ML20211K461
| ML20211K461 | |
| Person / Time | |
|---|---|
| Site: | Portsmouth Gaseous Diffusion Plant |
| Issue date: | 08/31/1999 |
| From: | Toelle S 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-99-0153, NUDOCS 9909070226 | |
| Download: ML20211K461 (45) | |
Text
E USEC j
. A Global Energy Company August 31,1999 GDP 99-0153 Dr. Carl J. Paperiello Director, Office of Nuclear Material Safety and Safeguards Attention: Document Control Desk U.S. Nuclear Regulatory Conunission Washington, D.C. 20555-0001 Portsmouth Gaseous Diffusion Plant (PORTS)
Docket No. 70-7002 Certificate Amendment Request - Update the Application Safety Analysis Report - Proposed Changes
Dear Dr. Paperiello:
By letter dated October 31,1997 (Reference 1), USEC submitted a certificate amendment request (CAR) containing the Safety Analysis Report Update (SARUP) required by Issue 2 of DOE /ORO-i 2027," Plan for Achieving Compliance with NRC Regulations at the Portsmouth Gaseous Diffusion Plant" (the Compliance Plan), for NRC review and approval. By letters dated April 30,1998 (Reference 2), October 19,1998 (Reference 3), November 20,1998 (Reference 4), May 10,1999 (Reference ij, June 1,1999 (Reference 6), and July 22,1999 (Reference 7), USEC submitted proposed changes to the SARUP certificate amendment request. This letter provides additional changes to the SARUP certificate amendment request. These changes to the SARUP certificate amendment request include changes that were discussed in USEC's letter to NRC dated February 26,1999 (Reference 8), changes in response to NRC questions, and changes made in accordance with item 5 of the Plan of Action and Schedule for Compliance Plan Issue 2. to this letter provides a detailed description of the proposed changes. Revised SARUP pages are provided in Enclosure 3. Revisions are noted by a revision bar in the right-hand page margin for the TSRs, and by a revision bar in the left-hand margin for the other SARUP sections.
. The conclusions stated in Enclosure 2 to Reference 1, that the proposed changes associated with the CAR are significant, are not affected by this revision and thus no significance determination is provided.
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6903 Rockledge Drive, Bethesda. MD 20817-1818 Telephone 301-564-3200 Fax 301-564-3201 http://www.usec.com
. Of0ces in Livermore, CA Paducah, KY Portsmouth, OH Washington, I)C
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.e Dr. Carl J. Paperiello.
August'31,1999;
. GDP 99-0153, Page 2 -
LAny questions regarding'this matter should be directed to Mark Smith at (301) 564-3244. -There are no new commitments contained in this submittal.
- Sincerely,
- s. A.
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Steven A. Toelle -
Nuclear Regulatory Assurance and Policy Manager
References:
- 1. Letter from James H. Miller (USEC) to Dr. Carl J. Paperiello (NRC), Certificate Amendment Request - Update the Application Safety Analysis Report, USEC
. Letter GDP 97-0189, October 31,1997.
- 2. Letter from James H. Miller (USEC) to Dr. Carl J. Paperiello (NRC), Certificate Amendment Request '- Update the Application Safety Analysis Report - Proposed
. Changes, USEC Letter GDP 98-0096, April 30,1998.
- 3. _ Letter from Steven A. Toelle (USEC) to Dr. Carl J. Paperiello (NRC), Certificate Amendment Request - Update the Application Safety Analysis Report - Proposed Changes, USEC Letter GDP 98-0212, October 19,1998.
- 4. Letter from Steven A. Toelle (USEC) to Dr. Carl J. Paperiello (NRC), Certificate Amendment Request - Update the Application Safety Analysis Report - Proposed Changes, USEC Letter GDP 98-0251, November 20,1998.
- 5. Letter from Steven A. Toelle (USEC) to Dr. Carl J. Paperiello (NRC), Certificate Amendment Request - Update the Application Safety Analysis Report - Proposed Changes, USEC Letter GDP 99-0076, May 10,1999.
- 6. Letter from Steven A. Toelle (USEC) to Dr. Carl J. Paperiello (NRC), Certificate Amendment Request - Update the Application Safety Analysis Report - Proposed Changes, USEC Letter GDP 99-0084, June 1,1999.
- 7. Letter from Steven A. Toelle (USEC) to Dr. Carl J. Paperiello (NRC), Certificate Amendment Request'- Update the Application Safety Analysis Report - Proposed Changes, USEC Letter GDP 99-0120, July 22,1999, p
p e.-
Dr. Carl J. Paperiello August'31,1999 GDP 99-0153, Page 3 -
References (continued):
- 8. Letter from Steven A. Toelle (USEC) to Dr. Carl J. Paperiello (NRC), Certificate Amendment Request - Update the Application Safety Analysis Report - Adoption of Current Application Technical Safety Requirements, USEC Letter GDP 99-i 0042, February 26,1999.
Enclosures:
- 1. Oath and Affirmation
- 2. United States Enrichment Corporation (USEC), Proposed Changes, Certificate 1
Amendment Request, Update the Application Safety Analysis Report, Detailed Description of Change
- 3. Proposed Changes, Certificate Amendment Request, Safety Analysis Report Update, Insertion /RemovalInstructions, August 31,1999 cc: Robert C. Pierson, NRC Patrick L. Ililand, NRC Region III Office David Hartland, NRC Resident Inspector - PORTS Kenneth O' Brien, NRC Resident Inspector - PGDP
' Randall M. DeVault, DOE -
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OATH AND AFFIRMATION l
1, Steven A. Toelle, swear and affirm that I am the Nuclear Regulatory Assurance and Policy Manager of the United States Enrichment Corporation (USEC), that I am authorized by USEC to 1
sign and file with the Nuclear Regulatory Commission these proposed changes to the Safety Analysis 4
Report Update for the Portsmouth Gaseous Diffusion Plant, as described in GDP 99-0153, 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.
n S. A.
LJ Steven A.Toelle On this 31st day of August 1999, the individual signing above personally appeared before me, is known by me to be the person whose name is subscribed 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.
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b 9) ga nct M. Boothe, N6tary Public l
Mitate of Maryland, Montgomery County My commission expires June 23,2003 l
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GDP 99-0153 Page 1 of 6 United States Enrichment Corporation (USEC)
Proposed Changes
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Certificate Amendment Request
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Update the Application Safety Analysis Report J
Detailed Description of Change 1.0 Purpose The purpose of this submittal is to provide revised pages to the Safety Analysis Report Update (SARUP) previously transmitted in USEC letters GDP 97-0189, dated October 31,1997 (Reference 1), GDP 98-0096, dated April 30,1998 (Reference 2), GDP 98-0212, dated October 19,1998 (Reference 3), GDP 98-0251, dated November 20,1998 (Reference 4), GDP 99-0076, dated May 10,1999 (Reference 5), GDP 99-0084, dated June 1,1999 (Reference 6), and GDP 99-0120, dated July 22,1999 (Reference 7) for NRC review and approval.
2.0 Description of Submittal The following changes are included in this submittal which modify the latest version of the SAR f
Update certificate amendment request. The revised pages are included in Enclosure 3.
A.
The Revision Log has been updated to reflect the changes included in this revision.
i B.
The List of EfTective Pages has been updated to reflect the changes included in this revision.
C.
SARUP Chapter 1, Appendix A, Section 1.1 has been revised to take exception to the ANSI N14.1,1995 requirements for the procurement of new cylinders.
D.
SARUP Chapter 1, Appendix A, Section 1.20 has been revised to delete the reference to the X-710 radiographic facility, which has been abandoned, and include a reference to the X-326 radiographic facility.
E.
SARUP Section 5.2, Appendix A, paragraph 7.12 has been revised to change the specified administrative controls and clarify the description of applicable waste streams for contaminated metals. The changes deleted the requirements for a supervisor verification of the contaminated metal and spacing requirements for certain wastes.
F.
SARUP Section 5.2, Appendix A, paragraph 7.13 has been revised to include references to enrichment and HEPA filter installation as nuclear criticality safety controls for small UF, release gulpers.
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i GDP 99-0153 Page 2 of 6 G.
SARUP Section 5.2, Appendix A, paragraph 7.15 has been revised to reflect the conclusions from a revised nuclear criticality safety approval (NCSA) that no nuclear criticality safety controls are required for potentially contaminated laundry.
H.
A proposed revision to SARUP Section 5.6.2 is being submitted to revise the organizational responsibilities for the receipt and shipment of hazardous materials from the Materials Management Manager to the Work Control Manager.
I.
SARUP Section 5.2, Appendix A, paragraphs 5.1 and 7.14 were revised to provide clarification of existing NCS controls. No substantive change in controls was involved. In paragraph 5.1, the description of a control was changed to delete specific reference to maximum possible fill limits and instead refer to analyzed conservative quantities when dealing with containers with unknown quantities of uranium. In paragraph 7.14, the controls for uranium analysis and sampling were clarified to indicate that spacing is not controlled by design and physical integrity of storage racks except in a few locations.
J.
SARUP Section 5.2, Appendix A, paragraph 7.6, Sample Cylinder Handling and Storage, was revised to delete the administrative controls that sample cylinders may not be stacked and that n.oderating materials be precluded from marked boundaries.
K.
SARUP Section 5.2, Appendix A, paragraph 1.7 was revised to change the minimum required temperature for the surge drums from 140 degrees F to 90 degrees F.
L.
SARUP section 3.8.7.3.1 was clarified to indicate that in response to an alarm, operators would investigate to verify a release occurred and if necessary evacuate the area affected by the release.
M.
SARUP section 4.2.5 and Table 4.2-5 provide an example of a hazard state, operating mode, and initiating event matrix combination. This example was revised to provide a more accurate presentation of the results.
N.
SARUP section 4.3.2.2.8 was revised to clarify one of the location descriptions for the scenario (from " process line" to " pigtail /line").
O.
SARUP TSR section 3.9.1.b was revised to add the requirement to have procedures for administrative controls described in SAR Chapter 4.
P.
SARUP Section 5.2, Appendix A, paragraph 1.6 was revised to reference passive design features as NCS controls in the cascade datum systems.
Q.
SARUP Sections 2.7.3,3.8.6.2.3,3.8.9.1.3,4.3.2.5.2, and Table 3.8-5 were revised to reflect the results of new wind analysis. This analysis concluded that the X-326 building, X-343 l
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.O GDP 99-0153 Page 3 of 6 building, and tie lines, which were previously assumed to fail or have capacities less than the assumed 250 year evaluation basis wind (EBW), will withstand the EBW without significant damage.
R.
SARUP Section 5.2, Appendix A, paragraph 3.13 was revised to specify that NCS controls previously in place for C-Area and Tunnel in the X-705 building now apply to general material handling and storage throughout X-705.
3.0 Basis for the Revision
[ltem C] The change to take exception to the ANSI N14.1,1995 requirements for the f
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_ procurement o new cy nders s necessary because of an erroneous change that was made to the 1995 standard. As part of the changes to ANSI N14.1 that were made for the 1995 edition, an attempt was made to clarify manufacturing practices associated with packing nuts for 1 inch cylinder valves. A paragraph was moved from section 6.15.2 to section 6.15.4 that specified stress-relieving requirements for packing nuts. Annealing temperatures and tolerances were specified for the three allowable materials. Finally, a new surface hardness requirement of RB70-RB80 was imposed. This last requirement was erroneously imposed on all three materials. The surface hardness requirement should only have been imposed on the 636 rnaterial. Ilowever, the 636 material is the least reliable choice from a safety standpoint. The other materials cannot meet the new hardness requirement imposed by the 1995 edition. Additionally, this section is being revised to indicate that 48 inch cylinders are purchased in accordance with the Certificate of Compliance for the Paducah Tiger Overpack. Thus, an exception to the 1995 edition of the standard for new 48 inch cylinders and new 1 inch valves is necessary. The change specifies that these cylinders and valves meet the 1990 edition.
[ Item D] The change to delete the reference to the X-710 radiographic facility is necessary since this facility is no longer in service. A reference to the X-326 radiographic facility has been inserted with identical commitments to ANSI N543 as were listed for the X-710 facility.
[ Item E] The change to SARUP section 5.2, Appendix A, paragraph 7.12 reflects changes in nuclear criticality safety approval (NCSA) administrative controls for contaminated metals. The changes deleted the requirements for a supervisor verification of the contaminated metal and spacing requirements for certain wastes. These controls are no longer necessary given the other controls that are in place including checks for visible contamination and free liquids. Double contingency j
continues to be maintained. This change also clarified the description of the applicable waste streams to state that certain items, such as building debris, concrete, glass, etc. are no longer limited l
to only ten percent of the total waste stream in the " contaminated metals" category.
[ Item F] The change to SARUP section 5.2, Appendix A, paragraph 7.13 reflects the addition of two controls to the NCSAs for small UF release gulpers. The additional controls involve 6
enrichment and HEPA filter installation.
1 GDP 99-0153 i
Page 4 of 6
[ Item G] The change to SARUP section 5.2, Appendix A, paragraph 7.15 is based on analysis that concludes that there are insignificant levels of contamination associated with laundry operations, except where there is visual contamination. In cases where personnel protective equipment (PPE) have visible (beyond fixed stains / films) amounts of uranium on or inside them or are contaminated with uranium bearing solution, they will be disposed ofin containers controlled under separate NCSA requirements.
[ Item H] The change to SARUP Section 5.6.2, to revise the organizational responsibilities for the receipt and shipment of hazardous materials from the Materials Management Manager to the Work Control Manager, is being made to provide for a more efficient management structure. This change also affects the Radioactive Material Packaging and Transportation Quality Assurance Program (PTQAP). A specific request for NRC review of this PTQAP change was submitted to the NRC in a separate correspondence (GDP 98-0234, dated November 16,1998). NRC approved this change to the PTQAP on November 24,1998 (Reference 8).
