ML20237A836

Rev 23 to USEC-02,application for Us NRC Certification for Portsmouth Gaseous Diffusion Plant.Rev Includes Changes to SAR

ML20237A836
Person / Time
Site:	Portsmouth Gaseous Diffusion Plant
Issue date:	08/06/1998
From:	UNITED STATES ENRICHMENT CORP. (USEC)
To:
Shared Package
ML20237A831	List: ML20237A829 ML20237A836
References
NUDOCS 9808170170
Download: ML20237A836 (51)

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APPLICATION FOR UNITED STATES NUCLEAR REGULATORY COMMISSION CERTIFICATION PORTSMOUTH GASEOUS DIFFUSION PLANT USEC-02 REMOVE / INSERT INSTRUCTIONS Revision 23 Remove Pages Insert Pages VOLUME 1 UPDATED LIST OF EFFECTIVE PAGES i / ii i / ii y / vi y / vi ix / x ix/x xiii / xiv xiii / xiv

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TABLE OF CONTENTS v / vi y / vi CHAPTER 1 1-5 through 1-10 1-5 through 1-10 SECTION 3.1 3.1-3 / 3.1-4 3.1-3 / 3.1-4 SECTION 3.2 3.2-41 / 3.2-42 3.241/3.2-42 SECTION 3.7 3.7-1 through 3.7-4 3.7-1 through 3.7-4 A-1 through A-10 A-1/ A-2 l

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APPLICATION FOR UNITED STATES NUCLEAR REGULATORY COMMISSION CERTIFICATION PORTSMOUTH GASEOUS DIFFU3 ION PLANT USEC ')2 REMOVE /INSEET INSTRUCTIONS Revision 23

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l SECTION 5.2 5.2-11 / 5.2-12 5.2-11 / 5.2-10 VOLUME 3 EMERGENCY PLAN List of Effective Pages, pages i / ii List of Sffective Pages, pages i / il 1-7 through 1-8b 1-7 through 18b FUNDAMENTAL NUCLEAR MATERIALS CONTROL PLA.N List of Effective Pages, pages i / il List of Effective Pages, pages i / ii Table of Contents, pages vii / viii Table of Contents, pages vii / viii 9-5 through 9-8 9-5 throug!; 9-8 2 of 3

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I SAR-PORTS April 15,1998 Rev.19 LIST OF FIGURES Chapter 2 (Continued)

Figure Eage i

2.6 7 "BLUME" PORTS Seismic Hazard Curves............................ 2.6-22

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2.6-8 Interpretation of the BLUME and Seismic Hazard Curve for PORTS.......... 2.6-23 1

2.6 Estimated Peak Acceleration Return Periods, Dames and Moore -

Seismic Hazard Curve.......................................... 2.6-24 2.6-10 TERA - PORTS Seismic Hazard Curve............................... 2.6-25 2.6-11 Superimposed Results for Fernald and PORTS......................... 2.6-26 2.6-12 Recommended Fernald and PORTS Seismic Hazard Curve................. 2.6-27 i

l TABLE OF CONTENTS Chapter 3 Eage 3.0 FACILITY AND PROCESS DESCRIPTION............................ 3.0-1

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3.1 CASCA D E SYSTEMS...................................... 3.1-1 3.1.1 Uranium Enrich =.ent Cascade........................... 3.1-1 3.1.2 Purge Cascade System............................... 3.1-98 3.1.3 Freezer / Sublimer Systems............................

3.1-120 3.1.4 Cold Recovery System...............................

3.1-129 1

3.1.S Freon Degrader System.............................. 3.1-152 j.

3.1.6 Cascade Systems Safety Systems, Design Features, and Administrative Controls.............................

3.1 - 162 l

l 3.2 UF, FEED, WITHDRAWAL, SAMPLING, HANDLING, AND l

CYLINDER STORAGE FACILITIES AND SYSTEMS............... 3.2-1 3.2.1 Cascade UF, Feed and Sampling Systems.................... 3.2-4 3.2.2 Talis and Product (ERP and LAW) Withdrawals............. 3.2-20 3.2.3 Top Product (PW) and Side Withdrawals.................. 3.2-42 l_

3.2.4 The High Assay Sampling Facilities (HASA)................. 3.2-44 3.2.5 UF, Cylinder Shipping and Receiving..................... 3.2-60 3.2.6 Cylinder Storage................................... 3.2-62 3.2.7 UF, Cylinder Transport.............................. 3.2-72 I

