance with the category of material that USEC seeks to access, use, or possess.

I USEC mus.t meet the general performance objectives of 10 CFR 73.67(a); submit a  ;

physical security plan per 10 CFR 9 73.67(c); and comply with the measures for physical j protection of SNM-LSS as required by 9 73.67(f) at plant sites and (g) for SNM-LSS intransit. Guidance for preparation of the Physical Protection Plans and the Transportation Protection Plans by USEC is provided in Nuclear Regulatory Commission Regulatory Guide 5.59, Revision 1, " Standard Format and Content for a Licensee Physical Security Plan for Protection of Special Nuclear Material of Moderate or Low Strategic Significance," February 1983.

l As part of the application for the PGDP, USEC submitted the Physical Protection Plan and I the Transportation Security Plan; because of their nature, these plans are not publicly available.

USEC has committed in the Physical Protection Plan and the Transportation Protection Plan for PGDP to implement procedures and measures to comply with the requirements of 573.67(f) and (g). The following paragraphs briefly address the program elements.

Storage and Use of Material. The entire facility is enclosed within a controlled access area bounded by a chain link fence. A clear zone surrounds the perimeter fencing. Access into the controlled access area is controlled and monitored by a security force and security patrols to allow entry of only authorized employees, visitors, delivery materials and products. There are procedures to control and account for SNM-LSS to assure this materialis used and stored only within authorized areas and only by authorized individuals.

Access Controls. Access through the perimeter fencing into the controlled access area is channelled through personnel and vehicle gates. Personnel entry points and vehicle gates are manned and monitored by security personnel when in use. Access into the site is l based upon established need and authorization. Uncleared visitors are escorted. Personnel and packages are searched for items which could be used for theft or sabotage. Access I control procedures include search of suspiciously acting persons and of packages. There i are procedures for control and accounting of SNM-LSS to assure that this materialis used l and stored only in authorized areas by authorized personnelin performance of their duties. l 1

Detection of Unauthorized Penetrations or Activities. USEC monitors the site perimeters, the controlled access areas, and storage areas by scheduled and by roving security patrols and from fixed posts at site gates. ~

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• l l penetrations or activities. The security patrols provide a detection and assessment capability, and form the primary response. The remaining on-duty security elements and, if necessary, local law enforcement authorities provide backup and support. Communication l between and among these elements is established, maintained, and exercised on a l periodic basis. PGDP has established and maintains liaison with local law enforcement authorities.

1 Personnel Trustworthiness. USEC requires that all personnel working at the facility be screened f.or personnel trustworthiness. Personnel are cleared at levels commensurate with their site access authorization and duties.

l Advance Notification and Confirmation of Shipment. PGDP has and implements l procedures for advance notification both prior to shipment and for confirmation after arrival at destination for each shipment of SNM-LSS.

Inspection of Shipments. PGDP has and implements procedures for use of containers, seals, and locks during transport and inspections prior to and on receipt of shipment of SNM-LSS.

l in-transit Physical Protection and Response Procedures. USEC is responsible prior to shipment to arrange or to assure by written agreement from the licensee (shipper or receiver) that appropriate measures for in-transit physical protection of SNM-LSS are in place; to establish and maintain response procedures for dealing with threats of theft or thefts of this material; to notify or make arrangements to be notified immediately upon I arrival of a shipment at its destination or when a shipment is lost or unaccounted for after the estimated time of arrival at its destination; to initiate an immediate trace investigation of such lost or unaccounted for shipment; and to notify the NRC Operations Center within one hour after discovery of such loss or of recovery or accounting for such lost shipment.

USEC has complied with the requirements of 9 73.67(f)(1) by providing procedures and measures for the storage and use of low enriched SNM only within a controlled acc'ess area.

USEC has complied with the requirements of 9 73.67(f)(2) by providing procedures and measures for monitoring of the controlled access areas to detect unauthorized penetrations or activities.

USEC has complied with the requirements of 9 73.67(f)(3) by establishing procedures and measures for response to unauthorized penetrations or activities using security patrols, locallaw enforcement and a communication capability.

USEC has complied with the requirements of 9 73.67(f)(4) by providing procedures and measures for response and maintaining those procedures and copies of superseded material for three years.

i USEC has complied with the requirements of 9 73.67(g)(1)(i),(ii), (iii)', and (iv) by providing l procedures and measures for advance notification prior to shipment, for confirmation of l

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arrival of a shipment, for inspection of containers and seals prior to shipment; and for use of tamper-indicating and sealed containers during transport.

USEC has complied with the requirements of 9 73.67(g)(2)(i) and (ii) by providing procedures and measures to check on the integrity of containers and seals on arrival of a shipment at destination and by notifying the shipper of receipt of the material.

USEC has committed to meet prior to shipment the requirements of 5 73.67(g)(1)(v),

73.67(g)(2)(iii), and 73.67(g)(3) by arranging for procedures and measures, or to assure by written agreement from the licensee (shipper or receiver), that provisions for in-transit physical protection have been made; that response procedures have been established and are maintained by the party responsible for arrangements for physical protection of the shipment; to verify that arrangements have been made for notification of the shipper immediately on arrival of the shipment at the destination, or for any such shipment that is lost or unaccounted for after the estimated time of arrival at its destination; and to verify that arrangements have been made to conduct a trace investigation of any shipment that is lost or unaccounted for after the estimated time of arrival; and to notify the NRC i Operations Canter within one hour after loss, recovery or accounting for such lost shipment.

1 in the Paducah Transportation Plan and the Paducah Physical Protection Plan, USEC has l committed to and has provided procedures and measures to meet requirements for physical '

protection and response for domestic shipments of SNM-LSS, both intransit and at the Paducah Gaseous Diffusion Plant, as specified in Sections 73.67 (g) (4),73.67 (g) (5), 1 73.73 and 73.74, and other applicable regulations cited and referenced in the Code of l Federal Regulations. )

USEC has chosen not to address or to commit to physical protection of export and import shipments of SNM-LSS in the Paducah Transportation Protection Plan. Therefore, USEC can not import to or export SNM-LSS from the Paducah Diffusion Plant until the NRC staff has reviewed and found acceptable an amendment to the Paducah Transportation Plan to address requirements for imports and export shipments of this material to and from this i site.  !

The staff concludes, with the exception noted above, that the Physical Protection Plan and Transportation Plan for the PGDP satisfy the performance objectives and system capabilities required by 10 CFR Part 76 and other applicable Parts cited and referenced therein, meet the regulatory requirements for physical protection of SNM-LSS both at this site and during domestic shipment to and from this site, and are, therefore, acceptable.

13.3 Classified Information The regulations in 10 CFR Part 76.60(i) require USEC to comply with the requirements of 10 CFR Part 95, " Security Facility Approval and Safeguarding of National Security Information and Restricted Data" in order to use, process, store, reproduce, transmit, transport, or handle National Security Information (NSI) and/or Restricted Data (RD) in connection with NRC-related activities. Additionally, in December 1993, the Chairman of the NRC and the Secretary of Energy signed a Joint Statement of Understanding on 113

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implementing the Energy Policy Act provisions on the regulation of gaseous diffusion l uranium enrichment plants. Paragraph No. 4 of the Joint Statement of Understanding states that " DOE will be responsible for the administrative determinations relating to granting, suspending, adjudicating, or denying a security clearance, and for reinvestigating an individual's background for continued access." USEC must also comply with the guidelines set forth in the " Joint Statement of Understanding," between DOE and the NRC. As part of the application, USEC submitted the " Security Plan for the Protection of I Classified Matter" for the Paducah facility; because of its nature, this pian is not publicly  !

available. l in August 1994, the NRC's Division of Security (SEC) accompanied DOE's Office of I Safeguards and Security on a security inspection of the Paducah gaseous diffusion plant. i During this inspection, SEC was able to observe the plant's security program for the l protection of classified matter. Other visits by SEC members were made to the plant in I fiscal year 94 as part of training courses being offered at the plant site and for general l reviews of the plant's security program, in fiscal year 95, SEC met with USEC staff on l several occasions to provide guidance and comments on a draft security plan for the  !

protection of classified matter at Paducah. On September 15,1995, as part of the Certification Application, USEC submitted a security plan for the protection of classified matter at the Paducah gaseous diffusion plant for formal review. Comments on the plan were provided to USEC on October 13,1995, and November 27,1995. During the period I of October 16-20,1995, representatives from NRC's Division of Security and NRC's Financial Management, Computer Security, and Adrninistrative Support Staff conducted a verification visit of the Paducah plant to confirm that their security plan, including computer security plans, accurately reflected the security program at the plant. The visit

• was favorable and only minor discrepancies were noted.

The NRC staff reviewed USEC's latest response to NRC's comments on the security plan for the protection of classified matter dated December 15,1995, and found it to satisfy the requirements of 10 CFR Part 95.

USEC has made commitments which meet the requirements of 10 CFR Part 95 by providing an acceptable security plan that establishes controls to ensure that classified matter is used, stored, processed, reproduced, transmitted, transported, and destroyed only under conditions that will provide adequate protection and prevent access by unauthorized persons.

Accordingly, the NRC staff concludes that the security plan for the protection of classified matter at the Paducah gaseous diffusion plant, when implemented and verified by the Commission, is acceptable in meeting the requirements of 10 CFR Part 95.

