ML20098A451
| ML20098A451 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, Catawba, McGuire  |
| Issue date: | 04/30/1995 |
| From: | Boyd T, Klarner R, Reiter D BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20098A448 | List: |
| References | |
| BWI-222-7693-LR, BWI-222-7693-LR-01, BWI-222-7693-LR-1, NUDOCS 9509260005 | |
| Download: ML20098A451 (160) | |
Text
{{#Wiki_filter:_ Page 1 of 160 DUKE POWER COMPANY MCGUIRE 1 & 2 AND CATAWBA 1 NUCLEAR GENERATING STATIONS REPLACEMENT STEAM GENERATOR 4 TOPICAL REPORT BABCOCK & WILCOX INTERNATIONAL BWI-222-7693-LR-01 APRIL 1995 REV.4 STATION: MNS/CNS , DPC SPECIFICATION: DPS-1201.01-00-0002 DPC PURCHASE ORDER: C23355-68 DPC DES. ENG. FILE NO.: MC/CN 1201.37 BWI CONTRACT NO: 222-7693n700n701 Prepared by: d Date: 004e Sf/Y' D. P. Reiter, Project Engineer V f ed by Date: '
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'It Klarner, P.Eng., [
Supervisor, Performance & Proposals Group Approved by: ,O- Date: 28 /9 < T. Boyd, P.Eng., Project Engineer f J I Approved by: 1 Date: bL ~24 / 9[ ; M. D! Lebs,[ I Manager, Nuclear Engineering 7693IJt08. SET 1. DUKE
- l 1
9509260005 950918 PDR ADOCK 05000413 l P _. - -_ PDR
r
- f. BWI-222-7693-LR-01. Page 2 !
i t LIST OF REVISIONS l
- r
- REV. ,
- NO. DATE ' PAGE . DESCRIPTION 1 0- Initial Release 1 Sept./94 Table 2.1-1 revised to delete proprietary data. !
~
2 Nov./94 Cover Revised to Rev. 2. i All Editorial change: Added header with ; page number to each page. Deleted
- page number at bottom.
- 4 through 11 Deleted Section 2.7.4. ' Revised page t numbers.
15,16 Revised. Title of Figure 2.2.5-5. Added , page numbers. i 17 Revised page numbers, j ~27 Deleted " typical"in 2 places and < j changed " Typical"to "The". l I L 4 29 Revised Figure 2.2.1-2. ' 1 30 Revised Figure 2.2.1-3. 1 31 Revised Figure 2.2.1-4. , 36 Revised text to specific DPC geometry. 37,38 Revised medium bar width and revised j text in 2.2.4.2.1 Deleted " typical" from
- 2.2.4.2.2and 2.2.4.2.3. Made reference 2
to Figure 2.2.4-4. 40 Text revised in Section 2.2.4.3.3. Added Ref. 3. 44 Revised Figure 2.2.4-4. I i 45 Revised Figure 2.2.4.-5. 52 Section 2.2.5.7 revised. 55 Revised text on Figure 2.2.5-2. ! 58 Revised text on Figure 2.2.5-5. , i
i BWI-222-7693.LR-01 Page 3 ! REV. NO. DATE PAGE DESCRIPTION . i 73 Section 2.2.8 revised. The word i
" typically" deleted from 2.2.9. -
77,78 Revised text in Section 2.2.12. 79 Deleted sentence in Section 2.2.13.1. 81 Deleted text in 2 places. 82 Changed " tests" to " experience". 84 Added last para, to 2.3.1. ; 85 Added Table references and 3rd l paragraph. 86 Clarified surface finish in 2.3.2.3 and ' moved last paragraph of 2.3.3 to Section 2.3.1. 95 Revised text of 2.4.1.2 to identify Code Edition. i 97 Added text for GDC 4. 103 Revised text. 105 Revised text for NUREG 0909. l 108 Added text under " Loose Parts". ! i 116 Revised solution to problems 32 and 34. i 136 Revised last paragraph of 2.7.2. Revised j 2.7.3. Deleted 2.7.4. , ! i 138 Deleted P,, design stress requirement and ) j defined S,. Added "and NB-3223". 139 Revised text in 1st paragraph and Level C requirements. 151 Revised text in 3.2.6.6 - . 3 Dec194 140 Revised S, value. 141~ Made typographical corrections and revised S, value. J 4 Apr.195 Cover Revised to Rev. 4 i
BWI-222-7693-LR-01 Page 4 REV. NO. DATE PAGE DESCRIFflON 4 through 160 Page numbers revised to reflect additional pages in List of Revisions. 7 through 14 Revised page numbers in Table of Contents. 18,19 Revised page numbers in List of Figures. 20 Revised page numbers in List of Tables. 21 Deleted "the same". 22 Item 4 changed "normally found in" to "found in presently operating". Changed "same" to similar. 26 Changed "I" to " Alloy". 29 Revised format. 39 Added to first sentence of first paragraph. Added description to lattice grid tube supports. Clarified text. 41 Added " tube lockup" and "via the J-tabs". 42 Changed " closing" to " operating". 43 Deleted " Preliminary" and changed
" indicates" to " indicate". Changed "41S" to "410S".
44 Added date to reference 3. 45 Replaced Figure 2.2.4.1. 52 Added "as well as... loads" to last sentence of 2nd paragraph. 53 Revised next to last paragraph for clarity. 55 Changed "AT CNS" to "For example," and "is" to "can be". . 3 e e
BWI-222-7693-LR-01 Page 5-f i i REV. l NO. DATE PAGE DESCRIPTION ,
- 56 Added " internal". Revised text in 2nd :
! paragraph of 2.2.5.5 for clarity. Revised text in 2.2.5.6 for clarity. . Added last sentence to first paragraph , of 2.2.5.7. 57 Added "the addition of" to the 1st sentence of 2.2.5.8. f l 58 and 59 Revised Figures 2.2.5.1 and 2.2.5.2.
, 63 Revised text for clarity under item 3 of r design features and under item 3 of features that preclude blockage.
66 Revised text for clarity in 3rd paragraph of 2.2.7.1.
- 68 Revised text for clarity in 2.2.7.7 and
- 2.2.7.8.
70 Revised Figure 2.2.7.1. l
- ~ 1 73 Revised Figure 2.2.7.4. l 75 and 76 Revised Figure 2.2.7.6 and 2.2.7.7.
77 Changed " addition" to " additional" in i first paragraph. i .
- 86 Deleted " typically" from item 3.
88 Revised last sentence of 1st paragraph j of 2.3.1 for clarity. 4 89,90 and 91 Revised Section 2.3.2.2 for clarity. Added "or 690" and changed "0.02" to "0.10" in Section 2.3.2.3. Revised i Section 2.3.3 for clarity, j 92 and 93 Changed "316N" to "F316N/316LN". 96 Revised material specifications. ~
]
97 Revised material speci6 cations and added components. 98 Added J-tabs. l' 4
~ a m,a l BWI-222-7693-LR-01 Page 6 REY.
NO. DATE PAGE DESCRIPTION 103 Revised last sentence under RG 1.50 for clarification. 111 Changed "close" to " minimize" and
" control boiling... tube bundle" to " low qualities at the tubesheet".
I12 Added "with the alloy 690 interface" and "All threaded... contained.". 118 thm 121 Added " Alloy 690 material" to solution of problems 8,9,12, and 15. Revised
" sulphate" to " sulfate" in problem 15.
Deleted problem 30 and renumbered following problems accordingly. 122 Deleted problem 30 from Figure 2.5-1 and renumbered following problems accordingly. 140 Revised " sulphate" to " sulfate". 156 Deleted "after installation" from Section 3.2.6.6, item 2.
-. _~_ . -
1 BWI-222-7693-LR-01 Page 7 j A ! i i DUKE POWER COMPANY REPLACEMENT STEAM GENERATOR TOPICAL REPORT TABLE OF CONTENTS Pages
?
LIST OF ACRONYMS AND GLOSSARY . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-17 LIST OF FIG URES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18-19 LIST O F TABLES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 EXECUTIVE
SUMMARY
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
- 1. INTRODUCTION AND PURPOSE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22
- 2. REPLACEMENT STEAM GENERATOR DESIGN . . . . . . . . . . . . . . . 23-146 2.1 GENERAL DESCRIPTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23-28 2.1.1 Qualifications of the Steam Generator Supplier . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23-24 l
2.1.2 Millstone 2 Replacement Steam Generator Design and Experience . . . . . . . . . . . . . . . . . 24-25 2.1.3 Comparison with Existing Design . . . . . . . . . . . . . . . . . . . . 25 2.2 STEAM GENERATOR DESIGN HIGHLIGHTS . . . . . . . . . . . . . 29-87 ! 2.2.1 Pressure Boundary Design . . . . . . . . . . . . . . . . . . . . . . . 29-35 2.2.1.1 Tubesheet Assembly and Primary Divider Plate . . . . . . . . . . . . . . . . . . . 29 2.2.1.2 Closure Design .......................30 2.2.1.3 Shell and Nozzle Design . . . . . . . . . . . . . . . . . 30
-- 1
BWI-222-7693-LR-01 Page 8 A 2.2.2 Steam Generator Tube Design . . . . . . . . . . . . . . . . . . . . . . 36 2.23 Tube-to-Tubesheet Joint . . . . . . . . . . . . . . . . . . . . . . . . 36-38 2.23.1 Tube-to-Tubesheet Welding . . . . . . . . . . . . . 36-37 2.23.2 Hydraulic Expansion . . . . . . . . . . . . . . . . . . . . 37 2.233 Inservice Inspection . . . . . . . . . . . . . . . . . . . . . 37 2.2.4 Tube Bundle Support System . . . . . . . . . . . . . . . . . . . . . 39-51 , 2.2.4.1 Lattice Grid Tube Supports . . . . . . . . . . . . . 39-40 2.2.4.2 U-Bend Supports . . . . . . . . . . . . . . . . . . . . 40-42 2.2.4.2.1 Design , Configuration . . . . . . . . . . . . 40-41 2.2.4.2.2 Flow Characteristics . . . . . . . . . . . . . 41 2.2.4.23 Flexibility and : Thermal Motions . . . . . . . . . . . 41 2.2.4.2.4 Support of FUR Assembly . . . . . . . . . . . . . . . 41-42 2.2.4.3 Design to Minimize Flow Induced Vibration (FIV) . . . . . . . . . . . . . . . 42-44 i 2.2.4.3.1 Clearances . . . . . . . . . . . . . . . . 42 . l 2.2.4.3.2 Bar Width . . . . . . . . . . . . . . . . 42 2.2.4.33 Fretting Assessment . . . . . . . . . . . . . . 42-43 , 2.2.5 Internal Feedwater System . . . . . . . . . . . . . . . . . . . . . . . 52-62 4 2.2.5.1 The Water Hammer Mechanism in Feedwater Piping . . . . . . . . . . . . . . . . . . 52-53 2.2.5.2 Design to Preclude Water Hamm er . . . . . . . . . . . . . . . . . . . . . . . . . . . 53-54
4 a 4ad t A . w. .nw..._.4- A .. aw4...r_ a,__ A ~ ._ a. _M BWI-222-7693-LR-01 Pzgs 9 i TABLE OF CONTENTS b ! i
- (continued) \ ,
. Pages ,
. 2.2.5.3 Thermal Stratification .
- Mechanism . . . . . . . . . . . . . . . . . . . . . . . . . 54-55
] 2.2.5.4 Design to Minimize Strati 6 cation ! Susceptibility . . . . . . . . . . . . . . . . . . . . . . . . . 55 i. j 2.2.5.5 Thermal Sleeve . . . . . . . . . . . . . . . . . . . . . . . . 56 l 2.2.5.6 Feedwater Distribution . . . . . . . . . . . . . . . . . . 56 2.2.5.7 Maintenance Features . . . . . . . . . . . . . . . . . 56-57 1 2.2.5.8 Auxiliary Feedwater' System . . . . . . . . . . . . . . . 58 l 2.2.6 Blowdown System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 63-64 : 4 ! 2.2.7 Moisture Separator System . . . . . . . . . . . . . . . . . . . . . . 66-76 ' 2.2.7.1 Prima y Separators . . . . . . . . . . . . . . . . . . . . . 66 4 2.2.7.2 Secondary Separators . . . . . . . . . . . . . . . . . 66-67 l 4 , ' l 2.2.7.3 Separator Performance . . . . . . . . . . . . . . . . . . 67 l i 2.2.7.4 Sensitivity to Operating I Pressure Fluctuations . . . . . . . . . . . . . . . . . 67-68 i ) 2.2.7.5 Sensitivity to Water Flow i Fluctuations . . . . . . . . . . . . . . . . . . . . . . . . . . 68 1 2.2.7.6 Sensitivity to Water I2 vel Fluctuations . . . . . . . . . . . . . . . . . . . . . . . . . . 68 2.2.7.7 Steam Carryunder . . . . . . . . . . . . . . . . . . . . . . 68 1 2.2.7.8 Separator Design Life . . . . . . . . . . . . . . . . . . . 68 2.2.8 Minimized Weld Design . . . . . . . . . . . . . . . . . . . . . . . . . . 77 2.2.9 Integral Flow Restrictor . . . . . . . . . . . . . . . . . . . . . . . . . . .' 77 2.2.10 Nozzle Dams . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 77-79 i
BWI-222-7693-LR-01 Page 10 TABLE OF CONTENTS b (continued) Pages 2.2.11 Provisions for Inspection . . . . . . . . . . . . . . . . . . . . . . . . 80-81 2.2.12 Electro-chemical Polishing . . . . . . . . . . . . . . . . . . . . . . . 81-82 2.2.13 Provisions for ALARA . . . . . . . . . . . . . . . . . . . . . . . . . 83-85 2.2.13.1 Material Selection and Design to Minimize Personnel Exposu re . . . . . . . . . . . . . . . . . . . . . . . . . 83-84 2.2.13.2 Minimization of Inspected Welds ..............................84 2.2.13.3 Water Chemistry . . . . . . . . . . . . . . . . . . . . . . . 84 2.2.13.4 Minimization of Personnel Exposu re . . . . . . . . . . . . . . . . . . . . . . . . . . 84-85 2.2.14 Provisions for Chemical Cleaning . . . . . . . . . . . . . . . . . . 85-86 2.2.15 Water Level Stability and Control . . . . . . . . . . . . . . . . . 86-87 2.3 STEAM GENERATOR MATERIALS . . . . . . . . . . . . . . . . . . . . . 88-98 2.3.1 Pressure Boundary Materials . . . . . . . . . . . . . . . . . . . . . . . 88 2.3.2 Critical-to Function Materials . . . . . . . . . . . . . . . . . . . . 88-90 1 2.3.2.1 RSG Tube Material . . . . . . . . . . . . . . . . . . 88-89 2.3.2.2 Tube Support Materials . . . . . . . . . . . . . . . . 89-90 2.3.2.3 Corrosion Resistant Cladding . . . . . . . . . . . . . . 90 2.3.2.4 Feedwater Headers . . . . . . . . . . . . . . . . . . . . . 90 2.3.3 SG Internals Materials . . . . . . . . . . . . . . . . . . . . . . . . . 90-91 2.3.4 Archive Samples . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 91 I
BWI-222-7693-LR-01 Page 11 TABLE OF CONTENTS b (continued) i t 2.4 RSG DESIGN BASES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 99-110 ;
'2.4.1 Codes and Standards . . . . . . . . . . . . . . . . . . . . . . . . . . 99-100 2.4.1.1 Code Reconciliation . . . . . . . . . . . . . . . . . . . . 99 i 2.4.1.2 ASME Certified Design .
Specification . . . . . . . . . . . . . . . . . . . . . . . . . . 99 r 2.4.1.3 Tests and Inspections . . . . . . . . . . . . . . . . . . 100 ) 2.4.1.4 N-Stamp . . . . . . . . . . . . . . . . . . . . . . . . . . . . 100 2.4.2 Comparison to NRC Guidance . . . . . . . . . . . . . . . . . . 100-109 2.4.2.1 NRC General Design Criteria . . . . . . . . . . 100-102 l 2.4.2.2 NRC Regulatory Guides . . . . . . . . . . . . . . 102-106 , i 2.4.2.3 Comparison to NRC Standard Review Plans . . . . . . . . . . . . . . . . . . . . . . 106-107 ! l 2.4.2.4 Generic Letters, Bulletins, l Notices and NUREGs . . . . . . . . . . . . . . . . 107-109 1 3 2.4.3 Seismic Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . 109 ) I 2.4.4 Performance Requirements . . . . . . . . . . . . . . . . . . . . . . . 109 l 2.4.5 Accidents and Transients . . . . . . . . . . . . . . . . . . . . . . . . . 110 ' 2.5 DESIGN IMPROVEMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . 111-128 , I 2.5.1 Minimization of Corrosion . . . . . . . . . . . . . . . . . . . . . . 112-115 j- 2.5.1.1 M'aterials Employed to Minimize Corrosion . . . . . . . . . . . . . . . . . 112-114 2.5.1.2 Design Features Employed to t Minimize Corrosion . . . . . . . . . . . . . . . . . 114-115 2.5.2 Minimization of Loose Parts . . . . . . . . . . . . . . . . . . . . . . 116 e
. :i BWI-222-7693-LR-01 Page 12 TABLE OF CONTENTS A (continued) i
! Pages 2.5.3 RSG Performance Improvements . . . . . . . . . . . . . . . . . . . 116 2.5.4 Maintenance and Reliability i Improveme nts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 116-117 2.5.5 Inservice Inspection Improvements . . . . . . . . . . . . . . . . . . 117 I 2.6 CONFIRMATORY ANALYSIS AND TESTING . . . . . . . . . . . . 129-139 2.6.1 Flow-Induced Vibration . . . . . . . . . . . . . . . . . . . . . . . . 129-131 i 1 2.6.2 Thermal-Hydraulic Performance . . . . . . . . . . . . . . . . . 131-134 i 2.6.2.1 Three-dimensional Thermal Hydraulic Analysis . . . . . . . . . . . . . . . . . . 132-133 2.6.2.2 The ATHOS Computer Code . . . . . . . . . . 133-134 l 2.6.3 Tube-to-Tubesheet Joint
- Qualification Program . . . . . . . . . . . . . . . . . . . . . . . . . 134-137
} 2.6.4 Separator Testing Experience . . . . . . . . . . . . . . . . . . . . 137-138 ! 2.6.5 Chemical Cleaning Qualification of Materials . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 8 2 2.7 OPERATING RESTRICTIONS WITH RSG DESIGN . . . . . . . . 140-141- ! 2.7.1 Removal of Temporary U-Bend Shipping '. Restrain ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 140 . 2.7.2 Primary Water Chemistry . . . . . . . . . . . . . . . . . . . . . . . . 140 2 l 2.7.3 Secondary Water Chemistry . . . . . . . . . . . . . . . . . . . . . . . 140 2.8 RSG STRUCTURAL EVALUATION . . . . . . . . . . . . . . . . . . . . 142-145 2.8.1 Tu bing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 143-145 2.9 STARTUP TESTING REQUIREMENTS . . . . . . . . . . . . . . . . . . .. . 146 1 1
BWI-222-7693-LR-01 Page 13 TABLE OF CONTENTS b' (continued) Pages
- 3. REPLACEMENT STEAM GENERATOR FABRICATION . . . . . . . . . 147-160 3.1 QUALITY ASSURANCE PROGRAM . . . . . . . . . . . . . . . . . . . . 147-149 3.1.1 Design Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 147-148 3.1.2 Docum ent Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 148 L 3.1.3 Corrective Action . . . . . . . . . . . . . . . . . . . . . . . . . . . . 148-149 3.1.4 Non-conforming Items . . . . . . . . . . . . . . . . . . . . . . . . . . . 149 i 3.1.5 Q A R eco rds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 9
! 3.1.6 Au dits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 9 4 't 3.2 FABRICATION CONTROL . . . . . . . . . . . . . . . . . . . . . . . . . . . 150-160 l 3.2.1 Control of Purchased Items and Services . . . . . . . . . . . . . 150 3.2.2 Control of Manufacturing Processes . . . . . . . . . . . . . . . 150-151 4 3.2.3 Control of Consumables . . . . . . . . . . . . . . . . . . . . . . . 151-152 3.2.4 Control of Specialized Processes . . . . . . . . . . . . . . . . . 152-154-3.2.4.1 Tube-to-Tubesheet Welding . . . . . . . . . . . . . . 153 3.2.4.2 Hydraulic Tube Expansion . . . . . . . . . . . . . . . 153 j 3.2.4.3 Electro-polishing . . . . . . . . . . . . . . . . . . . . 153-154 l 3.2.5 ' Material Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 154 l l
, 3.2.6 Shop Tests and Inspections . . . . . . . . . . . . . . . . . . . . . 154-156 1
3.2.6.1 Test and Inspection Equipment . . . . . . . . . . . 154 3.2.6.2 Tests and Inspections of Forgings . . . . . . . . . . . . . . . . . . . . . . . . . . 154-155 .
BWI-222-7693-LR-01 Page 14 TABLE OF CONTENTS b (continued) 4 Pages , 3.2.6.3 Tests and Inspection of Tubing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 155 3.2.6.4 Welds ............................. 155 i 3.2.6.5 Steam Generators . . . . . . . . . . . . . . . . . . . . . 155 3.2.6.6 Baseline Eddy Current ( Inspection . . . . . . . . . . . . . . . . . . . . . . . . . 155-156 1 3.2.7 Handling, Storage and Shipping . . . . . . . . . . . . . . . . . . 156-158 3.2.7.1 Cleanliness . . . . . . . . . . . . . . . . . . . . . . . . 156-158 3.2.7.1.1 Procedures . . . . . . . . . . . . . . . 156 3.2.7.1.2 Clean Room . . . . . . . . . . . . 157-158 3.2.8 Receipt Inspection Requirements . . . . . . . . . . . . . . . . . 158-160 3.2.8.1 Preparation for Shipment . . . . . . . . . . . . 15 8- 15 9 3.2.8.2 Handling and Shipping . . . . . . . . . . . . . . . . . 159 3.2.8.3 Inspection at Jobsite . . . . . . . . . . . . . . . . . . . 159 3.2.8.4 Storage ............................. 160 i M J
l l BWI-222-7693-LR-01 Pags15 LIST OF ACRONYMS AND GLOSSARY
- Terms and acronyms used in this report are defined the first time they are used in the text.
i The more significant and widely used acronyms and terms are defined below. Acronyms l ALARA As Low as Reasonably Achievable
- ANS American National Standard i
ANSI American National Standards Institute , ASME American Society of Mechanical Engineers ! B&W Babcock & Wilcox { BTP NRC Branch Technical Position (appended to SRPs) BWI ~ Babcock & Wilcox Industries i CANDU Canadian Deterium Uranium heavy water reactor design i; CDS Certi6ed Design Specification { CFR Code of Fedeal Regulations
~ CG Center of Gravity j CMS Corrosion Monitoring System
- DBE Design Basis Earthquake ECT Eddy Current Test
, EP Electro-chemical Polishing EPRI Electric l'ower Research Institute l F Degrees Fahrenheit FEI Fluid Elastic Instability
- FIV Flow Induced Vibration
! FW Feedwater j FSAR Final Safety Analysis Report ' FUR Flat bar U-bend restraint
- GDC NRC General Design Criteria
; GTAW Gas Tungsten Arc Welding
- hr Hour I.D. Inside Diameter 4 IGA Intergrannular Attack
! ISI In-Service Inspection kips Thousand pounds (load) ksi Thousand pounds per square inch LBB Leak Before Break LBLOCA Large Break LOCA LOCA Loss Of Coolant Accident MIG Metal Inert Gas welding process
. MFW Main Feedwater system
- MSLB Main Steam Line Break l MP2 Millstone Plant, unit 2 4
NRC United States Nuclear Regulatory Commission OBE Operational Basis Earthquake Y
BWI-222-7693-LR-01 Page 16 O.D. Outside Diameter OSG Original Steam Generator Owner Utility psi Pounds per square inch psia Pounds per square inch, absolute psig Pounds per square inch, gauge PWHT Post Weld Heat Treatment PWSCC Primary Water Stress Corrosion Cracking QA Quality Assurance RA Roughness Average, in micro-inches RCS Reactor Coolant System RG NRC Regulatory Guide RSG Replacement Steam Generator RUB Reverse U-bend SCC Stress Corrosion Cracking scfm Standard cubic feet per minute SBLOCA Small Break LOCA SG Steam Generator SMAW Shielded Metal Arc Welding SRP NRC Standard Review Plan (collected in NUREG 0800) SSE Safe Shutdown Earthquake Tech Spec Technical Specification (s) TFL Tube Free Lane TS Technical Specifications UA Heat transfer capacity (BTU /hr 'F) UT Ultrasonic test USNRC United States Nuclear Regulatory Commission e 6 e
BWI-222-7693-LR-01 Page 17 Glossary Circulation Ratio - the ratio of steam generator tube bundle (riser) flow to steam flow. Denting ' steam generator tube deformation caused by corrosion product interference at tube support plates. Downcomer - the annular space between the tube bundle shroud and shell that channels recirculated water to the base of the tube bundle. Moisture carryover - the percentage of steam mass flow that is entrained as liquid water. Recirculation Ratio - the ratio of liquid flow separated from the riser flow to steam flow (equal to circulation ratio minus one). Riser - the flow path through the steam generator tube bundle to the steam separator inlets. i Steam carryunder - the percentage of downcomer mass flow that is entrained steam, i l 1 l 4
BWI-222-7693-LR-01 Page 18 b LIST OF FIGURES PAGE i 2.1-1 BWI STEAM GENERATOR PHYSICAL COMPARISON . . . . . . . 28 4 2.2.1-1 WELDED IN DIVIDER PLATE . . . . . . . . . . . . . . . . . . . . . . . . . . 31 2.2.1-2 PRIMARY MANWAY CLOSURE DESIGN . . . . . . . . . . . . . . . . . 32 2.2.1-3 SECONDARY SIDE MANWAY DESIGN ' . . . . . . . . . . . . . . . . . . . 33 2.2.1-4 PRIMARY HEAD MANWAYS . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 . 2.2.1-5 SECONDARY SIDE MANWAYS, HAND HOLES, AND INSPECTION PORTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 1 2.2.3-1 JOINT GEOMETRY AT THE SECONDARY FACE OF THE TUB ESHEET . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 8 2.2.4-1 DETAILS OF BWI LATTICE GRID TUBE SUPPORT . . . . . . . . . 45 2.2.4-2 BWI LATTICE GRID ASSEMBLY . . . . . . . . . . . . . . . . . . . . . . . . 46 2.2.4-3 DIFFERENTIAL RESISTANCE LATTICE GRID . . . . . . . . . . . . . 47 2.2.4-4 FLAT BAR U-BEND RESTRAINT CONFIGURATION . . . . . . . . 48 2.2.4-5 FLO W PATTERN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49 2.2.4-6 U-BEND MOTION (EXAGGERATED) - COLD TO FULL POWER 50 l 2.2.4-7 ARCH B AR ASSEMBLY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 51 . 2.2.5-1 STFAM DRUM FEEDWATER DISTRIBUTION SYSTEM . . . . . . 58 3 2.2.5-2 GOOSE NECK DESIGN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 59 i 2.2.5-3 "J"-TUBE DES IGN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 60 ! 2.2.5-4 STRESS INTENSITY vs. RATE OF LEVEL INCREASE . . . . . . . . 61
~2.2.5-5 AUXILIARY FEEDWATER NOZZLE AND HEADER ASSEMBLY 62 l
2.2.6-1 TUBESHEET BLOWDOWN CONFIGURATION . . . . . . . . . . . . . 65
l BWI-222-7693-LR-01 Page 19
- b LIST OF FIGURES
- (continued)
PAGE ! 2.2.7-1 STEAM DRUM ARRANGEMENT . . . . . . . . . . . . . . . . . . . . . . . . 70 2.2.7-2 PRIMARY AND SECONDARY SEPARATOR . . . . . . . . . . . . . . . 71 2.2.7-3 TYPICAL OPERATIONAL CIRCULATION RATIO AND FLOW CONDITIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 72 l l 2.2.7-4 SEPARATOR PERFORMANCE FOR POWER SERIES TESTING . 73 l 2.2.7-5 SEPARATOR PERFORMANCE FOR PRESSURE . SENSITIVITY TESTING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 74 2.2.7-6 SEPARATOR PERFORMANCE FOR WATER FLOW SENSITIVITY TESTING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 75
)
i 2.2.7-7 SEPARATOR PERFORMANCE FOR WATER LEVEL ' SENSITIVITY TESTING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 76 2.2.10-1 RSG NOZZLE DAM DETAILS . . . . . . . . . . . . . . . . . . . . . . . . . . . 79 j
\
2.5-1 INDUSTRY-WIDE STEAM GENERATOR PROBLEMS . . . . . . . 122 i i j 2.5-2 STRESS vs. DETECTION TIME FOR SCC . . . . . . . . . . . . . . . . . 123 2.5-3 CONSTANT LOADING STRESS CORROSION CRACKING TEST RES ULTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 124 l 2.5-4 STEAM QUALITY CONTOURS . . . . . . . . . . . . . . . . . . . . . . . . . 125 4 2.5-5 GAP VELOCITY CONTOURS . . . . . . . . . . . . . . . . . . . . . . . . . . 116 2.5-6 TUBESHEET HOT LEG FLUID QUALITY PROFILE . . . . . . . . 127 2.5-7 TUBESHEET HOT LEG GAP VELOCITY PROFILE . . . . . . . . . 128 2.6-1 PERFORMANCE MARGIN FOR B&W SEPARATOR l EQUIPMENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 139
BWI-222-7693-LR-01 Page 20 b LIST OF TABLES l PAGE i l 2.1-1 STEAM GENERATOR COMPARISON . . . . . . . . . . . . . . . . . . . 26-27 l j 2.3-1 CHEMICAL ANALYSIS REQUIREMENTS FOR PRESSURE j BOUNDARY MATERIALS . . . . . . . . . . . . . . . . . . . . . . . . . 92 2.3-2 MECHANICAL PROPERTIES OF PRESSURE BOUNDARY MATERIALS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 93 j 1 2.3-3 CHEMICAL ANALYSIS REQUIREMENTS OF l CRITICAL-TO-FUNCTION MATERIALS . . . . . . . . . . . . . . 94 ; I 2.3-4 MECHANICAL PROPERTIES OF CRITICAL-TO-FUNCTION , MATERIALS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 95 2.3-5 REPLACEMENT STEAM GENERATOR MATERIALS . . . . . . 96-98 j i , 2.5-1 INDUSTRY STEAM GENERATOR PROBLEM AREAS AND RSG MEASURES TO ADDRESS THEM . . . . . . . . . . . . 118-121 j 4 1 2 a 4 4 4 1
! BWI-222-7693-LR-01 Page 21
- RSG TOPICAL REPORT EXECUTIVE
SUMMARY
4 The Replacement Steam Generator Topical Report describes the design and manufacture of the BWI Replacement Steam Generators (RSGs) for use in Duke Power Company's 4 McGuire 1 and 2 and Catawba 1 nuclear stations. This report discusses aspects of the RSG
- design that provide improved reliability from the existing design, addressing failure modes normally found in steam generators and describes features which eliminate or reduce the
, effects of the failure modes. In addition, it provides general information related to the
- RSGs describing design characteristics and discusses design criteria including the analysis
- executed to address the speci6ed requirements. .
! I The RSGs are manufactured by Babcock & Wilcox International (BWI) in Cambridge, l
- . Ontario, Canada. The RSGs are designed, manufactured and tested in accordance with the :
1986 Edition (no addenda) of Section III of the ASME Code, and will be N-stamped by i BWI prior to shipment. The design, procurement, and manufacturing process is performed l
- under a QA Program that complies with the requirements of Appendix B to 10CFR50, and i i
complies with current NRC requirements that relate to steam generator design. i l l RSGs including the Millstone Unit 2 (MP2) RSGs have been manufactured by BWI, f
- successfully certified by the Authorized Nuclear Agency and are performing satisfactorily.
i The steam generators described in this report employ similar design features and corrosion b g resistant materials as the MP2 design.
'Ihe success of the MP2 steam generator ,
l replacement confirms RSG design methods and expected ja situ operational performance. 1 ;
- The RSGs occupy the same physical envelope as the Original Steam Generator. Differences l
l between the OSG and RSG designs are identified in this report. There are no changes to ; i interfaces with the reactor coolant, main steam systems, or component or piping supports l with the exception of the RSG feedwater nozzle which has been relocated to accommodate ! ] replacing the integral preheater of the OSG with an internal feedwater header on the RSG - l and also relocation of the sample taps and the auxiliary feedwater nozzle for CNS. Normal , l operating conditions and plant transients have been requalified for the RSG design, thus, i associated design bases are not affected. l l e
- t 4
4 !
BWI-222-7693-LR-01 Page 22 RSG Topical Report
- 1. INTRODUCI' ION AND PURPOSE *
, This Topical Report describes the McGuire 1 and 2 and the Catawba 1 replacement J recirculating nuclear steam generators constructed by Babcock and Wilcox International ; (BWI) of Ontario, Canada. It describes the superiority of the BWI RSG design and manufacture with respect to generic steam generator failure modes and reliability.
Modifications that may be performed during the steam generator replacement outage or the replacement process itself are beyond the scope of this report. BWI has extensive nuclear steam generator design and fabrication experience, founded on , more than a hundred years of heavy-vessel manufacturing capability and experience for the fossil power and petroleum industries. The service record of BWI recirculating steam generators has been excellent. These issues are addressed in Section 2.1.1. The RSG design is described in Sections 2.2, 2.3, and 2.4 of this report. The RSGs incorporate many improvements. These are discussed in Section 2.5 of this report. , 4 Confirmatory analyses and tests, RSG operating restrictions, stress evaluations, and start-up j testing are discussed in Sections 2.6,2.7,2.8 and 2.9 respectively. BWI controls RSG design and fabrication to maintain high quality and to maintain the existing plant's design and licensing bases. The BWI RSG design and the quality assurance ! controls used in RSG construction conform to NRC requirements. The BWI quality plan ' is described in Section 3 of this report. l l i The principal objectives of this report are to
- 1. Describe BWI capability to design and build RSGs for pressurized water reactors,
- 2. Describe RSG design features, materials, methods of analysis, QA measures, l fabrication controls, and demonstrate physical, structural, and thermal-hydraulic compliance with the design requirements,
! 3. Identify the RSG design criteria and standards employed, including NRC guidance, and describe conformance to them, 4.' Describe industry steam generator problems and issues considered in RSG design and fabrication and discuss design features that provide improved reliability considering failure modes found in presently operating steam 8 generators. Section 2.1.2 d.iscusses the Millstone 2 RSG design. These RSGs were designed and fabricated by BWI (not the OSG manufacturer), installed under the provisions of
- 10CFR50.59 and 10CFR50.90, and approved by the NRC. The RSGs described in this
; report employ similar corrosion-resistant materials, and similar design features to the d l Millstone 2 steam generators.
4
BWI-222-7693-LR-01 Page 23 l l 1 t ( 2. REPLACEMENT STEAM GENERATOR DESIGN i i 2.1 . GENERAL DESCRIPTION ! l The McGuire and Catawba replacement steani generators (RSGs) described in this report ! i are of the recirculating non-preheater U-tube design. They have the following design i features: '
- 1. Stainless steel (410S) lattice grid tube supports.
i 2. Stainless steel (410S) flat bar U-bend supports. ,
- 3. High capacity primary and secondary cyclone separators.
- 4. Circulation ratio of 5.7. :
! 5. Feedwater headers which minimize potential water hammer and thermal 1
- stratification effects.
4 6. Minimum-radius tube U-bends of five times tube diameter or more.
- 7. Triangular tube pitch.
! 8. Thermally treated Inconel 690 tubes. Rese and other important aspects of the RSGs are described in the following sections of l
- this report. The BWI RSGs accommodate high internal circulation flows with acceptable i levels of tube vibration and effective steam separator performance. High internal circulation benefits steam generator performance and longevity by promoting flow 3 penetration across the tubesheet and reducing fluid quality and zones oflow velocity thereby
. reducing sludge accumulations. Through fabrication of steam generators for Canadian i heavy water reactor (CANDU) plants and for Millstone 2, and through performance of steam generator repairs and cleaning, BWI has demonstrated its capability to design,
- manufacture, and maintain steam generators with triangular pitch tube arrangement. ;
- 2.1.1 Qualifications of the Steam Generator Supplier l BABCOCK AND WILCOX INTERNATIONAL i
; Babcock & Wilcox International (BWI), located in Cambridge, Ontario, Canada, has , ;
i fabricated fossil-fueled boiler components for over 100 years and has fabricated nuclear ! system components since the late 1950's. Although most of the nuclear system components manufactured have been recirculating steam generators for CANDU nuclear plants, the ! ! RSGs are comparable in materials, water chemistry, and fabrication methods. As shown r in Figure 2.1-1 the size of these units is also comparable. Therefore, BWI's experience in
, supplying over 200 CANDU steam generators is directly applicable to the RSGs described i in this report. ; i ! . BWI has strong Project Management, Engineering, Manufacturing, Production Control, +
Purchasing, and Quality Assurance Departments. Dese provide close control of the quality
- of replacement steam generator design, procurement, fabrication, and documentation.
; Continuous work in the nuclear industry has enabled BWI to maintain a well qualified ;
steam generator design group. Engineers involved with the design and analysis of steam
BWI-222-7693-LR Page 24 generators have a thorough knowledge of design by analysis methods and are familiar with ! the application of the ASME code to nuclear pressure vessel design and analysis. BWI i holds ASME certificates of authorization for N, NA and NPT symbol stamps. Subcontractors for material supply and fabrication are all fully qualified under the requirements of the BWI Quality Assurance Program. He BWI quality assurance program is described in Section 3.1. BWI steam generator manufacturing experience to date includes: CANDU Steam Generators: { Lattice Grid Type 82 Wolsong 3 and 4 South Korea (under construction) 2
- Broached Plate Type J2ji i Total 209 PWR Steam Generators: ,
Lattice Grid Type Northeast Utilities (Millstone 2) 2
- 12 Duke Power Co. (under construction) i Florida Power & Light (under construction) 2 i Rochester Gas & Electric (under construction) 2 Commonwealth Edison (under contract) .i 1
- Total 22 l BWI recirculating steam generators have more than 20 years of operating history. The performance and reliability of BWI steam generators has been excellent. In over 200 steam generators, containing more than 600,000 tubes and having in excess of 6 million tube-years j 'of operation, less than one percent of the tubes had been plugged as of July,1993.
! Additional information on steam generator tube opening experience and BWI measures to preclude primary water stress corrosion cracking, intergranular attack and sludge ] accumulation are contained in Section 2.5.1. 4 4 2.1.2 Millstone 2 Replacement Steam Generator Design and Experience l Two BWI RSGs are in service at Millstone 2. The steam generators described in this report have many features in common with the Millstone 2 design. These include Alloy 690 tubes and other corrosion resistant materials, weld overlay of all primary side carbon steel surfaces with stainless steel or Inconel, tight packing of tubes, full depth hydraulic expansion of,the
- tubes in the tube sheet, and measures to minimize water hammer and vibration. Unlike the i Millstone 2 steam generators, the RSGs are complete replacements, shipped intact to the :
2 plant. The Millstone 2 heat exchanger (lower) sections were shipped to the site for use with i
', BWI-222-7693-LR-01 Page 25 a 4 the existing steam drums (upper sections). 4 Successful completion, licensing, installation and startup of the Millstone 2 steam generators demonstrates BWI design and fabrication capability, and the overall acceptability of the
- RSG design.
2.1.3 Comparison with Existing Design l
)
Parameter changes from the existing (OSG) design are provided in Table 2.1-1 and include j differences in steam generator weight, inventory, operating conditions, and major ! geometrical features. The differences potentially affecting plant safety (water inventories, primary side flow resistance, shell stiffnesses and RSG weight) are beyond the scope of this i i report. i The RSG is designed, fabricated and analyzed to minimize differences with respect to form, fit, and function as compared to the existing steam generator. Physical comparison of the
, RSG and OSG are discussed in this report. Compatibility of primary and secondary side
, materials with the existing design is generally demonstrated in Section 2.3. i 4 4 W 4
)
i N u t a l' . i 3 !
BWI-222-7693-LR-01 Page 26 l . TABLE 2.1-1 STEAM GENERATOR COMPARISON PARAMETER RSG DATA OSG DATA Primary side volume: no tubes plugged (nozzle 1229.1 955 dams in place) (ft') 8 Secondary side mass: 9 0 % full power (lbm) 144.5 x 10 116 x 10* O full power (Ibm) 8 124.6 x 10 104 x 10* Full power steam flow 3.78 x 10' lbm/hr 3.78 x 10' lbm/hr Primary Pressure drop across 5/G (unplugged) 9 33.0 psid 33.3 psid 37.0E6 lbm/hr Primary side design pressure psia 2500 2500 Secondary side design pressure psia 1200 1200 Primary side design temperature (* F) 650 650 Secondary side design tanperature (* F) 600 600 Primary side operating pressure (psta) 2250 2250 Steam outlet conditions: pressure (psia) 1020 1020 maximum carryover 0.25% 0.25% (Guarantee) l Feedwater temperature 9 full power (* F) 440 440 I Heat transfer rate 9 full power (W) 857.5 857.5 Steam Outlet Flow restrictor flow area (ft') 1.374 1.39 Primary side heat transfer surface area: 70,460 42,500 no tubes plugged (based on avg. I.0.) (ft') Secondary side heat transfer surface area: 79,800 48,000 no tubes plugged (based on avg. 0.0.) (ft8 ) Number of tubes 6633 4674 Tube 0.0.. 0.6875" nom. 0.750" nom. 2 upper tolerance +0.0" lower tolerance -0.005" Tube wall thickness: nominal 0.040" 0.043" tolerance 2 0.004" - Tube material: 58-163, Code Case N-20 Alloy 690 Alloy 600 Tube thermal conductivity: 9 400*F 8.92 Btu /f t-hr
- F 10.1 Btu /f t-hr
- F 9 500* F 9.54 Btu /ft-hr ' F 10.6 Btu /f t-hr
- F 0 600'F 10.167 Btu /f t-hr
- F 11.1 Btu /ft-hr *F G
e
BWI-222-7693-LR-01 Page 27 I ' t TABLE 2.1-1 (cont'd)
- STEAM GENERATOR COMPARISON PARAIETER R$G DATA OSG DATA
, Tube pitch trianguter square l Tde minisun strength (per ASME Code and code case 40 ksi 35 kol
- N20): yletd tensite 80 kal 80 ksi =
l Steam outiet Nottle Diameter (in) 29.469 29
- Shett Side Manueys (No. - Die.(in)) 1 21 2 16 i
Primary Side Manusys (No. - Die.(in)) 2 21 2 16 , l
- Primary Intet Norstes (No. - Die.(In)) 1 31 1 31 !