[ Item I] The changes to SARUP Section 5.2, Appendix A, paragraphs 5.1 and 7.14 involved 3
only minor clarifications in the SARUP wording to provide consistency with the applicable NCSAs.
Double contingency controls continue to be satisfied. No actual change to operational controls was involved.
[ Item J] The changes to SARUP Section 5.2, Appendix A, paragraph 7.6 deleted requirements that were determined by the NCSA to be unnecessary in demonstrating double contingency.
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[ Item K] The change to SARUP Section 5.2, Appendix A, paragraph 1,7 to revise the minimum surge drum temperature from 140 to 90 degrees F was made to more accurately reflect the NCSA control of moderation and to provide more operating flexibility. The temperature of 90 i
degrees F represents the maximum temperature that HF can condense assuming a 25 psia surge drum pressure (based on the maximum booster outlet pressure of 25 psia).
[ Item L] This change was necessary to reflect actual plant operating practices. Unless the release is observed visually, operators investigate alarms before evacuating the area.
[ Item M] The revision to the example matrix is informational only and is not of any safety consequence.
[ Item N] The revision to one of the location descriptions (from " process line" to
" pigtail /line") is necessary to provide consistency with the scenario and supporting analysis. No changes were made to any of the scenario assumptions.
[ Item 0] The revision to the TSR administrative controls adds a new requirement.
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GDP 99-0153 Page 5 of 6
[ Item P] The revision to SARUP section 5.2, Appendix A, paragraph 1.6 reflects changes made to the NCSA for the cascade datum systems. The new NCSA takes credit for passive design features, such as the volume of the pump, and no longer solely relies on administrative controls.
[ Item Q] The revision to SARUP to reflect the results of new wind analysis was necessary to incorporate changes to the 250-yecr return wind speed and to restate the ability of certain site facilities to withstand that wind speed. This new wind analysis was prepared by the Center for Natural Phenomena Engineering (LMES, Oak Ridge, TN). New wind loading and resultant building capacities were developed for the X-343, X-326, and process building tie lines. The results of this analysis is that no failures are predicted for these two buildings and tie lines.
[ Item R] The revision to SARUP Section 5.2, Appendix A, paragraph 3.13 was necessary to support changes to the applicable NCSA which now applies NCS controls for potentially uranium bearing storage / handling ofequipment previously applicable to only C-Area and the Tunnel to other areas ofX-705. This change did not result in a significant change in the handling of uranium bearing material since uranium bearing material (in equipment and containers) is present in virtually all areas of X-705 and is analyzed / controlled under other NCSAs.
In each of the above cases where nuclear criticality safety controls were revised, the specific NCSA is available onsite for review.
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References 1.
Letter from James IL Miller (USEC) to Dr. Carl J. Paperiello (NRC), Certificate Amendment Request - Update the Application Safety Analysis Repont, USEC Letter GDP 97-0189, October 31,1997.
2.
Letter from James IL Miller (USEC) to Dr. Carl J. Paperiello (NRC), Certificate Amendment l
Request - Update the Application Safety Analysis Report - Proposed Changes, USEC Letter GDP 98-0096, April 30,1998.
3.
Letter from Steven A. Toelle (USEC) to Dr. Carl J. Paperiello (NRC), Certificate Amendment Request - Update the Application Safety Analysis Report - Proposed Changes, USEC Letter GDP 98-0212, October 19,1998.
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4.
Letter from Steven A. Toelle (USEC) to Dr. Carl J. Paperiello (NRC), Certificate l-Amendment Request - Update the Application Safety Analysis Report - Proposed Changes, USEC Letter GDP 98-0251, November 20,1998.
5.
Letter from Steven A. Toelle (USEC) to Dr. Carl J. Paperiello (NRC), Certificate Amendment Request - Update the Application Safety Analysis Report - Proposed Changes, USEC Letter GDP 99-0076, May 10,1999.
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E GDP 99-0153 Page 6 of 6 6.
Letter from Steven A. Toelle (USEC) to Dr. Carl J. Paperiello (NRC), Certificate Amendment Request - Update the Application Safety Analysis Report - Proposed Changes, USEC Letter GDP 99-0084, June 1,1999.
7.-
Letter from Steven A. Toelle (USEC) to Dr. Carl J. Paperiello (NRC), Certificate Amendment Request - Update the Application Safety Analysis Report - Proposed changes, USEC Letter GDP 99-0120, July 22,1999.
8.
Letter from Patricia L. Eng (NRC) to Steven A. Toelle (USEC), Revision No. 6 to the Quality Assurance Program Approval for Radioactive Material Packages No. 0832, November 24,1998.
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1 GDP 99-0153 Page 1 of 35 SAFETY ANALYSIS REPORT UPDATE CERTIFICATE AMENDMENT REQUEST August 31,1999 REVISION Remove Pages Insert Pages SARUP Revision Log '
SARUP Revision Log None iii SARUP List of Effective Pages SARUP List of Effective Pages SARUP-1 through -4, SARUP-6 through -10 SARUP-1 through -4, SARUP-6 through -10 SARUP Chapter 1, Appendix A SARUP Chapter 1, Appendix A A-1, A-6 A-1, A-la, A-6 SARUP Chapter 2 SARUP Chapter 2 2.7-4 2.7-4 SARUP Section 3.8 SARUP Section 3.8 3.8-54,3.8-60,3.8-67, Table 3.8-5 3.8-54,3.8-60,3.8-67, Table 3.8-5 SARUP Chapter 4 SARUP Chapter 4 4.2-13, 4.3-92,4.3-135, 4.3-136,4.3-137, 4.2-13, 4.3-92, 4.3-135, 4.3-136, 4.3-137, Table 4.2-5 Table 4.2-5 SARUP Section 5.2A SARUP Section 5.2A 5.2A-5, -6, 21, -27, -31, -34, -35 and -36 5.2A-5, -6, 27, -31, -34, -35 and -36 SARUP Section 5.6 SARUP Section 5.6 5.6-1 5.6-1 SARUP TSRs SARUP TSRs None 3.0-6 f
O August 31,1999 United States Enrichment Corporation Portsmouth Gaseous Diffusion Plant Safety Analysis Report Update REVISION LOG (continued) 8/31/99 Submittal to revise Chapter 1, Appendix A Section 1.1 to take exception to the ANSI N14.1,1995 requirements for the procurement of new cylinders; revise Chapter 1 Appendix A, Section 1.20 to delete the reference to the X-710 radiographic facility and include a reference to the X-326 radiographic facility; revise Section 5.6.2 to change the organizational responsibilities for the receipt and shipment of hazardous materials from the Materials Management Manager to the Work Control Manager; revise Section 5.2, Appendix A, paragraph 7.12 to change the specified administrative controls and clarify the description of applicable waste streams for contaminated metals; revise Section 5.2, Appendix A, paragraph 7.13 to specify additional nuclear criticality safety controls for small UF release gulpers; revise Section 5.2, Appendix A, paragraph 7.15 to 6
delete nuclear criticality safety control requirements for potentially contaminated laundry; revise Section 5.2, Appendix A, paragraph 5.1 to delete specific reference to maximum possible fill limits and instead refer to analyzed conservative quantities when dealing with containers with unknown quantities of uranium; revise Section 5.2, Appendix A, paragraph 7.14 to clarify the controls for uranium analysis and sampling to indicate that spacing is not controlled by design and physical integrity of storage racks except in a few locations; revise SARUP Section 5.2, Appendix A, paragraph 7.6, Sample Cylinder Handling and Storage, to delete the administrative controls that sample cylinders may not be stacked and that moderating materials be precluded from marked boundaries; revise SARUP Section 5.2, Appendix A, paragraph 1.7 to change the minimum required temperature for the surge drums from 140 degrees F to 90 degrees F; revise SARUP section 3.8.7.3.1 to indicate that in response to an alarm, operators would investigate to verify a release occurred and if necessary evacuate the area affected by the release; revise SARUP section 4.2.5 and Table 4.2-5 to provide a more accurate presentation of the results; revise SARUP section 4.3.2.2.8 to clarify one of the location descriptions for the scenario (from " process line" to " pigtail /line");
revise SARUP TSR section 3.9.1.b to add the requirement to have procedures for administrative controls described in SAR Chapter 4; revise SARUP Section 5.2, Appendix A, paragraph 1.6 to reference passive design features as NCS controls in the cascade datum systems; revise SARUP Sections 2.7.3, 3.8.6.2.3, 3.8.9.1.3, 4.3.2.5.2, and Table 3.8-5 to reflect the results of new wind analysis; revise SARUP Section 5.2, Appendix A, paragraph 3.13 to specify that NCS controls previously in place for C-Area and Tunnel in the X-705 building now apply to general material handling and storage throughout X-705.
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SARUP-PORTS August 31,1999 LIST OF EFFECTIVE PAGES REVISION LOG CIIAPTER 2 (Continued)