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Chapter 14 DECOMMISSIONING The regulations in 10 CFR 76.35(n) require that "a description of the funding program to be established to ensure that funds will be set aside and available for those aspects of the ultimate disposal of waste and depleted uranium, decontamination and decommissioning,"

which are the financial responsibility of USEC. The regulations further state that "The Corporation shall establish financial surety arrangements to ensure that sufficient funds will be available for the ultimate disposal of waste and depleted uranium, and decontamination and decommissioning activities which are the financial responsibility of the Corporation."

1 As part of the application USEC submitted a Decommissioning Funding Program Description. It addresses the scope of USEC's financial responsibility for decommissioning, a cost estimate and basis, and the funding mechanism.

Under the lease agreement with DOE and the Energy Policy Act, USEC is not responsible for the decontamination and decommissioning of the leased premises. DOE retains responsibility for the decommissioning, including decommissioning of any capital improvements (i.e., new buildings or equipment). USEC is financially responsible for the disposal of low-level radioactive waste and mixed waste generated by USEC, and for the cost for disposition of the depleted uranium generated from the enrichment process.

l Costs of disposal of low-level and mixed waste are considered a production cost, assuming l the waste is removed from the site during the year generated. The waste generated during decommissioning activities will not be USEC's responsibility. Currently USEC has a backlog of waste stored at the site. To come up with a cost estimate for disposal of its low-level waste, USEC utilized a weighted average cost based upon existing contract ,

prices or upon prices from contracts being re-negotiated. USEC projects that its liability on l September 30,1996 will be $.7 million. This is based on the assumption that most of the backlog will be removed from the site as of that date. Cost estimates for disposal of the mixed waste are $.1 million. Again this is based on contract prices and the assumption that much of the waste will be disposed of prior to September 30,1996. This is acceptable to the staff. USEC will re-evaluate its cost estimates in October of each year.

If the backlog is not disposed of in the time frame currently expected, adjustments to the fund will be made at that time.

The disposition of the depleted uranium tails is the major cost factor for decommissioning costs. USEC has utilized the same cost basis for disposition of depleted uranium tails as Louisiana Energy Services (LES). The total unit cost for the disposition of depleted uranium is estimated to be approximately $5.27 per kilogram of depleted uranium ($4 per kilogram of uranium for conversion to U3 0s, $1 per kilogram of U3 0, for disposal, and $.27 per kilogram for transportation.) Due to inflation, the average cost is $5.30 per kilogram. The staff finds this estimate to be acceptable. The staff does caution USEC that the cost of disposition of the depleted uranium tails for LES is an issue before the Atomic Safety and Licensing Board, if the Board rules that a higher (or lower) cost should be imposed, the staff willimpose the same cost on USEC. An additional factor in USEC's future liability is the fact that DOE will be responsible for the ultimate disposition of the depleted uranium,if DOE determines how it will disposition the tails, the staff will consider imposing that cost 115 l

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basis on USEC. USEC estimates that approximately 30,048 metric tons of depleted uranium will be generated by the Paducah operations from July 1,1993 through September 30,1996. The total cost would be $159.3 million.

USEC is currently a government corporation, and as such has submitted a statement of intent to assure funding for USEC's decommissioning liabilities. The USEC Board of Directors issued resolutions that stated that it was the Board's intent to have funds  ;

available when necessary for decommissioning activities. USEC will review the  !

decommissioning cost estimates and associated funding levels in October of each year. )

Adjustments will be made as necessary. '

USEC has agreed that at the time the corporation is privatized that it submit an executed

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sinking fund arrangement, a standby trust agreement, and a payment surety bond in ,

executed form for NRC review. This item is addressed in Compliance Plan issue 39.

The staff concludes that USEC's Decommissioning Funding Plan meets the requirements of  ;

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Chapter 15 COMPLlANCE PLAN The regulations in 10 CFR 76.35(b) require USEC to submit a " plan prepared and approved by DOE for achieving compliance with respect to any areas of noncompliance with the NRC's regulation." The plan must include a description of the areas of noncompliance, a plan of action and schedule for achieving compliance, and a justification for continued operation. As part of the application, USEC submitted a Compliance Plan that was prepared and approved by DOE. Since these items are necessary for initial compliance, the items contained in the Compliance Plan will not be subject to the backfit provisions of 576.76.

, The PGDP Compliance Plan contains 57 issues that had areas of noncompliance. A portion l

of the Compliance Plan (issues related to safeguards, issues A.1 - A.7)is considered proprietary and is not publicly available. Each issue contains sections on the requirements, i l what the commitment was, a description of the noncompliance, a justification for l continued operation, and a plan of action and schedule. Most of the noncompliance issues are discussed in other sections of this CER; the issue of the SAR upgrade, the seismic

! issue, and the transition from DOE to NRC regulation are discussed in the following i

paragraphs. Compliance Plan issues 7,12,15,18,32,33, and 34 have been completed by USEC; Issues 28,47, A.5, and A.6 were deleted.

l Compliance Plan issue 1 addresses the transition from DOE regulation to NRC regulation.

l According to issue 1, the transition to NRC regulatory oversight is scheduled to occur 120 days after the directors decision on the certification is issued. This will allow tirne for USEC to revise procedures and train the operators on the TSRs. However, by letter dated August 16,1996, USEC has requested that the 120 day period be extended such that the transition period would end on March 3,1997. The staff has no objection to the extension of the transition period. A listing of open commitments under DOE will be provided to the l

NRC on the date NRC assumes regulatory authority to assure the commitments are not lost in the transition. This approach and timing is acceptable to the staff.

In order to establish the transition period, and the effective date of the certificate, the staff recommends the following condition:

This Certificate of Compliance shall become effective on March 3,1997.

The NRC shall assume regulatory jurisdiction from the Department of Energy at 12:01 AM on March 3,1997.

Compliance Plan issue 2 addresses the Safety Analysis Report Upgrade. The SAR submitted as part of the certification application is based, in part, on the 1985 Final Safety Analysis Report (FSAR). The 1985 FSAR has a number of areas which need to be updated with respect to the description of hazards, description of plant SSCs (structures, systems, and components), human activities, and supporting safety analyses, including the following: (1) the "as-exist" plant configuration does not match plant descriptions in the  ;

I 1985 FSAR; (2) assumptions used for the initiating events need to be reevaluated; (3) the i expected response of SSCs to events may be different than previously assumed (e.g.,

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DOE is in the process of upgrading the FSAR.

The DOE site-wide safety analysis upgrade (SAR Upgrade) was initiated to address the above deficiencies, changes to the plant configuration implemented since the previous analysis was performed, and the revised safety requirements issued since development of the 1985 FSAR. From the comprehensive analysis of all credible initiating events based on "as-exist" . plant configurations, and the consequences from these accidents, the SAR Upgrade will provide more clearly the technical basis for safety boundaries (i.e., safety systems, equipment, components, etc.) and human activities relied upon to ensure safety.

In addit;on, the SAR Upgrade will provide descriptions of the various safety programs (based primarily from the certification application). This final " safety basis", derived from the SAR Upgrade, will provide the necessary safety baseline from which future plant modifications can be made pursuant to 10 CFR 76.68. Since the SAR Upgrade is required to achieve compliance with NRC requirements, any safety improvements called for by the SAR Upgrade would not be subject to the backfit provisions in 10 CFR 76.76.

Pending completion of the SAR Upgrade and achievement of complaince, the current safety basis for continued operation is the plants' implementation of, and adherence to, safety requirements in OSRs (Operational Safety Requirements) and plant procedures, developed over 100 years of combined operating experience for ths 3 GDPs. This safety l basis will continue to be utilized by the GDPs until completion of the SAR Upgrade and implementation of the approved amendment request addressing results of the SAR )

Upgrade. The TSRs, which will take effect upon NRC assuming regulatory oversight of the GDPs, are based on the safety requirements in OSRs and additional safety requirements clarified by the NRC staff during the certification process. Current safety procedures will remain in effect after NRC assumes regulatory oversight of the GDPs. These safety l procedures can be changed only in accordance with the procedure TSR which provides the necessary safety assessment and both management and technical review before a procedure is changed or replaced. TSRs cannot be changed without prior NRC approval.

The staff concludes that the combination of the use of experienced plant personnel; use of the certification SAR; implementation of, and adherence to, TSRs (which supersede the OSRs); current as well as new GDP safety procedures; and commitments to compensatory measures and/or interim regulatory commitments will ensure that the GDPs continue to operate safely until the SAR Upgrade and accompanying GDP modifications for safety are approved by the NRC and implemented at the GDPs.

Compliance Plan issue 36 addresses the seismic analysis contained in the SAR, and specifically the seismic capability of buildings C-331 and C-335. The SAR submitted as part of the application for certification, contains an evaluation of an accident, and resulting consequences, caused by an Evaluation Basis Earthquake (EBE). The EBE chosen for the analysis is an earthquake with an expected return period of 250 years. This earthquake l

produces a peak ground acceleration of 0.189 at the site. The evaluation presented the conclusion that although minor structural damage to process building's would occur, the buildings would remain intact. However, the analysis predicted that several expansion i

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joints that connect process piping across building sections would f ail. This failure was estimated to cause the release of significant quantities of UFe. Since the time this analysis was completed, the majority of the expansion joints expected to fail have been replaced.

