ML18043A377
| ML18043A377 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 01/03/1979 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML18043A375 | List: |
| References | |
| NUDOCS 7901090247 | |
| Download: ML18043A377 (304) | |
Text
{{#Wiki_filter:PALISADES PLANT U.S. NUCLEAR REGULATORY COMMISSION DOCKET NO. 50-255
e Q_ _ PALISADES PIANT STEAM GENERATOR REPAIR REPORT TABLE OF CONTENTS LIST OF EFFECTIVE PAGES LOEP-1
- 1. 0 INTRODUCTION,
SUMMARY
, AND CONCLUSIONS 1-1
- 1. 1
SUMMARY
OF S'IEAM GENERATOR REPAIR PROGRAM 1-2 1.1.1 REPAIR ALTERNATIVES 1-2
- 1. 1. 2 REHOVAL Ai.~D REPIACEMENT OF THE STEAM 1-3 GENERATORS FROM CONTAINMENT 1.1.3 STEA.i.~ GENERATGR CHARAC~E~ISTICS 1-3
~-
1.1.4
'1.1.5 SAFETY-RELATED CONSIDERATIONS ALARA CONSIDERATIONS 1-'4 1-4 1.1.6 OFFSITE RADIOLOGICAL CONSEQUENCES 1-4 1.1.7 UNIQUE ASPECTS OF PROGRAM *1-4a 1.1*.8 STEAM GENERATOR DISPOSAL 1-5 1.2 IDENTIFICATION OF PRINCIPAL AGENTS 1-5
- 1. 3 10 CFR 50. 59 CONSIDERATIONS 1-6
1.4 CONCLUSION
S 1-6 2.0 REPLACEMENT COMPONENT DESIGN 2-1 2.1 COMPARISON WITH EXISTING COMPONENT DESIGN 2-1 2.2.1 PARAMETRIC COMPARISON 2-1 2.1.2 PHYSICAL COMPATIBILITY WITH EXISTING 2-2 STE.Ai.'1 GENERATOR AND SYSTEi-iS 2.1.3 ASME CODE APPLICATION 2-2 ,* (~ __ j 2.1.~ REGULATORY GUIDE APPLICATION 2-3 i MARCH 1979 REV. 1
PALISADES PIANT SGRR 2.2 ~OMPONENT DESIGN IMPROVEMENTS 2-5 2.2.1 DESIGN FEATURES TO IMPROVi: PERFORMANCE 2-6
- 2. 2. 2 DESIGN FEATURES TO IMPROVE MAINTENANCE 2-11 AND INSPECTION 2.3 SHOP TESTS AND INSPECTIONS 2-12 2.4 STORAGE CRITERIA FOR NEW STEAM GENERATORS 2-12 3.0 BALANCE-OF-PLANT SYSTEM MODIFICATIONS 3-1
- 3. 1 BLOWDOWN SYSTEM 3-1 3.2 RECIRCULATION S~EM 3-1 3.3 §.M'!PLING SYSTEM 3-2 3.~ PRIMARY HEAD DRAINS 3-2 3.5 ~IDE RANGE LE~ INDICATICN 3-3 3.6 MAIN STEAM ISOLATION VALVE CLOSURE SIGNAL 3-3 4.0* REPLACEMENT PROGRAM AND PROCEDURES 4~1 4.1 CONSTRUCTION CONSIDERATIONS LJ-1 4.1.1 SITE PREPARATION 4-1 4.1.2 RIGGING 4-5 4.1.3 RIGGING LOAD SU~PORTS 4-9 4.1.4 CONSTRUCTION-REIATED INCIDENTS 4-9 4.1.5 CONTAINMENT STRUCTURAL CX>NSIDERATIONS 4-10 4.2 EQUIPMENT AND MATERIAL REMOVAL AND REPLACEMENT 4-16 4.2.1 MECHANICAL EQUIPMENT 4-16 4.2.2 INSTRUMENTATION 4-16 4.2.3 ELECTRICAL EQUIPMENT 4-17
- 9'
\ . __ ,,,,/' 4.2.4 PIPING 4-22 MARCH 1979 ii REV. 1
PALISADES PLANT SGRR 4.2.5 CONCRETE AND STRUCTURAL STEEL 4-22
- 4. 2~ 6 COATINGS 4-23 4.3 RADIOLOGICAL PROTECTION PROGRAM 4-23
- 4. 3. 1 SUPPLEMENTAL ACCESS CONTROL 4-23 4.3.2 LAUNDRY 4-25
- 4. 3. 3 CONTROL OF AIRBCRNE RADIOACTIVITY 4-25 AND SURFACE CONTk~INATION 4.3.4 SUPPLEMENTAL PERSONNEL MONITORING 4-26 REQUIREMENTS
- 4. 3. 5 GENERAL ALARA CONSIDERATIONS 4-26 4.3. 6 MISCELLANEOUS WASTE DISPOSAL 4-29 4.3.7 MAN-REM ASSESSMENT 4-31 4.4 DISPOSITION OF OLD STEAM GENERATORS 4-*37
- 4. 4. 1 OBJECTIVES OF HANDLING/DISPOSAL 4-38 OPERATIONS 4.4.2 ONSI'IE STORAGE 4-38 4.4.3 OFFSITE DISPOSAL 4-39 4.4.4 MAN-REM ASSESSMENTS 4-41 4.4.5 RADIOACTIVE RELEASES AND DOSE ASSESSMENT 4-41 ASSOCIATED WITH OFFSITE DISPOSAL
- 4. 4. 6 RADIOACTIVE RELEASES AND DOSE ASSESSMENT 4-41 ASSOCIATED WITH ONSITE STORAGE 4.4.7 ACCIDENT CONSIDERATIONS ASSOCIATED 4-42 wITH ONSITE STORAGE 4.
4.8 CONCLUSION
S 4-43 4.5 PLANT SECURITY 4-43 - 4.6 PLANT SYS'IEMS LAYUP AND STARTUP METHODS iii 4-44
PALISADES PLANT SGF.R 4.7 QUALITY ASSURANCE 4-47 4.7.1 CONSUMERS POWER COMPANY QUALITY 4-q7 ASSURANCE PROGRAM 4.7.2 BECHTEL POWER CORPORATION QUALITY 4-47 ASSURANCE PROGRAM 4.7.3 COMBUSTION ENGINEERING POWER SYSTEM 4-47 GROUP NUCLEAR QUALITY ASSURANCE PROGPAM 4.8 REGULA~ORY GUIDE APPLICABILITY TO REPAIR 4-47 PROGRAM 4.9 SCALE MODEL OF THE PALISADES PLANT CONTAINMENT 4-55 5.0 RETURN .IQ_SERVICE TESTING 5-1 6.0 SAFETY EVALUATIONS 6-1 6.1 FSAR EVALUATIONS 6-1 6.
1.1 INTRODUCTION
6-1 6.1.2 NON-LOCA ACCIDENTS 6-1 6.1.3 LOSS-OF-COOLANT ACCIDENT EVALUATION 6-1
- 6. 1. 4 CONTAINi'1ENT PRESSURE ANALYSIS 6- 1 6.1.5 FSAR EVALUATION CON::LUSION 6-1 6.2 CONSTRUCTION-~TED EVALUATIONS 6-2 I 6.2.1 HANDLING OF HEAVY OBJECTS 6-2 6.2.2 OFFSITE RADIOACTIVE RELEASES AND 6-2 DOSE ASSESSMEN'l' 6.3 FIRE PROTECTION 6-4 6.3.1 EXISTING FIRE PROTECTION 6-4 6.3.2 FIRE PROTECTION DURING THE REPAIR 6-5 PROGRAM 6.
3.3 CONCLUSION
6-8 iv
PALISADES PLANT SGRR 7.0 ENVIRONMENTAL ASPECT§ OF THE REPAIR PROGRAM 7-1 7.1 ~ENERAL 7-1
- 7. 2 RESOURCES CO.NMITT£;.Q 7-1 7.2. 1 NONRECYCLABLE BUILDING MATERIALS 7-~
7.2.2 LAND RESOURCES 7-2 1.2.3 WATER RESOURCES 7-2
- 7. 3 WASTE WATER 7-3 1.3. 1 SANITARY FACILITIES 7-3 7.3.2 LAUNDERING OPERATIONS 1-3 7.4 CONS'IRUCTION 7-3 7.4. 1 NOISE 1-3
- 7. 4. 2 DUS'I* 7-4 7.4.3 OPEN BURNING 7-4 7.5 RADIOLOGICAL MONITORit!Q 7-4 7.6 RETURN TO OPERATION 7 .6. 1 wATER USE 7-4 7.6.2 OPERATIONAL EXPOSURE 7-5 7.6.3 RADIOLOGICAL RELEASES 7-5 8.0 EVALUATION OF ALTERNATIVES 8-1
- 8. 1 INTRODUCTION 8-1 8.2 CONTINUED TUBE PLUGGING AND PLANT DERATE 8-2 8.3 IN-PLACE TUBE SLEEVING 8-2 8.4 IN-PLACE TUBE REPLACEMENT 8-3 8.5 REPLACEMENT WITH COMPLETE UNII§. 8-4 v
- 8.6 PALISADES PLANT SGRR REPLACEMENT OF STEAM GENERATOR 8-6 EVAPORATOR SECTIONS 8.7 MAN-REM CONSIDERATIONS 8-6
8.8 CONCLUSION
S 8-7 9.0 COST BENEFIT ANALYSIS FOR THE DECONTAMINATION, 9-1 STORAGE, AND DISPOSAL OF THE OLD STEAM GENERATORS CONSIDERING ALARA
9.1 INTRODUCTION
9-1 9.2 STEAM GENERATOR IN-PLACE DECONTAMINATION 9-1 9.3 STEAM GENERATOR STORAGE AND DISPOSAL 9-2 9.3.1 LONG-TERM STEAM GENERATOR STORAGE 9-2 ONSITE 9.3.2 IMMEDIATE SHIPMENT BY BARGE 9-2
- 9.3.3 9.3.4 SHORT-TERM STORAGE WHILE UNITS ARE CUT UP FOR SHIPMENT (WITHOUT DECONTAMINATION)
CONCLUSIONS 9-3 9-3
- 10. 0 REFERENCES 10-1 APPENDIX A - RESPONSE TO NRC QUESTIONS OF 4/17/79 A-l(a)-1 2
APPENDIX B - RESPONSE TO NRC QUESTIONS OF 4/19/79 B-l(a)-1 APPENDIX C - RESPONSE TO NRC QUESTIONS OF 5/16/79 C-1-1 I3
- vi Revision 3 July 1979
1
- PALISADES PLANT SGRR LIST OF TABLES TABLE NO. TITLE 2.1-1 STEAM GENERATOR COMPARISON DATA 2.1-2 REPLACEMENT STEAM GENERATOR DATA 2.2-1 STEAM GENERATOR MATERIALS 4.2-1 ELECTRICAL EQUIPMENT AND INSTRUMENTS TO BE TEMPORARILY REMOVED 4.2-2 480 V MOTOR CONTROL CENTER B09 LOAD TABULATION 4.2-3 TEMPORARY AND PERMANENT ELECTRICAL LOADS ASSOCIATED WITH STEAM GENERATOR REPAIR AND THEIR CORRESPONDING POWER SOURCE 4.3-1 TYPICAL PORTABLE SURVEY INSTRUMENT SPECIFICATIONS 4.3-2 MAN-REM ASSESSMENT FOR REPLACEMENT 4.4-1 ACTIVATED CORROSION PRODUCTS AFTER SHUTDOWN 4.4-2 MAN-REM ASSESSMENT FOR OFFSITE DISPOSAL 6.2-1 ESTIMATES OF AIRBORNE RELEASES TO ENVIRONMENT DURING STEAM GENERATOR REPAIR
- 6.2-2 6.2-3 6.2-4 EFFORT COMPARISON OF GASEOUS EFFLUENT RELEASES ESTIMATED SPECIFIC ACTIVITY OF LAUNDRY WASTEWATER ESTIMATED RADIOACTIVE LIQUID EFFLUENT RELEASED DURING THE STEAM GENERATOR REPAIR 6.2-5 COMPARISON OF RADIOACTIVE LIQUID EFFLUENT RELEASES 8.4-1 STEAM GENERATOR REPAIR ALTERNATIVE COSTS 8.8-1 COMPARISON OF MAJOR REPAIR ALTERNATIVES B-2-1 .GENERAL CORROSION RATE IN FAULTED VOLATILE CHEMISTRY MODEL BOILER ENVIRONMENTS B-2-2 TUBE SUPPORT MATERIALS COMPARISON CORROSION 2 TESTING AT HEAT TRANSFER/SUPPORT LOCATION (MODEL BOILER TESTING)
C-1-1 MANUAL WELDING OF RC PIPE WITH MACHINE CLADDING C-1-2
SUMMARY
OF MANHOURS FOR ALL TASKS BY LOCATION C-1-3 MACHINE WELDING OF RC PIPE WITH REMOTE VIEWING 3 C-1-4
SUMMARY
OF MANHOURS FOR ALL TASKS BY LOCATION C-1-5 MAN-REM ESTIMATE C-7-1 MAN-REM ESTIMATE (PREPARATION, INSTALLATION, REMOVAL, AND STORAGE)
- vii Revision 3 July 1979
. --*-*- --- -*-*-------- - - ----*------- ----- -.---~-------------------*-"""'------------~-~-- ----------------~-~ ---*~---...,.--~----~------- -
- PALISADES PLANT SGRR LIST OF FIGURES FIGURE NO. TITLE 2.2-1 REPLACEMENT STEAM GENERATORS 2.2-2 DELETED 2.2-3 DELETED 2.2-4 BOTTOM BLOWDOWN DUCT ASSEMBLY 2.2-5 TUBE SUPPORT TYPES 2.2-6 EGGCRATE ASSEMBLY 2.2-7 BEND REGION TUBE SUPPORT 2.2-8 TUBE SUPPORT 2.2-9 UPPER ASSEMBLY 2.2-10 STEAM GENERATOR - FLOW RESTRICTOR NOZZLE 2.2-11 PRIMARY HEAD DRAINS 3.1-1 EXISTING BLOWDOWN AND RECIRCULATION SYSTEM 3.1-2 MODIFIED BLOWDOWN AND RECIRCULATION SYSTEM 3.3-1 EXISTING TURBINE ANALYZER PANEL FOR SAMPLING SYSTEM
- 3.3-2 3.4-1 3.5-1 4.1-1 4.1-2 4.1-3 MODIFIED TURBINE ANALYZER PANEL FOR SAMPLING SYSTEM PRIMARY HEAD DRAIN SYSTEM WIDE RANGE LEVEL TRANSMITTER SITE PLAN BARGE SLIP OLD STEAM GENERATOR STORAGE FACILITY PLAN VIEW 4.1-4 OLD STEAM GENERATOR FACILITY SECT.IONS AND DETAILS 4.1-5 CONTAINMENT LAYDOWN AREAS 4.1-6 GENERAL ARRANGEMENT PLAN VIEW, SHEET 1 4.1-7 GENERAL ARRANGEMENT PLAN VIEW, SHEET 2 4.1-8 GENERAL ARRANGEMENT SECTION A-A 4.1-9 GENERAL ARRANGEMENT SECTION B-B 4.1-10 DOWN-ENDING STEAM GENERATOR ONTO SLEDS 4.1-11 LOWERING STEAM GENERATOR FROM ELEVATOR PLATFORM ONTO TRANSPORTERS 4.1-12 STEAM GENERATOR IN HOISTED POSITION, SECTION VIEW 4.1-13 STEAM GENERATOR ON TRANSPORTER BETWEEN STORAGE AND CONTAINMENT 4.1-14 STEAM GENERATOR ON TRANSPORTER BARGE TO STORAGE 4.1-15 OFF LOADING STEAM GENERATOR FROM BARGE, PLAN VIEW 4.1-16 OFF LOADING STEAM GENERATOR FROM BARGE, viii Revision 3 July 1979
PALISADES PLANT SGRR LIST OF FIGURES ELEVATION VIEW 4.1-17 OFF LOADING STEAM GENERATOR FROM BARGE, SECTION VIEW 4.1-18 CONTAINMENT INTERNALS - RIGGING DESIGN LOADS
~ 4.1-19 CONSTRUCTION OPENING DETAILS I
4.1-20 CONSTRUCTION OPENING AND TENDON DETENSION/REMOVAL 4.1-21 FINITE ELEMENT MODEL FOR CONTAINMENT SHELL ANALYSIS 4.1-22 THREE DEMENSIONAL PLOT OF CONTAINMENT SHELL (WITHOUT OPENING) 4.1-23 MEMBRANE STRESSES VERSUS HEIGHT (WITHOUT OPENING) 4.1-24 MERIDIAN STRESSES VERSUS HEIGHT (WITHOUT OPENING) 4.1-25 HOOP STRESSES VERSUS HEIGHT (WITHOUT OPENING) 4.1-26 MEMBRANE STRESSES VERSUS AZIMUTH (WITHOUT OPENING) 4.1-27 MERIDIAN STRESSES VERSUS AZIMUTH (WITHOUT OPENING) 4.1-28 HOOP STRESSES VERSUS AZIMUTH (WITHOUT OPENING) 4.1-29 THREE DIMENSIONAL PLOT OF CONTAINMENT (WITH OPENING) 4.1-30 MEMBRANE STRESSES VERSUS HEIGHT (WITH OPENING) 4.1-31 MERIDIAN STRESSES VERSUS HEIGHT (WITH OPENING) 4.1-32 HOOP STRESSES VERSUS HEIGHT (WITH OPENING) 4.1-33 MEMBRANE STRESSES VERSUS AZIMUTH (WITH OPENING) 4.1-34 MERIDIAN STRESSES VERSUS AZIMUTH (WITH OPENING) 4.1-35 HOOP STRESSES VERSUS AZIMUTH (WITH OPENING) 4.1-36 MEMBRANE STRESSES VERSUS HEIGHT (OPENING CLOSED) 4.1-37 MEMBRANE STRESSES VERSUS AZIMUTH (OPENING CLOSED) 4.2-1 TEMPORARY ELECTRICAL POWER SUPPLIES - ALTERNATIVE-1 4.2-2 TEMPORARY ELECTRICAL POWER SUPPLIES - ALTERNATIVE-2 4.2-3 PLANT SINGLE LINE DIAGRAM lX
PALISADES PLANT SGRR LIST OF FIGURES 4.2-4 PRIMARY COOLANT PIPING CUT POINTS 4.2-5 MAIN STEAM PIPING CUT POINTS 4.2-6 FEEDWATER PIPING CUT POINTS 4.2-7 BLOWDOWN PIPING CUT POINTS 4.2-8 STEAM GENERATOR UPPER SUPPORT DETAILS 4.3-1 ACCESS CONTROL (EL 590'-0" AND EL 611'-0") 4.3-2 ACCESS CONTROL (EL 649'-0") 4.3-3 PRIMARY COOLANT PIPING CONTACT RADIATION SURVEY 4.3-4 AVERAGE RADIATION FIELDS 10 WEEKS AFTER SHUTDOWN 4.3-5 RADIATION SURVEY (EL 607'-6") 4.3-6 MAXIMUM DOSE RATE INSIDE STEAM GENERATORS 4.3-7 GENERAL RADIATION FIELD NEAR STEAM GENERATOR PIPING A-l(b)-1 STEAM GENERATOR REPLACEMENT SCHEDULE 2 A-l(b)-2 STEAM GENERATOR RETUBING ~ C-3-1 PRIMARY COOLANT PIPING CUT POINTS C-3-2 MAIN STEAM PIPING CUT POINTS C-3-3 FEEDWATER PIPING CUT POINTS 3 C-3-4 BLOWDOWN PIPING CUT POINTS
- x Revision 3 July 1979
(- PALISADES PLANT STEAM GENERATOR REPAIR REPORT LIST OF EFFECTIVE PAGES Page Latest Identification Amendment i 1 ii 1 iii 0 iv l v ~ 0 vi 0 vii 0 viii 1 ix 0 x 0 LOEP-1 1 LOEP-2 1 LOEP-3 0 ce LOEP-'J LOEP-5
- L<5:EP-=-6-- 1 0
<I*:..
1-1 0 1-2 0 1-3 0 f-4 ---- :t 1-4a 1 1-5 0 1-6 f 1-7 1 2-1 0 2-2 0 2-3 0 2-4 0 2-5 l 2-6 3 2-7 :t 2-8 0 2-9 0 2-10 0 2-11 1 2-12 0 2-13 0 Tbl 2.1-1 1 \*e Tbl 2.1-2 Tbl 2.2-1 Fig. 2.2-1 1 l 1 Fig. 2.2-2 i Fig. 2.2-3 1 Fig. 2.2-4 1 MARCH 1979 LOEP-1 REV. 1
<9"-. ___ , PALISADES PLAN'! SGRR Paqe Latest Identification Amendment Fiq. 2.2-5 0 Fiq. 2. 2-6 0 Fiq. 2.2-7 0 Fig. 2. 2-8 0 Fiq. 2.2-9 0 Fig. 2. 2-10 1' I Fig. 2.2-11 0 3-1 0 3-2 0 3-3 1 3:..::4 ------------ l Fig. 3.1-1 0 Fig. 3.1-2 0 Fig. 3.3-1 0 Fig. 3.3-2 0 Fig. 3.4-1 0 r*e \..:_-
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0 4-3 0 4-4 0
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\
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4-26 4-27 0 0 4-28 0 4-29 0 LOEP-2 MARCH 1979 REV. 1
- PALISADES PLANT SGRR Paqe Latest Identification Amendment 4-30 0 4-31 0 4-32 0 4-33 0 4-34 0 4-35 0 4-36 0 4-37 0 4-38 0 4-39 0 4-40 0 4-41 0 4-42 0 LJ-43 0 4-~4 0 4-45 0 4-46 0 4-47 0 4-48 0 4-49 0 4-50 0 4-51 0 4-52 0 4-53 0 4-54 0 4-55 0 4-56 0 Tbl 4.2-1 sh. 1 0 Tbl 4. 2-1 sh. 2 0 Tbl 4.2-2 0 Tbl 4.2-3 sh. 1 0 Tbl 4. 2-3 sh. 2 0 Tbl 4.2-3 sh. 3 0
'Tbl 4.2-3 sh. 4 0 Tl:l 4.2-3 sb. 5 0 Tbl 4. 2-3 sh. 6 0 Tbl 4.3-1 sh. 1 0 Tbl 4 .3-1 sh. 2 0 Tbl 4.3-1 sh. 3 0 Tbl 4.3-2 sh. 1 0 Tbl LJ.3-2 sh. 2 0 Tbl 4.4-1 0
- Tbl 4. 4-2 Fiq. 4.1-1 Fig. 4. 1-2 LOEP-3 0
0 0
PALISADES PLANT SGRR Page Latest Identification Amendment Fiq. 4.1-3 0 Fig. 4.1-4 0 Fig. 4 .1-5 0 Fiq. 4.1-6 0 Fiq. 4.1-7 0 Fig. 4.1-8 0 Fiq. 4. 1-9 0 Fig. 4.1-10 0 Fig. 4.1-11 0 Fiq. 4.1-12 0 Fig. 4.1-13 0 Fiq. 4. 1-14 0 Figa 4. 1-15 0 Fig. 4.1-16 0 Fig. 4.1-17 0 Fig. 4. 1-18 0 Fig. 4.1-19 0 Fig. 4 .1-20 0 Fig. 4.1-21 0 Fig. 4.1-22 0 Fig. 4.1-23 0 Fiq. 4. 1-24 0 Fig. 4.1-25 0 Fiq. 4.1-26 0 Fig. 4. 1-27 0 Fig. 4.1-28 0 Fig. 4 .1-29 0 Fig. 4. 1-30 0 Fiq. 4 .. 1-31 0 Fiq. 4. 1-32 0 Fig., 4. 1-33 0 Fig. 4.1-34 0 Fig. 4.1-35 0 Fiq. 4.1-36 0 Fig. 4.1-37 0 Fiq .. 4.2-1 0 Fiq. 4. 2-2 0 Fig. 4.2-3 0 Fig. 4.2-4 o. Fiqo 4. 2-5 0 Fig.,. 4 .. 2-6 0 0 Fig. 4. 2-7 Fig. 4. 2-8 0 Fig. 4.3-1 0 Fig. 11.3-2 0 LOEP-4
l*e I, ~ . PALISADES PLANT SGRR Paqe Latest Identification Amendment Fiq. 4 .3-3 0 Fig. 4. 3-4 0 Fig. ci.3-5 0 Fig. 4. 3-6 0 Fiq. 4. 3-7 0 5-1 0 5-2 0 6-1 1 6-1a 1 6-1b 1 6-1c 1 6-1d 1 6-1e l 6-1f l 6-1g l 6-1h 1 / 0 (*e 6-2 6-3 6-4 0 0 6-5 0 6-6 0 6-7 0 6-8 0 Table 6.2-1 0 Table 6.2-2 0 Table 6.2-3 0 Table 6.2-4 0 0
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Table 6.2-5 7-1 0 7-2 0 7-3 0 7-4 0 7-5 0 8-1 0 8-2 0 8=3 0 8-4 0 8-5 0 8-6 0 e
~ .* . ,. LOEP-5 MARCH 1979 REV. 1
PALISADES PLANT SGRR Page Latest Identification Amendment 8-7 0 8-8 . 0 Table 8.4-1 0 Table 8.8-1 0 9-1 0 9-2 0 9-3 0 10-1 0 LOEP-6 MARCH 1979 REV. 1
PALISADES PLANT STEAM GENERATOR REPAIR REPORT
1.0 INTRODUCTION
SUMMARY
, AND CONCLUSIONS During the operating history of the Palisades Plant, the steam generators have been afflicted by a number of corrosion-related phenomena. (The problems associated with wastage, intergranular attack, and 11 denting 11 at the Palisades Plant have been well documented elsewhere and will not be discussed in this report.) As a consequence of those problems, a substantial portion of the excess heat transfer capacity of the Palisades Plant steam generators has been removed. Although major progress has been made toward arresting or retarding the various corrosion mechanisms through changes in plant equipment, secondary chemistry, and operating procedures, denting remains an unresolved issue and threatens to result in additional steam generator problems. With the uncertainty that exists with regard to future plugging requirements and the expected useful lives of the existing steam generators, Consumers Power Company has evaluated those practical major repair alternatives that are available to restore the steam generators to their original condition.and regain the initial heat transfer capability. Complete replacement of the steam generators has been determined to be the preferred method of repair if major repair becomes necessary. Recognizing the length of the procurement lead times associated with repair components and the potential time required to obtain regulatory approval for a program of major repair, Consumers Power Company has embarked upon a contingency planning effort designed to mitigate possible adverse consequences of continued corrosion. The major elements of this contingency plan include: procurement of replacement steam generators and other long lead time materials, engineering studies and detailed engineering necessary to support the construction aspects of repair, preparation of work plans associated with repair activities, and preparation of applications for permits needed from various regulatory bodies to conduct the repairs. At this time, no date has been established for commencing the repair effort described in this report and, although Consumers Power Company is encouraged by the results of recent eddy current testing of the existing units, i t is considered only prudent to proceed with contingency planning while doubt exists about the continued degradation of the 1-1
PALISADES PLANT SGRR tubing. Sleeving, plugging, and other qualified methods will continue to be employed in the future to extend the lives of the original units as long as practical. The decision to ultimately replace the steam generators will be based upon system reserve margins, inspection requirements, the condition of the old units with respect to derating, and other relevant factors. The following report discusses the elements of a steam generator repair program, based upon complete replacement ©f the existing units, which is considered t9 offer the optimum solution for the Palisades Plant if major repairs become necessary. The primary emphasis of the report is on the safety-related aspects of the repair; however, other important aspects of the program are also addressed. The information presented in this report reflects the most current design information at the time of preparation. Since the design work for the program is currently underway, it may not be possible to present detailed information for all phases of the project. For those cases where engineering or design information is not av~ilable, the basis and criteria are presented. If there are alternatives under consideration, a summary of these is presented. 1.1
SUMMARY
OF STEAM GENERATOR REPAIR PROGRAM 1.1.1 REPAIR ALTERNATIVES Of the current options available to restore full steam-generating capability of the Palisades Plant, replacement of the existing steam generators with complete new units is the economic as well as the technical choice. The other major repair alternatives considered were the retubing of units in situ and repair limited to the replacement of theevaporator portions of the original units, including the primary heads (see Section 8.0). The key decision variables for selection between alternatives were the man-rem exposure associated with the repair activities and the plant outage requirement; complete replacement is the clear choice on the basis of both of these criteria .
- 1-2
1 ' PALISADES PLANT SGRR 1.1.2 REMOVAL AND REPLACEMENT OF THE STEAM GENERATORS FRO~ CONTAINMENT The feasibility of various schemes for the removal and replacement of the steam generators has been examined. Three schemes were evaluated in detail:
- a. Steam generator removal and replacement through the southeast containment wall
- b. Steam generator removal and replacement .through the northeast containment wall
- c. Steam generator removal and replacement through the containment dome The three schemes are equally feasible; however, from the standpoint of sc~edule and accident considerations, the first scheme is the most desirable. Therefore, the
- construction-related evaluations addressed in this report are only for the pathway through the southeast containment wall.
A 1/2-inch to 1-foot scale model of the containment, including the construction opening~ has been constructed to aid the repair program (see Section 4.9). The model has been used extensively to detail the removal and replacement sequences of the steam generators from the containment. 1.1.3 STEAM GENERATOR CHARACTERISTICS The replacement steam generators are designed to physically match the essential parameters of the existing steam
- generators and be compatible with the performance characteristics utilized in the Palisades Plant Final Safety Analysis Report (FSAR) and the license for operation at 2530 MWt. Although the plant safety analysis is now done for a power level of 2530 MWt, the replacement steam generators are designed for operation at 2650 MWt.
Improved design features are being incorporated in .the replacement steam generator design to increase long-term integrity and reliability. These improved features will have no significant adverse impact on the plant safety analysis. The shop-fabricated replacement steam generators will be designed and manufactured to updated manufacturing 1-3
PALISADES PLANT SGRR techniques and American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) editions. 1.1.4 SAFETY-RELATED CONSIDERATIONS The potential impacts of the replacement steam generators on the design basis events, analyzed in the FSAR, have been evaluated and are described in Section 6.1. Those evaluations conclude that ~he replacement steam generators will have no significant adverse impact on the design basis events. Construction-related incidents pertaining to the transportation and handling of the steam generators have been evaluated and are described in Sections 4.1.4 and 6.2. Those evaluations conclude that the construction activities will have no significant adverse impact on the safety of the Palisades Plant. The tire prevention and protection program to be in operation during the steam generator repair program is described in Section 6.3. It is concluded that these measures provide reasonable assurance that the Palisades Plant is protected from significant damage due to fire during the repair activities.
- 1.1.5 ALARA CONSIDERATIONS Personnel exposure will be maintained "as low as is reasonably achievable" (ALARA) throughout the steam generator repair program.
Estimates of the exposures to personnel involved in the various repair alternatives have been developed using projections of work activity durations, manpower levels, and expected radiation levels. 1.1.6 OFFSITE RADIOLOGICAL CONSEQUENCES Radiological evaluations of the gaseous and liquid releases attributable to the steam generator repair have been conducted. The effects of the releases are less than those associated with normal operation of the facility on the basis of the discussion in Section 6.2.2. 1-4 MARCH 1979 REV. 1
PALISADES PLANT SGRR 1.1.7 UNIQUE ASPECTS OF PROGRAM As presently contemplated, there are no unique engineering or construction aspects of the Palisades Plant steam generator repair program. The repair program, including the fabricatio~ of replacement units, will utilize conventional nuclear industry manufacturing and construction methods. The shop fabrication of the steam generators will be conducted in accordance with standard shop practices. The closure of the temporary construction opening in the containment will be performed in a manner similar to that used to close the original containment construction opening. The transport and rigging of the steam generator will utilize proven techniques. In short, the repiir program will.rely on fabrication and construction practices or techniques which h~ve been previously qualified for similar applications. MARCH 1979 REV. 1
PALISADES PLANT SGRR 1.1.8 STEAM GENERATOR DISPOSAL The repair activity and ultimate disposal of the existing steam generators are separate issues. This report discusses the various means by which the steam generators can be disposed of to demonstrate the feasibility of disposal. Economic and ALARA considerations will determine the method to be utilized. Because of the uncertainty of the timing of the repair and availability of offsite disposal facilities, the ultimate disposition of the old units cannot be finalized at this time. 1.2 IDENTIFICATION OF PRINCIPAL AGENTS Consumers Power Company, hereinafter called Consumers, is a public utility incorporated to do business under the laws of the State of Michigan. Consumers is the sole owner and operator of the Palisades Nuclear Plant. Consumers will have overall responsibility for the steam generator repair program . Bechtel Power Corporation, a Nevada corporation (hereinafter called "Bechtel" except in respect to the performance of services of a professional engineer), and Bechtel Associates Professional Corporation, a professional corporation authorized to practice professional engineering in the State of Michigan (hereinafter called "Bechtel" in respect only to the performance of services of a professional engineer), have been retained by Consumers to provide construction planning, engineering, procurement, and cost and scheduling services for the steam generator repair program. Combustion Engineering, Inc., a Delaware corporation authorized to do business in the State of Michigan (hereinafter called Combustion), will manufacture the replacement steam generators. Combustion also provided the existing steam generators. Because of the uncertainty of the timing of the repair, the principal agent for the installation portion of the repair program cannot be identified at this time. 1-5
e PALISADES PLANT SGRR 1.3 10 CFR 50.59 CONSIDERATIONS Repair or replacement of equipment at a power plant, performed in accordance with appropriate procedures, is a maintenance activity that is routinely conducted. Because of the scope of the steam generator repair, it was considered prudent to evaluate this activity to determine:
- a. If the proposed repair activity would involve a change in the technical specifications incorporated in the licence.
- b. If the proposed repair activity would involve an unreviewed safety question per the requirements of 10 CFR 5 0. 5 9 (a) ( 2) .
Each design basis event FSAR accident analysis has been evaluated, and it has been concluded that the replacement steam generators would not alter the conclusions reached in the FSAR. The evaluation indicates that the steam generator repair does not involve an unreviewed safety question. A change in the plant technical specifications is required to incorporate the main steam line isolation on high containment pressure discussed in Section 3.6 of this report. The construction incident potential has been evaluated to determine the presence of any new or unique accidents and the potential impact on cooling spent fuel. The evaluation indicates that the steam generator repair activity does not involve an unreviewed safety question. Additionally, before replacement of the Steam generators, the effective Palisades Plant Technical Specifications will be reviewed again and revised as necessary.
1.4 CONCLUSION
S The fundamental conclusions reached are that the steam generator repair can be conducted utilizing proven manufacturing and construction techniques and that the repair program does not result in any significant adverse impact on the plant safety analysis and the ability to maintain a safe configuration and cool stored fuel. Additionally, current FSAR safety analyses are applicable to the replaced steam generators. The detailed bases 1-6 MARCH 1979 REV. 1
PALISADES PLANT SGRR supporting these concluqions are provioed in the report that follows. 1-7 MARCH 1979 REV. 1
PALISADES PLANT SGRR 2.0 REPLACEMENT COMPONENT DESIGN Combustion is designing and shop-fabricating two replacement steam generator units shown in Figure 2.2-1. The design and performance of the replacement steam generators will match that of the steam generators being replaced such that:
- a. The replacement steam generators' physical parameters are essentially equal to those of the original units.
- b. The replacement steam generators' operating characteristics and parameters are compatible with the plant safety analysis at 2530 MWt.
Although the plant safety analysis is for a power level of 2530 MWt, the replacement steam generators are designed for operation at 2650 MWt (see Table 2.1-2). The replacement units incorporate many improved design features derived from both plant operating experience and development programs directed toward design of steam generators for long-term integrity and reliability. This section discusses the design and manufacture of these replacement units. 2.1 COMPARISON WITH EXISTING COMPONENT DESIGN 2.1.1 PARAMETRIC COMPARISON The replacement steam generators for the Palisades Plant will have physical and mechanical characteristics similar to the original design documented in the FSAR (see Table 2.1-1). These characteristics provide the replacement steam generators with a thermal performance consistent with the original steam generators. The original steam generators were fabricated to the 1965 Edition of the ASME Code, including all addenda through Winter 1966. The replacement steam generators are being designed and fabricated to the 1977 edition of the ASME Code. The stress report for the* replacement steam generators will be based on the 1977 edition of the ASME Code. The fabrication and analysis requirements for the replacement steam generators will be at least equivalent to those utilized for the original steam generators .