Eagg RAC/ Revision /Date Eage RAC/Pevision 3
i July 9,1999 2.3-7 RAC 97-X0248 (RO) ii July 9,1999 2.3-8 RAC 97-X0248 (RO) iii Angust 31,1999 2.3-9 RAC 97-X0248 (RO) 2.3-10 RAC 97-X0248 (RO)
CIIAPTER 1, APPENDIX A 2.3-11 RAC 97-X0248 (RO) 2.3-12 RAC 97-X0248 (RO)
Eagg RAC/ Revision 2.3-13 RAC 97-X0248 (RO)
A-1 RAC 97-X0506 (RO) 2.3-14 RAC 97-X0248 (RO)
RAC 98-X0050 (RO) 2.3-15 RAC 97-X0248 (RO)
A-la RAC 98-X0050 (RO) 2.3-16 RAC 97-X0248 (RO)
A-2 RAC 97-X0506 (RO) 2.3 17 RAC 97-X0248 (RO)
A-3 RAC 97-X0506 (RO) 2.3-18 RAC 97-X0248 (RO)
A-4 RAC 97-X0506 (RO) 2.3-19 RAC 97-X0248 (RO)
RAC 98-X0130 (RO) 2.3-20 RAC 97-X0248 (RO)
A-5 RAC 97-X0506 (RO) 2.3-21 RAC 97-X0248 (RO)
RAC 98-X0130 (RO) 2.4-2 RAC 97-X0248 (RO)
A-6 RAC 97-X0506 (RO) 2.4-6 RAC 97-X0248 (RO)
RAC 97-X0093 (RO) 2.4-7 RAC 97-X0248 (RO) l A-7 RAC 97-XO506 (RO) 2.4-8 RAC 97-X0248 (RO)
A-8 RAC 97-X0506 (RO) 2.4-9 RAC 97-X0248 (RO) l A-9 RAC 97-X0506 (RO) 2.4-11 RAC 97-X0248 (RO) l A-10 RAC 97-X0506 (RO) 2.6-1 RAC 97-X0248 (RO)
A-11 RAC 97-X0506 (RO) 2.6-2 RAC 97-X0248 (RO)
A-12 RAC 97-X0506 (RO) 2.6-3 RAC 97-X0248 (RO) 2.6-4 RAC 97-X0248 (RO)
CHAPTER 2, CONTENTS 2.6-5 RAC 97-X0248 (RO) 2.6-6 RAC 97-X0248 (RO)
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CIIAPTER 2 2.6-13 RAC 97-X0248 (RO) 2.7-1 RAC 97-X0248 (RO)
Eagg RAC/ Revision 2.7-2 RAC 97-X0248 (RO) 2.1-6 RAC 97-X0248 (RO) 2.3-1 RAC 97-X0248 (RO) 2.3-2 RAC 97-X0248 (RO) 2.3-3 RAC 97-X0248 (RO) 2.3-4 RAC 97-X0248 (RO) 2.3-5 RAC 97-X0248 (RO) 2.3-6 RAC 97-X0248 (RO)
SARUP-1
SARUP-PORTS August 31,1999 LIST OF EFFECTIVE PAGES CfIAPTER 2, (Continued)
CIIAPTER 3.8 (Continued)
Eagg RAC/ Revision Eagg RAC/ Revision 2.7-3 RAC 97-X0248 (RO) 3.8-15 RAC 97-X0506 (R1)
'2.7-4 RAC 97-X0248 (RO) 3.8-16 RAC 97-X0506 (R1)
RAC 99-X0079 (RO) 3.8-17 RAC 97-X0506 (R1) 2.7-5 RAC 97-X0248 (RO) 3.8-18 RAC 97-X0506 (R1) 2.7-6 RAC 97-X0248 (RO) 3.8-19 RAC 97-X0506 (RI) 2.7-7 RAC 97-X0248 (RO)-
3.8-20 RAC 97-X0506 (R1) 2.8-1 RAC 97-X0248 (RO) 3.8-21 RAC 97-X0506 (RI) 2.9-1 RAC 97-X0248 (RO) 3.8-22 RAC 97-X0506 (R1) 2.9-2 RAC 97-X0248 (RO) 3.8-23 RAC 97-X0506 (RI) 2.9-3 RAC 97-X0248 (RO) 3.8-24 RAC 97-X0506 (R1) 2.9-4 RAC 97-X0248 (RO) 3.8-25 RAC 97-X0506 (R1) 2.9-5 RAC 97-X0248 (RO) 3.8-26 RAC 97-X0506 (R1) 2.9-6 RAC 97 X0248 (RO) 3.8-27 RAC 97-X0506 (R1) 3.8-28 RAC 97-X0506 (R1)
SECTION 3.8, CONTENTS 3.8-29 RAC 97-X0506 (R1) 3.8-30 RAC 97-X0506 (R1)
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Eage BAC/Revisinn 3.8-38 RAC 97-X0506 (R1) 3.8-1 RAC 97-X0506 (R1) 3.8-39 RAC 97-X0506 (R1) 3.8-2 RAC 97-X0506 (R1) 3.8-40 RAC 97-X0506 (R1) 3 S-3 RAC 97-X0506 (RI) 3.8-41 RAC 97-X0506 (RI) 3.8-4 RAC 97-X0506 (RI) 3.8-42 RAC 97-X0506 (R1) 3.8-5 RAC 97-X0506 (R1) 3.8-43 RAC 97-X0506 (R1) 3.8-6 RAC 97-X0506 (R1) 3.8-44 RAC 97-X0506 (R1)
RAC 97-X0558 (RO) 3.8-45 RAC 97-X0506 (RI) i 3.8-7 RAC 97-X0506 (R1) 3.8-46 RAC 97-X0506 (R1)
RAC 97-X0558 (RO) 3.8-47 RAC 97-X0506 (R1) 3.8-8 RAC 97-X0506 (RI) 3.8-48 RAC 97-X0506 (RI) 3.8-9 RAC 97-X0506 (R1) 3.8-49 RAC 97-X0506 (R1) 3.8-10 RAC 97-X0506 (R1) 3.8-50 RAC 97-X0506 (R1) 3.8-11 RAC 97-X0506 (RI) 3.8-5 i RAC 97-X0506 (RI) 3.8-12 RAC 97-X0506 (RI) 3.8-52 RAC 97-X0506 (RI) 3.8-13 RAC 97-X0506 (RI) 3.8-14 RAC 97-X0506 (RI) l 1
1 SARUP-2
SARUP-PORTS August 31,1999 LIST OF EFFECTIVE PAGES CIIAPTER 3.8 (Continued)
CIIAPTER 3.8 (Continued)
P.agt RAC/ Revision Eagg RAC/ Revision 3.8-53 RAC 97-X0506 (RI) 3.8-83 RAC 97-X0506 (RI)
RAC 98-X0130 (RO) 3.8-84 RAC 97-X0506 (RI) l
~3.8-54 RAC 97-X0506 (R1) 3.8-85 RAC 97-X0506 (R1)
~ RAC 98-X0130 (RO) 3.8-86 RAC 97-X0506 (RI)
RAC 99-X0079 (RO) 3.8-55 RAC 97-X0506 (RI)
CIIAPTER 3.8, TABLES RAC 98-X0152 (RO) 3.8-56 RAC 97-X0506 (RI)
Eagg RAC/ Revision 3.8-57 RAC 97-X0506 (RI)
T3.8-1, Sht.1 RAC 97-X0506 (RI)
RAC 97-X0440 (RO)
T3.8-1, Sht. 2 RAC 97-X0506 (R1) 3.8-58 RAC 97-X0506 (RI)
T3.8-1, Sht. 3 RAC 97-X0506 (R1)
RAC 98-X0044 (RO)
T3.8-1, Sht. 4 RAC 97-X0506 (R1) 3.8-59 RAC 97-X0506 (R1)
RAC 98-X0152 (RO)
RAC 98-X0044 (RO)
T3.8-1, Sht. 5 RAC 97-X0506 (RI)
RAC 98-X0129 (RO)
RAC 97-X0408 (RO) 3.8-60 RAC 97-X0506 (R1)
T3.8-2, Sht.1 RAC 97-X0506 (R1)
RAC 98-X0044 (RO)
T3.8-2, Sht. 2 RAC 97-X0506 (RI)
RAC 99-X0077 (RO)
T3.8-2, Sht. 3 RAC 97-X0506 (RI) 3.8-61 RAC 97-X0506 (R1)
T3.8-2, Sht. 4 RAC 97-X0506 (R1) 3.8-62 RAC 97-X0506 (RI)
T3.8-2, Sht. 5 RAC 97-X0506 (RI) 3.8-63 RAC 97-X0506 (RI)
T3.8-2, Sht, 6 RAC 97-X0506 (RI) 3.8-64 RAC 97-X0506 (RI)
T3.8-2, Sht. 7 RAC 97-X0506 (RI) 3.8-65 RAC 97-X0506 (RI)
T3.8-2, Sht. 8 RAC 97-X0506 (RI)
.3.8-66 RAC 97-X0506 (R1)
T3.8-2, Sht. 9 RAC 97-X0506 (RI) 3.8-67 RAC 97-X0506 (R1)
T3.8-2, Sht.10 RAC 97-X0506 (R1)
RAC 99-X0079 (RO)
RAC 98-X0044 (RO) 3.8-68 RAC 97-X0506 (RI)
T3.8-2, Sht.11 RAC 97-X0506 (RI)
RAC 98-X0130 (RO)
T3.8-2, Sht.12 RAC 97-X0506 (RI) 3.8-69 RAC 97-X0506 (R1)
T3.8-2, Sht.13 RAC 97-X0506 (RI)
RAC 98-X0130 (RO)
T3.8-2, Sht.14 RAC 97-X0506 (RI) 3.8-70 RAC 97-X0506 (R1)
T3.8-2, Sht.15 RAC 97-X0506 (R1) 3.8-71 RAC 97-X0506 (R1)
T3.8-2, Sht.16 RAC 97-X0506 (RI)
RAC 97-X0440 (RO)
T3.8-2, Sht.17 RAC 97-X0506 (RI) 3.8-72 RAC 97-XO506 (R1)
T3.8-3, Sht.1 RAC 97-X0506 (R1) 3.8-73 RAC 97-X0506 (RI)
T3.8-3, Sht. 2 RAC 97-X0506 (R1) 3.8-74 RAC 97-X0506 (R1)
T3.8-3, Sht. 3 RAC 97-X0506 (RI) 3.8-75 RAC 97-X0506 (R1)
T3.8-4 RAC 97-X0506 (R1) 3.8-76 RAC 97-X0506 (R1)
T3.8-5 RAC 97-X0506 (R1) 3.8-77 RAC 97-X0506 (R1)
RAC 97-X0440 (RO) 3.8-78 RAC 97-X0506 (RI)
RAC 99-X0079 (RO) 3.8-79 RAC 97-X0506 (R1) 3.8-80 RAC 97-X0506 (RI) 3.8-81 RAC 97-X0506 (R1) 3.8-82 RAC 97-X0506 (RI) i SARUP-3
SARUP-PORTS August 31,1999 LIST OF EFFECTIVE PAGES CHAPTER 4, CONTENTS CIIAPTER 4, (Continued)
Eggg RAC/ Revision Eagg RAC/ Revision i
RAC 97-X0315 (RO) 4.2-13 RAC 97-X0315 (R1)
RAC 97-X0316 (RO)
RAC 99-X0041 (RO)
RAC 97-X0314 (R1)
RAC 99-X0077 (RO) 11 RAC 97-X0315 (RO) 4.2 14 RAC 97-X0315 (R1)
RAC 97-X0316 (RO) 4.2-15 RAC 97-X0315 (R1)
RAC 97-X0314 (R1) 4.2-16 RAC 97-XO315 (R1) iii RAC 97-X0315 (RO) 4.2-17 RAC 97-X0315 (RI)
RAC 97-X0316 (RO)
RAC 99-X0041 (RO)
RAC 97-X0314 (RI) 4.2-18 RAC 97-X0315 (R1) iv RAC 97-X0315 (RO)
RAC 98-X0165 (RO)
RAC 97-X0316 (RO) 4.2-19 RAC 97-X0315 (RI)
RAC 97-X0314 (RI)
RAC 98-X0165 (RO) v RAC 97-X0315 (RO) 4.2-20 RAC 97-X0315 (RI)
RAC 97-X0316 (RO) 4.2-21 RAC 97-X0315 (RI)
RAC 97-X0314 (RI) 4.2-22 RAC 97-X0315 (R1)
RAC 97-X0505 (RS) 4.2-23 RAC 97-X0315 (R1) vi RAC 97-X0315 (RO) 4.2-24 RAC 97-X0315 (R1)
RAC 97-X0316 (RO) 4.2-25 RAC 97-X0315 (R1)
RAC 97-X0314 (R1) 4.2-26 RAC 97-X0315 (RI) 4.2-27 RAC 97-X0315 (R1)
CHAPTER 4 4.2-28 RAC 97-X0315 (RI) 4.2-29 RAC 97-X0315 (RI)
Eage RAC/ Revision 4.2-30 RAC 97-X0315 (RI) 4.1-1 RAC 97-X0315 (RO) 4.2-31 RAC 97-X0315 (RI)
RAC 99-X0041 (RO) 4.2-32 RAC 97-X0315 (R1) 4.1-2 RAC 97-X0315 (RO)
RAC 99-X0041 (RO)
CIIAPTER 4.2, TABLES 4.2-1 RAC 97-X0315 (RO) 4.2-2 RAC 97-X0315 (RO)
Eagg RAC/ Revision 4.2-3 RAC 97-X0315 (RO)
T4.2-1 RAC 97-X0315 (RO) 4.2-4 RAC 97-X0315 (RO)
T4.2-2, Sht.1 RAC 97-XO315 (RO) 4.2-5 RAC 97-X0315 (RO)
T4.2-2, Sht 2 RAC 97-X0315 (RO) 4.2-6 RAC 97-X0315 (RO)
T4.2-3 RAC 97-X0315 (RO) 4.2-7 RAC 97-XO315 (RO)
T4.2-4 RAC 97-X0315 (RO) 4.2-8 RAC 97-X0315 (RO)
T4.2-5 RAC 98-X0044 (RO) 4.2-9 RAC 97-X0315 (RO)
RAC 99-X0041 (RO) 4.2-10 RAC 97-X0315 (RO)
RAC 99-X0077 (RO) 4.2-11 RAC 97-X0315 (RO)
T4.2-6 RAC 97-X0315 (R1) 4.2-12 RAC 97-X0315 (RO)
T4.2-7, Sht.1 RAC 97-X0315 (R1)
RAC 99-X0041 (RO)
T4.2-7, Sht. 2 RAC 97-X0315 (R1)
T4.2-7, Sht. 3 RAC 97-XO315 (RI)
T4.2-7, Sht. 4 RAC 97-X0315 (R1)
T4.2-7, Sht. 5 RAC 97-X0315 (RI)
SARUP-4 l
I
?i SARUP-PORTS August 31,1999 LIST OF EFFECTIVE PAGES CHAPTER 4.3 (Continued)
CIIAPTER 4.3 (Continued)
Eagg RAC/ Revision Eagg RAC/ Revision 4.3-26 RAC 97-X0316 (RO) 4.3-58 RAC 97-X0311 (R1) 4.3-27 RAC 97-X0316 (RO)
RAC 99-X0041 (RO) 4.3-28 RAC 97-X0315 (RO) 4.3-59 RAC 97-X0311 (R1) 4.3-29 RAC 97-XO315 (RO) 4.3-6C RAC 97-X0311 (R1)
'4.3-30 RAC 97-X0315 (RO) 4.3-61 RAC 97-X0311 (R1) j 4.3-31 RAC 97-XO315 (RO) 4.3-62 RAC 97-X0311 (R1) 4.3-32 '
RAC 97-X0315 (RO) 4.3-63 RAC 97-X0311 (R1) 4.3-33 RAC 97-X0315 (RO)
RAC 99-X0041 (RO) 4.3-34 RAC 97-X0315 (RO) 4.3-64 RAC 97-X0311 (R1)
RAC 97-X0311 (RI) 4.3-65 RAC 97-X0311 (RI) 4.3-35 RAC 97-X0311 (RI)
RAC 99-X0041 (RO) 4.3-36 RAC 97-X0311 (RI) 4.3-66 RAC 97-X0311 (RI)
RAC 99-X0041 (RO)
RAC 99-X0041 (RO) 1 4.3-37 RAC 97-X0311 (RI) 4.3-67 RAC 97 X0311 (RI)
RAC 99-X0041 (RO)
RAC 99-X0041 (RO) 4.3-38 RAC 97-X0311 (RI) 4.3-68 RAC 97-X0311 (RI)
RAC 99-X0041 (RO) 4.3-69 RAC 97-XO311 (R1) i 4.3-39 RAC 97-X0311 (RI) 4.3-70 RAC 97-X0311 (RI) 4.3-40 RAC 97-X0311 (RI) 4.3-71 RAC 97-X0311 (RI)
RAC 99-X0041 (RO) 4.3-72 RAC 97-X0311 (R1) 4.3-41 RAC 97-X0311 (RI)
RAC 97-X0312 (RI)
RAC 99-X0041 (RO) 4.3-73 RAC 97-X0312 (R1) 4.3-42 RAC 97-X0311 (R1) 4.3-74 RAC 97-X0312 (R1)
RAC 99-X0041 (RO) 4.3-75 RAC 97-X0312 (R1) 4.3-43 RAC 97-X0311 (R1) 4.3-76 RAC 97-X0312 (R1) 4.3-44 RAC 97-X0311 (R1) 4.3-77 RAC 97-X0312 (R1)
RAC 99-X0041 (RO) 4.3-78 RAC 97-X0312 (RI) 4.3-45 RAC 97-X0311 (RI)
RAC 99-X0041 (RO)
RAC 99-X0041 (RO) 4.3-79 RAC 97-XO312 (R1) 4.3-46 RAC 97-X0311 (R1)
RAC 99-X0041 (RO) 4.3 RAC 97-X0311 (R1) 4.3-80 RAC 97-X0312 (R1)
