ML20247K819
| ML20247K819 | |
| Person / Time | |
|---|---|
| Site: | Portsmouth Gaseous Diffusion Plant |
| Issue date: | 05/11/1998 |
| From: | John Miller UNITED STATES ENRICHMENT CORP. (USEC) |
| To: | Paperiello C NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| GDP-98-0100, GDP-98-100, NUDOCS 9805220213 | |
| Download: ML20247K819 (22) | |
Text
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- 9 A Global Energy Company JAMES H. MILLER Dim: (301) 564-3309 VICE PRESIDENT, PRODUCTION fax: (301)571-8279 May 11,1998 GDP 98-0100 Dr. Carl J. Paperiello Director, Office of Nuclear Material Safety and Safeguards Attention: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Portsmouth Gaseous Diffusion Plant (PORTS)
Docket No. 70-7002 Transmittal of Revision 20 to Portsmouth Certification Application
'p U/
Dear Dr. Paperiello:
In accordance with 10 CFR Part 76, the United States Enrichment Corporation hereby submits twenty (20) copies of Revision 20 (April 27,1998) to USEC-02, Application for United States l
Nuclear Regu:atory Commission Certification, Portsmouth Gaseous Diffusion Plant.
Revision 20 incorporates changes to the Safety Analysis Report (SAR) and Fundamental Nuclear l
Materials Control Plan (FNMCP). These changes were previously submitted for Nuclear Regulatory l
Commission (NRC) review in accordance with 10 CFR 76.45 and were approved as Amendment 10 to the Certificate of Compliance GDP-2 in your letter dated April 22,1998 (TAC NO. 32056).
Revision 20 was implemented effective April 27,1998. Also included with this transmittal is SAR page 6.1-15/16 which provides an improved copy of the Revision 19 version of SAR Figure 6.1-1.
The Revision 20 changes to the FNMCP contain confidential commercial or financial information or trade secrets that are exempt from public disclosure pursuant to Section 1314 of the Atomic Energy Act of 1954 (AEA), as amended, and 10 CFR 2.790 and 9.17(a)(4). In accordance with 10 CFR 76.33(e) and 2.790(b), the Revision 20 changes to this plan are being submitted under separate cover (USEC Letter GDP 98-0101).
9805220213 980511 PDR ADOCK 07007002
(
C PDR
.,)NOUk 6903 Rocidedge Drive. Bethesda, MD 20817-1818 Telephone 301-564-3200 Fax 301-564-3201 http://www.usec.com
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0$ces in Livermore. CA Paducah. KY Portsmouth. OH Washington. DC
l Dr. Carl J. Paperiello f3 May 11,1998 V
GDP 98-0100, Page 2 Should you have any questions regarding Revision 20, please contact Steve Routh at (301) 564-3251.
There are no new commitments contained in this submittal.
l Sincerely,
,,e
,' I Jamed., iller Vic'e President, Production
Enclosures:
- 1. Affidavit' l
- 2. USEC-02, Application for United States Nuclear Regulatory Commission l
Certification, Portsmouth Gaseous Diffusion Plant, Revision 20, Copy Numbers I through 20 cc: NRC Region III Office USEC-02, Copy Nos. 21,172 NRC Resident Inspector - PGDP USEC-02, Copy No. 22 j
NRC Resident Inspector - PORTS j
Mr. Joe W. Parks, DOE USEC-02, Copy Nos. 24 through 28 l
l l
o
=
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GDP 98-0100
/-
Page1of1 OATH AND AFFIRMATION
. I, James H. Miller, swear and affirm that I am Vice President, Production, of the United States L
Enrichment Corporation (USEC), that I am authorized by USEC to sign and file with the Nuclear Regulatory Commission Revision 20 (April 27,1998) to USEC-02, Application for United States Nuclear Regulatory Commission Certification, Portsmouth Gaseous Diffusion Plant, that I am familiar with the contents thereof, and that the statements made and matters set forth therein are true and correct to the best of my knowledge, information, and belief.
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[ James H. Miller LJ On this 1Ith day of May,1998, the officer signing above personally appeared before me, is known by me to be the person whose name is subsccibed to within the instrument, and acknowledged that he executed the same for the purposes therein contained.
In witness hereofI hereunto set my hand and official seal.
- amu /h.Nuehu-.