- 2-1
PALISADES PLANT SGRR The replacement steam generators will include a number of design improvements which are discussed in Section 2.2. Many of the advanced design features incorporated on the replacement steam generators are similar to features included on Combustion's System 80 steam generators. Data for the replacement steam generators at 2450 MWt is presented in Table 2.1-1, allowing comparison of these parameters between the original steam:generators and the replacement steam generators. The design data for the replacement steam generators at 2530 MWt and 2650 MWt is shown on Table 2.1-2. Materials used in the fabrication of the replacement steam generators are procured to the requirements of the 1977 edition of the ASME Code. These materials are identical to those used in the original steam generators except where specific design improvements have been incorporated or the fabrication practice has been improved. Table 2.2-1 enumerates applications of materials for the original steam generators and the replacement steam generators. 2.1.2 PHYSICAL COMPATIBILITY WITH EXISTING STEAM GENERATORS AND SYSTEMS The replacement steam generators (see Figure 2.2-1) are designed and fabricated, to the extent possible within the constraints of physical dimensions and design requirements, to preserve the existing plant mechanical interfaces, including the support structures and loadings. Interfaces between the steam generators and other plant components and systems are maintained. 2.1.3 ASME CODE APPLICATION The present operating steam generators were designed and fabricated to the reqliirements of the 1965 edition of ASME Code, Section III, including all addenda through Winter 1966. The replacement steam generators will be fabricated to the requirements of the 1977 Edition of the ASME Code. Design of the replacement steam generators is consistent with the design of the primary coolant system. Materials used in fabrication are being procured to the requirements of current codes. All material certification tests will be performed and recorded as required by the ASME Code .
- 2-2
PALISADES PLANT SGRR 2.1.4 REGULATORY GUIDE APPLICATION The compilation below addresses regulatory guides considered applicable to the fabrication of the replacement steam generators. It must be noted that these guides were issued after construction of this facility. The intent is to accommodate the guidance of these regulatory guides insofar as they provide an acceptable method to allow the licensee to comply with the requirements of 10 CFR 50.
- a. Regulatory Guide 1.29, Seismic Design Classification (February 1976)
Combustion's design of the replacement steam generators is consistent with design to withstand the effects of the safe shutdown earthquake (SSE) and the classification guidance of this regulatory guide.
- b. Regulatory Guide 1.31, Control of Stainless Steel Welding (May 1977)
Combustion's shop fabrication welding quality assurance is described in CENPD-210, Quality Assurance Program - A Description of the Combustion Engineering Nuclear Steam Supply System Quality Assurance Program.
- c. Regulatory Guide 1.34, Control of Electroslag Weld Properties (December 1972)
Where electroslag welding is utilized, Combustion requires its suppliers to follow the recommendations of this guide.
- d. Regulatory Guide 1.43, Control of Stainless Steel Weld Cladding of Low-Alloy Steel Components (May 1973)
Combustion's shop fabrication weld cladding of the replacement steam generators utilizes materials which are not susceptible to underclad cracking. These materials do not require the controls listed in this guide. 2-3
PALISADES PLANT SGRR
- e. Regulatory Guide 1.44, Control of the Use of Sensitized Stainless Steel (May 1973)
All of the unstabil1zed austenitic steels used for component parts of the primary coolant pressure boundary are utilized in the final heat treated condition required by the ASME Code, Section II material specification for that particular type or grade of alloy. Processing and fabrication are performed utilizing established techniques to avoid sensitization. *
- f. Regulatory Guide 1.48, Design Limits and Loading Combinations for Seismic Category I Fluid system Components (May 1973)
Combustion's design of the replacement steam. generators meets the requirements of general design Criterion 2 and the ASME Code. The loading combinations and design limits utilized in the stress report for the steam generators are consistent with those considered in the FSAR .
- g. *Regulatory Guide 1.50, Control of Preheat Temperature for Welding of Low-Alloy Steel (May 1973)
Combustion shop fabrication practices are in. agreement with Regulatory Positions C.l.a, C.3, and C.4. The soundness of all welds is verified by an acceptable examination procedure which per Regulatory Position C.4 meets all requirements of this regulatory guide.
- h. Regulatory Guide 1.71, Welder Qualification for Areas of Limited Accessibility (December 1973)
Combustion's shop fabrication welding complies with the fabrication requirements specified in ASME Code, Sections III and IX.
- i. Regulatory Guide 1.83, Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes (Rev 1, July 1975)
Combustion has designed the replacement steam generators to allow access to the tubes for inspection and plugging. 2-4
PALISADES PLANT SGRR
- j. Regulatory Guide 1.84, Code Case Acceptability -
ASME III Design and Fabrication (August 1977)
- k. Regulatory Guide 1.85, Code Case Acceptability -
ASME III Materials (August 1977) 2.2 COMPONENT DESIGN IMPROVEMENTS The replacement steam generator design incorporates the traditional Combustion design features and design improvements that have evolved through several generations of steam generator designs in response to the operational steam generator problems that have* occurred in the nuclear industry. The replacement steam generators will essentially duplicate the physical, thermal, and hydraulic characteristics of the original units while incorporating a combination of features proven in field operation and design improvements to mitigate operational problems. The heating surf ace has been selected to provide thermal performance which would match that presently installed and to respond to_ plant thermal transients in the same manner as does the existing unit. The design will provide improvements in thermal/hydraulics, notably in secondary flow distribution. These improvements are intended to minimize flow stagnation, steam blanketing, and harmful solids accumulations. It is important to avoid harmful solids deposits in contact with heat transfer tubing in the steam generators. The blowdown arrangements are designed to take advantage of the improved flow distribution, making significant improvements in the effectiveness of blowdown in removing harmful solids deposits. The tube sup.port system utilizes the traditional Combustion eggcrate tube support, with its low flow~resistance, support against vibration and wear, and resistance to tube denting or lateral tube deformation. The bend region tube support system also uses the standard Combustion approach, with double 90 degree bends and support assemblies of interlocking strips. The design provides positive restraint- against vibration and resistance to tube deformations during LOCA, steam line break, and seismic events. The tube support system provides rugged, positive support while minimizing flow resistances and the possibility of local dryout of steam blanketed regions. 2-5 MARCH 1979 REV. 1
PALISADES PLANT SGRR Access openings and inspection ports are provided to enable inspection of tubes, tubesheet, and support surfaces within the tube burldle as well as at the periphery of the bundle. 2.2.1 DESIGN FEATURES TO IMPROVE PERFORMANCE 2.2.1.1 Thermal Performance In order to minimize the effect on plant transient .!_ *- ~. performance, beat transfer tubes of 3/4 inch outside ~- . **.
.-~ diameter and .042 inch average wall thickness (consistent with Combustion's System 80 design) will be provided in such quantities and lengths on the replacement units that the product of their area (A) and heat transfer coefficient (U) equals the product of the original area and original coefficient, i.e., (UA) new = (UA) original. The total cross-sectional flow area of the heat transfer tubes will be equal to the cross-sectional flow area of the tubes in the original units. The control of these parameters on the replacement units allows the hydraulic impedance to primary flow to essentially correspond with that of the original steam generators.
2.2.1.2 Deleted
-- 2-6 MARCH 1979 REV. 1
PALISADES PLANT SGRR 2.2.l.3 Blowdown Capability The potential blowdown capabilities for the replacement steam generators are designed to take advantage of the improvements incorporated, which contribute to the improved secondary flow distribution and hydraulics within the operating steam generator. The recirculating fluid exits from the downcomer and flows radially across the tube bundle. The lowest fluid ~velocities occur near the center open region of the' tube bundle where dropout of solid particles may occur. In this region, free of heat transfer tubes, blowdown duct that takes suction in a circular pattern adjacent to the innermost tubes is provided. Figure 2 .2-4 shows the schematic arrangement of the blowdown duct and its relationship to the tube bundle. At the end of each circumferential section of the blowdown duct, a transport duct (with no blowdown openings) carries the blowdown fluid across the divider lane discharging through intersecting holes drilled in the tubesheet to a 6-inch Schedule 80 blowdown nozzle. See Section 3.1 for a description of the connecting blowdown system. The internal blowdown duct and tubesheet blowdown connection for the replacement units have been sized to accommodate future higher blowdown capabilities than those available on the original steam generators. L.-7 MARCH 1979 REV. 1
PALISADES PLANT SGRR The b~owdown capability provided on the replacement steam generators is able to provide effective continuous blowdown and potential periodic blowdown at high flowrates. 2.2.1.4 Tube Supports There are essentially three types of structures within the Palisades Plant replacement steam generators that provide support to the tubes. In addition, the tubesheet, into which the tubes are expanded through the full thickness, provides fixed support to the extremities of all tubes. The three types are (see Figure 2.2-5):
- a. H - Horizontal Grid or 11 Eggcrate 11 This grid may be a full circular structure or a partial circle bounded by the circumference and a chord. It is composed of slotted strips intersecting at an angle of 60 degrees and joined together at the outer and inner perimeters with a pair of square bars, top and bottom. Normally, the strips alternate between a 2-inch slotted one and a 1-inch unslotted one, both .090-inch thick. The structure is fabricated in a fixture and thereafte~
handled as a plate (see Figure 2.2-6).
- b. V - Vertical Grid Unlike the eggcrate support, this structure is assembled during the installation operation concurrently with the tubing operation. It is composed of vertical, slotted 2-inch strips intersecting with horizontal .5-inch strips, both
.090-inch thick. The assembly is bounded about the periphery by either square bars or custom-shaped plates depending on location (see Figures 2.2-7 and 2.2-8).
- c. B - Batwing Strips These strips provide out-of-plane support to the bend corners which would otherwise be quite
- flexible. The strips are 2-inch deep and .090-inch thick~ They rest in a slotted T-bar at their lower center and are joined together at their extremities by a wrapper strip which takes a sinuous shape when completely installed (see Figure 2.2-7).
2-8
PALISADES PLANT SGRR The tube support system (described above) selected for the Palisades Plant replacement units is particularly advantageous in that it provides sufficient strength while minimizing resistance to flow. The large open flow area available in this support system avoids the accumulation of boiler water deposits by eliminating local flow eddies and flat surf aces present in other commonly used tube bundle support systems. Avoiding the accumulation of corrosion products avoids the concentration of acid-producing chloride cells, which is believed to cause accelerated carbon steel support corrosion and subsequent tube denting. Further, if magnetite growth due to the chloride salt concentration does occur within the eggcrate, the geometry should not produce the denting phenomena. The thin strips which make up the support matrix do . not have sufficient rigidity to collapse the tube wall. *
- Both the horizontal and vertical grids are made from 409 ferritic stainless steel. This material was selected because of its high resistance to general corrosion and thinning. The ferritic stainless is preferable to austenitic stainless because the coefficient of thermal expansion is more compatible with carbon steel and Inconel material.
2.2.1.5 Feedwater Ring and Feedwater Nozzle Liner The f eedwater distribution ring for the replacement steam generators will consist of two half-ring pipes connected by a "goose neck" elbow arrangement to a header at the feedwater nozzle. A gap of approximately 20 inches will be provided at a location 180 degrees from the f eedwater nozzle to allow access to the area below the feed ring. Support for the ring is provided by attachments welded to the steam generator pressure shell. Distribution of feedwater to the downcomer will be accomplished by discharge nozzles spaced around the circumference of the ring. These nozzles will be 90 degree elbows (J-tubes) mounted on top of the ring. This design, as opposed to underside discharge ports, prevents immediate feed ring draining in the event of loss of feed flow, thereby minimizing the possibility of water hammer. Thermal fatigue protection for the feedwater nozzle will be provided by a nozzle thermal liner. It is the purpose of this liner to insulate the nozzle material from cold water transients. The li.ner also acts as the piping link between the feed nozzle and the feedwater ring. The design of the 2-9
PALISADES PLANT SGRR thermal liner also serves to minimize the leakage at the liner/header joint, thereby minimizing the potential for water hammer following feedwater flow interruptions. 2.2.1.6 Tube Lane Divider Plate On the replacement units, a divider plate is mounted in the tube lane between the hot and cold leg sides of the tube bundle beneath the flow distribution baffle. The divider plate performs a dual function of preventing preferential bypass of the tube bundle by recirculating water exiting from the downcomer along the tube lane and providing a positive support arrangement for the blowdown duct. The divider plate is attached to the lower shell by tongue and groove joints so that there is no structural interaction with the secondary shell under pressure and thermal deflections. 2.2.1.7 Integral Flow Restrictor Steam Outlet Nozzle The steam outlet nozzles on the replacement steam generators will incorporate an integral flow restriction equivalent to 70% of total flow area (see Figure 2.2-10). These nozzles are similar to those installed on Combustion's System 80 steam generator and will serve to limit the generator blowdown rate during a main steam line break accident. 2.2.1.8 Water Sampling Provisions Each replacement steam generator will be provided with 2 secondary water sampling connections (see Figure 2.2-10). They consist of a 3/8-inch diameter internal water sampling line which exits through a 3/4-inch instrument nozzle in the steam drum shell. Internally the sample line provides water samples from a sampling cup on the separator deck on the basis that recirculating water from the steam separators is most nearly representative of boiler water in contact with
- the heat transfer tubes. See Section 3. 3 for a description of the connecting sampling system.
Water sampling may also be taken from the bottom blowdown line external to the steam generator. 2-10
PALISADES PLANT SGRR
- 2. 2. 2 DESIGN FEATU.RES TO IMPROVE MAINTENANCE AND INSPECTION 2.2.2.1 Handholes The replacement steam generators will include four 6-inch handhole openings on the lower and intermediate shells to facilitate inspections. The lower two handholes will be positioned just above the tubesheet and have provisions for viewing through the tube bundle shroud.
The upper handholes are located just above the eggcrate in the tube lane and are adjacent to the bend region of the tube bundle. These handholes will also incorporate the provision for viewing through the tube bundle shroud (see Figures 2.2-1 and 2.2-10). 2.2.2.2 Inspection Ports Two 2-inch inspection ports will be added to the replacement units just above the tubesheet secondary face to provide accessibility to the tubesheet surface and to allow use of an inspection device such as a boroscope to observe tubes on either side of an opening between two particular tube rows. 2.2.2.3 Deleted 2.2.2.4 Primary Head Drains To facilitate draining of the steam generator primary head before maintenance or inspection activities in this area, the replacement units include a drain nozzle (see Figure 2.2-11) on both the inlet and outlet plenums of the primary head. See Section 3.4 for a description of the connecting drain system. 2-11 MARCH 1979 REV. 1
PALISADES PLANT SGRR 2.2.2.5 Recirculation and Chemical Cleaning Nozzle The replacement units will include a provision to allow recirculation of steam generator secondary water during standby or shutdown periods or for circulation of chemical cleaning fluids if chemical cleaning of the generator is undertaken. This capability is provided by a nozzle of 6-inch nominal pipe size in the steam drum shell to which is attached a piping system penetrating the steam separator deck and terminating in a sparger ring discharging in the riser space above the tube bundle. See Section 3.2 for a description of the connecting recirculation system. 2.2.2.6 Manways The replacement steam generators will have larger primary and secondary manways to improve access for personnel and equipment during inspection and maintenance activities. The manways will be 18-inch inside diameter and provide access to both the inlet and outlet plenums of the primary head, as well as the secondary steam drum area. 2.3 SHOP TESTS AND INSPECTIONS Combustion will perform all tests and inspection required by ASME Gode Section III during the fabrication of the replacement steam generators. Both primary and secondary side hydrostatic tests will be performed in Combustion's manufacturing facility in accordance with ASME Code Section III. The replacement steam generators will have an N-Code stamp upon delivery at the plant site. In conjunction with the ASME Code hydrostatic tests, a cyclic leak test of the secondary side will be conducted to ensure the integrity of the tube-to-tubesheet seal welds. Consumers will arrange source inspection and perform audit functions related to fabrication and shop testing. 2.4 STORAGE CRITERIA FOR NEW STEAM GENERATORS The following criteria are provided. for use in the event . that the new steam generators are not installed upon arrival at the site. Application of these criteria should maintain the integrity of the steam generators during storage.
- a. A strippable protective coating will be applied to the replacement steam generators before shipment.
The component surf ace will be protected by 2-12
PALISADES PLANT SGRR maintaining the integrity of this coating during storage.
- b. suitable supports should be provided for the replacement steam generators during horizontal storage.
- c. An internal positive inert gas atmosphere should be maintained in the primary and secondary sides of the replacement steam generators during storage and verified through periodic testing.
- d. The ambient temperature of the steam generators in storage should be maintained a minimum of 20F above the measured dew point of the internal gas.
2-13 l.
~ALISADES PLANT SGRR TABLE 2.1-1 STEAM GENERATOR COMPARISON DATA (1L Original Replacement Steam Steam A. Primary Side Generators Generators l . Thermal power, MWt 2450 2450 2* Design pressure, psi 2500 2500
- 3. Design temperature, OF 650 650
- 4. Cold leg temperature,. o.F 547.8 547.8 5* Hot. leg temperature, op 598.5 598.5
- 6. Coolant flow, 10~ lb/hr 6 2. 25 62.25 7* Calculated pressure drop, psid ' 30 .5 29.5 8* Normal ~perating pressure, psi 2100 2100 B. Secondary Side
- l. Design pressure, psi 1000 1000
- 2. Design temperature, op 550 550 3* Flow rate, 10 6 lb/hr 5.281 5.281
- 4. Steam outlet pressure, psi 770 770
- 5. Feedwater temperature, op 429.1 429.l e c. Dimensions
- l. Evaporator outside diameter, in 164 164
- 2. Steaffi' drum outside diameter, in 239-3/4 239-3/4 3* Overall length, in 709.78 742.00
- 4. Tubing outside diameter, in 0.750 0.750 5* Tubing wall thickness, in .048 .042 D. Hydrostatic Pressure i .* Primary, psi a 3125 3125 2 .. Secondary, psia 1250 1250 E. Weights and Volumes
- l. Complete vessel dry, lb 924,596 934,637 2
- Vessel c. G. dry, in 345.32 344.02 3
- Secondary fluid 0% power, lb 209,180 207,771 4
- Secondary fluid 100%
power, lb 129,164 147,288 Note: 1 (1 Values are per steam genera tor, except I tern A .1. MARCH 1979 REV. 1
PALISADES PLANT SGRR
., TABLE 2 .1-2 REPLACEMENT STEAM GENERATOR DATA (il Safety Analysis Design A. Primary Side
- l. Thermal power, MWt 2530 2650
- 2. Design pressure, psi 2500 2500
- 3. Design tempera tu re, '1.F 650 650
- 4. Cold leg temperature,°F 542.5 547.8 5* Hot leg tempera tu re, 0 .F 595.4 598.5
- 6. Coolant flow, 10 6 lb/hr 62.5 70.0 7* Calculated pressure drop, 29.6 36.l psid
- 9. Normal operating 2100 2250 pressure, psi B. Secondary Side
- 1. Design pressure, psi 1000 1000
- 2. Design tempera tu re, 0 .F 550 550
- 3. Flowrate, 10' 6 lb/hr 5.491 5.786
- 4. Steam outlet pressure, 770 770 psi 5* Feedwater temperature,°F 435 438 C. Weights and Volumes
- l. Complete vessel dry, lb 934,637 934,637
- 2. Vessel *c.G. dry, in 344.02 344.02 3* Secondary fluid 0% power 207,771 201,771 lb 4* Secondary fluid 100% 145,969 143,934 power, lb NOTE:
.**1 11 Values are per steam generator, except I tern A .1. MARCH 1979 REV. 1
i-i PALISADES PLANT SGRR TABLE 2. 2-1 STEAM GENERATOR MATERIALS Original Replacement Steam Generators Steam Generators Upper, inter- SA-302, Grade B SA-533, Grade A, mediate, and alloy steel Class I alloy cone shells steel Lower shell SA-516, Grade 70 SA-533; Grade A, carbon steel Class ,I alloy steel Tubesheet forging SA-508, Class II SA-508, Class III alloy steel alloy stee,;L Tube support plates SA-36 carbon steel Eggcrate tube supports A-570, Grade D/ A-176, Type 409 A-303-64, Grade D stainless steel carbon steel , Primary head SA-302, Grade B SA-533, Gi::ade B, alloy steel Class I.alloy steel Primary head clad Stainless steel Stainless steel Tubesheet clad Inconel Inconel Heat transfer tubing SB-163 SB-163 Inconel Inconel Secondary head SA-516, Grade 70/ SA-516, Grade 70 SA-302, Grade B carbon steel carbon steel/ alloy steel Nozzles/primary stay SA-508, Class II SA-508, Class III alloy steel alloy steel MARCH 1979 REV. 1
PALISADES PLANT STEAM GENERATOR REPAIR REPORT REPLACEMENT STEAM GENERATORS Figure 2.2-1 March 1979 Rev. 1
FIGURE 2.2-2 DELETED MARCH 1979 REV. 1
FIGURE 2.2-3 DELETED MARCH 1979 REV. 1
_BLOWDOWN
.-::-DUCT
-~SLOWDOWN NOZZLE b [k,_
- __ DIVIDER LANE
_BAFFLE -
*-~i~\
~
I /)
.-SECT\ON t\A..-A.11
~_:VESSEL -
- - SH ELL-----"
. ;-/
"~
-A--.
~'!v-
.SECONDARY BLOWDO\VN ASSY PALISADES PLANT STEAM GENERATOR REPAIR REPORT STEAM GENERATOR
- SECONDARY BLOWDOWN ASSEMBLY Figure 2.2-4 March 1979 Rev. 1
1' I
~
l\I\ B
. I.
~
II \
,,,~
f . ' 'i I
~
II I SUPPORT TYPES [fil HORIZONTAL CRID (£CC:CiATE) I (l] VERTICAL CRID c : (!] 8ATWIHG STRIPS I 111 , l I I I\ I II I 11 111 11 I I PALISADES PLANT I STEAM GENERATOR REPAIR REPORT
- ' '!UBE SUPPORr TYPES Figure 2.2-5 .
.A
- I PALISADES PLANT STEAM GENERATOR REPAIR REPORT Figure 2.2-6
PALISADES PLANT STEAM GENERATOR REPAIR REPORT BEND ROOION '!UBE SUPPORI'S Figure 2.2-7
PALISADES PLANT STEAM GENERATOR REPAIR REPORT
- TUBE SUPPORT Figure 2.2-8
PALISADES PLANT STEAM GENERATOR REPAIR REPORT UPPER ASSEMBLY Figure 2.2-9
PALISADES PLANT STEAM GENERATOR REPAIR REPORT STEAM GENERATOR FLOW RESTRICTOR NOZZLE Figure 2.2-10 March 1979 Rev. 1
\,
. I II I/
3 3
~ DRAIN ~ DRAIN
-- INCONEL WELD S.S. CLAD DING PRIMARY HEAD
// PALISADES PLANT 3 STEAM GENERATOR REPAIR REPORT.
~ SCH. 160 PRIMARY HEAD DRAIN (TYP.) PRIMARY HFAD DRAINS Figure 2.2-11 JI t.
\ '<#
/ \
.'~ PALISADES PLANT SGRR 3.0 BALANCE-OF-PLANT SYSTEM MODIFICATIONS 3.1 BLOWDOWN SYSTEM The existfng steam generator blowdown system is designed for a maximum ca'pabili ty of 100, 000 lb/hr. Each existing steam generator has\\two blowdown*connections, one for bottom blowdown and one for surface blowdown. Controls are provided to discharge both the surf ace and bottom blowdown to the flash tank for further processing in the blowdown system. Figure 3.1-1 shows a schematic representation of the existing blowdown system. The new steam generator design utilizes only bottom blowdown; a 6-inch diameter nozzle is provided for this purpose (See Section 2.2.1.3). The current intent is that the existing piping for the bottom blowdown will be connected to the new steam generator nozzle, and the present system blowdown capability will be maintained. The modified blowdown system is shown in Figure 3.1-2. The blowdown system is a nonsafety system, except the portion from the steam generator to the second isolation valve (CV-0771 for steam generator E-SOA and CV-0770 for steam generator E-SOB), which consists of seismic Catagory 1 and ASME Code Section III, Class 2 piping and valves. 3.2 RECIRCULATION SYSTEM The existing steam generator recirculation system is used as needed to maintain appropriate water chemistry while the steam generators are in wet layup or similar conditions. The system takes its suction from the bottom blowdown and discharges into the steam generator through the surface blowdown connections. The piping for the recirculation system is integrated with the blowdown system outside the containment. The steam generator blowdown pumps are used for maintaining the recirculation flow through the steam generator internals, and have a circulating capability of about 100 gpm. Figure 3.1-1 shows the existing recirculation system. In the new steam generator, a 6-inch nominal pipe nozzle is provided in the upper steam drum shell for steam generator recirculation and other purposes (See Section 2.2.2.5). The existing recirculation piping will be connected to the new steam generator recirculation nozzle for discharging water into the steam generator. The present connection for the recirculation suction line on the bottom blowdown line will 3-1
PALISADES PLANT SGRR be maintained. Figure 3.1-2 shows the piping arrangement for the modified recirculation system for the new steam generator. An additional pump will be provided in parallel with the existing pumps to increase the recirculation flow capability to approximately 150 gpm, which would provide a 6.4 hour steam generator turnover time. The recirculation system is a nonsafety system.
- 3. 3 SAMPLING SYSTEM
- Figure 3.3-1 indicates the parameters being sampled for the existing steam generator at the turbine analyzer panel.
These samples are drawn from the secondary side of the steam generator for monitoring the water chemistry. Additions to the existing secondary sampling system will be made to increase the capability of sampling the steam generators. The additional sample points will utilize the new steam generator sampling nozzles (See Section 2.2.1.8) and will continuously monitor pH, conductivity (specific and cation); and sodium. The modifications to the sampling systems are shown in Figure 3.3-2 .. The piping from the steam generator sampling nozzles will pass through the containment and will be provided with automatically controlled containment isolation valves. The sampling system is a nonsafety system, except the portion from the steam generator to the containment isolation valve, which consists of Seismic Catagory 1 and ASME Code Section III, Class 2 piping and valves. 3.4 PRIMARY HEAD DRAINS The existing steam generators do not have primary head drains to enable complete draining before entry for maintenance activities. The new steam generators have two 3/4-inch drain nozzles located in each primary head as described in Section 2.2.2.4. Each drain will be double valved as close as is practical to the steam generator. Th~ drains will be connected to the drain collection header, which discharges into the primary drain tank as shown on Figure 3.4-1. The portion of the primary head drain system from the steam generators to the second isolation valves consists of seismic Category 1 and ASME Code Section III, Class 1 piping and valves. The remainder of the system will be either Class 3 or nonsafety, with a transi,tion to Class 3 at the drain header .
- 3-2
PALISADES PLANT SGRR 3.5 WIDE RANGE LEVEL INDICATION A differential pressure type level transmitter will be added to each new steam generator to provide steam generator wide range level indication of about 44 feet. The top head of the new steam generators will have a 1-inch nozzle at el 671 1 , which will be used for the low-pressure sensing iine connection. The high-pressure sensing line will be connected to pressure taps at el 627' (see Figure 3.5-1). With this addition, an operator can determine the water level in the secondary side of each steam generator during wet layup (or similar operations) beyond the range measurable with the present level indicato~s (about 15 feet). The new level indicating system is functionally independent, both electrically and mechanically, of any safety-related systems. The sensing lines will be in accordance with ASME Code Section III, Class 2 and seismic Category I classifications. The transmitter will be located in a low radiation zone. The transmitter output will electrically connect to a new indicator in the main control room and will not be used to automatically initiate or terminate any action. 3.6 MAIN STEAM ISOLATION VALVE CLOSURE SIGNAL The existing circuitry provides for closure of the Main Steam Isolation Valves based on low steam generator pressure. This closure signal is provided to protect against excessively high releases of steam to the containment as a result of a steam line break. The replacement steam generators have steam nozzle flow restrictors to restrict the blowdown rate following a steam line break. The effect of the flow restrictors is to reduce the rate of steam generator pressure change during blowdown. Hence, following a design basis steam line break with the replacement steam generators, the high containment pressure trip setpoint will be reached before the low steam generator pressure trip is reached. The Main Steam Isolation Valve Closure signal will be modified to be actuated from high containment pressure as well as low steam generator pressure. This will reduce the mass/energy release following a steam line break and result in lower containment peak pressures. 3-3 MARCH 1979 REV. 1
PALISADES PLANT SGRR The containment pressure instrumentation and circuitry are safety grade. .1 3-4 MARCH 1979 REV. l
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STEAM STEAM GEN. GEN. E-SOA E-508 DRAINS y y FLOOR D~AIN FLOOR DRAIN DRAIN COLLECTION HEADER (/) x w PRIMARY SYSTEM DRAIN PALISADES PLANT TANK STEAM GENERATOR REPAIR REPORT PRIMARY HEAD DRAIN SYSTEM Figure 3.4-1 1*.;.
STEAM GENERATOR PALISADES PLANT STEAM GENERATOR REPAIR REPORT WIDE RANGE LEVEL TRANSMITI'ER Figure 3. 5-1
PALISADES PLANT SGRR 4.0 REPLACEMENT PROGRAM AND PROCEDURES This section discusses the engineering evaluation of various activities required to implement the steam generator repair. It should be noted that implementation methods and procedures may vary from that described below as engineering is finalized. The methods below are provided to demonstrate feasibility of implementation. 4.1 CONSTRUCTION CONSIDERATIONS 4.1.1 SITE PREPARATION The plant site will be prepared as necessary for the following activities pertaining to the replacement of the existing two steam generators. As described, none of these .activities will impinge on safety-related underground piping, conduits, or electrical duct banks.
- a. Receipt of Two New Steam Generators The new steam generators will be offloaded using a newly constructed barge slip which is described in Section 4.1.1.1.
- b. Storage of New Steam Generators If the new steam generators are not installed upon arrival at the site, they will be stored in compliance with the storage criteria listed in Section 2.4.
- c. Preparation of Outside Area Adjacent to Containment Building Construction Opening As shown in Figure 4.1-1, the foundations will be installed in this area to support the rigging equipment used for transporting the steam generators through the containment' wall construction opening. The rigging equipment description is given in Section 4.1.2.2. In order to accommodate the rigging equipment and to provide adequate space for handling the steam generators, an additional area at el 625'-0'Lwill be built-up with a structural backfill and held by a retaining wall. The containment building wall has been evaluated for the additional loads imposed due to the rigging equipment that is loaded with a steam 4-1
PALISADES PLANT SGRR generator to ensure that there are no adverse structural effects.
- d. Transportation of Steam Generators The steam generators will be transported between the steam generator storage area and the containment building along the access route as shown in Figure 4.1-1. The plant access road will be widened to accommodate the width required for the steam generator transporters. *Borings will be taken along the access road to confirm that the soil bearing capacity is adequate before steam generator transportation. There are no safety-related underground pipes, cables, etc, involved along the roufe, except for a fire protection line which is adequately protected by a minimum overburden of 5 1 -6 11 *
- e. Temporary Construction Facilities Adequate construction facilities will be provided as necessary for the labor force and material storage. Controlled access areas will be designated. Such facilities are shown in Figure 4.3-1 and are described in Section 4.1.1.3.
Supplemental access control is described in Section 4.3.1. Temporary construction fences will be installed to delineate construction areas and to provide construction security. 4.1.1.1 Barge Slip Facilities A temporary barge slip will be constructed within the existing plant area as shown in Figure 4.1-2. See Section 7.2.3 for dredging and construction details. A hopper type barge will be used to transport the new steam generators to the Palisades Plant jobsite. As presently contemplated, the steam generators will be offloaded from the barge by lifting them with a mobile jacking frame. The jacking frame loaded with a steam generator will be brought. to the dock at el 588'-0" (+). At this point the steam generators will be transferred to the waiting crawler transporters. The new steam generators will then be transported to the steam generator storage area pending installation. 4-2
PALISADES PLANT SGRR 4.1.1.2 Steam Generator Storage Facilities The storage requirements are different for the new and old steam generators. 4.1.1.2.1 Storage of New Steam Generators If storage of the new steam generators is needed, the storage facilities will conform to the criteria set forth in Section 2.4. 4.1.1.2.2 Storage of Old Steam Generators The requirements and design criteria for a storage building for the old steam generators are described in Section 4.4.2. If this building is necessary, adequate biological shielding will be provided in the form of 18-inch thick normal weight (150 lb/ft3) concrete walls all around its periphery (see Section 4.4.6). The steam generator storage building will have a removable metal deck roof to facilitate rigging and handling of the steam generators (see Figures 4.1-3 and 4.1-4). This building is located in a plant area where there are no safety-related structures or equipment. 4.1.1.3 Temporary Construction Facilities Existing plant warehouse and offices will be utilized, as far as possible, to support the construction force and store equipment and tools during the steam generator repair effort. Exact areas and details relating to temporary construction facilities will be developed before the construction phase. 4.1.1.4 Containment Preparation 4.1.1.4.1 Defueling of the Reactor All fuel assemblies will be removed from the reactor and stored in the spent fuel pool before the commencement of the repair activities that could affect plant safety. 4.1.1.4.2 Equipment and Material Modifications Equipment and material will be temporarily relocated to provide rigging clearances as discussed in Section 4.2. Rigging equipment inside the containment is discussed in Section 4.1.2. 4-3
PALISADES PLANT SGRR 4.1.1.4.3 Laydown Areas Laydown areas for the major equipment to be relocated inside the containment will be provided as shown in Figure 4.1-5 as follows:
- a. Reactor Vessel Thermal Shield The reactor vessel thermal shield which is presently stored at el* 649'-0" will be removed from the containment through the construction opening.
- b. Main steam Line The laydown area for the main steam lines will be located on the northwest side of the 649-foot level.
- c. Missile Shields The missile shields over the control rod drive mechanisms may be stored in their normal laydown areas on the 649-foot level or left in place over the reactor vessel.
It is noted that these laydown areas are provided to demonstrate laydown feasibility; the final design may alter the configuration and scope. 4.1.1.4.4 HVAC During Steam Generator Repair Program During the steam generator repair program, the containment will be kept heated, ventilated, and air conditioned as required by seasonal conditions. The existing air purge supply and exhaust system, containment air cooler recirculation fans, and unit heaters will be kept operational. Additional construction-related ventilation equipment, including portable fans, hoods and filters, will be used to remove fumes associated with welding and cutting operations. A temporary construction covering will be provided over the construction opening to reduce the infiltration of dust, sand, and water into the containment. Control of airborne radioactivity during the steam generator repair program is discussed in Section 4.3.3. 4-4
PALISADES PLANT SGRR 4.l.l.4.5 Floor Drains and Sumps Floor drains and sumps will be protected during the repair outage to prevent their becoming clogged with construction debris. On completion of the repair the drains will be inspected and cleaned as required. 4.1.2 RIGGING A rigging scheme has been developed to replace the steam generators which would avoid hazard to the fuel stored in the spent fuel pool. This scheme will also not adversely affect safety-related equipment, systems, and structures necessary to maintain the spent fuel pool cooling and makeup capability. This rigging scheme utilizes existing containment building structural supports and is described in the following ~ections.. The rigging scheme is subject to improvement or modification as the design progresses and/or to suit the equipment requirements of the selected rigging subcontractor; however, future modifications of the rigging scheme will not adversely affect the spent fuel pool cooling and makeup capability or the safety-related structures or equipment. 4.1.2.1 Rigging Inside Containment As presently contemplated, the rigging scheme developed for inside the containment to replace the steam generators is shown in Figures 4.1-6 through 4.1-10. The major pieces of the rigging equipment, as discussed in Section 4.1.2.3, will be brought into the containment through the construction opening. The center pole section is erected on a steel platform which is supported on the biological shield walls of the reactor cavity at el 649 1 -0 11
- The rotating temporary semi-gantry crane is supported by.the center pole section and the existing polar crane rail. The rptating temporary semi-gantry crane header beam with its lifting hardware is positioned over the old steam generator, and the lifting collars will be installed on the steam generator's trunnions.