RAC 99-X0041 (RO) 4.3-81 RAC 97-X0312 (R1) 4.3-48 RAC 97-X0311 (RI) -
4.3 82 RAC 97-X0312 (RI) 4.3-49 RAC 97-X0311 (R1) 4.3-83 RAC 97-X0312 (R1)
RAC 99-X0041 (RO)
RAC 99-X0041 (RO) 4.3-50 RAC 97-X0311 (RI) 4.3-84 RAC 97-X0312 (R1) 4.3-51 RAC 97-X0311 (Rl)
RAC 99-X0041 (RO)
RAC 99-X0041 (RO) 4.3-85 RAC 97-X0312 (RI) 4.3-52 RAC 97-X0311 (R1)
RAC 99-X0041 (RO) 4.3-53 RAC 97-XO311 (RI) 4.3-86 RAC 97-X0312 (RI) 4.3-54 RAC 97-X0311 (RI)
RAC 99-X0041 (RO)
RAC 99-X0041 (RO) 4.3-87 RAC 97-X0312 (RI) 4.3-55 RAC 97-X0311 (R1) 4.3-88 RAC 97-X0312 (R1) l RAC 99-X0041 (RO) 4.3-89 RAC 97-X0312 (RI) 4.3-56 RAC 97-X0311 (R1)
RAC 99-X0041 (RO)
RAC 99-X0041 (RO) 4.3-90 RAC 97-X0312 (R1) 4.3-57 RAC 97-X0311 (RI)
RAC 99-X0041 (RO)
RAC 99-X0041 (RO) 4.3-91 RAC 97-X0312 (R1)
]
i SARUP-6
SARUP-PORTS August 31,1999 LIST OF EFFECTIVE PAGES CIIAPTER 4.3 (Continued)
CIIAP'I'ER 4.3 (Continued)
Eagg RAC/ Revision Eagg RAC/ Revision 4.3-92 RAC 97-X0312 (RI) 4.3-122 RAC 97-X0312 (RI)
RAC 99-X0041 (RO) 4.3-123 RAC 97-X0312 (R1)
RAC 99-X0077 (RO) 4.3-124 RAC 97-X0312 (RI) 4.3-93 RAC 97-X0312 (RI) 4 1-125 RAC 97-X0312 (RI) 4.3-94 RAC 97-X0312 (RI) 4.3-126 RAC 97-X0312 (RI)
RAC 99-X0041 (RO)
RAC 98-X0044 (RO) 4.3-95 RAC 97-X0312 (R1)
RAC 99-X0041 (RO)
RAC 99-X0041 (RO) 4.3-126a RAC 98-X0044 (RO) 4.3-96 RAC 97-X0312 (RI) 4.3-127 RAC 97-X0312 (RI)
RAC 99-X0041 (RO) 4.3-128 RAC 97-XO312 (RI) 4.3-97 RAC 97-X0312 (RI)
RAC 99-X0041 (RO) 4.3-98 RAC 97-X0312 (R1) 4.3-129 RAC 97-X0312 (R1) 4.3-99 RAC 97-X0312 (R1)
RAC 99-X0041 (RO) 4.3-100 RAC 97-XO312 (RI) 4.3-129a RAC 99-X0041 (RO) 4.3-101 RAC 97-X0312 (R1) 4.3-130 RAC 97-X0312 (R1)
RAC 99-X0041 (RO)
RAC 98-X0044 (RO) 4.3-102 RAC 97-X0312 (R1)
RAC 99-X0041 (RO)
RAC 99-X0041 (RO) 4.3-131 RAC 97-X0312 (R1) 4.3-103 RAC 97-X0312 (RI)
RAC 97-X0313 (R1)
RAC 99-X0041 (RO)
RAC 98-X0044 (RO) 4.3-104 RAC 97-X0312 (R1) 4.3-132 RAC 97-X0313 (RO) 4.3-105 RAC 97-X0312 (R1) 4.3-133 RAC 97-X0313 (RO)
RAC 99-X0041 (RO) 4.3-134 RAC 97-X0313 (RO) 4.3-106 RAC 97-X0312 (RI) 4.3-135 RAC 97-X0313 (RO)
RAC 99-X0041 (RO)
RAC 97-X0440 (RO) 4.3-107 RAC 97-X0312 (R1)
RAC 99-X0079 (RO) 4.3-108 RAC 97-X0312 (R1) 4.3-136 RAC 97-X4313 (RO) 4.3-109 RAC 97-XO312 (R1)
RAC 97-X0440 (RO)
RAC 99-X0041 (RO)
RAC 99-X0079 (RO) 4.3-110 RAC 97-X0312 (R1) 4.3-137 RAC 97-X0313 (RO)
RAC 99-X0041 (RO)
RAC 99-X0079 (RO) 4.3-111 RAC 97-X0312 (RI) 4.3-138 RAC 97-X0313 (RO)
RAC 99-X0041 (RO)
RAC 97-X0506 (R1)
RAC 97-X0524 (RO)
RAC 99-X0041 (RO) 4.3-112 RAC 97-X0312 (RI) 4.3-139 RAC 97-X0313 (RO)
RAC 99-X0041 (RO)
RAC 97-X0506 (RI) 4.3-113 RAC 97-XO312 (R1) 4.3-140 RAC 97-X0313 (RO)
RAC 99-X0041 (RO)
RAC 97-X0506 (B.1) 4.3-114
. RAC 97-X0312 (R1) 4.3-141 RAC 97-I0713 (RO) 4.3-115 RAC 97-X0312 (RI) 4.3-142 RAC 97-X0314 (RI)
RAC 99-X0041 (RO) 4.3-143 RAC 97-X0314 (RI) 4.3-116 RAC 97-X0312 (RI) 4.3-144 RAC 97-X0314 (RI)
RAC 99-X0041 (RO) 4.3-145 RAC 97-X0314 (R1) 4.3-117 RAC 97-X0312 (R1) 4.3-146 RAC 97-X0314 (R1) 4.3-118 RAC 97-X0312 (RI) 4.3-147 RAC 97-X0314 (RI) 4.3-119 RAC 97-X0312 (RI) 4.3-148 RAC 97-X0314 (RI) 4.3-120 RAC 97-XO312 (RI) 4.3-149 RAC 97-X0314 (RI) 4.3-121 RAC 97-X0312 (RI) 4.3-150 RAC 97-X0314 (RI)
SARUP-7
l
' SARUP-PORTS August 31,1999 LIST OF EFFECTIVE PAGES CHAITER 4.3, TABLES '
CHAPTER 4.3, TABLES (Continued)
Eagg RAC/ Revision Eagg RAC/ Revision T4.3-1 RAC 97-X0315 (RO)
F4.3-28 RAC 97-X0312 (RI)
' T4.3 2 RAC 97-X0316 (RO)
F4.3-29 RAC 97-X0312 (RI)
T4.3-3 RAC 97-X0316 (RO)
F4.3-30 RAC 97-X0312 (R1)
T4.3-4 RAC 97-X0311 (RI)
F4.3-31 RAC 97-X0312 (RI)
T4.3-5 RAC 97-X0311 (RI)
F4.3-32 RAC 97-X0312 (RI)
T4.3-6 RAC 97-X0312 (RI)
F4.3-33 RAC 97-X0314 (RI)
T4.3-7 RAC 97-X0312 (R1)
T4.3-8 RAC 97-X0312 (RI)
CHAPTER 4.4 T4.3-9 RAC 97-X0312 (R1)
T4.3-10 RAC 97-X0314 (RI)
Eage RAC/ Revision T4.3 11 RAC 97-X0314 (RI) 4.4-1 RAC 97-X0315 (RO)
T4.3-12 RAC 97-X0314 (RI)
RAC 97-X0316 (RO)
T4.3 13 R' AC 97-X0315 (RO) 4.4-2 RAC 97-X0315 (RO)
F4.3-1 RAC 97-X0316 (RO)
RAC 97-X0316 (RO)
F4.3-2 RAC 97-X0316 (RO) 4.4-3 RAC 97-X0315 (RO)
F4.3-3 RAC 97-X0316 (RO)
RAC 97-X0316 (RO)
F4.3 RAC 97-X0316 (RO) 4.4-4 RAC 97-X0315 (RO)
F4.3 RAC 97-X0316 (RO)
RAC 97-X0316 (RO)
F4.3-6 RAC 97-X0316 (RO) 4.4-5 RAC 97-X0315 (RO)
F4.3-7 RAC 97-X0316 (RO)
RAC 97-X0316 (RO)
F4.3-8 RAC 97-XO316 (RO) 4.4-6 RAC 97-X0315 (RO)
F4.3-9 RAC 97-X0316 (RO)
RAC 97-X0316 (RO)
F4.3-10 RAC 97-X0311 (R1) 4.4-7 RAC 97-X0315 (RO)
F4.3-11 RAC 97-X0311 (RI)
RAC 97-X0316 (RO)
F4.3-12 RAC 97-X0311 (RI)
RAC 97-X0312 (RI)
F4.3-13 RAC 97-X0311 (RI)
RAC 97-X0314 (RI)
F4.3-14 RAC 97-X0311 (RI) 4.4-8 RAC 97-X0314 (RI)
F4.3-15 RAC 97-X0312 (R1) 4.4-9 RAC 97-X0314 (RI)
F4.3-16 RAC 97-X0312 (R1).
RAC 97-X0505 (RS)
CHAPTER 5 F4.317 RAC 97-X0312 (RI)
F4.3-18 RAC 97-X0312 (R1)
Eage RAC/ Revision F4.3-19 RAC 97-X0312 (R1) 5.2-5 RAC 97-X0506 (RO)
F4.3-20 RAC 97-X0312 (R1) 5.2 5a RAC 97-X0506 (RO)
F4.3-20 RAC 97-X0312 (RI) 5.2A-1 RAC 97-X0314 (RI)
F4.3-21 RAC 97-X0312 (R1) 5.2A-2 RAC 97-X0314 (R1) -
F4.3-22 RAC 97-X0312 (R1) 5.2A-3 RAC 97-X0314 (RI)
F4.3-23 RAC 97-X0312 (RI) 5.2A-4 RAC 97-X0314 (RI)
F4.3 24 RAC 97-X0312 (R1) 5.2A-5 RAC 97-X0314 (RI)
F4.3-25 RAC 97-X0312 (RI)
RAC 99-X0016 (RO)
F4.3 26 RAC 97-X0312 (RI) 5.2A-6 RAC 97-X0314 (R1)
F4.3-27 RAC 97-X0312 (R1)
RAC 99-X0023 (RO) 5.2A-7 RAC 97-X0314 (RI) 5.2A-8 RAC 97-X0314 (RI) l 5.2A-9 RAC 97-X0314 (R1)
SARUP-8
SARUP-PORTS August 31,1999 LIST OF EFFECTIVE PAGES CIIAPTER 5 (Continued)
TECIINICAL SAFETY REQUIREhlENTS Eage RAC/ Revision Eage RAC/ Revision 5.2A-10 RAC 97-X0314 (R1) vi RAC 97-X0505 (R5) 5.2A-11 RAC 97-X0314 (RI) vii RAC 97-X0505 (RS) 5.2A-12 RAC 97-X0314 (RI) viii RAC 97-X0505 (RS) 5.2A-13 RAC 97-XO314 (R1) ix RAC 97-X0505 (RS) 5.2A-14 RAC 97-X0314 (R1) x RAC 97-X0505 (RS) 5.2A-15 RAC 97-X0314 (R1) 1.0-8 RAC 97-X0505 (RS) 5.2A-16 RAC 97-X0314 (RI) 1.0-8a RAC 97-X0505 (R5) 5.2A-17 RAC 97-X0314 (R1) 2.1-3 RAC 97-X0505 (RS) 5.2A-18 RAC 97-X0314 (RI) 2.1-4b RAC 97-X0505 (R3) 5.2A-19 RAC 97-X0314 (R1)
RAC 97-X0505 (RS) 5.2A-20 RAC 97-X0314 (R1) 2.1-6 RAC 97-X0505 (RS) 5.2A-21 RAC 97-X0314 (RI) 2.1-8 RAC 97-X0505 (R4)
RAC 98-X0177 (RO) 2.1-10 RAC 97-X0505 (R4) 5.2A-22 RAC 97-X0314 (RI) 2.1-12 RAC 97-X0505 (RS) 5.2A-23 RAC 97-X0314 (R1) 2.1-14 RAC 97-X0505 (R4) 5.2A-24 RAC 97-X0314 (R1) 2.1-15 RAC 97-X0505 (R4) 5.2A-25 RAC 97-X0314 (RI) 2.1-17 RAC 97-X0505 (R4) 5.2A-26 RAC 97-X0314 (RI) 2.1-18 RAC 97-X0505 (R4) 5.2A-27 RAC 97-X0314 (RI) 2.1-19 RAC 97-X0505 (R3)
RAC 98-X0093 (RO) 2.1-20a RAC 97-X0505 (R4) 5.2A-28 RAC 97-X0314 (RI) 2.1-21 RAC 97-X0505 (R5) 5.2A-29 RAC 97-X0314 (RI) 2.1-22 RAC 97-X0505 (R4) 5.2A-30 RAC 97-X0314 (RI)
RAC 97-X0505 (RS) 5.2A-31 RAC 97-X0314 (RI) 2.1-23 RAC 97-X0505 (R5)
RAC 99-X0013 (RO) 2.1-24 RAC 97-X0505 (R4) 5.2A-32 RAC 97-X0314 (RI) 2.1-27 RAC 97-X0505 (R4) 5.2A-33 RAC 97-X0314 (R1) 2.1-28 RAC 97-X0505 (R4) 5.2A-34 RAC 97-X0314 (RI) 2.1-29 RAC 97-X0505 (R4)
RAC 98-X0108 (RO) 2.1-30 RAC 97-X0505 (R4) 5.2A-35 RAC 97-X0314 (RI) 2.1-30a RAC 97-X0505 (RS)
RAC 98-X0093 (RO) 2.1-30b RAC 97-X0505 (RS)
RAC 98-X0108 (RO) 2.1-30c RAC 97-X0505 (RS) 5.2A-36 RAC 97-X0314 (RI) 2.1-30d RAC 97-X0505 (RS)
RAC 98-X0037 (RO) 2.1-31 RAC 97-X0505 (R4)
RAC 98-X0108 (RO) 2.1-32 RAC 97-X0505 (R3) 5.2A-37 RAC 97-X0314 (R1)
RAC 97-X0505 (R4)
RAC 98-X0037 (RO) 2.1-33 RAC 97-X0505 (R4) 5.2A-38 RAC 97-X0314 (RI) 2.1-34 RAC 97-X0505 (RS) 5.4-2 RAC 97-X0506 (RO) 2.2-5 RAC 97-X0505 (R4) 5.4-3 RAC 97-X0506 (RO)
RAC 97-X0505 (RS) 5.4-6 RAC 97-X0506 (RO) 2.2-7a RAC 97-X0505 (R3) 5.4-7 RAC 97-X0506 (RO)
RAC 97-X0505 (RS) 5.6-1 RAC 97-X0506 (RO) 2.2-8a RAC 97-X0505 (R5)
RAC 98-X0141 (RO) 2.2-11 RnC 97-X0505 (R4) 5.6-6 RAC 97-X0506 (RO) 2.2-12 RAC 97-X0505 (R4) 5.6-7 RAC 97-X0506 (RO) 5.6-8 RAC 97-X0506 (RO)
SARUP-9
i SARUP-PORTS August 31,1999 LIST OF EFFECTIVE PAGES TECIINICAL SAFETY REQTS (continued)
TECIINICAL SAFETY REQTS (continued)
Eagg RAC/ Revision Eagg RAC/ Revision 2.2-13 RAC 97-X0505 (R4) 2.5-22a RAC 97-X0505 (R5) 2.2 RAC 97-X0505 (R4) 2.5-22b RAC 97-X0505 (RS) 2.2-16 RAC 97-X0505 (R3) 2.5-22c RAC 97-X0505 (RS) 2.2-18 RAC 97-X0505 (R3) 2.5-22d RAC 97-X0505 (RS) 2.2-19 RAC 97-X0505 (R4) 2.5-22e RAC 97-XO505 (R5) 2.2-20 RAC 97-X0505 (R4) 2.5-22f RAC 97-X0505 (RS) 2.2-21 RAC 97-X0505 (R4) 2.5-22g RAC 97-X0505 (RS)
I 2.2-23 RAC 97-X0505 (R4) 2.5-26 RAC 97-X0505 (R5) 2.2-24
- RAC 97-X0505 (RS) 2.5-27 RAC 97-X0505 (RS) 2.2-25 RAC 97-X0505 (RS) 2.5-23 RAC 97-X0505 (R4) 2.2-26 RAC 97-X0505 (R5) 2.5-24 RAC 97-X0505 (R3) 2.2-26a RAC 97-X0505 (R5)
RAC 97-X0505 (R4) 2.2-26b RAC 9-X0505 (RS) 2.5-25 RAC 97-X0505 (R4) 2.2-30 RAC 97-X0505 (R3) 2.6-5 RAC 97-X0505 (R3) 2.2-32 RAC 97-X0505 (R3) 2.6-6 RAC 97-X0505 (R3) 2.2-32a RAC 97-X0505 (RS) 2.6-7 RAC 97-X0505 (R3) 2.2-32b RAC 97-X0505 (R5) 2.6-9a RAC 97-X0505 (R3) 2.2-32c RAC 97-X0505 (RS)
RAC 97-X0505 (RS) 2.2-33 RAC 97-X0505 (R3) 2.6-11 RAC 97-X0505 (R3) 2.2 34 RAC 97-X0505 (RS) 2.6-13 RAC 97-X0505 (R3) 2.3-3 RAC 97-X0505 (R4) 2.6-14 RAC 97-X0505 (R3) i 2.3-5 RAC 97-X0505 (R4) 2.6-15 RAC 97-X0505 (R3) 2.3-6 RAC 97-X0505 (R3) 2.6-16 RAC 97-X0505 (R3) 2.3-7 RAC 97-X0505 (R4) 2.6-17 RAC 97-X0505 (R3) 2.4-4b RAC 97-X0505 (R3) 2.6-18 RAC 97-X0505 (R3)