Primary Outlet Nozzles (No. - Die.(in)) 1 31 1 31 .i j Feethseter Netzte Diameter (Nom.) (in) 16 16 ' Auxillery Feeduster Notate I.D. (in) 5.25 5.3 Bottom stoudoom Norstes (No.-Die.(in)) 23 22 4 1-Recirculation Mozate (No. Die.(in)) 13 N/A idater Level Taps (No. - Die. (In.)) 14 3/4 8-3/4 Nanctiot es (No. Die.) ((n) 10-6 26 1 3_ Inspection Ports (No. Dle.) (in) 12 2 42 I T 4 .5 1 i ( i i J ,k .
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h ( h (~ 63' 63* 63' i J _ !, D _k_ n BWICANDU ST..LUCIE 1 MILLSTONE 2 Romania BWI Replacement BWI Replacement
- 240 tons ca. = 527 tons ea' = 535 tons ca.
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M' r .__ L __ " ._ _; t R. E. GINNA MCGUIRE 1 & 2 & BWI CANDU BWI Replacement CATAWBA 1 Darlington
= 315 tons ca. BWI Replacement - 385 tons ca. I
= 406 tons ea. I BWI STEAM GENERATOR PHYSICAL COMPARISON -
FIGURE 2.1-1 l l
- BWI-222-7693-LR-01 Page 29 1 2.2 STEAM GENERATOR DESIGN HIGHLIGHTS 2.2.1 Pressure Boundany Design The reactor coolant pressure boundary and secondary side pressure boundary are critical to the safe and reliable operation of the RSG. This section describes the key design features of the RSG portions of these pressure boundaries except for the steam generator tubes which are described in Section 2.2.2. 'Ihe pressure boundaries withstand internal pressure, seismic, loss of coolant accident (LOCA), main steam line break (MSLB) loads and
] feedwater break loads. In addition, cyclic loading during normal operation creates the 4
potential for fatigue failures. The pressure boundary components are designed and
- documented to be in accordance with ASME Code requirements for Nuclear Pressure Vessels, Section III Division 1. Applicable codes and standards are described in Section l l
2.4.1. Pressure boundary materials are discussed in Section 2.3.1. l , l Pressure boundary design is analyzed by employing work-station based finite element ) i software. Finite element analysis is used as an analytical tool and a design tool. This . 3 i permits optimization ofimportant pressure vessel design features while minimizing stresses. , Critical design features and dimensions can be reviewed early in the design, accounting for i time dependent loads such as operational thermal transients. Two types of corrosion allowance are considered for design. Corrosion allowances for surfaces that are chemically cleaned include allowances for normal operation and for chemical cleaning. Corrosion allowances for surfaces that are not chemically cleaned include allowances for normal operation only. Allowances vary from zero to 0.0625 inches depending on material and application. Analyses for structural loads, pressure, flow, and flow-induced vibration were performed with corrosion allowances deducted. The corrosion values are verified as part of the BWI Chemical Cleaning Qualification Program. Key elements of this program are presented in Section 2.6.5. Preparation, revision and issue of design calculations and reports are governed by the BWI Quality Assurance Manual (described in Section 3.1). This e:sures that all design and analysis requirements are reviewed for adequacy and approved for release by authorized i personnel. 2.2.1.1 Tubesheet Assembly and Primary Divider Plate The tubesheet/ primary head assembly and primary divider plate arrangement is shown in l Figure 2.2.1-1. The divider plate is machined from Alloy 690 and welded around its entire d periphery to the tubesheet and primary head. At the tubesheet, the plate is welded to a machined Alloy 600 weld build-up along the tube free lane. Along the head, an Alloy 690
- . weld attaches the divider plate directly to the head base metal rather than to the stainless steel cladding. The stiffening effect of the divider plate is not taken into account when sizing the tubesheet thickness. '
l 1 l l
. - = - - -_ _- . ._ .
I BWI-222-7693-LR-01 Page 30 l I l
- 2.2.1.2 Closure Design i
i The RSG is fitted with removable closures at manways, hand holes and inspection ports located to provide access for inspection, repair and maintenance of steam generator i internals. Figures 2.2.1-2 and 2.2.1-3 show external and internal manway closure designs.
- The external cover design provides metal-to-metal contact with the gasket properly seated.
His is achieved by controlling the depth of the gasket groove in the inner diaphragm plate. 3 The metal-to-metal contact and use of long Dexible bolts reduces the' fatigue loading on the bolts during operation. The longer bolts also reduce bolt stress caused by pressure and thermal distortion of the opening. This design can be readily adapted to various stud tensioning systems. 4 Figures 2.2.1-4 and 2.2.1-5 show locations of primary side manways and secondary side , manways, hand holes and inspection ports. These provide access for inspection, maintenance and repair. ! 2.2.1.3 Shell and Nozzle Design
- The RSG shell is fabricated from forgings and plates. Forgings are used for the steam drum head including integrally forged steam outlet nozzle, the primary head including integrally
] forged primary nozzles and manways, the tubesheet and the conical transition section. Plate
- is used for the balance of the shell. By maximizing the use of forgings, the RSG design reduces the quantity of weld material requiring in-service inspection and the complexity of in-service inspection. The RSG is supported on support pads which are integrally forged into the primary head (channel head). The lower head and lower tube bundle shell sections are welded to the tubesheet forging.
The RSG primary and secondary side nozzles are the same sizes as those of the OSG. Primary and secondary side manways, however are twenty-one inches in diameter, i considerably larger than those on the OSG. This allows easier access to the channel heads l and secondary side. He RSG primary nozzles are integrally forged into the primary head. Safe ends are welded to the nozzles to accommodate RSG fit-up to the existing plant piping. The primary manways are integrally forged into the primary head. Stress concentrations are reduced by contouring all discontinuities and providing large blend radii in these areas. { A similar design is used for both the primary and secondary side manways. The upper head is a single forging which includes the main steam nozzle and an integral flow restrictor that limits internal RSG Duid velocities in the event of a main steam line break. The Dow restrictor design and function are discussed further in Section 2.2.9. I 1 l e i 4
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SECONDARY SIDE MANWAYS, HANDHOLES AND INSPECTION PORTS F]GURE 2.2.1-S
BWI-222-7693-LR-01 Page 36 2.2.2 Steam Generator Tube Design The RSG tubes are fabricated from thermally treated Alloy 690. This alloy ha:: better ! overall corrosion resistance than Alloys 600 or 800 in nuclear steam generator environments. ! Details of the Alloy 690 composition, heat treatment, and mechanical properties are provided in Section 23.2. Qualification of the tube-to tubesheet joining processes are l discussed in Section 2.63. Pressure stress limits and tube plugging criteria are discussed m - Section 2.8.1. The original and replacement tube bundle designs are geometrically compared. The [ comparison includes tube outside diameter, tube wall thickness, tube material, average , bundle surface area, number of tubes and tube thermal conductivity. Differences exist for the following reasons.
- 1. RSG bundle surface area is larger primarily due to the replacement of the :
integral preheater OSG with a non-preheater RSG. i
- 2. RSG tubes have lower thermal conductivity due to the change in tubing ;
material from Alloy 600 to Alloy 690.
- 3. To facilitate more tubes and corresponding larger surface area the tube O.D. j has decreased allowing a thinner wall.
i 2.23 Tube-to-Tubesheet Joint The RSG tubes are flush welded to the primary face of the tube sheet and hydraulically , expanded to maximize mechanical strength and to seal the tube to tubesheet crevice. This precludes crevice or stress corrosion in the tubesheet area. The tubes are installed into the [ tubesheet after the RSG lower shell and primary head assembly have been welded and received Post Weld Heat Treatment (PWHT). This precludes tube sensitization concerns. !
- The tubes are seal welded to the tubesheet and hydraulically expanded within the full l thickness of the tubesheet. Seal welding and expansion of the tubes after PWHT avoids l subjecting the tube to tubesheet joint to thermal stresses from these operations and ;
j eliminates concern over loosening of tubes or creation of crevices as a result of relaxation ! l of the expanded region. ; i r l The tube to tubesheet joint geometry at the secondary face of the tubesheet is shown in !
- Figure 2.23-1. The following paragraphs provide further information of the tube to
{ i tubesheet joint. Qualifications of the expansion processes are described in Section 2.63. l 2.23.1 Tube-to-Tubesheet Welding l l De flush tube-to-tubesheet weld has been applied successfully to twelve steam generators l for three 600 MWe power stations (Gentilly and Point Lepreau in Canada and Embalse, l Cordoba in Argentina). The generators have been in service since 1983 with no tube joint l problems reported. He smooth weld profile has a crown approximately 0.025 in. high and ; i : i . l l L _i
._ ._ _ _ - ~. _ _ _ _ _ _ _ _ _ _- _ _ _
l i BWI-222-7693 LR-01 Page 37 4 I l negligible tube diameter reduction. If necessary, the tube ends are sized by rolling to the minimum expansion diameter to allow subsequent use of tube repair or inspection ' l equipment. The tube-to-tubesheet weld is designed, analyzed, performed and examined in : l accordance with ASME Section III critena. > l 2.2.3.2 Hydraulic Expansion RSG tubes are hydraulically expanded through essentially the entire thickness of the ' tubesheet. The length of the expansion mandrel is determined by the thickness of the tubesheet with hydraulic seals positioned on the mandrel to control the length of tube expanded. The hydraulic seals are of clastomeric material and designed so that no metal ; parts are impressed upon the inside surface of the tube when the hydraulic pressure is applied. The position of the seal at the secondary face of the tubesheet is controlled to ensure that expansion of the tube is as close as possible to the secondary face of the . tubesheet without going past the face. This is detailed in Figure 2.2.3-1.
- For peripheral tubes, where access is limited by curvature of the primary head, expansion i is performed in two overlapping zones, using a shorter expansion mandrel The shorter l mandrel can access the peripheral tubes without interfering with the primary head. The
- expansion zones overlap near the center of the tubesheet to ensure full depth expansion.
, i 3 To ensure that all tube-to-tubesheet joining operations can be satisfactorily performed, a ,
- ten-tube sample is constructed. It simulates the full tubesheet thickness and uses materials I identical to those used in the steam generator. All processes, procedures and inspections !
i approved for use in manufacturing the tube-to-tubesheet joint are performed. Prior to RSG ' l fabrication, the sample is examined by sectioning to verify that manufacturing operations ;
- were correctly performed and results are satisfactory.
Tests on hydraulically expandedjoints made with Alloy 690 tubes, in closely fitted holes (the BWI practice) have shown that residual stresses exist in the transition region between the expanded and unexpanded tube. The hydraulic expansion process has been designed and quali6ed to minimize residual stresses while maintainingjoint integrity. Qualification of the hydraulic expansion processes is discussed in Section 2.6.3. BWI has successfully l hydraulically expanded approximately 334,000 tubes in thirty-eight steam generators. There j has been no case where a tube required plugging due to an expansion non-conformance. j After expansion the inside profile of each tube is measured through the entire expanded
- area of the tubesheet (including the transition) using an eddy current method and recorded.
The measurements indicate both the position and condition of the tube expansion, and
; become a baseline for subsequent inservice inspection. Section 3.2.6.6 describes the baseline
{ inspection. Test results are documented and supplied to the owner.
- 2.2.3.3 Inservice Inspection 4
The RSG desig'n provides the capability to perform inservice inspections in accordance with the requirements set forth in ASME Section XI. i i
BWI-222-7693-LR 01 P gs38 k* MIN.TOh' MAX. _ FINAL HYDRAULIC EXPANSION ZONE TO 11' PRIMARY OVER-LAP TACK EXPANSION SIDE CONDARY / SEAL WELD
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BWI 222-7693-LR-01 Page 39 \ 2.2.4 Tube Bundle Support System I i This section describes the RSG lattice grid and U-bend supports, and measures to minimize ; i flow induced vibration (FIV), dry out potential, flow resistance, wear susceptibility and discusses the structural integrity of the tube support systems. b :
- 2.2.4.1 Lattice Grid Tube Supports i !
- The RSG design uses a Type 410S stainless steel lattice grid tube support. BWI has l
l! experience with lattice grid and broached plate tube support designs. From this experience ! BWI concludes that the lattice grid is superior for a recirculating steam generator. The .
- lattice grid provides
- !
j e High circulation rates (through lower flow resistance). e Superior strength (capable of sustaining very high seismic loads, does not I requin: tie rods). l o Superior vibration restraint and fretting resistance. b
- e Lower tendency to accumulate deposits than a broache'd plate (line contact 2
with the tube rather than " area" contact provided by a broached plate).
- Reduced denting potential due to selection of stainless steel.
J I Examples of the success of this design include Pickering A, with lattice grids,20 years l 3 operation, and tubesheet sludge accumulation (the plant ran 3 years on phosphates) but no lattice grid deposit buildup (determined by visual inspection) and no under-deposit tube ; failures. ' i Figures 2.2.4-1 and 2.2.4-2 show the details of a lattice grid. The lattice grid is made up of
- ' a series of high bars (approximately 3 inches in width) oriented 30' and 150* to the tube i i free lane and located every sixth pitch, to accommodate the steam generator loading !
conditions. Low bars (approximately 1' inch width) are located at every pitch location between the high bars. All low bars flush to the top of the high bars are oriented at 30' ;
- to the tube free lane and all low bars flush to the bottom plane of the high bars are l oriented 150' to the tube free lane. The bar ends are fitted into precise slots of a specially '
designed peripheral support ring, which is then sandwiched by two outer retainer rings held l together by studs and lock welded acorn nuts. To further enhance stability of the grid, tube b free lane support beams and span-breaker bars are secured on the upper and lower surfaces of the grid. Lattice supports are positioned within the steam generator shroud at elevations selected to prevent flow induced vibration while not creating excessive flow resistance. The tubes'are b held in position within the diamond-shaped bar opening which provide line support contact. This minimizes the area of" crevices" between the tubes and bars which could trap corrosion
- BWI-222-7693-LR-01 Page 40 products and eliminates any stagnant spots responsible for " dry-out" caused by local
, superheat. All of the lattice supports are identical except that the lowermost lattice incorporates a differential resistance lattice grid (Figure 2.2.4-3) which resembles that of a regular grid. However, the low bars located toward the bundle periphery are replaced by medium bars
- (approximately 2%" in width). As a result, the flow passages through these regions offers
- more resistance to flow and the fluid is preferentially directed to penetrate into the central n
i region of the tube bundle. A drilled flow distribution baffle is not used. Since a distribution plate is simply a drilled plate with slightly oversized holes, it may accumulate l deposits and possibly become plugged. ; Tests conducted by BWI have shown that the in-plane strength of lattice grid supports is , ] higher than that of broached plate supports. This is important to seismic, shipping and handling requirements. Extensive laboratory testing and computer modelling have con 6rmed that the out-of-plane load handling capability of the BWI lattice grid is superior to the broached plate design. The tube bundle is analyzed to determine tube vibration characteristics and the effectiveness
- of lattice grids in suppressing vibration. The results show that lattice grids are the best ;
support system for damping tube vibration and minimizing tube wear due to fretting. Flow- ! induced vibration modelling is discussed in Section 2.6.1. BWI has refined the grid i manufacturing processes to allow very close tube-to-grid clearances so that tube vibration i and wear potential are further reduced.
- 2.2.4.2 U-Bend Supports !
l Like the lattice grid tube supports, the Type 410S stainless steel Flat bar U-bend Restraint (FUR) system provides effective, close clearance supports of the upper regions of the RSG i i tubes to prevent flow-induced vibration. 'Ihe potential for fretting is reduced by 1 compatibility with the tube material and longer contact length than is provided by an AVB-
- support system. All tubes are supported by FURS. The FURS provide open flow paths
)
and line contact support at all locations in the bundle, reducing the potential for sludge
- build-up.
4 l 2.2.4.2.1 Design Configuration j The FURS incorporate a series of flat bar fan assemblies on each side of the bundle,
- positioned between each layer of tubes as shown in Figure 2.2.4-4. The fans are positioned i
so that all U-bends are supported at close intervals. Fan assemblies consist of up to four fan finger bars at diagonal positions, connected at their lower ends to a nearly horizontal bar by full-penetration, heat treated welds. The nearly horizontal bar provides support for the smallest radius U bends. The wide FURS distribute contact force to minimize the j possibility of fretting. All U-bends are supported by the flat bars. The innermost tubes are installed with their U-l j 1
\
i BWI-222-7693-LR-01 Page 41 bends in a plane that is skewed with respect to the channel head divider and tube-free lane
- (TFL). This permits larger radius bends than if these tubes were installed perpendicular l . to the TFL. The FURS do not pass through this part of the tube bundle because of the skewed tube plane. The small-radius tubes are supported below their bend tangent points
] .
by the inner ends of the nearly-horizontal bars. J The FURS are made of Type 410S precision cold rolled steel that provides high resistance i to fretting wear, excellent strength and high resistance to corrosion-related tube denting. i Further information on selection of support system materials appears in Section 2.3.2. [ 2.2.4.2.2 Flow Characteristics l He FURS are designed with all spaces oriented with an upward slope. This promotes i continuous sweeping during operation. FURS do no: cross the bundle centerline. This
- avoids creation of spaces where deposits might collect.
1 Tubes are supported by line contact and bars are offset within each row to provide more
~ flow area than would exist with in-line bar placement. Bar array position generally follows
- the pattern of unobstructed U-bend flow. A flow diagram of an unobstructed U-bend and
! FUR array is shown in Figure 2.2.4-5. Cross flow is low and the FURS do not significantly . 4 impede or disrupt flow.
- 2.2.4.2.3 Flexibility and Thermal Motions Free expansion of the U-bend during operation is essentialin order to avoid tube stress and 4 potential tube damage. He FUR system allows free expansion of the U-bend tubes without sliding between the bars. He FUR assembly is supported by, and moves with the outermost layer of tubes rather than being anchored to the upper lattice support. The FUR
, and tube bundle move up and down together during heatup and cooldown. During power operation the tube hot- and cold legs have slightly unequal leg temperatures that create a
- slight angularity, shown (exaggerated).in Figure 2.2.4-6. Analyses show that for a U-bend
, assembly under the worst case conditions, tube-to-support angularity is easily accommodated i by the lattice supports without risk of tube damage, tube lockup, or loss of tube fixity at the d top support. ! 2.2.4.2.4 Support cf FUR Assembly l 4 The weight of the fan assemblies is supported by arch bar assemblies which transfer the
- weight to the outermost layer of tubes via the J tabs (see Figure 2.2.4-7). The FUR fan A finger bars (a) are notched at their upper ends. These bars are collected by a slotted
. clamping bar (b) which is attached by welded pins to the arch bar (c). The arch bars thereby collect all the weight of the fan assemblies. He weight of the fan assemblies is
' transferred to the outermost layer of the tubes by "J" tabs (d) installed after completion of
! the U bend assembly and positioned to uniformly contact the completed tube assembly. i Tube stress resulting from this weight is small. This is con 6rmed by a tube bundle / FUR interaction analysis. . l l
BWI-222-7693-LR-01 Page 42 l The arch bar/ fan finger assemblies are prevented from splaying apart under dead weight loads during operation by tie tubes that maintain arch bar spacing (Figure 2.2.4-4). I The arrangement described above accommodates all operating loads and motions. l Assembly, handling and shipping loads are supported by temporary restraints that are ! removed at the site by construction personnel after RSG installation, and prior to operating d
- the RSG Seismic tube bundle loads are supported by the FURS and lattice support. As
- there is no connection between the FUR assembly and the shroud, U-bend deflections
- during earthquake will not damage the tubing. The flat bars do not absorb the full seismic
- load, but moderate the deflection of the tubes relative to each other. Main steam line break j loads are insufficient to lift the FUR assembly. -
l 2.2.4.3 Design to Minimize Flow Induced Vibration (FIV)
- i Prevention of excessive FlV and fretting wear is achieved by a combination of design, 1
- analysis and testing. The FURS are arranged to meet the design limits established for Fluid j
- Elastic Instability (FEI) and for response to turbulence. These analysis methods and criteria ;
j are discussed further in Section 2.6.1. ; 4 2.2.4.3.1 Clearances Small U-bend support clearances are maintained while avoiding tube /bar interference ) problems (marking of the tubes by the bars, splaying of the bundle due to bar tolerance ! accumulation, or buildup of bundle thickness) as tubing progresses. The optimum range of , flat bar U-bend support clearance was verified by an air flow test (Reference 1). This test
- compared the effectiveness of flat bar U-bend supports to scalloped bar (360' drilled hole)
! supports. The test showed that flat bars with small clearances provided more effective !
- support than the scalloped bar design with larger clearances. Tests with larger clearances showed significant response in all directions, including in-plane (the " weak" direction), for either flat bar or scalloped bar supports. The flat bars more effectively suppressed
; instability and in-plane turbulence response.
? I 2.2.4.3.2 Bar Width i Fan finger bar width is sized to provide line contact that minimizes the potential for fretting. j Comparative autoclave fretting tests have shown that the wear rate is substantially reduced 1 as bar width increases (Figure 7 of Reference 2). ! l 2.2.4.3.3 Fretting Assessment Potential fretting is assessed by performing a FIV sensitivity analysis. FIV methods are I i discussed in Section 2.6.1. The FUR design is qualitatively compared to other designs by ,
; comparing the relevant U-bend support parameters (material selection, bar widths, support l clearance and span lengths). Design assurance is achieved by conservatively meeting the FIV analysis parameters for FEI and Random Turbulences Excitation (RTE), and then by assuring that support effectiveness, materials and clearances are optimum.
^
I
BWI-222-7693 LR-01 Page 43 2 1 Fretting is a major design consideration. The relevant parameters are: 1) U-bend flow
; loading,2) support positions,3) support material,4) support clearance, and 5) support i contact length. These parameters are considered in the design of the FUR which is shown schematically in Figure 2.2.4-4 and are addressed below:
Flow loading is determined for a given steam output by the circuhtion rate and the U-bend tube and support geometry. Having established the desired circulation rate, the velocity and quality distributions are determined by a 3 dimensional thermal hydraulic analysis code (See Section 2.6.2). De optimum geometry is one in which there is least interference with the free release of riser flow. The RSG design achieves this with its open flow configuration and bar orientation which is generally compatible with the flow direction. l j Optimization of the position of U-bend supports is based on the FIV analysis for FEI and l j for turbulence response (Section 2.6.1). The result is a design with short tube spans, high ! natural frequencies, small response to turbulence and large margin for instability. 4 Selection of Type 410S as the support material in combination with Alloy 690 tubes provides ) i a high degree of fretting resistance. The Alloy 690/410S combination has the lowest wear i rate of any of the available combinations. Fretting wear test results from AECL indicate A that the fretting wear rate for the Alloy 690/410S combination is essentially the same as that i for Alloy 800/410S and slightly better than Alloy 600/410S (Reference 3).- g The mean diametral U-bend support clearance has been set at a very low value. This clearance was selected based on comparative U-bend air flow testing (Reference 1) which indicated that flat bar U-bend supports with small clearances (0.003" to 0.010") provided
- good " pinned" support conditions and that the effectiveness of such a support was better than that of a scalloped bar support with a 0.020" clearance (even though the scalloped bar
! provided a 360' drilled support con 6guration). This mean clearance provides a snug overall j l design while still permitting thermal motions. l The RSG design provides substantial contact length compared to about 0.40" in other designs. This contact length reduces the contact stresses which result from ongoing i turbulent excitation. Comparative autoclave tests have shown that the wear rate is i substantially reduced with a greater bar width (Reference 2).
- The parameters noted above are the same or better than those used for the Millstone 2 RSG which is operating successfully.
l f 4 ]
t BWI-222-7693-LR-01 Pags 44 References for Section 2.2.4
- 1. "The Effects of Flat Bar Supports on the Crossflow Induced Response of Heat Exchanger U-Tubes", D. S. Weaver, W. Schneider, Journal of Engineering for Power, October,1983.
- 2. Third Keswick International Conference of Vibration in Nuclear Power Plants, England, May 1982, " Heat Exchanger Tube Fretting Wear: Review of Application to Design", P. L Ko, PhD.
- 3. AECL Research, Report RC-1314, "PWR Replacement Steam Generator Fretting-Wear", A. B. Chow and D. A. Grandison, November 1994. d I
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BWI-222-7693-LR-01 Page 52 - 2.2.5 Internal Feedwater System i This section describes the RSG feedwater distribution system and design measures to preclude damage from water hammer, thermal stratification, erosion and internal feedwater header collapse. Water hammer has affected more than half of the operational PWR plants i in the U.S. Thermal stratification has caused fatigue cracks on the inner surfaces of thermal sleeves, feedwater nozzles and feedwater piping. High flow velocities and abrupt changes in flow direction have caused erosion where the feedwater flow splits to enter the header rings, and in the feedwater discharge tubes located along the header. Feedwater headers , have collapsed due to external pressure. The RSG feedwater distribution system design recognizes these potential problems and includes features to address them. The RSG feedwater distribution system (shown in Figures 2.2.51,2.2.5-2 and 2.2.5-3) is a split ring design connected via a T-section to a " goose neck" assembly attached to the thermal sleeve in the feedwater piping. The feedwater header is supported by the thermal sleeve /feedwater piping weld interface, and by supporting lugs around the ring i circumference attached to the internal shroud. Support lugs are located ou ti4e header pipe at approximately 90' to the feedwater nozzle, and at the header ends (near the split in the header ring) opposite the feedwater nozzle. The header lugs are vertical plate structures ; which have elongated holes through which the header passes. This. provides support to restrain motion in the vertical direction, while allowing thermal growth in the horizontal plane. Lateral stability of the feedwater ring is accomplished by restricting motion at the split location in the direction perpendicular to the feedwater nozzle leg. This design provides a support system that accommodates thermal motions and potential loads due to water hammer or system pump pressure pulses as well as all operational, seismic and burst g
- pipe loads, i
i The RSG feedwater distribution system satisfies all current NRC recommendations with respect to water hammer, provides flow stratification mitigation and addresses industry concerns regarding corrosion, corrosion cracking, thermal fatigue and material erosion. ! 'A2.5.1 The Water Hammer Mechanism l Water hammer in steam generators resul's t from rapid condensation and collapse of steam ! pockets in the feedwater system. These can cause potentially damaging pressure pulses in j feedwater piping. Water hammer can occur under various combinations of operating i conditions and piping geometry. Most steam generator water hammer events have involved j feedwater headers that discharge downward from a header. During certain plant transients, j steam generator water level dropped below the feedwater header, and allowed the header
- to become partially filled with steam. Increased feedwater flow condensed the steam at the l steam-water interface. This caused a counter-flow of steam above the level of the
! feedwater. Turbulence trapped steam pockets which were condensed. Slugs of water driven i by pressure accelerated upstream to fill the void. This sequence is termed " steam-water slugging", and is generally accepted as the initiating mechanism of feedwater header water hammer events.
i BWI-222-7693-LR-01 Page 53 i 1 A total of 27 water hammer events were reported between 1969 and 1982 (Reference 1).
, In 1977, all recirculating steam generator PWR licensees were requested to submit hardware '
and procedural proposals to reduce steam generatorwater hammer susceptibility. The NRC
! declared water hammer to be an Unresolved Safety Issue in 1978. Design change l recommendations were made and implemented, and on-site testing of the new designs was documented for 38 operating plants. In 1982, the NRC considered the top-feeding steam i
generator water hammer issue to be resolved (Reference 2). Design recommendations addressing steam generator water hammer are described in NRC Branch Technical Position ASB 10-2 (Reference 3). These are addressed in Section 2.2.5.2. The following section discusses these recommendations relative to the RSG design and describes features of the RSG that reduce or eliminate the potential for water hammer. l 2.2.5.2 Design to Preclude Water Hammer NRC Branch Technical Position ASB-10-2, " Design Guidelines for Avoiding Water Hammers in Steam Generators", (Reference 3), for reviews of top-feed steam generator designs identifies four items which serve "to reduce or eliminate the potential for water ) i hammer in the feedwater system: l a) Prevent or delay water draining from the feed ring following a drop in steam l generator water level by means such as top discharge J-Tubes and limiting ! ] feed ring seal assembly leakage. l I b) Minimize the volume of feedwater piping external to the steam generator
- which could pocket steam using the shortest possible (less than seven feet) 4 horizontal run of inlet piping to the steam generator feed ring. i c) Perform tests acceptable to NRC to verify that unacceptable feedwater hammer will not occur using the plant operating procedures for normal and
- emergency restoration of steam generator water level following loss of normal
- feedwater and possible draining of the feed ring. Provide the procedures for these tests for approval before conducting the tests and submit the results l from such tests.
i d) Implement pipe refill flow limits where practical." ! Items (a) and (b) address steam generator and piping design, while items (c) and (d) address operating and test procedures. Items (c) and (d) have been resolved (see Reference ; i 2). The RSG feedwater header design incorporatesJ-tubes connected on top of the header to help prevent header draining and formation of steam pockets thus addressing item (a). g '
- Item (b) is addressed by minimizing the straight length within the feedwater nozzle prior to the gooseneck.
l Figure 2.2.5-1 shows the steam drum internals and the feedwater piping. Operating water levels and the primary separator deck locations are also shown. Design improvements include lowering the primary separator deck (below the normal water level) and the l l
BWI-222-7693-LR-01 Page 54 f i feedwater sparger ring, a " goose neck" inlet design (detail shown in Figure 2.2.5-2), a : schedule 80 header and schedule 160 J-tubes (detail shown in Figure 2.2.5-3), t i To show that the RSG design reduces the potential for water hammer damage, both the frequency and consequences of water hammer events must be considered. For water hammer to occur, there must be steam in the feedwater piping. This can occur if the steam j drum water level falls below a point of discharge or if a leak exists in the internal feedwater j piping system. The potential for these conditions is minimized by: i . 1. Reducing the chance of uncovering the feedwater header by: i
- a. Positioning the header as low in the steam drum as possible. .
- b. Providing a design that maximizes steam drum water inventory above
, the header.
- c. Avoidance of the transients that uncover the header.
- 2. Minimizing drainage of the header once it is uncovered by:
- a. Utilizing top discharge header with J-tubes. .
- b. Maintaining feedwater flow to keep the header full.
- c. Eliminating leakage throughout the header assembly (except at the J-
- tube discharge).
The BWI feedwater header design incorporates J-tubes, internals with maximum secondary } side water inventory between the header and the normal water level, and an all-welded , thermal sleeve / header assembly from the thermal sleeve /feedwater pipe interface to the J-
- tube exit. This eliminates the possibility of steam leakage into the feed ring through header
- joints.
} Because evaporation from the feedwater header during steam generator depressurization can cause steam accumulation, potential header dry-out cannot be totally precluded. If.a i steam pocket does form, the BWI design is less prone to serious consequences because the feedwater pipe goose neck will retard rapid condensation and water-slug acceleration better than a long, thin steam pocket, Additionally, the feedwater header is designed to prevent
- collapse if a large steam pocket were to condense and create a near vacuum.
- Operating BWI recirculating steam generators have not experienced water hammer 1 problems because the BWI feedwater header design meets NRC guidance and improves
- upon previous designs with respect to prevention of the occurrence and prevention of 1
damage from water hammer events. 2.2.5.3 Thermal Stratification Mechanism i l l
- BWI-222-7693-LR-01 Page 55 l . At low flow rates, thermal stratification of the feedwater may occur in the horizontal section of pipe through the feedwater nozzle. This has caused fatigue cracks on the inside surface of the nozzle and feedwater pipe in some steam generators (Reference 4). Dermal i stratification occurs at low power levels, when cold, incoming feedwater flows underneath i a warmer, less dense stagnant layer of water. With the low degree of mixing at these low j- flows, the division between cold and warm feedwater remains well defined. A feedwater #
flow of approximately 600 gpm of feedwater flow at 70"F is typical for hot standby, This ;
- flow is low enough to cause an uneven flow distribution across the horizontal portion of the
, feedwater pipe (Reference 4). Changes in local pipe wall temperatures associated with a fluctuating thesial layer cause l stress cycles that could lead to fatigue failure. NRC Bulletin 79-13 and Information Notice
- 91-28 describe thermally induced cracks found in many feedwater nozzle-to-pipe welds. A -
l similar concern exists for any horizontal sections of the external feedwater piping system. j 2.2.5.4 Design to Minimize Stratification Susceptibility
- The potential for flow stratification exists in any horizontal section of feedwater pipe, {
- including the nozzle. Mixing devices in these sections could reduce stratification, but could i l cause erosion / corrosion or loose parts at higher flow rates. The potential for flow j stratification can be reduced procedurally by preventing the introduction of cold feedwater.
4 For example, reverse purge can be employed to continually draw water out of the generator through the main feedwater nozzle during hot standby and low power operation, at which [ time flow is routed through the auxiliary nozzle. This procedure eliminates extreme shocks of cold feedwater, since when switching from the auxiliary to the main feedwater nozzle and
- stagnant slugs of cold water have been eliminated at an intermediate power, the feed flow has already been warmed.
The RSG incorporates a " goose neck" between the feedwater pipe and header (Figures ! 2.2.5-1 and 2.2.5-2). The goose neck limits the volume of pipe that can be filled with cold l water. This design minimizes the time to fill the horizontal runs of external feedwater' ! piping resulting in a rapid rise in the hot / cold dividing layer. This rise occurs quickly ! enough to prevent establishment of severe temperature distributions in the pipe wall. i Figure 2.2.5-4 shows the effect of increasing the fill rate on stress intensity with Braschel, 1 et al.'s graph of normalized stress intensity versus rate of elevation of the thermal dividing i layer (Reference 4). The vertical velocity of the thermal interface is an important factor in stress intensification. De faster rate of rise of the thermal interface afforded by the
; RSG design reduces stress intensity.
The BWI internal feedwater distribution design has considered the potential for thermal i stratification and incorporates features which minimize the risk of thermal stratification damage. The main feedwater distribution system goose neck design operates effectively to i alleviate thermal stratification. i 4
BWI-222-7693-LR-01 Page 56 , 2.2.5.5 Thermal Sleeve The BWI feedwater distribution design is an all-welded design. ' Die thermal sleeve (shown on Figure 2.2.5-2) is welded to the internal feedwater piping at the goose neck. The goose g neck is welded to a Tee that is welded to the feedwater split ring header components. This
- provides leak tight joints that protect against header drainage. The attachment point is located away from any pressure boundary thermal or geometric discontinuities to avoid j stress concentration. To prevent the attachment point between the thermal sleeve and j pressure boundary from thermal shock, an inner thermal sleeve attached to the feedwater j header downstream of the nozzle is employed. This double thermal sleeve design further protects the feedwater nozzle, the nozzle to shell juncture and the outer sleeve to nozzle
. juncture from any detrimental effects due to cold feedwater impingement, or other
- feedwater thermal variations.
Careful positioning of the attachment of the thermal sleeve to the feedwater piping allows i for a design which does not interfere with nozzle or pipe UT examinations. Placement of the feedwater distribution system within the downcomer at the relatively open conical
- elevation, and the placement of the nozzle within the steam drum access space, provide j
access for examination and inspection of the feedwater distribution header, goose neck and thermal sleeve components. Secondary and primary deck access doors and access panels g : in the shroud cone facilitate feedwater system access. i l 2.2.5.6 Feedwater Distribution 4 The J-tube discharge is below the gap between the feedwater header and the internal , shroud and oriented to avoid impingement of feedwater on internal surfaces. This reduces 1 the possibility of erosion. J-tube arrangement is shown in Figures 2.2.5-2 and 2.2.5-3. Feedwater distribution system materials are selected to optimize resistance to i erosion / corrosion, thermal fatigue and corrosion cracking. The Alloy 690 J-tubes provide L erosion resistance and withstand the effects of thermal gradients and thermal cycles. b ! Feedwater is distributed axisymmetrically around the downcomer to provide a homogeneous j temperature fluid to the bundle riser. I 2.2.5.7 Maintenance Features i The BWI steam generator is designed to provide access into the unit for inspection and , maintenance. An access tunnel, complete with ladder, is positioned immediately below the secondary manway located in the steam drum head. This provides direct access through the secondary deck and into the steam drum region down to the elevation at the primary deck. From the primary deck, the feedwater nozzle, thermal sleeve, gooseneck, and feedwater ring ! T-section are readily accessible. In addition a door is provided in the primary deck to i permit access to the top of the bundle inside the shroud. From this location access may be 4 l 'made by removal of seal welded windows through the shroud cone. I ! Access to the feedwater header for remote inspection is provided via a handhole through the pressure boundary. This is located at the elevation of the feedwater header and a a
BWI-222-7693-LR-01 Page 57 1 onented approximately 180' to the feedwater nozzle, at the header split location. Access into the header for inspection is made through the J-tube located at each end of the header
- ring sections (at the header split). This allows for remote fiber optic or camera inspection ,
of the entire header system from inside the header. ! 2.2.5.8 Auxiliary Feedwater System l 1 An auxiliary feedwater system is used for the addition of cooling water during upset d I emergency or faulted conditions. In addition, during normal operation feedwater may be ) , introduced through the auxiliary feed system at less than 25% power in order to relieve l thermal stresses on the main feedwater nozzle. The auxiliary feedwater system uses many , features similar to the main such as a welded thermal sleeve, erosion-corrosion resistant l material and an upturn in the header to prevent stratified flow. Unlike the main feedwater header, there are no J-tubes and fluid exits out the end of the header in an elevated section. , The auxiliary feedwater arrangement is illustrated in Figure 2.2.5.5. Reference for Section 2.2.5
- 1. Serkiz, A. W., Evaluation of Waterhammer Experience in Nuclear Power
. Plants - Technical Findings Relevant to Unresolved Safety Issue A-1. NRC Report NUREG-0927, Rev.1, March 1984. i
- 2. Anderson, N. and Han, J. T., Prevention and Mitiration of Steam Generator Water Hammer Events in PWR Plants. NRC report NUREG-0918, Nov.
1982. )
- 3. Branch Technical Position ASB 10-2,"Desian Guidelines for Avoiding Water j Hammers in Steam Generators". Revision 3, April 1984 (attached to Section 10.4.7 of Standard Review Plan for the rev'ew of Safety Analysis Reports for Nuclear Power Plants - LWR Edition. NUREG-0800).
~ 4. Braschel, R., et al., " Thermal Stratification in Steam Generator Feedwater Lines", Journal of Pressure Vessel Technology. February 1984, pages 78-76.
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BWI-222-7693-LR-01 . Pege 63 2.2.6 Blowdown System 1 The RSG's design for blowdown is through tubesheet holes and radial passages drilled m , the tube free lane. These are connected to nozzles at the tubesheet periphery. The RSG blowdown configuration is shown in Figure 2.2.61. ~ This arrangement provides the , capability for complete RSG drainage and does not obstruct the tubesheet or hinder steam l generator recirculation flow. Inspection operations are not hampered by internal blowdown ! piping or supports. The effectiveness of blowdown is increased by the.RSG con 6guration l and by high RSG recirculation and blowdown flow rates. 1 Corrosion product transport measurements (Reference 1) were made at Point Lepreau Generating Station during steady state operation with blowdown rates of 0.1 to 0.2% (of steam flow by weight). These tests showed that 29% to 45% of incoming iron corrosion products were removed via blowdown. This was at a plant with a high circulation rate and a well designed blowdown system but with a relatively low blowdown rate. The RSG is designed to have a continuous blowdown rate of at least 1% of full load'of steam flow without reducing steaming capacity below the specified value. This rate is nearly an order of magnitude greater than that of the Point Lepreau studies. Blowdown effectiveness may be enhanced by SG design and by rate of flow. Blowdown enhancement design features are as follows:
- 1. Blowdown is at the lowest point in the SG i.e. at the tubesheet level. !
- 2. Blowdown is via holes drilled down into the tubesheet and connected via l radial passages drilled into the tube free lane (TFL) to nozzles at the tubesheet periphery. This provides for the lowest possible takeoff point. It also provides for complete drainage of the SG for maintenance work, etc.
- 3. Provision of the above blowdown connections accommodates high rates of blowdown without exceeding erosion limits on the takeoff holes in the
! tubesheet. Even at a blowdown rate of 3%, the velocity in the two three-inch Schedule 160 blowdown headers will not produce excessive erosion-corrosion that jeopardizes the integrity of the tubesheet. d The design of the blowdown system incorporates features that preclude blockage. These include:
- 1. Recommendation that the plant operate using continuous blowdown. This helps prevent sludge build-up on the tubesheet face and over the blowdown ,
holes. l
- 2. The blowdown holes on the tubesheet face are accessible and can be cleaned.
- 3. The systent can accommodate nitrogen or water sparging by reverse flow, d
BWI-222-7693-LR-01 Page 64 l Reference for Section 2.2.6
- 1. Corrosion Product Transport Studies at Pt. Leoreau. G. Plume, W. Schneider, !