In 1995, as part of the SAR upgrade project, DOE identified structural weaknesses in two of the four main process buildings. Structural seismic loading capacity analyses that have been performed as part of the SAR upgrade have indicated that significant structural damage could occur at peak ground acceleration levels below those for the 250 year return period. At. peak ground acceleration levels above 0.05g, buildings C-331 and C-335 could suffer collapse of 20 ft wide sections of the roof and all floor levels above ground. These sections traverse each building in three places. This potential for loss of building integrity may invalidate the analysis of the consequences of the EBE. This constituted an unreviewed safety question and indicates that the SAR submitted as part of the application for certification does not meet the requirements cited above. DOE ordered USEC to make plant modifications to improve seismic capacity, and to implement compensatory safety measures until plant modifications are complete.

DOE has prepared, and USEC has submitted, a Plan of Action and Schedule to correct this noncompliance. Buildings C-331 and C-335 will be modified to strengthen building structures and increase the seismic capacity of the floor and roof sections. These modifications are scheduled to be complete by late 1997. Documentation for the design of the modifications will be provided to the NRC for review when it is issued.

Until the building modifications are complete, USEC has committed to operating buildings C-331 and C-335 at reduced power levels that will result in operating pressures below atmospheric pressure. Operations personnel will be instructed on the specific emergency procedures for shutting down the affected enrichment cascade equipment and building ventilation systems following a seismic event. Building access will be limited to only those workers essential to operations, inspections, or the completion of the building modifications.

These factors mitigate the expected consequences of a seismic event to a level that is acceptable for the period required for building modification. An analysis of the worst case

- total building collapse and release of the entire building inventory of UF,--indicates that a person located one mile from the building might inhale 139 mg uranium. This level of inhalation would be expected to result in significant renal injury but not to be life threatening. The consequences of a seismic event that resulted in the structural damage predicted for the EBE should be far below these levels.

The staff also has questions about DOE's seismic modeling techniques, and the fact that seismic data after 1985 were not considered. Therefore, the Compliance Plan also addresses the need to conduct an updated seismic hazard analysis. The new analysis, to be submitted for NRC approval by December 1,1997,is to consider all available regional and site-specific data published by the U.S. Geological Survey and provide an estimate of the peak ground acceleration for a seismic event with a 250-year return period, it is the staff's intent to consider the results from the new analysis and the potential need for further plant modifications based on the new analysis under the provisions of 10 CFR 76.76, "Backfitting."

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i The staff concludes that the justification for continued operation, the plan of action and the I schedule are acceptable.

The staff has reviewed all of the issues for the PGDP facility and concludes the actions, schedules and justifications for continued operation are acceptable and that the Compliance Plan should be approved.

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l Chapter 16 ENVIRONMENTAL REVIEW j issuance of the Certificate of Compliance for operation of the PGDP will not have a i

significant effect on the human environment. Regulation by the NRC will not result in any environmentalimpacts beyond those previously considered by DOE in its environmental reviews and which currently exist or would be expected to continue absent NRC regulatory j oversight. Therefore, in accordance with 10 CFR 51.22(c)(19), neither an environmental assessment nor an environmental imp 6ct statement is warranted for the certification of the PGDP. This determination only applies to those aspects that are in compliance with 10 CFR Part 76.

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! An Environmental Assessment (EA) was prepared as part of the Compliance Plan and l application review process. The EA concluded that the environmental effects of approving i the Compliance Plan will be insignificant. The EA further concluded that the Compliance i Plan is sufficient to ensure that, during the interim of noncompliance, plant operations related to areas of noncompliance will not significantly affect the quality of the human l environment.

The EA resulted in a Finding of No Significant impact (FONSI). The FL,NSI has been

forwarded with the Notice of Decision for publication in the ';ederal Reaister.

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i Chapter 17 AUTHORIZATIONS AND EXEMPTIONS USEC has requested authorization to release items for unrestricted use if the surface contamination is less than limits established in the SAR. The limits are consistent with those established in the NRC's April 1993 document entitled, " Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Termination of Licenses for Byproduct, Source, or Special Nuclear Material." This request is consistent with accepted industry practice and, therefore, approval of the authorization is recomm. ended.

l USEC has requested an exemption from the requirement of 10 CFR 20.1904," Labeling l Containers," which requires that each container of certified material bears a durable, clearly i visible label. In lieu of this requirement, USEC will post a sign stating that every container j may contain radioactive material. When containers are moved from contaminated areas, a survey is performed to ensure that contamination is not spread. During movement of UFe l cylinders, which are easily identifiable, radiological workers are in attendance. This l exemption is consistent with accepted industry practice and, therefore, approval is recommended.

To approve the special authorizations and exemptions, the staff recommends the following condition:

The United States Enrichment Corporation is hereby granted the special authorizations and exemptions in Chapter 1, Section 1.8 of the Safety

! Analysis Report, Revision 5.

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Chapter 18 TERM OF CERTIFICATE The initial certificate will be issued for an affective period of approximately 2 years, with an expiration date of December 31,1998. This is consistent with the new provision in Public Law 104-134,the USEC Privatization Act, which amended Section 1701(c)(2)of l the Atomic Energy.Act replacing the requirement for an annual application for a certificate

! of compliance with the requirement for an application to be filed " periodically, as l determined by the Commission, but not less than every 5 years." The staff believes that 2 l years is a reasonable period for the first certificate of compliance, in 2 years significant progress will occur in implementing plant improvements specified in the Compliance Plan.

Also within the next 2 years, USEC must address the findings in DOE's pending upgraded safety analysis report and prepare and submit an updated seismic hazard analysis.

The requirements in 676.31 and 676.36 for an annual application were based on the I previous statutory requirement for an annual application, which has been superseded.

Therefore, the exemptions from these requirements are justified under 676.23, which  ;

specifically allows the NRC to grant such exemptions from the requirements of Part 76 as  !

it determines are authorized by law and will not endanger life, property, or the common  ;

defense, and are otherwise in the public interest. The exemptions meet these criteria. l l

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l Therefore to accommodate a 2-year certification penod, the staff recommends the  ;

l following condition to grant an exemption from the requirement to submit an annual J l

application for certificate renewalin 1997: )

The United States Enrichment Corporation is hereby granted an exemption from the requirements in 10 CFR 76.31 and 76.36 requiring submittal of a renewal application for the year 1997. The next renewal application shall be filed by April 15,1998.

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A Chapter 19 CONCLUSIONS Upon completing the compliance evaluation of USEC's application, including the SAR, TSRs, program plans and Compliance Plan, the staff concludes that there is reasonable l

assurance that the plant will continue to be operated such that public health and safety will j be adequately protected, and that the common defense and security will not be l endangered. Furthermore, the staff determined that the application fulfills the requirements '

of 10 CFR Part 76. The staff recommends that USEC be issued a Certificate of Complianc.e in accordance with statements and representations contained in the SAR, l program plans, and TSRs. The staff further recommends approval of the Compliance Plan.

The staff recommends that the following conditions be part of the certification:  !

The United States Enrichment Corporation shall conduct its operations in accordance with the statements and representations contained in the Certification Application dated September 15,1995, and revisions dated January 19,1996, May 31,1996, July 26,1996, August 1,1996, and August 12,1996 and in the Compliance Plan dated July 12,1996, July 15, 1996, and July 18,1996, and revision dated August 1,1996.

The United States Enrichment Corporation shall conduct its operations in accordance with the Technical Safety Requirements that are contained in Volume 4, Revision 5 of the Application dated August 1,1996, as modified by Revision 6 of the application dated August 12,1996. Changes to the Technical Safety Requirements shall require NRC approval prior to implementation.

This Certificate of Compliance shall become effective on March 3,1997.

The NRC shall assume regulatory jurisdiction from the Department of Energy at 12:01 AM on March 3,1997.

The United States Enrichment Corporation is hereby granted the special authorizations and exemptions in Chapter 1, Section 1.8 of the Safety Analysis Report, Revision 5.

The United States Enrichment Corporation is hereby granted an exemption from the requirements in 10 CFR 76.31 and 76.36 requiring submittal of a renewal application for the year 1997. The next renewal application shall be '

filed by April 15,1998.

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l Chapter 20 ACRONYMS AND ABBREVIATIONS l ACR area control room l AFFF aqueous film-forming foam i AHJ Authority Having Jurisdiction '

AlHA American Industrial Hygiene Association ALARA As Low As Reasonably Achievable ALI annual limit on intake '

Am americium AMAD activity median aerodynamic diameter l ANS American Nuclear Society l ANSI American National Standards Institute i ARA airborne radioactivity area ASME American Society of Mechanical Engineers Be beryllium C Celsius CA contamination area CAAS criticality accident alarm system CCF central control f acility CCZ contamination control zone CEDE committed effective dose equivalent CER compliance evaluation report CFR Code of Federal Regulations Ci curie Cl chlorine -

CIF 3 chlorine trifluoride cm centimeter j DAC derived air concentration i DOE Department of Energy I dpm disintegrations per minute j EA environmental assessment EBE evaluation basis earthquake EOC Emergency Operations Center EPA Environmental Protection Agency ERPG emergency response planning guidelines F fluorine  !