The steam generator is- freed from all piping and restraints as discussed in Sections 4.2.4 and 4.2.5. The steam generator is then lifted clear of all obstruction to approximately el 660 1 -6 11 as shown in Figure 4.1-8. The semi-gantry crane is rotated about the center line of the containment building approximately 90 degrees until the steam generator is located directly over the radial center 4-5
PALISADES PLANT SGRR line of the construction opening. The steam generator is rotated approximately 47 degrees about its vertical axis and then lowered horizontally onto the tilt-up sled. The rigging tilt-up hardware is attached to the lower nozzle of the steam generator. The tilt-up sled is assembled with a vertical yoke which pins up to the ears on the steam generator's tilt-up hardware. As shown in Detail 2 of Figure 4.1-10, the tilting sleds are equipped with machinery dollies and travel on a runway system. As the steam generator is lowered, the tilt-up sled is progressively winched out of the construction opening along the runway as shown in the same figure. The steam generator down-ending procedure sequence is shown in Figures 4.1-9 and 4.1-10. In Sequence 1, two additional tilting sleds are positioned on the runway behind the No. 1 tilt-up sled. In Sequence 2, the lower section of the steam generator has been seated on the No. 2 sled and the lower nozzle tilting mechanism is free of the No. 1 sled. In Sequence 3, the lowering procedure has been completed and the large diameter portion of the steam generator now rests on the self-adjusting saddles of the No. 3 sled. At this point the steam generator is supported completely by the runway system and portion of the elevator beam outside the containment. The steam generator is freed from the semi-gantry crane hoists by removing the lifting collars from the steam generator trunnions, and the No. 1 tilt-up sled is also removed from the runway. Then the steam generator is slowly winched out of the containment onto the elevator beam portion of the runway. From this point the steam generator is lowered by the jacking frame onto transporters as described in Section 4.1.2.2 and moved to the steam generator storage laydown area. Removal of the old steam generators will serve to satisfy
- the load test requirements of rigging equipment inside the containment building. The old steam generators will be loaded to achieve the required test load.
Installation of the new steam generators will follow a reverse procedure similar to that given above for the removal of old steam generators. 4-6
PALISADES PLANT SGRR 4.1.2.2 Rigging Outside Containment The rigging scheme outside the containment as presently being considered is shown in Sequences 1 through 4 of Figure 4.1-11. The old steam generator is rolled out on the tilting sleds along the runway as described in Section 4.1.2.1. The old steam generator is positioned on the elevator beams supported by the jacking frames so that the lift points of the steam generator align with the jacking bars of the jacking frame as shown in Sequence 1. Once in position, the steam generator, together with elevator beam and sleds, is secured with lifting slings to the jacking system. Holddowns between steam generator and sleds and between sleds and elevator beams are attached, securing the steam generator in place as shown in Sequence 2. The lifting slings are raised until snug, allowing the elevator beams to be disconnected from the runway. In Sequences 3 and 4 the steam generators, sleds, and elevator beam assembly is then lowered and secured into the transporters that have been positioned under the jacking frame assembly. The slings attaching the steam generator assembly to the jacking system are removed. The steam generator is then ready for transport to the steam generator storage laydown area as described in Section 4.1.2.4. Installation into the containment of the new steam generators will follow a reverse procedure similar to that given above for removal of the old steam generators. 4.1.2.3 Rigging Equipment As presently contemplated and as shown in Figures 4.1-6 through 4.1-17, the following major pieces of rigging equipment will be utilized for handling, transporting, and removing/installing the steam generators. 4.1.2.3.1 Inside Containment Building
- a. Rotating semi-gantry crane consisting of pole section, box girders, header beam assembly, sill beams mounted on trucks, load blocks and hoisting lines, including swivel spreader beam
- b. One downending and 2 tilting sleds mounted on machinery dollies
- c. Runway system 4-7
PALISADES PLANT SGRR 4.1.2.3.2 Outside Containment Building
- a. Jacking system and elevator beams, including sleds
- b. Transporter system 4.1.2.3.3 At Barge Slip Area
- a. Travelling jacking system, including elevator beam
- b. Transporter system 4.1.2.3.4 At Steam Generator Storage Area
- a. Jacking frame, including elevator beams
- b. Transporter system 4.1.2.3.5 At Access Road
- a. Transporter system
- b. Tilting sleds 4.1.2.4 Steam Generator Transportation*
Before transporting the steam generators between .the containment building and the steam generator storage area, the access route will be completely surveyed and all obstructions removed. Soil borings will be taken to confirm that the access path has sufficient strength. The maximum soil bearing load under the crawler tracks is expected to be 5 ksf. This access path has been used frequently by heavy construction traffic for miscellaneous work within the plant area. The maximum grade is 7% with a minimum turning radius of 42'-0" feet along the access route. After removal from the containment building, the old steam generator will be moved by transporters to its laydown area near the steam generator storage area. The path taken by the transporters is shown in Figure 4.1-1. The new steam generator will be loaded on transporters and then transported along the north access road to the containment building for installation. The second old and new steam generators will be transported in a similar manner. 4-8
PALISADES PLANT SGRR The types of transporters that are presently being considered for transportation of steam generators are shown on Figures 4.1-13 and 4.1-14. There are existing underground fire lines, storm drains, and culverts involved at different places along the route. Adequate earth cover or other temporary protection will be provided to ensure safe transportation of steam generators over these underground utilities. 4.1.2.5 Rigging Controls Rigging operations and equipment associated with the steam generator repair program as described in Sections 4.1.2.1 and 4.1.2.2 will be closely monitored and prequalified to ensure the safe and efficient handling of steam generators. All rigging operations and procedures shall be developed and performed by experienced personnel. The rigging equipment will be load tested or prequalified in accordance with established industry codes and standards before the handling and installation of new steam generators. Throughout every phase of the rigging project detailed inspection and work procedures shall be used to ensure the timely and proper performance of all the rigging operations. Stringent administrative controls will also be employed to ensure overall safety and to preclude damage to equipment or structures during transportation, handling, and installation/removal of steam generators. A scale model of the containment and rigging equipment as described in Section 4.9 will be utilized to further verify the rigging sequence pertaining to rigging operations inside the containment building. 4.1.3 RIGGING LOAD SUPPORTS All design loads imposed by rigging equipment over existing containment building walls, floors, and other members are identified in Figure 4.1-18. 4.1.4 CONSTRUCTION-RELATED INCIDENTS The following unlikely incidents have been postulated during handling of steam generators, including the load testing:
- a. Dropping new steam generators while being offloaded from the barge at the barge slip area 4-9
PALISADES PLANT SGRR
- b. Dropping old/new steam generators during transportation
- c. Dropping old/new steam generators during rigging operations at the steam generator storage area
- d. Dropping old/new steam generators during rigging operations adjacent to the containment building
- e. Dropping old/new steam generators, rigging, or other equipment during construction activities inside the containment The locations of hypothetical incidents a. through c. are physically far enough removed so as not to affect any safety-related structures or equipment. Hypothetical incidents d. and e. could result in damage to the containment structure; however, they would not present a hazard to fuel pool cooling and makeup capability. None of the. fuel cooling and makeup system equipment is located inside the containment nor in areas which could be exposed to any of the postulated incidents. Since the reactor will J:>e defueled, any damage to safety-related equipment inside the containment as a result of the unlikely drop of steam generator, rigging, or other equipment during construction activities will not involve any safety considerations relative to the safe shutdown condition of the plant. Even though the above postulated events are not likely to occur, precautions will be taken to further minimize any damage to the containment building or safety-related equipment. In addition to the rigging controls described in Section 4.1.2.5, construction equipment, such as cranes required for/erection of rigging equipment outside the containment, will be positioned or rigged to preclude any possibility of their having a significant impact on Class I structures or safety-related equipment needed for fuel pool cooling.
- 4.1.5 CONTAINMENT STRUCTURAL CONSIDERATIONS The new steam generators will be similar in size and weight to the existing steam generators. The small changes in the centers of gravity and weights (see Table 2.1-1) for the new steam generators have an insignificant effect on previously calculated loads; therefore,_ no modifications to the support system will be necessary. The only structural considerations involved are those related to the support of temporary rigging loads for the removal and replacement of 4-10
PALISADES PLANT SGRR the steam generators, the construction opening in the containment building wall, and the laydown areas required by the construction activities inside the containment; however, because of new elevations of main steam lines, the main steam line supports will be modified. Preliminary structural analyses have been made for the containment shell and internals to ensure that structural integrity of the containment building will remain essentially the same as that of the existing containment structure. The design criteria for the containment building structural repairs will be based on the same criteria as outlined in Appendix B of the Palisades Plant FSAR for Class 1 structures. After the steam generator repairs are completed, an integrated leak rate test will be performed in accordance with the Palisades Plant Technical Specifications to ensure the leaktight integrity of the containment building. A modified structural integrity test, similar to that described in the FSAR, will be performed for the area affected by the containment construction opening. 4.1.5.1 Construction Opening in Containment Wall For purposes of rigging and handling the steam generators, a preliminary construction opening size was chosen as shown by dotted lines in Figures 4.1-19 and 4.1-20. The centerline of this opening is located at azimuth N 118° 40'. The bottom and top elevations are at el 650'-0" and 688 1 -0 11 , respectively. The width of the opening is 37 feet along the centerline. The containment structural shell analysis has been based on this opening size; however, the rigging scheme presently under consideration permits an opening smaller in size than the preliminary opening originally selected. The configuration and size of the latter opening is shown by solid lines in Figures 4.1-19 and 4.1-20. The bottom and top elevations of the new opening are at el 649' and el 679', respectively. The width of the smaller opening is 32 feet along the horizontal centerline. Results of the shell analysis based on the larger preliminary opening will not be significantly altered by the smaller opening. 4-11
PALISADES PLANT SGRR 4.1.~.2 Tendon Detensioning and Removal 4.1.5.2.1 Criteria for Detensioning/Removing Tendons Before the construction opening is made in the containment, the hoop and vertical tendons passing through the opening must be detensioned and removed. Moreover, additional vertical and hoop tendons must be detensioned in order to ensure (see Figure 4.1-20):
- a. That flexural and membrane stresses everywhere in the containment shell shall be within the allowable limits t~roughout the construction period
- b. That the prestress state at the opening area shall be approximately zero
- c. That the prestress distribution around the opening shall be small after the opening is made 4.1.5.2.2 Method for Detensioning/Removing Tendons Based upon th~ above considerations, the following scheme for tendon detensioning and removal has been developed.
- a. Detensioning and/or removing vertical tendons between azimuth N 48° 40' and N 188° 40' (70 vertical tendons)
- b. Detensioning and/or removing hoop tendons between buttresses located at N 85° and N 145° and.between elevation 644' and 684' (54 hoop tendons)
Vertical and hoop tendon detensioning sequences will start from the center and proceed symmetrically with respect to the centerlines of the construction opening. 4.1.5.2.3 Finite Element Modelling of the Containment Shell The BSAP computer program (Reference 1) was used to estimate the load combination of dead load and the prestress re-distribution due to tendon detensioning and/or removal. Shell elements were used to simulate the buttresses, wall and dome, and beam elements were used to simulate the ring girder. A total of 1,401 nodal points and 1,508 elements are used in the finite element modelling of the containment shell structure as shown in Figure 4.1-21. The isometric computer plot is shown in Figure 4.1-22. The sustained 4-12
/ PALISADES PLANT SGRR Young's modulus of elasticity, E = 2.1 x 106 psi, and Poisson's ratio, 0.17, were the concrete propeities used in the analyses of the containment shell. The effective prestress forces, which were obtained after considering all the prestress losses, are 370 Kips/ft per group, 290 Kips/ft, and 665 Kips/ft for dome, vertical, and hoop tendons, respectivel~. The tendon force was represented by the anchorage force applied at buttresses and the ring girder and a distributed pressure due to the curvature of the tendon. The wind and earthquake loads provided in the uniform building code are insignificant compared to the prestress forces. 4.1.5.2.4 Results of Containment Shell Analysis The membrane hoop and meridian forces as well as the meridian and hoop stresses at the inside and outside surf ace of the containment shell have been investigated. Results along vertical and horizontal cross sections through the center of the construction opening area have been presented in Figures 4.1-23 to 4.1-28. The compressive strength of the concrete f 1c is 5,000 psi. The allowable stresses for the concrete are as follows: Membrane tensile stress v'f1 = 70.7 psi Flexural tensile stress
~= 212 psi Membrane compressive stress . 3f c = 1,500 psi Flexural compressive stress .6f(: = 3,000 psi Results show that prestress levels anywhere in the containment shell are within allowable limits. The prestress level at the construction opening area is approximately zero.
4-13
PALISADES PLANT SGRR 4.1.5.3 Removal of Concrete and Liner Plate 4.1.5.3.1 Sequence of Material Removal After the necessary tendons are detensioned and/or removed, the construction opening will be made as follows:
- a. Chip the concrete in accordance with the opening size requirements
- b. Cut the rebar
- c. Cut and cap the tendon sheathing
- d. Cut the liner plate The cutting patterns of steel reinforcement, tendon sheathing, and liner plate are shown in Figures 4.1-19 and 4.1-20. These cutting patterns have been chosen in order to facilitate the cadwelding of rebar, splicing of tendon sheathing, and welding of liner plate for closing the opening.
4.1.5.3.2 Finite Element Analysis of Containment Shell With Opening As shown in Figure 4.1-21, shell elements in the construction opening area have been removed in the finite element analysis. The isometric computer plot is shown in Figure 4.1-29. Results of this computer analysis for estimating the str~ss re-distribution along the horizontal and vertical cross sections of the containment wall due to the presence of the construction opening are shown in Figures 4.1-30 through 4.1-35. These sections are taken along the centerlines of the construction opening. Results show that the stress levels anywhere in the containment shell are within the allowable limits and the stress level around the opening area is low. 4-14
PALISADES PLANT SGRR 4.1.5.4 Closing the Construction Opening 4.1.5.4.1 Creep Effect Consideration After the opening is closed, the replaced concrete will undergo creep. The maximum differential creep strain between the construction opening and other area of the containment can be derived from the FSAR as 0.22 x 10"6 in./in./psi. This value is obtained by assuming that the replaced concrete will undergo the largest creep strain, while the creep strain in the other area is zero. This uniform differential creep strain can then be represented by the equivalent thermal load as follows:
-a~T = 0.22 x 10~ x 1500 Where a (0.000005 in./in./F) is the ~hermal coefficient of the concrete and 1500 psi is the allowa~le compressive stress.
This leads to the temperature difference of ~T = -66F. T.o predict the creep effect,* a factor of 1.05 has been imposed, resulting in the temperature difference of 69.3F. The finite element modelling shown in Figure 4.1-22 has been used to estimate the maximum prestress loss due to the maximum differential creep strain. Resurts are shown in Figures 4.1-36 and 4.1-37. Additional reinforcement bars will be provided to carry the membrane tensile forces resulting from these creep effects. In this manner, the prestress loss due to creep effect will be avoided. 4.1.5.4.2 Sequence of Closing the Construction Opening After the steam generator replacement is completed, the construction opening will be closed and tendons replaced and re-tensioned. The final containment shell structure will have been restored to its original structural integrity. The general sequence to close the opening is as follows:
- a. Replace the liner plate
- b. Remove the sheathing caps and splice the tendon sheathing
- c. Place the creep reinforcement bars
- d. Restore the existing reinforcement bars 4-15
PALISADES PLANT SGRR
- e. Replace the concrete
- f. Replace and retension tendons
- g. Perform the modified structural integrity test
- h. Perform the integrated leak rate test 4.2 EQUIPMENT AND MATERIAL REMOVAL AND REPLACEMENT 4.2.1 MECHANICAL EQUIPMENT As presently being considered, no major mechanical equipment will have to be relocated because it interferes with the replacement of the steam generators.
As appropriate, equipment within the containment will be covered to ensure cleanliness during the repair. 4.2.2 INSTRUMENTATION The following instrumentation, sensing lines, and associated supports will be temporarily removed and relocated inside the containme*nt:
- a. Sensing lines on Steam Generator E-SOA for Pressure Transmitters PT-0751A through D, Level Transmitters LT-0751A through D, LT-0701 and LT-0702, and Sampling Point SX-0719
- b. Sensing lines on Steam Generator E-SOB for Pressure Transmitters PT-0752A through D, Level Transmitters LT-0752A through D, LT-0703 and LT-0704, and Sampling* Point SX-0718
- c. The sensing line support structures for the sensing lines described in a. and b.
- Disconnection of instrumentation cables to the above transmitters is discussed in Section 4.2.3. The open ends of lines will be capped to ensure cleanliness during the repair.
As appropriate, the instrumentation and sensing lines will be returned to service using standard procedures followed during routine plant maintenance programs. 4-16
PALISADES PLANT SGRR 4.2.3 ELECTRICAL EQUIPMENT 4.2.3.1 Temporary Removal and Relocation of Electrical Equipment Inside the Containment Table 4.2-1 provides a list of electrical equipment and instruments which will be temporarily removed and/or relocated inside the containment because their existing location would lead to interference with the steam generator repair operation. In addition, the removal of 480 V Motor Control Center B09 will result in the interruption of power supply to the loads normally supplied from this motor control center. These loads are listed on Table 4.2-2. 4.2.3.2 Temporary and Permanent Electrical Loads Associated With Steam Generator Repair and the Corresponding Power Source Table 4.2-3 provides a list of the temporary and permanent electrical loads associated with the steam generator repair and identifies the normal and temporary power sources for each load. It is to be noted that the table does not include other plant loads not directly or indirectly associated with the steam generator repair program even though such loads may continue to remain energized for the duration of the repair. As an exception, Table 4.2-3 includes the fuel pool cooling system pumps and associated auxiliary loads 'even though the fuel pool cooling system is not directly associated with the steam generator repair. 4.2.3.3 Temporary Alternate Electrical Power Supplies Temporary alternate electrical power supplies inside the containment will be arranged as follows:
- a. One or more 480 V/277 V power distribution panels
- b. One or more 208 V/120 V lighting and power distribution panels A temporary 2400 V or 480 V feeder will be installed to supply temporary load centers inside the containment as shown on the single lines, Alternates 1 and 2, Figures 4.2-1 and 4.2-2.
4-17
PALISADES PLANT SGRR The temporary alternate power supplies will be used to supply the following categories of loads:
- a. Miscellaneous rigging loads inside the containment during the repair
- b. Permanent loads required to be operational during the repair and supplied originally from the 480 V Motor Control Center B09. Motor Control Center B09 will be removed during the repair program (see Table 4.2-1).
- c. Permanent loads required to be operational during the repair but requiring temporary relocation and supplied originally from power sources outside the containment. Such loads will be powered from temporary alternate power supplies only in those cases where it is not practical to recable to the original penetrations connected to the normal power supply.
4.2.3.4 Relocation and Recabling of Permanent Loads, Instruments, and Devices Inside the Containment Required to be Operational During the Repair Program 4.2.3.4.1 Permanent Loads Normally Supplied from 480 V Motor Control Center B09 and Required to be Operational During the Repair Program As shown on Table 4.2-1, 480 V Motor Control Center B09 inside the containment will be deenergized and removed during the repair work. Therefore, permanent loads normally supplied from MCC B09 which are required to be operational during the repair program will be supplied from the temporary power or lighting distribution panels. Where required, temporary local starters will also be provided. The actual procedure will involve disconnecting all cabling to the motor control center, removing and relocating the motor control center, and recabling to the required loads with temporary cables from the temporary power or lighting distribution panels. 4-18
PALISADES PLANT SGRR 4.2.3.4.2 Permanent Loads and Instruments Connected to Power Sources Outside the Containment and Required to be Operational during the Repair Program The permanent loads and instruments are connected to power sources outside the containment and are required to be operational during the repair program. They fall into three categories:
- a. Devices which do not interfere with the steam generator repair. In addition, the associated cabling is routed via trays and conduits which do not interfere with the repair program.
- b. Devices which directly interfere with the steam generator repair and have to be removed and relocated
- c. Devices which do not directly interfere with the repair. However, the associated cables are installed in trays or conduits that interfere with the repair work and, therefore, will have to be removed.
Category a. devices will not be affected by the repair operation and will continue to function from their normal power sources. In the case of category b. devices, the associated cables will be disconnected and coiled back in the trays. The associated conduits, if any, will also be moved away. After the relocation of the load out of the interference path, the disconnected cable may be temporarily reconnected to the device, if possible, or a new cable may be installed from the penetration to the device. During the repair program, these loads will continue to draw power from the normal source. After the conclusion of the repair work, the devices will be moved back to the original location and reconnected using the old or new cable as the case may be. In the case of category c. devices, the cables will be disconnected and removed from the trays or conduits causing interference, and then the trays or conduits will be dismantled. The existing cables will be temporarily connected back to the devices via alternate routes or new temporary cables will be installed from the penetrations to 4-19 L
PALISADES PLANT SGRR the devices. During the repair program, these loads will also continue to draw power from the normal source. After the conclusion of the repair work, the trays or conduits will be reinstalled in the original position and the existing or new cables will be re-installed as per the original routing. 4.2.3.5 Power Sources 4.2.3.5.1 Avail~bility of Class lE Electrical Systems Before starting major repair activities within the containment, the reactor core will be offloaded and transferred to the fuel storage facility. Therefore, during the steam generator repair, it is not necessary to maintain the availability of the Class lE electrical systems requir~d to provide the capability to shut down the reactor. However, the full core of fuel elements will be stored in the spent fuel pool during the repair. To ensure safe storage of fuel under all foreseeable conditions and to protect against radiation release from irradiated fuel, the fuel pool cooling system and all associated support systems will be kept fully operational. During the repair program, the plant electrical system alignment will be such as to ensure at all times reliable and redundant power supplies to the fuel pool cooling system and all associated support systems. 4.2.3.5.2 Configuration of Offsite and Onsite Power Sources During the Repair Program During the repair program, the ~'quick disconnect" links between the main generator terminals and the isolated phase bus will be removed. This will enable continued power supply to the plant auxiliary power system buses from the 345 kV switchyard via the main transformer and station Power Transformers 1-1 and 1-2 (see Figure 4.2-3). In addition, the 345 - 4.16 kV and 345 - 2.4 kV Startup Transformers 1-1 and 1-2, respectively, will also be available to provide the second source of offsite power during the repair. Both 345 kV circuits, one to the main transformer and the other to the startup transformers, may be deenergized during 4-20
PALISADES PLANT SGRR certain construction phases of the repair program, during which time the diesels or an alternate 345 kV circuit will be utilized. 2400 V emergency diesel Generators 1-1 and 1-2 will be maintained in the "ready for operation" status to provide two sources of onsite power supply during the repair; 125 V de Batteries DOl and D02 and associated battery chargers will also be maintained operational to provide the 125 V de supplies and the preferred 120 V ac supplies through the inve~ters.
~.2.3.5.3 Operation of Systems Related to Plant Safety During Steam Generator Repair The following major systems related to plant safety will be maintained in fully operational status during the repair:
- a. Spent fuel pool cooling system
- b. Service water system, both critical and noncritical
- c. Containment air circulation and cooling system
- d. Compressed air system
- e. Component cooling system
- f. Fire protection system
- g. Radwaste systems
- h. Radiation protection and monitoring systems Reliable operation of these systems related to plant safety is enhanced by ensuring that the availability of electric power sources is not allowed to degrade to levels lower than that required for normal plant operation. This requirement will apply to offsite ac power sources, onsite standby ac
~ower sources, and onsite de power supplies.
The actual configuration of offsite and onsite power sources quring the repair has been discussed under Section 4.2.3.5.2. It is therefore concluded that the operational capability of systems related to plant safety is not co~promised to any degree during the repair program. 4-21
4.2.4 PIPING In order to accomplish the steam generator repair it will be necessary to cut portions of the following major piping systems:
- a. Primary coolant piping
- b. Main steam piping
- c. Main feedwater piping
- d. Steam generator blowdown piping Location of cut areas for reactor coolant system, main steam system, main feedwater system, and blowdown system piping are shown in Figures 4.2-4, 4.2-5, 4.2-6, and 4.2-7, respectively. As appropriate, the open ends of cut piping will be covered to ensure cleanliness during the repair.
The piping will be re-installed in accordance with FSAR criteria as close as possible to the original installation, except for changes required by the systems modifications discussed in Section 3.0 and below. Piping weld end preparations, welding, and nondestructive examination for the installation will be in accordance with the latest edition of the ASME Code. The horizontal portion of the main steam line will be raised 30.2 inches to accommodate the new steam generator flow restrictor nozzle. A spool piece of the same length will be used to connect the raised portion of the new main steam line with the existing steam pipe. Figure 4.2-5 shows the location of the new main steam line. The upward relocation of the steam line will have no significant effects on the stresses in the system. This conclusion is based on the conservative data used in the original calculations and the method of raising the main steam line by adding the additional height in a vertical line above a rigid support. The spool piece will be seismic Category 1 and ASME Code, Section III, Class 2. 4.2.5 CONCRETE AND STRUCTURAL STEEL The following structures or portions of structures within the containment will be removed and replaced to facilitate replacement of the steam generators.
- a. Steam generator upper guide steel supports (see Figure 4.2-8) 4-22
PALISADES PLANT SGRR
- b. Grating platform at el 657 1 -0 11 including supporting members
- c. Portion of peripheral concrete curb at el 649 1 -0 11 around the proposed opening will be removed as necessary to remove interference with temporary runway supports.
4.2.6 COATINGS All coatings removed during the repair program will be replaced with equivalent or better coatings utilizing Consumers' approved procedures. 4.3 RADIOLOGICAL PROTECTION PROGRAM The radiological protection program to be implemented for the repair effort will be in accordance with 10 CFR 20, the Palisades Plant Health Physics Procedures, and the ALARA practices described in Section 4.3.5 of this report. 4.3.1 SUPPLEMENTAL ACCESS CONTROL Facilities will be provided for the repair effort to accommodate the personnel involved. (See Figures 4.3-1 and 4.3-2). These facilities include:
- a. Outside Access Control Point
- 1. Radiological protection training facility
- 2. Craft change area
- 3. Locker area
- 4. Toilet
- 5. Protective clothing pickup area
- 6. Protective clothing dress-out area 4-23
PALISADES PLANT SGRR
- b. Inside Access Control Point
- 1. Radiation control point
- 2. Protective clothing undressi~g area
- 3. Storage area for protective clothing
- 4. Health physics area
- 5. Laundry area The following is a brief description of the access control procedure currently contemplated for entering and exiting the containment.
Personnel will enter the locker area, disrobe, pick up their protective clothing, and dress before proceeding through access control to the personnel air lock or the equipment hatch. The requirements for protective clothing are specified in the Palisades Plant Health Physics Procedures. Personnel leaving the containment will remove their shoe covers, gloves, and other protective clothing at the step-off pad and be frisked for residual contamination. They will then exit through the radiation control point and return to the locker area for their street clothes.* If an individual has been contaminated, he or she will be directed to don a clean set of protective clothing, as necessary, following removal of the contaminated clothing in the undressing area, and be escorted to the decontamination showers in the permanent health physics area in the auxiliary building. Decontamination methods and requirements will be in accordance with the Palisades Plant Health Physics Procedures. Additional health physics support will be provided to the Palisades Plant health physics organization in order to implement health physics related activities. Health physics technicians will be utilized for monitoring and assistance at the steam generators, personnel air lock, equipment hatch, and access control. In addition, television monitors will be utilized for reduction of personnel exposures when visual observations are required for work in high radiation areas. Personnel involved in work areas with a potential for high-level contamination will wear 2 sets of protective clothing. 4-24
PALISADES PLANT SGRR The outer set of protective clothing will be removed when leaving the work area and deposited in a container. The. second set will be removed at the step-off pads discussed above. 4.3.2 LAUNDRY The existing laundry is appropriately sized to accommodate the additional volume during the repair effort. Disposable protective clothing will be utilized when effective. Laundering of protective clothing and cleaning and sanitizing of respiratory equipment will be in accordance with the Palisades Plant Health Physics Procedures. 4.3.3 CONTROL OF AIRBORNE RADIOACTIVITY AND SURFACE CONTAMINATION Airborne radioactivity inside containment during the steam generator repair effort will be controlled, monitored, and ultimately released via the plant vent stack. Air will be drawn through the hatches and construction opening and exhausted by the purge system via the plant ventilation stack, thus precluding airborne radioactive particles or gases from leaving containment openings utilized for construction activities. This air will be conditioned, if necessary, for removal of airborne radioactivity by use of two installed recirculation filters (HEPA plus charcoal absorber) rated at 6,000 cfm each. The air being exhausted from the plant will be monitored as i t passes the existing sampling station located within the ventilation stack. It should be noted that even if credit were not taken for the purge system that the potential offsite dose would still be less than the 10 CFR 50, Appendix I, limits. In addition to bulk containment atmosphere control of airborne radioactivity, appropriate localized control will also be provided. Radioactivity generated during the cutting of the primary coolant pipes will be contained within specially designed contamination control envelopes, which will provide local high efficiency filtration. Personnel working inside these control envelopes will wear respiratory protection equipment, as required, described and implemented by the Palisades Health Physics Procedures. Section 4.3.1 describes the method of controlling the spread of surface contamination by personnel removing their outer set of protective clothing when leaving the control envelope. 4-25
PALISADES PLANT SGRR The radioactive release and dose assessment associated with cutting the primary coolant loop are provided in Sections 4.3.7 and 6.2.2. 4.3.4 SUPPLEMENTAL PERSONNEL MONITORING REQUIREMENTS 4.3.4.1 Monitoring of Airborne Radioactivity Mobile air monitors will be used, as required, to monitor the airborne radioactivity inside the contamination control enclosures and in other work areas inside containment. Airborne radioactivity samplers coupled with laboratory analyses will also be employed. 4.3.4.2 Monitoring of Workers for Ingested Radioactivity Workers who are planning to enter airborne radioactivity or contamination areas will be given an initial whole body count at the start of their employment. Subsequently, workers will be given whole body counts or bioassays, as necessary, to comply with requirements set forth in Palisades Plant Health Physics Procedures. 4.3.4.3 Personnel Monitoring All personnel entering the radiation controlled area will be provided with personnel dosimetry in accordance with the Palisades Health Physics Procedures. 4.3.4.4 Radiation and Contamination Surveys Detailed surveys which provide proper control of radiation and contamination will be performed, as required, throughout the repair effort. These surveys will be performed in accordance with the Palisades Plant Health Physics Procedures. 4.3.4.5 Portable Survey Instruments A description of typical portable survey instruments used at the Palisades Plant is included in Table 4.3-1. 4.3.5 GENERAL ALARA CONSIDERATIONS Personnel exposures will be maintained as low as is reasonably achievable (ALARA) in accordance with 10 CFR 20.l(c), Regulatory Guide 8.8 (Revision 2), and as defined in the Palisades Plant Health Physics Procedures. 4-26
PALISADES PLANT SGRR 4.3.5.1 Use of Scale Model in Radiological Protection Program A scale model of the Palisades Plant containment has been constructed in order to better assess structural and operational parameters that give rise to occupational radiation exposure (see Section 4.9). The model will provide a valuable tool in the continuing study of methods for reducing doses. Among the most important considerations are:
- a. Shielding or Equipment Removal The model, in conjunction with actual field survey data, will be used to study radiation fields to determine temporary shielding requirements. Where shielding may prove difficult or ineffective, the source of exposure will be considered for removal.
The model will aid decisions on shielding or removal of radiation sources including, but not limited to, heat exchangers, drain lines, tanks, and primary coolant pipe segments.
- b. Man-Rem Assessment The model aids in predicting the expected man-rem dose for activities in high radiation areas.
Decisions related to radiation exposure, such as employing local decontamination or determining the number of people required for an activity, can be made early in the design phase of the project in order to incorporate the most effective solutions to the reduction of exposures.
- c. Work Planning The model will be used to develop construction work plans to establish the most efficient procedures for performing work in high radiation areas.
4-27
PALISADES PLANT SGRR
- d. Craft Training The model will be used for the orientation and training of supervisory and key craft personnel to supplement construction work plans to achieve the most efficient utilization.
4.3.5.2 Temporary Shielding Shielding will be used, as necessary, to reduce the dose rates from components such as heat exchangers, valves, and temporary storage areas for contaminated pieces of pipe, cleanup materials, and tools. Temporary shielding will be used, as necessary, for the steam generator and associated piping while it is being cut out of the primary coolant loop and during weld-back operations. The steam generator shell will also help shield the more contaminated parts of the old steam generators. Openings in the old steam generators created by cutting the connecting pipes will be closed by welding plates over the openings. These plates will be supplemented, as required, with lead or other shielding to provide a minimum shielding value that is no less than that afforded by the original piping. 4.3.5.3 Local Decontamination Decontamination of localized areas within the steam generators and primary coolant piping may be performed~ Decontamination of other work areas will be performed periodically, depending on the contamination levels. Paper and plastic sheeting will be used to facilitate collection and cleanup of contamination. In all cases, the Palisades Health Physics Procedures will be followed. 4.3.5.4 Low Background Radiation Waiting Areas Low background radiation waiting areas will be established where workers must wait between tasks. Special signs, tape, or rope-off areas will be utilized to designate these areas. This technique will result in minimum stay times in high background radiation areas, yet provide for higher work efficiency than would the technique of waiting in an area outside of containment. Health physics personnel will work with the job supervisors to ensure that personnel not required in the work area remain in the waiting area. Television camera monitoring will provide supplemental coverage of high radiation areas, and it is anticipated that clothing and/or hard hat (where applicable) color codes will 4-28
PALISADES PLANT SGRR be utilized to provide rapid discrimination of the various work groups. Such coding will allow early detection of individuals entering areas not directly applicable to their work function and thereby reduce unnecessary exposures. 4.3.5.5 Training of Craft Personnel Selected craft personnel will be given a comprehensive course in radiological protection. This course will consist of instruction and demonstrations covering, in detail, the basic theory and practice of radiation protection principles, emergency planning, radiological protection program, and decontamination activity. Additional instructions and training will be provided for those individuals requiring the use of respiratory protective equipment and/or scheduled to work in high radiation areas. This training will involve system familiarization through review of the scale *model and practice with mockup equipment while wearing respiratory protective equipment, as .applicable. The minimum training required for all personnel will be successful completion of the orientation course described in the Palisades Plant Health Physics Procedures. 4.3.6 MISCELLANEOUS WASTE DISPOSAL 4.3.6.1 Concrete Disposal Concrete will be removed from the containment external walls before liner plate removal and will be disposed of as nonradioactive material. This concrete has an insignificant amount of transferable contamination (transferable contamination is considered insignificant if i t is less than 2200 dpm/100 cm2 per 49 CFR 173.397) without surface decontamination. The small amount of concrete removed from areas internal to containment will be considered contaminated and may either be decontaminated before cutting by vacuuming and/or scrubbing with detergent and water to reduce the amount of transferable contamination to as low as is reasonably achievable below 2200 dpm/100 cm2, or may be appropriately packaged for shipment. Following removal from the containment, contaminated concrete will be shipped as "low specific activity" (LSA) material to a licensed land burial site. 4-29
PALISADES PLANT SGRR 4.3.6.2 Miscellaneous Dry Waste Disposal Contaminated metal shavings from the various cutting operations and miscellaneous dry contaminated waste, such as paper and rags, will be put in standard shipping containers and shipped as LSA material to a licensed land burial site. 4.3.6.3 Liquid Radwaste Disposal There are three potential sources of radioactive liquid to be disposed of. These sources are:
- a. Water drained from the reactor coolant system
- b. Laundry wastewater
- c. Local decontamination waste fluids The radioactive releases associated with these sources are discussed in Subsection 6.2.2.4.
The primary coolant will be processed by the chemical and volume control system as described in Section 4.3 of the Palisades Plant FSAR. After appropriate sampling, the laundry wastewater may be discharged without processing through the radwaste system. The estimated activity level for laundry waste effluent is low, as indicated by the effluent total given in Table 6.2-3. Laundry wastes are treated by passage of waste through a series of filters (20 and 5 microns), followed by dilution and release. The filtration process serves to remove a major portion of the radioactivity adhering to laundered garments, particularly the cobalt, manganese, and iron isotopes which normally are present in insoluble particulate form. The small amount of liquid waste generated as a result of local decontamination will be treated either as part of the normal liquid radwaste processing scheme or solidified directly into a sodium silicate matrix on a batch basis if the solutions are incompatible with plant systems. The compatibility of decontamination solutions with existing processing equipment will be determined prior to their use. Sodium silicate solidification methods have been tested at the Palisades Plant for a broad spectrum of waste types, including oils, boric acid, and miscellaneous dirty wastes. 4-30
PALISADES PLANT. SGRR 4.3.7 MAN-REM ASSESSMENT 4.3.7.1 Man-Rem Assessment for Continuing Operation Assuming that the replacement steam generator tubes maintain their integrity during the remaining operating lifetime of the plant, radiation exposure attributed to steam generator work will be reduced. It is not expected to exceed 25 to 50 man~rem per year for a tube inspection operation in accordance with Regulatory Guide 1.83. It has been estimated that approximately 250 man-rem could be saved each year following the steam generator repair. ~.3.7.2 Man-Rem Assessment for the Repair Effort Health physics survey data have been reviewed for the period from November 1976 through March 1978 at various times after shutdown to determine trends in.dose rates and radionuclide contributors that affect operations in the vicinity of the steam generators. (Typical survey results and data are shown in Figures 4.3-3 through 4.3-5.) It is believed that this plant survey data is representative of conditions expected at the start of the repair activity, provided that appropriate dose rate increases due to activity buildup within the steam generators are considered (Reference 3)~ Survey data in the vicinity of the steam generators and primary coolant piping were averaged and used to project general field and dose rate estimates at various times after shutdown (see Figures 4.3-6 and 4.3-7). The man-rem assessment for the repair effort is shown in Table 4.3-2. 4.3.7.2.1 Radiation Field Uncertainties
- a. Radiation fields were taken from actual Palisades Plant surveys and adjusted for activity increase as a function of time before steam generator removal.