3.2.8 Safety Systems, Design Features for Safety, and Administrative Con trols......................................... 3.2-74 l

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SAR-PORTS A5ust 6,1998 Rev.23 TABLE OF CONTENTS Chapter 3 (Continued)

Eage 33 URANIUM RECOVERY AND CHEMICAL SYSTEMS............... 33-1 3.3.1 X-705 Decontamination and Recovery Facility

................33-1 33.2 X-705 Waste Water Treatment Facility.................... 33 42 333 Contaminated Storage Facilities......................... 33-52 3.5.4 Biodenitrification Facilities............................ 33-55 3.4 POWER AND UTILITY SYSTEMS............................ 3.4-1 3.4.1 Electrical Systems.................................... 3.4-1 3.4.2 Plant Water System................................. 3.4-7 3.43 Plant Nitrogen System................................ 3.4-21 3.4.4 Plant Air System................................... 3.4-27 3.4.5 Plant Steam and Condensate Systems..................... 3.4-31 3.4.6 Plant Waste Systems and Facilities....................... 3.4-33 3.4.7 HFIF: Systems..................................... 3.4-38 3.5 GENERAL SUPPORT FACILITIES AND SYSTEMS................ 3.5-1 3.5.1 Maintenance Facilities................................. 3.5-1 3.5.2 Laboratories and Pilot Plants........................... 3.5-20 3.53 Receiving and Storage Facilities......................... 3.5-29 3.5.4 Communications and Data Processing..................... 3.5-34 3.5.5 Administration Facilities.............................. 3.5-43 3.5.6 Health Protection Facilities............................ 3.5-46 3.6 FIRE PROTECTION AND RADIATION ALARM SYSTEMS AND ENVIRONMENTAL PROTECTION FACILITIES.................. 3.6-1 3.6.1 Fire Protection Systems................................ 3.6-1 3.6.2 Radiation Alarm Systems.............................. 3.6 12 3.63 Environmental Protection Facilities....................... 3.6-22 3.7 HEU AND MEU ACTIVITIES................................ 3.7-1 l

3.7.1 Descri ption........................................ 3.7-1 3.7.2 Temporary Conversion of Leased Areas of the X-705 Facility from NRC to DOE Regulation............................... 3.7-2 3.7.3 Organization and Responsibilities......................... 3.7-3 APPENDIX A (Deleted)

............................................A-1 l

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,q SAR-PORTS September 15, 1995

(_)

Rev.I from the cascade for maintenance service. Each cell consists of a number of converters and the supporting equipment, primarily the compressors and heat exchangers. The cell volumes are listed, by size of equipment, below:

a.

X-33 equipment (8 stages,26,000 to 28,000 ft.').

b.

X-31 equipment (10 stages,11,560 to 11,850 ft.').

c.

X-29 equipment (10 stages, 8,960 to 11,080 ft.').

d.

X-27 equipment (12 stages,2,210 to 2,250 ft.').

e.

X-25 equipment (6 or 12 stages, 940 to 1,630 ft.').

For practical operation, cells are grouped together to form units. Each unit consists of cells contammg the same equipment type and usually operate under the same conditions. This groupmg also allows additional flexibility in providing auxiliary service (e.g., lube oil and hydraulic oil).

i 3.1.1.1.3 Cascade Shane. Flow and Taner Three steel-framed, transite-covered, two-story buildings (X-326, X-330, and X-333) house the process equipment. The X-333 Building contains 640 stages (80 cells, 8 units) of X-33 (or 000) size equipment, which is inserted in the cascade flow between units 2 and 3 of X-31 (00) size equipment (located in the X-330 Building). He X-330 Building houses 500 stages (50 cells, 5 units) of X-31 size equipment, and

{

600 stages (60 cells, 6 units) of X-29 (0) size equipment. The X-326 Building houses 720 stages (60 cells,3 h

units) of X-27 equipment,1560 stages (130 cells, 6.5 units) of X-25 equipment, and 60 stages (10 cells, 0.5 V

unit) of purge cascade equipment. De location of any particular converter can be identified by specifying the equipment type, unit, cell, and stage. For example, the converter in stage one, cell five, unit three in the X-333 equipment would be specified by X-33-3-5.1.

l Figure 3.1.1.1-3 presents a graphic representation of the cascade and provides a simplified process

]

flow diagram. Diagrams of each type of cell are presented in Section 3.1.1.2 which provides a detailed l

description of the process flow in each building. Figure 3.1.1.1-4 provides a schematic of the cascade interplant flow.