F Fahrenheit FNMC Fundamental Nuclear Material Control FONSI finding of no significant impact F/S freezer / sublimer FS Fire Services FSAR final safety analysis report ft 2 square feet g gram g acceleration of gravity .

gaseous diffusion plant GDP HCA high contamination area HEPA high efficiency particulate air 125

HF hydrogen fluoride HP health physics HPFWS high pressure fire water system HRA high radiation area ICRP International Commission on Radiological Protection in inch JCO justification for continued operation k ,, criticality constant, effective km kilometer KPDES Kentucky Pollutant Discharge Elimination System I liter Ib pound Ibs/ min pounds per minute LCO limiting condition of operation LCS limiting control setting LES Louisiana Energy Services ifpm linear foot per minute LMUS Lockheed Martin Utility Services, Inc.

m meter m3 cubic meter m/s meter per second MAWP maximum allowable working pressure mg milligram MOU Memorandum of Understanding mrem millirem mrem /hr millirem per hour MW megawatt NCRP National Council on Radiation Protection NCS nuclear criticality safety NCSA Nuclear Criticality Safety Approval NCSE Nuclear Criticality Safety Evaluation NFPA National Fire Protection Association NIST National Institute of Standards and Technology Np neptunium NRC Nuclear Regulatory Commission NSI National Security Information OJT on-the-job training OSHA Occupational Safety and Health Administration OSR operational safety requirement PA public address pCi/g picocurie per gram PGDP Paducah Gaseous Diffusion Plant PHA process hazard analysis PNL Pacific Northwest Laboratory PORC Plant Operating Review Committee psia pounds pressure per square inch absolute '

psig pounds pressure per square inch gauge PSM process safety management 126

PSS Plant Shift Superintendent Pu plutonium QA quality assurance QAP Quality Assurance Program .

RA radiation area rad radiation absorbed dose RCRA Resource Conservation and Recovery Act RCW recirculating cooling water RD restricted data rem Roentgen equivalent man I RMA radiological material area  !

RMDC Records Management and Document Control  !

RP radiation protection  !

RWMP Radioactive Waste Management Plan RWP radiological work permit SAR safety analysis report SAT systems approach to training '

SFWS sanitary and fire water system SNM special nuclear material SNM-LSS SNM of low strategic significance SSC structure, system or component SWU separative work unit Tc technetium TEDE total effective dose equivalent l Th thorium l TLD thermoluminescent dosimeter TSR technical safety requirement TVA Tennessee Valley Authority U uranium U F, uranium hexafluoride UO2F2 uranyl fluoride U3 0, triuranium octoxide USEC United States Enrichment Corporation WKWMA West Kentucky Wildlife Management Area w/o weight per cent WSTS Westinghouse Standard Technical Specifications 127

I Chapter 21 REFERENCES l

American Industrial Hygiene Association (AlHA), " Emergency Response Planning l Guidelines: Hydrogen Fluoride," AlHA, Akron, OH,1988.

I American Society of Mechanical Engineers (ASME), " Quality Assurance Program Requirements for Nuclear Facilities," ASME NOA-1-1989, ASME, New York, NY,1989.

American Society of Mechanical Engineers, " Testing of Nuclear Air Treatment Systems,"

ASME N510-1989, ASME, New York, NY,1989.

Fisher, D.R., M.J. Swint, and R.L. Kathren, " Evaluation of Health Effects in Sequoyah Fuels Corporation Workers from Accidental Exposure to Uranium Hexafluoride," PNL-7328 (NUREG/CR-5566), Pacific Northwest Laboratory, Richland, WA, May 1990.

Fisher, D.R., T.E. Hui, M. Yurconic, J.R. Johnson, " Uranium Hexafluoride Public Risk,"

PNL-10065, Pacific Northwest Laboratory, Richland, WA, August 1994.

International Atomic Energy Agency (IAEA), " Monitoring Contamination on Surfaces,"

Technical Report Series No.120, IAEA, Vienna, Austria,1970.

International Commission on Radiological Protection (ICRP), " Recommendations of the Commission on Radiological Protection," ICRP Publication No. 26, Pergamon Press, Oxford, England,1977.

International Commission on Radiological Protection (ICRP), " Limits for intake of Radionuclides by Workers," ICRP Publication No. 30, Pergamon Press, Oxford, England, 1979.

Lockheed Martin Utility Services, Inc., " Validation of the Paducah Gaseous Diffusion Plant Nuclear Criticality Safety Code System for the ENDF/B-IV 27 Group Cross Sections," KY/S-221, Revision 1, LMUS, Paducah, KY, January 1996.

Martin Marietta Energy Systems, Inc (MMES), " Final Safety Analysis Report for the Paducah Gaseous Diffusion Plant," (KY-734), MMES, Paducah, KY, March 1985.

Martin Marietta Energy Systems, Inc (MMES), "Paducah Gaseous Diffusion Plant Environmental Report for 1992," ES/ESH-36 (KY/E-164), MMES, September 1993.

Moran, B.W., et al., " Recommendations to the NRC on Acceptable Content for the Fundamental Nuclear Material Control (FNMC) Plan Required for Low-Enriched Uranium Enrichment Facilities," NUREG/CR-5734, Oak Ridge National Laboratory, Oak Ridge, TN, November 1991.

National Council on Radiation Protection (NCRP), " Calibration of Survey instruments Used in Radiation Protection for the Assessment of lonizing Radiation Field ~s and Radioactive Surface Contamination," NCRP-112, National Council and Radiation Protection and Measurements, Bethesda, MD, December 31,1991.

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United States Enrichment Corporation,1996, Letter from G.P. Rifakes, USEC to C.J.

Paperiello, USNRC, " Lease Agreement Between the United States Department of Energy and the United States Enrichment Corporation," dated July 1,1993, April 15,1996.

U.S. Nuclear Regulatory Commission, " Health Phsics Surveys During Enriched Uranium-235 Processing and Fuel Fabrication," U.S. NRC Regulatory Guide 8.24, U.S. NRC, Washington DC, October 1979.

U.S. Nuclear Regulatory Commission (1983a), " Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," U.S. NRC  !

Regulatory Guide 1.145, U.S. NRC, Washington DC, February 1983.  !

l U.S. Nuclear Regulatory Commission, " Standard Format and Content for a Physical Security Plan for Protection of Special Nuclear Material of Moderate or Low Strategic Significance," U.S. NRC Regulatory Guide 5.59 (Revision 1), U.S. NRC, Washington, DC,

! February 1983.

U.S. Nuclear Regulatory Commission, " Chemical Toxicity of Uranium Hexafluoride  !

Compared to Acute Effects of Radiation," NUREG-1391, U.S. NRC, Washington DC, February 1991.

l U.S. Nuclear Regulatory Commission, " Guidelines for Decontamination of Facilities and  ;

Equipment Prior to Release for Unrestricted Use or Termination of Licenses for Byproduct,  !

Source, or Special Nuclear Material," U.S. NRC, Washington D.C., April 1993.

i U.S. Nuclear Regulatory Commission, " Material Control and Accounting for Uranium j Enrichment Facilities Authorized to Produce Special Nuclear Material of Low Strategic l Significance," U.S. NRC Regulatory Guide 5.67, U.S. NRC, Washington, DC, December j l 1993. j s

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Appendix A PUBLIC COMMENTS AND NRC STAFF RESPONSES l

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d PUBLIC COMMENTS AND NRC STAFF RESPONSES Introduction On Septernber 15,1995, the NRC received an application for certification from the U.S.

Enrichment Corporation (USEC) for the initial certification of the Paducah Gaseous Diffusion Plant (GDP) located near Paducah, Kentucky and the Portsmouth GDP located near Piketon, Ohio. A 45-day public comment period was provided which ended on November 6,1995. On November 6.1995 the NRC receive.; from USEC a compliance plan addressing the areas at the plants that were not yet in full compliance with 10 CFR Part 76. The 45 day public comment period for the compliance plan ended on December 29,1995. The NRC received a total of eleven (11) letters cornmenting on the application-or compliance plan. Copies of the public comments received are available for public inspection and copying at the Commission's Public Document Room (PDR)in the Gelman Building,2120 L Street, NW., Washington, DC 20555 and in the Local Public Document Rooms (LPDR) established for these facilities. The LPDR for the Paducah plant is located at the Paducah Public Library,555 Washington Street, Paducah, Kentucky 42003. The LPDR for the Portsmouth plant is located at the Portsmouth Public Library,1220 Gallia Street, Portsmouth, Ohio 45662.

The NRC also held public meetings near each site concerning the application for certification and the compliance plans for the Portsmouth and Paducah GDPs. These meetings were held to solicit public input on the initial certification of these facilities.

Transcripts of these meetings are also available at the PDR and LPDRs.

The following companies or individuals provided comments by letter or at the public meetings:

Comments Received by Letter Central Midwest Interstate Low-Level Radioactive Waste Commission Coalition for Health Concern Mark Donham and Kristi Hanson Environmental Protection Agency (EPA), Region 4 EPA, Region 5 Ronald Lamb Nuclear Energy Institute (NEI)

Occupational Safety and Health Administration (OSHA), Cincinnati, Ohio OSHA, Salt Lake City, Utah Craig Rhodes Velma M. Shearer A-2 I

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Comments Received at the Public Meetinas Portsmouth Public Meeting - November 28,1995 Speakers:

l Ms. Velma Shearer j Ms. Vina Colley

. Mr. Tim Mitchell Mr. Gerald Wilkins Paducah Public Meeting - December 5,1995 i

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Speakers:

Ms. Jotilley Dortch Mr. Ronald Lamb l Mr. A.B. Puckett Mr. Mark Donham The following is a listing of the comments received followed by the NRC staff response.

1. Commenter: Central Midwest interstate Low-Level Radioactive Waste Commission Date of Letter: December 19,1995

Reference:

Paducah Application 1.A. There are inconsistencies in the tables identifying low-level radioactive waste and mixed waste categories. USEC did not explain how they arrived at their projections for  ;

mixed waste, and their projections do not appear to include low-level radioactive waste l generated as a result of management of depleted uranium or radioactive tails that are generated during the enrichment process.