In developing the man-rem predictions, it was assumed that the radiation fields would not decay throughout the repair effort. Actual radiation fields will decrease with time. Therefore, the actual total job man-rem are expected to be lower than the calculated values.
- b. The effectiveness of temporary shielding or local decontamination will be further defined as dose rate survey data and primary system samples are gathered during future outages. Reduction factors as now estimated are indicated in Table 4.3-2.
4-31
PALISADES PLANT SGRR 4.3.7.2.2 Assumptions Used to Estimate Manhours by Area for Dose Calculations
- a. Nonwelding Operations in Radiation A;r:ea
- 1. 50% of manhours in radiation area
- 2. 30% of manhours checking in and out through Health Physics and Security
- 3. 20% of manhours in lower radiation area of containment
- b. Welding Operations in Radiation Area Welding operations in the radiation area are based on rotating welder and helper between work area and lower radiation area, as work operations dictat~,
to minimize welders' exposure.
- 1. 35% of manhours in radiation area
- 2. 30% of manhours checking in and out through Health Physics and Security
. 3. 35% of manhours in lower radiation area of containment
- c. Outside Work All manhours are outside of the contai~ent.
- d. Welding of Primary Coolant Pipe The following is based on rotating welder and welder's helper between work area and lower radiation area as work operations dictate to minimize welders' exposure.
4-32
PALISADES PLANT SGRR
- 1. 35% of manhours at primary coolant pipe broken down as follows:
Outside pipe 35% X 68% = 24% Inside pipe* 35% X 32% = 11% Total in place - = 35%
*Inside pipe manhours required to grind and clad inside of primary coolant pipe by conventional manual methods
- 2. 35% of manhours in low radiation area of containment
- 3. 30% of manhours checking in and out through Health Physics and Security
~* stress Relieving of Primary Coolant Pipe
- 1. 10% of manhours inside pipe
- 2. ~ 30% of manhours within 6 feet of pipe
- 3. 30% of manhours in lower radiation area of containment
- 4. 30% of manhours checking in and out through Health Physics and Security
- f. X~Ray and NDT of Primary Coolant Pipe A total of 2,600 hours is allowed to x-ray the primary coolant pipe. This is based on 4 hours availability per day for x-raying inside containment.
- 1. 37% of manhours checking in and out through Health Physics and Security
- 2. 20% of manhours inside of pipe
- 3. 30% of manhours within 6 feet of outside pipe
- 4. 13% of manhours in lower radiation area of containment 4-33
PALISADES PLANT SGRR
- g. Approximately 780 additional hours of time for x-ray technicians are included to x-ray the main steam and feedwater lines.
- h. Rigging
- 1. The following work operations were considered to be outside of the power plant building but inside of the security fence:
(a) Set up equipment to handle the new steam generators from the barge slip to the containment (b) Set up equipment external to the containment to handle the new and existing steam generators (c) Offload, move to storage, and later transport steam generators to containment (d) Remove all external rigging equipment from the site (e) Decontaminate and remove all internal rigging from the site (f) 21% of manhours allowed to install new steam generators
- 2. The following work operations were considered to be at the containment operating floor level or higher:
(a) Install rigging equipment inside containment (b) Decontaminate and remove rigging equipment from inside of containment (c) 60% of the manhours allowed to install new steam generators
- 3. Seventy percent of the manhours associated with 2. above were considered to be at the operating floor while 30% were required for checking in and out through Health Physics and Security.
4-34
PALISADES PLANT SGRR
- 4. Forty-nine percent of the manhours associated with removing the existing steam generators from the containment were considered to be within 6 feet of the primary coolant pipe or bottom of the steam generator, 21% were required to check in and out through Health Physics and Security, and 30% were next to the existing steam generators outside of the containment.
- 5. The manhours associated with moving the existing steam generators to storage were considered to be adjacent to the existing steam generators but outside of the containment.
- i. An assumption was made that 50% of the manhours required to cut the primary coolant pipe would be spent within 6 feet of the outside of the primary coolant pipe or bottom of the steam generators, with partial exposure to the inside of the primary coolant pipe before the steam generator removal.
Thirty percent of the remainder would be spent checking in and out through Health Physics and Security, and 20% would be spent in an area of low radiation.
- j. It was also assumed that 50% of the manhours required to level, line up, and tack the primary coolant pipe would be spent within 6 feet of the outside of the primary coolant pipe, with partial exposure to the inside of the primary coolant pipe after steam generator removal. Thirty percent of the remainder would be spent checking in and out through Health Physics and Security, and 20% would be spent in an area of low radiation.
- k. The manhours associated with miscellaneous piping operations were spread in accordance with the piping hours spent in each area.
- 1. Distributables
- 1. Welder tests and miscellaneous services were considered to be outside the plant buildings but inside the security fence.
4-35
PALISADES PLANT SGRR
- 2. Startup, cleanup, and scaffolding distributables were prorated on the basis of total manual manhours at each location, excluding the outside work.
- m. Nonmanual Labor The following assumptions were made on the nonmanual labor manhours:
- 1. All manhours, except for superinteµdents and engineers, would be expended outside plant buildings but inside the plant security fence (office work).
- 2. Fifty percent of the superintendents' and engineers' manhours for piping, electrical, civil, rigging, and safety would be outside the plant building but inside the security fence (office work).
- 3. The remaining 50% of the superintendents' and engineers' manhours were prorated to each of the work areas based on the total manual hours spent in each area by discipline. For instance, the piping superintendents' and engineers' manhours were prorated on the basis of the total of manual hours spent in performing the piping operations in each area.
4.3.7.2.3 Technique for Estimating Radiation Dose The total dose is dependent on the following factors:
- a. Dose rates (rem/hr) before shielding or decontamination
- b. Shielding or decontamination effectiveness
- c. Duration of tasks (hours)
- d. Manhours required to complete tasks
- e. Fraction of time the task is in radiation field of interest The entire repair program has been divided into discrete areas. The total personnel exposure in an area is the 4-36
PALISADES PLANT SGRR product of the dose rate and the manhours required to complete all tasks involved. The total exposure for the entire job is a summation of the exposure for all areas. Therefore: Ei = Di (rem/hr) x Mi (manhours) E = L Ei (man-rem) where: Ei = Total personnel exposure for area i (man-rem) Di = Average dose rate in area i (rem/hr) Mi = Manhours to complete all tasks in area i (manhour) E = Total personnel exposure for all areas (man-rem) 4.3.7.2.4 Confirmation of Man-Rem Estimate Daily radiation dose logs will be maintained for each worker . stationed within the higher dose rate (~10 mrem/hr) areas of the containment. Weekly, monthly, or quarterly records will be maintained for those working outside the containment and in other specially designated low dose rate areas. These actual doses will be tabulated by task category for confirmation of estimated doses provided in this report. 4.4 DISPOSITION OF OLD STEAM GENERATORS The disposal effort is independent of the repair and is evaluated on that basis. Because of the uncertainty of the timing of the repair and the availability of the offsite disposal facilities, the ultimate disposition of the old units cannot be finalized at this time; however, a variety of disposition alternatives has been investigated. The steam generators to be removed represent the single largest source of solid radioactive waste to be disposed of during the repair effort. The primary side internal surfaces of the steam generators are contaminated by a tenacious film of deposited radioactive corrosion products made up primarily of cobalt, manganese, and iron isotopes .. Isotopic analyses obtained from uncleaned 2-inch long sections of steam generator tubing indicate that at the time the steam generators are removed, each will contain approximately 30 curies of deposited gamma activity (see Table 4.4.l and References 2 and 3). The ~ctivity will_ decrease to approximately 2.8 curies per steam generator 2 years after shutdown, then continue to decay with the 5.6 year half-life _of Cobalt-60. 4-37
PALISADES PLANT SGRR 4.4.1 OBJECTIVES OF HANDLING/DISPOSAL OPERATIONS The objectives of handling/disposal operations are as follows:
- a. To dispose of the steam generators safely and economically
- b. To provide the means to handle/dispose of the steam generators so that radiation exposures to plant and contract personnel are as low as is reasonably achievable
- c. To minimize the release of radioactivity to the environment so as to keep radiation exposure to the public as low as is reasonably achievable-and within the limitations of 10 CFR 20 4.4.2 ONSITE STORAGE If i t is decided that the old steam generators will be stored onsite, a storage facility will be necessary (see Section 4.1.1.2.2).
Before removal from the containment, the openings in the steam generators will be sealed to prevent the release of radioactivity during transfer and subsequent onsite storage (see Section 4.3.5.2). Sealing will be performed by welding plates of steel over each pipe opening. The steel plates will be thick enough or supplemented by lead shielding, if required, so that external dose rates at the sealed opening are not higher than adjacent surface areas. The only significant radiological consideration associated with storage is the direct radiation from the steam generators (see Section 4.4.6). Shielding will be provided to ensure acceptable radiation levels external to the storage facility. Section 4.4.7 demonstrates that there are no credible accident considerations associated with onsite storage of the sealed steam generators that result in the release of radioactivity from the steam generators. 4-38
PALISADES PLANT SGRR Based on the above considerations, the required storage facility design criteria are:
- a. Appropriate shielding for direct dose
- b. Access for periodic surveillance of steam generator seal integrity using portable monitors
- c. Environmental protection in a weathertight and restricted-entry shelter 4.4.3 OFFSITE DISPOSAL The following three methods were investigated as alternative means of shipping the removed steam generators to a licensed land burial site:
- a. Shipment by barge in one piece
- b. Shipment by truck cut up
- c. Shipment by rail cut up 4.4.3.l Preparation for Shipment by Barge Barge shipment of the old steam generators is determined to be the most acceptable method from both environmental and occupational dose standpoints provided that routing and handling capabilities remain available at the time of shipment. The steam generators will be sealed before removal from the containment so that the radioactivity will be contained within a strong, tight package as required by 49 CFR 173. When the steam generators are to be shipped to a licensed land burial site, each one will be transported to the barge facility intact and shipped as low specific activity (LSA) material in accordance with applicable state and federal regulations.
4.4.3.2 Preparation for Shipment by Rail and/or Truck In preparation for shipment of the steam generators by rail or truck to a licensed land burial site, the generators would be cut into sections suitably sized for shipment. The cutup sections would then be packaged in strong, tight packages and shipped with appropriate shielding in accordance with applicable state and federal regulations. 4-39
PALISADES PLANT SGRR cutting operations on the steam generators would be performed in enclosure envelopes, as required, to minimize the spread of airborne radioactivity. The enclosure envelopes will be provided with a HEPA filtration system to reduce the potential release of radioactivity to the environment and will be designed to allow the use of remote cutting techniques to reduce personnel exposure to radiation during cutting. Temporary shielding will also be provided, as required, to further reduce personnel radiation exposure. Radiation detection and measurement during cutting operations will be in accordance with the Palisades Plant Health Physics Procedures. 4.4.3.3 Shipment If shipped by truck, potential disposal sites would be Sheffield, Illinois; Morehead, Kentucky; Barnwell, South Carolina; Beatty, Nevada; and Richland, Washington if in service at the time. Further, if shipped by truck, 14-foot maximum width dimension limits exist, and then only for escorted shipments. This width includes shielding and overpack material. Hence, the net width would be around 12 feet, with maximum length limited to about trail'er length, or 38 feet, without special hauling permits. Shipment by rail or barge limits disposal to Richland, Washington, as this is the only site with rail and ship offloading facilities. For rail, the maximum width, including shielding and overpack, is 10 feet, producing an effective maximum net width of the cutup material of about 8 feet. These maximum dimensions apply for rail and truck shipments only if additional large shields are constructed and receive U.S. Department of Transportation approval permits. Utilizing existing shielded casks, dimensions of the cutup material would be considerably smaller--on the order of 10 feet by 6 feet by 3 feet. If the steam generators were shipped by truck, with a net payload.of 10,000 pounds per shipment, approximately 90 shipments would be required for each of the 450-ton steam generators. Conceivably, a single shipment of both steam generators by barge can be made to Richland, Washington, through either
- the Illinois - .Mississippi Rive_r - Pa_nama Canal route or th~
St. Lawrence Seaway - Panama Canal route. The offloading facilities at Richland, Washington, can easily accommodate a single steam generator at 450 tons. 4-40
PALISADES PLANT SGRR 4.4.4 MAN-REM ASSESSMENTS If the steam generators are shipped by rail and/or truck, they must be cut into suitably-sized sections before shipment. The man-rem associated with this operation will vary depending on the length of time the steam generators are in storage before the actual cutting operation. If barge transport is employed, cutting is not required and handling of the steam generators would be minimized. Based on radiation survey data and anp.logous manhour estimates established in Section 4.3.7, the man-rem associated with barge transport (1 to 5) is a small fraction of the man-rem associated with rail and/or truck shipments (575 to 750) (see Table 4.4.-2). 4.4.5 RADIOACTIVE RELEASES AND DOSE ASSESSMENT ASSOCIATED WITH OFFSITE DISPOSAL The openings in the steam generators will be sealed before the steam generator is removed from the containment building (see Section 4.3.5.2). Since the steam generators will be sealed during storage and eventual shipment by barge, no airborne or liquid radioactive releases are associated with offsite disposal. 4.4.6 RADIOACTIVE RELEASES AND DOSE ASSESSMENT ASSOCIATED WITH ONSITE STORAGE If onsite storage is necessary, a suitable storage facility would be constructed before the removal of the old steam generato~s (see Section 4.1.1.2). Since all openings in the steam gen~rators will be sealed before removal from the containment, no airborne or liquid radioactive releases are expected as a result of onsite storage. As discussed in Section 4.4.7, the radioactivity within the steam generators is immobile. Thus, if seal integrity was lost, releases to the environment would not be likely. Nonetheless, a surveillance program will be implemented comprised of periodic visual inspection of the external surfaces of the lower assemblies, area radiation surveys, and random swipes of the welds sealing the covered openings in the lower assemblies. This surveillance program will provide further assurance that there are no unanticipated releases of radioactivity to the environment. 4-41
PALISADES PLANT SGRR The only contribution, therefore, to the annual dose equivalent to any member of the public is from direct radiation emanating from the storage facility. The storage facility would be shielded, as required, in order to limit the dose rate at the outside of the storage facility to 1.0 mR/hr. The resulting dose equivalent to an individual at the NNE site boundary (@ 2,200 feet) for a full year would be approximately 1. 9 x 10-3 mrem, which is considered an insignificant contribution to the offsite dose. Furthermore, i t is highly unlikely that an -individual would be continuously exposed for a period of 1 year at the site boundary; therefore, the actual annual dose equivalent to any individual at this location will be lower than that given above. 4.4.7 ACCIDENT CONSIDERATIONS ASSOCIATED WITH ONSITE STORAGE The primary concern associated with accidents involving the onsite storage of the old steam generator is the remote possibility for the release of radioactivity to the environment. The majority of this radioactivity is on the primary side surfaces of the lower assembly in the form of a protective corrosive film of metal oxides which is very adherent and refractory. As discussed in Section 4.4.6, an additional measure of radioactivity confinement will be attained by welding cover plates over all pipe connection openings in the old steam generators. Radioactivity could conceivably be released to the environment only if both of the conditions below occurred:
- a. Radioactivity is dislodged from the primary side surfaces.
- b. The lower assembly primary side boundary is breeched.
There are three mechanisms which could potentially dislodge the corrosion film:
- a. Thermal shock
- b. Chemical/corrosive attack
- c. Mechanical shock 4-42
PALISADES PLANT SGRR The old steam generator storage facility would provide a weathertight environment and minimize temperature extremes so that dislodging of corrosion by thermal shock is considered unlikely. Because the steam generators will be drained and sealed against moisture, chemical and corrosive attack is not likely to occur. The possibility of mechanical shock during storage is not great since the storage building would be an engineered structure and not subjected to general use. Even if thermal or mechanical shock is assumed, the tenacious nature of the corrosive film is such that it would not dislodge a significant amount of radioactivity. In addition to the fact that it is highly unlikely for a significant amount of radioactivity to become dislodged from a primary side internal surface, breeching the lower assembly primary side boundary is considered an extremely remote possibility because of the minimum steel thickness of approximately 4 inches. Based on the above, it is concluded that there are no realistic accident scenarios which would result in the release of radioactivity from the generators during the onsite storage interval. 4.
4.8 CONCLUSION
S The steam generators will ultimately be disposed of in a licensed land burial site or decommissioned with the plant. ALARA considerations, economics, and burial site availability will be the factors determining the storage, handling, and shipping techniques employed. 4.5 PLANT SECURITY Appropriate security measures will be implemented to ensure that the security program currently in effect at the site is not degraded during that portion of the steam generator repair program in which nuclear fuel is in the reactor vessel. Appropriate security measures of a reduced scope will be implemented for that portion of the program during which the nuclear fuel is completely contained in the spent fuel storage pool so as to ensure security of the fuel and fuel storage pool auxiliary systems. Pursuant,to Section 2.790(d) 10 CFR 2, the specific security, measures to be implemented will be addressed in a separate submittal withheld from puplic_disclosure and are not included herein. 4-43
PALISADES PLANT SGRR 4.6 PLANT SYSTEMS LAYUP AND STARTUP METHODS Because of the long outage that will be required to perform the steam generator repair, it will be necessary to take measures during layup and startup to minimize the introduction of corrosion products to critical systems or components. The methods currently considered for achieving this goal are presented in the following discussion. In the secondary system the tube side of the feedwater heaters, including the drain coolers, gland seal condenser, and inter- and after-condensers will be placed in a wet layup with condensate quality water having 50 mg/1 hydrazine (N 2H 4 ) and ammonia to maintain a pH range of 9.0 to 9.4. The system will be sampled and analyzed at least weekly for residual hydrazine and pH. Additional hydrazine will be added if the residual decreases to 45 mg/l. The feedwater recirculating system will be used to add and circulate the chemicals only; continuous use would result in aeration of the solution in the condenser. All heater bypass line valves are to be opened during recirculation periods. The condensate polishing demineralizers will be backflushed to clean the septums before being placed in wet layup with the rest of the feed system. The moisture separator drain tank and the shell sides of the f eedwater heaters will be drained as completely as possible for dry layup. All heaters, with the exception of the El and E2 heaters, which are located in the neck of the condenser and are nonisolable, will be purged weekly and blanketed with nitrogen. During startup, the layup water on the tube side of the feedwater heaters will be drained and replaced with condensate quality water. The feedwater is to be recirculated through the condensate polishing demineralizers until normal startup chemistry specifications are met before allowing any water to enter the steam generators. The shell drains of the ES and E6 heaters will be routed to the condenser until they meet normal feedwater chemistry specifications. The steam side of the main turbine condenser will be drained to a level compatible with the feedwater system recirculating requirements. Plastic sheeting is to be placed ov~r the turbine exhaust as it enters the condenser to serve as a vapor barrier isolating and protecting the turbine from moisture in the condenser. On the water side, 4-44
PALISADES PLANT SGRR the waterboxes will be drained and opened to allow the tubes to air dry. Before the tubes dry, the stainless steel air removal section tubes will be brushed or scraped to remove deposits, typical for these tubes, that may encourage pitting corrosion during the extended outage. Before startup, the steam side of the condenser will be hand-cleaned to remove loose oxide scale and debris, with particular attention being paid to the deaerating trays. No special layup provisions are anticipated for the primary system. The primary coolant water will become air saturated beginning with the defueling operation; as with any
*refueling outage it is not feasible to control the oxygen concentration below saturation. For the replacement of the steam generators it will be necessary to drain the primary coolant system to a level below the hot and cold leg nozzles; at this point existing means for recirculating or for feed and bleed of the primary coolant will not be available for control. It is anticipated that the reactor vessel head will be blocked in place above the reactor vessel and sealed with sheeting during the constructio~
activity; this will prevent contaminants from entering the
.vessel. It is likewise anticipated that some means will be employed to cover the vessel legs after they have been cut.
The pressurizer vessel will be drained along with the rest
.of the primary system; no layup requirements are planned.
' The quench tank will be drained as completely as possible and purged weekly with nitrogen. During startup, the normal primary coolant chemistry limits will be observed. Unless changed at a later date, the balance of the plant systems will be placed in a layup condition according to the respective normal plant operating procedures for refueling outages. Likewise, startup of these systems will follow normal procedures.
';rhe instrumentation sensing lines on the primary coolant system will be backflushed into the system with demineralized water after the primary coolant level has been lowered. The sensing lines are normally filled with demineralized water during power operation, but the flushing will guard against the precipitation of boric acid that may have diffused into the lines.
The instrument sensing lines on the shell sides of the feedwater system will be drained wherever possible. Instrumentation included in the wet layup of the tube side of the feedwater system will require no special provisions. 4-45
PALISADES PLANT SGRR 4.7 QUALITY ASSURANCE The quality assurance programs for Consumers Power Company, Bechtel Power Corporation, and Combustion Engineering, Inc. as applied to this project are described in this section. 4.7.1 CONSUMERS POWER COMPANY QUALITY ASSURANCE PROGRAM The Consumers Power Company quality assurance program is described in the Consumers Power Company Quality Assurance Program Topical Report (CPC-1-A). This report is an integral part of the Consumers Power Company Quality Assurance Program Manual for Nuclear Power Plants and will be invoked for those quality assurance activities within Consumers' scope of responsibilities on this project. 4.7.2 BECHTEL POWER CORPORATION QUALITY ASSURANCE PROGRAM Bechtel Power Corporation will perform its duties in accordance with Bechtel Topical Report BQ-TOP-1, Bechtel Quality Assurance Program for Nuclear Power Plants. This topical report will be invoked for quality assurance activities within Bechtel's scope of responsibilities on this project. Responsibility for the Palisades Steam Generator Repair Project has been assigned to the Ann Arbor office. 4.7.3 COMBUSTION ENGINEERING POWER SYSTEM GROUP NUCLEAR QUALITY ASSURANCE PROGRAM The quality assurance program used by Combustion during the design and shop fabrication of the replacement steam generators is described in CE-NPD-210, Quality Assurance Program - A Description of the CE Nuclear Steam Supply System Quality Assurance Program. 4.8 REGULATORY GUIDE APPLICABILITY TO REPAIR PROGRAM Section 2.1.4 discusses regulatory guide compliance during the manufacture of the steam generator units. This section discusses regulatory guide applicability to repair program activities other than steam generator manufacture.
- q. Regulatory Guide 1.31 (5/77), Control of Ferrite Content in Stainless Steel Weld Metal Control of stainless steel welding complies with interim position on Regulatory Guide 1.31 (Branch 4-46
PALISADES PLANT SGRR Technical Position MTEB 5-1, dated 11/24/75) except as discussed below.
- 1.
Reference:
Paragraph B.l.b of the Regulatory Guide Austenitic stainless steel welding filler materials used in the fabrication and installation of ASME Section III, Class 1, 2, and 3 components are controlled to deposit from 8 to 25% delta ferrite except for 309 and 309L welding filler materials, which are controlled to deposit from 5 to 15% delta ferrite and are used only for welding carbon or low alloy steel to austenitic stainless' steel. Use of 309L welding filler material is further limited to the overlay deposit on the carbon or low alloy steel component nozzles or connecting pipe when postweld heat treatment is required. These limits for delta ferrite in austenitic stainless steel welding materials comply with Regulatory Guide 1.31 since the upper limit of 20% delta ferrite does not apply for welds that are not heat treated after welding (Paragraph 3b), except for solution heat treatment. Solution heat treatment, although not required after welding, is permitted in order to avoid sensitization. The procedure for determining the amount of delta ferrite in each heat or lot of austenitic stainless steel welding material does not comply with the regulatory guide. Determination of delta ferrite is in accordance with ASME Section III, Division 1, 1974 Edition, Paragraph NB-2433, except that an undiluted weld deposit is required for each heat of bare wire used with the gas metal arc (GMA) process.
- 2.
Reference:
Paragraph B.2 of the Regulatory Guide This--paragr-aph is complied with for all tests and examinations required by ASME Section 1 III, Division 1, 1974 Edition. 4-47
PALISADES PLANT SGRR
- 3.
Reference:
Paragraph B.3.a of the Regulatory Guide This paragraph is not complied with. Magnetic measurement of production welds for delta ferrite is unnecessary when austenitic stainless steel welding materials are controlled to deposit 8 to 25% delta ferrite based on chemistry, except for 309 and 309L welding materials, which are controlled to deposit 5 to 15% delta ferrite based on chemistry.
- 4.
Reference:
Paragraph B.3.b of the Regulatory Guide This paragraph is complied with for welding material certification.
- 5.
Reference:
Paragraphs B.4.a, b, and c of the Regulatory Guide These paragraphs are not complied with since measurement of production welds for delta ferrite is not performed.
- b. Regulatory Guide 1.43 (5/73), Control of Stainless Steel Weld Cladding of Low-Alloy Steel Components The restrictions of Paragraph C.1.a of the regulatory guide are observed. The remaining requirements of Regulatory Guide 1.43 are complied with when high heat input welding processes such as*
submerged arc-welding (SAW) and gas metal arc-welding (GMA) are used by Bechtel suppliers to clad SA 508 Class 2 forgings or to plate material of equivalent composition. Regulatory Guide 1.43 is not complied with when low heat input welding processes such as shielded metal arc-welding (SMAW) and gas tungsten arc-welding (GTAW) are used in the field by Bechtel or Bechtel subcontractors to clad completed welds and adjacent SA 508, Class 2 material during installation. It is noted in Part B of the regulatory guide that underclad cracking has not been observed in SA 508, Class 2 material welded with the "low heat input" processes. 4-48
PALISADES PLANT SGRR
- c. Regulatory Guide 1.44 (5/73), Control of the Use of Sensitized Stainless Steel
- 1. General Subject to the following statements, the use of unstabilized austenitic stainless steel complies with Regulatory Guide 1.44 for components.which are part of the following:
(a) The reactor coolant pressure boundary (b) Systems required for reactor shutdown (c) Systems required for emergency core cooling (d) Reactor vessel internals that are required for emergency core cooling (e) Reactor vessel internals that are relieq upon to permit adequate core-cooling during any mode of normal operation or postulated accident conditions
- 2.
Reference:
Paragraph C.1 of the Re9qlatory Guide Contamination of austenitic stainless steels (Type 300 series) by compounds that could cause stress-cor~osion cracking is avoided during all stages of fabrication and installation. Except for trichlorotrifluoroethane (TCTFE), which meets the requirements of Military Specification MIL-C~8130 2B, cleaning is limited to solutions that contain not more than 200 ppm of chlorides. Rinsing or flushing is with water containing less than 200 ppm of chlorides. Special rinsing techniques are used to ensure complete removal of TCTFE when crevices or undrainable areas are present. Foreign substances in contact with austenitic stainless steel (die lubricants, penetrant materials_,_ marking materials, masking tape, etc) are controlled so as not to contain more than 200 ppm of chlorides, or are removed immediately following the operation in which 4-49
PALISADES PLANT SGRR they are used. Crevices and undrainable areas are protected before using materials containing more than 200 ppm of chlorides. All substances in contact with austenitic stainless steel are removed before any elevated temperature treatment. In the field, austenitic stainless steel components are stored clean and dry to prevent contamination. System hydrostatic tests are performed with water which contains less than. 200 ppm of chlorides. The influent water quality during final flushing or preoperational testing of the completed system is at least equivalent to the. quality of demineralized water as defined in ANSI N45.2.l - 1973, Cleaning of Fluid Systems and Associated Components During the Construction Phase of Nuclear Power Plants.
- 3.
Reference:
Paragraph C.2 of the Regulatory Guide All grades of austenitic stainless steels (Type 300 series) are required to be furnished in the solution heat treated condition before fabrication or assembly into components or systems. The solution heat treatment varies according to the applicable ASME or ASTM material specification.
- 4.
Reference:
Paragraph C.3 of the Regulatory Guide All austenitic stainless steels are furnished in the solution heat treated condition in accordance with the material specification. For material solution heat treated by the material manufacturer, testing to determine susceptibility to intergranular corrosion is performed only when required by the material specification. During fabrication and installation, austenitic stainless steels are not permitted to be exposed to temperatures in the range of 800 to lSOOF, except for welding and hot forming. Welding practices are controlled to avoid severe sensitization, as described in 6. below, and solution heat 4-50
PALISADES PLANT SGRR treatment in accordance with the material specification is required following hot forming in the temperature range 800 to 1500F. Unless otherwise required by the material specification, the maximum length of time for cooling from the solution heat treat temperature to below 800F is specified in the equipment specification. Corrosion testing in accordance with ASTM A 262-70, Practice A or E, or ASTM A 393 may be required if the maximum length of time for cooling below 800F is exceeded or the solution heat treat condition is in doubt.
- 5.
Reference:
Paragraph C.4 of the Regulatory Guide Use of low carbon (0.03% maximum) unstabilized austenitic stainless steel is not required since the reactor coolant meets the water chemistry requirements at temperatures over 250F.
- 6.
Reference:
Paragraph C.5 of the Regulatory Guide Heat treating austenitic stainless steel in the temperature range 800 to 1500F is not permitted and solution heat treatment is required following hot forming. Since ' sensitization is avoided, testing to determine susceptibility to intergranular attack is not performed.
- 7.
Reference:
Paragraph C.6 of the Regulatory Guide Welding practices are controlled to avoid severe sensitization in the heat-affected zone of unstabilized austenitic stainless steel as described below. Unless otherwise stated, the position applies to both Bechtel suppliers and subcontractors. 4-51
PALISADES PLANT SGRR (a) Weld Heat Input Bechtel controls weld heat input during field installation by using shielded metal arc-welding (SMAW) and gas tungsten arc-welding (GTAW) processes only and by limiting the size of electrodes for each process to 5/32-inch and 1/8-inch diameter maximum, respectively. In addition to these two processes, Bechtel suppliers and subcontractors are permitted to use automatic submerged arc-welding (ASAW) and gas metal arc-welding (GMAW). Hardsurfacing operations are not included. When either automatic submerged arc-welding (ASAW) or gas metal arc-welding (GMAW) is used, or shielded metal arc-welding (SMAW) or gas tungsten arc-welding (GTAW) is used with electrodes larger than those specified above, testing in accordance with ASTM A 262, Practice A or E is required unless welding is followed by solution heat treatment. (b) Interpass Temperature The interpass temperature is controlled so as not to exceed 350F. (c) Carbon Content Susceptibility to sensitization is reduced significantly by selecting material with the lowest reported carbon content. (d) Solution Heat Treatment Solution heat treatment in accordance with the material specification, although not required after welding, is permitted in order to avoid sensitization. Severe sensitization is avoided by not permitting heat treatment in the temperature range of 800 to 1500F following welding. This requires a 4-52
PALISADES PLANT SGRR special technique when welding stainless steel safe ends (transition pieces) to carbon or low-alloy steel component nozzles or piping. Specifically, a 309L stainless steel overlay or an Inconel weld overlay is deposited on the component and the component is postweld heat treated. Following final postweld heat treatment of the component, the stainless steel safe end is welded to the weld overlay using 308 or 308L austenitic stainless steel or Inconel type welding materials. Intergranular corrosion testing is not performed on a routine basis. Performing an intergranular corrosion test for each welding procedure serves no useful purpose when welding practices and reactor coolant water chemistry are controlled as described above.
- d. Regulatory Guide 1.48 (5/73), Design Limits and Loading Combinations for Seismic Category I Fluid System Components Regulatory Guide 1.48 is applicable to the primary head drain lines. These lines shall be analyzed to the requirements of the 1977 edition of the ASME Boiler and Pressure Vessel Code, Section III. The 3/4-inch NPS Class 1 portion of these lines shall be designed in accordance to Subsection NC as per paragraph NB-3630(d)(l).
- e. Regulatory Guide 1.50 (5/73), Control of Preheat Temperature for Welding of Low-Alloy Steel Preheat for welding of low-alloy steel is controlled in accordance with Regulatory Guide 1.50, except as described below.
- 1.
Reference:
Paragraph C.1.a of the Regulatory Guide The regulatory position is complied with when impact testing in accordance with Subarticle 2300 of Section III, Division 1, of the ASME Boiler and Pressure Vessel Code is 4-53
PALISADES PLANT SGRR required. The maximum interpass temperature shall be 500F unless otherwise specified. When impact testing is not required, specification of a maximum interpass temperature in the welding procedure is not necessary in order to ensure that the required mechanical properties are met. The minimum preheat temperatures of Appendix D of Section III of the ASME Boiler and Pressure Vessel Code are required to be met regardless of whether impact testing is required or not.
- 2.
Reference:
Paragraph C.1.b of the Regulatory Guide The regulatory position is not complied with since the welding procedure qualification requirements of Sections III and IX of the ASME Boiler and Pressure Vessel Code are considered to be more than adequate. 3 .
Reference:
Paragraph C.2 of the Regulatory Guide The regulatory position is complied with for Class 1 pressure vessels with nominal thicknesses greater than 1 inch. Maintenance of preheat beyond completion of welding until postweld heat treatment (PWHT) is not required for thinner sections, since experience has indicated that delayed cracking in the weld or heat affect zone is not a problem. current usage of low-alloy steel in piping, pumps, and valves is minimal and normally is limited to Class 3 construction. When low-alloy steel piping, pumps, and valves are used, preheat is maintained until welding is complete but not until postweld heat treatment (PWHT) is performed, since the conditions which cause delayed cracking in the weld or heat affected zone (HAZ) are not present.
- 4.
Reference:
Paragraph C.4 of the Regulatory Guide The regulatory position is complied with when the positions stated above are not met. 4-54
PALISADES PLANT SGRR
- f. Regulatory Guide 1.71 (12/73), Welder Qualification for Areas of Limited Accessibility
- 1.
Reference:
Paragraph C.l of the Regulatory Guide Performance qualifications for personnel who weld under conditions of limited access, as defined in Regulatory Position C.l, are maintained in accordance with the applicable requirements of ASME Sections III and IX. Additionally, responsible site supervisors are required to assign only the most highly skilled welders *to limited access welding. Of course, welding conducted in areas of limited access is subjected to the required nondestructive testing and no waiver or relaxation of examination methods or acceptance criteria because of the limited access is permitted.
- 2.
Reference:
Paragraph C.2 of the Regulatory Guide Requalification is required when any of the essential variables of ASME Section IX are changed, or when any authorized inspector questions the ability of the welder to perform satisfactorily the requirements of ASME Sections III or IX.
~.
Reference:
Paragraph C.3 of the Regulatory Guide Production welding is monitored and welding qualifications are certified in accordance with a. and b. above. 4.9 SCALE MODEL OF THE PALISADES PLANT CONTAINMENT A sc~le model of the containment has been developed to be utilized in the licensing, design, construction, and startup pµases of the Palisades Plant steam generator repair program. The model extends the full height of the containment with detailed presentation of those areas that are directly affected by the steam generator repair and associated systems modifications. The scale is 1/2-inch to 1 foqt. 4-55
PALISADES PLANT SGRR The three-dimensional presentation of the model will provide many beµefits to the repair program, including the following:
- a. The repair scheme will be examineq in detail to ensure that the proper clearances and replacement sequences are maintained.
- b. The model will be used to ensure that the repair program radiation exposures are ALARA (see Section 4.3.5.1).
- c. Communication will be ill)proved on all levels.
- d. The moaei will provide a management tool f9r planning an9 executing construction ope+ations.