The Portsmouth GDP has the capability of producing reactor grade (currently up to 5.0 % U-235 with a design capability to produce up to 10% U-235) product. In order to enrich uranium from a feed of 0.711 %

U-235 to product material at 5.0% U-235, the number of diffusion stages between the feed point and the top product is about 2200. This section of the cascade is called the " enriching' section. Additional stages are required to strip the U-235 isotope from normal feed material so that the depleted material or tails will contain about 0.2 to 0.3 % U-235 (this depends upon power load and marketing considerations). This section of the cascade, called the " stripping" section, is made up of about 500 stages. The Portsmouth GDP consists of 2700 isotopic stages. He cascade also contains 60 purge stages that separate and purge the light gas contam' ants m

that leak into the system, here are several factors that influence the cascade " shape." A cascade producing at a high product rate must have higher flow rates between stages than one producing at a lower product rate. To utilize power efficiently, the stages in the middle (or near the feed point) must have higher flow rates than those at the ends.

O This is usually represented or drawn in a diamond shape to illustrate this flow i

3.1-3

l SAR-PORTS August 6,1998 Rev.23 distribution. The Portsmouth cascade is tapered by appropriate sequencing of the five different equipment sizes and by reducing the pressures across each size equipment.

Tapering is accomplished as follows: The middle part of the cascade, where feed is introduced, contains the largest, highest pressure cells resulting in the maximum interstage flow rate uear the feed point.

The flow taper is achieved in units of the same equipment size by gradually decreasing the pressure from cell to cell and unit to unit. At approximately 320 stages away from the feed point, the equipment size is reduced (to X-31 size) and the pressure is raised to compensate for the smaller equipment. Pressure tapering is then resumed as the equipment size is further reduced; pressures would again be tapered down toward the cascade ends in order to achieve efficient flow distribution.

Fuel for light water reactors can be produced by diffusion cascades having only two, or possibly three equipment sizes, while a longer cascade, such as the Portsmouth GDP, is designed to produce a more highly enriched product for special reactors and, therefore, is made up of five equipment sizes (illustrated in Figure 3.1.1.1-5). As seen in this figure, cascade feed is made up of multiple feed streams (typically three and sometimes four streams), and are withdrawn for commercial reactor assay streams (currently 3.2 to 4.95 %

U-235, with a capability up to 10% U-235) and the tails stream. Average high-side pressures and assays are given on either side of the schematic.

Most of the plant load is consumed by the low-assay X-29, X-31, and X-33 size equipment. Each block shown in the schematic represents a unit, with the number of cells and stages shown in the right side of the block. The diamond represents the taper flow of an ideal cascade of this type. The sizes of electrical components which provide power to the stage compressors must also be tapered in accordance with the flow taper in the UF. system. He sizes of coolant condensers, and the recirculating cooling water flow rates, must also conform with the UF. flow taper. The assays and feed and withdrawal locadons shown in Figure 3.1.1.1-5 are not precise, but are typical for current market conditions.

The product withdrawals from the Portsmouth cascade are usually made up of two different streams.

They are removed at the Extended Range Product Withdrawal (ERP) and Low Assay Withdrawal (LAW) stations. The cascade feed is made up of two (sometimes three, including HEU Refeed) streams of different assays. He lower feed stream is usually normal assay material; the higher feed consists of slightly enriched product from the Paducah GDP. At every point where there is c feed or withdrawal, the flow taper required for the most efficient power utilization and assay control for withdrawal operations must take into account the sudden change in cascade flow conditions. A taper for projected typical operation of the Portsmouth cascade at full power is illustrated in Figure 3.1.1.1-5; the pressures and assays are approximate. A different taper would be required for other feed and withdrawal locations. The number of cells may vary as the lower product assay changes. Normally, cells are started and shut down and the cascade retapered to maintain high efficiency when the lower product assay is changed or as power availability dictates.

He major cascade feed streams are supplied through feed headers connected to the feed autoclaves in the X-342 and X-343 Feed Facility.

I O

3.1-4

SAR-PORTS April 15,1998 Rev.19 A

Power Failure b

When a total power failure occurs, a diesel-driven generator supplies power to the withdrawal stations MOV's. Should the diesel generator fail to start or supply power, tie breakers can be closed to another generator to supply backup power for closing the MOV's.

