ML20217M311
| ML20217M311 | |
| Person / Time | |
|---|---|
| Site: | Portsmouth Gaseous Diffusion Plant |
| Issue date: | 10/15/1999 |
| From: | UNITED STATES ENRICHMENT CORP. (USEC) |
| To: | |
| Shared Package | |
| ML20217M301 | List: |
| References | |
| NUDOCS 9910270192 | |
| Download: ML20217M311 (15) | |
Text
L GDP 99-0171 Page1of3 l
United States Enrichment Corporation (USEC) l Proposed Changes Certificate Amendment Request Update the Application Safety Analysis Report 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 l
(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), GDP 99-0120, dated July 22,1999 (Reference 7), and GDP 99-0153, dated August 31,1999 (Reference 8) for NRC review
{
uld approval.
l 2.0 Description of Submittal The following changes are included in this submittal which modify the latest version of the SAR l
Update certificate amendment request. The revised pages are included in Enclosure 3.
i A.
The Revision Log has been updated to reflect the changes included in this revision.
B.
The List of Effective Pages has been updated to reflect the changes included in this revision.
C.
SARUP Section 5.2, Appendix A, paragraph 4.1, X-700 Converter Disassembly and Repair Area, has been revised to allow for the repair and storage of Uncomplicated Handling (i.e.,
those with safe mass) converters and coolers in the X-700 building. The previous SARUP text limited the discussion to converters only and also required the removal of all uranium deposits from the equipment before entering the area.
i
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l D.
SARUP Section 5.2, Appendix A, paragraph 3.4, Handtable Operations, was revised to clarify that only one of the installed handtable overflow drains is credited as a nuclear criticality safety control in the Nuclear Criticality Safety Approval (NCSA).
E.
SARUP Section 4.3.2.2.14, Cylinder Failure Inside Autoclave, was revised to reflect the results of new analysis of autoclave head-to-shell 0-ring leakages.
F.
SARUP TSR Section 2.2.3.3, Basis, was revised to reflect modifications that installed additional UF smoke detectors in building tie-lines from feed facilities to the process 6
buildmgs.
1 9910270192 991015 PDR ADOCK 07007002 C
1 GDP 99-0171 Page 2 of 3 G
SARUP Section 5.2, Appendix A, paragraph 1.4 was revised to change the references to passive and active engineered controls that are required for criticality safety, consistent with the newly revised NCS A. For example, gamma spectrometers and steel plate covers are no longer credited in the NCSA while raw circulating water controls valves are credited.
3.0 Hasis for the Revision
[ Item C] The change to add the cooler equipment to the discussion clarifies the description of the actual plant equipment undergoing repair. The change from " trace" quantities of uranium deposits to " Uncomplicated Handling" quantities is not significant. A "UH" quantity of uranium is a safe mass that cannot achieve criticality. The change is the result of a revision to the associated NCSA. There was no change made to the types of controls utilized by the NCSA to control nuclear criticality safety, namely mass control and moderation control.
[ Item D] The change to the description of the controls for handtable operations clarified that only one overflow drain, the one that drains to an overflow container, is credited as a nuclear criticality safety control in the respective NCSA. The change is the result of a revision to the associated NCSA. While a second overflow drain still exists, its design does not qualify it to serve as a nuclear criticality safety control and therefore it was removed from the discussion. The handtables continue to satisfy the double contingency principle.
[ Item E] This change incorporates the results of a new analysis regarding autoclave head-to-shell 0-ring leakage assumed in SARUP section 4.3.2.2.14, Cylinder Failure Inside Autoclave. As part of the initial certificate amendment request to update the SAR (SARUP). USEC made a commitment to re-evaluate the calculated leakage through the autoclave head-to-shell 0-ring by more accurately defining the reaction process. New laboratory tests were performed to estimate the reaction rates and resulting temperatures and pressures that would be experienced due to a UF6 release into a closed autoclave. Analysis of these test results found that the SARUP autoclave models used to assess autoclave conditions durir.g the proposed accident were overly conservative with respect to predicted temperatures and pressures. Based on these results, the UF release rate 6
was calculated from a closed autoclave following a 14-ton cylinder failure inside based on a hole in the autoclave seal of equivalent size to result in the TSR acceptable leakage rate. The release rates were determined for two different cases: (1) for a hole in the autoclave seal toward the top of the autoclave (resulting in a UF. vapor release); and (2) fbr a hole in the bottom of the autoclave seal (resulting in a liquid UF release). These results are different than that presented in SARUP section 4.3.2.2.14 source term discussions. Specifically, the previous single release rate of 3 lb/ min is being replaced by two release rate cases: 3.5 lb/ min (vapor release) and 8.3 lb/ min (liquid release). This change is revising the reported release rates discussed in section 4.3.2.2.14 of the SARUP. The basic conclusions from the scenario are not changed by this new analysis in that: (1) the release is still bounded; and (2) the release from the closed autoclave would be maintained below Evaluation Guidelines for this type of accident. The calculations used for this analysis are available onsite fbr review.