RAC 97-X0505 (RS) 2.6-19 RAC 97-X0505 (R3) 2.4-7 RAC 97-X0505 (R4) 2.7-4 RAC 97-X0505 (R4) 2.4-8 RAC 97-X0505 (R3)
RAC 97-X0505 (RS) 2.4-9 RAC 97-X0505 (R3) 2.7-6a RAC 97-X0505 (R3) 2.4-11 RAC 97-X0505 (R4)
RAC 97-X0505 (RS) 2.5-3 RAC 97-X0505 (R4) 2.7-8a RAC 97-X0505 (R4) 2.5-5a RAC 97-X0505 (R3) 2.7-9 RAC 97-X0505 (R4)
RAC 97-X0505 (RS) 2.7-10 RAC 97-X0505 (R4) 2.5-7 RAC 97-XO505 (R4) 2.7-11 RAC 97-X0505 (R4) 2.5-8 RAC 97-X0505 (R4) 2.7-13 RAC 97-X0505 (R4)
I 2.5-9 RAC 97-X0505 (RS) 2.7-14 RAC 97-X0505 (R4) 2.5-10 RAC 97-X0505 (RS) 2.7-15 RAC 97-X0505 (R4) 1 2.5-11 RAC 97-X0505 (R3) 2.7-16 RAC 97-X0505 (R4)
I 2.5-13 RAC 97-X0505 (R3) 2.7-17 RAC 97-X0505 (R4) 2.5-15
' RAC 97-X0505 (R3) 2.7-19 RAC 97-X0505 (R4) 2.5-16 RAC 97-X0505 (R4) 2.7-21 RAC 97-X0505 (R4) 2.5-17 RAC 97-X0505 (R4) 2.7-25 RAC 97-X0505 (R3)
{
i 2.5-18 RAC 97-X0505 (R4) 2.7-27 RAC 97-X0505 (R3)
RAC 97-X0505 (RS) 2.7-28 RAC 97-X0505 (R3)
- 2.5-20 RAC 97-X0505 (R4) 2.8-Sa RAC 97-X0505 (R3) 2.5-21 RAC 97-X0505 (R4)
RAC 97-X0505 (RS) 2.5-22 RAC 97-X0505 (R4) 3.0-6 RAC 99-X0077 (RO)
SARUP-10
SAR-PORTS PROPOSED August 31,1999 RAC 97-X0506.(RO),98-X0050 (RO)
Appendix A Applicable Codes, Standards, and Regulatory Guidance This Appendix lists the various industry codes, standards, and regulatory guidance documents which have been referenced in certification correspondence. The extent to which PORTS satisfies each code, standard, and guidance document is identified below, subject to the completion of applicable actions required by the Compliance Plan.
1.0 American National Standard; Institute (ANSI) l 1.1 ANSI N14.1, Uranium Hexaflouride Packaging for Transport,1995 Edition PORTS satisfies the requirements of this standard, except for those portions superseded by Federal Regulations, with the following clarifications:
New 48-inch cylinders are purchased to ANSI N14.1 - 1990. 48-inch cylinders that were a.
already owned and operated by PORTS and were not purchased to ANSI N14.1 - 1990 were manufactured to meet the version of the ANSI standard or specification in effect at the time of the placement of the purchase order. In addition, all 48-inch cylinders satisfy Sections 4,5, 6.2.2 - 6.3.5,7, and 8 of ANSI N14.1 - 1995, except 48X cylinders used in the Paducah Tiger overpack which satisfy Sections 4,5,6.2.2 - 6.3.5,7, and 8 of ANSI N14.1 - 1990.
b.
New cylinders (other than new 48-inch cylinders) satisfy the requirements of ANSI N14.1 -
1995Property "ANSI code" (as page type) with input value "ANSI N14.1 -</br></br>1995" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process.. Cylinders (other than 48-inch cylinders) that were already owned and operated by PORTS and were not purchased to ANSI N14.1 - 1995 were manufactured to meet the version of the ANSI standard or specification in effect at the time of the placement of the purchase order and satisfy only Sections 4,5,6.2.2 - 6.3.5,7, and 8 of ANSI N14.1 - 1995.
New 1-inch cylinder valves are purchased to ANSI N14.1 - 1990. (1-inch cylinder valves are c.
not procured to ANSI N14.1 - 1995 because a wording error identified in ANSI N14.1 - 1995 relating to the allowable hardness of valve bonnet material restricts the use of all acceptable material.) 1-inch cylinder valves that were already in use at PORTS and were not purchased to ANSI N14.1 - 1990 were manufactured to meet the version of the ANSI standard or specification in effect at the time of the placement of the purchase order.
d.
New cylinder valves (other than 1 inch) are purchased to ANSI N14.1 - 1995. Cylinder valves (other than 1-inch) that were already in use at PORTS and were not purchased to ANSI N 14.1 -
1995 were manufactured to meet the version of the ANSI standard or specification in effect at the time of placement of the purchase order.
A-1 j
SAR-PORTS PROPOSED August 31,1999 RAC 97-X0506-(RO),98-X0050 (RO)
Tinning of cylinder valve and plug threads: ANSI N14.1 - 1995 requires the use of ASTM B32 c.
50A, a 50/50 tin / lead solder alloy described in the 1976 and previous editions of the ASTM standard. Cylinder valve and plug threads are tinned with solder alloys meeting the requirements of ASTM B32. Tinning is performed with nominal 50% tin alloy or with a mixture of alloys with nominal tin content from 40% to 50%, with a lower limit of 46% tin in the mix.
l f.
Section 5.2.1 - For U.S. Department of Transportation 7A Type A packaging, satisfy U.S.
Department of Energy (DOE) evaluation document DOE /RL-96 57, Revision 0, Volume 1, which supersedes DOE /00053-Hl.
g.
Use of steel or aluminum-bronze plugs in UF, cylinders is acceptable at PORTS for the following operations: heating, feeding, sampling, filling, transferring between cylinders and onsite transport and storage.
l For references to this standard, see SAR Table 3.2-1, Section 3.3.1.3.2.4, Section 3.8, Section 4.3.2.2.
1.2 ANSI /ANS 3.1, Selection, Qualification, and Training of Personnel for Nuclear Power Plants,1987 Edition PORTS satisfies only the following section of this standard:
Section 4.3.3 - The qualifications of the Radiation Protection Manager identified in SAR Section 6.1 satisfy the requirements of this section of the standard.
1.3 ANSI /ANS 3.2, Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants,1994 Edition i
I A la
SAR-PORTS PROPOSED August 31,1999
. RAC 97-X0506 (RO), 97-X0093 (RO)
Appendix C - Do not use manifold design guidelines.
Appendix D - The manifold qualification program uses this appendix as guidance only.
For references to this standard, see SAR Section 5.1.4.
1.19 ANSI N510, Testing of Nuclear Air Treatment Systems,1989 Edition New and existing fixed HEPA filter systems that satisfy the requirements of ANSI N509 and are needed to ensure compliance with release limits or to control worker radiation exposure satisfy the requirements of this standard with the following exceptions and clarifications:
Section 6.0 - Only satisfy this section for new seal-welded duct systems or for connections to a system where this section has been previou::ly applied.
Section 7.0 - Do not use guidance for monitoring frame pressure leak tests.
Existing fixed HEPA filter systems that do not satisfy the requirements of ANSI N509 will be tested using the requirements of this standard or another industry accepted standard as guidance only.
For references to this standard, see SAR Sections 5.1.4 and 5.3.2.10.
1.20 ANSI N543, General Safety Standard for Installations Using Non-Medical X-Ray and Sealed Gamma-Ray Sources, Energies up to 10 MeV,1974 Edition l
PORTS satisfies Sections 3.2,7, and 8.1.2 of this standard for the X-326, Radiographic Facility, as they apply to Enclosed Installations.
l For references to this standard, see SAR Section 3.5.1.6.2.
2.0 American Society of Mechanical Engineers (ASME) 2.1 ASME NQA-1, Quality Assurance Program Requirements for Nuclear Facilities,1989 Edition PORTS satisfies the requirements of this standard, including Basic and Supplementary Requirements, with exceptions and clarifications identified in the Quality Assurance Program Description. See also SAR Sections 6.6.12,6.8.1 and 6.8.2 and Section 7.5 of the Emergency Plan.
2.2 ASME Boiler and Pressure Vessel Code,1995 Edition PORTS satisfies the following sections of this code as clarified below:
Section VIII-The following pressure vessel components and systems satisfy the A-6
n:
SAR-PORTS PROPOSED August 31,1999 RAC 97-X0248 (RO), 99-X0079 (RO)
The wind hazard curve is shown in Figure 2.7-3 (Coats and Murray,1985). The evaluation basis wind (EBW) return period to be used for the site was specified by DOE to be 250 yr (Jackson 1995).
Another wind hazard analysis was prepared by the Center for Natural Phenomena Engineering (LMES, Oak Ridge, TN) in 1998. The study concluded that although the straight wind hazard decreased some, the current hazard curve was retained. The 250 year return period EBW was determined to be a 90-mph,3-second, peak gust with the comparable fastest straight wind speed of 75 mph. The report concluded that i
there was no change required to evaluation of PORTS facilities for high winds. Based on a return period of 250 yr EBW has a wind speed of 75 mph.
j The justification for use of an approximate 250-year return period for wind was based on the j
rationale for the seismic return period. Wind damage at the plants is less likely to result in a significant release of hazardous material than the direct failure of cascade equipment under seismic loading. Wind is more likely than seismic loads to cause exterior damage to the buildings without extensive damage internally. In addition, high winds will rapidly disperse any hazardous material released as well as reduce exposuie times down wind. Therefore the risk of serious injuries and/or deaths is substantially lower for high winds than an equivalent seismic event. Given the much lower risk of public health consequences with high wind damage than from a seismic event and the short life of the facilities, modifications would not achieve significant benefit.