The air-operated valves are designed to fail in the safe position; consequently the withdrawal cylinders, condensers, accumulators, and compression loops would isolate and the vent valve would open, when either an air or electrical failure occurs.

The electronic instrumentation on the LAW control panel in ACR-1 is supplied power from the backup generator. ERP and Tails instruments are pneumatic and therefore not affected by a power failure.

Phnt Air Failure Failure of the plant air system requires shutdown of the withdrawal facilities. During an air supply failure there would be time for an orderly shutdewn and isolation. The compressor recycle valves would open protecting the compressors. The RCW control vzives to the cooling ' condensers would close causing the compressors to trip on high coolant pressure or temperature, except in the Tails Area.

As described in Section 3.2.2.2, the air operated valves in the HPV system would fail in the safe position. The air operated pigtail valve has a reserve volume of air sufficient for valve closure.

Over a prolonged outage, the condensers and accumulators can be drained to a withdrawal cylinder and evacuated to surge drums 3.2.2.10 Nuclear Criticality Control The accumulator size is primarily the limiting factor when determining the maximum assay that may be withdrawn through a particular withdrawal loop. (Section 3.2.2.2) However, the 10-ton and 14-ton cylinders that are used at these stations are normally limited to 5 % U-235 assay. Cylinder fill and assay limits are discussed in Section 3.2.1.1.2.

l When the Tails loop is being used for product withdrawal, double block valves between loops are closed and chained to prevent mixing of assays. Continuous assay monitoring is required. The assay is verified by the laboratory analysis before the withdrawal is started. Some valves on the withdrawal headers may be chained and/or locked to prevent misvalving.

Liquid cylinders are moved by crane (Section 3.2.1.1.4), one at a time, and are not carried over other cylinders in the temporary storage yard, where they are stored until the solidification criteria defined in SAR Section 3.2.7 have beer satisfied.

The heated housings around lower liquid distribution headers, condenser and accumulator housings have 1-inch holes drilled in their bouoms to provide immediate drainage or vaporization in case of a UF. leak j

in those areas. The 8-inch i.d. accumulators at the ERP Station are exceptions. They are pd l

3.2-41 l

l

SAR-PORTS August 6 1998 Rev.23 located outside the heated housing and they are insulated with fiberglass and covered with an aluminum outer 3

j sheath. This insulation has an o.d. of 13 inches. To ensure that there is no leakage of UF, into the insulation, l

the 8-inch accumulators are leakrated annually. Should a small leak occur, the system would be subcritical, at an H/U-235 ratio of 10, at 10% assay.

3.2.2.11 Safety Structures _ Systems and comnonents The information contamed in this section has been deleted. Safety structures, systems and components are delineated in Chapter 4, Accident Analysis, of this SAR.

3.2.2.12 Administrative controls The information in this section has been deleted. Administrative controls required for the safe operation of systems, structures and components have been specified in the TSR portion of the Application.

1 3.2.2.13 Surveillance The information in this section has been deleted. Surveillance requirements required for the safe operation of systems, structures and components have been specified in the TSR portion of the Application.

3.2.3 Top Product (PW) and Side Withdrawals Both permanent (fixed) and portable facilides and equipment are available to withdraw enriched UF.

from any point in the enrichment cascade.

O l -

'Ihe Product Withdrawal Facility (PW) is a fixed facility located in the south west corner of the X-326 Building that was used to withdraw HEU product. The UF was removed from the cascade by routing it through headers from the desired cell and stage to a manifold in the PW facility (see Figure 3.2-13).

l Side withdrawal facilides make use of a portable unit consisting of a refrigeration unit and bath mounted on a scale. A schematic of a typical portable unit is shown in Figure 3.2-16. A cylinder is immersed in the bath and connected by pigtails to the line recorder manifold of the unit (or half unit) from which the withdrawal is to be made on the cell floor. The withdrawal point is selected by putdng the line recorder selector switch on the appropriate cell. Control samples can be obtained from the line recorder manifold lines in the ACR. The UF flow is established by the pressure differential between the suction and discharge pressures of the withdrawal stage compressor. As the UF, enters the cold cylinder, it condenses to a solid and the non-condensible gases return to the cascade through the line recorder return line. Line recorder manifold withdrawal connections are located in each unit of X-326 O

3.2-42

n

~SAR-PORTS August 6,1998 j

(J Rev.23 3.7 HEU AND MEU ACTIVITIES l

3.7.1 Description The Regulatory Approach for Post NRC Certification of Gaseous Diffusion Plants (IW Parks to GP Rifakes, October 11,1995), DOE will retain regulatory authority over HEU, except for residual HEU that is held up in equipment subsequent to DOE HEU cleanout performed as part of the HEU suspensian project and Category III quantities (or less) of other HEU (these small quantities of HEU may be handled incidental to general enrichment activities).