C. Stauffer, CNS Water Chemistry and Materials Conference, October 1986. i i i i t i i 6
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BWI-222-7693-LR-01 Page 66 2.2.7 Moisture Separator System The RSG moisture separators are located in the steam drum. They separate steam from : the circulating steam / water mixture from the heat exchanger section. The RSG moisture l separator assembly (shown in Figure 2.2.7-1) consists of a " curved-arm" primary stage and ) a secondary " cyclone" stage. Both are centrifugal type separators and operate with the high' i steam generator recirculation rates to produce relatively moisture-free steam at the steam l generator outlet. Design maximum moisture carryover is 0.25% (by weight), as speci6ed ! by the contract specifications. . Most of the water is removed from the steam / water mixture in the primary stage, resulting in an exiting quality greater than 90% to the secondary cyclone separators. He remaining water is then removed in the secondary stage. The compact separator design distributes flow more evenly over a larger number of separators, minimizing the potential for overload of any single separator. Their small size allows full scale testing of a single separator pair, facilitating design optimization and confirmation testing at prototypic conditions. 2.2.7.1 Primary Separators The primary separator (Figure 2.2.7-2) consists of a riser, four sets of curved-arms, and a return cylinder. The return cylinder extends above the top of the curved-arms where there are several perforations and a retaining lip, which are used to improve the water removal capabilities of the separator at high steam and water flows. The perforations are oriented to accommodate arranging the separators in a square pattern within the steam drum. The steam / water mixture exiting the tube bundle enters the primary riser at the bottom of the support deck. The mixture enters the curved-arms where a fihn of water develops on the inner wall of the return cylinder and spirals downward for recirculation. The steam exits i the top of the primary separators into the inter-stage region, which distributes the steam i prior to the secondary cyclones. The RSG primary separator provides a steam / water mixture to the secondary separator at l greater than 90% quality. Separation ofliquid and steam occurs in the region of the curved - arms, above the drum water level and minimizes the potential for steam entrainment in the steam drum water inventory. The small inter-stage space and compact separation region d l allow the primary separators to be positioned relatively high in the steam drum. This allows a higher water level and higher driving head in the downcomer, increases circulation ratio, reduces the chance of feedwater header draining, and reduces the chance of uncovering the tubes. The relatively large flow area through the curved-arms eliminates the need for periodic cleaning. The absence of narrow flow passages which could collect deposits reduces pressure drop and lengthens service life. 2.2.7.2 Secondary Separators i The secondary separator (Figure 2.2.7-2) also operates on the principal of centrifugal separation. The cyclone separator does not have the flow velocity limitations of a scrubber 1 i
- BWI-222-7693-LR-01 Page 67 1
s separator. This allows much higher steam flow per unit area. The steam enters the cyclone through tangentialinlet vanes at the bottom of the cyclone which spin the steam. The liquid in the steam is forced to the cyclone wall where it passes through exit vanes and drains back to the main steam drum for recirculation. Flow holes in the top plate of the secondary l l compartment provide a small steam flow through the secondary skimmer slots, improving j 2 separator performance. I ! The secondary separators are arranged with each separator in its own compartment and with its own drain tube. If unequal separator flow and inlet quality occur, steam exiting each separator is precluded from adversely affecting the performance of the others. This
! arrangement best matches the BWI separator test configurations. Preliminary moisture i i
carryover information from Millstone 2 RSG has veri 6ed the adequacy of this design. l i 2.2.7.3 Separator Performance The primary and secondary separator have been extensively evaluated at the B&W Alliance
- Research Center. The results of these evaluations show the BWI RSG design to produce
- relatively dry steam over a range of operating conditions. j 1 l l Steam generator circulation ratio is defined as the mass of mixture entering the steam l
! separators (riser flow) divided by the mass of steam exiting the steam generator (steam ' l flow). High circulation ratio improves water level stability by maintaining a lower void i fraction in the steam generator inventory. High circulation ratio also minimizes deposit i build-up and tube corrosion. Circulation ratio is increased in the BWI design by raising
- downcomer head and reducing flow losses. The low primary separator pressure drop 1
increases circulation in the BWI steam generators. Figure 2.2.7-3 shows the relation of circulation flow to steam flow. RSG circulation ratio is highest at low steam flow and
- decreases as steam flow increases. Riser flow is low at low loads and increases with increasing steam flow, becoming approximately constant between one-third and full load.
! Moisture carryover is the amount of liquid exiting the secondary cyclone expressed as a percentage of steam flow by weight. RSG specifications typically require moisture carryover to be less than 0.25%.
- Figure 2.2.7-4 illustrates moisture carryover performance versus steam flow for a BWI l separator pair at a saturation pressure of 880 psia. The moisture carryover is shown to j remain well below 0.25% by weight over the range of tested flows. The following j subsections describe the effect of operating pressure, water flow, and water level fluctuations
, on RSG performance, the effect of steam carryunder, and the design life of the BWI steam separators. 4-1 2.2.7.4 Sensitivity to Operating Pressure Fluctuations i Test results show BWI steam separators to be insensitive to operational pres'sure fluctuations (Reference 1). Figure 2.2.7-5 shows moisture carryover to be insensitive to
- operating pressure changes for one pair of BWI separators operating at a (high) circulation
BWI-222-7693-LR-01 Page 68 I ratio of 6.0 for three nominal steam flows.
- 2.2.7.5 Sensitivity to Water Flow Fluctuations Test results show BWI steam separators to be insensitive to water flow fluctuations, and therefore insensitive to steam drum flow imbalances (Reference 1). Figure 2.2.7-6 shows
, that moisture carryover remained insensitive to flow increases up to approximately 160% i of full flow. t
- 2.2.7.6 Sensitivity to Water Level Fluctuations l
l Figure 2.2.7-7 shows moisture carryover versus water level at two different steam flows, and j shows the steam separators to be insensitive to changes in water level below the primary a separator curved arms (Reference 1). This r.llows latitude for water level changes. , 1 { 2.2.7.7 Steam Carryunder j Steam carryunder is steam that becomes entrained in downcomer flow. It is due to l ineffective separation of the steam-water mixture from the separators allowing the return of water with entrained steam to drain from the separator back into the downcomer. This ] reduces the downcomer fluid density by creating steam void which results in a reduced d
- driving head for circulation. It contributes to the swelling potential of the steam drum
! water inventory and can adversely affect water level control. Carryunder can also increase l the downcomer temperature and cause the primary to secondary differential temperature
- to be reduced. This can impact the steam generator performance.
j Under normal operating conditions, the BWI separators have been shown to produce
- insignificant carryunder. They are highly efficient and con 6gured such that the separation i
of the steam-water mixture remains above the water level of the steam drum during normal operation. This minimizes the potential for steam entrainment and swelling in the steam j drum water inventory. Hus, it improves water level control of the RSG. 4 4 l 2.2.7.8 Separator Design Life t i BWI primary and secondary steam separators are designed to last for the life of the steam b
- generator without maintenance or periodic cleaning. The primary separators have large
! flow passages that preclude plugging even if deposition occurs. The secondary separator
- inlet body and outlet passages are also large. The relatively small skimmer and vent hole passages of the secondary separators are swept by flow during operation. In the unlikely
; event of skimmer or vent h' ole pluggage, they can be cleaned by water lancing. Access is provided from above the secondary separator deck without separator disassembly. . Separator design, sizing, material selection and water chemistry control minimize the . potential for corrosion to help ensure that the separators will function without maintenance l or replacement for the life of the steam generator.
( i n --
BWI-222-7693-LR-01 Page 69 Reference for Section 2.2.7
- 1. "High-Efficiency Separator Equipment for Use in Recirculating Steam !
Generator", M. J. Reed, W. P. Pructer, P. Caple, T. Boyd, ASME Winter Annual Meeting, New Orleans, LA, November 18 - December 3,1993. l l 1 l 2 l l l 6 I
BWI-222-7693-LR-01 Page 70 A ii" l 4 l ti
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BWI-222-7693-LR-01 Page 72 30 Typical Operational Circulation Ratio - 25 k 20
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BWI 222-7693-LR-01 Pags 73 b BWI SEPARATOR TESTING CAP-3 & MODULAR SECONDARY SEPARATOR POWER SERIES TESTING,-20" WL B&W ARC 1000 PSIA FACILITY, MARCH 1994,62414 DPC 100% Power Operating Condition 880 PSIA 0.5 *****O**** PROTOTYPE coa.mer spain.4 unimum (e.2s
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BWI-222-7693-LR-01 B&W Separatar Perf rmance SW ! Pressure Sensitivity Test Results Circulation Ratio = 6.0 1 87f STEAM FLOW 0.5 IO5r STEAM FLOW CustomerSpecirnd Maalmum(0.25% by welsht) " i o
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BWI-222-7693-LR-01 Pege 75 A B&W Separator Per.formance Water Flow Sensitivity Test Results Sat 6 ration Pressure = 880 psia Nominal Steam Flow = 41,640lbm/hr 1 1 0.5 CustomerSpecirwd Maximum (0.25% by weight)
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4 BWI-222-7693-LR-01 P:ga 76 A B&W Separator Performance Water Level Sensitivity Test Results Saturation Pressure = 880 psia Circulation Ratio = 6.0 B7% STEAM FLOW 0 20 - 105r STEAM FLOW
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BWI 222-7693-LR-01 Page 77 , 2.2.8 Minimized Weld Design In general, the RSG reduces the amount of ISI required to be performed. One exception is the additional ISI at the nozzle-to-safe end welds. Specifically, the RSG is provided with d safe-ends on the feedwater and primary inlet / outlet nozzles to facilitate field fit-up and , reduce post weld heat treatment exposure of vessel nozzle welds. The vessel, however, is designed to reduce the number of welds as far as practical and the subsequent ISI as , appropriate. Specifically:
- 1. The RSG uses a one piece forged steam drum head with an integrally forged
- nozzle. This eliminates the nozzle-to-head weld used in the OSG and the ISI 3
inspection of this weld. 4 i 2. The RSG conical transition section is a single forging. The OSG conical section is fabricated from three formed plates welded together with ' longitudinal seams. Thus, three longitudinal welds are eliminated by the RSG. Furthermore, the conical section of the RSG is forged with integral 1 cylindrical sections on each end of the cone. This allows the circumferential ' cone to-shell and cone-to-drum welds to be cylinder-to-cylinder con 6gurations rather than cone-to-cylinder configurations as on the OSG. This simpler geometry facilitates the volumetric ISI examinations reil uired by Section XI
- of the ASME Code.
1 l 2.2.9 Integral Flow Restrictor l Certain plant designs include flow restrictors in each main steam line near the steam generator. Their function is to limit steam line break flow for breaks downstream of the j flow restrictor and to provide a flow measurement signal. The RSGs include steam flow restrictors as an integral part of the steam generator outlet nozzle to limit steam flow during
- any steam line break accident, but are not used for flow measurement. The original flow measurement device is retained for that purpose.
. If a double ended rupture of the main steam line were to occur, steam flow would become choked at the flow restrictor rather than a't the break location. Under this condition, steam flow depends on flow restrictor area rather than break area. Reduction in steam line break i flow limits piping loads and energy release rates. BWI designs the flow restrictor to limit
- internal mass flow rates to four times their normal full load value. The flow restrictor is j designed to minimize normal operating pressure drop while ful611ing its flow-limiting function. Steam generator internal pressure differentials during a postulated steam line
] break are limited to acceptable levels.
)
1 2.2.10 Nozzle Dams The RSG design includes nozzle dams which isolate the RSGs from the hot and cold leg
- piping to permit maintenance and inspection within the primary head during refuelling
- operations. The nozzle dams are held in place by locking devices that engage the nozzle dam retention ring. Figure 2.2.10-1 shows RSG nozzle dam details.
a
BWI-222-7693-LR-01 Page 78 The RSG nozzle dams are similar to those in use on the OSG. The nozzle dams are designed so that primary seal failure does not result in catastrophic dam failure. They are designed and fabricated according to the requirements of an owner approved nozzle dam technical specification which provides details of the dams' design, service and test loadings as well as requirements for materials, fabrication, Q.A., cleanliness, and packaging. Although the nozzle dams and rings are not ASME Section III components, their design generally follows Code design philosophy. The nozzle dam vendor provides a design report that details nozzle dam structural design calculations. The nozzle dam rings are manufactured from Alloy 690 materialin accordance with ASME SB 166. They are welded to the head cladding surface concentric with the nozzle by full
~
penetration welds that comply with the requirements of the ASME Code for Section III Class 1 vessel attachment welds. A structural analysis of the weld is performed according to Section III and included in tiie RSG Code stress report. l
BWI-222-7693-LR-01 Page 79 1 l 3 1 DAM SEGMENT (3) i
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- . DIAPHRAGM RSG N0ZZLE DAM DETAILS '
4 FIGURE 2.2.10-1 e
BWI-222-7693-LR-01 Page 80 2.2.11 Provisions for Inspection 1 Adequate provisions for RSG inspection for maintenance and repair are provided. These
; include primary side manways, secondary side manways, hand holes, inspection ports, . ; internal access hatches and ladders. 'Ihe number and location of primary manways, .
j secondary manways, hand holes and inspection ports are described below and shown on Figures 2.2.1-5 and 6. Internal RSG access is provided by access hatches in the primary and
; secondary separator decks and access ladders.
Twenty-one inch manways are provided in the hot and cold leg channel heads to provide access to the primary side of the tubesheet. Covers for these manways use a stud design compatible with efficient simultaneous multi-stud hydraulic tensioners, enhancing l accessibility to the channel heads. One manway is provided on the secondary side in the 1-steam drum head to provide access to the RSG internals.
- The channel head mar'vay covers are provided with a manipulator to ease the handling of j covers after stud de-tensioning.
! A total of ten (10) six inch handholes provide access to the secondary side of the RSG. , j Eight of these are located in the shell between the tubesheet and the lowest lattice grid ! { providing access to the ends of the tube free lane, at 90' to the tube free lane and at four l
~
other circumferential locatioris. The internal shroud separating the downcomer and riser < sections terminates above the secondary face of the tube sheet, and the blowdown system is completely contained within the tube sheet thickness. This provides unobstructed access to the entire tube bundle at the tube shut elevation and cases inspection and maintenance. One(1) handhole provides access to the steam drum and is located at the approximate ! elevation of the lifting trunnions. One(1) handhole is located in the conical shell section and provides access to the feedwater header in the vicinity of the header split location. Access to the tube bundle is provided by a total of twelve (12) two inch inspection ports located at the ends of the tube free lane at the intermediate tube support grids. There are no internal obstructions such as shroud feet, tie rods, lane blockers or blowdown header. This maximizes access for inspection and maintenance. l The two inch inspection ports and hand holes allow inspection of the lattices. These access ! ports allow fiber-optic lane-by-lane inspection and direct lane-by-lane water lancing of supports. Fiber-optic and water lancing equipment reaches the tubes through the inspection : holes at the ends of the tube free lane. Since there are no tie rods or other obstructions i within the bundle, one 90' tool can access the intertube lanes on the hot and cold leg sides j of the lattice supports. ! j The internal steam separating subassembly is an all welded design constructed with . permanent manway tunnels to the top of the tube bundle and to the steam drum head.
- internal hinged doors are secured by a captured, bolted clasp arrangement during generator operation. Steam separation equipment removalis not required for access.
i . l
4 BWI-222-7693-LR-01 Pega 81 ~ He RSG is designed to accommodate accessibility for ASME Code ISI. Lugs and insulation brackets are arranged to avoid interference with pre-service and in-service inspection and ultrasonic scanning. ; 2.2.12 Electro-chemical Polishing i A major source of radiation exposure during inspection and service of nuclear steam generators is caused by radioactive nuclides on the internal primary side surfaces. Particles of radioactive meterial are incorporated into the metal surface oxide films. They are also
- trapped in the asperities of the metal surfaces. RSG contamination build-up is reduced by
- ' polishing the channel head surfaces to a very fine finish. A chemical process, termed electro-chemical polishing, or electropolishing, produces micro-smooth channel head surfaces.
Tests have proven that this is an effective way to reduce occupational exposure (Reference l .1). -In these tests, manway seal plates showed a contact dose rate reduction by a factor of 7 when the surfaces were electro-polished prior to insertion into the primary system. i French steam generator channel heads are routinely electro-polished prior to being put into service as new or replacement units. In France all new or replacement steam generators l have electropolished channel head surfaces. In the United States, Palisades, Millstone 2 and other RSGs have electropolished channel head surfaces. i A fine surface is first produced by mechanical polishing. Electro-chemical polishing is then 4 performed to reduce surface " micro-roughness". This process reduces surface asperities by L applying an electric cunent to the metal surface through an electrolyte. Since the surfaces are finished to 63RA (roughness average in micro-inches) by mechanical polishing before
- electropolishing, the electropolishing does not appreciably alter surface roughness. It does, l- however, reduce microscopic roughness. The process removes very little material, typically
- 0.001 to 0.003 inch. The surface layers of metal are removed in a predictable and controllable manner. Electropolished surfaces contain less total surface area, and less tendency to accumulate contamination.
- The entire surface of the divider plate and primary head including nozzles and manways are
! electropolished. Because of the complex geometry, the exposed tubesheet clad face and j tube welds are not electropolished. Particular attention is given to manway cladding to minimize personnel exposure. Qualification of the electropolishing process complies with EPRI-NP-6617 and 6618. This is further discussed in Section 3.2.4.3. i
- Because of the facilities required, electropolishing is applied to the RSGs after final assembly and hydrostatic testing have been completed.
! Electropolishing is performed on the steam generator primary side surfaces by a qualified subcontractor to qualified procedures. He electropolishing procedures are qualified to verify and establish the parameters that are used for electropolishing the heads. This includes metallographic evaluations to establish optimum application parameters and to l check for any deleterious effects such as intergranular attack, dendrite attack, surface cracks, l
BWI-222-7693-LR-01 Page 82 and degree of metalloss. References for Section 2.2.12
- 1. EPRI TR-100059 Volume 1 and 2, "Effect of Surface Treatments on Radiation in Steam Generators", Project 2758-02, Final Report, November, 1991. >
1 l 4 4
; BWI-222-7693-LR-01 Page 83 2.2.13 Provisions for ALARA The BWI RSGs are designed to reduce radiation exposure to operations and maintenance personnel"As Low As Reasonably Achievable" (ALARA). Measures employed to minimize ,
radiation exposure include selection of materials and design to minimize the number of ! welds that must be periodically inspected, water chemistry that minimizes sludge accumulation, sufficient hand holes and manways to permit personnel access for inspection, ; i maintenance and repair, and provision for substantial drainage of the channel head. l Remote maintenance equipment and procedures can also minimize personnel entry time. l
- Examples of measures employed to minimize personnel exposure and off-site releases are i j described in the following sections.
2.2.13.1 Ma*erial Selection and Design to Minimize Personnel Exposure
- The BWI steam generators have been designed tc zcorporate many features to enhance access and maintainability. Steam generator cladding and tube materials are selected to minimize cobalt content. Detailed information on steam generator materials is presented in Section 2.3. Manway and manway closure design, a steam drum manway tunnel, hand holes, inspection ports, an enhanced blowdown arrangement, and internal shroud support design significantly contribute to ease of access. Efficient design of welded seams, electro-polished channel head, provision for nozzle dam support, flush tube-to-tube sheet welds, in-head tube identification system, welded construction of all internals including locking j devices of threaded fasteners, access for sludge lancing and removal, and design
- compatibility for chemical cleaning contribute to the efficiency of maintenance.
- The enhanced blowdown system is a drilled hole configuration at the tube free lane located i within the tube sheet thickness - provides the required blowdown,100% drainage of the l secondary side without additional drain connections, and unobstructed access along the tube l
free Jane on the secondary side of the tube sheet. ' ! The removal of sludge from the secondary side of the steam generator is a significant.
- outage activity that can be minimized by careful management of secondary side water
! chemistry. The RSGs have features to facilitate access and this maintenance activity:
- 1. Access to the secondary face of the tube sheet is completely unobstructed
- from the face to the first grid support by the design decisions to incorporate
- the blowdown system completely within the tube sheet thickness; to support the shroud with attachments to the lower shell course; and to support the
- tube grids solely from the shroud. These design features eliminate 1
obstructions such as blowdown headers, shroud support legs, tube support tie rods, and lane blocker devices which would interfere with effective sludge removal. l 2. The BWI enhanced blowdown arrangement permits drainage of all sludge-laden water.
. _ _ ~ _ . _-__ _ __ _ __ __ . _ . . __. - _ . _ _ _ _ _
BWI-222-7693-LR-01 Page 84
- 3. A total of eight hand holes are conveniently located at the secondary side tube sheet face, four of which are on the axes of the RSG.
2.2.13.2 Minimization of Inspected Welds The RSG design reduces the number of welds that require inservice inspection. This
. reduces personnel radiation exposure and required inspection time. Examples of design to reduce inspected welds are addressed in Section 2.2.8.
2.2.13.3 [ Water Chemistry Primary system water chemistry requirements minimize corrosion products and formation of activated material. Water chemistry is discussed in detail in Section 2.7. 2.2.13.4 Minimization of Personnel' Exposure ] i Reduction of personnel entry time required for steam generatorinspection and maintenance l activities reduces personnel exposure. Improved access and facilitation of the actions required are incorporated into the RSG design. These include: )
- 1. . Nozzle dam retaining rings have been incorporated in'o t the channel head 4
designs. Dese contribute to plant outage efficiency by simplifying nozzle dam installation and permit' channel head access for steam generator inspection and maintenance activities concurrent with other outage activities. A nozzle
- dam can be installed manually in 1 to 2 minutes. Nozzle dams are described j further in Section 2.2.10.
- 2. Large manways are provided with a stud design compatible with simultaneous multi-stud hydraulic tensioning. This enhances accessibility to the channel heads and steam drums.
- 3. The design of external manways incorporates weldable seal diaphragms as a
{ back-up to the gasket seal. This permits a readily available alternative sealing method in the unlikely event of damage to gasket sealing surfaces. .
- 4. The channel head manway covers are provided with a manipulator / hinge which is mounted to the head and capable of being readily detached from the manway. This facilitates handling of manway cover for removal and installation. Internal steam drum manway covers are provided with a hinge assembly.
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- 5. The tube-to-tube sheet welds are of the " flush" design without reduced tube
; diameter. This enhances visibility and access to tube ends.
- 6. The tube sheet face is permanently stamped with multiple tube identification symbols. Dis enhances manual location of tubes and calibration and
< ?
- __ _ _ _ . _ _ _ . _ _ _ _ _ _ _ __.~. _ . . _ _ _ _ _ _ _ _ . _ . _
4 BWI-222-7693-LR-01 Page 85 { j verification of remotely positioned tooling. 1
- 7. Hand holes and inspectica ports are provided to ease inspection and maintenance. Their number and location are described in Section 2.2.11 and shown in Figure 2.2.1-5. ,
i !
- 8. The internal shroud separating the downcomer from the riser boiler sections '
- terminates above the secondary face of the tube sheet. It does not interfere l with tube inspection or maintenance.
l 9. The internal steam separating subassembly is an all welded design with !'
! permanent manway tunnels and ladders to the top of the tube bundle and to
- the steam drum head. This simplifies access for inspection and maintenance. >
Removal of steam separating equipment is not required for access to the tube , bundle. I
- 10. Electro-polishing steam generator channel head sur6ces reduces occupational 1 i radiation exposure. Section 2.2.12 describes the electropolishing process to
- be employed on the RSGs. I i
t
- 2.2.14 Provisions for Chemical Cleanmg
- Tube deposits degrade heat transfer and, may promote tube corrosion. Chemical cleaning techniques are effective at removing tube deposits. The RSGs are capable of being
! chemically cleaned to remove deposits that cannot be removed by sludge lancing. ! i , The RSGs have sufficient corrosion allowance to accommodate six chemical cleanings using l 2 the Electric Power Research Institute (EPRI) iron-copper solvent. All RSG materials are j quali6ed for this process (see Section 2.6.5). These materials include the vessel shell, the
- internals, the tubes and the lattice supports. BWI guidance ensures chemical cleaning in
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accordance with component requirements and EPRI Steam Generator Owners Group (SGOG) guidelines. i ! Provisions for chemical cleaning include compatibility of the generator design with the l cleaning process and appropriate access. The RSG design allows full solvent. access, free ! drainage and free venting of all surfaces. Access is provided by hand holes and inspection ports for fill and drain connections, monitoring of metal-removal, and inspection. RSG connections available for chemical cleaning are as follows: 4
- 1. Eight six-inch tubesheet hand holes for fill / drain connections.
2. Tubesheet blowdown holes (or tubesheet level suction line via above hand hole) for final drain-down. ,
- 3. Two six-inch tubesheet hand holes for corrosion monitoring system probe positioning and inspection.
BWI-222-7693-LR-01 Page 86 j 4. Upper and lower connection points for level measurement. '
- 5. Upper level connection points for venting of nitrogen blanket. ,
2.2.15 Water Level Stability and Control
- Circulation Stability Instability (of water level, circulation flows, etc.) occurs for flow systems with a high degree of two-phase losses. Resistances in the circulation loop that have high steam quality j associated with them (e.g., upper support plates, U-bend region, steam separators) are the
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most sensitive to changes in pressure, by virtue of the significant effect of quality on flow i resistance. A change in pressure results in some change of flow resistance, and hence 1 circulation ratio, water level, etc. The degree of this effect, and whether oscillations amplify, depend upon many factors. The most notable is the amount of stabilizing single-
- phase loss in the circuit, as this type of loss does not vary directly with changes in pressure.
; Single phase losses in the circulation loop include the downcomer entrance loss at the primary deck, the loss associated with flow around the feedwater header and cone region, friction loss in the downcomer, and tube bundle entrance loss. For RSG designs, the l downcomer friction and tube bundle entrance loss are the most significant.
, Based on experience at B&W's Nuclear Equipment Division in Barberton, Ohio, a
- conservative circulation stability rule was developed, namely single chase losses > 0.20 two phase losses 3 I
This rule was derived in the wake of field observations ofinstability. Field observations of i instability in older equipment were corrected by adding downcomer resistance. l l Stability RSG Design i RSG design for circulation stability proceeds as follows: 1 1. The downcomer annulus is sized to be as small as possible to maximize { bundle size, but not to create unacceptable flow velocities, unacceptably low i circulation ratios, or difficulties in shop assembly.
- 2. All flow loss coefficients are calculated and put into the B&W design program CIRC.
- 3. The single-phase /two-phase loss ratio is computed by CIRC. The stability
- ratio, based on the design procedure described above, demonstrates that no A
, downcomer orifice is required for this design. I e
m i BWI-222-7693-LR-01 Page 87 ' l There is no doubt that over decades of operation, fouling and deposits in the SG can
- change flow patterns and ultimately steam generator circulation stability. A case in point l is Bruce Nuclear Station A, Unit 2, Boiler 3. !
Signi6 cant oscillations in water level were first observed at Bruce A for power levels above 95% in October,1986. By November,1988, the threshold power for observations of the oscillation was 78%. Various analyses, including simulation and site measurement , con 6rmed that the problem was caused by excessively high blockage of the uppermost i broached plates (75% for the top broached plate). The problem was eventually solved by water lancing the uppermost three broached plates. i There are two points to this example. First, the kind of blockage experienced at Bruce A l
- is one of the main reasons that BWI has chosen the lattice grid support structure for the 3 RSGs. BWI experience with lattice grid supports (at Pickering), and the experience of other
- manufacturers (C-E and KWU) have proven the superiority of the lattice grid design due
- to its open flow area. Second, BWI's model of the Bruce design simulating the pluggage 4
prior to water lancing yields a ratio of single-phase to two-phase flow of less than 0.05 at
- the threshold of instability. After waterlancing, which reduced the blockage to 55%, the unit operated stably with a calculated stability ratio of 0.1. This confirms that the design rule of single-phase losses to two-phase losses .>_0.2 is conseivative.
The factors that contribute to potential water level instability in the steam generator are i identified and understood. Provisions are made in the RSG design to maximize water level i stability. Operation and consequent deposit build-up can alter flow patterns and decrease water level stability. Operating experience has shown the superiority of the lattice grid i design in resisting problems that lead to water level instability. Therefore, RSG water level 1 will be stable, and will remain stable during plant operation. l l l l l a i. } 4 l 1 b a 4 l d
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l l BWI-222-7693-LR-01 Page 88 I 1 l l 2.3 STEAM GENERATOR MATERIALS l This section discusses RSG pressure boundary, critical-to-function, and steam generator ' internals materials. These categories are defined below, and the materials in each category i are described. RSG materials are listed in Table 2.3-5. 2.3.1 Pressure Boundary Materials { RSG pressure boundary materials consist of ferritic steels, either carbon steel or low alloy steel, and weld material to join them. The chemical analysis and mechanical properties of
- these materials are contained in Tables 2.3-1 and 2.3-2. Low alloy steels such as SA-508
. Class 3 and SA-533 Type B Class 1 are supplied in the quenched and tempered condition, ; and are qualified on the basis of their mechanical properties after a simulated post-weld , heat treatment (PWHT). His simulated PWHT conservatively bounds all anticipated l PWHTs and complies with the requirements of the Certified Design Specification. b , Welding materials are selected on the basis of their mechanical properties after a simulated PWHT, and exhibit equivalent or higher strength levels than the base metals theyjoin. l t Strength and toughness are the critical selection criteria for all pressure boundary materials. , Materials comply with the ASME Boiler & Pressure Vessel Code Section II and Section III. :
- Toughness prevents brittle fracture during high load conditions, such as hydrostatic testing l
l and upset conditions. RSG material toughness is specified by a reference temperature RT i' NDT (reference temperature for nil-ductility transition temperature) defined by ASME III, Subsection NB-2300. RT NDTs depend on impurity levels, such as sulphur, material fabrication, such as rolling or forging, and heat treatment, including the simulated PWHT. t He RT NDT for each RSG pressure boundary plate, forging or weld is equal to or less
- than 0*F. Typically, these temperatures range from -70*F to -20*F. j Materials for bolting applications are ASME Section II SA-193 Gr. B7 for bolts and studs, i and SA-194 Gr. 7 for nuts. These materials are supplied in the quenched and tempered.
l condition and have adequate strength and toughness for their application. 2.3.2 Critical to. Function Materials 1
; He term " Critical-to-Function material" is applied to materials used in components essential ;
to preserve RSG internal structural integrity or emergency heat removal capability. Rese ' ! components include steam generator tubes, tube supports, cladding and feedwater headers. Chemical analysis requirements and mechanical properties of materials used in these l components are provided in Tables 2.3-3 and 2.3-4 respectively. )
. 2.3.2.1 RSG Tube Material RSG tube material is a nickel chromium-iron alloy, ASME Section II SB-163, Code Case N-20-3, Alloy 690, that exhibits high resistance to corrosion and stress corrosion cracking in primary and secondary side environments. Code Case N-20-3 permits the use of Alloy 4
BWI-222-7693-LR-01 Page 89 690 in the construction of Class I components in accordance with Section III, Division I of the ASME Code and gives requirements for strength and design stress intensities to meet Code requirements. The chemical composition and mechanical properties of Alloy 690 are shown in Tables 2.3-3 and 2.3-4. Corrosion resistance is derived primarily from the higher chromium content and heat treatment that produce a corrosion resistant microstructure. The cobalt content of the RSG tubing and cladding material is a major contributor to the radioactivity level in the primary head. In accordance with ALARA guidelines, cobalt content is limited in the tubing material speci6 cation to an average of 0.014% with a maximum of 0.016%. The tubes are procured from a BWI approved supplier who must demonstrate by manufacture and examination of pre-production tube samples the ability to manufacture tubing to BWI's requirements. These requirements are delineated in detail in a tubing specification to which the tube manufacturer must comply. This specification describes all
' technical requirements which must be met for the tubing, including the bent regions. In addition to the material requirements, criteria are specified for dimensions, properties, cleanliness, heat treatment, defects and surface finish as well as inspection, testing, Q.A.
non-destructive examination and packing / shipping. RSG tubes receive a 100% volumetric ultrasonic inspection and a 100% eddy current surface inspection designed to detect indications of 0.002 inches deep. Tubes are rejected if they have one or more flaws in excess of 0.002 inches deep. This is more stringent than the Code requirement of 0.004 inches for indication depth and contributes to long tube life. , RSG production tubes also receive an eddy current examination (ECT) for signal-to-noise ratio. Strip chart recordings of the test form part of the documentation. The criteria for acceptance is a minimum signal-to-noise ratio of 15:1 in the straight lengths. This allows detection of small flaws during inservice inspections and monitoring of flaw growth over time. The ability to detect small flaws and monitor their growth aids in planniug , maintenance activities. Tube bend geometry restricts use of the signal-to-noise criteria to ) straight runs of tube. The EPRI " Guidelines for Procurement of Alloy 690 Steam . 1 Generator Tubing", Report NP-6743 L, Vol. 2 are used as a basis to develop procurement requirements for RSG tubes. 2.3.2.2 Tube Support Materials The tube support material for the RSGs is SA-240 Type 410S, a 12% chromium martensitic stainless steel (see Tables 2.3-3 and 2.3-4). It is supplied in the quenched and tempered, cold rolled, stress relieved condition. The tube support material resists corrosion, has
, adequate strength to support design loads and effectively resists wear when coupled with Alloy 690 tubing. Type 410S is compatible with manufacturing operations such as welding, machining, and assembly operations of the U-bend tube support and the lattice grid tube A
support structures as applicable. This material forms a tight, adherent oxide in secondary side water which is not greater in volume than the original metal. This greatly reduces the
BWI-222-7693-LR-01 Page 90 potential for tube denting. After any applied welding processes, Type 410S stainless steel j is stress relieved to reduce the hardness of the weld joint and to maintain adequate stress corrosion resistance. Yield strengths above 50,000 psi are easily achieved with 410S. g 2.3.2.3 Corrosion Resistant Cladding
- All primary side ferritic steel surfaces (primary side of the tubesheet and inside surfaces of l
the primary head) are clad with austenitic stainless steel (Type 308L Type 309L) or Alloy ! 600 or 690 weld metal to prevent corrosion. Tables 2.3-3 and 2.3-4 show the chemical d i L composition and mechanical properties of these materials. Critical aspects of this cladding ! I are the thickness, cobalt content and surface finish. A maximum cobalt content of 0.10% d j is specified to reduce residual radioactivity in areas where maintenance personnel will be ( working. A smooth surface finish of 63 microinches roughness average (RA) is specified by the customer to reduce accumulation of radioactive material on primary side surfaces. 4 f The tubesheet is clad with Alloy 600 and has.a minimum thickness of 0.312 inches in accordance with customer speci6 cations. The primary head is clad with Type 308L and 309L i stainless steel with a minimum thickness of 0.20 inches in accordance with customer ! } ' speci6 cations. 2.3.2.4 Feedwater Headers l
+
' RSG feedwater headers and associated components (gooseneck, thermal sleeve, header and j J-tubes) carry feedwater at high velocities and accommodate temperature gradients that j occur between the feedwater, the RSG water, and the RSG shell. The material chosen must ' resist erosion and corrosion, and be thermally compatible with the nozzle and feedwater piping materials. The use of low alloy (2% Cr 1 Mo) steel for the gooseneck, thermal sleeve, and header provides erosion and corrosion resistance and thermal compatibility with ! the nozzle and attached piping. Alloy 690 is used for the J-tubes to resist erosion because t fluid velocities are high. i '233
.. SG Internals Materials
! Materials for various internal components ' uchs as shrouds, decks, lugs and steam separators, i are carbon steels in a variety of product forms. These materials are readily weldable and , have the required strength for their specific applications. Internal components in close i i proximity with tubing such as tube support components are fabricated from stainless materials. High chrome materials are also employed for components potentially susceptible ' ! to flow assisted corrosion such as some downcomer and feedwater components. When the ; material is being selected, sufficient corrosion allowances are applied to ensure that the components are compatible with all operating / shutdown conditions including six chemical g cleanings during the 60 year design life. The selected materials have performed well in other steam generators. 1 Materials for bolting applications are ASME Section II SA-193 Gr. B7 for bolts and studs, and SA-194 Gr. I or SA-194 Gr. 8 M.A. for nuts. These materials are supplied in the 4
BWI-222-7693-LR-01 Page 91 quenched and tempered condition and have adequate strength and toughness for their applications. To minimize the potential for loose parts bolting applications have been limited. 2.3.4 Archive Samples BWI provides archive samples of selected materials for later reference. Archive materials include pressure boundary base metals, pressure boundary weldments, selected internals, and tube samples. Pressure boundary base metal archives are provided from the steam drum head forging, secondary shell plate, the conical transition forging, tubesheet, forged primary head, and various forged inserts such as the handholes, manways, and feedwater nozzles. Weldment archives are provided for representative pressure boundary long seam welds. Single archive samples of the U-bend anti-vibration bars and lattice tube support bars are provided. He following archive samples of tubing are provided:
- 1. U-bends with 3 ft straight legs, two per row (ie. radius) from each of the 119 rows, for a total of 238 pieces, per station.
- 2. One foot sections of continuous straight tubing from 'different tubes from !
each of the approximately 44 heat / lot combinationsjust prior to the thermal treatment. l l
- 3. One foot sections of continuous straight tubing from different tubes of each of the approximately 44 heat / lot combinations after the thermal treatment.
- 4. Two simulated tubesheet pieces each including 77 tubes fully expanded and seal welded into the tubesheet block (Mockup Requirement).
l The tubing archive samples have undergone all the manufacturing process steps (m-process , cleaning, grit blasting, acid cleaning, stress relief, etc.) applied to the production tubing i ! installed in the generators except as noted in Item 2 above. l 2 1 } i 4 a 1
BWI-222-7693-LR-01 Page 92 Table 23-1 CHEMICAL ANALYSIS REQUIREMENTS FOR PRESSURE BOUNDARY MATERIALS' CHEMICAL' ANALYSIS, %( l SPECIFICATION C Mn P S Si Ni Cr Mo V Cu Others l Base Material SA-508 Cl. 3 0.24 1.20-1.50 0.010 0.005 0.15-0.40 0.40-1.00 0.25 0.45-0.60 0.01 0.1 Al 0.04 SA-533 Tp B CL1 0.24 1.15-1.50 0.035 0.005 0.15-0.40 0.40-0.70 - 0.45-0.60 - - l SA-106 Gr C 035 0.29-1.06 0.048 0.058 0.10 - - - - - - SA-336- 0.030 2.00 0.040 0.030 1.00 10.0-14.0 16.0-18.0 2.0-3.0 - N 0.10-0.16 l 1 F316N/316LN2 d l Small Nozzles l SB-166 N06690 0.05 0.50 - 0.015 0.050 58.0 27.0-31.0 - 0.5 Fe 7.0-11.0 (min.) 0 Weld Consumable SFA 5.23 (EF2) 0.12- 1.70-2.40 0.025 0.010 0.20 0.40-0.80 - 0.40-0.65 - 03 - 0.18 5 SFA 5.5 (E8018-C3) 0.12 0.40-1.25 0.03 0.010 0.80 0.80-1.10 0.15 0.15-0.35 0.05 - - UNS W86152 0.05 5.0 0.03 0.015 0.75 Bal. 28.0-31.5 0.50 -
.50 Cu 0.5 Fe 7.0-12.0 Ti0.5 Col.1.0-2.5 SFA 5.28 (ER80S- 0.07- 1.60-2.10 0.025 0.01 0.50-0.80 0.15 -
0.40-0.60 - 0.5 D2) 0.12 0
' Maximum values or range unless stated otherwise.
2 Procured to 316LN chemistry requirements.
Reference:
ASME Code Section II,1986.