GDP 99-0171 Page 3 of 3
[ Item F] The addition of UF. smoke detectors to building tie-lines provides additional capability to detect system leaks.
[ Item G] The change in passive and active nuclear criticality safety controls for the withdrawal facilities does not reduce the level of safety. The double contingency principle continues to be met a sd multiple failures would need to occur in order for a filled cylinder to reach a critical configuration.
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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GDP 99-0171 Page1 of12 i
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SAFETY ANALYSIS REPORT UPDATE-CERTIFICATE AMENDMENT REQUEST-
' October 8,1999 REVISION Remove Pages Insert Pages SARUP Revision Log SARUP Revision Log None iv SARUP List of Effective Pages SARUP List of Effective Pages SARUP-1, -7, -8, -9 S ARUP-1, -7, -8, -9 SARUP Chapter 4 SARUP Chapter 4 4.3-114 4.3-114 SARUP Section 5.2A SARUP Section 5.2A 5.2A-3, -4, -14, -26 5.2A-3, -4, -14, -26 SARUP TSRs SARUP TSRs 2.2-8a '
2.2-8a
October 8,1999 United States Enrichment Corporation
- Portsmouth Gaseous Diffusion Plant Safety Analysis Report Update REVISION LOG (continued) 10/08/99' Submittal to revise section 5.2, App ndix A, paragraph 4.1, X-700 Converter Disassembly and Repair Arc, t:, allow for the repair and storage of Uncomplicated Handling (i.e., those with safe mass) converters and coolers in the X 700 building;
. revise Section 5.2, Appendix A, paragraph 3.4, Handtable Operations, to clarify that only one of the installed handtable overflow drains is credited as a nuclear l
criticality safety control in the Nuclear Criticality Safety Approval (NCSA); revise Section 4.3.2.2.14 to update release rate information for the cylinder rupture inside an autoclave accident analysis scenario based on new data and analysis; revise the Basis statement for TSR 2.2.3.3 to reflect modifications that installed smoke l
detectors in additional building tic-lines; revise Section 5.2, Appendix A, paragraph 1.4 to change the references to passive and active engineered controls in the ERP, I
LAW, and Tails facilities that are credited in the respective NCSAs.
l iv
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i S4RUP-PORTS October 8,1999 LIST OF EFF"CTIVE PAGES REVISION LOG -
CHAPTER 2 (Continued) hgg -
RAC/ Revision /Date hgg RAC/ Revision i
,i July 9,1999 2.3-7 RAC 97-X0248 (RO)
)=
ii July 9,1999 2.3-8 RAC 97-X0248 (RO) 1 iii August 31,1999 -
2.3-9 RAC 97-X0248 (RO) iv October 8,1999 2.3 10 RAC 97-X0248 (RO) i 2.3-11 RAC 97-X0248 (RO) i CHAPTER 1. APPENDIX A-2.3-12 R.AC 97-X0248 (RO) 2.3-13 RAC 97-X0248 (RO) hgg RAC/ Revision 2.3-14 RAC 97-X0248 (RO)
A-1 '
RAC 97-X0506 (RO) 2.3-15 RAC 97-X0248 (RO) -
RAC 98-X0050 (RO) 2.3-16 RAC 97-X0248 (RO)
A-1a RAC 98-X0050 (RO)
.2.3-17 RAC 97-X0248 (RO)
A-2 RAC 97-X0506 (RO).
2.3 18 RAC 97 X0248 (RO) l A-3 '
RAC 97-X0506 (RO) 2.3-19 RAC 97-X0248 (RO) j 1'
A-4 RAC 97-X0506 (RO) 2.3 20 RAC 97-X0248 (RO)
RAC 98-X0130 (RO) '
2.3-21 RAC 97-X0248 (RO)
)
- A-5 RAC 97-X0506 (RO) 2.4-2
- RAC 97-X0248 (RO)
RAC 98-X0130 (RO) 2.4-6 RAC 97-X0248 (RO)
A-6 RAC 97-X0506 (RO) 2.4-7
. RAC 97-X0248 (RO)
RAC 97-X0093 (RO) 2.4-8 '
RAC 97 X0248 (RO) i j
- A RAC 97-X0506 (RO) 2,4 9
- RAC 97-X0248 (RO)
A-8 RAC 97-X0506 (RO) 2.5 11
- RAC 97 X0248 (RO)
~A-9,
RAC 97-X0506 (RO) 2.6-1 RAC 97-X0248 (RO)
A-10 '
RAC 97-X0506 (RO) 2.6-2 RAC 97-X0248 (RO)
A-11 RAC 97-X0506 (RO) 2.6 3 RAC 97-X0248 (RO) l:
A-12 RAC 97-X0506 (RO) 2.6 RAC 97-X0248 (RO) 2.6-5 RAC 97-X0248 (RO)
CHAPTER 2, CONTENTS.