Extreme wind dominates in the 250-year frequency range for the PORTS. As noted above, the extreme wind value is used in this study and tornado wind loadings are not considered. Tornados do occur in Southern Ohio; however, specific analyses of the frequency of tornados in the region show that they are rare. Recent analyses covering a 32-year period for the United Stater show an estimated strike frequency within the fenced area of the plant of approximately I event per 30,000 years at PORTS. Although tornados are extremely destructive in a localized area, the actual damage expected to cascade internal equipment and structures is also expected to be substantially less than the seismic event and may be minimal on the cell floor due to the large reservoir of air between the building roof and the cell floor of each building. Thus given the short operating life of the plants and the expectation of risk far less than a seismic event, a 250-year return period excluding tornados is believed justified.
2.7-4 l
1 SAR-PORTS PROPOSED August 31,1999 RAC 97-X0506.(RI),98-X0130 (RO),99-X0079 (RO) and testing requirements). In addition, the cranes are designed so that when the controls are released (e.g.,
i evacuation of facility event), some small additional movement occurs due to momentum after the crane drive mechanism stops and brakes are applied but these movements have no safety significance. The cranes were evaluated to assess their ability to withstand natural phenomena events. The analyses indicated that the cranes will not have any structural damage, will remain in place, and will not release their loads l during an evaluation basis earthquake and wind. Floods do not reach the elevation of the facility to threaten crane integrity.
Failure of the crane lifting components or load braking system while lifting liquid-filled UF.
cylinder could result in dropping the cylinder and rupturing the cylinder. Therefore, a load test is performed periodically.
1 l
Based on the analysis, the cranes can accomplish the required safety function.
3.8.6.2.4 System Classification The liquid UF cylinder handling cranes are required to perform the following safety function:
Prevent dropping a liquid-filled cylinder that could result in a cylinder failure event.
The cylinder failure event is classified as an EBE whose consequences could exceed the off-site EGs if the cranes were to fail in a manner that resulted in the drop and failure of a liquid-filled cylinder.
Therefore, the liquid UF handling cranes meet the criteria for classification as a Q system.
3.8.6.2.5 Boundary The Q boundaries for the liquid UF. cylinder handling cranes, including associated lifting fixtures, are defined in Table 3.8-1.
3.8.6.3 Liquid UF Cylinder Handling Equipment 3.8.6.3.1 Safety Function The liquid UF cylinder handling equipment shall not fail in a manner to cause UF primary system failure. The liquid UF cylinder handling equipment provides for the safe movement of liquid #F cylinders.
3.8.6.3.2 Functional Reauirements The liquid UF cylinder handling equipment shall be designed in accordance with the following functional requirements to ensure the capability to accomplish the required safety function:
3.8-54
4 SAR-PORTS PROPOSED August 31,1999 RAC 97 X0506 (R1),98 X0044 (RO),99-X0077 (RO) 3.8.7.2.5 Boundarv l
The AQ boundary of the high pressure fire water system is defined in Table 3.8-2.
3.8.7.3 UF, Release Detection System 3.8.7.3.1 Safety Function The UF release detection system shall detect and annunciate in the ACR, Uf releases in any enrichment cascade operating equipment that is operated above atmospheric pressure. The compressor UF.
outleakage detection system and the UF. release detection systems in the feed vaporization facilities and the toll enrichment services facility shall detect UF. releases to the atmosphere and provide an alarm to l alert personnel to take appropriate action (i.e., investigate to verify a release occurred and, if necessary, l evacuate the area affected by the release). Other systems that perform alarm and mitigation functions are discussed in Sections 3.8.2.2, 3.8.4.1 and 3.8.5.2.
3.8.7.3.2 Functional Requirements Each of the UF detection systems in the areas of the enrichment cascade that are intended to be operated above atmospheric pressure, the withdrawal, the feed vaporization, and the toll enrichment services facilities shall be designed in accordance with the following functional requirements to ensure the capability to accomplish the required safety function:
The system shall monitor the designated areas of the facility for UF. releases outside of the UF.
primary system.
The system shall provide, in the ACR, an alarm indication of a UF release from the UF primary system.
3.8.7.3.3 System Evaluation Enrichment cascade. The safety function of the system is to detect a UF release from the UF.
primary system and provide an alarm to alert on-site personnel in the ACR. This facilitates early detection by_ the operators allowing them to initiate required actions to minimize the release. The system is designed to detect releases in those areas that have the potential for a UF, release and provide an alarm in the ACR.
The detector heads are located in areas that are intended to be operr.ted above atmospheric pressure in the "00", "000", and interbuilding booster stations. Operation of these detector heads is required duri.ng a UF release. The detectors heads would be subjected to an environment associated with the release of UF 6 and its reaction products. However, the response time is relatively quick once the smoke is detected b:. sed on operational history. Once a detection signal is generated, the alarm circuit will be sealed in and operator action will be required to clear the alarm. Therefore, the environmental conditions during an event should not cause failure of the detection system. Additionally, there are multiple detector heads in each area to provide detection capability. Normal operation environments can also. cause some spurious operations due to various causes and result in detector failures. These are typically detected during the testing process and the detector head will not reset. However, these are typically limited to one detector at a time. With 3.8-60
SAR-PORTS PROPOSED August 31,1999 RAC 97-X0506,(RI),99-X0079 (RO) which do not affect the analysis include openings in the cell floor access hatches or access doors, or openings below the cell floor.
The process buildings are also required to prevent a large release of UF. resulting from evaluation basis natural phenomena events. This safety function is accomplished by requiring the building support structures to maintain structural integrity during evaluation basis natural phenomena events to the degree l needed to prevent failure of the UF primary system. Table 3.8-5 summarizes the results of the natural l phenomena evaluations for the process buildings. As indicated in the table, the process buildings and tie line structures will not experience any structural damage that would damage the UF primary system during l an evaluation basis natural phenomena event. As indicated in Table 3.8-5, Buildings X-326, X-330, X-333 and X-705 could experience some inleakage of water that may develop because of local ponding on the roof from heavy rainfall events (10,000-yr event). The analysis assumed all of the normal drainage paths were clogged, allowing no draining except over the parapets. The event would occur over a period of time, which allows for additional operator intervention prior to any inleakage occurring. The impact of water inleakage in a typical process building was reviewed, and no adverse impacts (i.e., loss of primary system integrity) were identified. Any inleakage to the cell floor would typically run to the open stairwells and floor drains, which would drain to the lower elevations. Some electrical equipment may be affected because of the water flow path. However, the likelihood of the combination of a heavy rainfall, all drainage paths being clogged, no operator intervention, and multiple electrical failures was not considered credible for these large buildings. It should be noted that compressor trip capability could also be accomplished from the switchyard if necessary to mitigate the effects of this event if it were to occur. The process buildings help minimize the consequences to on-site and off-site personnel and thus help ensure that the EGs for the facility are not exceeded to the extent possible as a result of any of the events that are postulated for the facility. Based on these requirements and supporting evaluations, the X-326, X-330 and X-333 process buildings can accomplish the required safety function.
Buildings X-342-A, X-343, and X-344-A are required to maintain structural integrity during evaluation basis natural phenomena events to the degree needed to prevent failure of the UF primary system. This safety function is accomplished by requiring the structures to withstand evaluation basis natural phenomena events that could result in failures of the UF primary system should the structure fail to maintain structural integrity These buildings will not see any struuural damage due to evaluation basis earthquake or flood events. However, some of the siding at X-342-A may be pulled off due to winds l greater than 50 mph, and X-343 could experience structural damage at winds greater than 70 mph. The siding is not a threat to the UF primary system integrity since it is pulled away from the interior of the l building and would not impact any equipment. The structural damage for X-343 will not result in failure l of the structure nor impact any UF systems. Based on these requirements and evaluations, these buildings l can accomplish the required safety functions.
3.8-67 I
1 l
l SAR-PORTS PROPOSED August 31,1999 RAC 9l-X0506,(R1), 97-X0440 (RO), 99-X0079 (RO)
Table 3.8-5. Natural Phenomena Capacities of Buildings.
l Wind OK > 75 mph (121 km/h)
Seismic Structure (OK > 0.05 g)
Structure Components Flood X-300 OK OK OK OK l
X-326 OK K
60'(97)
Inleakage' X-330 OK OK 65'(105)
Inleakage" l
X-333 OK OK 60'(97)
Inieakage'
-l X-343 OK OK 70'(113)
OK l
X-344-A/342 OK OK 70'(113)
OK l
X-345 OK OK OK OK l
X-705 OK OK 70*(113)
Inleakage 6
l X-710 OK OK OK OK l
X-720 OK OK OK OK l
XT-847 OK OK OK OK l
Tie lines OK OK OK OK a.
Siding pulls off,
- b. Roof ponding.
l c.
Limited roof displacement.
m SAR-PORTS PROPOSED August 31,1999 RAC 97X0315 (R1), 99X0041 (RO), 99X0077 (RO) systematic method of identifying the complete spectrum of hazards, hazard states, operating modes, and initiating events that may result in accidents with consequences of interest. The hazard matrix provides the foundation for detailed analysis to determine the TSRs, system classifications, and accident analyses. Each specific combination required analysis to determine whether protective action is required to prevent exceeding an Evaluation Guideline. An example of one combination taken from Table 4.2-5 follows:
Process parameter of interest-pressure increase.
Initiating event-steam control valve fails to open.
l*
Operating modes-heating and feeding, transfer or sampling.
l Hazard state for each mode-all states (heating); liquid / gas (feeding, transfer, or sampling).
The matrix combinations were evaluated to determine the minimum set of controls that could prevent exceeding the Evaluation Guidelines should the event occur. The combination that results in the most severe consequences (i.e., bounds all other initiating events by consequence) is identified as the limiting initiating event for that combination.
Each combination of initiating event, hazard state, and operating mode was reviewed as described for the operationa'l analysis task (Section 4.3.1.1). Defining the limiting initiating events for a facility required consideration of the integrated response of the facility to each initiating event without any i
mitigative action. If the initiating event could result in a parameter change with unmitigated consequences that exceed any Evaluation Guideline for the applicable frequency category, the initiating event was a
)
candidate for a limiting initiating event. The definition of a limiting initiating event described above was used to finalize the set of limiting initiating events. The resulting set of limiting initiating events was subjected to the detailed accident analysis described in Section 4.3.2. Initiating events that exceed the PSOA thresholds and are not limiting initiating events are documented in the analysis for the facility with justification provided for them-being bounded by the limiting events along with the controls necessary to support meeting the Evaluation Guidelines.
-4.2.6 Hazard Analysis Results This section presents the results of the hazard analysis. Most of the results are given in summary form. Details of the analyses are provided in the PrHA reports for the respective facilities.
One of the key elements of the hazard analysis was to identify historical events that resulted in accidents ofinterest. Table 4.2-6 identifies historical release events from 1961 through 1993 at all three GDP sites. Historical release events as well as discussions with operational personnel were a key input to all of the hazard analyses.
A comprehensive listing of all facilities that were included in the hazard analysis is presented in Table 4.2-7. The table shows the level of analysis that was applied to each facility based on the graded approach. It indicates whether a facility required a PHS, PrHA, and/or PSOA review. The hazard cate-i gorization based on DOE-STD-1027-92 is also presented. Facilities that were screened out by the PHS did not receive any additional review and are not addressed in the following sections. Facilities that
{
exceeded PHS thresholds but did not exceed PrHA thresholds are documented by an analysis statement (Table 4.2-8). The analysis statement serves as the safety analysis for these facilities, and they are not 4.2-13
f SAR-PORTS PROPOSED August 31,1999 RAC 97X0312 (R1), 99X0041 (RO), 99X0077 (RO) l d.
Comparison With Guidelines Further analysis is not required because this event is bounded by the cylinder failure inside autoclave event, Section 4.3.2.2.14.
e.
Summuy of SSCs and TSR Controls
. Based on the results of this analysis, the essential controls for this event are summarized as follows:
Administrative control to verify cylinder pressure less than or equal to 10 psia (69 kPa) prior to heating-maintain initial condition (normal operation, EG 5 only);
UF. cylinders, pigtails, and primary system piping outside the autoclave-maintain primary system
=
integrity (EGs 1,2, and 6 only);
Autoclave shell and associated isolation vah es-maintain primary system integrity (EG 3 only); and j
UF. cylinder high pressure autoclave steam shutoff system-minimize potential for primary system integrity failure (EGs 1,2, and 6-on-site only).
Based on the above essential controls, the resulting important to safety SSCs and TSRs are as follows:
l The UF. cylinder high pressure autoclave steam shutoff system, UF cylinders, pigtails, primary a
system piping outside the autoclave, and the autoclave shell and associated isolation valves are identified as important to safety SSCs. See Section 3.8 for details including safety classification.
l TSRs are provided for the UF cylinder high pressure autoclave steam shutoff system, pF cylinders, and administrative requirements for procedures and training of workers for evacuation actions.
4.3.2.2.8 Heavy Equipment Drop (Primary System Integrity)
The heavy equipment drop event is limited to the cell floor of the process building in the area of the withdrawal system equipment. Drops of heavy equipment in the cylinder handling and autoclave areas are discwed in other accident analysis scenarios. These scenarios include: (1) pigtail /line failure outside 1 autoclave (see Section 4.3.2.2.10), (2) the pigtail /line failure at the withdrawal station (see Section 4.3.2.2.11), and (3) the cylinder failure outside the autoclave (see Section 4.3.2.2.15). The administrative controls, and frequencies of occurrence for the heavy equipment drop accident scenario in the withdrawal facilities are bounded by the discussion of the heavy equipment drop accident scenario for the cascade facilities (see Section 4.3.2.1.8). Any required engineering controls for mitigation are the same as those described for the process line failure at compressor discharge in the withdrawal facilities (see Section l
4.3.2.2.12). Therefore, no additional analysis of this event is provided.
l t
4.3-92
SAR-PORTS PROPOSED August 31.1999 RAC 97X0313 (RO), 97X0440 (RO), 99-X0079 (RO) concern is not addressed further in this analysis.) Based on this analysis, this event was not identified as causing any impact on the leased facilities.