J!,aurie M. Knisley, Notary Publicj State of Maryland, Montgomery County p
My commission expires March 1,2002 v
APPLICATION FOR UNITED STATES NUCLEAR REGULATORY COMMISSION CERTIFICATION PORTSMOUTH GASEOUS DIFFUSION PLANT USEC-02 REMOVE / INSERT INSTRUCTIONS REVISION 20 Remove Pages Insert Pages VOLUME 1 LIST OF EFFECTIVE PAGES i/ii i/ii xiii/xiv xiii/xiv l
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LIST OF FIGURES l
Chapter 2 (Continued)
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h East 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 2.611 Superimposed Results for Fernald and PORTS........................ 2.6-26 l
2.6-12 Recommended Fernald and PORTS Seismic Hazard Curve................ 2.6-27 TABLE OF CONTENTS Chapter 3 East 3.0 FACILITY AND PROCESS DESCRIPTION..........................
3.0-1 3.1 CASCADE SYSTEMS....................................
3.1-1 l
I 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 1
3.2 UF, FEED, WITIIDRaWAL, SAMPLING, HANDLING, AND CYLINDER STORAGE FACILITIES AND SYSTEMS.............
3.2-1 3.2.1 Cascade UF, Feed and Sampling Systems................... 3.2-3.2.2 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 l
3.2.5 UF, Cylinder Shipping and Receiving.................... 3.2-60 I
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 V
SAR-PORTS April 27,1998 Rev.20 TABLE OF CONTENTS l
Chapter 3 (Continued)
{
fait 3.3 URANIUM RECOVERY AND CHEMICAL SYSTEMS.............
3.31 3.3.1 X-705 Decontamination and Recovery Facility............... 3.3-1 3.3.2 X-705 Waste Water Treatment Facility................... 3.3-42 3.3.3 Contaminated Storage Facilities........................ 3.3-52 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 Plant Water System
................................. 3.4-7 3.4.3 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 HF/F Systems.................................... 3.4-3 8 2
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 3.5.3 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 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 3.6.3 Environmental Protection Facilities..................... 3.6-22 3.7 HEU DOWNBLENDING ACTIVITIES......................... 3.7-1 3.7.1 Descri ptio n....................................... 3.7-1 3.7.2 Organization and Responsibilities........................ 3.7-3 l
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l SAR-PORTS April 27,1998 Rev.20 After a cylinder has been fed, it is removed from the feed position and weighed to determine the amount of uranium fed to the LEU enrichment process. A relatively small amount of non-volatile uranium typically remains in the cylinders after feeding. This " heel" is removed by a cleaning process conducted in a DOE-regulated X-705 Small Cylinder Cleaning area or shipped offsite for cleaning. Solutions resulting from the cleaning process are blended with solutions containing normal, depleted or LEU 'o reduce the assay to less than 10 wt-% 25U. 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 oneration of the gaseous diffusion process, cells are treated with oxidant gases to remcve deposits of urang fluoride and other compounds from the 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 uranium from deposits. The treatment gases, including any uranium 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 was left in USEC process equipment following completion of the DOE cleanup process. This may occur at any point during the remaining operational life cf the enrichment cascade. The liberated HEU material will mix with the LEU material in the process equipment and surge drums and the treatment 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 enrichments. This process ensures that the blended stream remains within the usU possession limits defined 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.
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 leased areas contain moderately (10-20 wt-% "5U) enriched or highly enriched uranium due to the presence of residual deposits of material that were not completely removed during the HEU Suspension program. On those limited occasions when this occurs, the equipment will be disassembled and decontaminated in an area in the X-705 Building which is placed temporarily under DOE regulation with appropriate safeguards in place. Material removed which exceeds 10 wt-% usU will be retained by DOE or will be blended with LEU solution until the overall enrichment is less than 10 wt-% "'U. DOE regulation and associated safeguards will cease to be applied when material equal to or greater than 10 wt-% "'U is no 1,onger present. The blended-down solution would be processed through uranium recovery as described above.
Another situation requiring the use of the X-705 building is for replacement of inoperable valves on HEU cylinders. In this case, the X-705 South Annex will be temporarily converted from NRC regulation to DOE regulation, the cylinders transferred to the area, the valves replaced, the cylinders transferred back to a DOE regulated storage or refeed area, and the X-705 South Annex will be transitioned back to NRC regulation. No fissile material will be removed from the cylinders during this process except any removable material that is in, or on, the replaced valves or contamination removed from the outside surface of the HEU cylinders. While the X-705 South Annex is temporarily converted to DOE 3.74 I
SAR-PORTS Rev.20 April 27,1998 9
regulation, access to the DOE regulated areas is controlled in accordance with the DOE Regulatory Oversight Agreement (ROA).