4-56
PALISADES PLANT SGRR TABLE 4.2-1 (SHEET 1) ELECTRICAL EQUIPMENT AND INSTRUMENTS TO BE TEMPORARILY REMOVED Number Item Description
- 1. B09 480 V motor control center
- 2. HT-1812 Containment humidity trapsmitter
- 3. TE-1815 Containment temperature element
- 4. HT-1815 Containment humidity transmitter
- 5. TE-1812 Containment temperature ele~ent
- 6. RE-2316 Fuel handling area radiation monitor
- 7. RA-2316 Fuel handling area radiation monitor
- 8. Q-15 Radiation monitor speaker (count-rate)
- 9. RA-2317 Fuel handling area radiation monitor
- 10. RE-2317 Fuel handli~g area radiation monitor
- 11. J-302 Combustion's refueling disconnect panel
- 12. V-49A,B Control rod drive mechanism blowers
- 13. J-301 Refueling disconnect panel
- 14. U-113 Evacuation siren
- 15. TV-2 Security camera
- 16. J-395 Junction box for security TV 17~ L-28 Lighting panel and flqod light di~er
- 18. X38 Lighting transformer
- 19. .POS-3040 Safety Injection Tank T-82A nitrogen valve position switch
- 20. POS-3067 Safety Injection Tank T-82A vent valve position switch
- 21. POS-3044 Safety Injection Tank T-82B nitrogen valve position switch
- 22. POS-3065 Safety Injection Tank T-82B vent valve position switch
- 23. POS-3050 Safety Injection Tank T-$2D nitrogen valve position switch
- 24. POS-3051 Safety Injection Tank T-82D vent valve position switch
- 25. CV 652 Electrical cable tray
- 26. CP 652 Electrical cable tray
- 27. CV 700 Electrical cable tray
- 28. CP 700 Electrical cable tray
- 29. cu 663 Electrical cable tray
- 30. CK 663 Electrical cable tray L
PALISADES PLANT SGRR TABLE 4.2-1 (SHEET 2) ELECTRICAL EQUIPMENT AND INSTRUMENTS TO BE TEMPORARILY REMOVED Number Item Description
- 31. c 4070 Electrical conduit
- 32. c 3010 Electrical conduit
- 33. c 3011 Electrical conduit
- 34. c 3012 Electrical conduit
- 35. c 3013 Electrical conduit
- 36. c 3014 Electrical conduit
- 37. c 3007 ;Electrical conduit
- 38. c 3008 E;1ectrical conduit
- 39. c 4003 Electrical conduit
- 40. c 4004 Electrical conduit
- 41. c 4020 Electrical conduit
- 42. c 4021 Elect]'."ical conduit
- 43. c 4050 Electrical conduit
- 44. c 4051 Electrical conduit
- 45. c 4054 Electrical conduit
- 46. c 4055 Electrical conduit
- 47. c 4060 Electrical conduit
- 48. c 4061 Electrical conduit
- 49. c 4064 Electrical conduit
- 50. c 4065 Electrical conduit
- 51. c 2350 Electrical conduit
- 52. c 3001 Electrical conduit
PALISADES PLANT SGRR TABLE 4.2-2 480 V MOTOR CONTROL CENTER B09 LOAD TABULATION Number Item Description
- 1. MO 3041 Safety Injection Tank T-82A outlet valve
- 2. H6 Fuel tilting device
- 3. Refueling machine
- 4. X38 Lighting transformer
- 5. MO 3045 Sqfety Ihjecti9n Tank T-82B outlet valve
- 6. SV 1043B Pressurizer power relief valve
- 7. MO 3049 Safety Injection Tapk T-82C outlet valve
- 8. MO 3052 Safety Injection Tank T-82D outlet valve
- 9. Fuel transfer machine
- 10. Ll Reactor building crane
- 11. V66 Air room recirculation fan
- 12. Welding receptacles
- 13. Vl6A Pressurizer shed cooling fans and Vl6B
- 14. V49A Control rod drive mechanical cooling fan
PALISADES PLANT SGRR TABLE 4.2-3 (SHEET 1) TEMPORARY AND PERMANENT ELECTRICAL LOADS ASSOCIATED WITH STEAM GENERATOR REPAIR AND THEIR CORRESPONDING POWER SOURCE Power Source Load Normal During Repair Program Miscellaneous rigging Temporary 480 V power equipment inside con- distribution paneis tainment inside the containment Welding receptacles 480 V Motor Temporary 480 V power inside the containment Control Center distribution panels B09 inside the containment Containment cooler recirculation fans: Fans VlA, V2A, V3A 480 V Load No change Center Bl2 Fan V4A 480 V Load No change Center Bll Fans VlB, V2B, V3B 480 V Motor No change Control Center B03 Containment purge 480 V Load No change exhaust Fan V-35 Center Bll Main exha~st Fan V6A 480 V L9ad No change Center Bl2 Main exhaust Fan V6B 480 V Load No change Center Bll Containment purge 480 V Load No chaµge supply Fan V-5 Center B12 Air space purge 480 V Motor No change Fan V-46 Control Center B07
PALISADES PLANT SGRR TABLE 4.2-3 (SHEET 2) TEMPORARY AND PERMANENT ELECTRICAL LOADS ASSOCIATED WITH STEAM GENERATOR REPAIR AND THEiR CORRESPONDING POWER SOURCE Radwaste area supply 4BO V Motor No change Fan V-10 Control Cente,r B07 Radwaste area exhaust 4BO V Motor No change Fan, V-14A Control Center B07 Radwaste a+ea exhaust 4BO V Motor No change Fan V-14B Control Center BOB Waste gas Compressor 480 v Motor No change CSOA Control Center B07 Waste gas Compressor 4BO V Motor No change CSOB Control Center BOB Air room recirculation 4BO v Motor Temporary local Fan V6 Control Center starter supplied from B09 temporary 4BO V power distribution panels inside the containment Spent fuel pool 4BO V Motor No change cooling Pump P51A Control Cent.er B07 Spent fuel pool 4BO V Motor No cpange cooling P~mp PSlB Control Center BOB Fuel pool recirc . 4BO V Motor No change booster Pump PB2 Control Center BOB service water Pumps 2400 v Bus lD No change P7A and P7C
PALISADES PLANT SGRR TABLE 4.2-3 (SHEET 3) TEMPORARY AND PERMANENT ELECTRICAL LOADS ASSOCIATED WITH STEAM GENERATOR REPAIR AND THEIR CORRESPONDING POwER SOURCE Service water Pump P7B 2400 v Bus lC No change Component cooling 2400 V Bus lC No change Pumps P52A and PS2C Component cooling Pump 2400 v Bus lD No change PS2B Screen wash
,,- Pump P4 480 V Bus B14 No change
/
Instrument and service 480 V Bus Bll No change air Compressors C2A and C2C Instrument and service 480 V Bus B12 No change air Compressor C2B High-pressure air Com- 480 V Motor No change pressor C6A (to supply Control center air-operated valves) B07 High-pressure air Com- 480 V Motor No change pressor C6B (to supply control Center air-operated valves) BOB Fire Pump P9A 480 V Bus B13 No change Fire system jockey 480 V Motor No change Pump P-13 Control Center BOS Diesel-driven fire 120/208 V No change Pump P9B controls lighting Panel and battery charger L20 (supplied from 480 V Motor Control Center BOS)
P~LISADES PLANT SGRR TABLE 4.2-3 (SHEET 4) TEMPORARY AND PERMANENT ELECTRICAL LOADS ASSOCIATED WITH STEAM GENERATOR REPAIR AND THEIR CORRESPONDING POWER SOURCE 2400 V emergency 480 V Motor No change diesel Generator Control Center 1-1 auxiliaries BOl 2400 V emergency 480 V Motor No change diesel Generator Control Center 1-2 auxiliaries B02 2400 V emergency diesel 120 V de Load No change Generators 1-1 and 1-2 Centers DlO controls and D20 2400 V and 480 V load 125 V de Loqd No change center ACBs controls Centers DlO and D20 Control power supply 125 v de Load No change for control panels Centers DlO, (various) D20 Power supply for annun- 125 V de Load No change ciators (various) Centers DlO, and D20 Radiation m.oni toring 120 V ac pre- No change system f erred supply Panels YlO, Y20, Y30, and Y40 Reactor building Crane 480 V Motor Temporary 480 V power Ll Control Center distribution panels B09 Containment jib crane 480 V Motor No change Control Center B04
PALISA.D~S PLANT SGRR TABLE 4.2-3 (SHEET 5) TEMPORARY AND PERMANENT ELECTRICAL LOADS ASSOCI~TED WITH STEAM GENERATOR REPAIR AND THEIR CORRESPONDING POWER SOURCE Containment lighting 120/208 V Temporary lighting lighting dis- distribution panel.
*. tribution Panel L28 (supplied "f:i;:-om Motor Con-trol C~nter B09) 120/208 V No change lighting dis-tribution Panel L27 (supplied from Motor Con-trol Center 808) 120/208 V No change lighting dis-
.. tribution Panel
.L29 (supplied from Motor Control Center B02)
Reactor buil~ing light- 120/208 V No change ing (oth~r than con- lighting dis-tainment building) tribution Panel L24 (supplied f :i;om Motor Control Center B07) 120/240 V No change
.lighting dis-tribution Panel L40 (supplied from Motor Control Center l304)
PALISADES PLANT SGRR TABLE 4.2-3 (SHEET 6) TEMPORARY AND PERMANENT ELECTRICAL LOADS ASSOCIATED WITH STEAM GENERATOR REPAIR AND THEIR CORRESPONDING POWER SOURCE Motor-operated valves Motor control No change (various) centers (various) Solenoid valves 125 V de Load No change (various) Centers DlO I D20 Fuel handling area 480 V Motor No change . supply Fan V7 Control Center B07 Fuel handling area 480 *v Motor No change exhaust Fc;:in V8A Control Center B07 Fuel handling area 480 V Motor No change exhaust Fan.V8B Control Center BOS 125 V de battery 480 V Motor No change Chargers Dl5 and Dl8 Control Center BOl 125 V ac battery 480 V Motor No change Chargers Dl6 and Dl7 Control Center B02
PALISADES PLANT SGRR TABLE 4.3-1 (SHEET 1) TYPICAL PORTABLE SURVEY INSTRUMENT SPECIFICATIONS I RADGUN (A68 - 10 KG-SR - VICTOREEN)
- 1. The radgun is a highly sensitive portable instrument designed to measure beta and gamma ra<liation dose rate from background levels to 10,000 R/hr.
- 2. The detector is a Neher-White ionization chamber filled with 10 atmospheres of argon.
The greater mass of argon at pressure makes the detector more efficient than unpressurized air chambers.
- 3. The detector chamber has a beta window which can be covered by the beta shutter. Because of the chamber thickness needed to contain the pressurized gas, it is not very efficient for beta.
- 4. The radgun contains a 60 µCi Kr 85 check source to check the low scale operation.
~- The radgun has three logarithmic ranges, 0.01 to 10 mr/hr, 0.01 to 10 R/hr and 10 to 10,000 R/hr.
II MODEL E-520 (EBERLINE)
- 1. The E-520 is a portable beta-gamma survey instrument.
- 2. Two different detector tubes are utilized:
both are GM tubes.
- 3. The external tube is located in the hand probe and used for lower range detection (0-0.2, 0-2, 0-20, 0-200 mr/hr). The other tube is located within the case and used in a range of O to 2,000 mr/hr.
- 4. The tube in the external probe has a thin wall (30 mg/cm2), and is used with a rotating beta shield.
PALISADES PLANT SGRR TABLE 4.3-1 (SHEET 2) TYPICAL PORTABLE SURVEY INSTRUMENT SPECIFICATIONS III MODEL R0-2 EBERLINE
,, 1. The R0-2 is a box-shaped, top. handled,
- I
'I portable beta-gamma survey instrument.
- 2. The detector is an unpressurized ion chamber.
- 3. The detector has a thin, one mil, beta window with a sliding beta shield. This instrument is a good choice for beta measurements.
- 4. The R0-2 has a basic scale of 0-5 with selector switch positions of 0-5 mr/hr, 0-50 mr/hr, 0-500 mr/hr, and 0-5,000 mr/hr.
IV MODEL PAC-1 SAGE ALPHA COUNTER (EBERLINE)
- 1. The PAC-1 SAGA is a box-shaped portable alpha and gamma detecting instrument used with the external AC-3 detector. This instrument is primarily for alpha contamination surveys.
- 2. The gamma detector is an internal, small GM tube. This is only used with the selector switch in the "2r" position reading the bottom meter scale of 0-2 r/hr. This instrument should not be used to survey gamma fields.
The gamma reading is basically to determine the alpha background reading.
- 3. The AC-3 alpha detector has a ZnS (Ag) scintillation crystal used with a photo
- multiplier tube. The window thickness is 1.5 mg/cm2 of aluminized mylar with an active area of 59 cm2.
- 4. Alpha range meter has basic upper scale of O - 2 K cpm used with selector switch multipliers of xl.O, xlO, xlOO, and xlK, which gives a total range of 0 - 2,000,000 cpm.
- 5. There is a reset button on the handle for returning the meter reading to zero for rapid
PALISADES PLANT SGRR TABLE 4.3-1 (SHEET 3) TYPICAL PORTABLE SURVEY INSTRUMENT SPECIFICATIONS recheck of readings and decreasing recovery time when changing scales. V MODEL PIC-6A
- 1. Th PIC-6A is a portable instrument which measures the dose rate from gamma radiation.
- 2. The detecting element is a gas-filled ionization chamber operating in the proportional (gas multiplication) region.
- 3. Six decades of dose rate, from 1 mr/hr to 1000 r/hr, are measured in two ranges of three decades each.
- 4. A beta window in the bottom of the instrument provides for the detection of energetic beta particles.
VI CD-V700 (VICTOREEN)
- 1. The CD-V700 is a box-style, portable survey instrument with an external GM probe. It is primarily used for low level gamma surveys or for beta contamination detection.
- 2. The detector is an external, plug in GM tube.
The detector probe has a beta window with a rotating beta shield.
- 3. Instrument range - the basic scale reads O - 0.5 mr/hr on the top and O - 300 cpm on the bottom. Both scales are used with the range switch positions xl, xlO, and xlOO.
PALISADES PLANT SGRR TABLE 4.3.2 (SHEET 1) MAN-REM ASSESSMENT FOR REPLACEMENT (The manhour and man-rem estimates have been revised refer to Table C-1-1 to C-1-5.) Average Area <11 Reduction Area Estimated<11 Radiation Man-Rem Factor Man-Rem Manhours in Field Dose (Man-Rem) (Shielding and/or Dose Work Area Radiation Field (rem/hr) Unshielded Decontamination) (Man-Rem)
- 1. Outside of power plant building but within security fence 213,300 .5xl0-6 1.06 1.0 1.06
- 2. Checking in and out through security and health pysics, as well as time spent suiting up, cleaning up, and moving to and from work area for personnel working in radioactive areas 55,300 .0025 138.25 1.0 138.25
- 3. Inside containment near new construction opening 3,550 0.001 3.55 1.0 3.55
- 4. Within 6 feet of outside of reactor coolant pipe or bottom of steam generator before removal of steam generators 5,050 0.03 151.5 1.0 15i.5
- 5. Within 6 feet of outside of reactor coolant pipe after steam generator's removal 19,100 0.01 191.0 1.0 191.0 23,400 Al 234.0 Al 234.0 Al
- 6. Within 6 feet of outside of reactor coolant pipe or bottom of steam generators with partial exposure to inside of reactor coolant pipe before steam generator's removal 750 1.0 750.0 0.05 37.5
- 7. Within 6 feet of outside of reactor coolant pipe with partial exposure to inside of reactor coolant pipe after steam generator's removal 4,400 1.0 4,400 0.05 220.0
- 8. Inside reactor coolant pipe 4,500 40,500 4070.0 200 Al 9.0 1,800 Al 0.1 180.0 Al 300 A2 2,700 A2 270.0 A2
- 9. Low radiation area within containment 41,250 0.001 4L25 1.0 41.25 4,200 A2 .005 *A2 21. 0 A2 21. 0 A2 Revision 3 July 1979
PALISADES PLANT SGRR TABLE 4.3.2 (SHEET 2) MAN-REM ASSESSMENT FOR REPLACEMENT (The manhour and man-rem estimates have been changed refer to Table C-1-1 to C-1-5.) Average Area 111 Reduction Area Estimated 111 Radiation Man-Rem Factor Man-Rem Manhours in Field Dose (Man-Rem) (Shielding and/or Dose Work Area Radiation Field (rem/hr) Unshielded Decontamination) (Man-Rem)
- 10. Within 6 feet of top half of original steam generators 1,100 0.005 5.5 1.0 5.5
- 11. Within 6 feet of top half of new steam generators 8,050 0.001 8.05 1.0 8.05
- 12. Operating floor of containment 15,800 0.005 79.0 0.2 15.8
- 13. Inside containment, above polar crane 1,150 0.001 1.15 1.0 1.15
- 14. Auxiliary building near clean resin tank and cooling water tank 750 0.001 0.75 1.0 0.75
- 15. Auxiliary building near blowdown tank 6,700 0.001 6.7 1.0 6.7
- 16. Spent fuel pool floor 2,750 0.005 13.75 1.0 13.75
- 17. Within 6 feet of the bottom half of new steam generators 3,700 0.010 37.0 1.0 37.0
- 18. Within 6 feet of the outside of the reactor vessel 50 1.0 50.0 1.0 50.0
- 19. Next to the existing steam generators outside of the containment 1,000 0.02 20.0 1.0 20.0 Total 4,993 Al 1,666 A2 1,193 Note:
111 The three man-rem estimates given for work area 8 (inside of reactor coolant pipe) are presented because of three welding techniques under consideration. The ALARA considerations will be the factor determining which technique is eventually used. The total time estimated for cutting, welding, and inspecting inside the reactor coolant pipes is 4,500 manhours. One alternative (Al) is a technique, presently being investigated for feasibility, utilizing manual welding from the outside of the piping. Using this method, only 200 manhours out of a total of 4,500 would be required inside primary coolant piping. The remaining 4,300 manhours would be spent within 6 feet of the reactor coolant pipe (work area 5). The second alternative (A2) is an automatic welding technique for the cladding from inside the piping utilizing remote viewing. This method required.300 manhours inside the piping, and the remaining 4,200 manhours would be spent in a low radiation area in the containment (work area 9). Revision 3 July 1979
PALISADES PLANT SGRR TABLE 4.4-1 ACTIVATED CORROSION PRODUCTS AFTER SHUTDOWN ACTIVITY IN (CURIES x 10-6 )/(2 11 SAMPLE )111 (2l Isotope 0 days 42 days 140 days 200 days 470 days Cr-51 5.85 2.06 0.807 0.04 0.0 Mn-54 1.33 1.23 1.00 0.86 0.47 Co-57 0.60 0.54 0.428 0.36 0.18 Co-58 341.10 228.50 117.18 48.95 3.56 Fe-59 3.08 1.70 0.733 0.14 o.o Co-60 6.16 6.06 5.85 5.73 5.18 Nb-95 0.31 0.14 0.0594 0.01 0.0 Zr-95 0.37 0.24 0.116 0.04 0.0 Total 358.8 240.47 126.24 56.13 9.39 e, Notes 111 The activities established are an approximation which assumes that the majority of the activity is concentrated in the tubesheet. 121 The following technique was used to find approximate activity per steam generator at the time of removal (-200 days):
- 3.5 x 10 5 inches of tube/tube sheet (3.5 x 10 5 inches of tube)(56.13 x 10~ Ci/2 inches of tube) = 9.82 curries/steam genrator at 200 days This should increase by a factor of 2-3 for additional operation of from 3 to 5 effective full power year (Reference 3).
Therefore: (9.82 curies) x 3 ~29.5 Ci/steam generator at
-200 days
PALISADES PLANT SGRR TABLE 4.4-2 MAN-REM ASSESSMENT FOR OFFSITE DISPOSAL 1--~~~~~~~~~~~~~~~~~~Es-"t-ima-ted-Raai-a-"t-ien-F-i~~a~~~~~~~~~~~~~ Manhours (Rem/hr) Man-Rem !3 ) Barge Shipment 220 (l)
.015-.020 1-5 (near term) cutup and shipment by truck and/or 37, 500( 2 ) .015-020 575-750 rail (near term)
NOTES: (1 l Assumes 10% contact with steam generators for length of trip. (21 This estimate is based on 750 manhours required for cutting primary coolant piping (Table 4.3-2, Area 6). The manhour estimate was increased by a factor of 50 to account for the increased cutting time to reduce steam generators to the minimum of 90 pieces required for shipment~ (3) Range in man-rem was included to reflect possible variations in manhour requirements and radiation field uncertainties.
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~.* Figure 4.1-11 --~
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NOTES-POLAR OPENING POLAR 1 11 CRANE AZ. 118°40' CRANE 1.1N1T1AL DOWEL LENGTH AS SHOWN (1'-3 OR 2'-6 ) IS BRACKET 1 BRACKET REQUIRED FOR ADEQUATE TOLERANCE TO AVOID INTERFERENCE WITH OTHER REINFORCEMENT WHEN LOCATING CADWELD SPLICES. 8' IS THE FINAL 37°-o' (PRELIMIN Y CONST. OPNG.) MINIMUM DOWEL LENGTH REQUIRED FOR CADWELD INSTALLATION AND INSPECTION.
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~70° PRESTRESS REMOVAL SCHEMATIC PLAN 1".. 20 1 PALISADES PLANT
_STEAM GENERATOR REPAIR REPORT CONSTRUCTION OPENING AND TENDON DETENSION/REMOVAL Figure 4.1-20
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- 85° 270° CROSS SECTION OF CONTAINMENT WALL (TYPICAL) v**-**-o*
FINITE ELEMENT MESH AT DOME 1/e-1:0* o* z5* 85° 1118°40' 145° 205° 265° I I I I I I FINITE ELEMENT MESH AT CLOSED OPENING ye*-1*-o* I\ ~\ \! / \ \ / I
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PALISADES PLANT STEAM GENERATOR REPAIR REPORT THREE DIMENSIONAL PI.OT OF CONTAINMENT SHELL (WITHDUr OPENING) Figure 4.1-22
I CJ Cl MEMBRANE STRESSES DUE 'IO DEAD PLUS PRESTRESS IDADS Ln ru VERTICAL CROSS SECTION THRU CENTER OF OPENING TENIXlNS DETENSIONED, WALL WITHOUT OPENING MERIDIAN AND HOOP STRESSES IN PSI, HEIGHT IN FEET Cl Cl D 20.DD YD.DD 60.DD IDD.tJD IYD.D 16D.DD ifllJ.00 HEIGHT Cl x CJ Ln ru N205° -D I n
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Ln I TYPICAL HORIZONTAL £ROSS-SECTION Ul w rr Cl OPENING I- CJ Ln u-i r wI z \ -140'
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X-H~~p- EL.588~6° CJ
,CJ ELEVATION Lil ru CJ Cl CJ Lil PALISADES PLANT STEAM GENERATOR REPAIR REPORT MEMBRANE STRESSES VERSUS HEIGHT (WITHOUT OPENING)
Figure 4.1-23
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.MERIPIAN STRESSES DUE TO DFAD PWS PRES'IRESS IDADS VERI'ICAL CROSS SOCTION THRU CENTER OF OPENING TENOONS DE'T.ENSIONED, WALL WIT.HOUT OPENING STRESSES AT WALL SURFACFS IN PSI, HEIGHT IN FEET CJ CJ CJ IJJ N
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WCI _J IJJ LL I CJ CJ ELEVATION CJ ru CJ CJ CJ LC PALISADES PLANT STEAM GENERATOR REPAIR REPORT MERIDIAN STRESSES VERSUS HEIGHT (WIT.HOUT OPENING) Figure 4.1-24
CJ CJ HOOP STRESSES DUE 'ID DEAD PIDS PRESTRESS LOADS
~,
VERTICAL CROSS SECTION THRU CENTER OF OPENING TENDJNS DETENSIONED, WALL WITHOO'I' OPENING STRESSES AT WALL SURFACES IN PSI, HEIGHT IN FEET m N205°
!ElJ.O!J itaD.DD H~I5HT OPENING N118° 40' TYPICAL HORIZONTAL CROSS-SECTION x OPENING
-140'
-120'
-100' OPENING so.5'
[+-l_NNER X-0UTER CJ CJ ELEVATION CJ n.J CJ CJ C! t.n PALISADES PLANT STEAM GENERATOR REPAIR REPORT HOOP STRESSES \7.ERSUS HEIGHT (WITHOOT OPENING) Figure 4.1-25 'l.,. ..
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CJ CJ MEMBRANE STRESSES DUE TO DEAD PWS PRFSTRFSS LOADS CJ l HORIZONTAL cross SECTION THRU CENTER OF OPENING ID TENOONS DEI'ENSIONED, WALL WITHOUT OPENING MERIDIAN AND HOOP STRESSES IN PSI, AZIMUTH IN DEGREES Ci CJ m CJ N205° c;-1-~~~~~~~~~~--..~-1-~~~~~~+-~~-..--~~~~~...--~~~~--.---~~~~--,.~~~~~--.~~~~----. D *.r 0 YD.OD 80.0 l60.!JD 200.lJlJ 2YD.DlJ 280.00 320.00 360.lJlJ rl AZIMUTH
)K Ci Ci Lil ....; OPENING N118° 40' Wm TYPICAL HORIZONTAL CROSS-SECTION Lil I Lil *w OPENING i:r Ci ~ Ci Lil c:i LD wi z
II i:r CJ OPENING 80.5 1 i£l !:::! L.~ w, L: +-MERIII. X-HIZ![ZJP EL.588~6* CJ CJ ELEVATION CJ ru CJ CJ w U1 PALISADES PLANT STEAM GENERATOR REPAIR REPORT
.MEMBRANE STRESSES VERSUS AZIMOTH (WITHOur OPENING)
Figure 4.1.:..26
'\ MERIDIAN STRESSES DUE 'IO DEAD PWS PRESTRESS ID.AUS HORIZCNTAL CROSS SECI'ICN THRU CENTER OF OPENING TENDONS DEI'ENSIONED, WALL WITHOUT OPENING STRESSES AT WALL Su"'RFACES IN PSI, AZIMUTH JN DEGREES U1 m r:::; D D YD.OD 20Ll.DD 2YlJ.DD 280.0D 320.0D 36lJ.DD AZIMUTH N205° Ci 0 U1 m l D OPENING N118° 40' rl
)i:: TYPICAL HORIZONTAL CROSS-SECTION Ci Ln !'.:!
Ci OPENING Wr Ln I Ln w ---~~+-~~---h--160°
!Yo -140' I-~ -120' Lf1 U1 C1 -100 1 .J OPENING 80.5° II 0:: -1 Ci -o Xr::::i EL.588°-6° w :T .J I
T ELEVATION LL +-I~~i'~ER X-!ZlUTER D Cl U1 Ci Ci Ci nJ l PALISADES PLANT STEAM GENERATOR REPAIR REPORT MERIDIAN STRESSES VERSUS AZIMUTH (WITHOUT OPENING) Figure 4.1-27
HOOP STRFSSFS DUE 'IO DEAD PLUS PRESTRFSS LOADS HORIZOl\!"TAL CROSS SECI'IOO THRU CENTER OF OPENil-JG TENOONS DEI'ENSIONED, WALL WITHOUT OPENING STRESSES AT WALL SURFACES IN PSI, 1'...ZIMUI'H IN DEGREES YD.OD 160.DD 2DlJ.DIJ 2YD.D!J 280.!JIJ 320.DD 360.DD AZIMUTH Ci N205° Ci U1
;n DI 1+-INNE'.:R rl X-BUTER
)!:': OPENING N118° 40 1 CJ Ln ~ TYPICAL HORIZONTAL CROSS-SECTION w~ in l OPENING in w lY CJ I-~ .+---~___,f--~----1' - 160. in U1 CJ -140'
-120'
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- J CJ x~
w~ 1- EL.588~6* - l LL ELEVATION CJ CJ Ul r PALISADES PLANT STEAM GENERATOR REPAIR REPORT HOOP STRESSES VERSUS AZIMUTH (WI'IHOUT OPENING) Figure 4.1-28
,1 \
PALISADES PLANT STEAM GENERATOR REPAIR REPORT THREE DIMENSICNAL PI.DI' OF CCNrAINMENT (WITH OPENING) Figure 4.1-29
\
Ci c:J Cl MEMBRANE STRESSES DUE 'IO DEAD PWS PRESTRESS IDADS lD VERTICAL CROSS SECI'ION THRU CENTER OF. OPENING TENDCNS DEI'ENSIONED, WALL WITH OPENING
.MERIDIAN AND HOOP STRESSES IN PSI, HEIGHT IN FEE'I' Cl m
N205° 120.00 !YO.OD 160.DIJ 180.DD HEIGHT OPENING N118° 40' Cl tn~ TYPICAL HORIZONTAL .CROSS-SECTION w:;:: tn l OPENING tn w 1.1 I-~ .+----,--,,-----,r---.--r--i - 160° tn ci -140' UJ wi -120' z a: OPENING 80.5 1 1.1 Cl d]~ L:~ w L:I ELEVATION Cl Cl c:J ru Cl Cl Cl. U1 PALISADES PLANT STEAM GENERATOR REPAIR REPORT MEMBRANE STRESSES VERSUS HEIGHT (WITH OPENING) Figure 4.1-30
\ MERIDIAN STRESSES DUE 'ID DEAD PIDS PRESTRFSS I..OADS VERTICAL CROSS SECTION THRU CENTER OF OPENING TENl:X:lN'S DE"IENSICNED, WALL WITH OPENING STHESSES AT WALL SURFACES IN PSI, HEIGHT IN FEET CJ CJ CJ m N205° 313.IE! 136.11 155.56 115.00 HEIGHT c:; Ln~ w~ OPENING N1l8° 40' Ln I TYPICAL HORIZONTAL: CROSS-SECTION Ln w tr c:; OPENING I- c:; Ln r::i UJ
.!--~~--1-~~1-- 160°
.J I
-140'
(( tr -120'
- Jo x~ 1 we; [J1 OPENING 80.5
.J I Li...
ELEVATION c:; CJ Cj iJ1 PALISADES PLANT STEAM GENERATOR REPAIR REPORT MERIDIAN STRF.sSES VERSUS HEIGHT (WITH OPENING) Figure 4.1-31
&_ 1'.'W.1 HOOP STRESSES DUE TO DEAD PllJS PRESTRESS LOADS VERTICAL CROSS SECTION THRU CENTER OF OPENING TENIXJNS DEI'ENSIONED, WALL WITH OPENING STRESSES AT WALL SURFACES IN PSI, HEIGHT JN FEE:!' Cl CJ C1 m N205°- 20.DD YD.DD 120.DD !YD.DD i6Li.DD lBIJ.DD HEIGHT OPENING N118° 40' TYPICAL HORIZONTAL CROSS-SECTION OPENING
.+---~~-+-~~--;,-- 16Q'
-140'
-120' 1
OPENING 80.5 1+-INNER X-i::JUTER EL.588~6* Cl Cl ELEVATION CJ ru r:I I CJ Ci U1 PALISADES PLANT STEAM GENERATOR REPAIR REPORT HOOP STRESSES VERSUS HEIGHT (w.rm OPENING)_ Figure 4.1-32
')*
*\
---t'
.MERIDIAN STRESSES DUE 'ID DEAD PIDS PRESTRFSS WADS HORIZONTAL CROSS SECTION THRIJ CENTER OF OPENJNG TENI::CNS DETENSIONED, WALL WITH OPENING STRESSES AT WALL SURFACES Il~ PSI, AZIMUTH lN' DEGREES CJ Ci Ci m
N205° YD.D!J BD.!J 160.D!J 2[JIJ.DD 2YD.D!J 290.DD 320.DD 361J.DIJ AZIMUTH Ci in I::! OPENING N118° 40" w~ in I TYPICAL HORIZONTAL CROSS-SECTION in w OPENING 11 CJ I- Ci in c:i LC
.+--~---<-~------ 16Q' w'
z
-140'
--120'
(( Ir CJ 1 dJ C! OPENING 80.5
- L Ci Wm
- L I I+-MERIJJ.
X-H~fZIP CJ w ELEVATION r=i ru Cl CJ Cl i.n PALISADES PLANT STEAM GENERATOR REPAIR REPORT
/
MEMBRANE STRESSES VERSUS AZIMUI'H (WITH OPENING) Figure 4.1-33
CJ Cl MEMBRANE STRESSES DUE TO DEAD PLUS PRESTRFBS LOAOO Cl LC HORIZCl-lTAL CROSS SECI'ION THRIJ CENTER OF OPENING TENDONS DEI'ENSIONED, WALL WITH OPENING MERIDIAN AND HOOP STRF.sSES IN PSI, AZIMUTH IN DErnEES Cl Cl CJ m Cl gC!_-t-="o____Y1D-.-D-D--~~~~~~*'?l~~. .----,2*a-o-.o-0___2~4-D-.D-D---2~8-D-.D-D---3~2-0-.D-D----.36D.DD N205° i AZIMUTH Cl in ~ w~ 4 OPENING N118° 40' in I TYPICAL HORIZONTAL CROSS-SECTION Lfl w Ir Cl OPENING I- Cl in U]c:i _J I .t--~~1---..--.--t- 160° a: -140' Ir -120'
- J Cl x~
wo _J C'1 OPENING 80.5 1 LL. I
+-INNER X-~UTER Cl Cl Cl ELEVATION ru Ci Ci Cl U'1 PALISADES PLANT STEAM GENERATOR REPAIR REPORT MERIDIAN STRFSSF.S VERSUS AZIMO'IH (WITH OPENING}
Figure 4.1-34
c::::; CJ HOOP STRESSES DUE TO DEAD PIDS PRESTRESS LOADS CJ cI:! HORIZONTAL CROSS SECI'ION THRIJ CENTER OF OPENING TENDCNS DE'I'ENSIOOED, W..LL WITH OPENING STRESSES AT WALL SURFACES IN PSI, AZIMUTH IN DffiREES CJ CJ Cl
- r r::l Ci N205° 0 D YO.DD l6D.DD 200.00 2YO.DD 280.00 32lJ.OD 36[LDO rl RLIMUTH
)K CJ Lil ~
w~
. OPENING N118° 40' Lil I j+-INNER X-BUTER TYPICAL HORIZONTAL .CROSS-SECTION Lil e w Ir f- g OPENING Lil c:i
[JJ
.r--~~+--~~-n-- 16Q' w'
z -140' [I -120'
!! CJ dJ~ CJ OPENING 80.5 1
L: ni w-
- L I Cl Cl CJ ELEVATION lD CJ CJ Cl Cl ru I PALISADES PLANT STEAM GENERATOR REPAIR REPORT HOOP STRESSES VERSUS AZJMUTH (WI'IH OPENING)
Figure 4.1-35 I~ ..
('/ Cl r:::J MEMBRANE STRESSF.S DUE 'IO CREEP EFFECI' r:::J HORIZONTAL CROSS SECTICN THRU CENTER OF OPENJNG TENOONS TENSIONED, WALL WITHOOT OPENING MERIDIAN AND HOOP STRESSF.S IN PSI, AZIMUl'H IN DEGREES N
/10')
Cl r::J
./~/
CJ U1 N205' Cl CJ CJ
- m OPENING N118° 40' D
r-1 TYPICAL HORIZONTAL CROSS-SECTION
~
OPENING lJl CJ WC! lJl ~
- lJl w
Ir I-lJl wl Cl c: CJ 320.DIJ AZIMUTH 360.IJIJ OPENING ao.o' z II Ir Cl dJ !:::!
- r: Ci Wm ELEVATION .