The Fundamental Nuclear Materials Control Plan, submitted as part of this SAR and the associated programs and plans, describes the accounting methods for HEU activities.

l Heel quantines of HEU material are stored in 5", 8" and 12" cylinders in the X-345 facility which is not leased by USEC. The HEU is moved under DOE oversight from the X-345 facility to a cylinder cleaning facility in a non-certified portion of the X-705 Building.

l l

After a cylinder has been fed, a relatively small amount of non-volatile uranium typically remams in the cylinders. This " heel" is removed by a cleaning process conducted in a DOE-regulated X-705 Small 0

Cylinder Cleaning area or shipped offsite for cleaning. Solutions resulting from the cleaning process are blended with solutions contanung normal, depleted or LEU to reduce the assay to less than 10 wt-% "U. The solution is then transferred to the uranium recovery area where it is converted to uranium oxides; finally the oxides are stored for future disposition. The cleaned cylinders and any cylinders destroyed during the cleaning process are returned to DOE.

As a part of the normal operation of the gaseous diffusion process, cells are treated with oxidant gases to remove deposits of uranyl fluoride and other compounds from tile cascade equipment surfaces in a manner described in Section 3.1.1.12. Generally, these treatments liberate a few hundred to several thousand grams of ur:: mum from deposits. The treatment gases, including any uramum liberated from deposits as UF., are evacuated to surge drums and then returned to the enrichment cascade at a point near its origin.

Cell treatment may result in the liberation of small quantities of residual HEU that svas left in USEC process equipment following completion of the DOE cleanup process. This may occur at any point during the remaining operational life of the enrichment cascade. The liberated HEU material will mix with the LEU material in the process equipment and surge drums and the treatmt.nt gases will be returned to the cascade, where it will be mixed with the much larger quantities of uranium present in the interstage flow at LEU ennchments. This process ensures that the blended stream remains within the "U possession limits defm' ed in Table 1-3. Analysis of uranium enrichment is not performed prior to returning the mixtures to the cascade.

Any changes in uranium inventory due to " recovery" of the relatively small amounts of HEU would be reflected in USEC's enrichment cascade Inventory Difference (ID) during periodic inventories.

G 3.7-1

)

1

\\

t i

L__

b

l SAR-PORTS August 6,1998 Rev.23 In addition to the HEU downblending activities, there may be occasions when equipment or components removed from the LEU cascade, X-705 Building or other Ic& sed areas contain uranium greater than USEC's possession limits due to the presence of material left from previous DOE operations. This includes equipment or components in the X-326 facility that need to be removed for maintenance or other operational pmposes and contain retained inventory of UF. plated out on the inside surfaces (includes both shutdown and operating equipment in the X-326 facility); material and equipment such as alumina traps, seal exhaust oil and GP contr% from always-safe vacuums that are generated es part of ongoing operations in X-326, or equipment in X-705 that needs to be removed for maintenance purposes. On those occasions when material is discovered in uninstalles equipment in any USEC-leased and NRC-certified area of the PORTS plant that exceeds USEC's possession limits, the PORTS NRC Resident Inspector will be notified of the situation; and the equipment, components, or items will be moved to a DMSA within seven days after removal for storage or further processing. Equipment and components may subsequently be disassembled and decontammated in an area in the X-705 Building which is placed temporarily under DOE regulation with appropriate safeguards in place (see the description of the process in Section 3.7.2). Material removed which exceeds USEC's possession limits will be retained by DOE or will be blended with LEU solution until the overall enrichment is less than USEC's possession limits. DOE regulation and associated safeguards will cease to be applied when material equal to or greater than USEC's possession limits is no longer present. The blended-down solution would be processed through uranium recovery as described above. While the X-705 area is temporarily converted to DOE regulation, access to the DOE regulated areas is controlled in accordance with the DOE Regulatory Oversight Agreement (ROA).

3.7.2 Temporary Conversion of Leased Areas of the X-70S Facility from NRC to DOE Regulation Temporary conversion ofleased areas of the X-705 facility from NRC regulation to DOE regulation will be accomplished as follows:

1.