4.3.2.5.2 IIigh Wind (External Event)
All of the facilities' Main Wind Force Resisting Systems (MWFRS) meet the requirements for high l winds (see Table 3.8-5). Other types of structural damage associated with X-330, X-333, X-344-A, l X-342-A, X-343, and X-705 due to high wind were identified, but were associated with loss of siding only l and limited roof displacement in X-343, and not the buildings' MWFRS. The evaluation indicated that some of the siding for these buildings could be pulled from the building, which would have no effect on the l equipment inside the buildings, except for wind impact. The limited roof displacement in X-343 would have l no effect on the equipment inside the building. Piping and equipment were reviewed in several facilities at the plant site as indicated in Section 3.8, and no failures were identified because of wind loading. Therefore, l these facilities were determined to be capable of withstanding the EBE high wind.
I a.
Scenario Descriotion Natural phenomena evaluations included consideration of evaluation basis wind loads on the major UF facilities. As indicated previously, the facilities that were analyzed satisfy the 250-year return interval l EBE of 75 mph (121 km/hr). Based on this analysis, this event was not identified as causing a release of l radioactive material (primarily UF ) in the leased facilities.
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4.3-135
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SAR-PORTS PROPOSED August 31,1999 RAC 97X0313 (RO), 97X0440 (RO), 99-X0079 (RO) l l
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i SAR-PORTS PROPOSED August 31,1999 RAC 97X0313.(RO), 99-X0079 (RO) _
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Source-Term An alvsis l
Since no rel' ease of radioactive material is postulated, no source term analysis is required.
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-Con =aanence Analysis l.
Since no release of radioactive material is postulated, no off-site consequence analysis is required.
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1 4.3-137
' SAR-PORTS PROPOSED August 31,1999 RAC 97X0315 (RO), 98X0044 (RO), 99X0041 (RO), 99X0077 (RO) l Table 4.2-i Example Initiating Event-Operating blode-Ilazard State Alatrix' l
Cylinder /
l Pigtail
- Feeding, l
Ope. tions Shutdown /
Transfer or l
Anticipated operating Containment IIeating Sampling initiating event mode operating mode operating mode operating mode llazard state Pressure increase l
Autoclave steam All states Liquid / Gas control valve fails open Temperature increase l
Fires (X-343 only)
All states All states All states Liquid / Gas Primary system integrity l
Minor leaks of UF, All states All states Liquid / Gas inside autoclave l
Minor leaks of UF.
All states Liquid / Gas to atmosphere l~
1.
Table represents X-342A and X 343
s 9
f SAR-PORTS PROPOSED August 31,1999 RAC 97-X0314. (R1), 99-X0016 (RO) instruments to control a constant air pressure. The extremely accurate pressure provided by a datum system, referred to as the datum pressure or datum, is the reference used to maintain desirable cascade operating pressures in the cascade buildings.
1 1
1 Nuclear criticality safety of the EBSs depends primarily on cascade parameters and moderation control. The cascade datum systems utilize a combination of administrative controls (primarily during pump maintenance) and a passive barrier consisting of the datum exhaust pump volume.
In order for a criticality to be po'ssible, multiple contingency events would have to occur l
simultaneously. For instance, loss of EBS system integrity allowing inleakage of wet air would need to j
occur (passive barrier failure), and an operator required to be in the control room would need to fail to respond to the resulting compressor motor load alarm (administrative control failure). For the datum systems, the volume (passive barrier) of the exhaust pumps is low enough that the pumps remain subcritical for maximum physically credible amounts of uranium in the oil for up to 10% enriched uranium. Multiple f
administrative controls would have to fail for a criticality to occur during datum maintenance activities, j
For example, the operators would have to bring a container of uranium bearing material into the room near the pump and the pump would have to contain uranium bearing material (unlikely event) before criticality would be possible. Administrative controls in the NCSAs for containers of uranium bearing material to remain spaced from other uranium bearing material would also have to fail.
The double contingency principle is met for most situations. For certain cases of moderator inleakage to the cascade, the double contingency principle is not strictly met for the EBS system (See Section 1.2 of this appendix).
l There are no AEFs identified as part of the cascade support systems.
l l
1.7 Surge Drums l
l UF can be stored in the gaseous phase in the cascade surge drums located in the cascade cell 6
servicing areas. The surge drums serve several functions:
(1)
To ensure stable operation of the cascade, large volumes of " lights" (N, 0, and R-114 2
2 mixtures) and UF may be removed from the cascade. The IJF is stored, and later 6
returned to the assay match point at a controlled rate.
(2)
To store large air and/or R-ll4 inleakages which have been removed from the cascade.
(3)
To store reaction products resulting from cell drying and unplugging treatment until they can be cold trapped.
(4)
For inventory control in a major disturbance.
For cell maintenance activities, the surge drums are used in obtaining a cell negative. Various banks of the surge drums are used in reducing pressure in the cells as much as possible.
5.2A-5 l
l
s SAR-PORTS PROPOSED August 31,1999 RAC 97-X0314 (R1), 99-X0023 (RO)
Moderation, concentration / density, and interaction are the controls used to ensure criticality safety for the surge drums. Under normal conditions, the UF is maintained in a vapor state by controlling the 6
pressure and temperature in the surge drums. UF,in a gaseous phase cannot achieve criticality at PORTS.
A number of administrative controls, unlikely events and passive barriers exist in the surge drum operation to prevent a criticality from occurring. The administrative controls include maintaining the j temperature of the surge drums to at least 90*F, limits on the enrichment of uranium at a given location (i.e. building X-330, X-333 and X-326) and ensuring that the maximum surge drum pressure specified in the NCSA is not exceeded. The passive barrier consisting of surge drum physical integrity prevents wet air inleakage (moderation con;rol). Administrative controls such as periodic monitoring of surge drum temperatures and pressures provide control over concentration / density.
Should wet air (H O) leak into a drum, a reaction with the UF.will form UO F pnd HF, producing 2
a higher pressure in the surge drum. The pressure rise in the surge drum would have to go unnoticed in both the ACR and X-300 for the abnormal condition to continue. However, if left unattended, the UF.
would continue to react with the wet air and form UO F and HF. The HF will not condense at the 2 2 elevated surge drum temperatures, thus producing no moderation problem. If wet air enters the surge drum while it is down for maintenance, the pressure is reduced to a level below that at which liquid water can exist before placing the surge drum back in service.
The double contingency principle is met and there are no AEFs identified for surge drums usage.
1.8 Cold Recovery Operations The X-330 and X-333 process buildings house cold recovery / trapping systems. Cold recovery systems provide a means of recovering UF from large amounts of lighter gases which have entered the 6
cascade through preparing equipment for maintenance, servicing equipment for return to the cascade, or through other abnormal circumstances. Refer to SAR Section 3.1.4 for further details of the operation.
The nuclear criticality safety of the cold recovery chemical traps is based primarily on geometry control of the traps. The criticality safety of the cold traps in the X-330 and X-333 cold recovery areas is based primarily on controlling the enrichment of the material entering the cold traps, and on the dimensions of the cold traps. The cold trap dimensions are designed to be safe for the expected enrichments, including sufficient margin to allow for contingency events. Criticality safety is also based on interaction control, since the cold traps are fixed in place. The nuclear criticality safety of the Cold Recovery Holding Drums is based primarily on enrichment and moderation control. Moderation control is maintained through controls on temperature and pressure.
A number of administrative controls and passive barriers exist for the cold recovery holding drums, chemical traps and cold traps to prevent a criticality from occurring. In order for a criticality to be possible, multiple contingency events would need to occur simultaneously. For instance, controls on the operating temperature and pressure of the holding drums ensure that the UF material present in the holding 6
drums remains in the gaseous phase. UF in the gaseous phase cannot be made critical. Therefore, another 6
loss of control coupled with large quantities of enriched material being placed in the 5.2 A-6 j
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o SAR-PORTS PROPOSED August 31,1999 RAC 97 X0314.(RI),98 X0177 (RO) l 3.13 General Material Handling and Storage in X-705 l
Material handling and storage in X-705 involves the general storage and handling of UH equipment in I
l three main storage locations. The three main storage locations are designated as the Large Parts Storage, East Tunnel Storage, and the Large Parts Cage.
The Tunnel Basement contains solution storages, associated piping and pumps used to recirculate and transfer solutions to and from various parts of the X-705 Building, and systems used to collect solution leaking from pumps, piping, and storages. The large parts cage is an enclosed area for storage of equipment.
l Criticality safety is maintained through spacing and mass controls for material handling and storage in 4
l X-705. Criticality safety in the Tunnel Basement is provided by 1) retaining uranium-bearing solutions inside equipment that is geometrically favorable and 2) secondary geometrically favorable containment.
4 Administrative controls and passive barriers have been incorporated to prevent a criticality from occurring. Administrative controls include inspecting the status of the Tunnel Basement floor and equipment, along with checking storage prior to each solution transfer to ensure that sufficient volume is available to avoid l overCow. Spacing controls are used throughout X-705 during movement of uranium-bearing equipment. Passive engineered controls are found in the piping systems including line elevations to prevent backflow or misdirected flows; for example, the transfer line leading from the Rinse Booth to the Overhead Storage is higher than the overflow columns and their vents. The favorable geometry and primary system integrity of the overhead storage and all installed equipment in the tunnel that contains uranium-bearing materials are passive engineered features.
Drains are provided in all electrical connection boxes to avoid an unfavorable geometry should a leak occur.
In order for a criticality to be possible, multiple contingency : vents would need to occur simultaneously.
For instance,if uranium-bearing solution leaks to the tunnel basemen floor from other locations in the building (failure ofpassive barrier) it is unlikely that the solution would collect t) an " unsafe" depth because the floor is checked every four hours for collection of solution. The double contingercy principle is met for these operations.
l There are no AEFs identified for General Material Handling and Storage in X-705.
3.14 Truck Alley Cleaniag / Oil and Grease Removal The Truck Alley, located along the northem wall of the X-705 building, is an area used for cleaning oily or greasy equipment. Equipment is dismantled into its component parts and any accumulated oil, grease, or dirt is scraped off. Equipment may be cleaned with a steam gun that normally uses a mixture of steam and a detergent solution. Solutions generated by the cleaning operations are rinsed into any of three geometrically favorable floor drains, then pumped to the Intermediate Storage 5.2A-21
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l SAR-PORTS PROPOSED August 31,1999 RAC 97-X0314 (RI), 98-X0093 (RO) 4 5.1 Fissile Material Operations in X-710 Laboratories Several different individual laboratory facilities are located in the X-710 building. These laboratories handle solutions and/or solids that contain or potentially contain uranium in varying enrichments, concentrations, and quantities.
A number of administrative controls and passive engineered features exist in the labs to prevent a criticality from occurring. The administrative controls include limits on the quantity and/or geometry of uranium at a given location in the lab, and on the minimum spacing between a container or array of containers and other uranium-bearing materials in the lab. When the quantity of uranium in a container l of solution or a solid is unknown, conservative quantities are analyzed in establishing NCS controls. The passive engineered features include the design and physical integrity of containers to exclude moderating
)
materials, and the design and physical integrity of storage racks or receptacles to control spacing of multiple containers.
1 In order for a criticalir i be possible, multiple contingency events would need to occur. For
{
instance, multiple containers of:uterial bearing a total quantity of uranium equal to or exceeding a critical i
mass would need to be brought into a lab and placed in proximity to one another (failure of administrative controls over spacing and/or geometry), and water or some other moderating material would need to have leaked into one or more of the containers (failure of passive barrier to achieve moderation control).
Therefore, the double contingency principle is met and there are no AEFs identified for fissile material operations in the X-710 laboratory facilities.
6.0 Waste Storage and Support Facilities Nuclear criticality safety associated with fissile material operations conducted in the waste storage and support facilities and operations are summarized in this section.
6.1 Waste Storage and IIandling Low level radioactive waste (LLRW) and mixed waste handling and storage operations are located throughout plantsite. RCRA 90-day storage areas and satellite accumulation areas are located near the origin of the waste. Many different types of containers are used for the wastes - ranging from small diameter containers to large waste boxes. Waste is transported to the XT-847 facility to be temporarily l
stored until it is either shipped off-site for treatment and/or disposal, or staged prior to treatment and/or storage in on-site facilities.
For containers that are not geometrically favorable, the mass of U2" in containers or groups of containers in these areas is administratively controlled to a safe limit. Where required, spacing and stacking limits are imposed as additional controls.
In order for a criticality to be possible, multiple contingency events would need to occur. For instance, a drum in the heavy metals sludge storage in X-705 would need to be double-batched, that is, it would need to contain twice the container mass limit (first administrative control failure), and one or 5.2A-27
I SAR-PORTS PROPOSED August 31,1999 RAC 97-X0314 (R1),99-X0013 (RO) geometrically favorable small containers that exclude moderators, that contain only limited amounts of UF.,
and that are kept in safely spaced and geometrically controlled configurations. Specific enrichment limits are also applicable for certain storage areas. The administrative controls include limits on the number of l each type of container that may be grouped together and spacing between groups of containers. The design and integrity of sample cylinders and valves and the separation provided by shelves in storage cabinets constitute passive barriers.