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Response

The tables are not inconsistent. Nine waste streams are described as the principal streams; each of those waste streams consists of one or more waste categories. Although some individual categories have projected no generation of waste, all nine of the principal waste streams have an estimated volume. The basis for the 60,000 ft' and the 860 ft8 estimates is contained in Tables 1 and 2. It is correct that the estimates do not contain the depleted UF, tails. The depleted tails are covered by the Depleted Uranium Management Plan.

1.B. The application does not indicate that information concerning off-site shipment is entered into the computer-based tracking system used for tracking movement of waste.

Response

Waste is tracked from the point of origin to final disposal in the waste tracking system.

f-inal treatment and/or disposal is entered into the database for the tracking system.

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J l 1.C. The application allows wastes to be released if the total activity is less than 30 pCi/g for either a solid or liquid.

Response

This has been deleted from the application.

1.D. Is it correct that no treatment wastes are shipped directly to DOE facilities for disposal since section 6.6 indicates that wastes can be disposed of at DOE facilities?

Response

l Although treatment residues are not shipped directly to DOE for disposal, this does not prohibit disposal at DOE facilities.

l 1.E. The application does not provide information concerning disposal volumes or activities l and does not indicate that disposal is in compliance with requirements.

Response

Waste volumes are provided in Section 3. USEC is responsible for packaging and transporting the waste in accordance with the regulations and the requirements of the disposal site. The facility receiving the wastes for disposal is responsible for the actual disposal. Onsite disposal is not authorized in the NRC regulated areas.

1.F. It is not clear in the application who operates the waste tracking system. Also, not l enough detail is provided concerning the identification number that is assigned to each l

waste container.

Response

The Waste Management Group Manager is in charge of the waste tracking system. A l

unique identification number is assigned to the waste upon completion of the request for l disposal form.

1.G. The Compliance Plan should indicate when USEC will begin implementing a quality j control program. (This comment actually refers to the Radioactive Waste Management Program.)

Response: l The Radioactive Waste Management Program merely indicates that some areas of the plan are not being met. The specifics are contained in the Compliance Plan. The Compliance Plan contains the detailed information on the items of noncompliance; it includes a justification for continued operations and a plan of action that provides the completion dates.

2. Commenter: Coalition for Health Concern Date of Letter: December 28,1995

Reference:

Paducah Application l 2.A. The Paducah plant has contaminated the site and surrounding area and should no longer operate.

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Response- I The Paducah plant is in compliance with applicable regulations and is operated in a manner that should assure the public health and safety and protection of the environment.

l Therefore, there is no reason to shut down the facility. The cleanup activities at the site will continue under DOE and will not be under NRC jurisdiction.  :

2.B. Seismic activity is minimized in the application.  ;

Response:.

USEC and DOE have identified weaknesses in the seismic analysis portion of the application. DOE is in the process of completing an upgraded Safety Analysis Report that will address these weaknesses. The Compliance Plan commits USEC to strengthen Paducah plant structures and increase the capacity to withstand a seismic event. Until the modifications are complete, USEC will be required to operate the Paducah plant at a reduced level of production that limits internal pressures to below atmospheric pressure.

This reduced pressure willlimit the possible consequences of a release of hazardous l materialin the event of an earthquake. I l

2.C. Nuclear criticality has not been adequately addressed.

l Response: '

The staff has reviewed the nuclear criticality safety program and found it to be acceptable.

The commenter was not specific as to the exact nature of the concern.  ;

2.D. Contamination from the contents of 6000 drums of radioactive waste has not been addressed.

l Response: i it is not clear what waste the commenter is referring to; the staff assumes the commenter  !

is referring to the depleted uranium tails cylinders that are stored at the site. DOE retains l responsibility for all tails and for any other waste that was generated prior to July 1,1993 when USEC took over operation of the enrichment facility. USEC has a radioactive waste management plan and a depleted uranium management plan that describes how the facility deals with its waste.

3. Commenter: Mark Donham Kristi Hanson Date of Letter: December 22,1995

Reference:

Paducah Application 3.A. Cumulative effects from past operations should be included in the accident assessment.

Response

Cumulative effects from past operations are not part of an accident analysis. The primary hazard of this facility is the inadvertent release of UFe; the pathway of concern is inhalation. Exposure due to accumulation in the environment would be very small.

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i l 3.B. Accident scenarios should be considered which might result in large releases of l hazardous material.

Response

The staff has reviewed the USEC accident analysis and found that it does adequately address potential accidental releases.

3.C. What is the plan for the 6,000 stored cylinders of depleted uranium?

Response

Most of the cylinders stored at the facility are the responsibility of DOE; only the tails cylinders filled after July 1,1993 are the responsibility of USEC. USEC plans to continue to store its cylinders at the site, pending development of a disposal solution by DOE.

l l 3.D. The offsite risk of a criticality accident is not addressed.

- Response:

The criticality safety program has been adequately described in the application. A criticality accident is not discussed in the context of an offsite analysis because the impact of a criticality is very localized; no significant impact would be expected offsite.

3.E. How much plutonium is at the plant?

Response

At one time recycled uranium (containing some radioactive impurities) was used as feed 1

material for the enrichment process. Contamination in piping and equipment still remains.

The very small quantity of plutonium which exists as contamination, and 0.5 Ci as calibration sources and lab chemicals, are the only plutonium which USEC will be l

authorized to possess. Recycled uranium is no longer used as a feed material.

l 3.F. Liquid effluents are not adequately addressed.

Response.

USEC does sample the surface streams that receive effluent and runoff from the facility.

The monitoring program is described in section 5.1 of the safety analysis report.

3.G. The earthquake risk is not adequately addressed.

l l Response:

l See response to 2.B.

3.H. Information on production of depleted uranium should be made public.

Response

The waste management plans are contained in volume 3 of the applidation and have not been withheld from public disclosure. The actual volume of tails generated was withheld from the original application at the request of USEC; this information is now publicly A-6 l - . -

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available. The information on production of depleted uranium tails has been made publicly  ;

i available in the Public Document Room, and the Local Public Document Room for the  !

l Paducah plant.

1 3.1. The emergency plan is inadequate. Are local authorities trained for nuclear situations? l Are the hospitals equipped to handle radiation or chemical poisoning? Public warning sirens that can warn the public cannot perform this service, and there are no maps or evacuation routes depending on wind direction.

Response

The emergency plan for the Paducah plant meets the requirements in 10 CFR 76.91,

" Emergency planning." Offsite response organizations are invited to participate in the biennial exercises that test all or most of the basic elements within the emergency plan. If offsite fire fighting assistance is needed to fight a fire involving radiological / hazardous materials, knowledgeable members of the plant emergency response organization provide radiological /toxicologicalinformation and assistance. The hospitals that have agreed to provide assistance in the event of an emergency are equipped to handle contaminated injured individuals. The plant may also provide radiation protection personnel from the plant to assist the hospital. Both hospitals have Nuclear Medicine and Radiography departments. The existing public warning sirens do not provide total coverage of the immediate notification area; however, plans are in place to replace the existing sirens with a new siren system to ensure better coverage of the immediate notification area. The scheduled completion date for the new system to be installed and operational is March 15,

• 1997. An emergency broadcast system message is an alternate method to provide public j warning and is currently available and functioning. It should be noted, however, that USEC is not required by Part 76.91, " Emergency planning," to have a public waming system and that the inclusion of the public warning system in the emergency plan for the Paducah j plant was a voluntary decision made by USEC. No scenario has been identified that would i require evacuation of members of the public.

4. Commenter: Environmental Protection Agency (EPA)

Region 4, Atlanta, GA Date of Letter: December 7,1995

Reference:

Paducah Application Comment:

The application appears to reflect compliance status with the various permits and regulations under EPA jurisdiction.

Response: No response needed.

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5. Commenter: Environmental Protection Agency Region 5, Chicago, IL Date of Letter: February 29,1996

Reference:

Portsmouth Application 5.A. The application is documented to meet the requirements under the Clean Air Act and National Emission Standards for Hazardous Air Pollutants for Radionuclides (NESHAPs).

The Compliance Plan seems to satisfy the NESHAPs requirements.

Response: No response needed.

5.B. If it is true that the ambient air pressure is greater in the storage drum room than on the process operating floors, then any leak in the storage drum room could potentially leak to the lower pressure area of the process operating floors. This is not what Section 3.1.1.6.8.8, Secondary Confinement System, on page 3.1-78, seems to be stating.

Response

It is true that in the event of a leak from a cylinder, there would be a tendency for the l released gas to flow from the area of higher pressure (the storage drum room) to the area i

of lower pressure (the process operating floor). In the event of a leak, features to protect safety of the workers on the process floor include: (1) the drum storage room is well sealed; (2) an alarm would sound, alerting operators to prepare to move to a safe location; and (3) fans in the vicinity of the entrance to the drum storage room would tend to l disperse the gas before it could escape from the drum storage room. The staff concludes l that the combination of features provides adequate worker protection against harm from leaks in the drum storage room.

t 5.C. The first paragraph on page 3.1-87 appears to indicate that any slings with test dates less than one year old need certification, while those that would have tests dates over one year old were fine.

Response

The commitment in the application is acceptable as written. The commenter misinterpreted the commitment.

6. Commenter: Ronald Lamb, Kevil, KY Date of Letter: December 28,1995

Reference:

Paducah Application l

6.A. The Paducah plant does not have a sufficient public warning system nor an effective i evacuation plan, l

l Response:

See response to 3.1.