2.6-6 RAC 97-X0248 (RO) 2.6-7 RAC 97-X0248 (RO) bgg.
' RAC/ Revision 2.6-8 RAC 97-X0248 (RO) ii RAC 97-X0248 (RO) 2.6 9 RAC 97-X0248 (RO) i iii' RAC 97-X0248 (RO)'
2.6-10 RAC 97-X0248 (RO) iv.
RAC 97-X0248 (RO) 2.6-11 RAC 97-X0248 (RO) v RAC 97-X0248 (RO) 2.6-12 RAC 97-X0248 (RO) 2.6-13 RAC 97-X0248 (RO)
I CHAPTER 2 2.7-1 RAC 97-X0248 (RO) l 2.7-2 RAC 97 X0248 (RO)
. hgg RAC/ Revision 2.1-6 RAC 97-X0248 (RO)-
2.3-1 RAC 97-X0248 (R0) i 2.3-2
'RAC 97-X0248 (RO) i.,
2.3-3 RAC 97-X0248 (RO) 2.34 RAC 97-X0248 (RO) 2.35-RAC 97-X0248 (RO) 2.3-6 RAC 97-X0248 (RO)
SARUP-1 l
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SARUP-PORTS.
October 8,1999
[
LIST OF EFFECTIVE PAGES CHAPTER 4.3 (Continued)
CIIAPTER 4.3 (Continued)
RAC/ Revision hgg RAC/ Revision
~ bgg
- 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 (R1) 4.3 93 RAC 97-X0312 (R1) 4.3-125 RAC 97-X0312 (RI) l
'4.3-94.
RAC 97-X0312 (R1) 4.3-126 RAC 97-X0312 (R1) 1 RAC 99-X0041 (RO)
RAC 98-X0044 (RO) 4.3-95
- RAC 97-X0312 (RI)
RAC 99-X0041 (RO)
RAC 99-X0041 (RO) 4.3 126a RAC 98-X0044 (RO) 1 4.3 96'
. RAC 97-X0312 (R1) 4.3-127 RAC 97-X0312 (R1)
RAC 99-X0041 (RO) '
4.3-128 RAC 97-X0312 (RI) 4.3-97 RAC 97-X0312 (RI)
RAC 99-X0041 (RO) -
l 4.3-98 RAC 97 X0312 (RI) 4.3-129 RAC 97-X0312 (RI) l 4.3-99
~ RAC 97-X0312 (RI)
RAC 99-X0041 (RO)
L
' 4.3-100..
RAC 97-X0312 (RI) 4.3-129a RAC 99-X0041 (RO)
~ 4.3-101
. RAC 97 X0312 (RI) 4.3-130 RAC 97-X0312 (RI)
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 (RI)
RAC 99-X0041 (RO)
RAC 98 X0044 (RO) 4.3-104 RAC 97-X0312 (RI) 4.3-132 RAC 97-X0313 (RO) 4.3-105 RAC 97-X0312 (RI) 4.3 133 RAC 97-X0313 (RO)
/:
RAC 99-X0041 (RO):
4.3-134 RAC 97-X0313 (RO) s L 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 (RI)
RAC 99-X0079 (R0).
l
' 4.3 108 RAC 97-X0312 (R1) 4.3-136 RAC 97-X0313 (RO) -
4.3 109 RAC 97 X0312 (RI)
RAC 97-X0440 (RO) l RAC 99-X0041 (RO) -
RAC 99-X0079 (RO) 4.3-110 RAC 97-X0312 (RI) 4.3-137 RAC 97-X0313 (RO)
RAC 99-X0041 (RO)
RAC 99-X0079 (RO) 4.3-111 RAC 97-XO312 (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) l-RAC 99-X0041 (RO)
RAC 97-X0506 (R1)
. 4.3.-l 13 ~
RAC 97-X0312 (RI) 4.3 140 RAC 97-X0313 (RO)
RAC 99-X0041 (RO)
RAC 97-X0506 (R1) l-4.3-114 RAC 97-X0312 (RJ) 4.3-141 RAC 97-X0313 (RO)
RAC 99-X0086 (RO) 4.3-142 RAC 97-X0314 (RI)
.4.3-115 RAC 97-X0312 (RI) 4.3-143 RAC 97-X0314 (RI) l
. RAC 99-X0041 (RO) 4.3-144 RAC 97-X0314 (RI) 4.3 116 RAC 97-X0312 (RI) 4.3-145 RAC 97-X0314 (RI)
RAC 99 X0041 (RO) 4.3-146'.