USEC will send a written request to NRC and DOE to temporarily convert an area of X-705 to DOE regulation for a specific activity in order to remain in compliance with USEC's possession limits. This nonfication will include a description of the work planned to be done, why USEC's possession limits may be exceeded if the work was done under NRC regulation, anticipated start and end dates for the temporary conversion, and a justification that NRC regulated activides will not be impacted by the temporary conversion.

2.

USEC will receive written approval from NRC and DOE to proceed with temporary conversion to DOE ROA regulation prior to beginnmg the work activity, along with the approved start and end dates.

3.

USEC will complete the subject activity (or suspend the activity if an unacceptable impact on other NRC activities have been identified) and then perform a static special inventory and a security sweep of the specific area to provide a high degree of assurance that materials are not present that would cause USEC to exceed its possession limits after conversion to NRC regulation.

O 3.7-2

SAR-PORTS August 6,1998 Rev. 23 sg 4.

USEC will notify NRC and DOE in writing that the activity is complete (or has been suspended if an unacceptable impact on other NRC regulated activities has been identified), that USEC requests conversion to NRC regulation, and that converting the area to NRC regulation will not result in USEC exceeding it possession limits.

5.

DOE will notify NRC and USEC in writing of any open findings or issues related to the subject area of the facility to NRC, and that DOE agrees with the conversion to NRC regulation.

6.

DOE and NRC will agree to a date when all regulatory authority will convert from DOE regulation to NRC regulation, and NRC will noufy USEC and DOE of this date in writing.

7.

The area will convert to NRC regulation on the date specified by NRC.

3.7.3 Organization and Responsibilities When HEU quantities greater than Category Ill are eliminated from PORTS (other than residual l

holdup that may be encountered during cell treatment and equipment removal described in Section 3.7.1 above), NRC will be the regulator for the entire USEC operated site and the DOE Regulatory Oversight Agreement (ROA) will expire. However, should remaining quantities of HEU or MEU greater than Category III be physically removed from USEC equipment subsequent to expiration of the DOE ROA, the material will

[]

be the responsibility of DOE.

.V The DOE ROA will cease to apply to all facilities or activities for which NRC assumes regulatory authority. This may occur on the entire site or on a facility, area, or activity basis. The DOE ROA will continue to be used for regulation after NRC certification for leased uncertified facilities, areas or activities after NRC assumes regulatory oversight. Such facilities and areas include the X-705 Small Cylinder Cleaning area; such activities include the movement of HEU along USEC leased roadways.

The boundaries between DOE and NRC regulation will not coincide with the USEC/ DOE lease during the transition period (HEU activities after NRC certification) from DOE to NRC oversight. The boundaries where the DOE ROA will apply after initial NRC certification of PORTS are described in the SAR and the associated programs and plans and the DOE Compliance Plan.

l A segregated DOE-regulated area in X-705 has been modified to allow cleaning of all sizes of cylinders following refeed. Processing of HEU materials in other areas of X-705 is complete. The DOE ROA will apply to identified, segregated areas in X-705 for these remanung refeed support activities. However, i' refeed support activities involving quantities of HEU greater than Category III levels continue, which are not confined to the segregated areas, then the DOE ROA will cover the entire X-705 facility.

l l

t O

3.7-3 e

i

SAR-PORTS May 31,1996 Rev.3 l

l l

l Blank Page O

9

' SAR-PORTS August 6,1998 Rev. 23 Appendix A - Deleted.

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

' SAR-PORTS April 15,1998 i

Rev.19 j

1 LIST OF FIGURES Chapter 2 (Continued) j 1

Figure Eage 2.6-7 "BLUME" PORTS Seismic Hazard Curves............................ 2.6-22 2.6-8 Interpretation of the BLUME and Seismic Hazard Curve for PORTS.......... 2.6-23 2.6-9 Estimated Peak Acceleration Return Periods, Dames and Moore -

Seismic Hazard Curve.......................................... 2.6-24 2.6-10 TERA - PORTS Seismic Hazard Curve............................... 2.6-25 2.6-11 Superimposed Results for Fernald and PORTS......................... 2.6-26 4

2.6-12 Recommended Fernald and PORTS Seismic Hazard Curve................. 2.6-27 l

TABLE OF CONTENTS Chapter 3 Eage 3.0 FACILITY AND PROCESS DESCRIPTION............................ 3.0-1 a

3.1 CiSciD E SYSTEMS...................................... 3.1-1 3.1.1 Uranium Enrichment Cascade........................... 3.1-1 3.1.2 Purge Cascade System

...............................3.1-98 3.1.3 Freezer / Sublimer Systems...........................