L:I Cl r::J Ci U1 l c:J Cl r::J r PALISADES PLANT I STEAM GENERATOR REPAIR REPORT MElvlBRANE STRESSES VERSUS HEIGHT (OPENING CTDSED) Figure 4.1-36 .A
CJ CJ MEMBRANE STRESSES DUE 'IO CREEP ~ CJ r VERTICAL CROSS SECTION" THRU CENTER OF OPENING TENDONS TENSIONED, WALL WITHOUT OPENING MERIDIAN J\.ND HOOP STRESSES n;i PSI, EEIGHT IN FEEI' CJ CJ CJ U1 CJ N205° CJ CJ
- m D
rl
)I(
OPENING N118° 40' Ulo TYPICAL HORIZONTAL CROSS-SECTION WC! Lil~ lJl, OPENING w + Jr 1-0 D se :n 11.18 91.22 !55.56 115.DD Lil c: CJ HEIGHT .+--~~+--~~-- - - 160'
-140' w I -120' z -100' II OPENING ao.s'
[1 CJ dl !::! rom W1 EL.588'..6"
'L ELEVATION CJ CJ Cl U1 I
CJ CJ CJ r I PALISADES PLANT STEAM GENERATOR REPAIR REPORT MEMBRANE STRESSES VERSUS AZIMOTH (OPENING CLOSED) ,. Figure 4.1-37
'1 **~** .1 I
.480 V BUS r)
PENETRATION r- - - - - - - , 480V 1 I
) : LOAD CENTER I
I ,,-...... I I l480-208V I - I r- -- ~-- ---12os/12ov L ,,...........,. :LIGHTING I L- _-_-- -=---_-_:-:_1 ' I 1DISTR I I PANEL LOCAL STARTERS
--~--t j----ryJ-
--~--tr--JXr-
- r--
1 l-------- ____ _J I I I ,
-- --il--JX.r-- PALISADES PLANT STEAM GENERATOR REPAIR REPORT TEMPORARY EI:..ECI'RICAL PCWER
' SUPPLIES - ALTERNATE-1 Figure 4.2-1 l '
- I-'
2400 V BUS r) PENETRATION
)
~2400-480V r -_ - - - .._ ----, 480 V 1 ) : LOAD CENTER I I I ~ I I ~~~-1~~~--i
_,._. : ~480-208V I I r- -~ ~--
---12Qs/12ov I I I ,,,-....... I I ....,____- _ :LIGHTING I - - - - - - - _I I jDISTR L- ------- I r'\
!PANEL
--~-t~ I LOCAL I STARTERS __.,.-...-t 1----'Xt- I I
/
PALISADES PLANT 1-- - - - - - - - - _ _J STEAM GENERATOR REPAIR REPORT
- -=-'- --rx.r- TEMPORARY ELECI'RICAL PCmER SUPPLIES - ALTERNATE-2 .
Figure 4.2-2
t
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0 MAIN TRANSFORMER N'i I START-UP TRAN5FORMER NQ 1-1 STARJ-.UP TRA~FORMfR N~ 1-2
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1 1 l
\PREFE'RR£D A-C BUS Nfl 4 (Y40)
PR£.1£RR£D AC BUS NO 3 (Y30) Figure 4.2-3 ..,.
STM. 6£'N. E*SOA ( £50-13 IS TYP.J BIJT OPPOSITE HAND)
//
CVT POINT-- A \ CUT POINT NEARSIDE", 42 /.D. P/,P/N6__,,{-- FAR SIDE 11 30 1.D. PIP/#6,;... NEARSIDE l/ ~- FAR.SIDE -~ ClJT POINT NC:ARSJDE !). FARS/OF ELEVATION PALISADES PLANT STEAM GENERATOR REPAIR REPORT PRIMARY COOLANT PIPJNG CUT POIN.rS Figure 4.2-4
FUTUR£ LOCATION (\J~ d PRESENT LOCATION (Y)
-+----
CUT POINT---r--i--; CUT POINT 11 36 MAIN STEAM STM. GEN. E-SOA ( E-508 15 TY.i?; 8LJT OPPOSITE HAND) ~ ELEVATION PALISADES PLANT STEAM GENERATOR REPAIR REPORT MAIN STEM1 PIPING CUT POINTS Figure 4.2-5
ci' 5TM. GEN. E- SOA T (£-SO 13 IS TYP.; 1 13JJT OPPOSITE HAND) CUT POINT Cl.JT POINT 11
/8 FE£DWA'T£1?
PLAN PALISADES PLANT STEAM GENERATOR REPAIR REPORT FEEDWATER PIPING CUT POINTS Figure 4.2-6
.....,:f' CUT POINT 2 SVRFACE
'=== <f. 81..0WDOWN LINE SIM. GEN. £-50A
£L£VATION STM. G£N. E-50 8 IS TYP.) BUT OPPOSITE HANC CUT POINT I cb 2' 1301/0M BLOWDOWAI I.IN*~
PALISADES PLANT EL£VATION STEAM GENERATOR REPAIR REPORT ROTATC-D .90° CLOCKWISE' BI.OWDCWN PIPING e** CUI' POINTS Figure 4.2-7
~.. /
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-& I STEAM Gl!N~RATO/l COLO POSITION 5TEAM GEN£RATOI< UPPER PALISADES PLANT
--~(j["Jg_ 5_(.j_fPORT STEAM GENERATOR REPAIR REPORT
,,. .. ' - o~
STF.AM GENERA'IDR UPPER SUPPORI' DE'TAIIS Figure 4.2-8
..I
~)
i FJ'\t.LOUT *ttU.TER I I *.o* I l:L.,14'--c" II EL. 611-0\ I SPE*T F'UEL POOL u.c..u".o* ZA l'N' 3""'6A ID GG°NElfAL f'uNPO.S£ CONSTAIJCTl~N <StJll..DIAIG cxnnAtNMEMT 01------.; TUR81NL AZ.De PALISADES PLANT STEAM GENERATOR REPAIR REPORT ACCESS CONTROL (EL 590'-0 11 AND EL 611 '-0") Figure 4.3-1
I t - - - - *,,_*..:.~_*---+-------~~o~------ J~J:
/ -.r*i i
J.. 4 C.ONPtWl NCI'. LUNCt4 ROO"' l ~~~~:.m~
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\
N* \
. ~ErD£LIN~ t~
v " 0 ----- ) PRIMARY PUMP S[llVIC[ P1'T COVER ACC.f.SS MATCH 'TO [L.590'-0"
'"---~
~
l Roof' EL. CiZS'*d" PALISADES PLANT STEAM GENERATOR REPAIR REPORT
-~------+---
ACCESS CONTROL (EL 649'-0") Figure 4.3-2
(APPROXIMATELY 10 WEEKS AFTER SHUTDOWN DURING 1978 REFUELING OUTAGE) DATE: TI:-IE~:OO INS'i'RU~IENT:~SEiUiU. NO. 2.Z..JD 6 HOURS 24 HOURS DAYS "A" STEAM GENERATOR:
- 1. Bottom of Hot Leg .3S- mr/hr
- 2. Bottom of Cold Leg 3S' mr/hr
- 3. Man ~.Jay of Cold Leg t.:J.~O mr/hr
- 4. Man Way of Hot Leg l.fJDO mr/hr "B" STEAM GENERATOR:
- 5. Bottom of Hot Leg ~ mr/hr
- 6. Bottom of Cold Leg-----~;z=~"""='..-----mr/hr
- 7. Man Way of Cold Leg /Ot:JD mr/hr
- 8. Han Way of Hot Leg //O() mr/hr PRTI-rAR.Y COOLMlT PUNP SUCTION SIDE:
9. 10.
"A" "B"
Primary Primary Coolant Coolant Pump Pump zr
---~2.-8--~rnr/hr mr/hr
- 11. "C" Priraary Coolant Pump 3'0 mr/hr
- 12. Pu;np -~~-2.=-'-/3~~-mr/hr "D" P:::-imary Coolant
- 13. Regenerative Heat Exchanger Line bt:J() mr/hr 5 Feet Above Floor
- 14. Letdown Heat Exchanger Line----.:--"-=--- /7() rnr/hr Bottom of Inlet Piping PRIHARY COOLANT LOOP D?~t..E LI:\E:
- 15. "A" Prim<1ry Coolant Pump Loop lA D:::-<::.in Line mr/hr
- 16. "B" Prinary Coola-:tt Pump Loop B Drain Line -------mr/hr
- 17. "C" Primary Coolant Pump Loop 2A D::-ain Line mr/hr
- 18. "D" Primary Coolant Pu:r.p Loop 2B Drain Lir..e -----~-mr/hr
- 19. Primary Coolant Pump Seal Housing (A and D Pumps) _____A mr/hr D nr/hr
- 20. Pressurizer Spray Valves (Above Shut-Do>-m Cooling Lines~ZS-O_ _mr/hr
- 21. Pressurizer Surge Lines - Two Points !fl mr/hr f.!2 mr/hr
- 22. Control Rod Drive (lf points) Above Insulation Heat #1 mr/hr
#2- - - - - - -mr/hr U3- - - - - -mr/hr #4- - - - - -rnr/br Contact Reading Coll.ected by: ~- ~~
TuCJmic i2~S1grlat ure Reviewed by Radiation Protection Sup_ervisor: ~
* ~ature PALISADES PLANT STEAM GENERATOR REPAIR REPORT PRIMARY COOIANT PIPING CONTACT RADIATION SURVEY Figure 4.3-3
- 1. 0 R/hr (cover off)
Primary Inlet Nozzle 1.2 R/hr View B-B Primary Manway Primary Inlet Primary Outlet 35 mr/hr 35 mr/hr Section A-A PALISADES PLANT STEAM GENERATOR REPAIR REPORT
- AVERAGE RADIATION FIELllS 10 WEERS AFTER SHlJTOOll1N Figure 4.3-4
NOI'E: RFADlliGS REPRESENT RADIATION LEVEIS APPROXIMATELY ONE DAY POST-SHUTDCWN 2 R/hr CONTACT AS-FOUND (NOI' ADJUSTED FOR PRQJECTED INCREASE AT THE.TIME OF THE STEAM 50-200 mr/hr CONT ACT GENERATOR REPI.ACEMENT). 800 mr/hr CONTACT 1.2 R/hr CONTACT AT VALVE 1.5 R/hr CONTACT 800 mr/hr CONTACT
----250 mr/hr GENERAL
- - 40 mr/hr GENERAL 40 mr/hr GENERAL 100 mr/hr GENERAL 50 mr/hr GENERAL 1.5 R/hr CONTACT AT VALVE 100 mr/hr GENERAL PALISADES PLANT STEAM GENERATOR REPAIR REPORT RADIATION SURVEY (EL 607'-6")
Figure 4.3-5
I= 1 *0 1tt111mii1itttHttttttnHtttttttttttttttttttttm~~~1t~t-t+ti-+++-ttt1~*H+IB-H-H+ftmtH~i~-1+1+Hm+ooam1t-+-H--+-+-H+~-W-+-Wu++wmmm+lllll~UDIDIITl11J]nnnnn 10 100 I~ 1,000 PALISADES PLANT DAYS AFTER SHUTDOWN BEGIN PIPE CUTTING STEAM GENERATOR REPAIR REPORT MAXIMUM IX>SE RATE INSIDE le STEAM GENERATORS Figure 4.3-6
- ~;.
1poo t-
... ~
100
" ~
10 1 10 100 1,000 PALISADES PLANT DAYS AFTER SHUTDOWN BEGIN PIPE CUTTING STEAM GENERATOR REPAIR REPORT GENERAL RADIATION FIELD NEAR STF.AM GENERA'IOR PIPING Figure 4.3-7
PALISADES PLANT SGRR 5.0 RETURN TO SERVICE TESTING As a part of the steam generator repair program, a preoperational, startup, and hot functional test program will be conducted to ensure that the plant is safely returned to full power operation. To meet this objective, the prog~am will require testing of all newly installed equipment, as well as testing of those pieces of equipment that have been affected by the steam generator replacement field construction efforts. Additionally, this test program will include tesT.ing of the safety-related equipment in accordance with the Technical Specification's requirements and other testing that is routinely performed after normal fuel reloadings and before return to full power operation. The test program would include the following:
- a. Preoperational checks and inspections to ensure that all newly installed safety-related equipment and equipment that was affected is prepared for functional testing. This would include such things as flushing and cleaning, leak checks, electrical continuity checks, visual checks, instrument calibration checks, verification of valve lineups, and hand rotating pumps.
- b. Functional checks of equipment that has been newly installed. This would include such things as steam generator water level instrument checks, steam generator blowdown performance testing, and steam generator performance testing.
- c. Functional checks, during and after construction, of any equipment that has been affected by repair.
On the basis of a system-by-system review, this would include starting, running and monitoring of pumps, valves, and ancillary systems.
- d. Surveillance of equipment in accordance with the technical specifications current at the time of the steam generator replacement, such as valve operability and pump operability
- e. Startup testing which would normally be performed between routine fuel loading and return to full power operations, such as rod drop tests and low power physics tests 5-1
PALISADES PLANT SGRR
- f. Review of jumper log, etc, to ensure that any temporary jumpers, etc, have been properly dispositioned
- g. Performance of an integrated leak rate test and any other testing that may be necessary to return the containment building to serv~ce
- h. A final overall review of the plant and systems to ensure readiness for return to service and power operati0n 5-2
~ALISADES PLANT SGRR 6.0 SAFETY EVALUATIONS 6.1 FSAR EVALUATIONS 6.
1.1 INTRODUCTION
The-purpose of this section is to evaluate the impact, if any, of the replacement steam generators on the accident analysis transients for the Palisades Plant. Under the guidelines specified in 10 CFR 50.59, such an evaluation is required to verify that no unreviewed safety concerns or changes to the Palisades Plant Technical Specifications occur. This section provides a qualitative discussion of the effect on the accident analyses of steam generator parameter changes resulting from steam generator repair. Conclusions are made in this section concerning the validity of the original FSAR to the repaired units. Consistent with the requirements of 10 CFR 50.59, licensing regulations and guidelines of the original licensing of the Palisades Plant are assumed to apply, and only changes in the safety analyses due to the equipment changes are considered. The relevant plant operating parameters and steam generator design parameters have been compared for the original and replacement steam generators in Section 2.1. While incorporating design improvements that will improve the flow distribution and tube bundle accessibility and reduce secondary side corrosion, the replacement steam generators continue to match the design performance of the original ste~~ generators. It may be noted from Section 2.1 that there is very little effect on plant operating parameters due to the replacement of the steam generators. It is, the~efore, to be anticipated that the impact on the accident analyses will be insign~ficant. The results of the accident evaluation show that no unreviewed-safety concerns exist because of operation with the replacement steam generators. 6.1.2 NON-LOCA ACCIDENTS Analyses of the following non-LOCA 9esign basis events were originally presented in the Palisades Plant FSAR. ~hese events were evaluated to determine the effect, if any, of the replacement steam generators on the plant transient response.
- a. Control rod withdrawal
- b. B_oron dilution e- 6-1 MARCH 1979 REV. 1
r . PALISADES PLANT SGRR
- c. Full-length control rod drop d.* Malpositioning of the part-length control rod group
- e. Loss of coolant flow
- f. Idle loop startup
- g. Excessive feedwater
- h. Excessive load increase
- i. Loss of load
- j. Loss of feedwater flow
- k. Steam line rupture inside containment
- 1. Steam generator tube rupture
- m. Control rod ejection The excessive feedwater, excessive load increase, loss of load, loss of feedwater flow, steam line rupture, and steam generator tube rupture events are discussed in the succeeding subsections. The remaining events are primarily core-related and are not significantly affected by the replacement of the steam generators.
6.1.2.1 Excessive F~edwater An excessive feedwater transient may be caused by a decrease in feedwater temperature or by an increase in feedwater flo~. These conditions primarily affect reactor coolant parameters due to the resulting excessive heat removal from the primary- system. Since the feedwater system f lowrates have not been altered; the replacement of the steam generators has no effect on the results for this t~ansient and the consequences will be no more adverse than those determined in analyses for the FSAR or the power uprating submittal. 6.1.2.2 Excessive Load Increase The excess load transient may be initiated by an inadvertent opening of the turbine control valves, atmospheric steam dump valves, and/or steam bypass valve. The ensuing transient causes a high power level trip to protect the 6-la MARCH 1979 REV. 1
PALISADES PLANT SGRR reactor core. At hot standby conditions, there may be an excessive reduction in the steam generator water inventory. However, the time required to boil the steam generators dry is in excess of that time predicted to empty the steam generators during a loss of feedwater flow transient. Since there are no.changes in the valves identified above as the potential initiating mechanisms, and since an excess load transient is much less severe than a steam line break, the consequences of this transient are no more adverse with the replacement steam generators than those results reported for previous analyses. In addition, the results of these analyses are bounded by those of the main steam line* break and loss of feedwater flow events. 6.1.2.3 Loss of Load A loss of load transient leads to a rapid (and large) reduction in the power demand while the reactor is operating at full power. There is a corresponding reduction in the rate of heat removal from the primary coolant system. This leads to elevated pressurizer and steam generator pressures that cause the pressurizer and steam generator safety valves to open to minimize the peak primary and secondary pressures. Additional protection is provided by the high pressurizer pressure trip. As a result of the actuation of the reactor trip and the opening of the valves, the peak primary and secondary system pressures are no more adverse with the replacement steam generators than they were when the analyses were performed using the characteristics of the original steam generators. 6~1.2.4 Loss of Feedwater Flow A complete loss of feedwater flow may be initiated by a rupture of the feedwater crossover line downstream of the main feedwater pumps or a condensate pump failure, which would cause low suction pressure on both feedwater pumps. The primary consequence of this accident is the reduction in, and eventual loss of, the primary coolant system heat sink. An analysis of this transient assuming the installation of the replacement steam generators shows that the consequences of this accident are no more adverse than those reported for analyses using the original steam generator characteristics. In fact, the increase in secondary water inventory for the replacement steam generators increases the predicted time to empty the steam generat~rs. 6-lb MARCH 1979 REV. 1
PALISADES PLANT SGRR 6.1.2.5 Steam Line Break A rupture in a main steam line would increase the rate of heat extractions of the steam generators and cause a rapid temperature reduction in the primary coolant. The fastest blowdown and the most rapid reactivity addition are associated with a break located at a steam generator nozzle. The replacement steam generators were analyzed for the full power Main Steam Line Break (MSLB) event using the NSSS simulation code version which is consistent with the analytical methods use for the corresponding FSAR analysis. Although the replacement steam generators have the increased water inventory (see Tables 2.1-1 and 2.1-2), the effect of including the steam nozzle flow restrictors is to decrease maximum power during the full power MSLB for the replacement steam generators. Therefore, the results of the full power MSLB event with the replacement st~am generators are no worse than the results reported in the FSAR except for a slight increase in the total blowdown flow. The effect of this slight increase of blowdown flow is discussed in Section 6.1.4. 6.1.2.6 Steam Generator Tube Rupture A steam generator tube rupture is a penetration of the barrier *between the primary coolant system and the main steam system. A double-ended (guillotine) rupture of a steam generator U-tube at the tubesheet is postulated. Integrity of the barrier between the primary coolant system and main steam system is radiologically significant, since a leaking steam generator tube allows transport of primary coolant into the main steam system. Radioactivity contained i~ the reactor coolant can then mix with shell-side water in the affected steam generator and be expelled to the atmosphere. During normal plant operations, some of this radioactivity is transported through the turbine to the condenser, where the noncondensable radioactive materials are released via the condenser air ejectors. The steam generator tube rupture event has been analyzed because the tube inside diameter is larger for the replacement steam generators than for the original steam generators. For a double~ended tube rupture within the replacement steam generators, the larger break area could result in a higher primary-to-secondary leak rate. The re-analysis confirms that the fluid leak rate with the replacement steam generators is higher, during the initial stages of this transient. However, the escalated decrease 6-1 c MARCH 1979 REV. 1
PALISADES PLANT SGRR in the primary coolant inventory leads to an earlier reactor trip on low-pressurizer pressure and an earlier emptying of the pressurizer for this transient. Because during this transient, the reactor remains at power for a considerable period of time, a noticeable reduction in the time to trip the reactor causes a reduction in the total primary coolant activity transferred to the secondary side of the steam generators. Thus, a reduction in time to trip the reactor also reduces the total curie content transferred from the secondary side of th'e steam generators to atmosphere via the atmospheric dump valves or steam generator safety valves. The overall impact is that the radiological releases from the steam generator tube rupture event are no worse than the values reported in the corresponding FSAR analysis. 6.1.3 LOSS-OF-COOLANT ACCIDENT EVALUATION A major primary coolant system pipe break would result in a rapid depressurization of the primary coolant system and subsequently in reactor trip and safety injection system actuation on either low pressurizer pressure or high containment pressure. The reactor trip and safety injection systems serve to mitigate the consequences of the event in the following ways:
- a. Reactor trip and borated water injection, in addition to void formation as a r~sult of the depressurization, cause a rapid reduction in core power to the fission product decay heat level.
- b. Water injected by the safety injection system provides for core cooling and prevents excessive fuel and clad temperatures.
Safety injection system water is supplemented by the injection of borated water from the safety injection bottles. The safety injection tanks passively actuate when the primary coolant system pressure drops below 200 psia (plus the elevation head in the injection lines and bottles). The safety injection tanks affect a rapid refilling of the reactor vessel due to large capacity and, hence, strictly limit the period of time during which the reactor core remains uncovered. The emergency core cooling system (i.e., the safety injection system in combination with the safety injection tanks) is designed so that the reactor can be safely shut down and the essential heat transfer geometry of the core 6-ld MARCH 1979 REV. 1
P~LISADES PLANT SGRR preserved following the LOCA. More specifically, when the emergency core cooling syst:em (ECCS) is degraded by the most severe active single failure, it is designed to meet the ECCS acceptance criteria as stated in Reference 4. An evaluation was performed to determine the effects of the replacement steam generators on ECCS performance and is summarized below. The most recent loss-of-coolant accident analysis submitted for the Palisades Plant was used as the reference analysis in evaluating these effects. The most recent analysis, as documented in Reference 5, used the currently approved Exxon Nuclear Company WREM-II PWR Evaluation-Model. For this evaluation, sensitivity studies were conducted to determine' the effect of changes in significant primary system operating parameters and steam generator characteristics on peak cladding temperature (PCT) for the most limiting large break LOeA as determined by the reference analysis. The methods used for these studies are identical to those used in the reference analysis. The sensitivity of .PCT to s~eam generator tube plugging, primary system pressure, arid core inlet temperature were evaluated. The results of .this study are summarized below. Change in Change in Parameter Parameter PCT Plugged tubes + 850 *tubes +25F Core inlet + 8.5F -18F temperature Primary system + 90 psi +54F pre'.3sure 1 S:as ed on these sensitivity studies and noting that the reference analysis assumed a total of 4,175 plugged tubes, it can be deduced that replacement of the steam generators without changing plant operating conditions should result in a net reduction in PCT with respect to the reference case of about 125F. This improvement is primarily due to reduced steam generator flow resistance during the reflood phase of the transient and, hence, higher core reflooding rates. However, because of higher expected primary system flowrates and higher expected secondary steam pressures with the new steam genera tors as compared to the referenced analysis, it is expected that the Palisades Plant will be able to operate 6-le MARCH 1979 REV. 1
.r PALISADES PLANT SGRR at a .. slightly higher core inlet temperature (+SF) and a slightly higher primary system pressure (+40 psi) as compared to the core inlet temperature and pressure at which the reference analysis was performed. The incr.ease in core inlet temperature should result in a slight reduction in PCT (-lOF), whereas the increase in pressure should result in a slight incr~ase in PCT (+2SF). Taking all of these changes into account, the replacement steam generators should result in a net improvement in PCT for the limiting lar.ge break LOCA of about llOF. Change in Change in Parameter Parameter PCT Plugged tubes -417S tubes -12SF Core inlet +SF -lOF temperature Primary system +40 psi +2SF Net -llOF Hence, it can be concluded that the replacement steam generators will have a benef 1cial effect on ECCS performance and that the ECCS acceptance criteria (1) will be met with the new steam generators installed in the Palisades Plant. 6.1.4 CONTAINMENT PRESSURE ANALYSIS The effects of the replacement steam generators upon the containment pressure response analysis have been evaluated by assessment of the mass/energy releases to containment during the main steam line break (MSLB) and the loss-of-coolant accident (LOCA). FSAR Section 14.18 and Answer 14.11 of Amendment 14 (FSAR) indicated that for the original steam generators the MSLB at full load would be more severe from a containment pressure point of view than either the LOCA or the MSLB at no load. The LOCA mass/energy release for the replacement steam generators at full power (2S30 Mwt) was compared to the LOCA mass/energy for the original steam generators. It was concluded based on this comparative evaluation that the peak containment pressure following a LOCA would be slightly less than that predicted, in the FSAR (Sl.O psig). 6-lf MARCH 1979 REV. 1 J
.\. .r PALISADES PLANT SGRR The full power MSLB mass/energy release for the replacement
.steam generators wai analyzed using analytical methods comparable to those used in the preparation of the FSAR, except that credit was taken for the steam nozzle flow restrictors in the replacement steam generators. To obtain full benefit of the flow restrictors, the analysis included a Main Steam Isolation System (MSIS) actuation on high containment pressure (5.75 psig) as described in Section 3.6. The containment response to a full power MSLB was analyzed using the version of the COPATTA computer program which is described in FSAR Section 14.18.1.
Containment initial conditions, engineered safeguard equipment actuation times and containment heat sink data used for this analysis were identical to those presented in FSAR Section 14.18.1. Although the mass/energy data were developed by conservatively assuming the availability of off-site power, the containment response analysis conservatively assumed the loss of off-site power. Consequently, the single active failure assumed for the containment response analysis was a diesel-generator failure. This postulated active failure minimizes the engineered safeguard equipment available during the accident and maximizes containment pressure. A peak containment building pressure of 47.6 psig was calculated for the full load MSLB with the replacement steam generators; this value is less than that predicted in the FSAR (51.8 psig). Since the zero power inventory for the replaGement steam generators is slightly less than that in the original steam generators, the peak containment pressure for the no-load MSLB would be less for the replacement steam generators than that predicted in the FSAR. Based on the foregoing, it is concluded that the peak containment pres.sure following either a MSLB or a LOCA would be no more severe for the replacement steam generators than for th~ original ste~m genetators. 6.1.5 FSAR EVALUATION CONCLUSIONS The conclusions based on the safety evaluation of the design basis events for the replacement steam generators are as follows:
- a. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased.
6-lg MARCH 1979 REV. 1
( ...,.. { PALISADES PLANT SGRR
- b. The possibility for an accident or malfunction of a different type than any of those evaluated previously in the safety analysis report is not created.
- c. The margin of safety as defined in the basis for any present technical specification is not reduced.
- d. The Palisades Plant, equipped with the replacement steam generators, may be safely operated without presenting any undue hazard to the health and safety of the public.
MARCH 1979 REV. 1
PALISADES PLANT SGRR 6.2 CONSTRUCTION-RELATED EVALUATIONS 6.2.1 HANDLING OF HEAVY OBJECTS Most of the equipment and construction-related tools required for the steam generator repair program will be brought into the containment via the construction and supplemental access routes as shown in Figures 4.3-1 and 4.3-2. Some equipment or tools which cannot be practically or conveniently transported through the access route described above will be transported through the fuel building floor hatch at el 649'-0". This equipment will be transported over the tilt pit areas away from the spent fuel pool. Movement of equipment or tools in the spent fuel area will be governed by the Palisades Plant Technical Specifications in effect at the time the repair is performed. Construction-related incidents concerned with the handling of the steam generators are discussed in Section 4.1.4. 6.2.2 OFFSITE RADIOACTIVE RELEASES AND DOSE ASSESSMENT Radioactive airborne and liquid offsite releases have been evaluated for the repair effort using conservative bounding parameters and assumptions. 6.2.2.1 Airborne Releases Radioactive airborne effluent releases to the environment resulting from the repair effort have been estimated using the following assumptions and parameters:
- a. Airborne releases are assumed to occur during the cutting operations.
- b. The reactor coolant pipes and the steam generators are expected to be contaminated primarily by deposited corrosion products. Typical corrosion product activities expected on the primary side surfaces of the steam generators are given in Table 6.2-1. These activities have been increased by a factor of 3 for approximately 5 effective full power years (Reference 3) of additional reactor operation.
- c. It is conservatively assumed that all the activity present in the vicinity of each cut will become 6-2
\
PALISADES PLANT SGRR airborne and be available for release to the environment.
- d. One hundred forty days of radioisotope decay were assumed before cutting operations, on the basis of the earliest reasonable time as dictated by the repair effort. No credit was taken for radioisotope decay during cutting operations.
- e. All primary coolant piping cuts are assumed to be made in specially-designed contamination control enclosures which will provide high efficiency filtration. The enclosures are assumed to be 90% efficient for capturing particulates. An additional 99% efficiency is assumed for the stack filter (HEPA) through which all plant ventilation flows. Further reductions in airborne radioactivity will occur through use of the two internal recirculation filters which are discussed in Section 4.3.3; however, no additional environmental release credit is assumed for these filters since their primary purpose is to reduce occupational doses and minimize personal respiratory protection devices. Radioactive airborne effluent releases to the environment based on the above assumptions are approximately 5.9 x 10-5 Ci. Details of the airborne effluent release by isotopes are given in Table 6.2-1.
6.2.2.2 Comparison with Observed Gaseous Releases and Estimated Doses During Normal Operation The estimated releases of radioactive airborne effluents during the repair effort are found to be much smaller than the observed effluent releases at the Palisades Plant during 1977. Observed airborne effluent releases during 1977 are compared with estimated releases during the repair effort in Table 6.2-2. The estimated critical organ dose for the repair program was found to be less than 1.0% of the calculated critical organ dose for 1977. 6.2.2.3 Liquid Effluent Releases Liquid effluent releases resulting from the repair effort were estimated using analogous data from previous refueling and steam generator inspection outages. The total radioactive effluent release estimated for the repair activity is shown in Table 6.2-4. The total includes 6-3
PALISADES PLANT SGRR laundry waste effluents expected during repair activities and the small amount of liquid waste generated as a result of local decontamination. The estimated specific activities of laundry wastewater are shown in Table 6.2-3. A description of the laundry waste treatment system is included in Section 4.3.6.3. A comparison of the average release for the repair effort is shown on Table 6.2-5. 6.2.2.4 Comparison with Observed Liquid Releases during Normal Operation Observed liquid effluent releases during 1977 are compared with the estimated releases for the repair effort in Table 6.2-5. The total body and significant organ doses for the repair effort were roughly equivalent to the dose from liquid effluents during the year 1977. 6.2.2.5 Conclusion The combined effect to the offsite dose from gaseous and liquid releases is less than that expected for a year of normal operation. The estimated dose to an individual in an unrestricted area from all pathways of exposure is much less than the limits specified in 10 CFR Part 50 (Appendix I). 6.3 FIRE PROTECTION A fire prevention and protection program is currently in effect at the Palisades Plant. A complete fire analysis has been conducted for the plant and a report has been submitted to the Nuclear Regulatory Commission. Work associated with the steam generator repair activities will be performed in containment and the yard area east and southeast of containment. 6.3.1 EXISTING FIRE PROTECTION All permanent fire protection systems and equipment that can be maintained in its present mode of operation and position without being disrupted, removed, or relocated by the construction will be maintained as is throughout the construction period. The present fire emergency operations plan will continue to be in effect during construction with inclusions pertaining to fire protection around the actual construction site. 6-4
PALISADES PLANT SGRR The Consumers' Property Protection Department will be notified before altering the fire protection system, equipment, or plan. During the construction progress, inspections will be made by members of the Consumers' Property Protection Department and N.M.L. fire protection engineers used by Consumers to check on compliance with established permanent and temporary fire protection systems, equipment, and policies. 6.3.2 FIRE PROTECTION DURING THE REPAIR PROGRAM This section contains policies and procedures pertaining to fire protection that will be in operation and enforced during the repair program. 6.3.2.1 Portable Fire Extinguishers Portable fire extinguishers will be located around the work areas to support construction activities. The types of fire extinguishers to be used are pressurized water, carbon dioxide, and dry chemical. Each fire extinguisher location would be determined before the construction process so that each fire extinguisher would be in the most accessible location and conspicuous to view for locating by workers. The type and size of fire extinguisher selected for each location will be dependent on the class (or classes) of fire that could be expected to occur at that location or in that area. Fire extinguishers will be inspected and maintained according to N.F.P.A. Pamphlet 10. Any fire extinguisher discharged or damaged will immediately be removed from service to be recharged or repaired. During its absence, a fully charged unit of identical size and type will be placed in the location. 6.3.2.2 Fire Hose A 2-1/2-inch gate valve will be attached to the fire hydrant at Hose House 5. During construction, 100 feet of 2-1/2-inch hose will be kept ready for use at the hydrant at the area of construction near containment. Two 100-foot lengths of 1-1/2-inch hose will be available for attaching to a gated wye (two 1-1/2-inch male x one 2-1/2-inch female) 6-5
PALISADES PLANT SGRR attached to the 2-1/2-inch hose at the construction site. Each length of 100-foot 1-1/2-inch hose will have attached a 1-1/2-inch water fog nozzle. These 1-1/2-inch hoses will be available to be used on either side of the construction site. Fire hose is also located at Hose Houses 4, 6, and 7 near the area of construction. Two internal 1-1/2-inch fire-fighting standpipe hoses are located near the containment lock if a hose is needed inside of containment. New hoses for fire-fighting and fire protection inside of containment on the existing house service water system will be installed in the future. 6.3.2.3 Fire Emergency Reporting A fire emergency reporting phone will be located in the immediate vicinity of the working area on the construction site. This communications system will be tied into the plant system for immediate notification of the plant fire brigade. All construction workers on site will be instructed in fire response/reporting procedures. Portable bullhorns will be available at or near the work site. 6.3.2.4 Combustible Materials A minimum amount of combustible materials will be used at the work area. Combustible materials not in use will be stored away from the work areas. This separation will provide a natural fire break. All construction areas will be kept as free and clear of rubbish and combustible materials as possible. Metal containers will be strategically located around the construction areas for disposal of materials. 6.3.2.5 Welding and cutting All combustible materials to the extent practical at welding and cutting locations will be relocated before welding and cutting starts. Combustible materials that cannot be moved will be covered with noncombustible tarps, if possible, before welding and cutting and will remain covered until all welding and cutting operations are completed. Frequent tours by supervisors to inspect areas where welding and cutting is being done will help minimize the number of welding and cutting fires that could occur. 6-6
PALISADES PLANT SGRR 6.3.2.6 Flammable and Combustible Liquids All flammable and combustible liquids used on the construction site will be handled, transferred, and stored in accordance with established Consumers' fire protection practices. Transfer and storage of flammable and combustible liquids will be in areas away from ignition sources. 6.3.2.7 Smoking Smoking will only be permitted in designated areas. 6.3.2.8 Electrical Wiring All wiring used to provide electricity to construction equipment or regular lighting at the construction site will be of UL approved 3-wire type designed to be used in outdoor locations and able to handle the electrical load placed on it. The actual wire size and type will be selected by the contractor doing the work. Wiring will be installed in a professional manner and located to prevent possible damage. All wiring will be properly fused. 6.3.2.9 Work Site Enclosure Any work site enclosure erected to provide a work area protected from the elements will be constructed of fire-retardant material whenever possible. Exit passages and/or doors will be provided for easy exit in case of emergency. 6.3.2.10 Emergency Lighting Emergency lighting will be provided at the work site for use in case of emergency or accident. 6.3.2.11 Access Into All Work Site Areas An access road will be maintained into and through the construction area up to containment for access by vehicles (fire apparatus, ambulance, etc) and personnel. 6-7
PALISADES PLANT SGRR 6.3.2.12 Fire Brigade The plant fire brigade will be available at all times in case of emergency. Construction workers will also be available for fire brigade operations, if needed. 6.
3.3 CONCLUSION
The fire protection measures presently in effect at the Palisades Plant, augmented by the special temporary measures discussed above, provide +easonable assurance that potential fires can be readily detected and extinguished if they occur without causing significant damage to the facility. 6-8
PALISADES PLANT SGRR TABLE 6.2-1 ESTIMATES OF AIRBORNE RELEASES TO ENVIRONMENT DURING STEAM GENERATOR REPAIR EFFORT ( 1) (2) Activity of Corrosion Products Total Release Isotope at 140 days (µCi/in2) (µCi) Cr-51 0.589 0.378 Mn-54 0.730 0.468 Co-57 0.312 0.200 Co-58 85.5 54.8 Fe-59 0.535 0.343 Co-60 4.27 2.74 Nb-95 0.040 0.026 Zr-95 0.085 0.054 TOTAL 92.2 59.1 NOTES: ( 1) These are the activities presented in Table 4.4-1 converted to µCi/in2. The activities were increased by a factor of 3 to account for the expected activity build-up. (2) The following technique was used to estimate the activity from each isotope released during cutting operations: Airborne Enclosure and Activity = Area x Activity of x Number x Stack Filter Near cut of Cut Corrosion of Cuts Penetration (µCi) (.5 in 2 ) (µ Ci/in2) (No. ) ( .1) ( .01) The total number of cuts on primary coolant piping which were assumed to total 12 where: 4 cuts (42-inch ID pipe) 8 cuts (30-inch ID pipe)
PALISADES PLANT SGRR TABLE 6.2-2 COMPARISON OF GASEOUS EFFLUENT RELEASES Estimated Release Average During SG Isotope 1977 Release (Ci) Repair Effort (Ci) Noble Gases 59.89 Negligible Iodines 1. 51 x 10*2 Negligible
* (1I Particulates 1.1 x 10"3 .059 x io*3 Tritium 2.21 Negligible NOTES:
111 Approximately 29. 0 and 26% of the total particulate release during the year 1977 are Co-58 and Co-60, respectively . . ..\
PALISADES PLANT SGRR TABLE 6.2-3 ESTIMATED SPECIFIC ACTIVITY OF LAUNDRY WASTEWA'IER ISO'.OOPE §.fECIFIC ACTIVITY Ci/cc< 1 > Co-57 8.96 x 10-7 Cs-134 4.64 x 10-s Cs-13 7 1.03 x 10-4 Co-58 3.85 x 10-4 Mn-54 2.63 x 10-s Co-60 7.22 x 10-s Fe-59 3.27 x 10-6 Zn-65 4.09 x 10-7 Zr-95 4. 15 x 10-6 Nb-95 7. 2 x 10-6 sr-90 3.66 x 10-s Ni-63 2.65 x 10-s NOTE:
<1> Time averaged specific activity during a period of 365 days.