BWI-222-7693-LR-01 Page 93 Table 2.3-2 MECHANICAL PROPERTIES OF PRESSURE BOUNDARY MATERIALS 1 Ultimate : - Reductio 7'. . Tensile
- Yield Stangth n LMax.8RTy iSpeelficationl :SG A J(psi); i Elong. : 1in - NDT
((psi) f L(%); Ama 1(*F)'
- (%)+
Base Material SA-508 Cl. 3 80,000-105,000 50,000 18.0 38.0 0 SA-533 Tp B Cl. I 80,00-100,000 50,000 18.0 - 0 SA-106 Gr. C 70,000 40,000 20 - 0 SA-336-F316N/316LN 80,000 35,000 25.0 45 0 d Small Nozzles SB-166 N06690 85,000 35,000 30.0 - - Weld Consumable SFA 5.23 (EF3) 80,000-100,000 68,000 20.0 - 0 SFA 5.5 (E8018-C3) 80,000 68,000-80,000 24.0 - 0 ' UNS W86152 80,000 - 30.0 - 0 SFA 5.28 (ER 80S-D2) 80,000 68,000 17.0 - 0 f* Minimum properties unless a range is shown. 2 Typical customer requirement. 1 1 ) i f e i 1
!- BWI-222-7693-LR-01 Page 94 Table 2.3-3 CHEMICAL ANALYSIS REQUIREMENTS OF CRITICAI-TO-FUNCrlON MATERIAlf g g. - ~r-- - , ( SPECplCATION; l Ni Cr Fe om Me C Si 5 P Ce N Al B Me T1 Nb
+ Ta Tubes SB-163 ABoy 690 58A 28531.0 8A11.0 0.5 0.50 OA15-0.025 0.50 0.00 OA1 0.01 0A3 0.50 OAO 0.40 0.2 - 0.1 Code Case N-20-3 main. 2 5 6 4 Tube Support Materiais SA-240 Tp 4105 0.60 11513.0 1.00 0.05.Q06 1A0 0.03 0.04 0 0 Ci.dm. u ma SFA 5.9 ER 308L 9A11.0 18A21.0 0.75 0.05-2.5 0A3 1.0 0.01 0.03 0A2 0.75 0
SFA 3.9 ER309L 12A14.0 22.0-25A 0.75 1A2.5 OR2 1A 0.01 0A3 0.02 0.75 0 N06082 67A 18A22A 3A 0.5 2.5-3.5 0.10 0.50 DA1 0.03 0.75 sein 5 N06052 Rem. 28A31.5 7.0-11.0 0.3 1A 0.04 0.50 OA1 0.02 1.1 1A 0.5 5 0 W86152 Rean. 28.0-31.5 7A12.0 0.5 5A 0.05 0.75 Oh! 0.03 93 03 0.5 5 0 Feodomeer Hender Materini SA-335 P22 1.9-2.6 0.30460 0.15 0.50 0A3 0.03 0.87-0 0 1.13 Maximum values unless a range is shown. 2 As-deposited weld metal Table 2.3-4
BWI-222-7693-LR-01 Page 95 TABLE 2.3-4 MECHANICAL PROPERTIES OF CRITICAL-10-FUNCTION MATERIALS' Ultimate Tensile Yield ~- Elong. : Reduction of Haniness Strength. Str % 1 (%) Areal Rockwell "B"-' (ksi) ' (ksi): (%) Tubes SB-163 Alloy 690 89.0 40-55 30 - 95 Max. Tube Support Materials SA-240 Tp 410S 80.0 50.0 18 - 95 Max. Feedwater Headers SA-335 P22 60.0 30.0 30 - - Minimum properties unless a range is shown. l l I I
BWI-222-7693-LR-01 Page 96 TABLE 2.3-5 ' TYPICAL REPLACEMENT STEAM GENERATOR MATERIALS , Component Material Pressure Boundary Primary Head SA-508 C13 Tubesheet SA-508 O 3 Boiler Shells SA-533 Tp B C11 Cbnical Shells SA-508 O 3 Steam Drum SA-533 Tp B Q 1 4 Steam Dome SA-508 0 3 Tubes SB-163 Code Case N-20-3 Alloy 690 , Primary Head Components Primary Nozzles SA-508 0 3 ; Primary Nozzle Safe Ends Primary Manway Cover Plate SA-336-F316N/316LN SA-533 Tp B Q 1 A l Primary Manway Studs SA-193 Gr B7 l Primary Manway Diaphragm SB-168 UNS N06690 Primary Manway Gasket Seating Surface Primary Head Cladding SFA 5.4E3091/308L SS Weld Build Up (Equivalent to 304L) SFA 5.9 ER.309IJER 308L d (equivalent to 304L) Tubesheet Primary Side Cladding SFA 5.14 ER Ni-Cr3 (equivalent to Alloy 600, Inco 82) Primary Divider Plate SB-168 N06690 - Nozzle Dam Rings SB-168 N06690
BWI-222-7693-LR-01 Pege 97 TABLE 2.3-5 (cont'd) TYPICAL REPLACEMENT STEAM GENERATOR MATERIALS b Secondary Shell Components Handhole Forging Inserts SA-508 C13 l Handhole Studs SA-193 Gr B7 l Handhole Diaphragm SB-168 N06690 Handhole Gasket Seating Surface SFA 5.4 E309I/308L SS Weld Buildup (Equivalent to 304L) Handhole Cover Plate SA-533 Tp B Cl 1 Inspection Ports SFA 5.5 (E8018-0) or SFA 5.23 (EF2) L Inspection Port Studs SA-193 Gr B7 Inspection Port Diaphragms SB-168 N06690 Inspection Port Cover Plate SA-533 Tp B C11 Inspection Port Gasket Seating Surface E3091/308L SS Weld Buildup (Equivalent to 304L) l l Secondary Manway Forging Insert SA-508 C13
- Secondary Manway Studs SA-193 Gr B7 Secondary Manway Gasket Seating Surface SFA 5.4 E3091/308L SS Weld Buildup (Equivalent to 304L)
Secondary Manway Cover Plates SA-533 Tp B Cl 1 l Small Nozzles (3/4 in) SB-166 N06690 and SFA 5.5 (E8018-0) Weld Build-Up i 3" Recirculation Nozzle SFA 5.5 (E8018-0) and UNS W86152 l Blowdown Nozzles SFA 5.5 (E8018-0) and UNS W86152 l Steam Outlet Nozzle / Flow Restrictor SA-508 C13 (integral with Dome) l . Steam Flow Restrictor Venturies SA-312 304L Steam Outlet Nozzle Safe End SA-106 Gr. C Auxiliary Feedwater Nozzle SA-508 C13 l Auxiliary Feedwater Nozzle Safe End SB-166 UNS N06690 Main Feedwater Nozzle SA-508 C13 Main Feedwater Nozzle Transition Piece SA-508 CL 3 Secondary Shell Lugs SA-106 Gr. B Vessel Lifting Trunnions SA-508 C13 i- Lateral Support Trunnion SA-106 Gr. B
BWI-222-7693-LR . Ptge 98 , TABLE 2.3-5 (cont'd) TYPICAL REPLACEMENT STEAM GENERA'lT)R MA'mRIALS Secondary Internals Wrapper (Shroud) SA-516 Gr 70 Shroud Pins SA-193 Gr B7 Lattice Ring SA-516 Gr 70 Lbttice Bars SA-240 Tp 410 S Lattice Ring Studs SA-193 Gr B7 U-Bend Flatbars SA-240 Tp 410 S U-Bend Archbars SA-516 Gr 70. , J-Tabs SA-240 TP316L A [; l Feedwater Nozzle Thermal Sleeve SA-335 P22 Feedwater Distribution Ring SA-335 P22 i i , l Auxiliary Nozzle Thermal Sleeve SA-234 P22 Class 1 i Auxiliary Feedwater Distribution System SA-335 P22 l Steam Drum Internals Structural Components SA-516 Gr 70 Primary Cyclone Moisture Separators A569/A36/A500 Gr B .t Secondary Cyclone Moisture Separators A569/A36/SA 106 Gr B 1
BWI-222-7693-LR-01 Page 99 2.4 RSG DESIGN BASES The following sections describe development of the RSG design bases, compliance with design codes and standards, and conformance with regulatory guidance. 2.4.1 Codes and Standards This section describes the analysis of key design features of the RSG pressure boundary and outlines the experience and capability of BWIin design and analysis of pressure boundaries capable of withstanding internal pressure, seismic loads and loads from postulated accident events and cyclic loading that occur during operation. The RSGs are designed in accordance with the ASME Code requirements for Nuclear Pressure Vessels, Section III Division 1 (Subsection NB) and supported by comprehensive documentation. BWI has considerable experience in designing and analyzing pressure vessels to meet Subsection NB, having performed the design and analysis on over 200 nuclear steam generators, in addition to other nuclear pressure vessels for domestic and foreign markets. In 1991, BWI completed the ASME Design Report for the Millstone 2 Replacement Steam Generator project, meeting the'ASME Code, regulatory, and customer specification design requirements. This section describes reconciliation of the code versions applied to the replacement steam generators with th: existing plant code requirements. It also describes the certified design specification proceis, tests and inspections, and application of the ASME N-Stamp. 2.4.1.1 Code Reconciliation ' The RSG specification requires that the steam generators be purchased to the latest edition of the ASME code and applicable addenda approved by 10CFR50.55a at the time of the purchase order issue date. ASME Section XI requires reconciliation of the design, fabrication, and examination for use of later codes on replacement components. Relevant ASME III code technical changes from the edition to the OSG through the edition l applied to the RSG are evaluated for significance to RSG fabrication. The results of these evaluations will be documented, and typically conclude that all ASME Section III changes that pertain to steam generators from the OSG edition up to and including the RSG edition [ are reconciled. The detailed reconciliations for technical changes that are different or less restrictive include technical resolutions of each difference. 2.4.1.2 ASME Certified Design Specification
. The RSGs are designed, fabricated, inspected and tested according to the requirements of 'the 1986 ASME Boiler and Pressure Vessel Code, Section III (with no addenda) as defined in the Certi6ed Design Specification provided by the owner in accordance with paragraph NCA-3250 of the Code.
i
BWI-222-7693-LR-01 Pege 100 : 2.4.1.3 Tests and Inspections i The RSGs are tested and inspected according to the requirements of the applicable ASME ; Boiler and Pressure Vessel Code, Sections III and V and applicable addenda. ,
- Code-required tests include materials, fabrication and leak tests. Materials testing includes ,
chemistry and mechanical property testing, and may include other tests such as ultrasonic, !' i hydrostatic and surface finish tests depending on the material form and intended { application. These are performed by BWI or material suppliers. i 1 Fabrication testing includes magnetic particle, dye penetrant, radiographic, eddy current, ,
< and ultrasonic testing depending on the materials used, their form, application, and the ; specified manufacturing processes and sequence. ,
i !
; Leak testing includes hydrostatic testing of tube material and of the completed RSG. t j Additionally, BWI performs a helium leak test to confirm tube seal weld leak tightness.
2.4.1.4 N-Stamp [ The RSGs are designed and fabricated at BWI's facilities located in Cambridge, Ontario, , . Canada. This facility holds the ASME N-Stamp and complies with the requirements and rules governing its use. The RSGs are inspected by the Authorized Nuclear Inspector j according to the requirements of NCA-5000 of the ASME Code. The RSGs will be given j an ASME Code N-Stamp prior to their release to ship from the Cambridge facility. The , Certified Date Report is submitted to the owner. l ! l i 2.4.2 Comparison to NRC Guidance l I l This section compares the RSG design to NRC General Design Criteria, NRC Regulatory 1 Guides, applicable portions of the NRC Standard Review Plan, and other NRC information. l ~2.4.2.1 NRC General Design Criteria l The RSG design complies with the NRC General Design Criteria as follows: l GDC 1 - Quality Standards and Records - i The RSG design meets the requirements of GDC 1. The RSGs are designed, . fabricated, and tested to quality standards commensurate with their safety functions. The codes and standards used are defined elsewhere in this
- section, and have been selected on the basis of their applicability and adequacy. The BWI quality assurance program provides adequate assurance that the RSGs will satisfactorily perform their safety functions by meeting
- ASME NOA-1, and the requirements of 10CFR50, Appendix B. Approp'riate records of RSG design, fabrication, and testing will be maintained, as
- described in Section 3.2 of this report.
l 1
BWI-222-7693-LR - Pag 2101 [ GDC 2 - Design Bases for Protection Against Natural Phenomena - The RSG desip meets the requirements of GDC 2. The RSGs are designed l to withstand the effects of earthquakes and are protected from the effects of > other natural phenomena by their location in the reactor building. RSG i i seismic design is described in Section 2.4.3 of this report. He RSG seismic design envelopes the existing plant seismic design basis. GDC 4 - Environmental and Dynamic Effects Desip Basis The RSGs are designed to accommodate the effects of and be compatible ! with the normal operation, maintenance, testing and postulated accidents . including loss-of-coolant accidents. Protection against dynamic effects, including the effects of missiles, pipe whipping and discharging fluids that may i result from equipment failures and from events _and conditions outside the
- nuclear power unit are provided by the existing plant structures and systems. :
The RSG meets the requirements of GDC 4. ! GDC 14 - Reactor Coolant Pressure Boundary - l The RSG design meets the requirements of GDC 14. The RSG portions of the Reactor Coolant Pressure Boundary are desiped, fabricated, and tested to have an extremely low probability of abnormal leakage, rapidly propagating failure or gross rupture. The RSG design meets ASME Section III ; requirements, and complies with 10CFR50.55a. ; GDC 15 - Reactor Coolant System Design - l The RSG desip meets the requirements of GDC 15. The RSG portions of j the reactor coolant system are designed with sufficient margin to assure that RSG limits are not exceeded during any condition of normal operation or anticipated operational occurrences. GDC 30 - Quality of Reactor Coolant Pressure Boundary - The RSG design meets the requirements of GDC 30. RSG portions of the reactor coolant pressure boundary are designed, erected, and tested to the highest practical quality standards by meeting the ASME Code and 10CFR50, Appendix B. Detection and identification of the location of RSG leakage is through existing plant instrumentation and procedures. GDC 31 - Fracture Protection of Reactor Coolant Pressure Boundary - The RSG design meets the requirements of GDC 31. The RSG portions of the reactor coolant pressure boundary are designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and l l i l
I 4 BWI-222-7693-LR-01 Page 102 b postulated accident conditions (1) the boundary behaves in a non-brittle manner and (2) the probability of rapidly propagating fracture is minimized. He design considers service temperatures and other operating, maintenance, testing and postulated accident conditions; uncertainties in determining
- material properties; affects of radiation on material properties; residual, steady state and transient stresses; and size of flaws.
GDC 32 - Inspection of Reactor Coolant Pressure Boundary - l The RSG design meets the requirements of GDC 32. RSG portions of the 4 reactor coolant pressure boundary are designed to permit periodic inspection and testing of important areas and features to assess structural and leak-tight
- integrity. RSG provisions for inspection are provided in Section 2.2.11.
- 2.4.2.2 NRC Regulatory Guides :
i l ! NRC Regulatory Guides were reviewed to identify those potentially applicable to the RSG J replacement. Regulatory Guides identified as applicable to the licensing basis are met by 1 l the RSG design except as speci6ed below: i RG 1.29, Rev. 3,9/78, " Seismic Design Classi6 cation" j The RSG design complies with the NRC regulatory position. The l replacement steam generator is a seismic Category I component. Itis ! .l designed to withstand the effects of the SSE event and still perform its safety functions. RG 1.31, Rev. 3,4/78, " Control of Ferrite Content in Stainless Steel Weld Metal" J ! i l The RSG design complies with the regulatory position. All austenitic stainless ! steel welding consumables are procured and quali6ed in conformance to the
- ASME Code Section III, as well as this Regulatory Guide. In addition, j stainless cladding welding procedure qualifications are subject to the i minimum and maximum Ferrite Numbers of this Regulatory Guide and Section III.
a j RG 1.37, 3/73, " Quality Assurance Requirement for Cleaning of Fluid Systems and i Associated Components of Water-Cooled Nuclear Power Plants" The RSG design complies with the NRC regulatory position. His NRC
, Regulatory Guide applies to the tubing material used in the RSG. The i Regulatory Guide's positions are imposed on the tubing supplier in the RSG tube speci6 cation. Note that BWI uses ANSI 45.2.1 - 1980 rather than ANSI 45.2.1 - 1973 as referenced in RG 1.37.
S s 9
- _ - - I
BWI-222-7693-LR-01 Page 103 j RG 1.38, Rev. 2, Sn", " Quality Assurance Requirements for Packaging, Shipping, Receiving, t Storage, and Hardling of Items for Water-Cooled Nuclear Power Plants" > The RSG design complies with the NRC regulatory position. This Regulatory i Guide applies to the tubing from the tube mill to BWI as well as to the ! 1 completed RSG itself. RG 1.43,503, " Control of Stainless Steel Weld Cladding of Low-Alloy Steel Components" The RSG design complies with the NRC regulatory guidelines. The ferritic ! ! base metals which are clad (SA508, Cl.3, equivalent to SA533 Grade B, Cl.1, i and SA508 Cl.1), are procured to fine grain practice and are not considered susceptible to underclad cracking. Weld procedure quali6 cation is performed ~ on material of the same speci6 cation (or equivalent) as used in production. BWI performs a 70 degree longitudinal UT examination for underclad l cracking on all primary side clad inside RSG surfaces. l RG 1.44,593, " Control of the Use of Sensitized Stainless Steel" 4 The RSG design complies with the NRC regulatory position. Sensitized stainless steels are used for cladding in the primary head assembly, for cladding on gasket surfaces or for diaphragms, for the feedwater ring assembly and for the channel divider plate. In each instance the cladding is not a pressure retaining component and is L grade material on all primary surfaces; the feedwater assembly on the secondary side is not subject to post-
- weld heat treatment temperature and the divider plate is L grade material and is not subject to post-weld heat treatment temperatures.
RG 1.50,593, " Control of Preheat Temperature for Welding of Low-Alloy Steel" 1 The RSG design complies with the NRC regulatory position. This NRC ; Regulatory Guide applies to weld fabrication for low alloy components i speci6ed in Sections III and IX of the ASME Code. For production welds, BWI does not maintain preheat temperature until post-weld heat treatment i ! as required by regulatory position C.2. Instead either the maximum interpass l temperature is maintained four hours or the minimum preheat temperature ; is maintained eight hours after welding. However, as permitted in C.4, this ; deviation from requirement C.2 is permitted since the soundness of the welds A ' l~ are veri 6ed by an acceptable examination procedure appropriate to the weld l under consideration. I l RG 1.60, Rev.1,12/93, " Design Response Spectra for Seismic Design of Nuclear Power Plants" BWI complies with the NRC regulatory position. l I
BWI-222-7693-LR-01 Pags 104 RG 1.61,10/93, " Damping Values for Seismic Design of Nuclear Power Plants" BWI complies with the NRC regulatory position. RG 1.71,12/73," Welder Qualification for Areas of Limited Accessibility" BWI complies with the regulatory position. RG 1.83, Rev.1, 7/75, "In-Service Inspection of Pressurized Water Reactor Steam Generator Tubes" The RSG design complies with the regulatory position with the following clarifications: The Regulatory Guide addresses bo6 new and in-service components. The RSGs are new components and as such comply with the appropriate sections of this regulatory guide. Specifically C.I.a, C.1.b, C.2, C.3.a, and C.4.a. A 100 percent baseline inspection of the RSG is performed prior to the unit being put into service. BWI acceptance criteria exceeds the NRC guidelines for wall thickness reductions in that BWI limits wall thickness reductions to no more than 15% versus 20% allowed in the NRC guidelines. RG 1.84,4/92," Design and Fabrication Code Case Acceptability ASME Section III Division 1" This regulatory guide lists those Section III ASME Code Cases oriented to design and fabrication that are generally acceptable to the NRC staff for implementation in the licensing oflight-water-cooled nuclear power plants. The following Code Case is used on the RSG: N-411-1 Alternative Damping Values for Response Spectra Analysis of Class 1,2 and 3 Piping Section III, Division 1. 1 RG 1.85, Revision 28,1992, " Material Code Case Acceptability ASME Section III, l Division 1" I i The'RSG design complies with the regulatory position. The following code l cases are used on the RSG's: l N-10 This Code Case addresses the use of ultrasonic examination following RSG PWHT in lieu of radiographic examination. l
'Ihis Code Case was incorporated into ASME Section III by the 1986 addenda and endorsed by the NRC in RG 1.85, Revision 26, dated 1986. Through inclusion into the ASME Code, and annulment as a Code Case. N-10 is now listed in paragraph C.2 of RG 1.85, and deleted from the list of acceptable Code i
BWI-222-7693-LR-01 Page 105
-l Cases. However, paragraph D.3 of RG 1.85 permits components ordered to a previously approved Code Case to-remain unchanged. 1 N-20-3 This Code Case addresses the use of Nickel-Chromium Iron Alloy 690 tubing in the construction Class I components in ;
accordance with ASME Section III. , l l N-71-15 This Code Case addresses additional materials for subsection NF Classes 1,2,3 and MC Component Supports fabricated by Welding Section III Division 1. N-474-1 This Code Case addresses the use of Nickel-Chromium-Iron , UNS N06690 material (Inconel 690) with a minimum yield ' strength of 35 KSI in the construction of Class 1 components l' in accordance with ASME Section III. 2142 This Code Case was approved by ASME on November 25,1992 and addresses the use of Nickel-Chromium-Iron UNS W86052 welding filler metal (this material is similar to Inconel 690 material) in the construction of Class I components in accordance with ASME Section III. 2143 This Code Case was approved by ASME on November 25,1992 and addresses the use of Nickel-Chromium-Iron UNS W86152 weld rod (this material is similar to Inconel 690 material) in the construction of Class I components in accordance with ASME Section III. RG 1.92, Rev.1,1976, " Combining Modal Responses and Spatial Components in Seismic Response Analysis" The RSG design complies,with the regulatory position. Modal responses ! which are not closely spaced (greater than 10 percent) are combined by using the square root of the sum of the squares (SRSS) method to determine the
' maximum response for the seismic direction being considered. Dynamic
- systems that exhibit closely spaced modes (if any) are analyzed in accordance
- with the Regulatory Guide position.
RG 1.121, 8/76, " Bases for Plugging Degraded PWR Steam Generator Tubes" l The RSG design complies with the regulatory position. i l e e
i BWI-222-7693 LR-01 Page 106 RG 1.147, Rev. 5,8/86, "In-service Inspection Code Case Acceptability ASME Section XI, i Division 1" , He RSG design complies with the regulatory position. Code Case N401 is i used on the RSG to permit digitized collection and storage of data for permanently recording eddy current examination of pre-service exam.
. l RG 8.8,1982, "Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Reasonably Possible" i i
The RSG design complies with the regulatory position as it relates to the i design and supply of the RSG component. Design features are incorporated to minimize access time, to allow for rapid removal of equipment, to permit use of robotics, to provide surfaces which minimize crud traps, and to minimize welds requiring in-service inspection.
. 2.4.2.3 Comparison to NRC Standard Review Plans ,
NUREG 0800 was reviewed to identify Standard Review Plan (SRP) acceptance criteria that j potentially affect RSG design or fabrication. The relation of each potentially applicable SRP acceptance criteria to RSG design or fabrication is described below. This section is intended to acknowledge existence of SRP guidance, and to document review and evaluation ; of potentially relevant items. Other than the statements below, blanket commitment to SRP acceptance criteria is not intended. Reactor Coolant Pressure Boundary Materials (SRP 5.2.3) The RSGs meet the requirements of GDC 1 and 30 as described in Section 2.4.2.1 and RGs 1.31,1.43,1.44,1.71 and 1.85 as described in Section 2.4.2.2. Low alloy and carbon steels used as pressure retaining components are clad with austenitic stainless steel or Alloy 600. The fracture toughness requirements of 10CFR50, Appendix G are met. Steam Generator Materials (SRP 5.4.2.1) i RSG materials are those permitted by Appendix I of Section III of the ASME Code and speci6ed in detailin Parts A, B and C of Section II. RSG materials meet the requirements l of RG 1.85 as described in Section 2.4.2.2, and meet the requirements of Appendix G to
, 10CFR50 as augmented by sub article NB-2300, Section II of the Code, and Appendix G,
- Article 2000 of the Code. The RSGs are designed to minimize crevice areas where the
- tubes pass through the tubesheets as described in Section 2.2.3, and to minimize crevice areas where the tubes pass through the lattice bar supports as described in Section 2.2.4.l.
j Compatibility of the RSG tube material with primary and secondary side fluids is discussed in Section 2.3.2. l l I i 1 i l
BWI-222-7693-LR-01 Page 107 > !, Steam Generator Tube Inservice Inspection (SRP 5.4.2.2) l I He design of the RSGs provides access for an msemce inspection program of all tubes. ! His is to permit detection ofimperfections in the tube wall. A baseline tube inspection is ! scheduled prior to startup as described in Section 3.2.6.6. '
- i
? 2.4.2.4 Generic'Ietters, Bulletins, Notices and NUREGs i !
- Indexes of NRC Generic Letters, Bulletins, Information Notices and NUREGs were i
! reviewed to identify those potentially applicable to SG replacement. Documents potentially ! applicable to RSO replacement were evaluated. The relation of each potentially applicable 4
- document to RSG design or fabrication is described below. His section is intended to ;
acknowledge existence of these NRC documents, and to document review and evaluation ;
- of potentially relevant information. Other than specific statements below, blanket '
] commitment to recommendations or other provisions contained in the listed documents is ; not intended. NRC Generic Letters: 81-16 Steam Generator Overfill 4 This generic Letter discusses the potential for steam line and safety valve damage from events that fill the main steam lines with water. It describes a i potential concern that B&W (once through) steam generators may be subject ; to failure of weakened tubes by thermal transients resulting from overfill { events. His generic letter does not affect RSG design, construction or i installation. 81-28 Steam Generator Overfill i ! This generic letter is a re-issue of GL 81-16 and attaches a report from the
- Office of Analysis and Evaluation of Operating Data (AEOD). The report i identifies steam generator overfill events as an Unresolved Safety Issue. The i AEOD report focused on operating procedures and analyses rather than i design changes. This generic letter does not affect RSG design, construction l
! or installation. ' NRC Bulletins: i
- 79-13 Cracking in Feedwater System Piping NRC Bulletin 79-13 required thorough testing of feedwater-to-steam
; generator nozzle welds as a result of cracks detected in this area on numerous i
CE and Westinghouse plants. These inspections were conducted and results reported to the NRC. NRC Information Notice 91-28 (see below) concluded - that because the prescribed inspections appeared to reliably detect cracking l 4
,. . -- . - . - . - _ - - _ _ - - _ ~
BWI-222-7693-LR-01 Page 108 at the feedwater piping-to-nozzle weld no expanded inspections or other actions were necessary. The required inspections can be performed on the : RSGs. No further RSG design or fabrication actions are required to respond to the concerns identified in Bulletin 79-13. 88-02 Rapidly Propagating Cracks in Steam Generator Tubes Bulletin 88-02 described a tube rupture attributed to high cycle fatigue cracking at the top tube support plate in a Westinghouse steam generator. The cracking was caused by flow induced vibration (FIV). Owners of l Westinghouse plants with carbon steel tube support plates were requested to : review inspection data for evidence of tube denting, and implement a program to minimize the probability of rapidly propagating fatigue failure. This Bulletin is not directly applicable to the BWI RSGs because they have stainless steel support bars rather than tube support plates. However, . tube denting is addressed in Section 2.5 of this report, and FIV of SG tubes is addressed in Section 2.6.1. NRC Information Notices: 91-28 Cracking in Feedwater System Piping Information Notice 91-28 describes NUREG/CR-5285, a document that closed out Bulletin 79-13. NUREG/CR-5285 discusses feedwater piping-to-steam generator nozzle weld cracking, inspections confirming that cracks existed at many plants, repairs, and augmented inspections in the future. The NRC concluded that because the inspections appear to reliably detect degradation in feedwater piping that no additional actions are needed. The RSG design includes a thermal sleeve that is welded to the feedwater piping and prevents header drainage. The weld is located to avoid geometric and j thermal transitions in the pressure boundary that could concentrate stress. Sections 2.2.5.3 and 2.2.5.4 discuss causes of cracks and the design features l incorporated into the RSG in consideration of minimizing these causes of l cracking.
- 93-20 Thermal Fatigue Cracking of Feedwater Piping to Steam Generators i
Information Notice 93-20 states that additional feedwater line cracks have l escaped detection although augmented inspections are being performed. It states that ASME Section XI does not appear adequate to detect these thermal fatigue feedwater piping cracks. It also states that replacement of
- piping with the same material does not solve the cracking problem. To facilitate periodic inspection, the RSG thermal sleeve is positioned to avoid
- . interference with feedwater pipe and nozzle weld inspection. Also a detailed i fatigue analysis in accordance with ASME Section III is performed to analyze the effects of both mechanical and thermal induced stresses such as l
I i
BWI-222-7693-LR-01 Page 109 ) stratification. i NRC NUREGs: 0909 Ginna Event I Review of this report revealed no specific RSG design impacts. The report includes a discussion of steam generator tube operating history, with past tube plugging information, historical inspection results, and descriptions of foreign particles found during inspections of both stenm generators after the tube i rupture event. The cause of the Ginna event appears to have been tube . j j.. fretting caused by foreign material that had accumulated in the lower :
- tubesheet area. The thickness of the ruptured tube at the break was only five l
- percent of its original nominal value. Steam generator design was not an J
- issue in the Ginna event. During fabrication, cleanliness procedures are ;
! employed to ensure that debris and foreign materials are excluded from the : o RSG. Full accountability is maintained of all tools and loose parts used l
~
during assembly. An inspection for foreign objects is performed just prior to final closure before shipment, to further minimize the potential for loose
- i. secondary side parts (see Sections 3.2.7 and 3.2.8).
2.4.3 Seismic Requirements ? ! The design basis for seismic requirements is detailed in the Certified Design Specification t (CDS) according to the requirements of the ASME Code Section III. The CDS provides 1 4 the seismic, performance and transient loadings which are used to design the RSG. Included are the mechanical design data, thermal-hydraulic design data, service loads and test loads. The loadings are identified by service level (levels A, B, C, D or test) and service I limits are applied according to the requirements of Section III. These service levels include i requirements for OBE and SSE conditions and are further described in Section 2.8. In i_ order to Code stamp the vessel, the requirements of the ASME Code must be met and j verified by the Authorized Inspector. i 2.4.4 Performance Requirements o The' design basis for performance requirements is detailed in the Certified Design l Specification (CDS) according to the requirements of the ASME Code Section III. The CDS requires that the RSGs generate steam that meets or exceeds its' pressure,
; temperature, and flow rate requirements when supplied with reactor coolant and feedwater at the full load conditions. It requires that the RSGs not exceed the design basis limit of
.0.25% moisture content when operating at full load conditions.
The RSG performance requirements duplicate or exceed those of the original steam
' generator design. Additional RSG performance considerations include minimization of flow n induced vibration and avoidance of loose parts.
1 . 5
BWI-222-7693-LR-01 Pege 110 2.4.5 Accidents and Transients
- The design basis for accidents and transients is detailed in the Certi6ed Design Specification (CDS) according to the requirements of the ASME Code Section III. The CDS provides the seismic, performance and transient loadings which are used to design the RSG.
Included are the mechanical design data, thermal-hydraulic design data, service loads and test loads. The RSG design transients duplicate or exceed those used for the OSG. The loadings are identified by service level (levels A, B, C, D or test) and service limits are applied according to the requirements of Sec6on III. In order to Code stamp the vessel, the requirements of the ASME Code must be met and verified by the Authorized Inspector. b i 1 4 I + l l l
BWI 222-7693-LR-01 Page 111 2.5 DESIGN IMPROVEMENTS , BWI studies world-wide steam generator performance to identify industry problems and potential steam generator design improvements. BWI has identified 45 potential steam generator problem areas. Most are PWR issues, fewer are heavy water reactor problems. Each potential problem area was considered in designing the RSG. Table 2.5-1 summarizes these problem areas and identifies the RSG features that preclude or minimize each. Figure 2.5-1 lists the problem areas and identifies the potentially affected part of the steam generator. The following paragraphs highlight the more important problem areas and identify RSG design measures to preclude or minimize their occurrence. e Tubesheet Cladding Separation is avoided through the use of MIG cladding processes, ultrasonic testis.g of applied cladding, and a channel head divider e plate that does not impose high stresses on the cladding at the tubesheet or primary head connections.
- Tube to-Tubesheet Crevice IGA is avoided by selection and control of the tube alloy and tube expansion to minimize the crevice at the tubesheet secondary A face.
e Tube-to-Tubesbeet Crevice and Primary Side SCC is avoided by using low residual stress expansion techniques. BWI employs full depth hydraulic expansion and monitors tube and tubesheet hole tolerances and tubesheet thickness variations. Additionally, once expanded the tube is not subjected to thermal stress by post weld heat treatment processes that could relax the tube recreating the crevice.
- Tube Sensitization at the tube-to-tubesheet weld is avoided by stress relieving the pressure boundary of the steam generator, including the primary head-to-
! tubesheet weld (except for the closing seam with the steam dome portion)
- prior to tubing the generator. The closing seam, located in the conical j transition section connecting the steam drum and tube bundle shell section is stress relieved after tubing the RSG. Stress relief of this weld is performed locally, using a controlled process and does not affect the tubesheet area.
l e The Tubesheet Sludge Pile is minimized through the use of a high circulation ratio in the steam generator, high cross flow at the tubesheet secondary face, , low qualities at the tubesheet, high capacity blowdown capability, water g , chemistry limits, and provision of access ports for sludge lancing.
- e Tube Support Crud Accumulation and consequent undesirable increases in tube support pressure drop over the 60-year design life of the steam generator is avoided through the use of"open flow" lattice grids.
BWI-222-7693-LR-01 Page 112' e Denting by Tube Supports is precluded by open-flow lattice grid supports, line contact with tubes, high circulation flows to keep the tube-to-support areas clean, and select on of a tube support material (410S) that resists corrosion. i e Tube Vibration Fretting Wear at lattice grid and U-bend supports is avoided by maintaining small clearances, installing U-bend restraints as the tubing process proceeds, applying conservative analytical predictive techniques, and selecting tube support material that resists wear with the Alloy 690 interface. d e U Bend Cracking of Inner Row Tubes is avoided by use of large minimum radius U-bends and stress relieving the small, tight-radius inner row bends. e Water llammeris minimized by avoiding potential sites for the steam pockets that cause water hammer. Feedwater inlet piping volume is limited with discharge through inverted "J-tubes". e Loose Parts are avoided by minimizing the number of RSG parts and by securely capturing the applied parts that are used. All threaded connections are encapsulated such that hypothetically severed studs remain contained. Loose parts and control of parts during fabrication are discussed further in Sections 2.5.2., 3.2.7 and 3.2.8. e Steam Separator Molsture Carryover is reduced by testing prototype steam separators to 130% of required capacity prior to release for manufacture. These tests are described in Section 2.6.4. e Pressure Boundary Weld Cracking is prevented by selection and testing of shell material and weld consumables for resistance to cracking, by control of welding processes, and by control of preheat and post-weld heat treatment. The RSG is designed to resist the steam generator failure mechanisms that have been identified through industry operating experience. By following industry problems, the failure mechanisms listed above were identified and are considered in the RSG design. The following subsections describe design measures that further contribute to minimization of these problems, maximization of performance and reliability, and facilitation of maintenance. 2.5.1 . Minimization of Corrosion
- 2.5.1.1 Materials Employed to Minimize Corrosion PWR steam generators have experienced primary side stress corrosion cracking (SCC) in the small-radius U-bends and in the expanded zones of tubes. These problems have been less severe in heavy water reactors, perhaps partly due to the lower operating temperatures in these plants. But temperature differences do not fully explain the SCC differences.
l Typical PWR steam generator tubing heat treatment practices of the late 1960's and early
BWI-222-7693-LR-01 Page 113 i
. 1970's used lower annealing temperatures than those used for heavy water reactor steam ;
- generators, and did not use subsequent heat treatment. Typical PWR values are compared '
below with values for a typical heavy water reactor of the same vintage (Bruce A): :
; Typical PWR Bruce A ; , Steam Generator j l Tubing Alloy 600 600 >
Anneal Temperature 1760*F 1850*F (min) l e Subsequent Heat Treatment None Stress relief in , the vessel at 1125*F , The higher annealing temperature (improves grain structure) and stress relief (improves grain boundary carbide precipitation) improved the primary SCC resistance at Bruce A. l Laboratory results show that heat treatment of the Alloy 600 material significantly increases 1 its resistance to pure water SCC. Higher annealing temperature and stress relief, assisted
- by the lower heavy water reactor temperatures have yielded excellent tube performance.
4 Comparison of the Bruce A (Alloy 600) steam generator tubes to those of a typical PWR l steam generator, shows the following temperature differences: flant T-hot T-Cold T-U-bend Maximum i hes bu Bruce A 580 507 535 107,000 Typical PWR 610 550 570 105,000 Heat treatment of Alloys 690 and 600 for optimum SCC resistance involve a mill annealing at temperatures sufficient to put all of the carbon into solid solution, followed by a thermal
- treatment to precipitate carbides on the grain boundaries in the tube metal microstructure.
l Resistance to SCC is greatest when the grain boundaries are well decorated with carbides. I The development of Alloy 690 steam generator tubing was driven by stress corrosion cracking (SCC) of Alloy 600 in both primary side and secondary side water environments. These nickel-chromium-iron alloys differ in that the Alloy 690 has 27-31% chromium and l Alloy 600 has 14-17% chromium. A considerable amount of work has gone into evaluating j the effects of thermo mechanical processing and environment on the SCC resistance of Alloy 690 and two other popular steam generator materials - Alloy 600 and Alloy 800. i The SCC testing has demonstrated that Alloys 690 and 800 are highly resistant to cracking i in primary side water environments, while Alloy 600 is susceptible to primary water stress corrosion cracking (PWSCC). Alloy 690 resists SCC as well or better than Alloy 600 or Alloy 800 in most secondary side water environments (Reference 3). Alloy 690 has somewhat greater SCC resistance than Alloy 600 in concentrated caustic environments. Alloy 690 resists pitting and general corrosion as well or better than Alloy 600 or Alloy 800. i
BWI-222-7693-LR-01 Page 114 l Many tests have been performed which compare the PWSCC behaviour of candidate steam generator tubing. Figure 2.5-2 displays test results on statically loaded Reverse U-Bend 1 (RUB) specimens which were presented at the 1985 EPRI Workshop on Alloy 690. These results indicate that both the mill annealed and thermally treated conditions of Alloy 600 are susceptible to PWSCC. Cracking was observed in the mill annealed specimens in approximately 300 hours while 800 hours were required to crack the thermally treated
- material. Alloy 690 and Alloy 800 RUBS did not exhibit PWSCC in this 12,000 hour test i
(Reference 1). I Figure 2.5-3 presents data from statically loaded tube tensile specimens tested in 680*F
- primary water. In this test, both Alloys 600 and 800 exhibited PWSCC within 2,300 hours.
- Alloy 690 did not exhibit PWSCC in this test after 7,000 hours (Reference 1).
- Additional results collected on highly stressed Alloy 690 specimens tested in a variety of pure and primary water environments for times up to 31,000 hours indicate that Alloy 690 j is highly resistant to PWSCC (References 2 through 6).
t i Alloy 690 specimens tiere also compared to Alloy 600 specimens in " steam tests" which ! produce accelerated PWSCC. Steam tests are performed in 760*F steam produced from ! hydrogenated pure water. As in the water tests, Alloy 600 cracked within 1,000 hours while j . Alloy 690 displayed no PWSCC after exposure times up to 6,000 hours (References 4 and ! 5). i The above results indicate that Alloy 600 is susc'ptible e to PWSCC while Alloy 800 has
- generally displayed resistance to PWSCC. PWSCC of Alloy 690 has not been reported in j the open literature.
2.5.1.2 Design Features Employed to Minimize Corrosion
- Several design features that have been developed and included in the BWI steam generator design are directed toward avoiding . corrosion problems which have been observed in l recirculating steam generators. These features include the selection of corrosion-resistant 2
materials of construction, an improved tube support design, improvement of fluid dynamics i for introduction of feedwater, high circulation ratio, blowdown header design and high ! blowdown rate capability. All wetted surfaces of the steam generator primary side are , constructed of stainless steel or Alloy 690, or are clad with weld-deposited stainless steel or Alloy 600 (tubesheet cladding). RSG materials are described in Section 2.3. Material compatibility is evaluated as part of the BWI Chemical Cleaning Qualification Program (Section 2.6.5). 4 Tube-to-tubesheet crevice IGA is avoided by material selection and expansion of the tube to close the crevice at the tubesheet secondary face. Primary Side SCC is avoided by using
' low residual stress expanding techniques detailed in Section 2.5.1.1. Tube sensitization due to welding stress relief has been eliminated by stress relief prior to commencement of the j tubing operation.
BWI-222-7693 LR-01 Page 115 The Tubesheet Sludge Pile is minimized through high circulation, use of a special two-zone high/ low resistance lower lattice grid, high capacity blowdown and water chemistry i recommendations. The dual resistance lower lattice grid has an outer region of higher ' resistance to vertical cross flow through it which enables the downcomer flow to penetrate deep into the tube bundle along the secondary face of the tubesheet. The benefit of deeper penetration is to minimize the zone of net boiling and low velocities, that contribute to sludge pde formation. Three-dimensional flow simulation is used to quantify the boiling and low velocity zones. This is accomplished by the ATHOS Computer Code discussed in Section 2.6.2.2. Such zones can be graphically illustrated in figures similar to those shown in Figures 2.5-4 through 2.5-7. These design measures are complemented by periodic cleaning of the tubesheet. Tube support crud accumulation and consequent increases in : tube support pressure drop is avoided through the use of "open flow" lattice grids. References for Section 2.5.1
- 1. T. Yonezawa, et al., " Evaluation of the Corrosion Resistance of Alloy 690",
EPRI NP-4665S-SR Proceedinas: Workshoo on Thermally Treated Allov 690 Tubes for Nuclear Steam Generators. Pittsburgh, Pennsylvania, June 26-28, 1986, paper no.12.
- 2. R.M. Rentler, " Laboratory Corrosion Test Results of Alloy 600 and 690 Steam Generator Tubing Exposed to Faulted Secondary Chemistry Environments" EPRI NP-4665S-SR Proceedinas: Workshop on Thermally i Treated Allov 690 Tubes for Nuclear Steam Generators. PittsSurgh, )
Pennsylvania, June 26-28,1986, paper no.11.
- 3. RJ. Jacko, A.W. Klein and C.E. Sessions, "An Overview of L.aboratory Te:.is Conducted by Westinghouse to Qualify Alloy 690 Steam Generator Tubing" l
' EPRI NP-4665S-SR Proceedines: Workshop on Thermally Treated Allov 690 Tubes for Nuclear Steam Generators. Pittsburgh, Pennsylvania, June 26-28,
- 1986, paper no. 9a.
l 4. R.G. Aspden, T.F. Grand and D.L Harrod," Corrosion Performance of Alloy . 1 690" EPRI NP-6750-M Proceedines: 1989 EPRI Allov 690 Workshop. New Orleans, Louisiana, April 12-14, 1989. i 5. G. Santarini, et. al., " Alloy 690: Recent Corrosion Results" EPRI NP-6750-M i Proceedines: 1989 EPRI Allov 690 Workshoo.New Orleans, Louisiana, April ! 12-14, 1989.
- 6. K. Norring, J. Engstorm, and H. Tornblom, "Intergranular Stress Corrosion Cracking of Steam Generator Tubing: 25,000 Hours Testing of Alloy 600 and
- Alloy 690" Proceedines of the Fourth International Symoosium on Environmental Derradation of Materials in Nuclear Power Systems - Water Reactors. Jekyll Island, Georgia, August 6-10,1989, pg.12-1.
I I
t BWI-222-7693-LR-01 . Page 116 { f I 2.5.2 Minimization of Loose Parts ; i ! The existence of a loose part within a steam generator, whether originating from outside or ! within the unit, can cause signi6 cant damage to the steam generator particularly the tube >
- bundle components. The RSG design employs speciSc criteria to minimize the potential !
l for loose parts, namely, I
- 1. The total number of parts making up the steam generator assembly is ;
minimized within the overall design constraints. The only installed parts on- l ! the primary side are the divider plate, and nozzle dam retaining rings. The { i retaining rings are attached by full penetration welds to the head. The divider l 4 plate is attached to the tubesheet and primary head by full penetration weld ! j around its entire periphery. On the secondasy side, most installed parts are , ! weldments, others are securely captured using the measures described below, i . i
- j. 2. Where possible welded joints are used in lieu of bolted joints.
! 3. Where fasteners must be used, the design fastener material is specified with
- a chemistry that permits lock-welding of the component. !
4 j 4. Where internal fasteners that are not lock welded are used, the bolt or nut i is locked in place with a corrosion resistant locking tab. To avoid cracking ! of high strength bolting materials, fasteners with ultimate tensile strengths
- over 150 ksi are not used.
i j 5. Manufacturing and quality control procedures (described in Section 3.2.7.1) ensure that loose parts are not left in the steam generator before shipment. The above criteria ensures that the completed steam generator has minimal loose parts !,. potential. Typically, each steam generator is equipped with provisions for mounting acoustic sensors for detection of loose parts during operation. There are pads on the inlet side of , the channel head and on the shell side of the RSG for mounting the acoustic sensors in the
- same locations as on the OSG.
s 2.5.3 RSG Performance Improvements l . Several sections of this report discuss improvements in the RSG compared to the OSG.
; RSG performance is typically improved by increased heat transfer surface, higher circulation rate, lower moisture carryover and better water level stability during transient conditions.
j The speci6 cation and design changes that produce these improvements are discussed in
- Section 2.2.2 (higher heat transfer surface), Section 2.2.7.3 (high circulation rate), Section j - 2.2.7 (lower moisture cariyover) and Section 2.2.15 (water level stability).