RAC 97-X0314 (RI) 4.3-117 RAC 97-X0312 (RI) 4.3-147 RAC 97-XO314 (RI) 4.3-118 RAC 97-X0312 (RI) 4.3 148 RAC 97-X0314 (Rl) 4.3-119 RAC 97-X0312 (RI) 4.3-149 RAC 97-X0314 (RI) 4.3 120 RAC 97-X0312 (R1) 4.3-150 RAC 97-X0314 (RI) 4.3 121-RAC 97-X0312 (RI)
SARUP-7
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i SARUP-PORTS 1 October 3,1999
)
LIST OF EFFECTIVE PAGES
- CHAPTER 4.3, TABLES CHAPTER 4.3, TABLES (Continued)
Engg RAC/ Revision Pace RAC/ Revision l
T4!-l RAC 97-X0315 (RO)
F4.3-31 RAC 97-X0312 (R1)
. T4.3-2 RAC 97-X0316 (RO) ~
F4.3 32 RAC 97-X0312(RI)
T4.3-3 RAC 97-X0316 (RO)
F4.3-33 RAC 97-X0314 (RI)
T4.3-4 RAC 97-X0311 (R1)
T4.3 ' RAC 97-X0311 (RI)
. CHAPTER 4.4 T4.3-6 RAC 97-X0312 (RI)
T4.3-7 RAC 97-X0312 (RI)
Pace RAC/ Revision T4.3-8 RAC 97-X0312 (RI) 4.4-1 RAC 97-X0315 (RO)
T4.3-9 RAC 97-X0312 (RI) ~
4.4-2 RAC 97-X0315 (RO)
RAC 97-X0316 (RO).
T4.3-10 RAC 97-X0314 (RI)
T4.3-11 RAC 97-X0314 (R1)
RAC 97-X0316 (RO)
T4.3-12 RAC 97-X0314 (RI) 4.4-3 RAC 97-X0315 (RO)
- T4.3-13 RAC 97-X0315 (RO)
RAC 97-X0316 (RO)
F4.3-1 RAC 97-X0316 (RO) 4.4-4 RAC 97-X0315 (RO)
' F4.3 2 RAC 97-X0316 (RO)
RAC 97-X0316 (RO)
F4.3-3 RAC 97-X0316 (RO) 4.4-5 RAC 97-X0315 (RO)
F4.3-4 RAC 97-X0316 (RO)
RAC 97-X0316 (RO)
F4.3 5 RAC 97-X0316 (RO) 4.4-6 RAC 97-X0315 (RO)
F4.3-6 RAC 97-X0316 (RO)
RAC 97-X0316 (RO)
F4.3-7 RAC 97-X0316 (RO) ~
4.4-7 RAC 97-X0315 (RO)
- F4.3-8 RAC 97-X0316 (RO)
RAC 97-X0316 (RO)
F4.3-9 RAC 97-X0316 (RO)
RAC 97-X0312 (RI)
F4.3-10 RAC 97-X0311 (RI)
RAC 97-'X0314 (RI)
F4.3-11 RAC 97-X0311 (RI) 4.48 RAC 97-X0314 (RI) l F4.3-12 RAC 97-X0311 (RI) 4.4-9 RAC 97-X0314 (RI)
F4.3-13 RAC 97 X0311 (RI)
F4.3-14 RAC 97-X0311 (R1)
CHAPTERS F4.3-15 RAC 97-X0312 (RI)
' F4.3-16 RAC 97-X0312 (RI)
Egge RAC/ Revision RAC 97-X0505 (RS) 5.2-5 RAC 97 X0506 (RO)
F4.3-17 RAC 97 X0312 (RI) 5.2-5a RAC 97-X0506 (RO)
F4.3 18 RAC 97-X0312 (RI) 5.2A-1 RAC 97-X0314 (RI)
F4.3-19 :
RAC 97-X0312 (RI) 5.2A-2 RAC 97-X0314 (RI)
F4.3-20 RAC 97-X0312 (RI) 5.2A-3 RAC 97-X0314 (RI)
F4.3-20 RAC 97-X0312 (RI)
RAC 99-X0005 (RO)
F4.3-21 RAC 97-X0312 (RI) 5.2A 4 RAC 97-X0314 (RI)
F4.3 RAC 97-X0312 (RI)
RAC 99-X0005 (RO)
F4.3-23 RAC 97-X0312 (R1) 5.2A-5 RAC 97-X0314 (RI)
F4.3-24 RAC 97-X0312 (RI)
RAC 99-X0016 (RO)
- F4.3-25 RAC 97-X0312 (Rl) 5.2A 6 RAC 97-X0314 (RI)
F4.3-26 RAC 97-X0312 (RI)
RAC 99 X0023 (RO)
F4.3-27
. RAC 97-X0312 (RI) 5.2A RAC 97-X0314 (RI)
F4.3-28 RAC 97-X0312 (RI)'
5.2A-8 RAC 97-X0314 (R1)
F4.3-29.