In order for a criticality to be possible, multiple contingency events would need to occur simultaneously. For instance, containers would need to be batched beyond allowed limits (first administrative control failure), and moderation control would need to be lost through container failure or placement (passive barrier failure). Therefore, the double contingency principle is met.
There are no AEFs identified for sample cylinder handling and storage.
7.7 Portable Cart-Mounted Sampling, Test, and Analysis Equipment Portable cart-mounted sampling, test and analysis equipment generally consist of a small non UF.
gas cylinder, vacuum pump, chemical traps, pressure transducer, analytical apparatus, and the interconnecting piping and valves. The analytical apparatus may be any of several types of devices including sample collection tubes, infrared, and ultraviolet gas analyzers.
The only components of concern to criticality safety are the small chemical traps and vacuum pumps. These components, while installed on the carts, do not present a criticality hazard due to the small size, spacing between the components, and the configuration of the equipment. Limits are in place to ensure that the configuration of the buggy, sample carts, and analyzers-specifically, size and spacing of components-are maintained during use and all maintenance activities. Components removed from the carts are handled and stored with spacing requirements to ensure criticality safety. Additionally, the carts are serviced such that the spent trapping media and waste oil are collected and treated as potentially fissile waste in accordance with the plant waste management program.
The nuclear criticality controls for the use of the sampling carts and buggies are based on volume, geometry, moderation, and interaction.
Several administrative controls over volume, geometry, moderation and interaction have been incorporated to prevent a criticality from occurring.
In order for a criticality to be possible, multiple contingency events would need to occur. For instance, a criticality would be nossible through altering of the volume by replacement with larger traps or a larger pump (failure of ahnistrative control) and excessive uranium and moderating material would need to be deposited in the traps before a criticality would be possible (failure of second control). These operations and equipment meet the double contingency principle and there are no AEFs identified for the portable cart-mounted sampling, test, and analysis equipment.
5.2A-31 1
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r SAR PORTS PROPOSED August 31,1999 RAC 97-X0314.(RI),98-X0108 (RO) 7.12 Contaminated Metal Uranium contaminated metal is generated throughout the site and includes metal parts, scrap, equipment and 55 gallon or smaller drums that are potentially contaminated extemally and internally with uranium.
Examples ofmetalinclude alloys / metals in waste streams, structural metals, process metals and fencing. Other metals, which cannot be released for unrestricted use, are included in the disposal profile of radiologically contaminated metal.
l This waste stream may consist of building debris, including glass, concrete, masonry, brick, other man-made materials, and incidental dirt / oily residues. Scrap metals are placed in various sized containers. The most common containers are SS-gallon drums, B-25 boxes, and hoppers.
Hoppers are large dumpsters used in the X 705 facility and occasionally in the X-720 building. Hoppers have slots drilled in all four sides near the base. Hoppers are sorted for proper disposal of all contents. Posting on the drums prohibits disposal of visible uranium contamination into the drums.
l NCS of contaminated metal is based on mass control and moderation control for most cases. Volume l
l and interaction control can also be used on parts known or suspected of contamination. These controls are accomplished through a ccmbination of admmistrative controls and passive barriers. The primary administrative l controls are a visual inspection for uranium by the operator and assurance that the material has no free liquids.
There are other administrative controls such as limits on the number of full or partially full containers that can be in motion at any one time, limits on the minimum spacing of equipment or other large pieces with hidden l
volumes, limits on the minimum spacing of uranium-bearing equipment from uncharacterized containers, limits i
on uncharacterized B-25 and 6x6x8 boxes minimum spacing, and limits on non-empty damaged metal drums l
separation spacing. Administrative controls are in place for hoppers, which consist of not plugging the drain holes, spacing ofhoppers from other uranium-bearing materials and keeping the lids closed on hoppers when left unattended. A passive barrier is provided by the structural integrity and design of 55 gallon drums, B-25 and 6x6x8 boxes and hoppers, including: drains holes in the 55 gallon drums, holes in the hoppers, lids on some hoppers, and lids on allemate smaller contaminated metal receptacles.
In order for a criticality to be possible, multiple contingency events would need to occur simultaneously.
Therefore, double contingency is met and there are no AEFs identified for contaminated metal.
7.13 Always-Safe Portable Small UF, Release Gulpers The small UF gulpers are utilized for contamination control of certain operational and maintenance activities.
The controls for these units are primarily passive engineered controls on geometry and interaction (spacing considerations). In addition, there are administrative controls on spacing from other containers of l uranium materials, enrichment, HEPA filter installation, and on transportation considerations.
5.2A-34 l
N SAR PORTS PROPOSED August 31,1999 RAC 97-X0314. (R1), 98-X0093 (RO), 98-X0108 (RO) l J
l The operations utilizing this equipment meets the double contingency principle using primarily passive engineered controls and some administrative controls. There are no AEFs associated with this equipment.
7,14 Uranium Analysis and Sampling In order to demonstrate that process conditions are within NCS limits, process material or waste is analyzed for uranium and/or U2" content. Frequently, a procedure or specific NCS controls require two sample analyses to satisfy the double contingency principle. When lab analysis is used, only a sample of the uranium-bearing material is analyzed.
When non-destructive assay (NDA) measurements of characteristic uranium emissions are used, withdrawing samples is usually not necessary because U " is 2
analyzed directly in the container or equipment. The lab analysis process involves withdrawing a small j sample, transporting the sample to a laboratory, chemically preparing the sample, analyzing the sample, and logging and reporting the results. NDA measurements can involve qualitative measurements, such as l
the gamma scan, to determine if the amount of uranium present has increased significantly, and/or quantitative measurements, which can estimate U " mass.
2 The nuclear criticality safety of sampling operations is maintained through vohime limits and geometry control. Sample containers are limited in volume and may be handled in groups whose total volume is a factor of two below the subcritical volume limit. In some cases, sample containers are permitted to be stored on approved storage racks where the fixed geometry constitutes a passive design feature. Interaction among sample storage racks, groups of sample containers, and other uranium-bearing material is administratively controlled through spacing requirements.
In order for a criticality to be possible, multiple contingency events would need to occur simultaneously. For instance, a group of sample containers would need to exceed the combined volume limit (first administrative control failure), and the minimum spacing limit from the group would need to be violated (second administrative control failure). Therefore, the double contingency principle is met.
There are no AEFs identified for uranium analysis and sampling.
7.15 Laundry The X-705 laundry facility is divided into potentially contaminated and uncontaminated areas and laundry destined for each is usually kept separate within the laundry areas. The uncontaminated laundry is accumulated in hampers located in various locker rooms around the plant. These items are not expected to be contaminated since either they were not inside a controlled contamination area when worn or were surveyed by the user upon exit from a contaminated area.
There are nc NCS controls applied to potentially contaminated laundry because of the insignificant levels of contamination associated with laundry operations. For areas or activities which are expected to result in contamination, PPE are generally required. Areas to be serviced for maintenance and/or repair l
are decontaminated of visual surface contamination prior to any maintenance and/or service activities and decontamination activities are performed in a manner such that minimal uranium-bearing material will be on the PPE worn by the chemical operations personnel. Outer protective clothing worn in a contaminated l
area is removed at the contamination area exit point. These PPE are removed by the individual and placed into plastic and/or water soluble bags.
5.2A-35 1
o SAR-PORTS PROPOSED August 31,1999 RAC 97-X0314 (R1), RAC 98-X0037 (RO),98-X0108 (RO)
The accumulation of uranium in the washer over a number of wash cycles is non-credible since there are no cavities in the washer and the number of water reloads per wash cycle tends to flush away possible pockets of accumulation. Accumulation of uranium in the dryer is also non-credible since any significant uranium contamination on the laundry items would be washed out by the washer.
Except for instances of visual contamination, the initiating event of exceeding the safe mass is non-credible and no NCS controls are required. In cases where the PPE have visible (beyond fixed stains / films) amounts of uranium on or inside them or are contaminated with uranium bearing solution, they will be disposed of in containers as described in section 6.2 of this appendix.
There are no AEFs identified for the laundry facility.
7.16 Cylinder Valve Replacement UF, cylinder valves occasionally need to be replaced due to some form of malfunction (plugged, seized, etc.). The cylinder may be partially or entirely filled with solid UF. at the time of the replacement.
A number of administrative controls are in place to prevent criticality, such as maintaining plugs readily available during the operation. The plugs are used in the event that neither of the two replacement valves can be installed after removal of the old valve. Administrative controls require the replacement valves be prepared and ready for installation. Administrative restrictions limit the number of used cylinder valves that can be grouped together for storage or transport. The primary criticality controls for UF.
cylinder replacement are mass, enrichment, moderation and interaction.
In order for a criticality to be possible, multiple contingency events would need to occur simultaneously. For instance, if more than the permitted number of used cylinder valves are gathered together in a group (violation of an administrative control), a spacing requirement between groups is maintained. Additionally, a passive barrier limits the internal free volume of the cylinder valve. All of these requirements prevent a criticality from occurring. Therefore, the double contingency principle is met and there are no AEFs identified for the UF cylinder valve replacement operation.
6 7.17 Building Decontamination Activities A number of decontamination activities are routinely conducted in support of building operations.
Field decontamination is conducted in order to safely and expediently remove potential health hazards, to recover uranium material so that it may be recycled back into the enrichment process, and to prepare equipment for repair or discard. Because uranium-bearing materials enriched in U may be present on the surfaces to be decontaminated, decontamination activities have the potential to generate unsafe gi'antities of U*. Decontamination activities generally fall into two categories: dry and wet.
Dry decontamination comprises the removal of contamination through mechanical action on contaminated compounds with brushes, rags, shovels, dry vacuum cleaners, strip coat, dry ice and other blasting, etc. Dry decontamination using brushes, shovels, and other non-vacuum equipment is generally used outside of containment enclosures only when necessary due to the potential for generating airborne particulates. Any of the dry decontamination methods can be performed in negative pressure containment enclosures erected to minimize the potential for generating airborne particulates.
5.2 A-36
o SAR-PORTS PROPOSED August 31,1999 RAC 97.X0506,(RO),98X0141 (RO) 5.6 CHEMICAL SAFETY United States Enrichment Corporation (USEC) operations at PORTS require radioactive, hazardous, and toxic chemicals to support the basic process of uranium enrichment. The enrichment process uses, consumes, combines, and manufactures various hazardous, flammable, reactive, and toxic chemicals. Pursuant to 10 CFR Part 76.87(c)(ll), the TSRs include appropriate references to address chemical safety. The Safety Analysis Report (SAR) describes the technical basis program requirements for chemical safety, the integration of chemical safety with uranium enrichment operations, and describes the management systems used by PORTS for chemical safety. The TSRs identify those requirements for control of chemicals and commits to the chemical safety controls described in this section.
5.6.1 Introduction Chemical safety at PORTS consists of the integration of environmental, safety and health management systems to address chemical safety. Chemical safety controls are designed to mitigate the adverse effects of the toxic materials used in the enrichment process to workers, the public and the environment. To achieve this objective, sal'ety analyses, process hazard analyses, and industrial hygiene and safety programs are utilized as described below.
5.6.2 Organization and Administration Section 6.1 identifies the roles and responsibilities for the Safety and Health program.
I The Safety and Health Assurance and Policy Manager has responsibility and authority for the 1
Safety and Health program. The chemical safety program is functionally coordinated by a program manager appointed by the Environmental, Safety and Health Manager. The Environmental, Safety and Health Manager is designated as the senior site manager having responsibility for safety and health matters.
l The Work Control Manager is designated senior site manager responsible for the receipt and shipment of hazardous materials. The Production Support Manager is the senior site manager having responsibilities for environmental matters. The Fire Services Manager is the senior fire protection officer. This group of PORTS personnel provides the management framework for chemical safety.
Chemical safety utilizes existing plant programs rather than developing new or specifically tailored programs. These referenced programs may or may not contain direct chemical safety references.
Chemical safety is described _ in the SAR and incorporates technical and administrative control to l manage risk. In Section 5.6.13, the chemical safety control strategy is discussed further and the additional controls and requirements utilized to protect workers and the general public from chemical hazards are identified.
The PORTS chemical safety controls are limited to non-radiological materials. Radiological materials, safety analyses, and the toxicity of uranium are addressed in Chapter 4 and Section 5.3.
5.6-1
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o TSR-PORTS PROPOSED August 31,1999 RAC 99-X0077 (RO)
SECTION 3.0 ADMINISTRATIVE CONTROLS
' 3.6 TECHNICAL SAFETY REQUIREMENTS (TSR) BASES CONTROL Changes to the TSR Basis statements shall be reviewed and approved in accordance with the plant change control process as described in Section 6.3 of the SAR.
3.7 EFFECTS OF NATURAL PHENOMENA Emergency response procedures shall be established, implemented, and maintained to prescribe plant response to the following natural phenomena events:
' Earthquake Tornado /High Winds Flooding / Intense Precipitation 3.8 PROCESS VENTILATION AND OFF-GAS Control of radioactive emissions shall be established implemented and maintained as described in Section 5.1 of the SAR.
3.9 PROCEDURES 6
3.9.1 SCOPE Written procedures shall be prepared, reviewed, approved, implemented, and maintained (except for a limited time as specified in the Compliance Plan) to cover the following:
a.
Activities described in SAR Section 6.11.4.1 and listed in Appendix A to SAR Section 6.11; b.
Operator actions and administrative controls described in SAR Chapter 4 to prevent i or mitigate the consequences of accidents; and I
I c.
Programs specified and described in TSRs 3.11 through 3.19, and 3.23.
3.9.2 REVIEW AND APPROVAL a.
Each new procedure required by TSR 3.9.1 shall be reviewed by the PORC in accordance with TSR 3.10.
f 3.0-6 L