! 2.5.4 Maintenance and Reliability Improvements RSG design aspects that are expected to reduce maintenance or improve reliability include l , I I
4
BWI-222-7693-LR-01 Page 117
- reduced corrosion (leading to tube wall thinning or cracking), less release of material to the reactor coolant that could become activated, and secondary side design and materials selection that reduce sludge accumulation and promote sludge removal. The paragraphs below describe the more significant maintenance and reliability design improvements.
Steam generator reliability can be reduced by tube thinning and cracking. He Alloy 690 tubing used in the RSGs is less prone to stress corrosion cracking than the Alloy 600 tubing used in most current PWR steam generators. Additional information on Alloy 690's
- resistance to stress corrosion cracking is presented in Section 2.3.2.
! Alloy 690 forms tightly adherent oxide films. General corrosion does not pose a large problem for Alloys 690, 600 and 800, however, in high temperature flowing ammoniated water that simulated secondary side water chemistry, the metal release is related to the alloy's chromium content. Therefore the metal release rate of Alloy 690 (30% Cr)is less than Alloy 800 (21% Cr) and less than Alloy 600 (15% Cr). ; Because of its high chromium content, Alloy 690 is expected to exhibit less general l corrosion than Alloys 600 or 800 in most secondary side environments. Testing performed i
, in an environment that simulated a condenser leak indicated that Alloy 690 had the greatest resistance to wastage, followed by Alloy 800 and Alloy 600. l The lower metal release rate of Alloy 690 benefits both the primary and secondary sides of
- the steam generator. On the primary side, the low metal release rate of Alloy 690 leads to l less cobalt transfer and lower radiation levels in steam generators tubed with Alloy 690 than those tubed with Alloy 600.
l l 2.5.5 Inservice Inspection Improvements i i The RSG design accommodates inservice inspection (ISI) with uncluttered design and ample l access. Section 2.2.11 describes access provisions for the RSG primary and secondary side j inspection, maintenance and repair. Section 2.2.8 describes the reduced number of pressure l boundary welds that require ISI. Reduced inspection requirements and improved , accessibility of areas requiring inspection reduce personnel time, expected dose and calendar l time for ISI. i i I
BWI-222-7693-LR-01 Page 118 Table 2.5-1 Industry Steam Generator Problem Areas and RSG Measures to Address Them )
\
1 Problem BWI Heaw Water S.G. BWI Solution to Generic Exnerience Industry Problem
- 1. Secondary None. Hydraulic expansion to .
SCC / IGA in All crevices closed. close crevice. i crevice Corrosion resistant 690 tubing.
- 2. Denting at None. 'losed C crevices. )
tube sheet High circulation ratio ' face to limit sludge buildup.
- 3. Tube SCC at None. Hydraulic expansion (low ID of expanded Stress-relieved Alloy residual stress). :
600 or 800 material. Alloy 690 material l region resists SCC.
- 4. Caustic IGA None. High circulation ratio l under sludge to limit sludge buildup. I Water chemistry :
recommendations. l Alloy 690 material i resists IGA.
- 5. Sludge Present in older steam High circulation ratio. ;
accumulation generators. Phosphate Enhanced bundle flow l and copper based. penetration. ! Accessibility for sludge ' lancing. l
- 6. Phosphate None despite phosphates Water chemistry wastage- in older units. recommendations.
sludge Limit sludge accumulation. J l 7. Secondary 3 minor incidents. Limit part count. loose parts Approximately 10 tubes All-welded structure. , damage (total) plugged. Rigorous tool control. j cleanliness check during j assembly. I
- 8. Open span None. Water chemistry pitting recommendations. Alloy 690 material.
- 9. Pitting Experience at Pickering Use lattice grids.
. under Reactor 5. Water chemistry i support Approximately 2000 tubes recommendations. Alloy A l plate plugged. 690 material. QV j deposits
- 10. Tube damage None.
at supports, No local stress reliefs Omproper , done.BWI manufacturing heat sequence treatment 4 1. 4
i BWI-222-7693-LR-01 Page 119 l
- 11. Support Extensive at Bruce A and Use lattice gride r
Plate '} Pickering B. Water chemistry ' clogging recommendations _!
\
- 12. Acid wastage None. Water chemistry I
' recommendations.. L High circulation ratio. ! ! Alloy 690 material. jfg q 13. Phosphate None. Water Chemistry ; , wastage recommendations.
- 14. Denting at Very minor, if any. Stainless lattice grids. l
- ~
tube support i plate l 15. Sulphur / sulf May have happened at Water chemistry l ! ate attack Bruce A. recommendations. Alloy A ! 690 material M ; 4 16. Secondary. None. Stainless lattice grids. j IGA / SCC, at Alloy 690 tubing. ; , tube support j
- 17. High cycle Very small occurrence in Ample U-bend support. -
fatigue early operation, Bruce Alloy 690 tubing. j A. j 6 i 18. Wastage None. Non-accumulating flat- l under anti- bar design. 1 vibration Water chemistry ! bar deposits recommendations. Alloy 690 tube material. 4
- 19. U-bend Bruce B scallop bar Flat-bara lattice.
, fretting designs. Ample support. Small clearances. ! i 20. Failures due None. Install FURS as bundle to is tubed. improperly Small clearances. : installed i anti-i vibration 4 bars 4 } 21. Waterhammer None. Gooseneck feedwater ring . , ! inlet. t J-tube header discharge. ' Low header elevation. t l 22. J-tube weld None. Large size J-tubes. j failures Substantial, heat '
- treated welds.
- 23. Stratificati None. Gooseneck inlet.
- on and Extended thermal sleeve.
cracking of feed nossle J
- 24. Moisture Some in oldest steam Cyclone separators.
1 carryover generators due to d clogging of scrubber [ . type driers. 4 e
}
BWI-222-7693-LR-01 Page 120 ~
- 25. Secondary some in o16. . steam Centrifugal drier steam
, . drier generators due to velocity precludes clogging clogging of scrubber deposits, type driers.
- 26. Water level some in older units. High circulation ratio trips operating procedures reduces level variation.
improved. Separators have wide
- j. tolerance range.
j 27. Water level Experienced at Bruce A, Balance single and two.
- j. stability after tube supports phase losses.
severely clogged. Non-clogging lattices maintain balance. 1
- 28. steam None. Cyclone separators carryunder eliminate.
j Feed ring design assists.
- 29. Pressure None. Forged shells. ,
boundary No corner welds. ; weld Material controls. . , failures Preheat and post-heat
- controls.
Multiple inspections. l 30. Secondary Bruce A, traced to acid Flat-bar lattice design j SCC / IGA at excursions. has no deposition sites. . I ' anti- Flat bars cannot dont or i vibration restrict tubes. bars Alloy 690 material. I o s
- 31. Lead induced Bruce A, units 1 and 2. Water chemistry ,
IGA /TGC recommendations. , Foreign material 4 exclusion controls. t
- 32. Impingement None. Considered N/A to RSG.
- Erosion Lattice grid'eupports.
- 33. Primary SCC, None. Large minimum radius
- tight bonds bends in innermost j tubes.
i stress relieve inner row tubes. thi 1 Tube material selection j (Alloy 690). 1 7
- 34. Tube support None. Stainless steel j plate lattices.
deterioratio n i j 35.. Surface Bruce A. High circulation ratio. fouling Water chemistry related. Water chemistry recommendations. t
- 36. Fretting at None. Flow mixing region top of design.
preheater 4
BWI-222-7693-LR-01 Piga 121 ! l 37. Preheater BWI not aware of any. Axial flow preheaters baffle using lattice grids as sludge tube supports. accumulation '
- 38. Combined None. Design to eliminate ;
vibration / vibration. corrosion Water chemistry recommendations. '
- 39. Tube None in BWI units. Inlet flow distribution fretting in Two loose part. fretting belt. !
preheater failures in Argentina. Analysis of entrance ! region. , conservative fretting ! criteria. !
- 40. Fretting None. RSG has no lane against lane blockers, blocker i
- 41. Primary side one incident at Gentilly Flush welds (less loose parts 2. susceptible to damage).
damage Primary side designed to , facilitate robotic repair.
- 42. Plug None. Recommend state-of-the failures art plugs, i RSG should need fewer i plugs.
- 43. Cladding None. Clad application !
Separation control. Stringent inspection. Low divider plate stress ! on cladding. !
- 44. Manway/handh one handhole gasket Advanced gasket designs.
ole failure at Bruce A. Covers and tooling gasket designed for effective failures installation. l l 4 l
~ . _ . . ._ __ _ ._ _ . . _ . _ . _ _ . _ . . _ . _ . _ _ _ _ . _ . _ . _ . _ _ _ _ _ _ . __ -._._.
BWI-222-7693-LR-01 Page 122 4
- 24. MDIS7tstE C#tPT0VER m b
[ t
- 25. SECOC@r UtlER CLOGGitG 6
) 26. WATDt LEvgt yngp3 a 1 .
- 27. WATDt LEvEt, gTA81LITT " ",
j b) o 5 J ; I j 20. STEAN CARRyyggg WELD L g 23. STRATIFICATION AND CRACKING 4
, j
- 22. J-Tunt WCLD FAILUIRES
- < 21. WATER HApMgt
- = rsmarie== v aa fam,wp-T !
'31. LEAD iteuCEp-iga b 18. U BEM FMTTise j , I p- "*.18. WASTAGE. DEPOSITS Avs s g 17. HIGH CTCLE rATICLE
- mBE '
18.. SEcoc@r laA/ scc n Tsp l . p, g,erscEsem ElioSim
'. 8
._- [ 15 f . SLLFUR/SLt.PHATE ATTACK j
[ I4' 13.PRIMMtf SCC-T!Gwr 3 Dog IW AT TSP l j 3 l u.TggiglpgEn> o I-
! *5'H"E W^ STAGE l 3
g g f . *CIO WASTAGE 12 A
- 35. EarACE FOLLipo 11. SUPPORT PLATE CLOGGING i
- 35. MIETTINil TCP (F PfGEAfDt % 10. pAMA0g pqppgpgg tcAhTT h*
M j ,,,,,,,,, , _ , - - -
*2 ' L7p g g ,s = = '
i ,,* #CCLMULAfl A 38,CD4]DEO Vl8AAfitN/CENIlI0530N--r w, p ; 3. PITTIDS-tFEN SPAN l ry [m 7= SECDOWT LOOSE PAMS DAMAGE S. P)CEPHATE.MASTAGE ucER SLM
---5. 3.tcot ActigtAyges l
- 39. .TL8E FftETT!m. PIE > EATER '
,M, e ;- ut
- 40. FilETTIl4 AGAINST l.ANE , " 4. CAUSTIC IGA uCER SLUDGE
""C*'"
- 4 . g g Loose / \ . I.o.
3 scc Expueoo nEGION
- , , , , , , 2. DEmite a Tusts cET rAct i C
- 43. Ctacolm SEPARuim 8' I'w EPCEs10E IGAesec l 44 ***c/>woctE sezT r4Ituts !
- i I
. l l l . INDUSTRY WIDE STEAM GENEl(ATOR PROBLEMS i I l FIGURE 2.5-1 ' i l
> l l
l l \ l
I l BWI-222-7693-LR-01 l l 100
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BWI 222-7693-LR-01 pags 124 l 4 i RUPTURED CRACKEDi NOMWPTURED ' MATEMAL 9 O- M.A.Alioy 600 b=- T.T. Alloy 690 i C sp. Alloy 800L :
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w. w a w TEsim mE. m I CONSTANT LOADING STRESS CORROSION CRACKING TEST l RESULTS IN TEMPERATURE ACCELERATED (360*C) PRIMARY WATER FIGURE 2,.S-3 i l l 1 e
BWI-222-7693-LR-01 Page 125 K= 6 COLD SIDE HOT SIDE - ~. K- 5
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r-li l - gx. 2 i RT != 2 j 1 1 ANou.AR R.ME K= CROSS SECTIONAL (ELEVATION) PLANE STEAM OUALITY CONTOURS FIGURE 2.5-.4 .
BWI-222-7693-LR-01 Pege 126 Ks COLD SIDE HOT 510E ,. % K- 5 7 94 Km 4 i ll C6 em en teemaa 0
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!=AN3ULAR PLATE K= CROSS SECTIONAL (ELEVATION) PLANE GAP VELOCITY CONTOURS FIGURE 2.5-5
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TUBESHEET HOT LEG FLUID OUALITY PROFILE ;p i: FIGURE 2.5-6 @ _ _ _ _ _ . _ _ _ _ _.._____.-________m___ _ _ . _ ~ _ . .
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BWI-222-7693-LR-01 . Page 129 4 2.6 CONFIRMATORY ANALYSIS AhD TESTING i Extensive analysis and testing support the BWI RSG design. This section discusses analysis and test programs that support design, manufacturing and operational aspects of the RSGs. 2.6.I Flow-Induced Vibration 1 l ' l BWI performs a flow-induced vibration (FIV) analysis in order to confirm that the tube ; l bundle is adequately supported to avoid signi6 cant levels of tube vibration. FIV reports will i
- be provided to the owner.to verify that the vibration of the RSG internals does not result in excessive were or fatigue throughout the tube bundle and U-bend regions. This section l addresses the FIV methods and evaluations supporting the RSG design. RSG design .
f measures to minimize FIV are discussed in Section 2.2.4.3. i The three pertinent cross-flow FIV mechanisms in the RSG' are vortex shedding resonance, l; random turbulence excitation and fluid clastic instability. The FIV analysis veri 6es that ~ excessive tube vibration from these sources is avoided. Particular areas of emphasis are the tube bundle entrance and the U-bend region.
- Fluid clastic instability is a mechanism which causes the vibration amplitudes to increase
, sharply when a certain critical flow velocity is exceeded. The acceptance criteria for fluid clastic instability is that the maximum cross-flow velocity at any point in the bundle is less than the critical velocity. F Another FIV excitation mechanism is vortex shedding resonance. When fluid flows across a circular cylinder, the wake behind the cylinder contains vortices. The vortices detach from
- the cylinder in a regular manner, i.e. at a certain frequency, and cause the tubes to vibrate
- at the same frequency in a direction perpendicular to the flow direction. When, at a critical j cross-flow velocity, the vortex shedding frequency happens to be close to a tube natural j frequency, the vibration of the tube can organize the wake, causing it to synchronize (lock
- in) with the tube motion at the tube natural frequency. This phenomenon is called vortex shedding resonance, and on a plot of amplitude (y) vs. flow velocity (x) it would show up i as a hump. For flow in tube bundles, vortex shedding resonance has only been observed
- to occur at the first few (three or four) outermost tube rows, and is limited to single-phase
- cross-flows. This means that for a feed ring design RSG, vortex shedding resonance ::an only occur at the bundle entrance region, which is the region where the downcomer flow
- enters the base of the tube bundle. A vortex shedding analysis in this region is carried out i by first calculating the vortex shedding frequency from the equation
) f, = S* V l D l where f, = vortex shedding frequency (Hz) S = Strouhal number (proportionality constant, determined i experimentally)
- V . = maximum cross-flow velocity l
D = tube outside diameter (O.D.) l a
BWI-222-7693-LR-01 Page 130 Subsequently, this frequency is compared to the tube natural frequencies. Vortex shedding resonance is assumed to occur if for any mode of vibration the vortex shedding frequency
- is within 30% of the tube natural frequency.
For the modes where resonance is predicted, resonant' amplitudes of vibration are calculated. The maximum allowable vortex shedding amplitude is 2% of the tube outer
- diameter.
The third mechanism, random turbulence excitation, is the buffeting of the tubes primarily from the turbulence in the Dow, and is the " background" mechanism which accounts for tube d. vibration below fluid clastic instability (FEI) and outside regions of vortex shedding resonance. It results in relatively low levels of vibration which increase with increasing flow velocity, with amplitudes and mode shapes varying randomly in time and in direction. The maximum allowable amplitude is 10 RMS mils. i
- A three-dimensional analysis is performed to derive a detailed flow distribution in the U-
- bend area. From this analysis, velocity and density profiles are determined for each of the '
j five longest tube spans (i.e. longest span with five supports, longest with four supports, etc.). A finite element analysis is used to predict mode shapes for each case for the various mode - types and frequencies. j , The B&W FIV computer code EasyFIV (Reference 2), developed by B&W Alliance
- Research Center, and the finite element macro "MSC/ pal 2" are used to determine if the l FEI threshold velocity is avoided and to analyze response to random turbulent excitation.
4 These analyses are repeated until the optimum number and position of support locations
- is achieved that conservatively meet the design criteria.
Conservatism of the B&W FIV analysis was demonstrated in two_ projects with McMaster University. Hot leg bundle entrance region FIV was measured on a full-scale model of a Darlington steam generator tube bundle (with lattien grids). The FIV response of a full- ! scale model U-bend section, simulating Darlington conditions, was measured. In both cases, i the measured FIV responses were below those predicted by B&W. 1 ' The potential for fretting is assessed by FIV sensitivity analysis. The FIV analysis is used to con 6rm that the tube bundle is adequately supported to prevent excessive tube motion due to FIV excitation mechanisms.
- The RSG bundle design parameters that are the most important for controlling FIV are
- '1. Tube and support materials.
- 2. Tube ou tside diameter, thickness and pitch / diameter ratios, and diametrical clearance at the lattice bars.
- 3. Bundle height.
l D
BWI-222-7693-LR-01 Page 131
- 4. Radius of outermost tube.
I
- 5. Number of lattice grids. !
- 6. ' Number of U-bend supports.
- 7. Width of fan bar and high bars.
l
- 8. Steam flow at full power.
- 9. Circulation ratio. i 1
These parameters are compared for the RSG and previous BWI steam generators. Typically this comparison shows similarity with existing units and indicates that all regions of the tube bundle are adequately supported to prevent excessive tube motion due to. flow
. induced vibration. This comparison provides a basis on which to conclude that the RSGs will be adequately resistant to FIV. ;
h 2.6.2 Hermal-Hydraulic Performance Verification of the CIRC Computer Code l The computer code CIRC has been the principal design tool for nuclear steam generators at the Babcock & Wilcox International Division in Canada for nearly twenty years. The code is one-dimensional and was written primarily to allow rapid data manipulation and to allow the operator to run a multiplicity of cases of design alternatives to arrive,at an optimum design conclusion quickly. The code is capable of analyzing the following alternatives.
- 1. Steady-state heat transfer, any power level.
- 2. Steady-state circulation, any power level.
- 3. Feed ring or integral preheater type steam generators.
- 4. ' Heavy water or light water primary fluid.
The only difference to the CIRC code between heavy water and light water is that the code is directed to use one or the other of two tabulations of property values such as enthalpy, viscosity, etc. as a function of temperature. The secondary side characteristics are identical and vary only in accordance with actual RSG operating pressure. The CIRC code has been used in performing the thermal hydraulic design of the majority of the BWI steam generators operating in the field today. Dese include the Pickering B l steam generators, the Bruce B steam generators, the 600 Mw units (Point Lepreau 1, 1 Gentilly 2, and Embalse,. Cordoba), the Darlington and Cernavoda steam generators, and i
)
BWI-222-7693-LR-01 Pege 132 Millstone Unit 2. All of the operating steam generators have performed satisfactorily and have seen no adverse performance due to design aspects veri 6ed by the CIRC program. Further, operational characteristics determined by the CIRC program and used to calculate velocities for the flow-induced vibration analysis of the tube bundle have been more than adequate as evidenced by no tube failure due to vibration in the operating steam generators
! to date. ;
i i ne CIRC code has been used to simulate the operational parameters of many of the
- existing plants in operation today. In this capacity, the CIRC code uses measured field data
, and as-built steam generator geometric characteristics to simulate conditions measured in ! 'the operating unit. In general, the CIRC code accurately simulated the observed site ]
- performance thus con 6rming its applicability and accuracy as a design tool for new steam :
; generators. The above listed analysis covers the full range of Steam Generator designs with l l a wide range of geometric and performance parameters expected to bound all current and i future applications of the CIRC code.
k l 2.6.2.1 Dree-dimensional Thermal Hydraulic Analysis 4 A 3-D Thermal Hydraulic Analysis is performed to assess the following design features: l
- 1. Flow penetration at the tubesheet face.
l l 2. Steam quality at the tubesheet face.
- 3. Maximum bundle cross flow gap velocities.
- 4. Flow expansion at the shroud window exit.
Flow distribution in the U-bend region. ) i 5.
- 6. Primary cyclone entrance quality distribution.
- 7. Overall heat transfer rate and circulation ratio.
i i Design features 1 and 2 above are critical for estimating sludge pile size and rate of growth. As a general rule, the sludge pile is expected to form in regions of net steam quality and l low cross flow velocities at the tubesheet face. As expected, previous analyses indicate that
- these two regions are located at approximately the same location. .
; Design features 3 to 5 are important for assessing the ability of the design to preclude j . excessive tube motion due to flow-induced vibration. Feature 5 is also used to reduce flow resistance in the U-bend region by aligning the U bend supports with the U-bend flow
' streams.
i Design feature 6 is used to ensure adequate loading of the primary steam separation equipment. . i
BWI-222-7693-LR-01 Page 133 Design feature 7 is used as additional confirmation of the heat transfer analysis results ,
- obtained from the single-dimension thermal hydraulic (CIRC code) analysis.
! 2.6.2.2 The ATHOS Computer Code i He ATHOS (Analysis of Thermal-Hydraulics Of Steam Generators) computer program is used to analyze the thermal-hydraulics and the three dimensional flow distribution in steam generators at steady state conditions. The RSG riser is modeled in " volume elements" and , the "6 nite differences" method is used to solve the equations of mass, momentum and
- energy conservation. Extensive documentation is available for ATHOS and its applications.
t He flow field in the tube bundle is represented as a three dimensional model. To deal with two phase flow it is possible to employ either the homogeneous flow model or a model with . 1 slip in the steam generator axial direction according to a drift flux model. The homogenous j flow model is used. ATHOS uses " distributed resistances" to simulate pressure losses in the steam generator, particularly the tube bend region. The free flow area in the tube bundle is calculated with the aid of" porosity". The pressure losses across the differential resistance lattice grid, the typical grids, the steam generator downcomer and the moisture separators are treated as locally concentrated resistances and assigned a pressure loss coefficient. j Refer to Section 2.2.4.1 of this report for a description of both typical and differentiallattice i grid tube supports.
- The velocity field calculated with ATHOS shows the asymmetry between the hot and cold g legs of the riser due to dissimilar steam quality. Asymmetric profiles are illustrated in
- Figure 2.5-5. In keeping with the homogenous two-phase model, the velocities presented 1 i are mixture velocities (equal water and steam velocities). The reference area for velocity l
i is based on the volume porosities of the respective cells. The radial flow displacement in the tube bend region is correctly simulated by the program insofar as the pressure loss in
- each cell is considered as appropriate to the direction of incidence on the tube (axial flow,
) cross flow or a combination of both). The velocities corresponding to the volume porosity j provided by ATHOS give no information on the velocities in the narrowest gaps between
' adjacent tubes. However, these gap velocities are crucial to vibration excitation. For this reason, the gap velocities parallel and normal to the tube axis (axial flow gap velocity and
- cross flow gap velocity, respectively) are determined in a post-processing conversion routine. l
( l In addition to the geometric input data required by the ATHOS pre-processor programs, l ATHOS input includes process data for various load step conditions, pressure loss ; j coefficients, information on the type of flow model and on the sections for which flow
; parameters are required, initial values and convergence criteria. ATHOS output includes l a summary of the input data as well as a summary of the calculated thermal-hydraulic operating data, i
Much of the thermal-hydraulic sizing of the steam generator is performed using classical analysis techniques and one-dimensional thermal-hydraulic code analysis. Th'ree- ! dimensional analysis results are utilized in very specific areas that require detailed 2 knowledge of the various flow parameters. In addition the heat transfer results from the
, . - - ~ _ _ w ,... ..- - , - . - . ,-s.- - -y - -- ,-
BWI-222-7693-LR-01 Page 134 three-dimensional thermal-hydraulic analysis are used for additional confidence in the results obtained from the one-dimensional CIRC code results. An RSG area requiring details for the three-dimensional analysis is the tubesheet secondary face region. This is the most common location for sludge deposition. The three-dimensional analysis evaluates the effects of various bundle entrance geometries and differential resistance lower tube supports. It determines the flow distribution and can determine areas oflow cross-flow and potential stagnation. His allows minimization oflow cross-flow and stagnant areas. He deposition of sludge is affected by the amount of boiling at or near the tubesheet secondary face. Sludge deposition can occur at these locations due to the inability of the steam phase to transport corrosion products that drop out of solution and form sludge piles. By minimizing the areas of high' net quality and low cross-flow at the tubesheet secondary face, the size and rate of growth of the sludge pile can be minimized. From the three-dimensional analysis, flow velocities on the outermost tubes at the bundle entrance region, flow imbalances at the periphery of the bundle, and variations in vertical l velocity at the bundle entrance can be determined. These are important to the ability of i the design to preclude FIV. As the flow rises through the tube bundle and steam quality increases, interaction of the hot I leg and cold leg produce some flow imbalance. The U-bend tube support system, and flow resistance of the tube bundle in cross-flow tend to create flow imbalances as well. The three-dimensional thermal-hydraulic analysis aids the design process in two ways. First, the position of the U-bend supports can be optimized.so there is minimal effect on flow distribution. This reduces the pressure drop in the U-bend region and, by aligning U-bend supports with U-bend flow streams, additional flow imbalance is avoided. Secondly the i three-dimensional analysis identifies flow imbalance in the U-bend region of the final l design. The flow and local quality distribution (and therefore density), determine the U- l bend gap velocities. These are used in FIV analysis of the U-bend tubes. He flow forcing function distribution is used to quantitatively confirm that the design has no tendency for
- excessive FIV excitation.
! Directly above the U-bend region is the primary separator deck. The flow leaving the U-l bend region has a steam quality distribution and a mass flow distribution. This produces i variation in steam separator component loading. The three-dimensional thermal-hydraulic i analysis of quality- and mass distribution determines the range of expected water- and steam ; loadings, and circulation ratios. BWI uses this information to determine the appropriate j
- ranges for confirmatory testing of steam separation equipment.
4 2.6.3 Tube-to-Tubesheet Joint Qualification Program The function of the tube-to-tubesheet joint is to fasten the tube to the tubesheet such that it sustains the forces associated with pressure and thermal transients while not creating conditions which could cause the tube to degrade in service through crevice corrosion or stress corrosion mechanisms. Industry wide problems associated with this region of the steam generator include secondary stress corrosion cracking (OD SCC), intergranular attack 1 i
BWI-222-7693-LR-01 Page 135 ~ ! (IGA) in the crevice and ID SCC in the expanded region. BWI steam generators have to l date suffered no tube failure due to tube corrosion at this location. Several key factors have contributed to this success:
- 1. Water Chemistry Control.
i
- 2. Materials Selection.
- 3. Design Optimization.
Primary and secondary water chemistry are addressed in Sections 2.7.2 and 2.7.3. Materials selection for corrosion resistance is addressed in Section 2.5.1. The design integrity is i evaluated by BWI's tube-to-tubesheet joint qualification program. Based on growing knowledge and understanding of the failure mechanisms associated with industry problems with the joint, BWI has developed design parameters and fabrication
- techniques that are directed at minimizing potential forjoint problems.
Exoanded Joint Interrity Requirements: An assessment of the integrity of expanded tube-to-tubesheet joints considers the following:
- 1. Pullout Strength
' Pullout strength is the force required to pull the expanded tube through the full thickness of a tubesheet. Pullout may consider the unflexed tubesheet and/or the
- integrated effect of pullout resistance over the full thickness fully flexed tubesheet.
This relates to the ability of the tube to structurally resist pullout without benefit of the seal weld. l 2. Leak Tightness i Leak tightness is the ability of the expanded joint to resist net leakage of fluid I through the full thickness of tubesheet. Leakage may consider the unflexed tubesheet and/or the integrated effect of the full thickness / fully flexed tubesheet.
. This relates to the ability of the tube to resist primary to secondary side leakag'e without benefit of the seal weld. -
- 3. Secondary Fluid Inrress Resistance Secondary Guid ingress resistance is the ability of the tube to resist ingress / weeping
^ of secondary fluid into the crevice. Ingress resistance and the depth of ingress will be affected by dilation of the tube holes due to tubesheet flexing as well as by the basic parameters of the expanded joint. l
BWI-222-7693-LR-01 Page 136 This relates to the minimization of corrosion within the crevice area due to contaminants in the secondary side fluids. ,
- 4. Control of Residual Stresses I
- Surface residual tensile stresses are a by-product of the hydraulic expansion process that creates the joint by yielding the tube plastically into the elastically yielded l tubesheet hole. High tensile stresses on the outside diameter leave the tube vulnerable to failure through in-service stress corrosion cracking.
This relates to the ability of the tube to obtain design longevity. BWI fabrication processes and technologies applied to the joints address these design objectives and result in:
- 1. Full-depth expansion through tubesheet to eliminate crevices.
- 2. High SCC resistance to both primary and secondary water through selection of I-690 material. '
- 3. Minimization of microscopic crevice. l
~
1
- 4. Hydraulic expansion to minimize and accurately control residual stresses.
1 [ 5. Full-depth expansion to minimize strains, maximize pullout strength and leak tightness. . 6. Stringent QA standards and fabrication controls assure that design requirements are met. 1 The tube-to-tubesheetjoint qualification program consists of studies, evaluations, and testing that consider the following:
- 1. Measurement of residual tensile surface stresses on the outside diameter of the transition area (and resulting SCC susceptibility). Because of the uncertainties
- inherent in this type of measurement, independent techniques are applied
l i
- a. An X-ray diffraction residual stress survey of the transition region.
- b. Finite element predictions of the stress state in the transition region.
- c. A study of SCC susceptibility of the joint transition, seal welds, and scratched tubes.
~
The residual stress experiments are complemented by existing analytical data on the BWI tube-to-tubesheet joint.
BWI-222-7693-LR-01 Page 137
- 2. An evaluation of the degradation of the structural (pullout) load capacity and leak
- tightness and secondary fluid ingress as a result of experimentally simulated plant !
] heat up and cool-down transients, i 2.6.4 Separator Testing Experience j - He performance of the primary and secondary steam separators has been extensively ,
- evaluated at the B&W Alliance Research Center. The following paragraphs discuss >
) performance results for the BWI steam separation equipment, including the test facility
- used, instrumentation, and results. Testing experience shows the RSG steam separation
, - equipment to be effective and relatively insensitive to variations in operating pressure, water j flow, water level, and steam carryunder. Additional information on steam separator design ,
and performance is found in Section 2.2.7. At typical operating conditions, the moisture carryover was shown to remain below the speci6ed design value of 0.25 percent by weight. Testing at flows ranging from 15 to 30% i
- . above the design steam flow showed that the design value moisture content was achieved.
! The Darlington steam generators incorporate the same replacement steam generators except ,
- that the secondary separators are not enclosed in separate isolated compartments. The !
{ enclosures of the secondary separators ensures a greater compatibility between the full scale prototype test configuration where a single pair of primary and secondary separators are ; i tested and the actual in-service conditions. Also by providing isolated secondary separators 4 potential performance deficiencies due to flow maldistribution between the separators is ! avoided. The Millstone 2 RSGs have a isolated secondary separators and have ; i demonstrated moisture carryover well below 0.1 percent by weight. i Steam carryunder describes steam which is carried downward with the downcomer flow. l His reduces downcomer flow density and the available driving head for steam generator l circulation. Carryunder can also raise downcomer temperature, reducing steam generator i
- performance. Tests have shown the BWI steam separators to produce no appreciable l carryunder at normal operating conditions.
BWI performs steam separator testing at its facilities located in Alliance, Ohio. BWI separator development capabilities rely heavily on experimental research utilizing state-of-the-art measurement, diagnostic, and analysis equipment supplemented by advanced analytical modelling techniques using mainframe and microcomputer hardware. Experimental research includes both quantitative and qualitative assessments of separators 4 using both an Air / Water Test Facility and a Steam Test Facility. ' .The steam test facility is designed for a pressure of 1000 psia and 'a temperature of 544*F at steam flows up to 58,000 lbm/hr. Performance parameters such as pressure drop, i moisture carryover, water level, liquid film height, steam carryunder, temperatures, and , secondary cyclone drain Gow are measured. Technology available for testing includes a i gamma densitometer for determining local density and optical techniques for obtaining high-speed videos in the steam / water environment. S
BWI-222-7693-LR-01 Page 138 , Figure 2.6-1 illustrates the steam flow and water flow capacity curve for a BWI separator pair relative to a typical operating curve for a BWI steam generator. There are substantial steam and water flow margins beyond normal operating conditions before the moisture carryover limit of 0.25% by weight is reached. , 2.6.5 Chemical Cleaning Qualification of Materials BWI qualifies RSG materials for corrosion resistance considering the effects of normal operation and multiple chemical cleanings. This is done through a program of testing and
- assessments of the various RSG materials. This program is documented in the Chemical Cleaning Qualification Report. Key elements of the program include:
- 1. Laboratory screening tests of the tube (Alloy 690) and tube support material
- (SA240TY 410S) against materials already qualified in industry chemical cleaning i database (i.e. Alloy 690, carbon steel, and a SMAW electrode).
i
- 2. Assessment of the steam generator to identify materials, welds and/or joints.
- 3. Laboratory testing of materials, welds and/or joints not qualified within the BWI i Chemical Cleaning Database, but identified in the RSG material survey.
! 4. An estimate of expected corrosion losses for the RSG materials during multiple ! chemical cleanings, as well as during normal operation. S. Specification of corrosion allowance design values based on the estimates of expected l
- corrosion.
Reference for Section 2.6 1
- 1. " Veri 6 cation of the ATHOS3 Code Against Feedring and Preheater Steam i Generator Test Data", EPRI report NP-5728, Project 1066-10, May 1988.
- 2. Flow Induced Vibration Analysis Program EasyFIV Rev. 0 Verification Package.
e I n i e j
BWI-222-7693-LR-01 Pega 139 B&W Separator Performance at 880 psia 200% i i I 150% ~
- Performance 1 Margin
$100% -
I l g - Typical Operational Curve I i-- ' f 50% 8
- E O
0 SOY 100% PERCENT STEAM FLOW _ PERFORMANCE MARGIN FOR B&W SEPARATOR EOUIPM AT SATURATION PRESSURE OF 880 PSIA FIGURE 2.6-1
l BWI-222-7693-LR-01 Page 140 l 2.7 OPERATING RESTRICTIONS WITH RSG DESIGN , i 2.7.1 Removal of Temporary U-Bend Shipping Restraints f Temporary restraints'are applied to the RSG tube U-bends to prevent shipping damage. , These must be removed prior to RSG operation. Removal requires RSG secondary side l ! entry. The temporary supports are accessed through the steam separator deck. ! 2.7.2 Primary Water Chemistry There have been relatively few corrosion problems associated with the primary system chemistry environment. In recent years, however, a phenomenon called primary water stress , corrosion cracking (PWSCC) has been observed on the primary side of certain RSG Alloy i 600 tubing. However, this corrosion phenomenon has less to do with the primary water chemistry environment than the metallurgical condition and stress levels of the tubing. This i i phenomenon is most prevalent in Alloy 600 tubing having high residual tensile stresses, e.g. ; j U-bends and expansion and bending transitions. In addition, this phenomenon occurs most
' I frequently in tubing that has been mill-annealed at relatively low temperatures (1750F).
j Thermally trea'ted or stress-relieved Alloy 600 tubing has been proven to be far less
- susceptible to PWSCC. ;
I. All U.S. PWR plants use boric acid for chemical shim reactivity control and lithium i hydroxide to raise pH. Each nuclear fuel vendor normally provides the owner with ;
- guidelines relative to the proper concentrations of these chemicals, since the boric acid concentration must be reduced as burnup of the fuel progresses. EPRI has developed and j updated guidelines for PWR Primary Water chemistry (Reference 1). These guidelines i minimize the impact of the boric acid and lithium hydroxide on primary system materials l l and fuel cladding. Reference 1 presents principles for each plant to use in developing its !
own boron / lithium hydroxide control scheme. From this information, the owner can develop i , the optunum prunary system chemistry control scheme in consultation with their fuel vendor. dissolved hydrogen, and dissolved oxygen.Other Diagnostic parameters parameters include sulfate andthat are co
; suspended solids. A water chemistry control program which covers all modes of operation is documented in the station chemistry manuals.
2.7.3 Secondary Water Chemistry 2 The water chemistry requirements for the secondary system are dependent upon the operational mode of the plant, as well as system materials. The operational modes that j require environmental control are cold shutdown, heatup/startup/ hot standby, and normal power operation. A water chemistry control program which covers all modes of operation i is documented in the station chemistry manuals. e
+
1
BWI-222-7693-LR-01 Page 141 References for Section 2.7
- 1. NP-7077, "PWR Primary Water Chemistry Guidelines: Rev. 2", Electric Power Research Institute, November,1990.
- 2. NP-6239, "PWR Secondary Water Chemistry Guidelines: Rev. 2", Electric Power Research Institute, December,1988.
F 9 4
.. i BWI-222-7693-LR-01 Page 142 2.8 RSG STRUCTURAL EVALUATION i
Structural and seismic evaluation of RSG primary and secondary side pressure boundaries demonstrate that these components satisfy ASME III, Division 1, Class I design J requirements for service levels A, B, C, and D (normal, upset, emergency and faulted , conditions, respectively). Steam generator internal components are not governed by the- ! ASME Boiler & Vessel Code . However, ASME III Subsection NB for Class 1 i components is used as a guide for structural analysis of RSG internal components. RSG ; internal components are required to withstand all specified loadings to maintain heat transfer capability during and following a design basis earthquake. This helps to ensure that safe shutdown capability is maintained. The RSG structural evaluation will be documented in a Code Stress Report. Conservative hand calculations and finite element modelling (where required for pressure and thermal transients) are employed to prove that the components examined meet the ASME Code allowable stresses. For seismic loading, an equivalent static load analysis is performed to determine seismic loads on components for subsequent stress analysis. The design and hydrotest primary stresses in the RSGs meet the design and hydrotest allowables of ASME III as shown in the following sections. The requirements of Subsection NB-3221 for design stresses are met'as follows: P, < S at design temperature Pi < 1.5S, at design temperature P+P i 3 < 1.5S, at design temperature where: P, = General primary membrane stress i Pi = Local primary membrane stress l P3 = Primary bending stress
= Design stress intensity value
[ S. . S, = Yield strength 4 S, = Ultimate strength The criteria for normal and upset loads are the ASME level A & B allowables for the range of primary plus secondary stress. The requirements of Subsection NB-3222 and NB-3223 j are met as follows:
- Range of (P, + P3 + Q) < 3S, @ operating temperature
! I where: Q = Secondary Stress l l
BWI-222-7693-LR-01 Pege 143 i l l For pressure boundary components and tubing, it is also shown that the cumulative fatigue
, usage factor remains below 1.0 for all Level A, B, and test condition operating cycles.
i ! The criterion for level C loading conditions is to maintain integrity of tube, tube supports . l , (lattice grid) and steam drum internals for emergency conditions of level C. The {
- requirements of Subsection NB-3224 are met as follows
l l P, < greater of (1.2 S, or S,) at operating temperature ! l J ! P, + P, < greater of (1.8 S, or 1.5 S,) at operating temperature ! , The criterion for level D loading condition (combined main steam line break and design basis earthquake) is to ensure tube integrity by proving that tube rupture and leakage cannot occur. The requirements of Subsection NB-3225 are met as follows: i P, < lesser of (2.4S, or 0.7S,) at operating temperature i P, + P3 < lesser of (3.6S, or 1.05S,) at operating temperature
'Ihe requirements of Subsection NB-3226 for hydrotest are met as follows:
for P. < 0.67S, at test temperature
- P+P 3 < 1.35S, at test temperature for P, < 0.9S, at test temp. l P, + P3 < (2.15S,- 1.2P ) at test temperature 1
- 2.8.1 Tubing i
l The structural analysis demonstrates that for an instantaneous full rupture of the steam line a downstream of the steam outlet nozzle occurring during normal full power operation, the tube integrity is maintained. The structural evaluation of the tubing for level D is in i accordance with the ASME Boiler and Pressure Vessel Code Section III requirements as- 1 explained in this section. Furthermore, the tube material selection and size exceed the l strength requirements of the existing steam generators. l Comparison of the RSG Alloy 690 used in the RSGs with the typical Alloy 600 tube material shows that the RSG material strength characteristics are as good or better than ] those of the existing design. The existing steam generator has a nominal tube wall thickness of 0.043 inches, compared with the RSG nominal of 0.040 inches. The ASME III allowable stresses (in ksi) based on S., S,, and S, for Alloy 600 and 690 are: , H 4
BWI-222-7693-LR-01 , pcge 144 Allov 600 Allov 690 Temperature S. S, S, S. S, S, 70*F 23.3 35.0 80.0 26.6 40.0 85.0 (as ordered) 650*F 23.3 27.4, 80.0 26.6 35.2 85.0 (as ordered) , where: S, = Ultimate Strength Pressure-induced stress in the tubing is calculated as a, = PR/t. where: a, = Hoop stress t = Tube wall thickness P = Pressure Ri = Tube inside radius A stress margin, defined as the ratio of the ASME III allowable stress to the actual pressure-induced stress is expressed as S/a,. where: S = Allowable stress A comparison of the stress margins of the RSG and OSG is expressed as a stress margin ) ratio defined as: Stress Margin Ratio = (S/a,)(a30) , (S/a,)(a30)
=
t(aso) X Sc aso) x R;(oso3 t<oso) x S(oso3 x R ic asa) A stress margin value of 1.0 or greater indicates that the RSG tubing has strength ,
BWI-222-7693-LR-01 P:gs 145 characteristics with respect to pressure-induced stresses that are equal to or better than those of the OSG. ., For S , @ 70*F or 650*F: ; Stress Margin Ratio = 0.040 x 266 6 x Q.J)31 = 1.16 0.043 23.3 0.304 : For S, @ 70'F: Stress Margin Ratio = Q_04Q x 40.0 x Q)l2 = 1.16 0.043 35.0 0.304 For S, @ 650*F: Stress Margin Ratio = Q_Q4Q x 312 x 0 332 = 1.31 0.043 27.4 0.304
. For S, @ 70*F or 650*F:
Stress Margin Ratio = Q.Q4Q x $1 x QJ32 ='1.08 0.043 80 0.304 l d. 1 3 s 4 l - l l I 0 l
BWI-222-7693 LR-01 Page 146 2.9 STARTUP TESTING REQUIREMENTS After installation, tests will be performed to verify that the replacement steam generators i comply with the requirements of the equipment specification and are capable of l satisfactorily performing their intended function. These tests form the basis for determining l compliance with the terms of the replacement steam generator contract performance l requirements. The following parameters are to be veri 5ed by start-up testing: 1
- 1. Thermal and Hydraulic Performance.