3.1-120 '

3.1.4 Cold Recovery System...............................

3.1-129 3.1.5 Freon Degrader System..............................

3.1-152 3.1.6 Cascade Systems Safety Systems, Design Features, and Administrative Controls.............................

3.1 -162 3.2 UF FEED, WITHDRAWAL, SAMPLING, HANDLING, AND CYLINDER STORAGE FACILITIES AND SYSTEMS............... 3.2-1 3.2.1 Cascade UF Feed and Sampling Systems.................... 3.2-4 3.2.2 Tails and Product (ERP and LAW) Withdrawals............. 3.2-20 3.2.3 Top Product (PW) and Side Withdrawals.................. 3.2-42 3.2.4 The High Assay Sampling Facilities (HASA)................. 3.2-44 3.2.5 UF Cylinder Shipping and Receiving..................... 3.2-60 3.2.6 Cylinder Storage................................... 3.2-62 3.2.7 UF Cylinder Transport.............................. 3.2-72 3.2.8 Safety Systems, Design Features for Safety, and Administrative Controls......................................... 3.2-74 y

i

SAR-PORTS August 6,1998 Rev.23 TABLE OF CONTENTS Chapter 3 (Continued) 233C 3.3 URANIUM RECOVERY AND CIIEMICAL SYSTEMS............... 33-1 3.3.1 X-705 Decontamination and Recovery Facility................ 3 3-1 33.2 X-705 Waste Water Treatment Facility.................... 33-42 33J Contaminated Storage Facilities......................... 33-52 3.3.4 Biodenitrification Facilities............................ 3 3-55 3.4 POWER AND UTILITY SYSTEMS............................ 3.4-1 3.4.1 Electrical Systems.................................... 3.4-1 3.4.2 Plan t Water System.................................. 3.4-7 3.43 Plant Nitrogen System................................ 3.4-21 3.4.4 Plant Air System................................... 3.4-27 3.4.5 Plant Steam and Condensate Systems..................... 5.4-31 3.4.6 Plant Waste Systems and Facilities....................... 3.4-33 3.4.7 IIF/F Systems..................................... 3.4-38 3.5 GENERAL SUPPORT FACILITIES AND SYSTEMS................ 3.5-1 3.5.I Maintenance Facilities................................. 3.5-I 3.5.2 Laboratories and Pilot Plants........................... 3.5-20 3.5.3 Receiving and Storage Facilities......................... 3.5-29 3.5.4 Communications and Data Processing..................... 3.5-34 3.5.5 Administration Facilities.............................. 3.5-43 3.5.6 IIealth Protection Facilities............................ 3.5-46 3.6 FIRE PROTECTION AND RADIATION ALARM SYSTEMS AND ENVIRONMENTAL PROTECTION FACILITIES.................. 3.6-1 3.6.1 Fire Protection Systems................................ 3.6-1 3.6.2 Radiation Alarm Systems.............................. 3.6-12 3.63 Environmental Protection Facilities....................... 3.6-22 3.7 IIEU AND MEU ACTIVITIES................................ 3.7-1 l

3.7.I Descrip tion........................................ 3.7-1 3.7.2 Temporary Conversion of Leased Arers of the X-705 Facility from NRC to DOE Regulation............................... 3.7-2 3.7.3 Organization and Responsibilities......................... 3.7-3 A PPENDIX A (Deleted)............................................ A-1 l

VI

-SAR-PORTS January 19,1996 (sv)

Rev.2 Volume Limit Maximum 235U Enrichment (wt_ nercent) 55-Gallon 1.4 30-Gallon 2.0 22-Quart 5.5 14-Quart 10.0 5-Quart not limited The bases for the container sizes are provided in each NCSE prepared for those operadons requiring containers. Specific details of these bases can be obtained by referring to the particular NCSE of concern.

Interaction Interacdon is controlled by spacing items bearing fissile material when those items could result in a criticality accident if not properly spaced. The spacing necessary to maintain a safe array of fissile material units is determined in the NCSE performed for the array. The spacing requirements are documented in the NCSA for the operation. The amount of spacing needed between items is determined based on analysis of the normal and credible abnormal process upset condidons for the particular operadon. The basis for the spacing is documented in PORTS NCSEs. Other spacing requirements I

(~N are applied on a case-by<ase basis, depending on the results of a given NCSE.