RAC 97-X0312 (RI) 5.2A 9 RAC 97 X0314 (RI)
F4.3-30 RAC 97-X0312 (RI)'
SARUP-8
1 SARUP-PORTS October 8,1999 1
i LIST OF EFFECTIVE PAGES i
CIIAPTER 5 (Continued)
TECIINICAL SAFETY REQUIREMENTS
. hgg RAC/ Revision Pace RAC/ Revision 5.2A-10 RAC 97-XO314 (R1) vi RAC 97-X0505 (R5) 5.2A-11 RAC 97-X0314 (R1) vii RAC 97-X0505 (R5) 5.2A-12 RAC 97-X0314 (RI) viii RAC 97-X0505 (RS) 5.2A-13 RAC 97-X0314 (R1) ix RAC 97-X0505 (RS) 5.2A-14 RAC 97-X0314 (R1) x RAC 97-X0505 (RS)
RAC 99-X0022 (RO) 1.0-8 RAC 97-X0505 (RS) 5.2A 15 RAC 97-X0314 (RI) 1.0-8a RAC 97-X0505 (RS) 5.2A-16 RAC 97-X0314 (R1) 2.1-3 RAC 97-X0505 (RS) 5.2A-17 RAC 97-X0314 (RI) 2.1-4b RAC 97-X0505 (R3) -
3 5.2A-18 RAC 97-X0314 (R1)
RAC 97-X0505 (RS) 4 5.2A-19 RAC 97-30314 (RI) 2.1-6 RAC 97-X0505 (RS) 5.2A-20 RAC 97-X0314 (R1) 2.1-8 RAC 97-X0505 (R4) 5.2A-21 RAC 97-X0314 (R1) 2.1-10 RAC 97-X0505 (R4)
RAC 98-X0177 (RO) 2.1-12 RAC 97-X0505 (R5) 5.2A-22 RAC 97-X0314 (R1) 2.1-14 RAC 97-X095 (R4) 5.2A-23 RAC 97-X0314 (R1) 2.1-15 RAC 97-X0505 (R4) 5.2A-24 RAC 97-X0314 (RI) 2.1-17 RAC 97-X0505 (R4) 5.2A-25 RAC 97-X0314 (R1) 2.1-18 RAC 97-X0505 (R4) 5.2A 26 RAC 97-X0314 (RI) 2.1-19 RAC 97-X0505 (R3)
RAC 98-X0168 (RO) 2.1-10a RAC 97-X0505 (R4) 5.2A-27 RAC 97-X0314 (RI) 2.1 !!
RAC 97-X0505 (R5)
RAC 98 X0093 (RO) 2.1-12 RAC 97-X0505 (R4) 5.2A 28 RAC 97-X0314 (R1)
RAC 97-X0505 (RS) 5.2A-29 RAC 97-X0314 (RI) 2.1-23 RAC 97 X0505 'R5) 5.2A-30 RAC 97-X0314 (RI) 2.1 24 RAC 97-X0505 (R4) 5.2A-31 RAC 97-X0314 (R1) 2.1-27 RAC 97-X0505 (R4)
RAC 99-X0013 (RO) 2.1-28 RAC 97-X0505 (R4) 5.2A-32 RAC 97-X0314 (RI) 2.1-29 RAC 97-X0505 (R4) 5.2A-33 RAC 97-X0314 (R1) 2.1 30 RAC 97-X0505 (R4) 5.2A-34 RAC 97-X0314 (R1) 2.130a RAC 97-X0505 (R5)
RAC 98-X0108 (RO) 2.1 30b RAC 97-X0505 (RS) 5.2A-35 RAC 97-X0314 (R1) 2.1-30c RAC 97 X0505 (RS)
RAC 98-X0093 (RO) 2.1-30d RAC 97-X0505 (RS)
RAC 98-X0108 (RO) 2.1-31 RAC 97-X0505 (R4) 5.2A-36 RAC 97-X0314 (RI) 2.1-32 RAC 97-X0505 (R3)
RAC 98-X0037 (RO)
RAC 97-X0505 (R4)
RAC 98-X0108 (RO) 2.1 33 RAC 97-X0505 (R4) 5.2A-37 RAC 97-X0314 (RI) 2.1-34 RAC 97-X0505 (RS)
RAC 98-X0037 (RO) 2.2-5 RAC 97-X0505 (R4) 5.2A-38 RAC 97-X0314 (R1)
RAC 97-X0505 (RS) 5.4-2 RAC 97-X0506 (RO) 2.2-7a RAC 97-X0505 (R3) 5.4-3 RAC 97-X0506 (RO)
RAC 97-X0505 (R5) 5.4-6 RAC 97-X0506 (RO) 2.2-8a R AC 97-X0505 (R5) 5.4-7 RAC 97-X0506 (RO)
RAC 99-X0003 (RO) l 5.6-1 RAC 97-X0506 (RO) 2.2-11 RAC 97-X0505 (R4) l RAC 98-X0141 (RO) 2.2-12 RAC 97 X0505 (R4) 5.6-6 RAC 97 X0506 (RO) 5.6-7 RAC 97-X0506 (RO) 5.6-8 RAC 97-X0506 (RO) i i
SARUP-9 I
SAR-PORTS PROPOSED October 8,1999 RAC 97X0312 (RI),99X0086 (RO) excessive temperature in relation to the amount of UF in the cylinder. A cylinder is assumed to fail at some point 6
above its hydrostatic test pressure limit. Once the primary system fails, the release of UF material into the autoclave environment ofsaturated steam initiates the exothermic UF and H O reaction, which generates HF and 6
2 UO:F and causes a pressure increase inside the autoclave. The source term for this event will address the amount ofmaterial that could be released prior to autoclave isolation and subsequent leakages / relief paths once isolation is completed.