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l 6.B. In a catastrophic emergency there may not be sufficient electricity to safely shut the operation down which could possibly lead to a criticality.

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! Response:

All plant structures, systems and equipment are designed to fail safely in the event of a i loss of power and therefore would present no immediate hazard. There would be no out l l leakage of UFe gas because UF e gas in the diffusion process would return to a l subatmospheric pressure and a criticality is very unlikely under those process conditions.

I Neither the Portsmouth or Paducah plants have ever sustained a complete loss of offsite I power.

l 6.C. There is an unresolved safety question regarding seismic risk.

Response: 1 See response to 2.B.  !

6.D. NRC is understaffed at the plant and civil penalties should be levied. '

Response

Two resident inspectors onsite is consistent with what NRC has at nuclear power plants.

There are also inspectors at the NRC regional office and at headquarters that inspect the facility periodically. Originally the NRC did not have the authority to levy fines, however on April 26,1996, President Clinton signed into law H.R. 3019, legislation which includes

! a subchapter entitled the "USEC Privatization Act." This legislation includes several provisions affecting regulation of USEC by NRC. One of these provides authorization for NRC to impose civil penalties on USEC or its successor for failure to comply with j regulatory requirements governing the gaseous diffusion plants.

6.E. USEC should not be allowed to keep the amount of waste produced at the plant confidential.

Response

See response to 3.H.

7. Commenter: Nuclear Energy Institute (NEI)

Date of Letter: November 6,1995

Reference:

Paducah and Portsmouth Applications Comment:

NRC's regulatory approach should be performance-based rather than prescriptive. Some of the NRC questions submitted to USEC indicate an overly prescriptive approach.

Response

The staff agrees that, to the extent feasible, performance-based regulation is appropriate.

However, for these plants, which have been operating for many years, the safety basis for operation is not well-defined in all cases. Therefore, the staff has required USEC l commitments to many existing safety procedures established over time by the Department of Energy. This has resulted in many prescriptive requirements, but the staff believes that ;

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8. Commenter: Occupational Safety and Health Administration Date of Letter: November 24,1995

Reference:

Paducah and Portsmouth Applications 8.A. When design changes affect general operating procedures, employees should be apprised of the changes and related training should be updated.

Response

This is currently included in the training program in SAR Chapter 6.6.  !

8.B. Threshold limits in 29 CFR 1910.119 for hazardous materials should be adopted.

Response

NRC does not have the authority to require USEC to comply with OSHA regulations.

8.C. Reference Paducah SAR, Emergency Plan, Page 1-3. Have less hazardous materials  !

other than chlorine been considered for water treatment process, such as hypochlorite

. solution or ozone?

Response: l Other water treatment process has been considered. A bromate treatment process is being l tested at Portsmouth. Depending upon its result, this process might be applied to l Paducah. l l

I 8.D. Reference Paducah Quality Assurance Program, section 2.7.3.6. For items which

• were obtained from an internal parts or equipment source, preinstallation inspection should be required.

Response: The "preinstallation" inspection is part of the receiving inspection.

8.E. Reference Paducah Technical Safetv Requiremer.t (TSR), section 1.2.4. The statement "No training is required to perform the fire patrol" should be deleted.

Response

USEC has deleted this statement.

8.F. Reference Paducah TSR, section 2.1.2.2. There is a potential for cylinder rupture during movement or handling, such as rolling, tilting, or suspending. Therefore, the first sentence is not inclusive and should be amended.

Response

The statement applies to cylinders that are not being mo /ed, and is appropriate in that context.

8.G. Reference Paducah TSR, section 2.2.3.1, Basis. The Basis stat'es that the test parameters were approved in December,1980. The safety committee should update or reevaluate test parameters.

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USEC has revised and updated the Basis statement.

8.H. Reference Paducah TSR, section 2.3.3.1, Basis, is there a pressure- l limiting cut-off switch or sensor near the discharge expansion joint that would avoid the over-pressurization possibility which is described?

Response

As described in Basis, dual high-pressure shutdown instrumentation is installed on the pump discharge pipe to prevent this type of accident.

8.1. In Paducah TSR, section 2.3.4.1, Condition C., a continuous smoke watch is initiated, it also should describe actions to be taken if smoke is observed.

Response

Actions to be taken when smoke is observed are described in detail in plant implementing procedures.

8.J. Reference Paducah TSR, section 2.3.4.1, Conditions G and H. The action instructions are incomplete.

Response: USEC has revised and completed the instructions.

8.K. Unplanned medical treatment at a medical facility of an individual is described in Paducah TSR Section 3.8. It should also describe what capability is available if a larger number of people need to be treated.

Response

This section of the TSR describes special reporting requirements. An event that requires unplanned medical treatment at a medical facility of an individual with-radioactive contamination on the individual's clothing or body must be reported to the NRC. USEC is not required to describe medical capabilities for different situations.

8.L. There is lack of employeas' involvement or participation in safety and health processes of various program developments.

]

Response-Many of the programs (Nuclear Safety Programs, Quality Assurance Program, etc.)

described in the application are part of USEC's management control program. Participation and involvement in these programs require some degree of experience and expertise which may not be possessed by non-supervisory employees. While program development is a management responsibility, it is anticipated that employees from all segments of the plant staff will be involved in program development as appropriate. The As Low As Reasonably l Achievable (ALARA) subcommittee has a representative from the Oil, Chemical & Atomic l workers bargaining unit. .

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9. Commenter: Occupational Safety and Health Administration Date of Letter: March 8,1996

Reference:

Paducah and Portsmouth Compliance Plans 9.A. Describe distinctions between small fires and major fires in order to know how to respond to each.

Response: This section describes available features for responding to small and major fires. Distinctions for the purpose of responding to fires are not necessary here.

9.B. Is there a procedure for closing out noncompliance items c'ompleted?

Response

For completed Compliance Plan (CP) items, DOE will conduct closure inspections to ensure that each item has been completed as committed. DOE's findings will be documented in its Site Safety Representative's inspection reports, and the CP will be updated to reflect the proper completion / closure. This procedure is documented in a letter from DOE to USEC, dated May 22,1996.

9.C. Reference Paducah CP issue 6, Commitments,5.2.2.6 Procedure Requirements.

Operation procedures should be written to cover not only normal, emergency, and temporary operations but also for start up following a turnaround and after an emergency shutdown.

Response

The section to which the comment applies is limited to nuclear criticality controls that are contained in procedures. Operational procedures are covered in other areas of the application. USEC does have emergency procedures.

9.D. Reference Paducah CP lssue 10/Portsmouth CP lssue 13, Commitments, item

2. The statement " Containers located within radiological areas are not required to be labeled." should be explained further even when considering item 3 and Description of Noncompliance.

Response

Section 5.3.1.7 of the application states that containers located in Radiological Areas within USEC leased areas are not individually labeled but instead are posted with signs that state any container may contain radioactive material.

9.E. Reference Portsmouth CP issue 25, Justification for Continued Operation, in addition to normal operating procedures, emergency operating procedures, and others listed, current operations program should include emergency shutdown procedures and start up following i a turnaround or emergency shutdown.  !

Response: .

The Compliance Plan focuses on operating procedures. Emergency and shutdown procedures are covered by the application.

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9.F. Reference Portsmouth CP. Is there involvement of employees in various safety related aspects of the operations including operating procedures, training, and emergency j response? l Response: j Employees are involved in safety related aspects of operations. See also response to i comment 8.L. l

10. Commenter: Craig Rhodes, Brookport, Illinois l Date of Letter: November 14,1995

Reference:

Paducah Application 10.A. The assumptions regarding an earthquake risk at the Paducah plant are too low and should be more detailed.

Response

See response to 2.B.

10.B. The plant lacks warning sirens in a two mile radius, there is no evacuation plan and the notification plan is inadequate.

Response

See response to 3.1.

10.C. Hazardous material cleanup is inadequate.

Response: The Paducah Gaseous Diffusion Plant is regulated by EPA and must comply with applicable EPA regulations.

11. Commenter: Rev. Dr. Velma M. Shearer Date of Letter: December 29,1995 )

Referer.ce: Portsmouth Compliance Plan l 11.A. Containers with radiological contents should indicate the amount in the container.

I Response -

In all restricted areas where containers may contain radioactive material, USEC is committed to posting caution signs stating that every container may contain radioactive material. USEC is also committed to provide appropriate safety training related to working in the vicinity and in handling of containers that may contain radioactive material, to all unescorted individuals entering restricted areas. In unrestricted areas,10 CFR 20.1904 requires USEC to label all containers containing radioactive material with suffici:nt information, including the radionuclides present and their quantities,,to permit wori.ers to j take precautions to minimize or avoid radiation exposures.

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11.B. USEC should monitor individuals at shift closing and provide individuals with copies of dose monitoring results upon completion of reports.

Response

USEC is required to provide radiation dose monitoring results to its workers and monitor them in accordance with 10 CFR 19 and 10 CFR 20, respectively.

11.C. Reference CP, issue 21, " Management Controls," the Protection of national security interests is noted as a management responsibility. However, there is no indication as to who makes national security interest decisions. Will Lockheed-Martin Utility Services have jurisdiction to make political decisions regarding world-wide sales of enriched uranium?