PALISADES PLANT SGRR TABLE 6.2-4 ESTIMAT~D RADIOACTIVE LIQUID EFFLUENT RELEASED DURING THE STEAM GENERATOR REPAIR ISOTOPE RELEASE Ci Co-57 6.85 x 10 4 Cs-134 3.54 x 10-2 Cs-137 7.85 x 10-2 Co-58 2.95 x 10- 1 Mn-54 2.01 x 10-2 co-60 5.5 x io- 2 Fe-59 2.5 x 10 -3 Zn-65 3.12 x 10 -4 2;r-95 3.17 x 10 -3 Nb-95 5.5 x 10 -3
*sr-90 2.79 x 10 -2 Ni-63 2.02 x 10 -2 TOTAL .544 H-3 1.91
PALISADES PLANT SGRR TABLE 6.2-5 COMPARISON OF RADIOACTIVE LIQUID EFFLUENT RELEASES ESTIMATED RELEASE DURING ISOTOPE AVERAGE .1977 RELEASE (Ci) SG REPAIR EFFORT (Ci) Fission .093 .544 and acti-vation products Tritium 55.8 1.91 Total 55.9 2.45
PALISADES PLANT SGRR 7.0 ENVIRONMENTAL ASPECTS OF THE REPAIR PROGRAM 7.1 GENERAL The following sections present information and the assessment of environmental impact of the proposed steam generator repair program. The estimated environmental impact of the repair activity and disposal of the removed steam generators is expected to be negligible and temporary. The proposed activity will cause little additional environmental impact over that of normal plant operation. Construction activities, particularly the barge slip preparation, will be carried out in conformance with local, state, and federal regulations. When the facility is returned to service after the repair, water use, occupational exposures, and radiological releases are expected to be less than those associated with current facility operation. It may be necessary to store the old steam generators onsite in an engineered storage facility that will provide shielding for the direct radiation. The steam generators will be sealed to contain any airborne radionuclides. The steam generators may be shipped offsite to a federally licensed storage facility, if available. If shipped offsite, all local, state, and federal regulations pertaining to the shipment of radiological materials will be followed. 7.2 RESOURCES COMMITTED 7.2.1 NONRECYCLABLE BUILDING MATERIALS The steam generator repair program at the Palisades Plant will require the commitment of various irretrievable building materials. The quantitative estimates for the nonrecyclable building materials are as follows: Concrete 1,000,000 pounds Structural steel 50,000 pounds Alloy or stainless steel 1,594,000 pounds Tendons (53) 150,000 pounds Cable (copper) 400 pounds Inconel 280,000 pounds Pipe 5,000 pounds Wood 15,000 board feet 7-1
PALISADES PLANT SGRR 7.2.2 LAND RESOURCES The steam generator repair program will have minimal impact on the existing site in terms of land use. The construction of facilities (see Section 4.1.1.2) to store the steam generators will require some excavation, leveling, and foundation work. If necessary, a building will be located near the temporary barge slip (see Figure 4.1-1). The extent of the disturbance will be temporary, negligible, and of minor impact. This area had been previously excavated during plant construction. 7.2.3 WATER RESOURCES During the repair effort, construction water will be supplied from existing Palisades Plant water sources. No requirements for commitments of new water sources have been identified for the repair effort. Since water consumption during the extended shutdown is expected to be less than during plant operation, water consumption during the repair effort will result in a reduction in plant water usage. A temporary barge slip will be constructed just north of the existing Palisades Plant (see Figure 4.1-2) to receive and offload two replacement steam generators. The temporary barge slip will be approximately 110 feet long, 50 feet wide, and 18 feet deep. The total dredged quantities are anticipated to be approximately 12,000 cubic yards based upon a lake elevation of 579 1 -0 11 (USGS). The dredged material will be disposed of according to Guidelines for the Pollution Classification of Great Lakes Harbor Sediments, U.S. EPA Region V. The dredging will be conducted according to specifications of the U.S. Corps of Engineers permit. Steel sheet piles or sunken ship hulls may be used to provide wave and scour protection for the barge. Depending on lake conditions, the barge slip may have to be periodically redredged to specifications if suspended materials (sand) are deposited there before delivery of the generators. The dredging activity will temporarily increase turbidity in the site vicinity and remove a small number of benthic organisms with the spoils. The piling or sunken hulls, if used, will provide substrate suitable for attachment of periphyton and filamentous algae, but large growths are not expected. The impacts of the dredging will only have a temporary impact upon the aquatic biota in the immediate plant vicinity. Removal of the temporary barge slip 7-2 e
PALISADES PLANT SGRR facility will result only in some temporary increase in turbidity along the shore line. 7.3 WASTEWATER 7.3.l SANITARY FACILITIES Since the repair activities will take place in locations near which permanent sanitary facilities are not readily accessible (e.g., the containment and laydown area), portable units will be used. There will be no modification to existing sanitary facilities as a result of the repair activity. I 7.3.2 LAUNDERING OPERATIONS ~ Laundry wastewater generated during the repair activities will be produced in the existing facility.- A description of the laundry waste processing scheme is included in Section 4.3.6.3. 7.4 CONSTRUCTION Construction activities at the time of the repair effort will satisfy applicable laws that are in force at that time. These activities will have a negligible effect on noise levels, dust, or smoke. 7.4.1 NOISE Values and calculation methods described in Reference 6 have been used to examine the maximum sound pressure level (SPL) expected at the site boundary (0.4 miles) because of construction noise. The maximum source SPL is expected to be less than 94 dBA at 50 feet, which will attenuate to less than 62 dBA at the site boundary. The site boundary maximum expected SPL is within the acceptable limits for permissible outdoor noise levels fo~ sleeping with open windows. Moreover, the site is located in a low population area. Oh the basis of these facts, it is concluded that the additional noise resulting from the repair program for the steam generators is expected to have negligible impact on the local area. To protect personnel located-on the site, Occupational Safety and Health Administration standards (state and federal) will be followed. 7-3
PALISADES PLANT SGRR 7.4.2 DUST Dust, if any, will be abated by periodically spraying with water or other dust control measures. The frequency of spraying and the quantity of water sprayed will be determined by visual inspection of the areas and will vary with the weather conditions. 7.4.3 OPEN BURNING Open burning is not anticipated during the steam generator repair effort. However, should the necessity arise, applicable county and state regulations for open burning will be followed. 7.5 RADIOLOGICAL MONITORING Radioactive effluent release points during steam generator repair activities will be the same as during normal plant operations; therefore, the plant radioactive process monitors will not be affected. Since releases of radioactive effluents during the repair program will be similar to those from the operating plant, and their potential exposure pathway will be the same as for existing plant operations, these effluents will be monitored in accordance with the existing environmental monitoring program at Palisades Plant. 7.6 RETURN TO OPERATION 7.6.l WATER USE Water c.onsumption during post repair plant outages is expected to be appreciably less than is currently required as a result of repairs to degraded tubing. Periodic plant shutdown for steam generator inspection consumes large quantities of pure water. During shutdown, the steam generator water .level is controlled on the low side (between 20 to 40% of operating band) to aid in chemical layup. Approximately 10,000 gallons of water are required to place the two steam generators into wet chemical layup. If a steam generator requires draining (for tube plugging, tube sleeving, or tube removal) during the inspection, i t would require an additional 24,000 gallons to refill. Other plant requirements for pure water include 70,000 to 85,000 7-4
PALISADES PLANT SGRR gallons for heater train, condensate polishers, and hotwell. This is the amount necessary to refill the systems if maintenance had been performed that required draining the systems. Approximately 75,000 gallons would be required for primary system dilutions to return to power. Depending on various chemical parameters, as much as 50% of this water could be recovered through the plant recovery systems, sucJ;i as clean radwaste system, boric acid recycle system, and steam generator blowdown recovery system. Following replacement of the steam generators, i t is expected that forced outages associated with steam generator tube plugging and/or tube sleeving will be essentially eliminated; however, it is not anticipated that the water consumption associated with the current inspection program will be significantly reduced because of the continuing requirement to inspect (eddy current test) the steam generator tubing at regular intervals. 7.6.2 OPERATIONAL EXPOSURES Section 4.3.7 discusses the future reduction in man-rem exposure as a consequence of the repair program. A potential savings of 250 man-rem/yr may be realized because of the expected elimination of the necessity to plug tubes in the repaired steam generators and the decrease in the number of inspections required (Regulatory Guide 1.83). 7 .,6. 3 RADIOLOGICAL RELEASES Although the Palisades Plant has experienced only one pri~ary to secondary leak from tube failure (1974), the repair of the steam generators should reduce the probability of future. secondary releases as a consequence of the same tube failure mechanism. 7-5
PALISADES PLANT SGRR 8.0 EVALUATION OF ALTERNATIVES
8.1 INTRODUCTION
In view of the uncertainty which existed with regard to the continued full-power operation of the Palisades Plant as a consequence of corrosion-related steam generator tube degradation, Consumers has investigated a number of potential repair concepts. Those alternatives that have been found feasible, as well as practical, have been classified as "minor" or 11 major. 11 Minor repair alternatives allow retention of the existing generators, whereas major repair alternatives would involve the replacement of substantial portions of the original units. In the former category, sleeving appears to offer the most promise, but the "denting" phenomenon has the potential for severely limiting its application. Chemical cleaning has also been extensively investigated as a method to restore the existing units, but its effectiveness and the absence of adverse side effects has not been satisfactorily demonstrated. Continued tube pluggi~g is obviously the least desirable alternative because of the eventual impact the plugging will have on plant output. - In the latter category, the field of practical repair alternatives available to restore the original steam generating capability of the Palisades Plant appears limited to the following:
- a. Retubing the original units within the containment
- b. Repair, limited to replacement of the evaporator sections only
- c. Repair utilizing complete replacement units The following discussions of these alternatives are based on feasibility studies which have been performed for Consumers during the period 1974 through 1977. The repair options have been evaluated and compared on the basis of the following factors:
- a. Plant outage requirements
- b. Direct cost of repair
- c. Radiological aspects 8-1
PALISADES PLANT SGRR
- d. Equipment procurement lead times
- e. Special tooling development requirements
- f. Plant structural consideration
- g. Equipment design improvement possibilities
- h. Site preparation requirements
- i. Radwaste disposal considerations On the basis of comparisons of the three alternatives, complete replacement of the existing steam generators with compatible spares appears to offer the optimum repair solution if a program of major repair becomes necessary at the Palisades Plant.
8.2 CONTINUED TUBE PLUGGING AND PLANT DERATE The effect of future tube plugging on plant power output with the current steam generators has been evaluated. No derating is projected before 500 additional tubes (a total of 4,175) are plugged. In the absence of mitigating measures, it is estimated that the plant would be derated from its currently rated level of 2530 MWt by approximately 14% if a total of 7,175 tubes were plugged, and by 30% if a total of 10,000 tubes were plugged. These estimates are based upon the most recent safety an~lyses for 2530 MWt generation and on an empirical relationship between core flow and number of tubes plugged, which was derived on the basis of actual plant measurements. 8.3 IN-PLACE TUBE SLEEVING Sleeving is the insertion of a thin-walled tube insert that is positioned in the vertical section of the tube, spanning a degraded area, that is hydraulically expanded in place. The feasibility of local repair, via sieeving, of steam generator tubes that have suffered external wall thinning has been previously demonstrated at the Palisades Plant. such technique will restore the structural integrity of the tube and yet avoid reducing heat transfer surface. The use of sleeving techniques for complete repair of the Palisades Plant steam generators is not, however, considered feasible at the present time. Sleeve insertion and inspection is a time-consuming process requiring extensive baseline and inservice inspections. Defects at certain positions cannot 8-2
PALISADES PLANT SGRR be repaired by sleeving because of the inability to install sleeves at those locations (bends, etc). The sleeving process, as presently applied, could not be used to repair areas which have experienced significant denting. Although those conditions do not currently exist in the Palisades Plant steam generators, and their occurrence in the future is not expected, these restrictions create a degree of uncertainty with respect to future application of sleeving. 8.4 IN-PLACE TUBE REPLACEMENT Because of the limited size of the Palisades containment building equipment hatch, retubing is the only major option that would not entail the construction of a temporary opening in the containment building in order to accomplish a major repair. The hatch diameter is 12 feet, whereas the steam generator, even at its narrowest, is more than 14 feet in width. The general concept for in-place retubing would be to remove the secondary head, abrasively cut and remove the old tubes and structural material, remove the tube stubs from the tubesheet, install new tubes and support structures, replace steam drying and separating equipment, and replace the secondary head. All equipment and material could be transferred to and from the containment through the existing equipment hatch. A new tube bundle shroud and new eggcrate supports would be shop-fabricated in sections and aligned and assembled within the containment before installation in the steam generator shells. The conceptual schedule for the retubing alternative is estimated at approximately 3 years, which does not allow for any decontamination beyond local measures. Activity levels within the containment and the primary coolant system will dictate the extent to which decontamination is required. Consumers has explored aspects of decontaminating the primary system, and the results of those studies suggest that full decontamination could add approximately 1 year to the retubing program, although for purposes of comparison with other alternatives, decontamination has been limited to localized efforts only. The success of any program of retubing is crucially dependent upon the prior development of special tooling that would be necessary to remove the tube stubs from the tube sheet and prepare the holes for the welding of the new tubes. Despite prior local decontamination and the use of 8-3
PALISADES PLANT SGRR special tooling, it is, nevertheless, expected that an appreciable number of manhours will be expended in the primary head, near the tubesheet, contributing significantly to the ,man-rem dose estimated for retubing. Aside from the lengthy outage required for such a program, an additional negative aspect of retubing is the limited ability to modify the existing steam generators in order to improve performance and maintainability. The estimated direct cost for a retubing program, assuming minimal decontamination is required, is approximately $75 million (see Table 8.4-1). 8.5 REPLACEMENT WITH COMPLETE.UNITS Because of the negative aspects associated with a program of retubing, the feasibilities of various steam generator replacement schemes were also investigated. These studies concluded that complete replacement could be accomplished in approximately two-thirds of the time required for retubing, and that even though a temporary construction opening in the containment was necessary, the removal and replacement of the steam generators could be performed without disturbing any important structural elements within the containment. Three alternative schemes involving construction openings were evaluated. Two programs were based upon construction openings in the side of the containment but at different locations. The third scheme was based upon an opening in the containment dome, through which the units could be removed and replaced. The preferred repair scheme, and the one for which NRC approval is being sought, is based upon removal and replacement through a temporary side opening located above the 649-foot level and centered horizontally about the containment radius at approximately 118 degrees (north is O degrees). The repair program based on that alternative is discussed at length in Section 4.0. Briefly, i t is planned that the new units would be shipped by barge and unloaded at a new temporary barge slip at the plant site. The containment would be partially detensioned and the construction opening cut. The old units would be cut free, rigged vertically from their cavities, lowered to a horizontal position, and removed from the containment. The installation would then follow in reverse order. Decontamination would be limited to local efforts in the 8-4
PALISADES PLANT SGRR area of major pipe cuts, supplemented with the wide use of temporary shielding. The estimated man-rem dose for a repair program based on such a replacement scheme would be in the range of 1,200 to 5,000 man-rems. The length of the repair outage associated with complete replacement is estimated at approximately 2 years, based upon the currently anticipated scope of work. The related direct cost in 1983 dollars is estimated at approximately
$75 million (see Table 8.4-1).
Some important direct benefits of a repair program based upon replacement relate to those steam generator design improvements that could be readily accommodated in new units but which would prove to be extremely difficult to incorporate into the existing generators under a program of retubing. Included in that category are:
- a. High capacity blowdown capability
- b. Recirculation/chemical cleaning capabilities
- c. Visual inspection provisions for critical areas
- d. Flow limiting nozzle design e e. Enlarged manways
- f. Primary head drains All of these features have been included in the design of the replacement units for the Palisades Plant and are discussed more fully in Section 2.0.
Since the procurement lead time for replacement steam generators is approximately 4 years, the assessment of the future operating performance of existing units is of crucial importance in the decision between alternatives and the decision as to when to proceed with planning for a repair program based on replacement. A parametric analysis of the various steam generator failure mechanisms was conducted, utilizing stochastic modeling techniques, to forecast future unit performance. As a result of that study, it was concluded that there is a significant probability that the existing steam generators will operate, without derating, until the time that replacement units are available. 8-5
PALISADES PLANT SGRR 8.6 REPLACEMENT OF STEAM GENERATOR EVAPORATOR SECTIONS The third major repair alternative considered for the Palisades Plant is a variation of the replacement concept discussed previously in Section 8.5. It is different in that the upper shell and head portion of each old unit would be reused. That variation would require a circumferential cut at some appropriate location on the shell in order to separate the evaporator section from the steam drum. The removal and installation procedure for the new tube bundle assemblies would closely follow that for the complete assemblies with the additional requirement of the completion weld on the transition cone. Other than the reduced scope of rigging related to the lower weight of the evaporator bundle, there is nothing to recommend this option over the complete replacement concept: the temporary containment opening is still required, the barge slip is still necessary, and the reduced cost of the hardware is more than offset by the additional time required to accomplish the completion weld on the cone. Furthermore, because of the interferences associated with the probable location of the completion weld, it appears that it might be necessary to perform all cutting, welding, and stress relieving on the transition cone in a separate facility outside containment. On that basis, an additional outage 0 period of 8 months would be required. 8.7 MAN-REM CONSIDERATIONS The detailed development of the man-rem dose estimate for steam generator repair utilizing complete units is presented in Section 4.3.7. The results of that assessment would suggest that a dose in the range of 1,200 to 5,000 man-rem could be expected as a consequence of replacing the steam generators. Although a comparable dose assessment of the evaporator replacement option was not performed, the projected dose is expected to be slightly greater because of the extended outage period. The man-rem dose associated with a retubing program was developed in a manner similar to that for replacement. In the absence of preliminary decontamination, and using worker residence times in contaminated areas comparable to those used in the man-rem assessment for replacement, a dose of 40,250 man-rems was estimated for retubing in-place. The major contributor to that estimate is the dose incurred while performing tube stub removal and welding activities at 8-6
\
PALISADES PLANT SGRR the tubesheet. Various combinations of decontamination and. mechanized welding programs could conceivably reduce that estimate by a significant amount, but considerable development is necessary in either case before they could be considered reliable options. Since the need for extensive steam generator inspection and tube plugging operations should be eliminated by the repair, yearly exposures presently incurred for these operations should be significantly reduced. It is estimated that at least 250 man-rem would be saved per year, or a total of 7,500 man-rem over the plant's remaining lifetime of approximately 30 years. The savings above the replacement dose of 1,200 to 5,000 man-rem equals 2,500 to 6,300 man-rem. Based on the biological savings of $1,000/man-rem (10 CFR 50, Appendix I), this* results in a savings to 'Society of $2,500,000 to $6,300,000. In-place decontamination before removal could possibly result in lowering the total man-rem incurred, but has been found impractical on a cost-benefit basis (see Section 9.2).
8.8 CONCLUSION
S As a result of Consumers' investigations and comparisons of alternatives, it is evident, from both an economic and a technical viewpoint, that complete replacement represents the optimum method of major steam generator repair for Palisades .Plant if repair becomes necessary. The key factor in the economic comparison of alternatives is the cost of replacement power associated with the plant outage required to perform the repair. In that category, replacement has the clear advantage. Similarly, from the standpoint of ALARA consideration, replacement is the preferred method among alternatives. On a technical basis, the installation of replacement steam generators would permit the inclusion of important steam generator design features that, in some instances, would be impossible to incorporate in the existing units in conjunction with a program of retubing. In addition, the success of a retubing program is dependent upon the development of special tooling, while the replacement option is not similarly constrained. 8-7
PALISADES PLANT SGRR The decision to replace, and the timing of the repair, will be based upon the conditions of the old units with respect to derating, the adequacy of system reserve margins, the future inspection requirements of the degraded generators, and other relevant factors. Consumers will continue to investigate sleeving and other qualified means of steam generator repair in order to extend the useful lives of the original units for as long as practical. A tabular comparison of the major repair alternatives is presented in Table 8.8-1. 8-8
PALISADES PLANT SGRR TABLE 8.4-1 STEAM GENERATOR REPAIR ALTERNATIVE COSTS (Costs in Millions of Dollars) Replacement Replacement Evaporator Retube Item Complete Uni ts* Sections In-Place Engineering and material 26.5 21.5 7.7 Construction 9.5 12.0 21.8 Licensing 5.0 5.0 2.5 Disposal 2.0 2.0 0.2 Consumers directs 1.0 1.0 0.8 Escalation 10.5* 12.8* 18.2* Subtotal - directs 54.5 54.3 51.2 Administrative and general 1.5 1.5 1.0 Insurance and taxes 1.0 1.0 1.0 Allowance for funds used during 11.5** 12.0** 9 .. 7** construction. (AFUDC) Subtotal - overheads 14.0 .14. 5 11. 7 Consumers' contingency and 6.5 11.2 12.1 rounding Total estimated project cost 75.0*** 80.0*** 75.0*** Notes:
*I.ncludes escalation for material and steam generator costs, as applicable.
**Includes AFUDC for material and steam generator payments, as applicable.
)
***Total estimated cost is for a commercial operation date of April 1, 1983.
PALISADES PLANT SGRR TABLE 8.8-1 COMPARISON OF MAJOR REPAIR ALTERNATIVES Replacement Replacement Re tube Complete Units Evaporator Sections In-Place
- 1. Length of repair outage (months) 24 30 36 (Local decontamination only)
- 2. Cost of replacement power ($xl0 6 ) 200 250 300
- 3. Direct cost of repair ($xl06) (1983) 75 80 75
- 4. Dose to workers (man-rem) 1,200-5,000 1,200-5,000 40,000
- 5. Equipment procurement lead times (yr) 4 3 Not critical
- 6. Special tooling requirements None None a) Tube stub removal b) Tube sheet weld prep c) Tube/tubesheet weld d) Tube expansion
- 7. Plant structural considerations Temporary containment Temporary containment None opening required opening required
- 8. Equipment design improvements a) SS egg crate tube a) Same a) Same which can be incorporated supports b) Flow distribution b) Same b) Same baffle c) Increased capacity c) Same blow down d) Inspection ports d) Same e) Primary head drains e) Same f) Larger primary f) Same manways g) Larger secondary manways h) Recirculation/chem-ical clean_ing pro-visions i) Steam nozzle flow restrictor
- 9. Site preparation requirements Temporary barge slip Same None
PALISADES PLANT SGRR 9.0 COST BENEFIT ANALYSIS FOR THE DECONTAMINATION, STORAGE, AND DISPOSAL OF THE OLD STEAM GENERATORS CONSIDERING ALARA
9.1 INTRODUCTION
The following are evaluated on a cost-benefit basis:
- a. In-place decontamination
- b. steam generator storage and disposal methods The cost of each method compared with benefits gained in the total man-rem reduction to workers is in accordance with the philosophy of reducing worker dose to levels which are ALARA.
9.2 STEAM GENERATOR IN-PLACE DECONTAMINATION A study has been performed by United Nuclear Industries (UNI) to evaluate the best method of decontamination of the Palisades steam generators. Its report (Reference 2) indicated that a combination acid/base flush with several rinses will lower fields inside the steam generators at the time of replacement to approximately 260 mrem/hr. Total costs of decontamination and waste disposal, including equipment, chemicals, and processing of solutions, is estimated at approximately $5 million. Decontamination time for one steam generator is optimistically estimated to be 17 weeks, with a total of 30 weeks required for both steam generators. A man-rem assessment was made for the repair effort considering in-place decontamination of the steam generators. The dose assessment was based on the radiation field information and tasks required for removal presented in Section 4.3.7.2. A dose rate reduction factor of 25 was obtained from analysis of the decontamination study as applied to fields in various areas within 6 feet of the steam generators. Assuming in-place decontamination before the initiation of the repair effort, it is estimated that the repair effort would require 200-300 man-rem, a savings of approximately 1,000-4,800 man-rem (see Table 4.3-2). The estimate does not consider man-rem incurred during the decontamination effort. Although the exposures from decontamination operations are difficult to quantify, i t is known that approximately 60 curies of radioactivity would be removed 9-1
PALISADES PLANT SGRR and require handling. Since a considerable volume of the solutions are not compatible with the installed radwaste processing system, much manual contact would be required for solidification and shipping of this waste. In addition, considerable work is required in high radiation fields to connect piping and modify recirculation paths to avoid material incompatibilities. Thus, the 1,000-4,800 man-rem saving in replacement exposures is negated by exposures associated with decontamination efforts. Considering that the estimated biological cost of a man-rem is $1,000 (10 CFR 50, Appendix I), the combined considerations of dose reduction, schedule penalty, and large capital cost associated with decontamination do not indicate a benefit for full-scale steam generator decontamination. Therefore, it is concluded that in-place decontamination of the steam generators is not practicable. 9.3 STEAM GENERATOR STORAGE AND DISPOSAL The old steam generators may be stored onsite, at least temporarily, before eventual disposal. The alternatives associated with the storage and disposal are addressed below. 9.3.1 LONG-TERM STEAM GENERATOR STORAGE ONSITE As discussed in Section 4.4.6, the steam generators would be sealed before storage to ensure complete encapsulation of residual contamination and placed in a storage facility. The storage facility would provide adequate shielding around the steam generators. Access control and monitoring measures would be implemented during the storage period. At . the end of the plant lifetime, disposition of the steam generators will be accomplished in conjunction with plant decommissioning. 9.3.2 IMMEDIATE SHIPMENT BY BARGE The immediate shipment of the steam generators in one piece by barge as discussed in Section 4.4.3.3 is currently considered to be a viable method of disposal. Immediately upon removal from the containment, the steam generators would be loaded on barges and shipped to a licensed depository.
- 9-2
PALISADES PLANT SGRR 9.3.3 SHORT-TERM STORAGE ONSITE WHILE UNITS ARE CUT UP FOR SHIPMENT (WITHOUT DECONTAMINATION) If the steam generators are to be cut up for shipment, additional contamination control measures, as well as shielding, would have to be employed as discussed in Subsection 4.4.3.2. For this option, an enclosure with appropriate controls for airborne and liquid effluents will be required in addition to the requirements set forth in Section 4.4.2. Techniques for cutting and packaging are not well established, making cost and dose calculations uncertain. 9.
3.4 CONCLUSION
S The present-day conceptual cost and man-rem estimate for each method of steam generator storage and disposal are summarized below. Approximate Cost Man-Rem
- a. Immediate shipment by barge $353,000 1-5 9,. b. Long-term storage with disposition during decommissioning $2,560,000 5-10
- c. Cut up and disposal near term with no decontamination $1,756,000 575-750 It is apparent that immediate shipment by barge would be benefiqial from the standpoint of both man-rem exposure and cost.
- 9-3
PALISADES PLANT SGRR
10.0 REFERENCES
- 1. BSAP - Bechtel Structural Analysis Program, Bechtel Power Corporation, June 1978 .
- 2. United Nuclear Industries,
- Jnc. Study, Decontamination of Palisades Steam Generators, November 1, 1975
- 3. Electrical Power Research Institute (EPRI 404-2)
Primary System Shutdown Radiation Levels at Nuclear Power Generating Station, p.56, D~ceroper 1975
.4. ~cceptance Grite+ia for Emergency Core Cooling Systems for Light Water-Cooled Nuclear Power R~actors, 10 CFR 50.46 and Appendix K of 10 CFR 50,. Federal Register, Volume.39, Number 3, January 4, 1974 !?- LOCA Analysis fo.r*Palisades at 2530 MWt Using the ENC WREM-II PWR ECCS Evaluation Model, Exxon Nuclear Co. XN-NF-77-24, July 1977.
6; L.L. Beranek, Noise and Vibration Control, McGraw-Hill Book Company, Cl?.apters 7 and 18, 1971 I 10-1
SGRR A-l(a)
- A-l(a)
PALISADES PLANT SGRR Evaluation of Alternatives Table 8.8-1, Comparison of Major Repair Alternatives, estimates the cost of replacement power during the 24-month period required for replacement of the steam generators at a total cost of $200 million. The report stated that a 4-year procurement lead time is required for delivery of the new steam generators. Give a detailed cost breakdown of the type and cost of replacement power during this 24-month period as well as for the time periods for alternate repairs. The replacement power cost estimates should be in mills/kWh and thousands of dollars per day. Based on required procurement times, the earliest time of replacement would occur in 1983. Include the re-ference year when replacement power will be required for each type of steam generator repair.
RESPONSE
The time estimates required to repair the Palisades steam generators by means of replacement with complete units, re-placement with evaporator sections, or in-place tube replace-ment are 24, 30, and 36 months, respectively. The repair outage in each case is assumed to start in 1981 and will con-tinue for the above stated time periods. The average cost of net replacement power required during these outages is estimated to be 27.0 mills/kWh, which is equivalent to $275,000 per day. This replacement power cost is based on a net plant output of 675 MW, a capacity factor of 0.75, and a scheduled outage of 3 months for every 18 months of normal operation. The source of replacement power will be a combination of CPCo system generation and purchased power from other sources, de-pending on cost, availability, and time of day *
- A-l(a)-1 Revision 2 June 1979
SGRR A-l(b) PALISADES PLANT SGRR A-l(b) Length of Repair Outages Sections 8.4, 8.5, and 8.6 give the estimated times of repairs as 36 months for retubing the original units~ 24 months for complete replacement of the units, and 30 months for repair of the evaporator sections only. Table 8.4-1, Steam Generator Alternative Costs, gives the total estimated project costs of all three alternatives. Complete replacement, $75.0 million; evaporator repair,
$80.0 million; and retubing, $75.0 million. The project costs are fairly close, however, the big difference is the cost of replacement power associated with plant shutdown. Because of the extensive period of plant outage and high cost of replacement power, include a detailed explanation for these required times. Are the 24, 30, and 36-month periods based on single shifts or have double shifts been considered in an attempt to reduce the high cost of replacement power.
RESPONSE
- The attached Figures A-l(b)-1 and A-~(b)-2 provide the bases for the nominal outage durations of 24, 30, and 36 months to repair the Palisades steam generators by means of replacement with complete units, replacement with evaporator sections, or in-place tube replacement. Minor adjustments to these schedule~ in the areas of post-construction testing or contingency requirements are responsible for any differences which exist between the outage duration shown on these figures and the nominal times used in the report. In each case, the schedule is based on conducting the repair on a 6-day workweek, consisting of two 10-hour shifts per day for critical path activities. As noted in Section 8.6 of the report, the evaporator replacement alternative is a variation of the repair concept based on replacement with complete units, with the additional work of cutting, welding, and stress relieving on the transition cone being responsible for the schedule extension of from 6 to 8 months *
- A-l(b)-1 Revision 2 June 1979
*"I~',.
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- 1. THE FOLLOWING DURATIONS ARE l!IASEO ON 2 SHIFTS OF 10 HOURS PEA DAY I DAYS
-TENDON
- _ _REMOVAL
_ _ _&_ INSTAUATION 5 PEA WEEK: LINE 10 MEASURE, ERECT* _ _ _ _ _ _ _ _ _ _....J_(___ ~- >-- WELD SIG PIPING: LINE 1* INSULATION: LINE 5 DETENSION & REMOVE TENDONS AND PULL TENDONS: LINE 19 ILRT TEST: CONTAINMENT OPENING 6 LINE 6 CHIP CONCRETE* CUT REBAR* LINER' PLANT SHUTDOWN ' I SH~TDOW1 ICPCol I'~ I I I I I 2. RIGGING ACTIVITY DURATIONS ARE BASED ON 7 A llO HAI QAYLIGHTI WORK WEEK I 1 1 ! I I I : 3. THE REMAINING DURATIONS ARE BASED ON A .&Q HOUR WORK WEEK
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STEAM GENERATOR ELECTRICAL & CONTROLS II I I ".It.. JREllO E l J l.lact ~Triru I I I t------------------+--J--+--+--i-+:c-.+~-f.-{~*l-+--_lj*~1*~1+--.J_!j:_--!.i1__!ll--"'"'_J._~ll.--E~~ REACTOR VESSEL ROUNDNESS MONITORING 12 I. I I
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CONTINGENCY 21 I I I 'CIJ--+CO-llT~l~G~EC~Y.+--'.nll I I I C EC& 0 T, TES IUT STARTUP IPRE-OPI 22 I r---~~--------------it--+---lr--+--+--t----l--t---l---l-~~+--i-_j__""-1--1--l---l-__jL-_J_ _L_L__j__+--+-+--l----+~>-- STEAM GENERA TORS Z3 {Ju' <llY 24 25 PALISADES PLANT 26 STEAM GENERATOR REPAIR REPORT 27 STEAM GENERATOR MILESTONES 28 REPLACEMENT SCHEDULE
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COMPLE1E ENGINEER I NEGOTIATIONS PROCURE ANO DELIVER RAW MATERIAL TO FABRICATION Pl.ANT DMW BE BUND E ANO P.O. FABRICATE AMO OEi.1..m; SHROUD sibONs TO SITE DESJ0N ~ ~ING f!!!!!Q!TE I 1JSNEA
- RI~ I IW&Kfiii I 'RINGS AND TIE STllAPS TO SITE PAOcURE MATERIAL FOR EOG CllATEB I
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?MIC WBJIElt9 OEVB.OP DECONTAMINATION PLAN DISP SE OF STEAM OisPose OF TUBING a eoo CRATES & r BEAMS
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DECISION SEPARATORS & DRYERS 41 mMpt m a PROO' 'DE uegRJoL FOR i;ftRA1cov ettp MNSIRYQI wqn1 DESION PLATFORM PLATFORM OVER REFUELING POOL
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. * * ,. 1* 110 111 114 11* 11' 111l IU IN i* i,. REii Mimi ll'ES baNst11ERS l'CMl!R CO. (lOtmW:IOR'll SHOP & llEllllGN ENlllEEllWIG CONTRACrOR'S FIELD WORK
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YCJ\IE INTO ANO DRYERS PALISADES PLANT STEAM GENERATOR REPAIR REPORT STEAM GENERATOR RETUBING Figure A-l(b)-2
*-*-------"'*-~-------
SGRR A-2 PALISADES PLANT SGRR A-2 8.7 Man-Rem Consideration The three alternate repair methods have comparable direct costs. The Palisades Plant Stearn Generator Repair Report, Docket Number 50-255, also shows the occupational exposure of 1,200-5,000 man-rem for complete replacement of steam generators and replacement of evaporator sections. Re-tubing the steam generators, in place, has an estimated 40,000 man-rem exposure, Table 8.8-1, Comparison of Repair Alternatives. Based on biological savings of $1,000/man-rem cost is $35,000,000, if Alternate 3 is selected over the other alternatives. Give a detailed breakdown of how the man-rem exposure was estimated for retubing the steam gen-era tors. Also, include a breakdown of man-rem exposures for the other alternatives~ The man-rem estimate should be of same detail as Table 17, Comparison of Exposure Estimates in NUREG/CR-0199, "Radiological Assessment of Steam Generator Removal and Replacement."
RESPONSE
The three alternate repair methods described in Table 8.8-1 were replacement (complete units), replacement (evaporator
- sections), and retube in place
- A detailed breakdown of how the man-rem exposure was estimated for replacement of the steam generators as complete units has been shown in the repair report as Section 4.3.7 and Table 4.3.2.