During operations in which the autoclave is closed for heating purposes, the only autoclave penetration that would allow release ofmaterial to the atmosphere is the condensate drain line. Any releases through this line would react with the condensate to form UO F and HF. Both of these compounds are sufficiently soluble in water 2
to allow very little, if any, release to the atmosphere. Once isolation of the autoclave is accomplished (typically less than 35 seconds for any significant release), the inflow of steam will be terminated to prevent additional moisture from entering the autoclave and reacting with the UF.. The pressure inside the autoclave will continue to rise until the limiting reactant (i.e., the moisture) is consumed. An analysis indicates that the pressure inside the autoclave will not reach the maximum allowable working pressure (MAWP) for any of the autoclaves given any of the autoclave / cylinder configurations. Therefore, no release due to autoclave over pressurization is considered.
Subsequent leakage of released material from the autoclave as a result ofleakage past valve seats and the autxlave shell seals is also considered. This leakage was quantified by assuming that the autoclaye is i functionally tested at 90 psig (0.72 MPa) with no more than 10 psi /h (69 kPa)/hr) or 12 scfm loss. Two cases I were considered. The first case assumed that the leakage occurs at the autoclave shell seal toward the top of the l autoclave resulting in a UF, vapor release. The release rate for this case was determined to be 3.5 lb/ min (1.6 l kg/ min). The second case assumed that the leakage occurs at the autoclave shell seal toward the bottom of the l autoclave resulting in a liquid UF release. The release rate for this case was determined to be 8.3 lb/ min (3.8 6
l kg/ min). These release rates are extremely low relative to the rates evaluated in Section 4.3.2.2.10, pigtail /line failure outside autoclave. Additional leakage through the autoclave seal may also be experienced if some 4
degradation occurs due to the environment inside the autoclave. However, the leak rate would have to be significant to result in any significant consequences beyond the immediate vicinity. This is not expected because less than 200 lb (91 kg) of HF is generated from the reaction with UF. Therefore, the releases from a closed 6
autoclave would be maintained below the EGs for the EBE frequency category.
c.
Consecuence Analysis -
l This is not applicable because the amount of release is bounded by the event evaluated in Section J
4.3.2.2.10.
d.
Comnarison With Guidelines l
The comparison with guidelines is subdivided to address the different receptors.
Local workcrs in the immediate area -Workers in the immediate area of the release could be exposed to a significant uranium dose or HF exposure. In the event of a release, the plant see and flee policy requires personnel to evacuate the area for their own protection. The essential method of detection for workers within these facilities is (1) visual indication of a " white smoke" (i.e., reaction products of UF and moisture) or (2) the 6
odor of HF, which is a product o.!the reaction of UF and moisture. The visual indication or the odor of HF will 6
provide indication of(1) the occurrence of a release and (2) the 4.3-114 i
e,
.i.
(
SAR-PORTS PROPOSED October 8,1999 RAC 97-X0314 (RI),99-X0005 (RO) and some ofits components are no longer in place. The WAESs include similar components as the seal exhaust l
systems; they are interconnected and use the same discharge path. In fact, they can be used as backups to the seal exhaust systems.
Nuclear criticality safety of the seal exhaust and wet air evacuation systems is based primarily on control j
of volume, interaction, and enrichment. Controls over these parameters are accomplished through a combination j
of administrative controls and passive barriers. The oil capacity of seal exhaust pumps is physically limited, using overflow lines if necessary, to a volume that is " safe" for the maximum uranium enrichment that could be present. Other passive baniers include the limited physical dimensions of alumina traps, roof vent oil traps, and piping, and the fixed spacing of pumps /demisters and alumina traps.
The double contingency principle is met and there are no AEFs identified for the seal exhaust and wet i
air evacuation systems.
1.4 Tails and Product (ERP and LAW) Withdrawal Stations The withdrawal stations for Extended-Range Prodcct (ERP), Tails, and Low-Assay Withdrawal (LAW) are located in the X-326, X-330, and X-333 buildings, respectively. The withdrawal stations are designed to remove gaseous UF. from the cascade, compress the gas, condense the gas to a liquid state via cooling, and drain the liquid into cylinders. Each withdrawal station contains one or more loops of equipment and multiple withdrawal positions. A withdrawal station equipment loop generally consists of a first stage compressor, a cooler, a second stage compressor, one or more condensers, accumulators, and a withdrawal header.