- 2. Moisture Carryover Testing.
l l
- 3. Reactor Coolant System Flow Rate and Pressure Drop Measurements. l
- 4. Primary to Secondary Leakage.
J Only minor differences are expected in shrink and swell of the RSG as compared to the l OSG. Therefore, the effect on the feedwater control system will be minor and no specific ! tests, beyond those required to verify warranted performance values, are required. DPCwill also perform additional tests as required by the ASME Code Section XI and to monitor startup performance. j I i e I i l \
BWI-222-7693-LR-01 Page 147
- 3. REPLACEMENT STEAM GENERATOR FABRICATION 3.1 QUALITY ASSURANCE PROGRAM This section describes the BWI Quality Assurance (QA) program and controls applied during RSG design and construction. The BWI QA program is implemented by the-
" Quality Assurance Manual for Nuclear Products" and by supporting procedures and instructions that govern the design and construction of nuclear steam generators and other components. The program conforms to the requirements of ASME Section III; the applicable sections of ASME NOA-1 invoked by Section III; 10CFR50, Appendix B; 10CFR21 and other international codes and standards. BWI holds ASME certificates of authorization for N, NA and NPT symbol stamps. BWI obtained ASME certification initially in 1986 and has successfully passed ASME surveys in 1988 and 1991.
3.1.1 Design Control BWI has established measures to ensure that applicable code and regulatory requirements and the owners' design specifications are correctly translated into BWI design documents (design analysis, design reports, drawings, etc.) and that the design documents are verified against the design speci6 cations. Design control measures include:
- 1. Assignment of a cognizant project engineer responsible for coordinating design document preparation.
- 2. Review for suitability and application of design methods, materials, parts, equipment and processes essential to RSG safety and performance functions.
i 3. Planned and controlled design analysis, using appropriate analysis plans, outlines or drawings; material specifications and review schedules, based on the complexity of
- the analysis and previous experience.
- 4. Legible documentation suitable for reproduction, filing, and retrieval, covering design analysis and required ASME Code design reports. Typical design report contents
. include:
- a. definition of objective, design inputs and their sources
- b. results of literature searches and other background data . c. identification of assumptions and indication of those that must be verified as
- the design proceeds
- d. the main body section of calculations including Acceptance Criteria, Loads, Material Properties, Boundary Conditions, Model Description, Computer i
l
- . - . . . - - . - - - - . . -- . _ . - - - - . _ _ - ~ -
1 BWI-222-7693 LR-01 Page 148 input, Analysis and Results as appropriate for the calculation. 1 ( e. identification of any computer type, computer program name, evidence of/or
- reference to computer program verification and the basis supporting the application of the computer program to the speci6c physical problem.
l f. review and approval by personnel other than the preparer.
- g. Design verification by a qualified engineer not responsible for the design.
! Verification may be by means of design review, alternate calculations or
- quali6 cation tests.
- h. Formal documented design reviews of first-of-a-kind features or features that are major extrapolations of BWI designs. These are conducted by
- experienced BWI or outside engineers not involved in the design process.
- i. Control of changes to design documents by the same controls used on the original documents, including necessary reviews' and approvals. Design changes are controlled by means of revisions to the original documents. ;
4 The as-built conditions are formally reconciled with the 'ASME Code Design
; Report by the project engineer.
- j. Identi6 cation and control of design interfaces in accordance with documented i
procedures. Information transmitted across an interface is controlled and documented with regard to the information transmitted and its status. 3.1.2 Document Control Documents that specify quality requirements or describe activities affecting quality (such as j QA program procedures, inspection and non-destructive examination procedures, inspection !' ' & test plans, manufacturing procedures, welding & heat treatment and other special process procedures, material ordering standards, drawings) are issued under a formal Document . Control system which ensures that all documents and revisions are reviewed for adequacy j and released by authorized personnel prior to use. ! 3.1.3 Corrective Action i Conditions adverse to quality detected during audits, inspections or other activities are i addressed under a formal corrective action program. The program requires for signi6 cant i conditions that the cause of the condition be determined and corrective action be taken to j preclude recurrence. [ The program documents the responsibilities for initiation, evaluation and acceptanc'e of corrective actions. It establishes time limits for these activities to ensure timely corrective action. . l l
BWI-222-7693-LR-01 . pcg3149 i Periodic reports are provided to BWI senior management documenting each condition
- adverse to quality, its cause and corrective action (s) taken.
3.1.4 Non-conforming Items l A program has been established to detect non-conformances to drawings and speci6 cations l l to prevent unauthorized use or shipment. Non-conforming items are segregated from the normal production flow. Further processing of non-conforming items is controlled, pending evaluation and disposition by authorized personnel. , The program defines the responsibility and authority of personnel responsible for the , disposition of non-conformances. Design Engineering, Manufacturing Engineering and j Quality Assurance are involved in the disposition of all non-conformances. Suppher non- { ! conformances are handled under the same program. ! 3.1.5 - QA Records l
- QA Records required by the applicable codes and standards, owner's speci6 cation or for j other reasons are generated, supplied and maintained under a formal records program. The i program ensures that records are legible, accurate, accepted by authorized personnel,
identifiable to the item or activity to which they apply and retrievable. 4 For each contract a Documentation Checklist identifies the records required, their classification, the personnel responsible for obtaining and storing them, and record retention i j and distribution requirements. i 3.1.6 Audits l
- Planned and periodic audits are conducted to verify compliance to and the effectiveness of l 3
the QA program. Audits are conducted by QA personnel quali6ed to the requirements of ASME NOA-1. Audits are performed in accordance with written procedures and checklists l j by personnel not having direct responsibilities for the areas being audited. Results are documented in audit reports that define audit scope, identify auditors, identify persons contacted, summarize audit results, assess program element effectiveness, and describe i findings. Audit reports are forwarded to management personnel including the Division General Manager. The QA Manager is responsible for the initiation of corrective action ) requests and other actions based on the audit findings. i Follow-up action including re-audit of deficient areas is performed as appropriate. Resolution of audit report findings are documented. l l i l
i BWI-222-7693-LR-01 Page 150 ' l 3.2 FABRICATION CONTROL ! l This section describes the measures used to control fabrication of the RSGs from purchase of materials through shipping to the owner. 3.2.1 Control of Purchased Items and Services ! I Procurement documents (material ordering standards, purchase orders, etc.) are prepared ! under a controlled program which ensures that the requirements of the design basis ! ! documents, specification requirements and applicable codes and standards are included and ! that procurement documents and revisions are reviewed and accepted by cognizant l ) personnel before use. ! 3 Purchased items and services are procured from vendors who meet applicable quality program standards. Vendor evaluation is performed by the QA Vendor Control Group
- based on reviews of vendor quality program manuals and audits of program implementation.
An " Approved Supplier List" is published and issued as a controlled document. It lists ,
- ' acceptable vendors and their approval status. ,
- i e Vendor performance is verified by a combination of surveillance, source inspection and ,
incoming inspection. The methods used to verify performance are selected and documented ! on BWI and vendor inspection and test plans. The vendor is required to submit inspection j and test plans, and manufacturing, inspection and test procedures for review and approval by cognizant BWI personnel. ! A historical file is maintaiced for each vendor. It contains survey and audit reports, source -
- inspection reports, non-conformance reports and other documentation relative to the vendor's items and services. Vendor performance is assessed at least annually and approved l supplier status revoked if quality is not acceptable.
3 BWI has implemented a Commercial Grade Dedication procedure to allow materials for i safety related components to be procured without imposing 10CFR21 and 10CFR50, Appendix B. This procedure is modelled after EPRI guideline NP-5652. In addition, procedures indicate that for safety related items, either 10CFR21 be imposed or commercial
- grade dedication be performed.
i All items and services for steam generator construction are subject to receipt inspection for conformance to procurement documents. Acceptance is documented to ensure that only j conforming items are used for construction. 3.2.2 Control of Manufacturing Processes Control of quality related manufacturing processes to ensure performance in accordance with documented procedures, instructions and drawings is achieved through a shop traveller (Route Sheet) system. The Route Sheet controls and documents the status of shop - operations and performs.the following functions: 4 e
BWI-222-7693-LR-01 Page 151
- 1. Lists the sequence of operations.
- 2. Describes each operation.
- 3. Identifies drawings and procedures & instructions to be followed with revision levels.
- 4. Provides space for indicating inspection, witness, documentation and hold points.
- 5. Provides space for sign-off of completion of fabrication operations and inspection points.
- 6. Documents the fabrication history of the product.
Input for the preparation of Route Sheets is obtained from Fabrication Outlines, Inspection
& Test Plans, drawings and lists of weld procedures prepared by cognizant engineers.
Route Sheets and revisions are reviewed by cognizant engineers before issue. Special processes such as welding, non-destructive examination and heat treatment are performed in accordance with documented procedures developed by technical specialists. Procedures and personnel are qualified as required by applicable codes and standards. 3.2.3 Control of Consumables Consumable products are nonmetallic, non-permanent products which come into contact with the RSGs during manufacture, inspection or testing. Because these products may contain materials that could be detrimental to the RSGs, the use of consumable products : is controlled. Limits are placed on the amounts of certain materials that may be present in consumable materials. Consumable materials include: I ! 1. Cleaning solvents and agents. 1 i I
- 2. Non-destructive testing compounds and agents.
- 3. Adhesives and adhesive tapes.
- 4. Insulation and refractory materials.
! 5. Cutting, drilling and tapping compounds.
- 6. Other consumables that are capable of transferring detrimental materials to nuclear hardware.
~ ~
All consumable products are legibly labelled with the product and manufacturer. Three levels of control are established for consumable materials: I
1 BWI-222-7693-LR-01 Page 152 l 1
- 1. Prohibited Products - These are materials that are not allowed to contact nuclear i hardware, because they contain elements and/or compounds known to be detrimental. Typical examples of prohibited materials are:
- a. Lead and lead based alloys.
- b. Copper and alloys containing more than 50% copper.
- c. High sulphur compounds, especially molybdenum disulphide.
- d. Alloys based on, or containing significant amounts of, cadmium, mercury, arsenic, zinc, antimony, bismuth and tin.
- e. Halogenated solvent, aerosol propellants or similar highly halogenated compounds.
- 2. Acceptable Products - shown to contain low levels of elements known to have deleterious effects on nuclear materials, especially nickel-based alloys and stainless steels. These elements include chlorides, fluorides, sulphates, mercury, lead, antimony, bismuth, copper, zinc, tin, arsenic and cadmium.
l Records.are kept on all acceptable products. These include the manufacturer's speci6 cations, certificates of a'nalysis, identification of low melting point constituents (where applicable) and special use restrictions (e.g. "must be removed if temperature i exceeds 200F"). These records also identify any special cleaning procedures that may be required to remove the material.
- Only items on the " Acceptable Products List" may be used in contact with corrosion
- resistant materials (nickel-based alloys and stainless steels) during RSG assembly.
l Only items on this list are allowed contact with final cleaned surfaces, or during l processes involving elevated temperatures (welding or heat treatment). i 3. Controlled Products - products that contain (or may contain) potentially detrimental materials in excess of the amounts allowed in the Acceptable Products List. Use of 3 controlled products is restricted to applications where there is either no transfer of the potentially detrimental material to nuclear materials, or where the potentially ! detrimental material can be removed and the surface condition can be verified. An l example of this latter condition is use of high sulphur cutting fluid. After machining, l the fluid is removed and verified to have been removed. Controlled products are discussed in Section 3.2.5. 3.2.4 Control of Specialized Processes l The following paragraphs describe control of specialized RSG manufacturing processes. l , l i l l l l
l BWI-222-7693-LR-01 Page 153 3.2.4.1 Tube-to-Tubesheet Welding Each heat of welding wire is tested for weldability before use. Each operator makes a test tube-to-tubesheet weld each shift. These are sectioned and examined to ensure a : satisfactory weld has been made. Should a test weld prove unsatisfactory, welding is halted and all welds made by the operator prior to the stop are subject to a non-conformance report for evaluation. The origin of the problem is determined and corrected, and another test weld is made by the operator before he resumes welding. The second test weld is examined in the same manner as the first. Sectioning and polishing equipment, and a metallurgical microscope are dedicated to this examination in the clean room. Each completed tube-to-tubesheet weld is visually and dye penetrant examined. 3.2.4.2 Hydraulic Tube Expansion Close control of the hydraulic expansion process is maintained throughout the operation. Detailed instructions are prepared and operators are trained to use the process before working on the steam generator. Quality control checks are made on all critical parameters, including:
- 1. Pressure measured at the expansion mandrel.
- 2. Thne of applied pressure.
- 3. Position of the hydraulic seal at the secondary face of the tubesheet.
i
- 4. Verification that all tubes have been expanded. !
3.2.4.3 Electro-polishing
'Ihe electro-polishing (EP) process, used to improve channel head surfaces, is quali6ed by
- performing the process steps on sample specimens and microscopically evaluating the l resulting surface finish. Qualification ensures that the entire EP process does not degrade
! the RSG primary side surfaces. The surfaces poli'shed include primary head, stay cylinder i and nozzle stainless steel weld overlay, and stainless steel divider plate material. Because
- of the complex geometry, the tubesheet cladding and tube-to-tubesheet welds are not electropolished.
l j' Qualification involves electro-chemical polishing of samples representing the primary side surface materials, using the electrolyte, polishing equipment and electrical parameters
- proposed for the RSG. Procedures define prerequisite operations, precautions to be taken,
- and the mechanical polishing parameters to prepve the surface. They list materials and
- chemicals that can be used, specify cladding thickness requirements, and specify cleanliness l requirements. During qualification, measurements of the treated surface are made. These include scanning electrpn microscopy to characterize surface pro 61ometry, amount of l
l l
BWI-222-7693-LR-01 Page 154 ' cladding removal, and dye penetrant testing to detect excessive metal removal or surface
, finish problems prior to production EP. These examinations ensure that the EP process will
- . not result in any degradation of the primary side surfaces.
l 3.2.5 Material Control. 1 l l , Measures are established to identify and control materials, parts, and . components to ensure l
- that only the correct materials are used and that proper records are maintained from initial receipt of the material through shipment of the finished component. Identification is ;
i maintained either on the component or on documentation traceable to the component. i BWI verifies material identification prior to shipment under the material control system. As required by the ASME Code or owner's Design Specification, material control measures ! ensure that materials'are traceable by heat and lot number, or by other appropriate means, l to the Material Test Reports.
- 3.2.6 SHOP TESTS AND INSPECTIONS The following paragraphs summarize shop tests and inspection requirements applied to key
- RSG components and tooling.
3.2.6.1 Test and Inspection Equipment f
- Tools, gauges and other measuring and test equipment used for activities affecting quality l are controlled to assure their calibration and adjustment to maintain accuracy within I
acceptable limits. Measuring and test equipment is calibrated by comparison to certified standards which are traceable to National Standards. Equipment found to be out of calibration tolerances are physically segregated until repairs are made. Equipment beyond j repair is replaced. Documented procedures establish the responsibility, calibration methods, frequency, and l notification requirements for calibration and the requirements for handling discrepant j equipment including validation of items checked with equipment. l Subcontracted calibration is performed by approved suppliers. Supplier approval is i described in Section 3.2.1. 3.2.6.2 Tests and Inspections of Forgings - Forgings used for steam generator pressure boundary components are examined and tested in accordance with ASME Section III requirements. Additional requirements are imposed by BWI on critical forgings such as tubesheets. These include restrictive chemistry requirements (Sulphur, Phosphorous, etc.) and additional ultrasonic and magnetic particle examination requirements. Their purpose is to ensure that critical forgings (primarily tubesheets) are free from inclusions or defects which could affect the structure, cladding, 4
- tube-to-tubesheet welds, or welds in highly stressed areas of the tubesheet and lead to in-
BWI-222-7693-LR-01 Page 155 service problems. 3.2.6.3 Tests and Inspection of Tubing Tubing quality is critical to long-term steam generator p rformance and integrity. For this reason, BWI tubing requirements significantly exceed A1 1E Code and industry standards. Tubing is procured to the requirements of ASME Section 111 and EPRI NP 6743-L, Volume 2 guidelines. BWI chemistry requirements are more restrictive than those required by either ASME or EPRI. Special requirements are placed on content of iron, carbon, sulfur, chromium and cobalt. In addition, BWIimposes requirements for chemistry, nondestructive examination (multi-directional UT with both outside- and inside diameter calibration notches), and rejectable defect size (max. 0.002 inches). Prior to manufacturing tubes for an order, the tubing vendor must qualify the manufacturing ;
-process and inspection techniques on a pre-production tubing lot. These tubes are examined by BWI using enhanced NDE techniques and destructive examination to assess 1 whether they meet BWI standards.
l At all times during tube manufacture, the vendor's processes are monitored by resident BWI inspectors. In addition a statistical sample of tubes from each lot are subject to enhanced NDE and destructive examination by BWI personnel to ensure that tubing quality is maintained. 3.2.6.4 Welds All pressure boundary welds are examined to ASME Code requirements using trained and qualified personnel. In addition significantly more stringent requirements are imposed on welds critical to long-term integrity and performance. Tubesheet overlay cladding is ultrasonically inspected to an acceptance standard that is more stringent than that required by Section V of the ASME Code. Tube-to-tubesheet welds are required to pass a no-indication acceptance standard for liquid penetrant examination. 3.2.6.5 Steam Generators l Nuclear steam generators are tested and examined in accordance with ASME Section III ~ requirements with additional requirements based on the experience of BWI and associated B&W divisions. 3.2.6.6 Baseline Eddy Current Inspection Eddy current inspection is performed on the RSG tubing prior to operation to document tube condition and to form a baseline for comparison with future (inservice) tube inspections. Important features of the RSG eddy current inspection are:
- 1. Each tube is inspected end to end with an internal bobbin probe prior to installation -
in the steam generator and after fabrication is complete.
BWI-222-7693-LR-01 l Page 156 l l ! 2. The inspection provides a complete cross section capable of showing flaw indications g ; 1 and wall thinning. All indications are reported and dispositioned as either being l acceptable or requiring removal, replacement or plugging.
- 3. A pro 61ometry inspection of each tube is made through the length of the tubesheet to assure proper and complete expansion. 1
- 4. Data is collected and stored on optical disk for future reference. <
He eddy current inspection equipment used for the baseline inspection is the MIZ 18/30 Eddy Net Acquisition and Analysis Systems. This equipment provides examination reliability in the presence of extraneous test variables and greater flexibility in data ' manipulation to provide thorough signal detection and analysis capability. Additional specialized equipment (such as Motorized Rotating Pancake Coil (MRPC)) is available to : inspect areas of special interest. Inspection techniques, personnel qualification and procedures are prepared using the guidelines identified in the EPRI Report Summary NP-6201, the ASME Sections V and XI, and NRC Regulatory Guide 1.83. 3.2.7 Handling, Storage and Shipping Detailed BWI procedures ensure that the handiing, storage, cleaning, packaging, shipping and preservation of items are controlled to prevent damage or loss and to minimize deterioration. Cognizant engineers provide drawings and instructions for critical operations. Important steps for handling, storage and shipping are indicated in the inspection and test plan. These activities are inspected and documented. 3.2.7.1 Cleanliness BWI combines a cleanliness policy, cleanliness procedures, assembly in a nuclear clean room, and a consumables control policy to ensure that RSGs are clean and free from contamination when shipped. The BWI cleanliness policy guides overall conduct of fabrication and material handling activities and maintains awareness of the importance of cleanliness. Procedures ensure all cquipment remains free of debris and potentially deleterious materials. The clean room is used exclusively for assembly of nuclear steam generators and similar equipment. Initial heavy fabrication operations, such as the welding of shells, tubesheets and heads is carried out in other areas of the BWI plant. BWI also implements a consumables control policy to ensure that expendable materials utilized during the manufacturing and assembly processes do not contaminate primary or secondary wetted surfaces. These measures ensure that the RSG meets the NRC and customer standards. He following paragraphs detail BWI cleanliness procedures, clean room and consumable material control.
- -, \
- -- _- -. .. . -= __- - - . . - _ - - -- -
BWI-222-7693-LR-01 Page 157 1 l 3.2.7.1.1 Procedures Cleanliness procedures ensure that debris and foreign material are excluded from the RSGs during assembly. These procedures were developed in conjunction with the manufacturing procedures and maintain and verify cleanliness during manufacturing, assembly and testing operations. Cleaning and cleanliness inspection points are incorporated into the shop { routing instructions and are used, with the appropriate shop instruction sheets, to ensure j that the components are clean prior to and during assembly and ensure that the final RSG
- meets cleanliness criteria.
De requirements of ANSI N45.2-1 (1980 edition) and NRC Regulatory Guide 1.37 are ! included in the cleanliness procedures. These procedures are designed to obtain N45.2-1 cleanliness level B for the primary (or tube) side and level C for the steam (or shell) side
- of the RSGs. Cleanliness inspections are conducted at critical stages of fabrication and l assembly including-i
- 1. After cleaning operations. I l
- 2. Prior to operations involving elevated temperatures (pre-heating before welding, and
! post-weld heat treatment). I
- 3. Prior to assembly operations, especially any operation which will result in a loss of access (fitting the lattice support grids into the shell assembly).
- 4. After completion of final assembly and prior to the sealing of openings in j preparation for shipment. !
Implementation of cleanliness procedures ensures that loss of cleanliness is rare, but would l be detected. The Quality Assurance Manual provides procedures to detect and rectify such a situation, and to reverify that the required degree of cleanliness is reestablished.
- Full accountability is maintained of all tools and loose parts used during assembly ~. This
- includes personal effects such as eyeglasses. Additionally, all hand tools, including .
- electrically or air powered tools, are maintained in a clean condition. This ensures that dirt, j debris, oil, etc. are not transferred from the tools to the RSG.
i 3.2.7.1.2 Clean Room Before the RSG or sub assembly is moved to the clean room, it is cleaned and inspected i to ensure that debris and foreign materials are not transferred into the clean room. Assembly operations, such as installation of the shroud and tube support grids, installation i . of tubes, and the tube-to-tubesheet welding are performed in the clean room. Filtered and i heated air is provided to the clean room to maintain a positive pressure relative to outside i ambient conditions. His prevents ingress of contaminants.
- Internal combustion engines are not permitted inside the clean room, eliminating the
! BW1-222-7693-LR-01 Page 158 L. potential for oil fumes, lead and other materials. Most welding operations are performed , before components are moved into the clean room. Tube-to-tubesheet welding is a Gas Tungsten Arc Welding (GTAW) process which generates a minimum amount of fumes. It is carried out within sealed and air-conditioned enclosures. Ifit is necessary to use another -< I welding process, such as Shielded Metal Arc Welding (SMAW), adequate precautions are ! taken, using temporary enclosures, fume hoods and extractors to prevent significant release of weld fume into the atmosphere. Clean-up procedures ensure that slag and debris are l contained and removed after completion of the welding operation. 1 The clean room is equipped with its own laboratory facility to monitor activities such as tube-to-tubesheet welding. It also has a dedicated document control center and tool crib ; t for storage of hand tools and consumables. The operation of these facilities is governed by
- procedures which are compatible with the clean room cleanliness requirements.
3.2.8 Receipt Inspection Requirements j 4 Packaging, shipping, receiving, storage and handling are in accordance with standardized
- . procedures which meet the requirements of ANSI N45.2.2-1972 as supplemented by
- Regulatory Guide 1.38 and customer specifications. The packaging procedure considers the 4
method of transportation and handling as well as possible storage environment. i
- 3.2.8.1 Preparation for Shipment Prior to shipment, the RSG tube and shell sides are cleaned. A foreign object inspection i is performed just prior to final closure of all openings. Equipment is stored, inspected, handled, installed, and cleaned by methods which ensure that harmful contaminants do not i remain on any component surface in contact with process Duids. Protection of internal cleanliness is achieved by sealing all openings with plugs, caps, or covers. All threaded '
plugs used to seal auxiliary nozzles are removed after site installation. These items are also
- protected to preclude damage that could result in loss of the nitrogen blanket or
~
contamination of RSG internal surfaces. Covers are designed and installed for removal
'without damaging the vessel or pipe nozzle weld preparation. The primary nozzle covers
- installed for the shop hydrostatic test are left in place for shipping.
RSG internal surfaces are required to meet the following criteria prior to shipment: 3
- 1. Surfaces having free access must pass visual, wipe test, leach sample, and rust examinations. Visu al techniques include boroscopes, mirrors, supplementary lighting, p or other aids when needed to properly examine hard-to-see surfaces:
- a. The surface must appear" metal clean" when examined without magni 6 cation
- under a lighting level of at least 100 foot-candles.
- b. The surface must be free of particulate contaminants such as sand, packing l
< materials, sawdust, metal chips, wire, weld spatter, tape, and tape residue. , I
* * . i f
13WI-222-7693-LR Page 159
- c. The surface must have no evidence of organic material or films such as oil, grease, paint, crayon, moisture, chemical residue, or preservatives. In addition to visual examination, the surface is wiped with a solvent-dampened, white, lint-free cloth,' using a clean portion for each wipe. A visible discoloration of the cloth is unacceptable unless it is established that the deposit is not detrimental.
- 2. If visual examination is not possible, but the surface is accessilile, inspection consists )
of wiping the surface with a dry, white, lint-free cloth. Visual discoloration on the l cloth is unacceptable unless it is established that the deposit is not detrimental.
- 3. The cause of rust shall be determined to prevent recurrence. ':
The RSG primary and secondary sides are drained and dried immediately after hydrotesting and cleanliness inspection, and are evacuated to eliminate residual moisture (dew point :s; - l 20*F). Each RSG is sealed and pressurized on both sides with dry nitrogen to a pressure l between 5 and 10 psig. The nitrogen is maintained at this pressure for shipping. l Each RSG is shipped with a connected nitrogen supply. Redundant pressure gauges are : used to indicate the nitrogen pressure in each circuit. Valved connections allow adding i nitrogen as necessary. Calibration requirements and gauge ranges to nionitor nitrogen , pressure, the nitrogen addition procedure for supplied valving, and cleaning controls for ; caps are also provided to help assure that nitrogen blanket is properly maintained. , 3.2.8.2 Handling and Shipping The RSG is designed to withstand the associated loads, including lifting and upending, and environments without damage during shipping, installation, and handling. Appropriate instruments on the carrier are provided to monitor and record vibration and shock to which the RSG is subjected during shipment. Continuously recording accelerometers are installed to measure accelerations in all three directions during transit. A report characterizing the loads and effects on the shipment is prepared and submitted to the owner.' The weight, center of gravity, and lifting points for all handling procedures are provided. ' Limitations imposed when the RSG is lifted or moved, including maximum allowable three-dimensional accelerations, including maximum internal ambient temperatures and pressures, j during shipping are also provided. These are included in the equipment Operation and Maintenance Manual (O&M Manual) which is provided to the customer. 3.2.8.3 Inspection at Jobsite 4 1
'Upon receipt of a steam generator at the jobsite, the surface shall be inspected by the '
owner to ensure that no damage has occurred. Nozzle caps are inspected for shipping damage and to ensure that a positive nitrogen pressure has been maintained on both the primary and secondary sides. i
BWI-222-7693-LR-01 Page 160 3.2.8.4 Storage Each RSG is prepared for long-term storage prior to shipment. Exterior surfaces are l protected against rust. Interior surfaces are protected against oxidation or corrosive attack by an inert dry nitrogen gas blanket. These storage provisions should be maintained at the jobsite until the RSG is installed . I 9 5 A
-d , o j , 5 Submittal ofNPDES Permit Renewal Application September 14,1995
- RBG-41958 Page 2 of 2 cc: Mr. J. Dale Givens Louisiana Department of Environmental Quality Office of Water Resources P. O. Box 82215 Baton Rouge, LA 70884-2215 U.S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 U.S. Nuclear Regulatory Commission Document Control Desk
' M/C PI-37
-Washington, DC 20555
'RC Sr. Resident Inspec'or Box 1051 ;
'ncisville, LA 70775
' man, Gen Mgr, Plant Operations, EOI I
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Submittal ofNPDES Permit Renewal Application September 14,1995 RBG-41958 Page 2 of 2 cc: Mr. J. Dale Givens Louisiana Department of Environmental Quality Of5cc of Water Resources P. O. Box 82215 Baton Rouge, LA 70884-2215 U.S. Nuclear Regulatory Commission RegionIV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 U.S. Nuclear Regulatory Commission Document Control Desk M/C PI-37 Washington, DC 20555 NRC Sr. Resident Inspector P.O. Box 1051 St. Francisville, LA 70775 Mike Sellman, Gen Mgr, Plant Operations, EOI O A.__-_._____.________.-. e- - - _ _ _ -__
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u . y if . APPLICATIONLFOR. RENEWAL OF NPDES
. PERMIT NO.TLA0042731
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[ , . [ [ [ . ENTERGY OPERATIONS, INC. . RIVER BEND STATION [ , [ APPLICATION FOR RENEWAL OF NPDES PERMIT NO. LA0042731 ; [ [ i l [ SEPTEMBER 1995 [ , i PREPARED BY: { C-K ASSOCIATES, INC. [ 17170 PERKINS ROAD BATON ROUGE, LOUISIANA 70810 l (504) 755-1000 C-K ASSOCIATES' PROJECT NO. 53-502 [- [ [
- 5312R01.DP
TABLE OF CONTENTS Section Page No. l
)
1.0 INTRODUCTION
................................. 1 l i
1.1 G eneral . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 l 1.2 Regulatory and Permitting Background . . . . . . . . . . . . . . . . . 1 2.0 SITE OPERATIONS AND PROPERTY DESCRIPTION . . . . . . . . . 2 3.0 EFFLUENT COLLECTION, TREATAIENT, AND DISCII A RG E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3.1 Permitted Outfalls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 l 3.2 Ancillary Water Systems . . . . . . . . . . . . . . . . . . . . . . . . . . 9 I i 4.0 STORh! WATER DRAINAGE, AIANAGEMENT, ' AND DISCII ARG E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 5.0 WASTEWATER AND STORA1 WATER SAMPLING AND ANALYTICAL CONSIDERATIONS . . . . . . . . . . . . . . . . . . 14 i 6.0 SUMA1ARY OF PRIOR BIOMONITORING REQUIREMENTS I A ND R ES U LTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 LIST OF TABLES Table 1 Inventory of Water Wells in the Vicinity of the Site 2 Inventory of Significant Materials in Storage and Unloading Areas 3 Inventory of Significant Materials in Oil Storage Areas 4 Analytical Data Summary for Outfall 102 5 DMR Summary for February 1993 - January 1995, Outfall 003 6 DMR Summary for February 1993 - January 1995, Outfall 005 7 DMR Summary for February 1993 - January 1995, Outfall 006 8 DMR Summary for February 1993 - January 1995, Outfall 007 9 DMR Summary for February 1993 - January 1995, Outfall 009 l 10 Biomonitoring Test Results 5312R01.DP i I
L l < i LIST OF FIGURES : r i L 1 1 Figure 1 Site Location Map I 2 Site Plan and Stormwater Drainage Map { 3 Station Water Flows 4 Surface Types of Stormwater Drainage Areas { (- LIST OF APPENDICES I Appendix A U.S. EPA Application Form 1 B U.S. EPA Application Form 2C C U.S. EPA Application Form 2F i D May 23,1995 Letter from LDEQ on Biomonitoring Testing i l [ [ l ( ( ( '
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l
1.0 INTRODUCTION
( 1.1 General l J The Entergy Operations, Inc. (Entergy), River Bend Station currently discharges under authority of National Pollutant Discharge Elimination System (NPDES) I Permit No. LA0042731 and Louisiana Water Discharge Permit System (LWDPS) Permit No. WP 0409. The NPDES and LWDPS permits were issued by the U.S. Environmental Protection Agency (U.S. EPA) and the Louisiana Department of Environmental Quality (LDEQ), respectively, and authorize discharge of facility wastewater /stormwater from nine final outfalls (001 - 009) and one internal outfall (102) to the Mississippi River. The following information and U.S. EPA Application Form 1 (General Information), Form 2C (Wastewater Discharge Information), and Form 2F (Storm l Water Discharges Associated with Industrial Activity) are being submitted in l connection with the renewal of the site's existing NPDES permit. . I U.S. EPA Forms 1, 2C, and 2F are included as Appendices A, B, and C, respectively. Section 2.0 provides a description of the site operations, location, and property boundaries. Effluent collection, treatment, and discharge are addressed in Section 3.0. Stormwater drainage, management, and discharge are discussed in Section 4.0. Section 5.0 includes pertinent information on wastewater and stormwater sampling and analyses conducted for this permit application. Section 6.0 addresses prior biomonitoring requirements and results. ) [ 1.2 Regulatory and Permitting Background The operation of the Entergy River Bend Station and the discharge of treated wastewater and stormwater are regulated by the NPDES permit program administered by the U.S. EPA. The regulations for the NPDES program as they apply to the Entergy facility are set forth in Title 40 of the Code of Federal Regulations (CFR), specifically at 40 CFR Parts 122,124,125,129, and 136. l Additionally, the State of Louisiana Water Quality Regulations [ Louisiana Administrative Code (LAC) at Title 33, Part IX, Chapters 1 through 15] as administered by the LDEQ apply to the Entergy facility. The establishment of l efnuent limitations in Entergy's NPDES and LWDPS permits is governed by the l aforementioned regulations and is based upon the effluent guidelines and standards for the Steam Electric Power Generating point source category at 40 [ CFR Part 423. 7 Entergy currently operates and discharges treated wastewater and stormwater l under the NPDES permit that was issued on February 15,1991 (effective March 16,1991 and expiring on March 15, 1996). In accordance with requirements at 40 CFR 6122.21(d)(2), this NPDES permit renewal application is being submitted at least 180 days prior to the expiration date of the currently effective permit. 5312R01.DP I b
Entergy currently operates under the LWDPS permit issued on May 28,1987 (effective on date of issuaned, with an expiration date specified as five years from the date ofissuance (May 27,1992). The LWDPS permit was subsequently modified on May 23,1991 as requested by Entergy. An LWDPS permit renewal application was submitted on January 27, 1992. Although the LWDPS permit does not presently include Outfall 009, Entergy did address this outfall in the 1992 permit renewal application. The LWDPS permit is currently in the process , of being reissued by the LDEQ and will be updated to reflect current site conditions (i.e., outfalls). 2.0 SITE OPERATIONS AND PROPERTY DESCRIPTION The Entergy River Bend site is a nuclear fuel steam generation facility, Standard Industrial Classification (SIC) Code Number 4911. The site received a full-power license from the U.S. Nuclear Regulatory Commission (NRC) on November 20,1985 and achieved commercial operation on June 16, 1986. The facility's generating capacity is - 934 megrmaits GCT) electri.:al. The commercial generation of electricity is provided by a General Electric BWR-6 reactor with Mark III containment. 1 The Entergy facility is located at 5485 U.S. Highway 61 in St. Francisville (West Feliciana Parish), Louisiana. It is situated on approximately 3,800 acres in Section 48, Township 3 South, Range 3 West; Sections 41,44,45,57,58,59,60,62,63, and 65, Township 3 South, Range 2 West; and Sections 45 and 66, Township 4 South, Range 2 West. Figure 1 is a Site Location Map showing the setting of the River Bend Station and the location of the designated discharge outfalls. Approximately 132 acres of the property have been developed for steam electric power generating activities. The facility is located between U.S. Highway 61 (on the northeast) and the east bank (left descending I bank) of the Mississippi River near River Mile 262. The northwest and southeast boundaries adjoin undeveloped land. The developed portion of the plant site has a topography with an average elevation of approximately 100 feet National Geodetic Vertical Datum (NGVD). Rolling hills occupy a considerable area of the Entergy property surrounding the developed portion, and the elevations for the entire property range from approximately 35 to 130 feet NGVD. The locations of permitted final Outfalls 001 - 009 and internal Outfall 102 are shown on Figure 1. Also shown on Figure 1 are the locations of active water wells in the near vicinity (one-mile radius) of the Entergy site that are registered with the louisiana Department of Transportation and Development (LDOTD), Office of Public Works. Wells shown include those used for industrial, domestic, fire suppression, and power generation purposes. Plugged wells, monitor wells, test holes, piezometers, observation wells, and recovery wells are not included. Summarized in Table 1 is relevant information on each water well in the LDOTD inventory shown on Figure 1. Figure 2 is a Site Plan and Stormwater Drainage Map depicting pertinent features of the Entergy River Bend Station. 5H2R01.DP 1
i 3.0 EFFLUENT COLLECTION, TREATMENT, AND DISCIIARGE This section addresses water use and wastewater generation, collection, treatment, and discharge (including stormwater management) at the Entergy River Bend Station. While overall stormwater management and discharge are discussed in this section, specific site information required by the NPDES stormwater permit application regulations (and U.S. EPA Form 2F) is addressed in Section 4.0. j 3.1 Permitted Outfalls l Water used in the facility for cooling purposes is obtained from the Mississippi River via a single intake structure. It is clarified before use in the cooling towers. Water used in the facility for potable, sanitary, fire suppression, process, and auxiliary boiler feed purposes is obtained from four on-site wells, the locations of which are shown on Figure 1. Some well water is treated by a reverse osmosis process (ion exchange) for plant use. Figure 3 depicts Station Water Flows. Components of each outfall and wastewater treatment, as j applicable, are described below. Outfnli 001 This is the River Bend Station's main water discharge outfall to the Mississippi River (Water Quality Management Basin Segment Number 070201). It consists of cooling tower blowdown and other wastewater streams previously monitored at designated outfalls. These other outfalls include the metal cleaning wastewater discharge (Outfall 102), the low-volume chemical wastewater discharge (Outfall 002), and the treated sanitary wastewater discharge (Outfall 004). Entergy redirected the treated sanitary wastewater (Outfall 004) from discharge to Grant's Bayou to Outfall 001 during the refueling outage in March 1992. Cooled water from cooling towers is pumped through the turbine condenser and service water heat exchangers, and the heated water is returned to the cooling l towers. Four eight-cell induced draft cooling towers reject heat from the turbine I condenser, and one five-cell induced draft cooling tower rejects heat from the ) service water heat exchangers. Water losses from drift and evaporation are ) replenished with clarified river water. Clarifier sludge is diluted with river water to approximately 4 % solids and returned to the Mississippi River (via a discharge line separate from Outfall 001) as shown on Figure 3. Cooling tower blowdown is accomplished by directing cooled water from the cooling tower fiume via a I portion of the condenser pumps' discharge to a common discharge header leading to Outfall 001. This diversion of pumpage is normally valved to provide a minimum of 2,200 gallons per minute or gpm (3.17 million gallons per day or l MGD) blowdown rate. During full power, hot weather operation of River Bend Station, cooling water blowdown occurs at approximately 3,500 gpm (5.04 MGD), but may occur at rates up to 7,000 gpm (10 MGD). c , { . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
Cooling tower blowdown, metal cleaning wastewater (described in more detail for Outfall 102), low-volume chemical wastewater (described for Outfall 002), and sanitary wastewater treatment effluent (described for Outfall 004) merge into a common discharge header for conveyance to the Mississippi River via a 2.6-mile long, buried pipeline (see Figure 1). The discharge volume of Outfall 002 constitutes approximately 10% of the flow from Outfall 001 for about three hours per day and less than 2% of the flow for the remainder of the day during full power operation. The discharge volume of Outfall 004 constitutes less than 2% of the flow through Outfall 001. Residual chlorine levels are reduced by treatment with ammonium (or sodium) bisulfite injection into the combined Outfall 001 effluent downstream of the common discharge header, prior to discharge to the Mississippi River. Permit compliance monitoring is performed at the exposed vacuum-break chamber of the 30-inch diameter buried pipeline approximately 300 meters before the pipeline enters the , floodplain. This pipeline emerges on the east bank of the river in the discharge l control structure located at approximately River Mile 262. The 30-inch diameter submerged discharge is located 610 feet downstream of the plant's river water intake structure (see Figure 1). l l i Outfall 002 i This outfall is the power station low-volume chemical wastewater discharge to the cooling tower flume or to the common discharge header leading to Outfall 001 ) which discharges to the Mississippi River. It consists of the treated water and ! wastewater from the following sources: (1) intermittent ion-exchange resin backwash, regeneration, and reverse osmosis reject waters from makeup water polishing (demineralized water production); (2) intermittent auxiliary boiler blowdown; (3) intermittent metal cleaning wastewater discharge (monitored as Outfall 102); (4) intermittent reverse osmosis wastewaters, filter backwash from service water polishing, and/or feed-and-bleed from the service water system and the standby cooling tower; and (5) intermittent wastewaters from floor washdown, equipment washing, personnel decontamination, laboratory drains, and treated wastewaters from low-level, solid radioactive waste dewatering (Note: These treated wastewaters are discharged when recycling to condensate storage, i demineralization, and reuse as boiler feed is not available). 1 5312R01.DP 4
There are two treatment systems associated with Outfall 002. The wastewaters described in items (1) and (2) above are always pumped, and the wastewater described in Item (3) above is pumped on an intermittent, as-needed basis, to one of two 30,000-gallon capacity treatment tanks for neutralization before discharge. A process monitor controls the discharge from these tanks, recirculating the tank contents until the pH is within preset limits, then allowing the diversion of the treated water through disposable filter cartridges to the common discharge header (to Outfall 001). If the process monitor senses an unacceptable shift in pH during discharge, the wastewater is diverted back to the tanks for further treatment. Neutralization, filtration, and other treatments may be provided by a contracted service or with temporary equipment for special projects, with treated effluent discharged to the cooling tower flume or directly to the common discharge header. Solids removed during wastewater treatment are sent for approved off-site disposal. With further regard to the wastewaters described in Items (1) and (2), polishing is necessary for well water used in the plant and the auxiliary boiler. Polishing is accomplished through reverse osmosis and ion-exchange systems. The auxiliary boiler is brought in by a contractor every 18 months or so when the reactor has been inactive and needs to be restarted. Boiler blowdown is routed to the non-radioactive, low-volume wastewater treatment system and Outfall 002. Approximately twice per year, the ion-exchange system is restored, and the resulting ion-exchange resin backwash and regeneration wastes are routed to the non-radioactive, low-volume wastewater treatment system and Outfall 002. During polishing of the makeup water, a reverse osmosis reject stream is produced. This wastewater is currently intermittently routed to the non-radioactive, low-volume wastewater treatment system and Outfall 002. By this application, Entergy is requesting authorization to reroute the reverse osmosis reject (ROR) from Outfall 002 to Outfall 006. The reverse osmosis reject water is intermittently produced, at 25 gpm during operation (24-hour period for three days, every two weeks); this results in a long-term average flow rate of 7,714 gallons per day (gpd). In order to facilitate process operation, Entergy wishes to reroute this wastewater from Outfall 002 to Outfall 006. Entergy believes that routing this concentrated well water back into the environment without treatment will have no adverse effects on the environment. Effluent characterization data are presented on Form 2C (as Outfall ROR). In a separate treatment system, low-level radioactive wastewater from the steam condenser system, reactor water cleanup system, and fuel pool system demineralizers' backwash, as well as solid radioactive waste dewatering, floor and lab drains., equipment washing / draining, and personnel decontamination [ Item
- (5) above] is collected in one of nine 25,000-gallon holding tanks for filtration and/or demineralization. Treated water collects in one of four 19,500-gallon recovery tanks for monitoring of boiler water quality and radioactivity. The
- station recycles this water whenever demineralization achieves boiler water l quality and sufficient tankage exists. Otherwise, the treated wastewater is metered to the common discharge header (to Outfall 001) at a rate ensuring
\ c , 5312R01 DP 5 l
compliance with 10 CFR Part 20 and 10 CFR Part 50 - Appendix I standards. When this treated wastewater must be discharged, the tank is sampled during recirculation to verify that all parameters are within permit limits. If the wastewater is not within permit limits, the tank is reprocessed for permit compliance prior to discharge. The ion-exchange resins used in these demineralization processes are replaced instead of regenerated. The station disposes of these resins and other solids removed during the treatment of these j low-volume wastes in accordance with NRC, U.S. EPA, U.S. Department of Transportation (DOT), and applicable state requirements. Permit compliance monitoring is performed on these two treated effluent streams ; before they are released to the river via the common discharge header (to Outfall i 001). The results of each are combined (flow-weighted) for reporting as Outfall 002. Outfnll 102 This outfall discharges the treated metal cleaning wastewater [ listed also as Item (3) under Outfall 002 above]. This wastewater is discharged on an intermittent , basis only. The cleaning and passivation stages use specialized chemicals ' designed to remove scale and corrosion products from iron, copper, zinc, and nickel surfaces. The cleaning / passivation stages are usually followed by a rinse with fresh or demineralized water. Treatment for this wash and rinse water typically consists of contracted services that may include biodegradation, precipitation of dissolved metals, filtration, and neutralization. If the quality of the treated water is suitable, it is chlorinated and recycled to the cooling tower makeup water system. Permit compliance monitoring is performed before the wastewater is recycled or discharged. If recycling is not available, compliance I monitoring is performed as the treated water is conveyed to the non-radioactive, I low-volume wastewater treatment system (Outfall 002) or pumped directly to the j i common discharge header (Outfall 001). This batch treatment may yield 100,000 gallons of treated water per day, discharged at up to 400 gpm. This discharge occurs very infrequently. This wastewater has only been discharged once since the River Bend Station became operational, and this was during a three-month period in 1992. 1 Outfall 003 This outfall discharges the non-radioactive floor drain wastewater and transformer yard wastewater /stormwater. Three oil / water separators discharge through the storm drain system to Outfall 003, then to Outfall 006, then to the East Creek, l and then to Grant's Bayou which ultimately discharges to the Mississippi River. l Two of the oil / water separators receive intermittent fire suppression water (from sprinklers) and stormwater runoff from within the River Bend Station electric power distribution transformer yards. The third oil / water separator receives wastewater from floor drains within power plant buildings not associated with the nuclear reactor, and therefore having no potential for radioactive contamination. I
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These non-radioactive floor drain wastewaters consist of well water, fire suppression water, and domestic (potable) water. During the refueling outage beginning in March 1992, the plant's cooling water system was modified to isolate the service water system from the condenser cooling system. To prevent this chemically treated service water from entering the storm drain system, these non-radiologically-contaminated floor drains have been isolated from the yard drain system. These floor drains were rerouted to the sanitary waste treatment system. OutInli 004 The wastewater discharged via this outfall is the treated sanitary wastewater from facilities throughout the River Bend Station. The existing sewage treatment plant and Outfall 004 currently discharge via Outfall 001 to the Mississippi River. A new sewage treatment plant is under construction. When it becomes operational, the location of Outfall 004 will change as shown on Figure 2. Construction is anticipated to be completed by late 1995. The following discussion addresses the design and operation of the new sewage treatment plant. Treatment consists of two parallel systems, one for the sanitary discharge from the power plant, and another system for all other sanitary discharges from the River Bend Station. No personnel decon water is allowed to the sanitary system from " radiologically-active" portions of the power plant. Both treatment systems are comprised of an aerated lagoon followed by a sedimentation pond and a rock filter basin. Influent l wastewater passes through a bar screen prior to entering the aerated lagoons. Undesirable microbial activity within the sedimentation pond will be removed by the rock filter basin. The design life for the sedimentation ponds and rock filter , l basins is 20 years. Efnuent from both systems drains by gravity to lift stations, where it is pumped to a common sand filter. Treated effluent drains by gravity through the sand filter and through an ultraviolet disinfection unit immediately l prior to monitoring for permit compliance. The treated effluent is pumped to a l sump that is normally pumped to the common discharge header. There, the effluent is combined with cooling tower blowdown and other monitored outfalls I for discharge via final Outfall 001 to the Mississippi River. During infrequent maintenance activities on the common discharge header, the treated sanitary effluent will temporarily be routed to Grant's Bayou via Outfall 005. Solids removed by sedimentation and tertiary filtration are sent for approved off-site disposal, i As described above, the plant's cooling water system was modified in March 1992 to isolate the service water system from the condenser cooling system. This isolated service water system contains a biocide as part ofits chemical treatment. To prevent this chemically treated service water from entering the storm drain I systems, these non-radiologically-contaminated floor drains, including one oil / water separator, were isolated from the yard drain system. These floor drains
, were rerouted to the sanitary waste treatment system (and will continue to be routed to the new sewage treatment plant), and as discussed above the effluent 5312R01.DP 7 ;
i
r l from the sanitary waste treatment system was rerouted from Grant's Bayou to the Mississippi River via the cooling tower blowdown common header (and Outfall 001). Outfall 005 Outfall 005 discharges stormwater runoff from the industrial materials storage l area and the Low Level Waste Storage Building area to Grant's Bayou as shown on Figure 2. As discussed above for Outfall 004, a new sewage treatment plant is under construction. Stormwater from the 3.3-acre area surrounding the new sewage treatment plant will be discharged through Outfall 005. Outfall 004, normally discharged to the Mississippi River, will be diverted to Outfall 005 during scheduled maintenance of the common discharge header / valves. Therefore, Entergy requests that Outfall 005 specifically be authorized by the renewal permit for (1) this additional source of stormwater and (2) the infrequent and temporary discharge of treated sanitary effluent. Outfnli 006 Outfall 006 includes the discharge of the drainage conveyances from the east side of the River Bend Station to the East Creek and then to Grant's Bayou as shown on Figure 2. It consists of stormwater from a significant portion of the power station; de minimis quantities of cooling tower drift / mist; condensate from oil-free, electric-driven and backup diesel air compressors; reverse osmosis reject (requested in this permit application); discharge from Outfall 003; and a portion of the discharge from Outfall 008. The station building roof and yard drain systems direct drainage to a ditch called East Creek, which receives stormwater runoff from the site. A relatively new source to Outfall 006 is condensate from recently installed air compressors associated with the Instrument Air / Service Air Systems. Six of the compressors are electric and oil-free and will operate continuously. There are two backup compressors which will operate only when the main units are out of service: one electric, oil-free compressor and one diesel-driven air compressor. The two backup compressors will also be tested weekly. It is estimated that flow l from condensate drains for these systems will be approximately 16 gpm (to Outfall 006) when operating. Entergy notified the U.S. EPA and the LDEQ of this discharge in letters dated February 14,1995 and May 16,1995. The LDEQ responded with an August 10,1995 letter of no objection regarding this discharge through Outfall 006. Entergy requests that the renewal permit specifically authorize the discharge of air compressor condensate through Outfall 006. As discussed previously, Entergy requests specific authorization to reroute reverse osmosis reject water from Outfall 002 to Outfall 006. Effluent characterization l data are presented on Form 2C (as Outfall ROR). { 5312R01.DP b
~
Outfall 007 l Outfall 007 includes the discharge of the drainage conveyances from the west and north sides of the plant to West Creek and then to Grant's Bayou as shown on Figure 2. It consists of stormwater from the west and north sides of the plant and L a portion of the discharge from Outfall 008. A network of small ditches from office areas, warehouse areas, materials storage areas, and equipment and vehicle p maintenance areas (including de minimis quantities of domestic water vehicle L rinsate) connect to a drainage ditch called West Creek which receives stormwater runoff from these areas of the site. Outfall 008 The discharges designated and monitored as Outfall 008 result from the hydrostatic testing and flushing of piping systems and vessels, including periodic required flushing and testing of the Fire Protection Water Supply System and the Automatic Sprinkler System. Wastewater from hydrostatic testing and flushing activities is usually conveyed from the power plant and support areas by hoses or temporary piping to yard drains or ditches for discharge to either East Creek (via Outfall 006) or West Creek (via Outfall 007) and then to Grant's Bayou. Some of these activities may also direct wastewater to the sanitary waste treatment system (to Outfall 004) via non-radiologically-contaminated plant floor drains or to the cooling tower flume for discharge to the river (via Outfall 001). Flushing and hydrostatic testing is usually performed with well water. Occasionally, demineralized water may be used, which, upon standing in storage, absorbs carbon dioxide resulting in pH levels sometimes as low as 5.6 standard units. Outfall 009 [ While this stormwater outfall is currently addressed and authorized in NPDES permit number LA0042731 for the River Bend Station, it is a proposed new outfall for LWDPS permit number WP 0409. This outfall is the stormwater discharge from part of the cooling tower yard to Grant's Bayou on the extreme , eastern side of the power station as shown on Figure 2. Stormwater runoff and de minimis quantities of cooling tower drift / mist drains by gravity from tne cooling tower area eastward to Grant's Bayou via Outfall 009. i 3.2 Ancillary Water Systems The cooling water treatment program to minimize scaling, biofouling, and corrosion of plant metallurgy consists of the following: Cooline Tower Water The following may be added to the river water intake pumps / piping and clarifiers ! providing cooling tower makeup for condenser cooling and service water cooling:
+ Cationic coagulant, occasionally supplemented with anionic flocculent during periods oflow river water turbidity, may be added to river water 3312R01.DP 9
clarifiers for silt and colloid removal. Control of pH may be undertaken subsequently to enhance this process.