3.7.2 Organization and Responsibilities DOE will retain regulatory autority over HEU, except for inaccessible residual holdup and Category III quantities (or less) of other HEU. Up to 50 kg of usU, contained in uranium enriched from 10 wt-% vp to 20 wt-% e5U, may be present in Units X-25-7 and X-27-2, interconnecting piping and the X-326 surge drums in the gas phase during routine operations, as a result of HEU refeed. This equipme its inventory and operation, is covered by the NRC Certificate and is under NRC regulatory autority; the material presence is covered by a Compliance Plan item which specifies that the enrichment will decrease to below Category III limits upon completion of the HEU refeed. When HEU quantities greater than Category III are eliminated from PORTS (other than residual 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 greater than Category III be physically removed from USEC equipment subsequent to expiration of the DOE ROA, de HEU will be the responsibility of DOE, per the
}
Memorandum of Agreement between USEC and DOE dated December 15, 1994, and the Regulatory Approach for Post NRC Certification of Gaseous Diffusion Plants (AV Parks to GP Rifakes, October 11, 1995).
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 120 days after initial certification. Such facilities and areas include the X-705 Small Cylinder Cleaning area; such activities include the movement of HEU along USEC leased roadways and HEU refeed.
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. Should the HEU suspension program (including the processing through X-705) extend beyond the point at which NRC assumes regulatory oversight, the DOE ROA will continue as 6e regulatory basis for suspension activities in X-326, and the entire X-705 facility; this possibility will be addressed in the DOE Compliance Plan.
As long as HEU refeed activities are conducted in X-326, the DOE ROA will be used for regulation of refeed activities in X-M6. A segregated DOE-regulated area in X-705 is being modified to allow cleaning of all sizes of cylinders following refeed. An additional segregated area in the X-705, temporarily converted to DOE regulation, will be used to change-out HEU cylinder valves. Processing of HEU materials in other areas of X-705 is expected to be complete before NRC assumes regulatory oversight, and after that the DOE ROA will apply to identified, segregated areas in X-705 for remaining l
refeed support activities. However, if refeed support activities involving quantities of HEU greater than Category III levels continue, which are not confined to se segregated areas, then be DOE ROA will cover l
the entire X-705 facility. These activities will be identified in the DOE Compliance Plan.
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XIV
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SAR-PORTS Rev.19 April 15,1998 m
i)
- V LIST OF FIGURES Chapter 2 (Continued)
Figure 1
g 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 Hazrrd Curve......................................... 2.6-24 l
2.6-10 TERA - PORTS Seismic Hazard Curve
- 2. 6-2 5 2.6-11 Superimposed Results for Fernald and PORTS........................ 2.6-26 l
l 2.6-12 Recommended Fernald and PORTS Seismic Hazard Curve................ 2.6-27 TABLE OF CONTENTS Chapter 3 l
l l
Pagt 3.0 FACILITY AND PROCESS DESCRIPTION..........................
3.0-1 3.1 CASCADE 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 I
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 1
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 Con trols........................................ 3.2-74 O
1 Y
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1 SAR-PORTS f
April 27,1998
{
Rev.20 TABLE OF CONTENTS Chapter 3 (Continued)
PER 3.3 URANIUM RECOVERY AND CHEMICAL SYSTEMS.............
3.3-1 3.3.1 X-705 Decontamination and Recovery Facility............... 3.3-1 3.3.2 X-705 Waste Water Treatment Facility................... 3.3-42 3.3.3 Contaminated Storage Facilities........................ 3.3-52 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 Plant Water System................................. 3.4-7 3.4.3 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 HF/F Systems.................................... 3.4-3 8 2
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 3.5.3 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 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 3.6.3 Environmental Protection Facilities...................... 3.6-22 3.7 HEU DOWNBLENDING ACTIVITIES......................... 3.7-1 3.7.1 Description....................................... 3.7-1 3.7.2 Organization and Responsibilities........................ 3.7-3 l
A P P END IX A................................................. A-1 O
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SAR-PORTS April 15,1998 Rev.19 i
i Safety, Safeguards and Quality Manager. From a technical perspective, the Plant Operations Review Committee (PORC) and its associated subcommittees provide the mechanism for evaluating and integratin E&H and SS&Q program elements from a plant design and program change perspective.
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