Nuclear criticality safety of the withdrawal stations is based primarily on mass, concentration, and moderation control. Controls over these parameters are accomplished through a combination of administrative controls, passive barriers, and AEFs. Limits on the mass of uranium in the scale pit (s), equipment housings, j
coolant system drain tank (Tails), and the exhaust duct HEPA fiher (LAW) are maintained through administrative l controls such as visual inspections and operator response to process gas releases. Passive barriers such as UF6 l pipes and cylinders also serve to control mass.. Administrative controls such as periodic solution sampling control fissile material concent stion, as does the passive banier formed by containing process gas and coolant in separate loops. Moderation control is achieved primarily by administratively limiting the pressure in the condensers to less than 60 psia, which ensures that the H/U ratio is s 0.088. Moderation control is also supported by passive barriers such as the design (including pressure differentials) and physical integrity of portions of the system containing process gas to prevent moderator entry.
In addition to NCS controls over the above parameters, enrichment is administratively corcrolled to a maximum of 10% lF" at ERP and LAW, and a maximum of 5% at Tails. Geometry is controlled at various points in the withdrawal station equipment loops by the physical dimensions of the condensers, accumulators, and piping. Interaction is controlled by the fixed relationship of equipment items to one another. Neutron absorption control exists in the form of borosilicate glass raschig rings in the scale pits.
l l
5.2A-3 i
l SAR PORTS PROPOSED October 8,1999 RAC 97-X0314 (R1),99-X0005 (RO)
In order for a criticality to be possible, multiple contingency events would need to occur simultaneously.
l For instance, product cylinders have assay (10% limit) and moceration controls when being filled. The assay is l normally monitored by spectrometers that monitor enrichment values. At up to 12% enriclunent, the maximum l cylinder size for UF, withdrawal is safely subcritical provided moderation control is maintained. All cylinders l are filled under moderation control, through controlling the maximum UF condensing pressure to 45 psig so that l gaseous HF will not go into the UF, liquid being withdrawn. Hence, multiple failures would have to occur for l a tilled cylinder to reach a critical configuration.
I l
Active Engineered Features are included for ERP and LAW to ensure proper functioning of NCS l controls. RCW control valves close upon receiving a signal from RCW and R-114 pressure switches. See SAR j Section 3.8 for details including safety classification.
1.5 Freon Degraders The operational Freon degradation units, or Freon Degraders, use an electrically heated combu.; tion vessel, with fluorine as an oxidizer, to decompose R-114 (Freon) coolant into lighter gases which can be more readily purged from the Cascade. The 5-inch diameter combustion vessel, or degrader vessel, is contained within an unfavorable geometry steel shell that is pressurized with a nitrogen buffer.
Administrative controls and passive barriers have been incorporated to prevent a criticality from occurring in the Freon Degraders. Administrative controls consist of procedural requirements for responding to instrumentation and/or alarms by shutting down the Freon degrader if a leak is detected or suspected. The nitrogen buffer pressure is maintained greater than or equal to 1 psig while the system is operating, and the process gas pressum is maintained below atmospheric pressure. The safe geometries of the degrader vessel and of the mixer, fiher columns, and all other piping and components, with the exception of the degrader vessel outer shell, constitute passive barriers.
In order for a criticality to be possible, multiple contingency events would need to occur simultaneously.
For instance, wet air inleakage to the geometrically unfavorable annulus surrounding the degrader vessel (first contingency) would need to occur in conjunction with leakage of UF from the process gas in the degrader vessel 6
out into the annulus (second contingency). Therefore, the double contingency principle is met.
There are no AEFs identified for the Freon Degrader.
1.6 Cascade Support Systems The evacuation booster stations (EBSs) and cascade datum systems are included as cascade support systems. The EBSs in the X 330 and X-333 buildings are used primarily for purging cells in the cascade so that maintenance can be performed. They are also used for evacuating fluorine, CIF, or other light gases from the 3
cells during conditioning operations as well as pumping the contents of one bank of surge drums to another, pumping from surge drums to the cascade, and senice auxiliary systems such as Tails, LAW, and :he holding drums. Datum systems are control systems made up almost entirely of 5.2A-4 h
_g
l SAR-PORTS PROPOSED October 8,1999 RAC 97-X,0314 (R1),99-X0022 (RO)
In order for a criticality to be possible, multiple contingency events would need to occur simultaneously.
For instance, in the " Blue Room" a non favorable geometry container is used to store solid material removed from l
equipment (failure of contingency control). The second control for this contingency is the requirement that the mass be limited based on number of parts to be cleaned. This control limits the quantities of uranium solid on parts that could be placed in a container. Double contingency is met for the above operations.
There am no AEFs identified for the operation of the small parts pit, blue room, seal dismantling room and the small cylinder rinse pit.