+ Clarifier clearwells may be shock chlorinated with sodium hypochlorite and possibly sodium bromide for control of algae and macrofouling agents such as the zebra mussel, Dreissena polymorpha.
+ Clarifier sludge is diluted with river water to approximately 4 % solids and returned to the Mississippi River.
.I + Sodium hypochlorite and possibly sodium bromide may be injected intermittently, or continuously at lower levels, into the river water intake .I at the river to control infestation of the intake pipeline by the zebra mussel. Alternatively, a non-oxidizing biocide, such as a quaternary I amine, may be added occasionally to the river water intake at the river to control infestation by zebra mussels. This occasional use of the non-oxidizing biocide is planned to occur on a 3- or 4-day per year basis,
- B depending on the entrainment of the zebra mussel larvae. This infrequent use of non-oxidizing biocide in the river water intake system is strictly for the protection of the buried pipeline to the intake water clarifiers. Its use is not expected to produce a detectable biocide residual in the cooling tower water or in cooling tower blowdown that is ultimately discharged via Outfall 001.
I The following may be added to the cooling towers / flumes: l + Zinc salts, and/or phosphate salts, blended with an anionic copolymer, and/or terpolymers may be added for mild corrosion control of steel l structures (piping, vessels, etc.). I + Tolyltriazole salts may be added for copper and brass corrosion control.
+ A polyacrylate polymer /hydroxyethylidene diphosphonate (HEDP) blend may be added for scaling control.
l
+ Sodium hypochlorite and possibly sodium bromide / surfactant blend may be added for biofouling control.
+ Sulfuric acid may be added for pH control.
I + The cooling tower system operation normally results in 4 to 6 cycles of concentration. The cooling tower blowdown is dechlorinated with ammonium (or sodium) bisulfite as needed before discharge to the river. l I I L m a m.or 10
Isolated Service _and Standby Cooline Water The isolated service water is made up with demineralized water to which may be added molybdate, nitrite, and tolyltriazole sodium salts for corrosion control, polyacrylate dispersant for scaling control, sodium hydroxide for pH control, very low levels of an antifoaming agent, and a broad spectrum biocide such as isothiazoline, glutaraldehyde, or dibromonitrilopropionamide. The standby cooling water is a reservoir of 6.5 million gallons made up from fresh well water and a multicell induced draft cooling tower to which may be added sodium hypochlorite and possibly sodium bromide / surfactant, hydrogen peroxide, and/or a broad spectrum biocide such as isothiazoline, glutaraldehyde, or dibromonitrilopropionamide for biological control. This system provides backup emergency cooling of nuclear safety related systems in the event that normal cooling becomes unavailable. During refueling outages, at 18-month intervals, this standby cooling tower is operated for several weeks with the isolated service water while the normal systems undergo maintenance. The water treatment chemicals listed above for the isolated service water system are added to the reservoir to maintain the corrosion and biological control attributes of the isolated service water. Cooling tower reservoir water level and water quality are controlled by feed-and-bleed with fresh well water, with the reject water discharged via Outfall 002. Auxiliary Boiler Water The following may be used for auxiliary boiler makeup: zeolite softeners for demineralization, sodium sulfite or hydrazine for oxygen removal, phosphate salts for scaling control, and sodium hydroxide for pH control. Fire Sunpression Water The following may be used for protection of the fire suppression water system: sodium hypochlorite and possibly sodium bromide or a biodegradable biocide for biofouling control, sodium hydroxide for pH control, and phosphate or molybdate / nitrite salts for corrosion control. With the exception of the zinc salts noted above, no chemicals which contain any of the priority palhttants listed in 40 CFR Part 423, Appendir A, are usedfor treatment ofcooling water or service waters discharged to the environment. 4.0 STORMWATER D.RAINAGE, MANAGEMENT, AND DISCHARGE In accordance with the requirements of the revised NPDES stormwater discharge permit application regulations under 40 CFR 6122.26, Entergy is presenting the following discussion on stormwater management at the River Bend Station. This discussion is presented in conjunction with the information required in connection with and provided on U.S. EPA Form 2F (Appendix C) as it relates to the currently permitted NPDES
$312R01.DP II
stormwater Outfalls 003,005, 006, 007, and 009 which discharge stormwater associated with industrial activity from the site. The drainage areas for these stormwater outfalls are described in Section 3.0. The quantitative analytical data characterizing the stormwater discharged through the stormwater outfalls are presented on Form 2F. Other nonanalytical information required by Form 2F is provided below for stormwater discharged through, and the dreinage areas served by, the stormwater outfalls at the site. Stormwater runoff at the Entergy site from all areas associated with industrial activity l (as denned by 40 CFR 6122.26) is discharged through Outfalls 003,005,006,007, and l 009. Stormwater runoff at the site from areas which are not associated with industrial l activity discharges from the site by either sheet flow or point sources which do not l require permitting under 40 CFR fl22.26. Figure 2 depicts features at the Entergy site l pertinent to stormwater. This figure illustrates the areas from which stormwater drains into the outfalls, direction of stormwater flow to these outfalls, intake and discharge structures, and structural control measures designed to reduce pollutants in stormwater. ! Also, Figure 4 shows surface types in the areas drained by the outfalls (i.e., impervious versus non-impervious). Hazardous waste storage units and areas where significant materials that are potentially exposed to stormwater are handled or stored are shown on Figure 2. Table 2 is an inventory of the significant materials storage / unloading areas and lists the containment associated with each area. Table 3 is an inventory of significant materials within oil storage areas; most of these areas are not shown on Figure 2 because , they are located inside of buildings and thus have no potential to impact stormwater. The i transformers listed in Table 3 are also not shown on Figure 2 because they are too numerous. Structural controls used to minimize the potential for stormwater contamination include containment dikes / berms around the toxic or hazardous materials handling areas, tanks, and the hazardous / nonhazardous waste storage areas. Sloping and grading of roads and
> lands are used to direct stormwater runoff to a storm drain where appropriate. The storm drain system of pipes and ditches provides a mechanism to contain and control runoff, facilitating the effective use of countermeasure plans in spill control.
Nonstructural measures employed at the site which aid in the management of stormwater include: 1
+ Stormwater Pollution Prevention Plan,
+ Spill Prevention Control and Countermeasure Plan,
+ Hazardous Waste Management Plan
+ Environmental Inspections,
+ Plant Emergency Response Plan, i + Employee Safety Training Programs, and
+ Equipment Preventive Maintenance Programs. l
~
3312R01.DP I2
These programs have definite schedules which encourage awareness of the importance of the program and require equipment operational tests and repairs which assist in minimizing the potential for contaminant releases. Entergy has no hazardous waste treatment or disposal units. Hazardous waste storage units are shown on Figure 2 and include a Hazardous Waste Storage Building (with a concrete berm inside) which is utilized for the purpose of 90-day or less accumulation of drums of hazardous wastes prior to their shipment for off-site disposal. Hazardous wastes stored in this area include paint waste, paint thinner, fuel operation waste, photographic waste, and waste varsol. The shop area has an outside, but under roof, hazardous waste satellite storage area for paints and solvents within concrete containment. Because the River Bend Station is a nuclear fuel electric power generation plant, very little process hazardous waste is generated. Most hazardous waste is generated from construction, maintenance, and other support actidties. Radioactive hazardous waste is generated inside the power plant and is thus contained within the confines of the radiologically-controlled area. The River Bend Station employs numerous operational practices to avoid and/or contain all potential releases of significant materials. Significant materials used in the process areas are stored or handled such that they will not impact stormwater runoff. All roads at the site are used for the transport of significant materials. Imading and unloading areas are shown on Figure 2. Entergy uses herbicides such as Roundup
- at the River Bend Station in limestone areas, landscape areas, and parking lots. Previous typical usage of Roundup
- was approximately three gallons per year. Herbicides are only used in areas which, if exposed to stormwater, are within the drainage of permitted outfalls. De minimis !
quantities of fertilizers, soil conditioners, and insecticides may be used in plant areas which, if exposed to stormwater, are within the drainage of permitted outfalls. l Significant leaks or spills of toxic or hazardous substances at the site during the last three l years are required to be reported in accordance with 40 CFR 6122.26(c)(1)(i)(D).
"Significant spills" are defined as the release within a 24-hour period of toxic or hazardous substances in excess of reportable quantities under Section 311 of the Clean Water Act and/or Section 102 of the Comprehensive Environmental Response, ;
Compensation and Liability Act (CERCLA). Reportable Quantities are predefined amounts of substances as listed in 40 CFR Part 117 and 40 CFR Part 302. There have been two reportable spills / leaks at the Entergy River Bend Station in the last three years, ; and both had minimal potential to be exposed to precipitation or the potential to drain to a stormwater conveyance. On October 20,1992, a spill involved 70 gallons of sodium ; hypochlorite (15.0%), and on March 10,1993, a spill involved 500 gallons of sodium hypochlorite (12.0%). Both releases involved sodium hypochlorite which was spilled l within the concrete berm of a tank in the Water Chemical Addition Area. The spilled ! material was recovered for normal usage. l l 4 A l
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i 5.0 WASTEWATER AND STORMWATER SAMPLING AND ANALYTICAL CONSIDERATIONS In accordance with the requirements of U.S. EPA Application Forms 2C and 2F, wastewater efnuent analytical data were obtained for each outfall discharge. Representative wastewater and stormwater samples from all of the permitted outfalls were collected as required by NPDES regulations at 40 CFR 6122.21 and 40 CFR 6122.26. Effluent characterization data are presented onF' orm 2C for non-stormwater outfalls or for non-stormwater components of an outfall (for those which discharge a combination of stormwater and non-stormwater sources). Flow rate data obtained from Discharge Monitoring Reports (DMRs) for the period February 1993 through January 1995 and analytical data for the period February 1994 through January 1995 for those parameters that are required to be monitored at the outfalls have been included on Form 2C, Part V. Because Outfall 102 has not discharged in recent years, it could not be sampled for this permit application. Instead, historical analytical data for Outfall 102 are summarized - on Table 4. Sampling activities were conducted at the other non-stormwater outfalls for the permit application as follows. For Outfall 001 (process waste. water), a 24-hour sampling event was conducted on June 22 through 23,1995 in order to obtain the required analytical data. Because this outfall discharges continuously, a 24-hour flow-weighted composite sample was collected for all analyses, except for those pollutants (oil and grease, pH, temperature, fecal coliform, total phenols, cyanide, and total residual chlorine) which require grab samples as speciGed at 40 CFR 6122.21(g)(7). During the June 22 through 23, 1995 sampling period, four discrete volatile organic compound (VOC) sample aliquots were manually collected; these aliquots were combined in equal volumes by the analyst in the laboratory immediately before analysis to prepare a single composite sample. A grab sample was collected at Outfall 001 on June 28, 1995 for fecal coliform analysis. A 24-hour composite flow-weighted sample was collected at Outfall 001 on August 28 through 29, 1995 for polychlorinated biphenyls (PCBs) analyses. Outfall 002 (process wastewater) was sampled by obtaining separate grab samples from the two low-volume wastewater treatment systems. Twenty-four-hour composite samples were not collected because both sources are intermittent, not continuous, discharges. The j sample at the low-volume waste treatment system (no low-level radioactivity contribution ' i sources) will be hereafter referred to as Outfall 002A, and the other sample at the low- l level radioactive, low-volume waste treatment system will be referred to as Outfall 0028. Grab samples were collected on June 21,1995 at Outfall 002B and on June 22,1995 at Outfall 002A, and analyzed separately. The results of the two samples were flow-weighted and combined to characterize the combined wastewater discharged through Outfall 002. Outfall 002A was resampled on August 29,1995 for mercury analysis. For Outfall 003 (nonprocess wastewater), the intermittent, dry-weather discharge from the oil / water separator which receives wastewater from non-radiologically contaminated power plant floor drains (consisting of well water, fire suppression water, and domestic potable water) was sampled on June 22, 1995. Only grab samples were collected m 1 s 5312R01.DP l4 J
because this source to Outfall 003 is an intermittent, not continuous, discharge. A i sample was collected on June 28,1995 for fecal coliform analysis. Because this outfall consists of two other stormwater sources, DMR data were not included on the Form 2C (which represents only the non-stormwater compenent); instead, a DMR summary is presented on Table 5 (representing all three sources to the outfall). For Outfall 004 (nonprocess and sanitary wastewater), a 24-hour sampling event was conducted on June 21 through 22,1995 in order to obtain the required analytical data. Because this outfall discharges continuously, a 24-hour flow-weighted composite sample was collected for all analyses, except for those pollutants which require grab samples as specified at 40 CFR Q122.21(g)(7). During the June 21 through 22, 1995 sampling I period, four discrete VOC sample aliquots were manual y collected; these aliquots were equally combined by the analyst in the laboratory immediately before analysis to prepare a single composite sample. A grab sample was collected on June 28, 1995 for fecal coliform analysis. 4 l f Outfall 008 (nonprocess wastewater) was sampled on June 22,1995. Only grab samples were collected because this is an intermittent, not continuous, discharge. The reverse osmosis reject source (referred to as Outfall ROR on the Form 2C) to Outfall 002 (which is being requested for rerouting to Outfall 006) was sampled on June 22,1995. Only grab samples were collected because this is an intermittent, not continuous, discharge. Effluent characterization data are presented on Form 2F for stormwater outfalls or for stormwater components of an outfall's discharge (for those which discharge a combination of stormwater and non-stormwater sources). Flow rate data obtained from-DMRs for the period February 1993 through January 1995 and analytical data for the period February 1994 through January 1995 for those parameters that are required to be monitored at the outfalls are summarized in Tables 5 through 9. Sampling activities were conducted at the stormwater outfalls for the permit application as follows. First-flush and composite samples for Outfalls 003, 005, 006, 007, and 009 were collected during a storm event on July 5,1995 which had a total rainfall of 0.32 inch and a duration of approximately four hours. The previous rainfall event with at least 0.1 inch of rainfall occurred on July 1,1995. Form 2F includes flow data for the discharge of stormwater through the outfalls during the sampling event and the areas which contribute to the total drainage area of the outfalls. I Outfall 003 has two stormwater sources from oil / separators associated with the l transformer yards (auxiliary and main). Stormwater samples were collected at only one of the stormwater oil / separators (the auxiliary), because it has been determined that the two oil / water separators discharge stormwater which is "substantially identical" [as allowed at 40 CFR Q122.21(g)(7)). H12R01.DP 15
f 6.0
SUMMARY
OF PRIOR BIOMONITORING REQUIREMENTS AND RESULTS 1 As required by 40 CFR fl22.21(g)(ll), information on biological toxicity tests conducted within the last three years on Entergy's discharges is included in this permit renewal application. Entergy performed toxicity tests during three molluscicide treatments of the Mississippi River intake water during the previous three-year period. A chronic elutriate toxicity test using Ceriodaphnia dubia and a chronic 10-day static, solid-phase, sediment toxicity test using Hyalella azteca were conducted on samples of sediment from the intake water clarifier collected prior to and during the first two molluscicide applications, January 6 and November 10,1994. Acute 48-hour static-renewal toxicity tests using Daphniapulex and Pimephales promelas was conducted on Outfall 001 effluent collected August 17, 1995. Each molluscicide application consisted of an approximate 8-hour period in which the non-oxidizing Calgon molluscicide H130M or Betz molluscicide CT-2 was injected into the Mississippi River water intake system. In the first two applications, samples of clarifier sediment were collected one day prior to molluscicide application (untreated sample) and during application (treated sample). During the third molluscicide application only Outfall 001 effluent, containing clarifier blowdown, was collected for toxicity testing. The chronic elutriate toxicity tests were conducted with three elutriate concentrations
, (25%,50% and 100%) and two control treatments (a sediment control and a water only control). Reconstituted moderately hard water was used as the dilution and control water. The chronic 10-day static, solid-phase sediment toxicity tests consisted of one treatment and a control with the overlying water consisting of reconstituted moderately hard water. The 48 hour acute static-renewal toxicity tests consisted of five effluent dilutions (0.2%, 0.3%, 0.4%, 0.6% and 0.8% effluent) in addition to two control treatments (laboratory and dilution water control). Dilution water consisted of Mississippi River water.
No Observed Effect Concentration (NOEC) vaues were calculated for the Ceriodaphnia dubia chronic elutriate toxicity tests and the Daphnia pulex and Pimephales promelas j acute toxicity tests. NOEC values are the highest concentration of effluent or clutriate to which organisms are exposed which causes no statistically significant adverse effect on organism survival or reproduction in comparison with the control (0% effluent,0% elutriate). In the Hyalella azteca solid-phase toxicity tests, percent survival and growth, as measured as average dry weight, were compared to the control for significant differences. Test results from the clarifier sediment toxicity tests are presented in Table 10. : Ceriodaphnia dubia survival in the January 6 and November 10, 1994 tests and reproduction in the November 10, 1994 tests were not significantly different from the control in either the untreated or treated clarifier sediment tests. Reproductive effects in the January 6,1994 tests could not be determined due to the poor performance in the control treatment. 5312R01.DP Ib
l l Hyalella azteca survival in the January 6 and November 10,1994 tests and growth in the January 6,1994 tests were not significantly different from the control in either the untreated or treated clarifier sediment tests. Growth was significantly different from the control in the untreated and treated tests conducted November 10, 1994. I In a letter to Entergy dated May 23,1995 (see Appendix D), the LDEQ stated that 48-hour acute toxicity testing would be sufficient to monitor effluent quality during l molluscicide application. A molluscicide application event was conducted on August 17, l 1995, and acute toxicity test results are presented in Table 10. The acute 48-hour survival NOEC value for both the Daphnia pulex and Pimephales promelas test species l was 0.8% effluent, which was the highest effluent dilution required to be tested. 1 I
)
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I r i l I J l 3 4 I i, e f' 1 M I 1 1 i e I 4 l ,d I l i, 1 i. 1 , TABLES 6 P a 4 4 l 4 1 i l 4
.I I
d i e Y i 1 4 4 4 ) > f P' A
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TABLE 1 ENTERGY RIVER BEND STATION INVENTORY OF WATER WELLS IN TIIE [ VICINITY OF TIIE SITE L Latitude - Well Number 5 Owner' . Longitude L. .Well Depth m wegg p,e ( 68 Ed Daniels 30'45'50" 91'18'58" 483 Domestic r 69 Ed Daniels 30'45'50" 168 Domestic
- L 91*18'59" l 82 J.E. Poche Jr. 30'46'02" 510 Domestic I 91'19'17" i
84 Ricks 30*46'12" 180 Domestic l y 91*20'34" L 85 Adda Markie 30'15'37" 103 Domestic
- 91'20'40" l l 87 Murphy Dreher 30*46'14" 497 Industrial 91'19'28*
, 91 H. Daniel 30*46'08" 485 Domestic L 91'i9'07-241 J. Rogers 30'46'13" 161 Domestic f 91'20'35" 246 Entergy River Bend Station 30'45'18" 1,821 Power Generation 91*19'52" l L 256 Entergy River Bend Station 30'45'19* 124 Fire Protection 91'19'50" ( 257 Entergy River Bend Station 30*45'19" 91*19'46" 1,815 Power Generation 266 Entergy River Bend Station 30'45'40" 500 Industrial 91*20'11* i l Well number assigned in LDOTD database. Depth of the well, in feet, measured from the bottom of the screen to the ground surface. Although this wc!! is listed as abandoned, it is included herein because it was not listed as closed, and snay therefore be used again in the future.
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TABLE 2 ENTERGY RIVER BEND STATION INVENTORY OF SIGNIFICANT MATERIALS IN STORAGE AND UNLOADING AREAS Item No.Ju Description Volume Containment i Standby Cooling Tower Hypochlorite Tank 1,000 gal Concrete Curb, Drains to Cooling Tower 1 Standby Cooling Tower Hypochlorite Unloading Unloading Curbed Concrete Pad 2* Emergency Diesel Generator Fuel Unloading Unloading Curbed Concrete Pad 3 CWS Treatment Chemicals Tanks
- - TTA (Nalco 9237) 3,000 gal. Concrete Floor & Walls
- HEDP (Nalco 1345) 6,400 gal. Concrete Floor & Walls
- Zine Chloride (Nalco 1360) 6,400 gal. Concrete Floor & Walls
- Sodium Bromide (Nalco 1338) 6,000 gal. Concrete Floor & Walls
- Sodium Hypochlorite 6,000 gal. Concrete Floor & Walls 3 CWS Treatment Chemicals Unloading Unloading Curbed Concrete Pad 4 Hypochlorite Tank 22,000 gal. Concrete Floor & Walls 4 Hypochlorite Unloading Unloading Curbed Concrete Pad i 5 WTA Sulfuric Acid Tanks (Two) 42,000 gal. ea. Concrete Floor & Walls 5 WTA Acid Unloading Unloading Curbed Concrete Pads 6 Ammonium Bisulfite 4,000 gal. Concrete Floor & Walls 6 Ammonium Bisulfite Unloading Unloading Curbed Concrete Pad 7 Fire Pump Diesel Fuel Unloading Unloading Concrete Curbed with Earthen Floor 8* Diesel Fuel Trailer Parking 2,750 gal. Concrete Curb & Sump (Largest Compartment) 9 Ilazardous/Non-hazardous / Oil Waste Storage & Drums Curbed Concrete Floor Unloading & Walls 10 Paint Shop Drum Storage & Unloading Drums Curbed Concrete Floor 11 "' Main Warehouse Drum Storage & Unloading Drums Concrete Floor, Walls &
Sump l 12
- Gasoline / Diesel Storage (Two) 6,000 gal. each Concrete Floor & Walls 12
- Gasoline and Diesel Storage Unloading Unloading Curbed Concrete Pad 13 Main Warehouse Hazardous Materials Storage Drums Curbed Concrete Floor, Building & Unloading Walls & Sump 14 Outside Oil Drum Storage Building & Drums Curbed Concrete Floor, Unloading Walls & Sump 5312R01.DP
I i ~ TABLE 2 ENTERGY RIVER BEND STATION INVENTORY OF SIGNIFICANT MATERIAIS IN STORAGE i AND UNLOADING AREAS (Continued) L ltem No.m ' Description Volume Containment 15 Turbine Building Oil Storage and Unloading Drums Curbed Concrete Floor
& Walls 16 Service Water Storage & Unloading (Three) 1,000,000 gal. Concrete floor, Walls &
each Lined Earthen Berm 17 Main Warehouse Paint /Flammables Storage Drums / Containers Curbed Concrete Floor, Walls & Sump L 18 SWP Treatment Chemicals Tanks
- Sodium Molybdate (Nalco 7357)(Two) 400 gal. each Concrete Floor & Walls l - Sodium Hydroxide & TTA (Nalco 1336) 400 gal. Concrete Floor & Walls
- Glutaraldehyde (Nalco 7338) 1,000 gal. Concrete Floor & Walls
- Isothiazoline (NALCO 7330)(Two) 400 gal. each Concrete Floor & Walls l - Sodium Hydroxide (NALCO 8073) 400 gal. Concrete Floor & Walls
- Sodium Nitrite Solution 400 gal. Concrete Floor & Walls
- Polyquarternary Amine (NALCO 8103) 5,000 gal. Concrete Floor & Walls r
L 19
- Field Administrative Diesel Generator Fuel 200 gal. None Tank r 20
- Drummed Oil 150 gal. (varies) Secondary Containment
, Item numbers correspond to those shown on Figure 2. Additional information is presented for this item on Table 3 (corresponds to Item 21 on Table 3). Additional information is presented for this item on Table 3 (corresponds to Item 5 on Table 3). Additional information is r. resented for this item on Table 3 (corresponds to item 6 on Table 3). Additional information is presented for this item on Table 3 (corresponds to Items 3 and 4 on Table 3). Additional information is presented for this item on Table 3 (corresponds to Item 22 on Table 3). Additional information is presented for this item on Table 3 (corresponds to Item 24 on Table 3). I k l s $382R01.DP
TABLE 3 ENTERGY RIVER BEND STATION INVENTORY OF SIGNIFICANT MATERIALS IN OIL STORAGE AREAS Volume Description (Gallons) Drainage Containment
- 1. Fire Protection Diesel Fuel Tank "l A" - Fire Protection Pump 300 Through oil water separator #2 into East Creek Inside a building House
- 2. Fire Protection Diesel Fuel Tank "lB" - Fire Protection Pump 300 Through oil water separator #2 into East Creek inside a building House 3.* Vehicle Gasoline Fuel Tank - Vehicle Maintenance Shop 6,000 On ground into West Creek Covered by a roof 4.* Vehicle Diesel Fuel Tank - Vehicle Maintenance Shop 6,000 On ground into West Creek Covered by a roof 5.* Auxiliary Diesel Fuel Tanker - Southwest of the Hazardous Waste 6,500 On ground into West Creek Yes Yard 6.* Drummed Oil - Warehouse Oil Storage Building i1,500 On ground into West Creek Covered by a roof (varies)
- 7. Drummed Used Oil - Hazardous Waste Yard Varies On ground into West Creek inside a building
- 8. Drummed EHC Fluid - Hazardous Waste Yard Varies On ground into West Creek inside a building
- 9. Lube Oil Containers / Drums - Lube Oil Storage Facility 1,600 Into a sump and then drummed for off-site disposal inside a building (varies)
- 10. Lube Oil Containers - Turbine Lube Oil Storage Facility 1,440 Into a sump and then drummed for radwaste Inside a building (varies) processing i1. Drummed Used Oil - Turbine Lube Oil Storage Facility 990 Into a sump and then drummed for radwaste inside an (varies) processing underground vault
- 12. Standby Diesel Generator Division I Fuel Tank - Diesel Generator 50,000 Through oil water separator #1 into sewage inside an Building treatment plant underground vault
- 13. Standby Diesel Generator Division II Fuel Tank - Diesel 50,000 Through oil water separator #1 into sewage Inside an Generator Building treatment plant underground vault
- 14. HPCS Diesel Generator Division III Fuel Tank - Diesel Generator 50,000 Through oil water separator #1 into sewage inside an Building treatment plant underground vault 5312RDI.DP
J TABLE 3 ENTERGY RIVER BEND STATION INVENTORY OF SIGNIFICANT MATERIALS IN OIL S1DRAGE AREAS (page 2 of 4) Volume Description (Gallons) Drainage - Containment
- 15. Standby Diesel Generator Division I Fuel Oil Day Tank - Diesel 535 nrough oil water separator #1 into sewage Inside a building Generator Building treatment plant
- 16. Standby Diesel Generator Division II Fuel Oil Day Tank - Diesel 535 Drough oil water separator #1 into sewage. Inside a building Generator Building treatment plant
- 17. HPCS Diesel Generator Division III Fuel Oil Day Tank - Diesel 535 nrough oil water separator #1 into sewage Inside a building j Generator Building treatment plant l 18. Standby Diesel Generator Division I Lube Oil Sump Tank - 514 Through oil water separator #1 into sewage Inside a building Diesel Generator Building treatment plant
- 19. Standby Diesel Generator Division II Lube Oil Sump Tank - 514 Drough oil water separator #1 into sewage inside a building Diesel Generator Building treatment plant
- 20. HPCS Diesel Generator Division ill Lube Oil Sump Tank - 514 Drough oil water separator #1 into sewage Inside a building Diesel Generator Building treatment plant 21.* Station Blackout Diesel Generator Fuel Tank - North of the 180 Drough stormwater drain into East Creek Yes Diesel Generator Building .
22? Field Administration Diesel Generator Fuel Tank - East of the 200 On ground into West Creek No Field Administration Building
- 23. Backup Air Compressor Diesel Generator Fuel Tank "C4" - 200 Through stormwater drain into East Creek Inside a building Southwest of the Turbine Building 24? Drummed Oil - East Side of Mechanical Maintenance Shop 150 Brough stormwater drain into East Creek Area covered by a (Varies) roof
- 25. Transformer ISTX-XNSI A - East Wall of Turbine Building 3,951 nrough oil water separator #3 into East Creek - Yes
- 26. Transformer ISTX-XNSIB - East Wall of Turbine Building 3,951 nrough oil water separator #3 into East Creek Yes
- 27. Transformer ISTX-XNSIC - East Wall of Turbine Building 3,405 nrough oil water separator #3 into East Creek Yes
- 28. Transformer IRTX-XSRIC - East Wall of Turbine Building 7,900 nrough oil water separator #3 into East Creek Yes
- 29. Transformer IRTX-XSRIE - East Wall of Turbine Building 15,300 nrough oil water separator #3 into East Creek Yes
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~ _ _ _ _ . _ _ _ _ _ . ..- _ _ - _ . - _ _ _ _ _ - _ _ _ .~ _ _. _ _ _ _ _ _ _ _ _ . . _ _ . 1 TABLE 3 - ENTERGY RIVER BEND STATION ' INVENTORY OF SIGNIFICANT MATERIALS IN OIL S'IURAGE AREAS' (page 3 of 4) Volume Description (Gallons) Drainage '. Containment !
- 30. Transformer IMTX-XM1 - East Wall of Turbine Building 16,733 Through oil water separator #3 into East Creek Yes ~
- 31. Transformer IMTX-XM2 - East Wall of Turbine Building 16,733 Through oil water separator #3 into East Creek Yes d
- 32. Transformer IRTX-XSRIF - Southwest of the Turbine Building 15,300 Through oil water separator #4 into East Creek Yes
- 33. Transformer IRTX-XSRID - Southwest of the Turbine Building 7,900 Through oil water separator #4 into East Creek Yes !
- 34. Transformer NJS-X2A - Cooling Tower A 234 Into a sump, and then on ground into East Creek Yes
- 35. Transformer NJS-X2B - Cooling Tower A 234 Into a sump, and then on ground into East Creek Yes j
- 36. Transformer NJS-X2C - Cooling Tower C 234 Into a sump, and then on ground into East Creek Yes
- 37. Transformer NJS-X2D - Cooling Tower C 234 Into a sump, and then on ground into East Creek Yes ;
- 38. Transformer NJS-X2E - Cooling Tower B 234 Into a sump, and then on ground into East Creek Yes i
- 39. Transformer NJS-X2F - Cooling Tower B 234 Into a sump, and then on ground into East Creek Yes
- 40. Transformer NJS-X2G - Cooling Tower D 234 Into a sump, and then on ground into East Creek Yes
~
- 41. Transformer NJS-X2H - Cooling Tower D 234 Into a sump, and then on ground into East Creek Yes
- 42. Transformer NJS-X3A - Clarifiers 197 Into a sump, and then on ground into East Creek Yes
- 43. Transformer NJS-X3B - Clarifiers 197 Into a sump, and then on ground into East Creek Yes
- 44. Transformer NJS-X3C - Service Water Area (Hypochlorite 200 Into a sump, and then on ground into East Creek Yes System)
- 45. Transformer NJS-X3D - Service Water Area (Hypochlorite 200 Into a sump, and then on ground into East Creek Yes System)
- 46. Transformer NJS-X4A - Service Water Area (Closed Loop 241 Into a sump, and then through Outfall 009 into Yes System) Grant Bayou
- 47. Transformer NJS-X4B - Service Water Area (Closed Loop 241 Into a sump, and then through Outfall 009 into Yes System) Grant Bayou
- 48. Transformer RCS-XI A - West Wall of Fuel Building 1,260 Into a sump, and then through a stormwater drain Yes (Recirculating MG Set Room) into East Creek 3312R0i.DP
TABLE 3 ENTERGY RIVER BEND STATION INVENTORY OF SIGNIFICANT MATERIAIS IN OIL S'IURAGE AREAS (Page 4 of 4) o Volume Description (Gallons) . Drainage ' Containment
- 49. Transformer RCS-XIB - West Wall of Fuel Building 1,260 Into a sump, and then through a stormwater drain Yes
- (Recirculating MG Set Room) into East Creek
- 50. Transformer STX-XS2A - Circulating Water House 1,490 Into a sump, and then on ground into East Creek Yes
[ 51. Transformer STX-XS2B - Circulating Water House 1,490 Into a sump, and then on ground into East Creek Yes
- 52. Transformer STX-XS3A - River Intake 620 On ground into Mississippi River Yes
! 53. Transformer STX-XS3B - River Intake 620 On ground into Mississippi River Yes
- 54. Transformer STX-XSSA - Service Water Area (Closed Loop 1,270 Into a sump, and then through Outfall 009 into Yes System) Grant Bayou
- 55. Transformer STX-XS5B - Service Water Area (Closed Loop 1,270 Into a sump, and then through Outfall 009 into Yes System) Grant Bayou
- 56. Transformer STX-XGNI A - Main Transformer Yard 100 Through oil water separator #3 into East Creek Yes
- 57. Transformer STX-XGNIB - Main Transformer Yard 100 Through oil water separator #3 into East Creek Yes
- 58. Transformer STX-XGNIC - Main Transformer Yard 100 Through oil water separator #3 into East Creek Yes
- 59. Transformer STX-XGNID - Auxiliary Transformer Yard 100 Through oil water separator #4 into East Creek Yes m
Corresponds to Item 12 on Table 2 and Figure 2.