Response

USEC is bound by regulatory requirements to provide adequate security and safeguards for information, material, and equipment that requires protection to safeguard national security interests. Decisions regarding implementation of USEC's safeguards and security programs are made at appropriate levels in USEC's chain of command, in accordance with applicab;e NRC requirements. USEC is free, to the extent provided by law and regulation, to continue to provide enrichment services on a world-wide basis.

11.D. There is a comprehensive " flow-down" system of authority. However, no plan within the system is in place for a " flow-up" from any worker who may be able to contribute improvements to the enrichment or system process.

Response: ,

The CP issue that is the subject of this comment relates to having an appropriate organizational structure and effective management control to assure that all safety, safeguards and security requirements are met. This willinclude responding to safety problems identified by workers, it does not require a plan for use of beneficial ideas and suggestions from employees which are not related to safety.

11.E. A 500 mrem per year exposure limit for individual workers is recommended.

Response:  ;

The annual whole body committed effective radiological dose limit contained in 10 CFR Part 20 for workers is 5 rem. In addition,10 CFR Part 20 does not require monitoring if  ;

the annual dose is not expected to exceed 10 percent of the limit, which is 500 mrem.

Also the radiological dose hazard at the Portsmouth plant is not significant. This is apparent based on monitoring data, for the years 1992,1993 and 1994, during which a total of five individuals out of over 11,700 monitored for external radiation, and none out of 5,000 monitored for internal radiation at the Portsmouth plant, may have exceeded this monitoring threshold level.

11.F. Procedures need to be developed to allow workers to report such things as a faulty valve to prevent spillage or serious exposures.

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Response

j There is a problem reporting program by which workers can report items like a faulty valve.

Some of the minimum procedural requirements by which faulty valves to prevent spillage or serious exposures will be identified and reported are in the areas of internal audits and inspections, investigations and reporting, quality assurance, equipment control (lockout /tagout), preventive maintenance, etc.

l 11.G. The need for Atomic Vapor Laser Isotope Separation (AVLIS), which uses laser beams for.the uranium enrichment process, at the Portsmouth plant is questionable.

l Response:

The NRC has not received an application to license an AVLIS facility . This comment is outside the scope of the certification review.

Comments from Transcriots of Public Meetinas

12. Portsmouth Public Meeting Transcript - November 28,1995 Commenter: Ms. Velma Shearer 12.A. Who is responsible for depleted uranium and low level mixed waste at the Portsmouth plant? Who is responsible for the decontamination and decommissioning costs at the plant.

Response

The President signed into law on April 26,1996, Pub.L.No. 104-134,110 Stat.1321 (1996). Title ll1 of this Act, namely the "USEC Privatization Act," among other things, deals with the disposal of low-level radioactive waste (LLRW), including depleted uranium, if it were ultimately determined to be low-level radioactive waste, to be generated by facilities regulated by the NRC. According to this Act, DOE is required to take possession  ;

of, at the request of the generator, and p. ovide for the ultimate disposition of all such material.

DOE is responsible for decontamination and decommissioning costs once the plant is returned to them.

12.B. Will USEC have to pay the disposal costs for its depleted uranium and other waste?

Response

According to the Privatization Act, USEC is required to reimburse DOE for disposal of low I level radioactive waste (LLRW) and depleted UF, (DUF6)-if it is determined to be LLRW -

in an amount equal to the costs incurred by DOE, including a pro-rata share of any capital  !

costs.

12.C. Who will receive and manage the radioactive and mixed waste?

! Response:

See response to 12.A. and 12.B.

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s 12.D. Who pays for decommissioning? It is unfair for the taxpayers to pay.  ;

1 Response: l DOE will pay for decommissioning. Current law requires DOE to pay any costs of decontamination and decommissioning (D&D) with respect to conditions existing before ,

July 1,1993. D&D of any supplemental contamination resulting from accidental spills, I leaks, releases, etc. after July 1,1993, would be the responsibility of USEC. I

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l 12.E. What is the anticipated market for enriched uranium?

1 Response.

NRC's mission does not require it to maintain current information on the anticipated market i for enriched uranium, and that question is not part of NRC's safety, safeguards or security 1 evaluations for the gaseous diffusion plants.

12.F. Two resident inspectors at the plant does not seem adequate. l

Response

See response to 6.D.

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12.G. Are NRC's health and safety standards available to the public?

Response

Yes. NRC's health and safety standards are contained in Chapter i of Title 10 of the Code of Federal Regulations (CFR). Specifically for the two operating GDPs, NRC's health and safety standards are contained in 10 CFR Part 76. ,

12.H. A copy of the USEC proposed compliance plan should have been distributed at the i public meeting.

Response

Due to the size of the Compliance Plan,it was not distributed at the public meetings. It is publicly available at the Public Docuruent Room (PDR),2120 L Street, N.W., Washington, D.C., and the Local Public Document Room (LPDR), Portsmouth Public Library,1220 Gallia Street, Portsmouth, Ohio. The final versions of the Compliance Plan are also available at these locations.

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12.1. How will NRC regulate AVLIS?

Response: See response to 11.G.

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l Commenter: Ms. Vina Colley i

12.J. Is there plutonium at the Portsmouth plant? Are there fluoride gas releases at the

! Portsmouth plant?

Response

l Plutonium is present in small amounts as contamination (from past processing of recycled uranium) and in laboratory sources.

There are fluoride gas releases at the Portsmouth plant. First, there are planned releases of fluoride gases up any of 16 stacks under carefully measured and controlled conditions.

These releases are regulated by the EPA. Additionally, there are unplanned fluoride gas releases from time to time as a consequence of accidents, human error, or component failure.

12.K. All classified, secret, and proprietary documents associated with the plant should be released to the public.

Response

Information may be classified and specifically exempted from disclosure by statute (e.g.,

the Atomic Energy Act of 1954, as amended) or authorized under criteria established by an Executive Order to be kept secret in the interest of national defense or foreign policy.

Section 2.790 (d) of the Code of Federal Regulations provides that correspondence and reports to or from the NRC which contain information or records concerning a licensee's or applicant's physical protection or material control and accounting program for special nuclear material, not otherwise designated as Safeguards information or classified as National Security information or Restricted Data, may be withheld from public disclosure.

12.L. Real time monitors should be set up at the site boundaries to detect chemicals and radiation.

Response

The facility has continuous vent monitors for 13 discharge points which are considered to be the potentially significant contributors to the total plant radionuclide emissions. The plant also maintains both onsite and offsite stations to collect ambient air samples continuously. Samples are collected and analyzed monthly. Thermoluminescent dosimeters are also located both onsite and offsite to monitor external gamma radiation.

This program is considered to be adequate. l 12.M. Exposures to workers should be as low as reasonably achievable (ALARA).

Response

USEC is required by 10 CFR Part 20 to maintain radiation exposures to workers ALARA. In addition, USEC has committed in SAR Section 5.3 to do so.

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s 12.N. NRC should have civil penalty authority.

Response: See response to 6.D.

12.0. NRC should have more than 2 inspectors onsite.

Response: , See response to 6.D.

12.P. The emergency plan should adequately reflect the presence of nearby schools and institutions.

Response: The Emergency Plan includes U.S. Geological Survey maps that identify the  ;

locations of schools and churches.

12.Q. The plant has uranium leaks in its containment.

Response

A large portion of the enrichment process is operated below atmospheric pressure so that loss of containment would result in inleakage of air into the cascade as opposed to outleakage of UFe gas. Significant inleakage, which could cause severe operational upsets, can be detected in many ways before it impacts safety. Where UF, exists above atmospheric pressure, sensitive smoke detectors are required to be installed and operable at all times, so that any release of significant quantities of UF, would readily be detected.

Also the Radiation Protection program requires periodic surveys to detect contamination.

12.R. The public is concerned about safety due to past problems at the plant. NRC should assure that workers and the public are protected. -

Response

The NRC's mission is to assure thcf USEC provides adequate protection of the workers and l the public health and safety and NA: will carry out that responsibility.

12.S. Fluorine used at the plant is extremely hazardous.

Response: ,

The consequences of a fluorine release have been analyzed and the results reported in the '

application SAR. No effects are expected to people off site. The fluorine concentrations that would result in the immediate location of a release do have the potential to result in a fatality if a worker is unable to escape. Therefore, workers are trained in the proper J operation of the fluorine system and are required to operate the system in accordance with i those procedures. j l

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l 12.T. The method of evacuation if there was an accident is inadequate because school buses might be used.

Response

The Emergency Plan does not rely on school buses to evacuate personnel from the plant area.

12.U. The plant possesses neptunium.

i Response:

l Neptunium is one of the many transuranic radioactive substances that the NRC staff has authorized USEC to possess in the form of sealed (0.5 Ci) and unsealed (1.0 Ci) sources and as contaminants resulting from previous operations and projected to result from l

processing uranium from the Former Soviet Union. The provisions of 10 CFR Part 20 require USEC to handle all radioactive substances in a manner that assures public, worker and environmental protection.

12.V. NRC needs to review the waste treatment process involving thermal absorption.

Response

The waste treatment process that USEC is using does not include a thermal absorption process.

12.W. Will NRC regulate AVLIS?

Response: See response to 11.G.

12.X. The plant operations involve both chemical and radiological hazards.

Response

USEC has both radiation safety and chemical cafety programs. The NRC staff has determined that USEC's proposed radiation protection program will ensure worker protection from all onsite radiological hazards as well as the toxicological hazards of uranium. NRC will also coordinate with OSHA to assure protection from all chemical hazards.

12.Y. How does NRC plan to measure neutron exposures to workers?