3.4 Handtable Operations A number of difTerem handtables are used throughout the X-705 building. These handtables include the Small Parts Handtables, the Leaching / Complexing Handtables and the Recovery Handtables. The handtables are primarily used for decontaminating seals and other small parts (i.e., pigtails, cylinder valves, monel filters, control valves, nuts and bolts), leaching uranium from various solid materials, complexing and acidifying solutions, pouring of uranium-bearing liquids from polybottles or any other safe geometry containers into the handtable, and cleaning bag filters from any of several locations in the building.
A number of administrative controls and passive barriers exist for the handtable operations to prevent l a criticality from occurring. The administrative control includes keeping the opening to the overflow draid free ofdebris and obstructions, limiting the depth of solids placed into the screen baskets, and maintaining edge-to-edge spacing as specified in the NCSA of all containers of uranium-bearing material, except when pouring materials into /onto a handtable. The physical design and construction of the handtable provides passive barriers.
l An overflow drain exists which channels flow to the overflow container located underneath the handtable. The overflow is at a height which prevents collection ofliquid in the table tray in the event the primary drain should l become blocked.
In order for a criticality to be possible, multiple contingency events would need to o: cur simultaneously.
l l For instance, the overflow drain becomes blocked by foreign material and allows accumulation of the solution l
(loss of geometry control) and a container of uranium-bearing materials is dropped cuto the handtable (loss of l interaction control). The double contingency principle is met by having one overflow drain, and the fact that the operator would have to violate procedural requirements by placing or dropping at least two containers onto the l Handtr.ble before a criticality is possible. There are no AEFs identified for handtable operations.
3.5 Waste Water Treatment (Microfiltration System)
The X-705 Microfiltration/ Pressure Filtration Treatment System, one of seven major pieces of equipment or systems for waste water treatment. A 7000-gallon geometrically safe storage system is provided in X-705 to l
collect waste water effluents from ten different sources of waste water, most of which are cleaning or rinse j
solutions. From the 7000-gallon storage system, the waste water is J
l 5.2A 14
~
t SAR-PORTS PROPOSED October 8,1999 RAC 97-XD314 (RI),98-X0168 (RO) 4.0 Maintenance and Sunnort Facilities Nuclear criticality safety controls associated with fissile material operations conducted in the X-700 and X-720 maintenance and support facilities are summarized in this section.
4.1 X-700 Converter Disassembly and Repair Area A weld shop in the West High Bay of Building X-700 is used for disassembly and repair ofleaking l converters and coolers from the process buildings. The weld shop area consists simply of floor space, welding l equipment, and a cart to rotate and transport the converter or cooler.
Administrative controls are in place to prevent a criticality in the X-700 converter disassembly and repair area by limiting both the mass of fissile material and the quantity of moderator available. Mass is controlled by l requiring that deposits of uranium-bearing materials in equipment be within safe mass limits. Moderators are l excluded from the converters and coolers by requiring process gas openings to be covered except when converters l are actually being worked on, and by limiting moderating materials in the equipment to small quantities such as a damp cloth or spray bottle for cleaning if necessary.
In order for a criticality to be possible, multiple contingency events would need to occur simultaneously.
For instance, a converter would need to be sent to the disassembly and repair area without first having a q
l significant (greater than safe mass) uranium-bearing deposit removed (first administrative control failure), and l an unsafe quantity of moderator would need to be allowed to enter the converter (second administrative control failure). Therefore, the double contingency principle is met and there are no AEFs identified for the X-70f, converter disassembly and repair operations.
l 4.2 Maintenance Shop Fissile Operations The Maintenance Shops that handle fissile operations do not handle large quantities of fissile material.
Normally, equipment has either been decontaminated prior to being brought to the Maintenance shops or there j
is little likelihood of significant contamination.
The controls are a combination of passive engineered features controlling geometry and spacing of solution handling equipment and administrative controls on the type of equipment to be cleaned or handled by the shop.
All of these operations meet :he double contingency principle with a combination of passive engineered and administrative controls and there are no AEFs identified for the maintenance shop fissile operations.
5.0 X-710 Laboratory Facility Nuclear criticality accident scenarios associated with fissile material operations conducted in the X-710 laboratory facility are summarized in this section.
5.2A-26
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TSR-PORTS PROPOSED October 8,1999 RAC 97XD505 (R5),99X0003 (RO)
SECTION 2.2 SPECIFIC TSRs FOR X-330 AND X-333 FACILITIES 2.2.3 LIMITING CONTROL SETTINGS, LIMITING CONDITIONS FOR OPERATION, SURVEILLANCES -
2.2.3.3 CADP UF Smoke Detection System (continued) j 6
(
BASIS:
In the event of a UF release in a cell, bypass housing, tie-line, or booster station an alarm will sound l
6 in the ACR notifying operating personnel that immediate investigation and action must occur. The CADP system is sensitive enough to detect very minor out gassings of UF and therefore will provide 6
operators sufficient time to take any actions necessary to minimize the amount of UF released [SAR 6
Sections 3.8.7.3, 4.3.2.1.1, 4.3.2.1.2, 4.3.2.1.3, and 4.3.2.1,7].
l
)
l 2.2-8a
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