.Q, Geometry Geometry control is applied by limiting equipment dimensions for those systems which depend on the geometry for criticality safety. He geometry is determined in the NCSE which is performed for each system and depends on the normal and credible abnormal process upsets condidons related to the specific system. For example, the following dimensions are used in geometry controls:

Volume Limit Maximum 235U Enrichment (wt percent) 5-Inch Depth 5

3.5-Inch Depth 10 1.5-Inch Depth not limited 10.25-Inch Inside Diameter (ID) 5 8.2-Inch ID 10 5-Inch ID not limited The inside diameters are for geometries of unspecified length. Containers with larger inside diameters j

and limited heights are used as evaluated in NCSEs. Geometry controls are specified in the NCSAs.

O 1

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SAR-PORTS August 6,1998

(

Rev. 23 Mass Mass controls are applied on a case-by-case basis deper. ding on the fissile material operation involved.

The acceptable mass is determined based on the specific NSE perfo1med for the operation. The safe mass value depends on many factors including the geometry, f, #U enrichment, composition, etc. The safe mass values are communicated to the operating personne' via the NCSAs.

A typical operating limit is 350 g "U, regardless of enrichment. A maximum mass of 760 grams "U is considered suberidcal, as recognized by ANSI /ANS-8.1. If under an abnormal condition the mass would not exceed 760 grams "U, the operadon would be considered to be subcritical.

Enrichment Uranium-contammg material on the plantsite with "U enrichment s I wt. percent is considered incapable of supporting a nuclear chain reaction, but interaction of such materials with materials of higher enrichment is taken into consideration in the specific NCSE for those operations which involve material enriched to sI wt. percent.

At PORTS, the maximum "U enrichment of product UF. is 10.0 wt. percent. Small quantities of higher enriched uranium may be present on the plantsite in the form of laboratory samples and standards for instrument calibration purposes and in Department of Energy (DOE) operations as described below.

DOE operations involving greater than 10.0 wt. percent "U on plantsite are ongoing. For NCS, at interfaces between uranium enriched to greater than 10.0 wt. percent "U and uranium of less than 10.0 wt.

percent"U, either specific controls are present to limit enrichment to less than 10.0 wt. percent "U, or the possibility of the higher enrichment is addressed in the NCS evaluation for the operation. When 5", 8", and I

12" cylinders with residual UF of greater than 10.0 wt. percent "U are cleaned in X-705, the HEU solutions generated are mixed with LEU solution and double sampled to confirm the "U enrichment is less than 10.0 wt. percent prior to transfer to a NRC-regulated area. In some cases, equipment removed from X-326 could contain residual deposits of HEU. The HEU deposits individually will not be of NCS concern since the HEU suspension project will reduce HEU deposits to below a minimum critical mass, including measurement error for the enrichment of the deposit. In X-705, this equipment would be disassembled, and the bulk of the deposit removed to favorable geometry containers. The equipment would then be decontaminated further in the tunnel cleaning operation. Small amounts of HEU may have entered the seal exhaust oil and trapping material in X-326 during initial HEU refeed and during HEU suspension activities. Due to the large quantity of LEU present, the enrichment would not exceed 10.0 wt. percent in these materials. The operations involved were originally approved for enrichments of up to 100 wt. percent "U.

The maximum "U enrichment for each operation is established by the specific NCSE. The NCSA shall i

specify the maximum acceptable enrichment for each operation. Credible process upset conditions which could alter the SU enrichment shall also be considered in the NCSEs. Many operations at PORTS were l

previously approved for enrichments higher than 10.0 wt. percent "U. Those operations evaluated assuming this higher enrichment will have a larger margin of safety than indicated in the evaluation, due to the lower enrichment of the uranium actually being processed.

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5-2 12 A-2 12 1-13 1

5-3 2

B-1 1

1-14 1

5-4 2

B-2 1

1-15 1

5-5 15 C-1 1

1-16 1

5-6 2

C-2 1

1-17 1

5-7 12 D-1 19 18 1

5-8 2

D-2 1

21 1

5-9 15 D-3 2

2-2 1

5-10 15 D-4 1

2-3 19 5-11 3

E-1 19 2-4.

12 5-12 1

E-2 19 O

e i

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