- Corresponds to Item 8 on Table 2 and Figure 2.
- Corresponds to Item 11 on Table 2 and Figure 2.
- Corresponds to Item 2 on Table 2 and Figure 2.
!
- Corresponds to item 19 on Table 2 and Figure 2.
- Corresponds to item 20 on Table 2 and Figure 2.
5312 mot.DP
TABLE 4 ENTERGY RIVER BEND STATION ANALYTICAL DATA
SUMMARY
FOR OUTFALL 102 EFTLUENT UNITS POLLUTANT MAXIMUM DAILV ' Lit.UE MAXIMUM 30 DAY VALUE l LONG TERM AVERAGE CONC. NO.OF MASS CONC. MASS l CONC. MASS l ANALYSES CONC. MASS l Flow "* VALUE 0.014 VALUE 0.002 l VALUE 0.0009l 92 MOD NA Iron
- 1.00 0.012 0.90 0.02 l 0.70 0.01 l 9 mg/L lbs/ day Copper
- 0.90 0.11 0.80 0.01 0.30 0.002 9 mg/L lbs/ day Temperature (Winter) Ambient
- NA Ambiera* NA 0 NA NA Temperature (Summer) Ambient
- NA Ambient
- NA 0 NA NA MINIMUM MAXIMUM MINIMUM MAXIMUM PH NA* NA* N/A N/A NA NA g NA = Not Applicable There was no discharge at Outrall 102 during the period fmm February t 199.1 hrough January 1995 (tl.e DMR summary period presented in this permit application for other outfalls at the sit all the data included in this table is incorporated from the Form 2C for Outfall 102 from Entergy's previous LWDPS permit application submitted to the LDEQ on Jamary 24,1992.
All flow rate values are based on calculations from three consecutive months ofintermittent discharge from this outfall during a reduced volume, procesa development tdal period. Masses calculated using now values mentioned in footnote (2).
- No heat input to this discharge.
i I 5312R01.DP
TABLE 5 ENTERGY RIVER BEND STATION DMR
SUMMARY
FOR FEBRUARY 1993 - JANUARY 1995 OUTFALL 003 EFTLUENT UNITS POLLUTANT MAXIN1031 DAILY VALUE l MAXI %fU5130 DAY LONGVALUE TER31 AVERAGE l NO.OF CONC. MASS l CONC. MASS l CONC. MASS l ANALYSES CONC. MASS Taal Suspended Solids (TSS) 11.2 0.55 8.6 0.42 l 2.2 0.08 l 140 mg/L lbs/ day oil & Grease 10.0 0.57 3.7 0.21 l 2.1 0.07 l 140 mg/L lbs/ day Flow '" VALUE 0.0707 VALUE 0.0082 VALUE 0.0033 482 MGD NA MINIMUM MAXIMUM pH 6.36 7.58 140 S.U. NA NA = Not Applicable The Maximum 30 Day Value and the long Term Average Value for now rates are calculated based on the days of discharge (days of zero discharge have not been included). l { l l l 5312R01.DP am- - -- _ _ _ - - - - - - - _ _ _ _ . - - - _ _ _ _ - - - _ . _ _ _ _ . - - - _ - - - - - - - _ _ _ - - - _ _ - _ _ _ _ _ _ _ - - - - - - - - _ _ . - - _ - - - _ - - - _ _ - - - _ _ - - _ - _ _ _ - - - - - - - - - -
TABLE 6 ENTERGY RIVER BEND STATION DMR
SUMMARY
FOR FEBRUARY 1993 - JANUARY 1995 OUTFALL 005 EFFLUENT UNITS POLLUTANT A1AXI.41U31 DAILY VALUE l htAXI%tU3130 DAY VALUE l LONG NO. OF TER31 AVERAGE l CONC. htASS l CONC. htASS l CONC. StASS l ANALYSES CONC. htASS Tota! Organic Carbon (TOC) 14.9 5.3 11.8 4. I 8.6 2.5 l 40 mg/L Ibs/ day Oil & Grease 4.9 1.19 l 2.4 1.04 l 1.5 0.45 l 40 mg/L lbs/ day , Flow "' VALUE O.465 VALUE 0.057 VALUE 0.035 l - 237 MGD NA MINIMUM MAXIMUM i pH 7.23 8.48 40 S.U. NA NA = Not Applicable The Maximum 30 Day value and the long Term Average Value for now rates are calculated based on the days of discharge (days of rero discharge have not been included). l l 53I2R01.DP
l t TABLE 7 ENTERGY RIVER BEND STATION DMR
SUMMARY
FOR FEBRUARY 1993 - JANUARY 1995 OUTFALL 006 i EFR.UENT UNITS POLLUTANT MAXIMUM DAILY VAll'E MAXIMUM 30 DAY VALUE LONG TERM AVERAGE l CONC. MASS CONC. l MASS l CONC. MASS l ANALYSES CONC. MASS Total Organic Carbon (TOC) 13.8 58.7 10.8 49.7 6.4 14.4 43 mg/L lbs/ day Oil & Grease 4.6 14.34 4.4 14.34 l 1.6 3.99 43 mg/L lbs/ day Flow "' VALUE 8.055 VALUE 0.718 VALUE 0.180 444 MGD NA MINIMUM MAXIMUM PH 7.20 8.66 43 5.U. NA l NA = Not Applicable l The Matinuam 30 Day Value and the f.ong Term Average Value for now rates are calculated based on the days of discharge (days of zero discharge have not been included). l 5312R01.DP
TABLE 8 ENTERGY RIVER BEND STATION DMR
SUMMARY
FOR FEBRUARY 1993 - JANUARY 1995 OUTFALL 007 EITLUENT UNITS M)LLUTANT MAXIMOI DAILY VALUE l MAXINID130 DAY VALUE l IDNG TERM AVERAGE NO.OF CONC. MASS l CONC. MASS l CONC. MASS l ANALYSES CONC. MASS Total Organic Carbon (TOC) 14.5 94.2 12.5 76.9 8.6 25.8 46 eng/L Ibs/ day Oil & Grease 13.t. 83.7 2.2 12.9 1.3 3.9 46 mg/L Ibs/ day flow "' VALUE 8.625 VALUE 0.862 VALUE 0.303 301 MGD NA MINIMUM MAXIMUM pH 7.84 8.99 46 S.U. NA NA = Not Applicable The Maximum 30 Day Value and the Img Term Average Value for flow rates are calculated based en the days of discharge (days of zero discharge have not been included). 5312R01.DP
MM W 'W W " unw - aus - wour - M M M W W W M M TABLE 9 ENTERGY RIVER BEND STATION DMR
SUMMARY
FOR FEBRUARY 1993 - JANUARY 1995 OUTFALL 009 EFT 1 CENT UNITS POLLUTANT MAXI %1UN1 DAILY VALUE NlAXIN1UN130 DAY VALUE l LONG TERN 1 AVERAGE NO.OF CONC. MASS l CONC. MASS l CONC. MASS l ANALYSES CONC. MASS Total Organic Carbon (TOC) 16.4 9.5 l 11.7 8.9 7.9 3.9 40 mg!L lbs/ day Oil & Grease 3.1 1.79 l 2.7 1.66 1.3 0.60 40 rng/L lbs/ day Flow "* VALUE 1.739 VALUE 0.176 VALUE 0.051 301 MGD NA MINIMUM MAXLMUM pH 7.49 8.78 40 S.U. NA NA = Not Applicable
"' The Matimum 30 Day Value and the long Term Average Value for flow rates are calculated based on the days of discharge (days of zero discharge have rut been included).
5312R01.DP
_ ww v- -- - , - - - . - _ ~ . -- - TABLE 10 ENTERGY RIVER BEND STATION BIOMONITORING TEST RESULTS Toxicity Test Results from the January 6 and November 10,1994 Molluscicide Applications Ceriodaphnia dubia Chronic Toxicity Test Hyalella azteca Solid-Phase Toxicity Test Survival NOEC Reproduction NOEC Percent Survival (%) Growth (Avg Dry Weight in mg) Sample I 4 Dates Untreated Treated Untreated Treated Control Untreated Treated Control Untreated Treated MMM - M MMM \ 01/06/94 100 100 N/A ' N/A' 84 75 68 0.065 0.072 0.053 11/10/94 100 100 100 100 98 80 98 0.223 0.190 0.166 Toxicity Test Results from the August 17,1995 Molluscicide Application Sample Date Daphnia pulex Survival NOEC Pimephales promelas Survival NOEC ' 8/17/95 0.8 % 0.8 % NOEC = No Observed Effect Concentration Control did not meet acceptable performance criteria for the reproduction test endpoint. 3312ROI .DP
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@ PERMITTED OUTFALL f .; WEST FELICIANA PARISH _
NOTES: THE LOCATION SHOWN FOR OUTFALL 004 WILL BE USED AFTER THE NEW SEWAGE 5' ASSOCIATES INC* 8 TREATUENT PLANT COMMENCES OPERATION BATON ROUGE, LOUISlANA BASE M AP TAKEN FROM U S G S 7 5 MINUTE TOPOGRAPHIC M AP ' ELM PARK LA- DRAWN MPC/MAC IAPPROVED MHS DATED 1965. PHOTOREVISED 1972. AND PORT HUDSON. LA DATED 1963 CHECKED DP 1DAIE SEPTE5[BEi11,I995 ij PHOTOREVISED 1980. AT A SCALE OF1 24.000 g!SHFFT OF i DWG NO A$1A02 06
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l . L, APPENDICES l l l l l l 5312R01.DP l
m' . M e APPENDIX A L U.S. EPA APPLICATION FORM 1 l t O A 5312R01.DF
r-Please pnne or type in the tenahaded seems only Form Appmved OMB No. 20dMM6 Approval expsres 7 3148 Ollike areas sparedpr stue type. i.e.,12 ekarerserskscin) l FO4f U.8. mVIRONhuNTAL PROTirTION AGl'NCY 1. EPA I.D. NUMilER
} {p GENERAL INFORMATION connandasal r,,mur r,o,va,,,
p LAD 070664818 l g
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WWAL thed she *Gearrellastnncuens* bepre snarw,gy i i nI ,. .. aAug. smt" GENERAL INSTRUCTIONS If a preprinted label has been provided, affix it in the designated I. EPA I.D. NUMBER space. Review the information carefully;if any ofit is incorrect. cross through it and enter the correct data i. the appropriate fill-in IH. FAOW N, area below. Also,if any of the preprinted data is absent (the arra PLEASE PLACE LABEL IN TIIIS SPACE '" d' A *.f e' 'ab 'Pa" ""' '^' l/*'"'d8" 'A 'ha"'d appear). please provide is tr> the proper fill-to area (s) below. If the V. FACILITY MAILING label is complete and correct you eined not complete items I, HI, V. ADDKFSS and VI (excepr Vf-B sskch muss be complesed regardless). Conglese all items if no label has been provided. Refer to the VI. FACILITY LOCATION instructions for detailed item descriptions and for the legal authorizations under which this data is collected INST RUCTIONS: Complete A through J to deternune whether you need to subnut any pernut applacation forms to the EPA. If you answer *yes' to any gecations. you must submit this form and the supplemental form listed in the parenthesis following the question. Mark *X* in the box in the third column if the supplemental form is cttached. If you answer *no* to each question, you need not submit any of these forms. You may answer *no" if your activity is excluded from permit requirements; see Section C of the instmetions. See also. Section D of the instructions for definitions of hold-faced terms. adAng *E' hiARE *E* SPECIFIC QUFATIONS na SPECIFIC QUESTIONS nu anerWrD AnArMD A. la the fud2y a put.hcly owned treaunant works whach resuhe en a tt Duas or wdl Gus incddy (auher ensung or pinpasaf) nachale a ducharge to waters of the U.S.? (FORM 2A) X concentrated maimal feeding operataan or aqueue ananal productaan X facilny which ruuks in a ducharge to waters of the U.S.? (10RM M .. x l C. la unas a lEsidy whsh curnatly results a decharges to waters of the y D. 4 tius a proposed facday (osher shan shase describai an A or b l g'*
' 9 mbove) whih wiu seauk sa a decharge to weiers of the U.S.?
U.S. ether than thoes deocnbod in A or B above? (10kM 2C) l y ,, , (10RM 2D) ,, , L. Duse or wdl than facday treat, esore, or depues of hazardous wassen? F. Do you er wdl you anject at thas facddy nulusinal or munscipal (10RM 3) X effbent below the lowennost stratum contammg. withm one quarter X mGe of the well bore, undergiound sources of drmkmg water? n ,. * (f0PM 4) i t , G. Do you or wdl you anject at uus facdity any produced weser or other H. Do you or wdl you mject at uus taudny ikala for specul pruccasca fluida wbkh are broug*it in the surf ace a connectaan wah y such as mmmg of sulfur by the Frasch pan cas, solutaan mining of y convenuonaled or natural gas producuan, mject ihads used for mmerals. in estu combusuon of fomed fuel or recovery of geothermal enluuned twovery of oil or natural gas, or inject ikkle for ohnuse of
)
energy?(FORM 4) I hquid bvdencartums?(I;ORM 4) , m , , , ,, l
- 1. Is the fa6ddy a proposal stauonary sourm whs6b as one of the 28 J. Is uus lacday a proposed ===ry source which a !%Ol' une of the mulusanal categones hated an the instructions anJ whib will y 28 industnal categones lasted in the metructama and which will p.nentaally emat 100 tons per year of any aar poButant regulated under y
potenusDy emit 250 tons per year of any air pollutant regulated under the Clean Aar Act and namy affect or be located in an annarunent area? the Clena Air Act and may affect or be locatal m an anamment area? eem (IUl(M 5) (R)RMS)
-hem h SKIP ENTERGY OPERATIONS, INC. - RIVER BEND STATION c
A. S AML A IIILL flast, first, a tutt) 8. PilONE farra code a no.) T HOLMES, JEROME, SUPERINTENDENT, CllEMISTRY 504 381 4602
.. i. .. . . . ., ., _
A. bl RELI OM P.O. DO A c T POST OFFICE BOX 220 n .. e l 11. CITY OR TOWN C. STATE D. ZIP CODE y ST. FRANCISVILLE LA 70775
- 5. I<Aril ITY I.OC ATION A. 5 . . O . 60. OM O alLR SPetitiC IDLNlitiLR c
T 5485 U.S. lilGHWAY 61 NORTH n i. , B. COUNTY NAME WEST FELICIANA PARISH C. CITY )R 11[)WN D. STATE E. ZIP CODE h CS$7E h ST. FRANCISVILLE LA 70775 063 u . . < r. . . . 413T01.DP Computer Reproduction EPA Form 3510-1 (Rev.1040) CONTINUE ON REVERSI
CONTINUED FROM Tile FRONT A. nust n. u nsu
- -L (specV1) l'pectfy) 49g3 L I7 ELECTRIC SERVICES - STEAM ELECTRIC 7 N/A I ,. u - e. .. i. . i.
I C. THIRD D.M WRTH Qr (spec (fy) ,J,,_ (spectfy) l l7 N/A 7 N/A l [ .. .. .. - .. A. N A%1E B. Is the name ILsted in item ENTERGY OPERATIONS, INC. * ** BYES NO N C. STATUS OF OPERATOR (Duer she aprovprsase leuer Inw she ansswr bar;if 'Osher*, spectfy.) D. YllONE (area code & no.) l F= FEDERAL M = PUBLIC (osher thanfederal or state) (spectfy) J,_ p $9y 35g 5999 5= STATE O = UTHER (spec {fy)
^
P = PRIV ATF = is i. - u i. m u v E. SI REEL 6k P.O. BOA POST OFFICE BOX 31995 F. CITY OR TOWN G. STALE II. ZIP CODE lA. INDIAN LAND l la the facshty located on indan lands? h JACKSON MS 39286 YES rm Lil NO es u . = m a c si a E F%IN'll%G ENVIRON %1FNTAl, PFR\1tTS E l A. NYDES (INscharges to Surface Mater) D. YSD (Air Emissions from Proposed Sources) j [ LA0042731 19 I4 99 19 W 19 le 79 18 - D B. UIC (Underground injection of Hakis) E. 011LER (specify) (spectfy] r v s e v 0 U 9 WP 0409 LOUISIANA WATER DISCI (ARGE PERMfT H M IT IS - N 88 le IT IS - N C. RCRA (lla:ardous %'astes) E. OTilER (specify) (spectf>) r e i r e i e g LAD 070664818 , RBC36201 CWA SECTION 404 (USAGE) u u et is - m as u is u - w Ottach to this spgdscatan a topogrugduc map of the area extending to at least one unile beyond propeny boundancs. 'the map must show the outhne of the bility, the locatina of each of its esisting and propned intake and discharge sinactures, caech ofits hasardous waste treatment, storage or dispaal facilities, and emb me4l where it isdicts fluids underground. Include all springs, rivers and other surface water bodies in the map area. See instructions for prwie Quiresnesats. XII. % ATl'RF UV Itt'NI%VNN tromnde a bnef desenntuns) COMMERCIAL GENERATION AND SALE OF ELECTRIC POWER. ( tenify under penalty oflaw that I have personally examined and am familiar avish the Wormation submitted in this application and all attachments and that, based p my inquiry ef those persons immediately responsiblefor obtaining the Wormation contained in the application, I believe that the Wormation it true, accurate and danplete.1 em amare that there are significant penaltiesfor submitting false Wormation, including the possibility offine and imprisonment. ). N AME & OtTICIAL TITLE (type or print) B. SIG N ATCRE C. DATE SIGNED s "lCll AEL B. SELIMAN ENERAL M ANAGER. PLANT OPERATIONS . , s ( W k( g ' [ 'cf i ' e ,. @ 3T01.DP Cornputer Reproduction EPA Fonn 3310-1 (Rev.10-80) Reverse
- ns omemnient Pnnung off.x avas wasso2wi
L-r u L F L [ [ [ [ APPENDIX B [ U.S. EPA APPLICATION FORM 2C [ [ [ [ [. ( [
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VPA 1.D. NUMnty twpyjmm jum j of yo, jj Fom 4p m f Omit k 204Mxu P>ame prms or type a the unshaded crees ce4 #P**l esmm 7 JI-M LAUD 70664818 yog U.S. INVIRONhn'NTAL PROTECTION ACDiCY APPLICATION IVR PERMIT TO DISCllARGE WASTEWATER 2C EXISTING MANUFACTURING, COMMERCIAL, MINING AND SILVICULTURAL OPERATIONS
- Comolidated Permits Program
[ L For each outfall, list the latitude and longitude ofits kwation to the nearest 15 seconds and the name of the receiving water. A.OUTFALL B. LATTTUDE C. LONGITUDE NUMBER l D. RECEIVING WATER (name; g,,,j I. Din 2. MIN. 3. 51%. 1. Din 2. WN. 3. $bC. 001 30 43 43 91 21 13 l MininippiRiver 002 30 45 21 91 19 46 l Mississippi River (via Outfall 001) 102 30 45 21 91 19 46 l Mininippi River (vis outfal1002/001) 003 30 45 20 91 19 49 l Granta Bayou (via East Creek and Outfall 006) then to Mississippi Ri ver 004 30 44 52 91 19 50 l Missinippi River (via Outfall 001) 005 30 45 06 91 19 38 l Granta Bayou then to Mississippi River A. Attach a line drawing showing the water flow through the facility. Indicate sources ofintake water, operations contributing wastewater to the effluent, and treatment umta labeled to currespond to the more detailed descriptions in item B. Construct a water balance on the line drawing by showing average Dows between intakes, operations, treatment unita, and outfalls, if a water balance cannot be determined (e.g.,for renain nunmg acavines), provide a pictonal description of the nature and amount of any sources of water and any collection or treatment measures See Figure 3. B. For each outfall, provide a description of: (1) All operations contributing wastewater to the effluent, including process wastewater, sanitary wastewater, cooling water, and stormwater runoff; (2) The average flow coninbuted by each operation; and (3) The treatment received by the wastewater. Continue on additional sheets if necessarv.
- 3. OUT. 2. OPf7 STloNN CONTRIBtT NG How 3. TRFATMFNT
^ "
4"') a. OPIR ATION dur) a. DFSCRII410N (mdade unirs) TABLE 2C-1 Coohng Tower Blowdow n (and 2144 gpm Dechlorination 2E 001 momtored Outfalls 002,102, and 004) Discharge to Surface Water 4A low-volume Treated Westewater 27.1 gpm Multmwdia Filtration; IQ (Interminent) Neutralization; lon-e xc hange; IT 2K Re-use/ Recycle of Treated Eftluent; 2J 4C Dixharge to Surface Water 4A Chemical Metalsleamng Wastewster 0.6 gpm Neutralization, Chemical Precipitation; 2K 2C ~ (Intermittent) Carbon Adsorption. 2A Vaeuum Filtration /Landfilhng of sludge; 5U SQ Dissharge to Surface Water 4A Nan radioastive Fhur Drams 2 3 gpm Oil / Water Separation; GO.1 and OfWater Separators. includmg (IntermittenO Disharge to surface water 4A Stormw ster Treated Samtary Westem sier 11.1gpm Screening; Pre-aeration. IT JE Activated Sludge; Slow Sand Filtration; 3A 2H 004 Disinfection (UV-light) IV Stormwater Runoff from Matenals 7.64 gpm Discharge to Surface Water 4A g Storage Area (Intermittent) Intermittent Treated Sanitary Wastewater Normally 0 See above for Outfall 004 Normally Routed through Outfall 004 OH1CIAL USE ONLY (cAu u guate/mes secuacgone.o Computer Reprmluction EPA Foriu 3510-2C (Res. 245) PAGE I OF 4 CONTINUE ON REVERSE 5313T02.DP
w EPA l.U. NLMuty gn,pyfmni hem / of fem 1) fom APPmwd OMB M Me&OOM Please prena or type m the uanhadalsnes caly. APPm*83 e2Poref 7-ll M MD()7()664818
# U1INVIRON311NTAL PROTECTION AGENCY 2c NYDt%
EPA APPLICATION R)R PERMIT TO DISCilARGE WASTEWATER sususo muumcrumsc. cosisitacm, muisc - siaicuuum oi cuuoss l Consolidated Permits Program i s For each outfall, list the lautude and longitude of its location to the nearcat 15 acconda and the name of the receiving water. l A. OLTTFALL B. LATfTUDE C. II)NGrTUDE NUMBER D. RECEIVING WATER (name) 4,g 1. Din 2. MIN. 1 $tr. 1. Din 2. Mj N, 3. SEC. l (K)6 30 45 12 91 19 29 Granta Bayou (via East Creek) then to Minaissippi River t l 007 30 45 02 91 19 50 Granta Bayou (via West Creek) then to Missianippi River l 008 30 45 21 91 19 46 Graata Bayou (via East or West Creek) then to Mississippi River l l 009 30 45 32 91 19 39 Granta Bayou then to Mississippi River l Note: Coordmates for Outfall 004 are for new location due to construction of new sewage treatment plant. l k 1 I I l l I A. Attach a ime drawing showing the water flow through the feedity. Indicate sources ofintake water, operations contributing wantemster to the ef!luent, and treatment unita labeled to correspond to the armre detailed descriptions in item B. Construct a water balance on the line drawing by showing average Cows ! between intakes, operations, treatment unita, and outfalla. If a water balance cannot be determined (e.g.,for cenam mmmg actmnes), provide a pictorial f description of the nature and annunt of any snurtes of water and any collection or treatment measures See Figure 3. i B. For each outfalt provide a description of: (1) All operationa contributing wastewater to the effluent, including pruccan wastewater, aanitary w antew ster, cooling water, and stormwster runoff; (2) The average now contributed by each operation; and (3) The treatment received by the wastewater. Continue on [ additional sheets if necessarv. l
- 1. OLIT. 2. OPf 7 4TIONN) CONTIMittTING Ft OW 'i. TRFAITIFN" l 4d a. Ol41A110N diar) a. DENCRIPTION 6md.de aush) TAltlE 2C-1 Stormwater Rurmff from East Side of 75.7 gpm Discharge to Surface Water 4A l
Plant (and nomtored (Intermittent) l 00n Outfalle 003 and 00M) l Air Compreamr Condennaie 16 gpm Discharge to Surface Water 4A l Reserse Osnmsis Reject 25 gpm (Interouttent) Dmharge to Surface Water 4A Stormwater Runoff from West SiJe of 86 8 gpm Discharge to Surface Water 4A l Plant (and momtored outfall 00x) (Intenmiteno 007 l l Mamtenance Hydrostatie Test and 5 gpm Sc reening; iT l Flualung of Pipmg Systema. Vessela, (intenmtteno Dmharge to Surface Water 4A 008 l and Automatie Spnnkler Systems l l Stonnwater Runoff from 14.6 gpm DmharFe to Surface Water 4A l 009 Coohng Tower Yard (Intennineno l l l I I l l l OillC1AL USE ONLY tefnuent guatelinee subicsones) Compter Reproduction EPA Forut 351tL2C (Res. 2-NS) PAGE I OF 4 (Continunh CONTINUE ON REVERSE 53l3T02.DP
CONTINUED FROM Tile FRONT - C. Except for anorm runoff, leaks, or spills, cre eny of the disch2rges diacribed in hems U-A or B inerminent or seasonal? O YES (complew thefolloweng sable) NO (go so Secsioss111)
- 3. 51 LEX)LWCY 4. gww I. OLTFAIL 2. OPERA 110N(s) a. DAYS b.MONT1H a. FIDW RATE b. TOTAL VOLUME NUMBER CONTit18ttrING PMW PER win Pt2 YEAR (iss m,/) c. DUR.
(M M wiss) (f**@ (***D i.umo 7:ans M4AXInnB4 gg,,
. ..) .-. ) a.a-uma tr.ans t anexu,dvas
- a-, -r r 002 Low-volume Treatad 7 12 0.052 0.497 52000 gals. 497000 gals. 365
! Wastewater 102 Metal Cleaning Wastewater 7"' 3'" 0.0009 0.014 900 gals. 14000 gals. 92 006 Reverse Osmnais Reject 1.5 12 0.007714 0.036 7714 gals. 36000 gals. 78 008 Hydrostatic T. sting and 0.25 12 0.0074 0.0638 7400 gals. 63800 gals. 12 r Flushing of Piping Systems l
"' Outfall 102 has only discharged for three consecutive months out of 10 years of facility operation.
f L A. Does an efnuent guideline limitation promulgated by EPA under Section 304 of the Clean Water Act apply to your facility? b YLS (complew sum fil-B) NO (go so Section IV) B. Are the limitations in the applicable e(Duent guideline expressed in terms of production (or odier measure of operation)? YES (complew hem Ill-C) $ NO (go to Section (V) C. If you answered "yes" to item Ill.D, list the quantity which repre3 cats an actual measurement of your level of production, exprissed in tie terms and units umi in the applicalde efflueet guideline, and indicate the affected outfalls.
- 1. AVERAGE DAILY PRODUCTION l 2. AHTCTED Ot3TALIE
& QUANTrrY PER DAY b. IfNIT1 or MEA $l'RE 8 "
- e. OPERATION. PRODtCT, MATERIAt ETC.
wem N/A r b r L IV. IMPROVEMENTS A. Are you now required by any Federal. State or local authority to meet any implementstion schedule for the construction, upgrading or operation of wastewater treatment equiprnent or practices or any other environmental programs which may affect the discharges described in this application? 7his includes, but is not limited to. permit conditions, administrative or enforcement orders, enforcenwnt compliance schedule letters, stipulations, court orders. and grant or kian conditions. 0 YES (complew thefolios.ing sable) NO (go so hem IV-8)
- 1. IDENTIFICATION OF CONDITION, 2. AFFECTED OUTFALLS 3. BRIEF DFSCRIPTION OF PROJECT 4.11NAL COM. l AGREEMENT, ETC. PLIANCE DATE l a h0. b. Sol'RCE OF DISCitARCE a. RE- b. PROW
( N/A QUIRED JECTED l B. OPTIONAL: You may attach additional sheets describing any additional water pollution control programs (or other environmentalprojeczs which may qfect your discharges) you now have underway or which you plan. Indicate whether each program is now underway or planned, and indicate your actual or planned schedules for construction. MARK "X" IF DESCRIPTION OF ADDITIONAL CONTROL PROGRAMS IS ATTACilED Computer Reproduction EPA Form 3510-2C (Rev. 2-85) PAGE 2 OF 4 CONTINUE ON PAGE 3 5313T02.DP
L ETA i.D. Nt%fM3t tasyfom han J of fem )) fem APPowd cult W 206040u CONTINUED ih0M PAGE 2 APPNd 88Pm 7-JIM LAD 0070664818 V. INTAKE AND EFFLUENT CilARACTERISTICS A, B. A C: See instructions before proceeding - Complete one act of tables for each outfall - Annotate the outfall number in the space provided. NOTE: Tabica V-A, V-B, and V-C are included on separate ahecta numbered V 1 through V-9. D. Une the space below to list any of the pollutanta listed in Table 2C-3 of the instructions, which you know or have reason to believe is discharged or may be discharged from any outfall. For every pollutant you list, briefly describe the reasons you believe it to be present and report any analytical data in your posseaaion.
- 1. POLLUTANT 2. SOURCE 1. POLLUTANT l l 2. SOURCE N/A L
F L VI. POTENTIAL DISCIIARGES NOT COVERED BY ANALYSIS la any pollutant listed in Item V-C a substance or a component of a substance which you curantly use or manufacture as an intermediate or final pnWuct or byproduct? YES Gus allsuch pollutams below) NO (go so Iwm U-B) N/A e L [ c E E [ [ Congvter Reproduction EPA Form 3510-2C (Rev.245) PAGE 3 OF 4 CONTINUE ON REVERSE 5313T02.DP
CONTINUED FROM TIIE FRONT Do you have any knowledge or nason to believe that any biological test for acute or chronic toxicity has been made on any of your discharges or on a nceiving j water in niation to your discharge within the last 3 years?
' D YES 6dentify Ae sessis) and describe sheir purposes below) NO (go so Secdon WII)
- See Section 6.0 of document and Table 10.
) ) Were any of the analyses reported in hem V performed by a contract laboratory or consuhing Grm? YES (list she name, address, and selephone number of. and pollutants NO (go so Secdon IX) analyzed by, each such laboratory or}rm below) A. NAME B. ADDRESS C. TELEPilONE D. POLLUTANTS ANALYLED (area code & no.) (!!st) Inchcape Testing Services 7979 GSRI Avenue (504) 76?-4900 All pollutants analyzed on Forrn Baton Rouge, LA 70820 2C except pH, TRC, and FAC and all pollutants at Outfall l 002B. l Barringer Laboratories, Inc. 15000 W. 6th Ave. (303) 277-1687 All pollutants at Outfall 002B, l Suite 300 except pH, TRC, and FAC. Golden, CO 80401 m l I L I J cersde underpenalty oflaw that shis derument and all anachments wre prepared under my directiore or supervision in accordance Mth a system designed *o assure has qual @ed personnelgaaer and evaluate de ht formation submisted. Based on my inquiry ofsie person orpenant who manage Ae syssem or Aase perwns direcdy responsiblepr gaAering at i$rmasion Ae iprmadon submined is, so the bess ofv y knem4 edge and belief, trw accurase, and complese. I am enwrr Ans Aers are sign @ cant penaisiespr submlaingfalse i$rmanton, including Ae posslNitr of)ine and imprisonmensjor knoMng nolagons. A. N AME A OFFICIAL TITLE (rype orprint) B. PHONE NO. (arra code & no.) Michael B. Sellman, General Manarer. Plant Operations [ (504) 381-4200 L C. 51 NATURE / D. DATE SIGNED f Casuputer Reproduction EPA Form 35 0% ' Y ' h I VW C (Rev. 2-8'5) ' PAG 9 $ OF 4 L 5313TD2.DP
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2b. l 2c TESTNeG NMyliD KMWED e teANIRSUtf D4ftY V ALE E EFFLENT 4. LSefTS E Wff ASEE (UPTIOetAL) b === So DAY WALE a LOceGTEms AWEMAEQd. 900. OF a b. a LTA VALE b. 900. OF Afe) CAS NR KQUIED PEKKT ABENT (1) CONC. (2) RAASS (1) CONCL ANAL M S mee (2) RAASS (1) CONCL (2) RAASS OJBOCL [1) CONC. (F) RAASS AfGALMS 2E D-n-Butyt Phthaps*e (S4-74-2) X NA NA NA NA NA NA 0 NA NA NA NA NA 778 2.4-Onitratoluene(171-14-2) X NA NA NA NA NA NA 0 NA NA NA NA NA 22 2.Q-Dnilrotoluene (506-20-2) X NA NA NA NA NA NA O NA NA NA NA NA 298 D-n-OctytPhthalate(117-84-0) X NA NA NA NA NA NA 0 NA NA NA NA NA 3W. 1.2 -CV.i, .ymyeaztne (122 7) X NA NA NA NA NA NA O NA NA NA NA NA 318. Fluoraiib . (208-44-0) X NA NA NA NA NA NA O NA NA NA NA NA 328. Fluorene (88-73-7) X NA NA NA NA NA NA 0 NA NA NA NA NA 32 Hexachlorotxarvene(118-74-1) X NA NA NA NA NA NA 0 NA NA NA NA NA 348. Hexachlorobutattene (87-88-3) X NA NA NA NA NA NA 0 NA NA NA NA NA 3"B. Hexachawy@adene(77-47-4) X NA NA NA NA NA NA O NA NA NA NA NA 3EB R64vdi ne (87-72-1) X NA NA NA NA NA NA 0 NA NA NA NA NA 378 Indeno (1.2.3-cq ry (193-39-5) X NA NA NA NA NA NA O NA NA NA NA NA 3EB. bvp;-mm (78-59-1) X NA NA NA NA NA NA 0 NA NA NA NA NA 3W. N.p;a;.nene(91-20-3) X NA NA NA NA NA NA O NA NA NA NA NA 42 Nevue: -= (98-95-3) X NA NA NA NA NA NA 0 NA NA NA NA NA 418 N-Ntrosod.oenyamine (82-75-9) X NA NA NA NA NA NA O NA NA NA NA NA 429. N-NititMod-n-Propytamme (821-04-7) X NA NA NA NA NA NA 0 NA NA NA NA NA 438 N-Nev=.4p;-ytamine (88-30-6) X NA NA NA NA NA NA 0 NA NA NA NA NA 448. Phenani =(85-01-8) X NA NA NA NA NA NA 0 NA NA NA NA NA 450 ry-m(129-00-0) X NA NA NA NA NA NA NA O NA NA NA NA em 13,4-Trichlorobenzene(120-82-1) X NA NA NA NA NA NA 0 NA NA NA NA NA Pane c- _n 1P. Aktin (309-00-2) X NA NA NA NA NA NA 0 NA NA NA NA NA 2P. alpha-BHC (319-84-8) X NA NA NA NA NA NA O NA NA NA NA NA 3P beta-BHC (319-85-7) X NA NA NA NA NA NA 0 NA NA NA NA NA 4P. gamma-BHC (58-89-9) X NA NA NA NL NA NA O NA NA NA NA NA SP. delta-BHC (319-88-8) X NA NA NA NA NA NA O NA NA NA NA NA 6P. Chlordane (57-74-9) X NA NA NA NA NA NA O NA NA NA NA NA TP. 4.4-DDT (50-29-3) X NA NA NA NA NA NA O NA NA NA NA NA SP. 4,4-CCE (72-55-9) X NA NA NA NA NA NA 0 NA NA NA NA NA 9P. 4,4-000(72-54-8) X NA NA NA NA NA NA O NA NA NA NA NA 10P. Dekttri (80-57-1) X NA NA NA NA NA NA 0 NA NA NA NA NA 11P. alpha-Endosuran(115-29-7) X NA NA NA NA NA NA 0 NA NA NA NA NA 12P. beta-Endosuran (115-29-7) X NA NA NA NA NA NA 0 NA NA NA NA NA 13P. Endosutan Scrate (1tIli-07-8) X NA NA NA NA NA NA 0 NA NA NA NA NA 14P. Endin(72-20-8) X NA NA NA NA NA NA O NA NA NA NA NA 15P. Endrtn Aldehyds (7421-93-4) X NA NA NA NA NA NA 0 NA NA NA NA NA 18P. Heptachk (78-44-8) X NA NA NA NA NA NA O NA NA NA NA NA 17P. HeptachlorEpoxide(1024-57-3) X NA NA NA NA NA NA 0 NA NA
.NA NA NA l 1BP. PCB-1242(53489-21-9) X NA NA NA NA NA NA 0 NA NA NA NA NA 19P. PCB-1254 (11097-09-1) X NA NA NA NA NA NA NA NA NA O NA NA 20P. PCB-1221 (11104-28-2) X NA NA NA NA NA NA 0 NA NA NA NA NA 21P. PCB-1232 (11141-18-5) X NA NA NA NA NA NA O NA NA NA NA NA 22P. PCB-1248 (12872-29-6) X NA NA NA NA NA NA O NA NA NA NA NA EPA Form 3510-2C@er 2-es) PAE Y-4 Cor*%md .
____m___m- - - ^ - '" -
s c ENTERGY OPERATIONS, INC. - River Bend Station - l EPA ID. IIUISER IN l lOUTFAliOE M MM 2 at 2 b. R EFLENT l 2 e. 4.455TS ' 5. WIT M P _j
- 1. POLLUTANT TESTeeG EtKDEDEELEMED a -- me DNLY WALLE b. tam at DAY VALIE et LDIOS TEgne meEftREETAS a. DOCL W a- b. -- tLTA M tE b. NO. OF AeE3 One ammst REQueED PESENT AstaENT (1) CX3NC. A GARSS (1) N $35IAA38 (1) N_ RfGAL M E COII(L (E) teRSS EARSS tt) - 55 IARAS AIGALMS z!IP. Pm-1280(11095-82-5) X - NA NA NA NA NA NA O NA NA NA NA NA 2eP, PW-1016(12674-11-2) X NA NA NA NA NA NA 0 NA NA NA NA NA 25P. Tonachone(8001-:5-2) X NA NA NA NA NA NA O NA NA NA NA NA
~ M 0uess 002 consise of weets strenne which are htermetenty routed to two low-vonsme waste treatment systems (non-radoacuse treatment syuum (Otx2A) and low-level redoersve sentment synesm (0029)].Tte -
docherges from each treatment system are separate 4 sampled and ana%2ed. The separate resuas are combined (flow-weghte4 and charadertand as tes finat renues br OutlaB 002. ' '
- M The hdMtssal anaggest resues are shown for tioth Outfans 002A and 002B (In parentheses) because t is inappropriate to flow-welpt the rueuts t;r this parameter.
O) These parameters are found on Tatde V of the U.S ErWronments Protection Agency Form 2C; however, they are not retsuired to be tested in atz:ordance wth 40 CFR 12221(g(7) and 40 CFR 122 Appencht 0 Tatile u.
'9 2-Chbrosthyhenyt Etter wee not - t is known to hydrolyze in the presence at csute acki Ncess: The defy average comtioned Roar rate of 0.033 MGD (0.018 MOD and 0015 MOD at Outfalls 002A and 0029, respectNefy) obtained daring the samping period trom 8/21/95 gloaB) and 6/22515 (002A) was veed to calcunste the mese for thoes parameters (esce# br mercury) br wNch on4 one labora*ory analysis was performed.
A sampts br mercury analysis was toended on IV2E95 at Outfat 002A. The flow rate - .g to this sample was 0022 MOD. For Mercury, the resuas tem the two separate samping events (one at 002A on IW29MI5, and the other at 00s on EV21/95) are comtsined (flow-weighted to get the final resuts at Outist 002. The montNy DMR forms br 24 months (February 1993 through January 1995) lbr flow data and 12 montts (February 1994 throu@ January 1995) br ad other parameters were used to calculate the Maulmum Dafy Value, neenimum 30 Day Value, and the Long Term Average Value for those parameters that are routinely monitored pursuert to Entergy's NPDES and LWDPS permes.
- As anahecal results reported wth s ' lass tharf sigi (c) were other (t) not detected in the ef!Iuent sample et or above the analytical method detection Ilmt achieved by the appecatHe luboratory analyGcal methoJ or (2) riot deluded and cpsentillable at the practical casanattlation Imt afflieved by the appleable laboratory analytcal methoti Also, f one of the terJits among the two temples (002A and 00PB) was teos than method detection Omt, the comtilned resut is reported with a *less tharf eigi (<). Further,I a parameter was detected at either 002A or 00EB, then the parameter le betsued present. ,
' NA = Testing not required; no data avanetWe.
- No anahtical data were aveEmble for Total Radum for Outfag 002B die to a laboratory error.
-l l.
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