Response

USEC must measure significant neutron doses to workers as required by 10 CFR Part 20.

10 CFR 20.1502 requires all NRC licensees to monitor personnel for radiation exposures if they are likely to receive in one year more than 10 percent of the annual dose limit. For the who!e body, the annual dose limit is 5 rem.

The doses that workers are likely to receive from neutron radiation orisite are well below l the monitoring criteria provided in 10 CFR 20.1502. For instance, in the tails cylinder I storage areas, USEC reports neutron dose rates between 0.1 and 0.3 millirem /hr. This A-19 L

dose rate is a small fraction Oess than 10 percent) of the dose rates from bremsstrahlung and other direct radiations emitted from full tails cylinders. Also, past history of operations at the GDPs has not indicated a significant neutron dose hazard. See CER Chapter 7.

12.Z. Workers should have a daily urinalysis by an independent monitor.

Response

USEC has proposed an adequate bioassay program designed to detect soluble uranium intakes below the weekly regulatory limit of 10 milligrams. This conclusion is based on the determination that the lower limits of detections for urine samples, in conjunction with the sample collection frequencies proposed by USEC, are adequate to ensure compliance with the weekly 10 milligram soluble uranium intake limit. Also, past history of operations at the GDPs has not indicated a significant internal dose hazard.

12.AA. The public is concerned about liquid effluents from the plant.

Response

It is not clear what the nature of the commenter's concern is. Discharges from the plant are made in accordance with the state issued discharge permit. USEC collects water samples as part of its monitoring program.

Commenter: Mr. Tim Mitchell 12.BB. Workers should not be wrongfully disciplined for contamination incidents.

Response

The NRC's primary responsibility at the gaseous diffusion plants is to ensure that workers and the public are protected from unnecessary or excessive exposure to radiation and that the facilities are operated in a safe manner. The NRC does this by establishing requirements in Title 10 of the Code of Federal Regulations and in the certificate issued to USEC. In general, the NRC does not regulate or otherwise have a role in the hiring and firing practices at the gaseous diffusion plants or other facilities it regulates, with the following important exception. Federal law prohibits an employer from firing or otherwise discriminating against an employee for bringing safety concerns to the attention of the employer or the NRC. Specifically a worker may not be fired or otherwise discriminated against because the worker (1) asks the NRC to enforce its rules against the employer; (2) refuses to engage in activities which violate NRC requirements; (3) provides or plans to provide information to the NRC or the employer about violations of requirements or safety concerns; or (4) asks for, testifies in, helps in, or takes part in an NRC or Congressional or Federal or State proceeding.

Commenter: Mr. Gerald Wilkins 12.CC. Will NRC certify the plant if there is no labor agreement?

Response

NRC regulations do not require a labor agreement to be in place for the certification.

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13. Paducah Public Meeting Transcript - December 5,1995 l

Commenter: Ms. Jotilley Dortch 13.A. Will there be a commerciallow-level waste storage disposal facility at Paducah?  ;

l Response: l No, there will not be a USEC commercial waste facility at Paducah.

13.B. USEC should not be permitted to withhold the waste storage plan from public  !

disclosure. l Response: i See response to 3.H.

13.C. There is a seismic risk at the Paducah plant. j Response:  ;

See response to 2.B. I 13.D. An accident could adversely affect the nearby inland waterway.

Response

It is possible for contamination to occur offsite as a result of an accident, but the risk is very small. However, most of the contamination is likely to occur near the vicinity of the release. USEC would be required to assess any contamination that occurred as a result of i an accident and decontaminate the area as appropriate.

Commenter: Mr. Ronald Lamb 13.E. NRC should have civil penalty authority.  !

l Response- l See response to 6.D.

13.F. Does NRC have adequate inspection coverage?

Response

See response to 6.D.

1 13.G. Information on waste should be available to the public.

Response: -

See response to 3.H.

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o Commenter: Mr. A.B. Puckett 13.H. Is the NRC an independent regulator, separate from DOE? Is the NRC going to be

, responsible for the public health and safety and assuring protection of the people near the

, plant?

! Response:

5 Yes, NRC is independent from DOE. NRC is responsible for assuring that USEC provides protection.for people living near the plant. See also response to 12.R.

j Commenter: Mr. Mark Donham 13.1. In the past, DOE has not adequately controlled disposal of radioactive and hazardous waste.

Response

USEC has a radioactive waste management program for safe handling of waste. The NRC will assure that USEC handles its waste safely.

13.J. Public notification about the meeting was not sufficient. People who have J expressed interest over the years should have received a notice in the mail.

Response

The meeting was announced in press releases, Federal Reaister Notices, and ads in the local paper. In addition, those individuals who have expressed interest in the NRC

• activities at the site did receive a copy of the Federal Reaister Notice announcing the meeting.

13.K. The people have a right to know the extent of contamination around the plant site.

Response

The application contains environmental monitoring data. In addition, DOE publishes an annual Environmental Report that summarizes environmental monitoring data.

13.L. Does the C-310 stack routinely vent radioactive gases?

Response

The Paducah plant operations result in gaseous effluent from several sources, including the C-310 stack. Emissions are within the regulatory limits. The highest dose to an offsite individual in 1994 was 0.016 mrem, compared to the 10 mrem limit.

13.M. Is there technetium in the groundwater and is neptunium the transuranic waste at the Paducah plant?

Response: ,

Yes, there is technetium contamination in the groundwater. Both the~ technetium and the transuranic wastes resulted from processing recycled uranium that contained these contaminants. Recycled uranium is no longer processed at the Paducah plant.

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. . qp 13.N. Are rail shipments coming from Portsmouth and Oak Ridge?

I Response- 1

The Paducah plant ships most of its product to the Portsmouth plant. Portsmouth ships i some of its uranium tails back to the Paducah plant for further removal of the uranium-235.

The Paducah plant does not routinely ship to or receive items from Oak Ridge. The Oak j Ridge facility is shut down and may on occasion ship equipment to the Paducah plant.

13.0. What type of warning system will the NRC require for the public?

i

Response

1' It should be noted that USEC is not required by Part 76.91," Emergency planning," to have a public warning system and that the inclusion of a public warning system in the

! emergency plan for the Paducah Gaseous Plant was a voluntary decision made by USEC.

} The existing public warning sirens do not provide total coverage of the immediate notification area, and plans are in place to replace the existing sirens with a new siren j system to ensure better coverage of the immediate notification area. The scheduled

completion date for the new system to be installed and operational is March 15,1997.

13.P. If NRC cannot find the plant in compliance, could it be turned back over to DOE?

! Response:

DOE has agreed to retain regulatory oversight over the gaseous diffusion plants until the initial certification is completed, and for an additional transition period to allow for an i orderly transition to NRC regulation. After NRC assumes regulatory jurisdiction, regulatory oversight will remain an NRC responsibility until enrichment operations are terminated and

- the plants are returned to DOE for decommissioning. After initial certification, the plants i must retain a valid certificate of compliance and/or an approved compliance plan to i continue to operate.

i j 13.0. NRC should do offsite background monitoring to assess any accumulation of i radioactive contamination.

Response

' NRC does not routinely conduct environmental sampling around similarly licensed facilities j and does not plan to conduct environmental sampling around the Paducah plant. As part l of the inspection program, samples are occasionally analyzed to verify the applicant's i sampling and analysis program.

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v Appendix B INTERAGENCY CONSULTATION RESULTS l

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c Interagency Consultation Results The Energy Policy Act of 1992 required the NRC to consult with EPA. NRC added this requirement in 10 CFR 76.53, which requires the NRC to consult with EPA and solicit EPA's written comments on the application. As part of the consultation process, NRC staff met with EPA Region 4 on November 29,1994 at their offices in Atlanta, Georgia.

Region 4 is the region with responsibility for the Paducah facility. Prior to receipt of the application the staff met with EaA headquarters on August 24,1995;the regional offices (Region 4 and Region 5) participated by a conference line. EPA Region 4 was provided a copy of the application by letter dated September 15,1995. The letter invited EPA to comment on the application, informed them of the public meeting, and offered to meet to discuss any questions. In order to inform EPA of changes in the application, by letter dated October 27,1995, the staff provided copies of USEC responses to questions in the environmental and waste management areas. By letter dated November 7,1995, a copy of the Compliance Plan was provided to EPA with a request for comment and an offer to meet at EPA's request. EPA responded by letter dated December 7,1995. EPA noted no major problems or omissions in the material provided and stated that the application appears to reflect compliance status with the permits and regulations under EPA jurisdiction.

Although not required by law, NRC staff also consulted with OSHA. OSHA was invited to comment on the application and Compliance Plan by letters dated September 15,1995, and November 8,1995, respectively. OSHA provided comments by Mtter dated November 24,1995;the response to those comments is provided in Appendix A.

The 1996 USEC Privatization Act requires NRC and OSHA to enter into a memorandum of ,

agreement, within 90 days of enactment of the Privatization Act, to govern the exercise of their authority over occupational safety and health at the GDPs. The staff held several meetings / discussions with OSHA officials in developing a Memorandum of Understanding (MOU); an MOU had been planned prior to the Privatization Act. The MOU describes the authorities of NRC and OSHA in implementing the Act and covers such topics as inspection, investigation, and enforcement. The MOU was signed on July 26,1996,and published in the Federal Reaister on August 1,1996.

The staff also offered to meet with state and local officials and held a meeting for this purpose on December 5,1995, No state or local official expressed any concern outside the issues already under consideration by the staff.

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