The manhours and corresponding man-rem estimates associated with the replacement of the steam generator evaporator sections closely follow the manhour estimates made in Table 4.3.2. The difference would be a man-rem increase in Work Areas 10 and 11 which would reflect the additional time involved for a circum-ferential cut and weld on the shell of the steam generators. Since the significant influence on the total man-rem for either replacement alternative is the welding and cutting technique used on the primary coolant piping (see footnote, Table 4.3.2), it is felt that the 1,200-5,000 man-rem range presented in Tables 8.8-1 and 4.3.2 is an adequate description of either replacement as a complete unit or replacement of evaporator sections only *
- A-2-1 Revision 2 June 1979
SGRR A-'2-2 PALISADES PLANT SGRR The remaining repair technique described in Table 8.8-1 is to retube the steam generators in place. The following is a breakdown of how the man-rem exposure was estimated for the re-rubing alternative. Items (A-F) were used as the basis for the man-rem estimate. A. TASK TIME, WEEKS
- 1. Tube removal 20
- 2. Tube removal, 2 feet from tube sheet 2 3* Tube stub removal 16
- 4. Tube replacement, rolling, welding, etc 27 Total 65 B. It bas been estimated that approximately 300 boilermakers would be required to complete the above tasks. It is assumed that all 300 craftworkers can be used interchange-ably *
- c. Tube removal would begin at Week 46 and tube welding and rolling would begin at Week 73. The radiation dose rate at the tubesheet is estimated as 0.8 r/hr at Week 73.
The radiation dose rate at the tube bend area is 5 mr/hr at.Week 46. The radiation dose rate remains constant with respect to height in the steam generators from the U-bend area to within 2 feet of the tubesheet at which point it increases to the dose rate specified at the tubesheet. D. It has been estimated that only 35% of the workers' time will be spent in the specific radiation field (see Repair Report, Section 4.3.7). The remaining time is spent in security, access control, and time spent donning and re-moving protective clothing in nonradiation areas. E. Craftworkers are available at 150/shift, working 10 hours per shift, 2 shifts per day for a 6-day week. F. Reduction factors used for local decontamination and shielding during tube stub removal and tube replacement are 0.2 and 0.1, respectively *
- A-2-2 Revision 2 June 1979
SGRR A-2-3
- PALISADES PLANT SGRR Using the above information, the integrated dose for the retubing alternative is as follows:
0.005 r/hr x 150 men x 20 hrs/day x 20 wks x 6 days/wk x .35 = 630 man-rem
+ 0.8 r/hr x 150 men x 20 hrs/day x 2 wks x 6 day/wk x .35 =
10,080 man-rem
+ 0.8 r/hr x 150 men x 20 hrs/day x 16 wks x 6 days/wk x .35 x 0.2 = 16,128 man-rem + 0.8 r/hr x 150 men x 20 hrs/day x 27 wks x 6 days/wk x .35 x 0.1 = 13,608 man-rem Total = 630 + 10,080 + 16,128 + 13,608 = 40,446 man-rem
- A-2-3 Revision 2 June 1979
SGRR B-l(a) PALISADES PLANT SGRR B-l(a) The original Palisades steam generators appear to be unusually prone to promoting hideout of impurities from the coolant. This may be a result of the extremely dense forest of tubes and relatively small tube-to-tube spacing. Therefore, indicate: (a) The number of spacing of tubes in the replacement generators and how they compare with the existing generators
RESPONSE
The replacement steam generators will have 8,182 heat transfer tubes of 0.750-inch OD and .042-inch average wall. The original steam generators have 8,519 heat transfer tubes of 0.750-inch OD and .048-inch average wall. The number of heat transfer tubes for the replacement steam generator was established, using an equivalent UA basis, i.e., (UA) replacement steam generators= (UA) original steam generator. This equivalent UA basis will provide the replacement steam generators with essentially the same thermal performance capability as the original steam generator. The tube bundle geometry is arranged using a 1-inch triangular pitch pattern for the tube holes and is identical to the spacing pattern utilize<l on the original generators .
- B-l(a)-1 Revision 2 June 1979
SGRR B-l(b) PALISADES PLANT SGRR B-l(b) Details of design changes made to improve the flow characteristics to eliminate or reduce the potential for sludge accumulation, and . * *
RESPONSE
Design features which enhanc~ the flow distribution and circulation which have been included in the design of the replacement steam generators are:
- 1. Eggcrate type tube supports
- 2. Tube lane divider plate
- 3. Increased blowdown capability The eggcrate type tube support system used on the replacement steam generator offers several advantages over the drilled tube support plates used on the original generators from a hydraulic standpoint. Since an eggcrate support is of a relatively opentype construction, localized crevices adjacent to tube surfaces are minimized. The eggcrate tube support system in the steam generator provides a minimum number of potential localized steam blanketed areas which might be present in the annular gaps between tubes and drilled support plate holes. The open flow area of each eggcrate support is calculated to be 69% of the total flow area. From a different point of view, the eggcrate support obstructs only 31% of the available secondary flow area. The large open flow areas available in the eggcrate support system avoids the accumulation of boiler water deposits by minimizing local flow eddies and flat surfaces which are present in other commonly used tube bundle support systems. The tube support design on the replacement steam generator has also been designed to reduce the number of supports, both horizontal and vertical grids in the tube bend region. In addition, the vertical tube pitch in the bend region has been increased from 1.0 inches to 1.75 inches which significantly reduces flow resistance in the bend region. The resultant effect is that the tube support system selected for the replacement steam generator is a design which reduces the potential for localized corrosion and increases internal recirculation.
On the replacement steam generator, a divider plate is mounted in the tube lane between the hot and cold leg sides of the tube bundle. This divider plate prevents the preferential bypass of the tube bundle by recirculating water exiting from the downcomer along the tube lane, thereby forcing the recirculation fluid from the downcomer to be directed across the tube bundle, ensuring relatively high radial fluid velocities throughout the region
- above the tubesheet
- B-l(b)-1 Revision 2 June 1979
SGRR-B-l(b)-2 PALISADES PLANT SGRR The blowdown capabilities for the replacement steam generators compliment the improved hydraulicp associated with the replace-ment generators for solids control within the steam generator. With the tube lane divider plate preventing flow bypass, the lowest radial recirculation fluid flow velocities exist in or near the open center region of the tube bundle where dropout of the solid particles occur. It is in this region that the blowdown ducts are located to take suction in a circular pattern *
- B-l(b)-2 Revision 2 June 1979
SGRR B-l(c)
- B-l(c)
PALISADES PLANT SGRR Indicate what water chemistry controls you plan to use to minimize degradation of the replacement generators.
RESPONSE
Primary Coolant System We plan to maintain our current program to control water chemistry in the primary coolant system. Primary coolant system operating parameters are to be maintained as follows: pH 4.5 to 10.2 Lithium 0.2 ppm to 1.0 ppm Cl <0.12 ppm Dissolved o2 <0.10 ppm F <0.10 ppm pH is controlled by a balance between boric acid concentration and lithium concentration maintained by purification flowrate via the letdown system. F and Cl will be controlled by purification flowrate via the letdown system and/or the makeup source (makeup deminera~izers). Dissolved o will be maintained as low as possible using a 2 continued excess of hydrogen during operation and will be con-trolled using hydrazine, as necessary, at system startup or shutdown. Other corrosion products may be controlled by purification flowrate via the letdown system. Secondary Systems We plan to maintain our current "all volatile chemistry" program to control water chemistry in the shell side of the steam gene-rators and in the secondary system. Steam generator operation parameters are to be maintained as follows: pH 8.2 to 9.2 Cl <O.l ppm Dissolved o <0.01 ppm 2 Specific conductivity <7 rnrnho/cm The "all volatile chemistry" program (in effect for over 3 years at Palisades) consists primarily of adding hydrazine and morpho-line for control of dissolved o and pH, respectively. Dissolved o 2 will be maintained as low as 2 practical. B-l(c)-1 Revision 2 June 1979
SGRR B-l(c)-2
- PALISADES PLANT SGRR Cl will be controlled by flowrate through the full flow conden-sate polishing demineralizers and/or the makeup source (makeup demineralizers).
During wet layup, the shell side of the steam generators will be recycled to facilitate mixing of chemicals and improve the turnover time of the water inventory in the steam generators (reference Page 3-1 of the SGRR). Additional hydrazine and morpholine will be used for control of dissolved o and pH, respectively. 2 Since our earlier problems with steam generator degradation, modifications have been completed for both the condensate and feedwater systems to improve secondary chemistry as follows:
- 1. To improve water purity during startup, shutdown, and abnormal operations, full flow condensate polishing demineralizers have already been installed with feed-water recirculation capability.
- 2. To reduce impurities and to maintain integrity of the secondary feedwater and condensate system(s),
stainless steel tubing has been installed in the Numbers 1 through 4 f eedwater heaters and the main condenser has been retubed using primarily 90-10 Cu Ni tubing with small quantities of stainless steel tubing in the periphery tubes and in the air removal section.
- 3. To improve steam generator chemistry, a new makeup demineralizer and water storage facility has been installed, which produces a better quality of water and provides for a large reservoir of water for blow-down considerations.
B-l(c)-2 Revision 2 June 1979
SGRR B-2
- B-2 PALISADES PLANT SGRR The eggcrate tube supports in the replacement steam generators will be fabricated of Type 409 stainless steel. Since Type 409 is only marginally a stainless steel, provide documentation of technical data showing its behavior under off normal water chemistry conditions.
RESPONSE
Combustion Engineering, in conjunction with the Electric Power Research Institute, has performed model steam generator testing in various faulted environments where ferritic stainless steels were included side by side with carbon steel. The test results demonstrate conclusively that 409 ferritic stainless steels perform better than carbon .steel in all environments tested. General corrosion of 409 is less than that of carbon steel by at least a factor of four and localized attack in heat transfer areas is orders of magnitude less for 409. The following table (B-2-1) provides general corrosion rates. Localized corrosion comparisons of carbon steel with 409 at heat transfer/tube support locations were conducted in a model boiler which operated for 282 days, utilizing intermittent seawater injec-tions at concentrations of from less than 50 ppb to 30 ppm chloride
- The average daily chloride concentration was approximately 4~8 ppm (conductivity 5 to 200 µmhos/cm). Test results are shown in the following table (B-2-2).
B-2-1 Revision 2 June 1979
SGRR B-2-2
- PALISADES PLANT SGRR TABLE B-2-1 GENERAL CORROSION RATES IN FAULTED VOLATILE CHEMISTRY MODEL BOILER ENVIRONMENTS Average Surface Corrosion Rate (Mils per year)
Faulted Chemistry Days Carbon 409 Condition Steaming Steel Stainless Fresh water 287 0.08 0.01 condenser leakage (12-15 µmhos/cm) Ammonia cycle 179 0.16 0.04 condensate polisher effluent (12-15 µmhos/cm) Intermittent sea- 179 0.11 0.03 -, wa te:r condenser leakage (5-200 µmhos/cm) Intermittent sea- 179 0.24 0.04 water condenser leakage (10-20 µmhos/cm) Condensate polisher 273 1.17 0.07 resin fines addition (20-60 µmhos/cm) NOTE: Normal steam generator conductivity is maintained at approximately 3.0 µmhos/cm with an alarm setpoint at 4.0 µmhos/cm Revision 2 B-2-2 June 1979
SGRR B-2-3
- PALISADES PLANT SGRR TABLE B-2-2 TUBE SUPPORT MATERIALS COMPARISON CORROSION TESTING AT HEAT TRANSFER/SUPPORT LOCATIONS (MODEL BOILER TESTING)
Carbon Steel 409 Parameter Support Plate Eggcrate Denting 8 mils radially None Support plate Extensive None cracking Corrosion of Extensive 6.4 mils support material maximum pitting B-2-3 Revision 2 June 1979
SGRR B-3 PALISADES PLANT SGRR B-3 Indicate what materials will be used for the batwing strips in the replacement generators.
RESPONSE
The batwing strips on the replacement steam generator will be fabricated from 409 ferritic stainless steel strips. All tube support structures, including eggcrate supports on the replacement steam generators, will be fabricated using 409 stainless steel material *
- B-3-1 Revision 2 June 1979
SGRR B-4
- B-4 PALISADES PLANT SGRR Discuss what effects the thinner tube walls will have on the existing plugging criteria.
RESPONSE
The tube plugging criteria established for the original Palisades steam generator was based upori a minimum tube wall thickness of .017 inches for the steam generator tubes and tube supports to retain structural adequacy during a hypothetical large pipe break. Utilizing this minimum tube wall thickness, it was determined that tubes having a local uniform degradation up to 64% of the nominal tube wall could withstand the postu-lated accident conditions and still meet the requisite code and regulatory guide criteria for faulted conditions. For the replacement steam generator, it is anticipated that the results of similar analysis*will demonstrate the structural adequacy of the tubes and tube supports, for the same accident conditions at 2,530 MW t, while retaining the same minimum tube wall thickness requirements as the original steam generator. Since the replacement steam generator will utilize tubes having a nominal wall thickness of .042 inches, the allowable degradation value, when correlated on a percentage basis, is anticipated to be 59% of the nominal tube wall *
- B-4-1 Revision 2 June 1979
SGRR C-1
- C-1 ALAR.~
PALISADES PLANT SGRR Provide the following additional information regarding considerations (Sections 1.1.5 and 4.3.5): (1) Duration of exposure associated with anticipated replacement/repair tasks (2) Repetition rate of the tasks (3) Numbers of work force exposed during each task (4) Occupational exposures associated with antici-pated replacement/repair activities
RESPONSE
Tables (C-1-1 through C-1-5) provide the requested additional information. The manhour and corresponding man-rem estimates have changed from the original presented in Table 4.3.2. The changes are based on modifications to the welding techniques, described as Alternative (A 1 ) and Alternative (A 2 ), and further development of work packages.
- The new exposure estimates are as follows: 1,547 to 2,808 man-rem (A ), based on manual welding the reactor coolant pipe carbon 1 steel portion and machine welding the cladding (with remote viewing), and 1,537 to 2,663 man-rem (A ) based on
- machine welding (with remote viewing) both the 2 carbon steel and cladding. The man-rem range reflects two analyses, for 140 and 42 days after shut down for the corrunencement of primary system pipe cutting. These estimates do not include a contingency.
It should be noted that Work Area 8, which represents manhours spent inside the reactor coolant pipe, has been expanded to appropriately differentiate radiation field levels before and after local decontamination *
- C-1-1 Revision 3 July 1979
- PALIS. PLANT SGRR TABLE C-1-1 (Sheet 1)
MANUAL WELDING OF RC PIPE WITH MACHINE CLADDING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL *LOCATION PERFORMED MANHOURS COMMENTS 1 Scaffold, cut, and remove con- 180 6 .& 11 1 1,080 st ruction opening liner plate. 83 6 2 1 498 2 Install new liner plate for con- 233 6 3 1 1,398 struction opening, including 100 6 2 1 600 fitup, scaffolding welding, etc. 3 Cover and uncover spent fuel pool 270 7 16 0 1,890 (protective cover). 116 7 2 0 812 240 7 1 0 1,680 4 Cut reactor coolant pipe. 15 2 4 12 360 Inside radiation 15 2 6 12 360 control envelope 17 2 2 12 408 12 2 9 12 288 5 Machine weld preparation on ends 58 2 7 18 2,088 12 weld preps of reactor coolant pipes. 35 2 2 18 1,260 would be on decon-23 2 9 18 828 taminated pipe and 6 on pipe attached to reactor. 6 Rig, fitup, line up, and tack in 28 5 7 12 1,680 place reactor coolant pipe closure 17 5 2 12 1,020
'I spools. 11 5 9 12 660 *Refer to Table c-i-5
PALISADES PLANT SGRR TABLE C-1-1 (Sheet 2) MANUAL WELDING OF RC PIPE WITH MACHINE CLADDING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL"' *LOCATION PERFORMED MANHOURS COMMENTS 7 Weld hot leg reactor coolant 131 4 5 4 2,096 Manual welding BC pipe (carbon steel). 2 1 BC 4 B level 131 4 9 4 2,104 113 4 2 4 l,BOO B Clad hot leg reactor coolant 2 1 . BC 4 B Utilizing machine pipe (stainless steel) 50 2 9 4 400 welding with re-30 2 2 4 240 mote welding 9 Weld cold leg reactor coolant B9 4 5 B 2,864 Manual welding pipe (carbon steel). 2 1 BC B 16 Manual welding 91 4 9 B 2,896 77 4 2 8 2,448 10 Clad cold leg reactor coolant 2 1 BC B 16 Machine welding pipe (stainless steel) 45 2 9 8 720 with remote 28 2 2 8 448 viewing 11 Stress relieve reactor coolant 22 4 5 12 1,056 pipe. 2 1 BC 12 24 22 4 9 12 1,056 20 4 2 12 960
*Refer to Table C-1-5
PALISADES PLANT SGRR
- TABLE C-1-1 (Sheet 3)
MANUAL WELDING OF RC PIPE WITH MACHIN~ CLADDING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL *LOCATION PERFORMED MANHOURS COMMENTS 12 X-Ray and NDT reactor coolant 1/2 1 B 4B 24 Leave pill guide pipe. 4 2 5 120 1,960 in place BC when 3 2 2 120 720 possible in order 1 2 9 120 240 to reduce exposure. 13 Install cleanliness plugs in 1/2 1 BA 6 3 reactor coolant pipe prior to 1 2 4 6 12 cutting pipe. 14 Clean inside of reactor coolant 1 1 BC 12 B pipe after welding. 15 Cover steam generator reactor 1 4 BB 6 24 coolant nozzles 1 4 1 6 24 1 4 9 6 24 4 2 7 6 4B (Seal weld only) 16 Cover ends of reactor coolant 1 4 BB lB 72 pipe spools (temporary). 2 4 7 lB 144 1/2 4 1 lB 36 1 4 2 lB 72 17 Rig reactor coolant pipes to 3 4 5 6 72 decontamination area. 1 4 2 6 24
*Refer to Table C-1-5
PALISADES PLANT SGRR TABLE C-1-1 (Sheet 4) MANUAL WELDING OF RC PIPE WITH MACHINE CLADDING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL *LOCATION PERFORMED MANHOURS COMMENTS 18 Measure reactor coolant pipe 4 3 5 12 144 closure spools. 1 3 2 12 36 19 Remove insulation from steam 38 4 2 2 304 generators. 32 4 4 2 256 25 4 9 2 200 32 4 10 2 256 20 Reinsulate new steam generator. 180 4 2 2 1,440 120 4 9 2 960 150 4 11 2 1,200 150 4 17. I .2 1,200 21 Remove and replace reactor 5 2 2 12 120 coolant pipe insulation. 3 2 4 12 72 5 2 5 12 120 3 2 9 12 72 22 Cut, remove_, bevel, erect, weld, 260 3 2 2 1,560 stress relieve, and insulate main 2.77 3 9 2 1,662 steam lines at top of steam 40 3 10 2 240 generator. 148 3 11 2 888 148 3 12 2 888 23 X-Ray and NDT main steam line. 40 2 2 2 160 (6 weld) 25 2 9 2 100 65 2 11 2 260
*Refer to Table C-1-5
PAL.ISADES PLANT SGRR TABLE C-1-1 (Sheet S) MANUAL WELDING OF RC PIPE WITH MACHINE CLADDING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL ~~LOCATION PERF.ORMED MANHOURS COMMENTS 24 Cut, bevel, erect, weld, stress 162 2 2 2 648 reileve, and insulate feedwater 107 2 9 2 428 line. 2S 2 10 2 100 . 24S 2 11 2 980 2S X-Ray and NDT feedwater line. 20 2 2 2 80 (4 welds)
. 4S 2 11 2 180 26 Remove and replace miscellaneous 60 4 2 1 240 small pipe near steam generators. so 4 4 1 200 40 4 9 1 160 so 4 17 1 200 27 Move component cooling water 7S 4 2 1 300 tank and clean resin tank out of so 4 9 1 200 way and reinstall. 12S 4 14 1 soo 28 Install blowdown, tank, pump, and 40 3 2 1 120 insulation. 20 3 9 1 60 67 3 lS 1 201 29 Install new blowdown piping 122 6 2 1 732 inside containment. 13 6 4 1 78 92 6 s 1 552 177 6 9 1 1,062 *Refer to Table C-1-5
PALISADES PLANT SGRR TABLE C-1-1 (Sheet 6) MANUAL WELDING OF RC PIPE WITH MACHINE CLADDING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL 1>LOCATION PERFORMED MANHOURS COMMENTS 30 Install new blowdown piping 193 6 2 1 1,158 outside containment. 128 6 9 1 768 322 6 15 1 1,932 31 Remove electrical inside con- 37 4 2 1 148 tainment so steam generator can 63 4 4 1 252 be removed. 25 4 9 1 150 32 Reinstall electrical inside 90 8 2 1 720 containment. 60 8 9 1 480 75 8 11 1 600 75 8 17 1 600 33 Install electrical for new 92 4 2 1 368 blowdown system. 12 4 9 1 48 173 4 15 1 692 32 4 17 1 128 34 Install and remove equipment 30 2 2 1 60 required to monitor position of 10 reactor and steam generators 130 1 9 1 130 during weldup of reactor coolant 20 2 17 1 40 pipe. 20 2 18 1 40
*Refer to Table C-1-5
PALISADES PLANT SGRR * . TABLE C-1-1 (Sheet 7) MANUAL WELDING OF RC PIPE WITH MACHINE CLADDING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL ~°'LOCATION PERFORMED MANHOURS COMMENTS 35 Mobilize, install cages. 1,075 20 l l 21,500 Remove dome and buttress facia, relax tendons, remove tendons, chip concrete, cut rebar and tendon sheathing for containment construction. opening. 36 Replace opening on containment, 1,275 15 l l 19,125 including the following: Replace tendon sheathing, re bar, concrete and tendons, stress tendons and replace dome and buttress facia, demobilize. 37 Construct and remove barge slip. 829 14 l l 11,606 38 Foundations for rigging equip- 692 12 l l 8,304 ment, including sheetpiling, earthwork, concrete foundations and removal of foundations (at containment building). 39 Mobile heavy lift rigger. 100 15 l l 1,500 40 Assemble 4 crawlers. 150 10 l l 1,500
*Refer to Table C-1-5
PALISADES PLANT SGRR TABLE C-1-1 (Sheet 8) MANUAL WELDING OF RC PIPE WITH MACHINE CLADDING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL *LOCATION PERFORMED MANHOURS COMMENTS 41 Assemble jacking frame at barge. 62 12 1 1 744 42 Assemble jacking frame at con- 78 12 1 1 936 tainment. 43 Preassemble equipment for inside 78 12 1 1 936 containment. 44 Install lifting equipment inside 112 15 2 1 1,680 containment. 158 15 12 1 2,370 105 15 13 1 1,575 45 Remove existing steam generators 23 15 1 2 690 from containment (rigging). 26 15 2 2 780 44 15 4 2 1,320 19 15 19 2 570 46 Transport and store existing 31 10 19 2 620 steam generators. 47 Receive and ballast barge. 20 5 1 1 100 48 Offload, store, load, and trans- 158 10 1 2 3,160 port new steam generators. 49 Rerig as required to install. 16 15 1 2 480
*Refer to Table C-1-5
PALISADES PLANT SGRR TABLE C-1-1 (Sheet 9) MANUAL WELDING OF RC PIPE WITH MACHINE CLADDING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL ~'(LOCATION PERFORMED MANHOURS COMMENTS 50 Install new steam generators. 19 15 1 2 570 21 15 2 2 630 37 15 12 2 1,110 12 15 17 2 360 51 Remove all external rigging 225 15 1 1 3,375 equipment from site. 52 Remove all rigging equipment 75 15 2 1 1,125 from containment. 105 15 12 1 1,575 70 15 13 1 1,050 53 Remove internal rigging equip- 63 15 1 1 945 ment from site. 54 Miscellaneous rigging 150 5 1 1 750 55 Steam generator storage building. 406 10 1 1 4,060 56 Cut and remove top support of 8 5 10 1 40 steam generator. 5 5 2 1 25 3 5 9 1 15 57 Remove shims other steam 6 4 10 1 24 generator top support. 4 4 2 1 16 2 4 9 1 8
PALISADES PLANT SGRR TABLE C-1-1 (Sheet 10) MANUAL WELDING OF RC PIPE WITH MACHINE CLADDING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL *LOCATION PERFORMED MANHOURS COMMENTS 58 Remove hydraulic snubbers. 4 5 10 8 160 3 5 2 8 120 1 5 9 8 40 59 Unbolt existing* steam generat.or. 4 3 4 2 24 3 3 2 2 18 1 3 9 2 6 60 Remove shims at steam generator 15 3 - 4 2 90 base. 9 3 2 2 54 6 3 9 2 36 61 Bolt down steam generators. 10 3 5 2 60 6 3 2 2 36 4 3 9 2 24 62 Reshim bottom steam generator 50 3 5 2 300 (sliding base). 30 3 2 2 180 20 3 9 2 120 63 Replace top steam generator 120 3 11 2 720 support. 72 3 2 2 432 48 3 9 2 288 64 Reshim top steam generator 40 3 11 2 240 supports. 24 3 2 2 144 16 3 9 2 96
*Refer to Table C-1-5
PALISADES PLANT SGRR TABLE C-1-1 (Sheet 11) MANUAL WELDING OF.RC PIPE WITH MACHINE CLADDING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL *LOCATION PERFORMED MANHOURS COMMENTS 65 Reinstall steam generator hy- 30 4 11 8 960 draulic snubbers. 18 4 2 8 576 12 4 9 8 384 66 Miscellaneous pipe operations 1 25,088 Welders tests, (welders tests, material 2 7,030 training, material hangling, scaffolding, training, 4 952 handling and fabri-hangers and supports, line 5 3,699 cation of tents are testing, cleanup, tents). 8 20 in Location 1. Re-9 6,982 mainder of manhours 10 552 were allocated on 11 1,088 the basis of piping 12 413 manhours in each 15 914 location. 17 87 67 Distributables (startup, 1 11,436 Welders tests and cleanup, scaffolding, welders 2 13,298 miscellaneous in tests other than pipe fitters, 3 1,003 Area 1. Remainder miscellaneous). 4 1,526 of manhours allo-5 4,883 cated on the basis 6 131 of direct manhours 7 1,613 excluding Area 1. 8 51 9 9,766 10 567 11 2,921
*Refer to Table C-1-5
- PALIS. PLANT SGRR TABLE C-1-1 (Sheet 12)
MANUAL WELDING OF RC PIPE WITH MACHINE CLADDING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL *LOCATION PERFORMED MANHOURS COMMENTS 67 (Continued) 12 2,616 13 1,090 14 218 15 1,526 16 785 17 1,090 19 480 68 Nonmanual 1 99,160 Office personnel and 2 6,511 50% of engineers and 3 53 supervision's man-4 1,110 hours are in Location 5 2,872 Ill. Remainder of man-6 16 hours by discipline 7 85 were allocated to the 8 14 proper location based 9 5, 770 on direct manhours 10 426 expended in that loca-11 1,176 tion by discipline. 12 895 13 149 For example; The man-14 11 hours for electrical 15 1,143 engineers and supts 16 141 were allocated based 17 401 on the electrical 19 67 direct hours expended on each task at each location.
*Refer to Table C-1-5
TABLE C-1-2
-(She.-;tl)
SUMMARY
01' MANllOllRS FOR ALL TASKS HY LOC:ATlON MANUAL W~:I.llfNG 01' Rl\AC:TOR COOLANT C.S. l'll'E WTTll MACll I NE C:LAIJll I NC~ AND REMOTE VI l\Wl NG TASK NO. 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 J8 19 1 498 1,080 2 600" l,398 3 l, 680 812 1,890 4 l10fl 360 360 288 5 1,260 2,0BB 828 6 1,020 l,680 660 7 1,800 2,096 8 2,104 8 2110 8 1100 9 2 ,448 2,864 16 2,896 10 11110 16 720 ll 960 1,056 24 l,056 12 720 960 211 240 13 12 3 14 8 15 24 48 24 24 16 36 72 1411 72 17 24 72 18 36 144 19 )()/1 256 200 256 20 1, 4110 960 1,200 1,200 21 120 72 120 72 22 l,560 1,662 240 888 888 23 160 100 260 24 648 428 100 980 25 BO 180 26 2110 200 160 200 27 300 200 500 "l
.!f. \*
- TABLE C-1-2
. (Sf1~Pl 2)
StlMMARY OF MANllOIJHS F!lH i\f.L TASKS BY LOCATION Mi\Nlli\L WET.Ill NG OF REACTOll COOLANT C. S. PI PE WITll t!AC!IJNE CI.i\Dlll NG /\Nil REMOTE VI EWING TASK NO. I 2 9 1.0 lJ 19 28 120 60 201 29 712 78 552 I ,062 30 I, 158 768 1,932 31 J /ill 252 150 32 720 41l0 680. 600
- n 368 48 692 1.28 34 60 130 40 1,0 35 21,~>00 36 19' 125 37 ll ,606 JS 8 ']()/,
39 I ,500 40 I ,~>00 41 71,1, 1,2 'l:l6 43 <J:l6 44 I ,680 45 (,CJ() 780 I ,320 2,'.HO I ,575 116 570 47 I 00 620 48 3' 160 49 /,(lO so 5 70 630 I, 110 ]60 51 '.l, 375 \
~
52 I, 125 I ,5 75 I ,050 5:J 91,5
T/\11!.li C-1-2
-(siu;rt -ff SUMM/\HY OF tl/\NHOIJllS FOR /\J.I, T/\SKS JIY l.OC/\TTON M/\NUA I. WF. Lil I NG OF HF.ACTOR COOL/\NT C. S. I' I PF. WI Tll tl/\ClllN~'. C:l./\IJlllNr. /\ND RF.MO'.tE VIEWING
~- -'
T/\SK NO.
**-----1,-
8 9 10 11 12 ---14--"'15--1-6____ 17-*---rn -- ---ff*--------- 5t, 150 ~5 1, ,or,o S6 1.5 1,0 57 8 24 16 58 120 40 160 59 18 21, 6 60 90 36 6L 36 ciO 24 62 180 *ino 1.20 63 1,*17. 288 720 61, 11,1, 96 2110 65 '.)7(, 38'* %0 66 2'i, OHR 7, Cl'JO 7.0 (,, 982 552 l,Cl88 1,13 914 87 Tola I !Ji rrrl s 107 ,O(d 12,:n1, 2,l17R 3,6t.o 11 ,'l23 360 q,960 ~123 21,662 1,372 7,116 6,J56 2,625 500 3,739 1,906 2,615 40 l, 190 67 11,11% IJ,'198 I ,OOJ 1,526 4,883 131 1,61.J 51 <J' 766 567 2,921 Z,616 1,090 218 1,526 785 l ,090 480 68 <J'J, I (,o 6,'i11 'i1 1, uo 2,872 16 85 11. 'i. 770 426 l, 176 , MS I l19 ll l,143 141 401 67 Tot:il 217,657 52,141 1,'i34 6,276 l'J,678 507 5,658 288 39,198 2,365 11,213 9,867 3,864 729 6,408 2,832 4,106 40 1 '737
PALISADES PLANT SGRR TABLE C-1-3 (Sheet 1) .II MACHINE WELDING OF R.C. PIPE WITH REMOTE VIEWING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL *LOCATION PERFORMED MANHOURS COMMENTS l* Scaffold, cut, and remove 180 6 3 3 l,080 construction opening liner plate. 83 6 2 1 498 2 Install new liner plate for con- 233 6 3 1 1,398 struction opening, including 100 6 2 1 600 fitup, scaffolding, welding, etc. 3 Cover and uncover spent fuel 270 7 16 0 1,890 pool (protective cover). 116 7 2 0 812 240 7 l 0 1,680 4 Cut reactor coolant pipe. 15 2 4 12 360 Inside radiation con-15 2 6 12 360 trol envelope 17 2. 2 12 408 12 2 9 12 288 5 Machine weld prep on ends of 58 2 7 18 2,088 12 weld preps would reactor coolant pipes. 35 2 2 18 1,260 be on decontaminated 23 2 9 18 828 pipe and 6 on pipe attached to reactor. 6 Rig, fitup, lineup and tack in 28 5 7 12 1,680 place reactor coolant pipe 17 5 2 12 1,020 closure spools. 11 5 9 12 660
*Refer to Table C-1-5
PALISADES PLANT SGRR TABLE C-1-3 (Sheet 2) MACHINE WELDING OF R.C. PIPE WITH REMOTE VIEWING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL *LOCATION PERFORMED MANHOURS COMMENTS 7 Weld hot leg reactor coolant 4 2 7 4 .1 32 Utilizing machine pipe (carbon steel). 3 1 8C 4 12 welding with remote 234 1 5 4 935 viewing. 146 3 2 4 1,753 273 3 9 4 3,276 8 Clad hot leg reactor coolant 2 1 8C 4 8 Utilizing machine pipe (stainless steel) 50 2 9 4 400 welding with remote 30 2 2 4 240 viewing. 9 Weld cold leg reactor coolant 4 2 7 8 64 pipe (carbon steel). 2 1 8 8 16 160 1 5 8 1,280 100 3 2 8 2,400 186 3 9 8 4,464 10 Clad cold leg reactor coolant 2 1 8C 8 16 Machine welding pipe (stainless steel). 45 2 9 8 720 with remote viewing. 28 2 2 8 448 11 Stress relieve reactor coolant 22 4 5 12 1,056 pipe. 2 1 8C 12 24 22 4 9 12 1,056 20 4 2 12 960
*Refer to Table C-1-5
PALISADES PLANT SGRR TABLE C-1-3 *I (Sheet 3) MACHINE WELDING OF R.C. PIPE WITH REMOTE VIEWING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL *LOCATION PERFORMED MANHOURS COMMENTS 12 X-Ray and NDT reactor coolant 1/2 1 8C 48 24 Leave pill guide in pipe. 4 2 5 120 1,960 place 8C when possi-3 2 2 120 720 ble in order to re-1 2 9 120 240 duce exposure. 13 Install cleanliness plugs in 1/2 1 8A 6 3 reactor coolant pipe prior to 1 2 4 6 12 cutting pipe. 14 Clean inside of reactor coolant 1 1 8C 12 8 pipe after welding. 15 Cover steam generator reactor 1 4 8B 6 24 coolant nozzles. 1 4* l' 6 24 1 4 9 6 24 4 2 7 6 48 (Seal weld only) 16 Cover ends of reactor coolant 1 4 8B 18 72 pipe spools (temporary). 1 4 7 18 144 1/2 4 1 18 36 1 4 2 18 72 17 Rig reactor coolant pipes to 3 4 5 6 72 decontamination area. 1 4 2 6 24
*Refer to Table C-1-5
PALISADES PLANT SGRR
- TABLE C-1-3 (Sheet 4)
MACHINE WELDING OF R.C. PIPE WITH REMOTE VIEWING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL *LOCATION PERFORMED MANHOURS COMMENTS 18 Measure reactor coolant pipe 4 3. 5 12 144 closure spools. 1 3 2 12 36 19 Remove insulation from steam 38 4 2 2 304 generators. 32 4 4 2 256 25 4 9 2 200 32 4 10 2 256 20 Reinsulate new steam generator. 180 4 2 2 1,440 120 4 9 2 960 150 4 11 2 1,200 150 4 17 2 1,200 21 Remove and replace reactor coolant 5 2 2 12 120 pipe insulation. 3 2 4 12 72 5 2 5 12 120 3 2 9 12 72 22 Cut, remove, bevel, erect, weld, 260 3 2 2 1,560 stress relieve, and insulate main 277 3 9 2 1,662 steam lines at top of steam 40 3 10 2 240 generator. 148 3 11 2 888 148 3 12 2 888
*Refer to Table C-1-5
PALISADES PLANT SGRR TABLE C-1-3 (Sheet S) MACHINE WELDING OF R.C. PIPE WITH REMOTE VIEWING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL *LOCATION PERFORMED MANHOURS COMMENTS 23 X-Ray and NDT main steam line. 40 2 2 2 160 (6 welds) 25 2 9 ' 2 100 6S 2 11 2 260 24 Cut, bevel, erect, weld, stress 162 2 2 2 648 relieve, and insulate feedwater 107 2 9 2 428 line. 2S 2 10 2 100 245 2 11 2 980 25 X-Ray and NDT feedwater line. 20 2 2 2 80 (4 welds) 45 2 11 2 180 26 Remove and replace miscellaneous 60 4 2 1 240 small pipe near steam generators. so 4 4 1 200 40 4 9 1 160 50 4 17 1 200 27 Move component cooling water 7S 4 \ 2 1 300 tank and clean resin tank out of so 4 9 1 200 way and reinstall. 12S 4 14 1 soo
*Refer to Table C-1-5
' ~
' I *'
PALISADES PLANT SGRR TABLE C-1-3 (Sheet 6) MACHINE. WELDING OF R.C. PIPE WITH REMOTE VIEWING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL *LOCATION PERFORMED MANHOURS COMMENTS 28 Install blowdown, tank, pump, 40 3 2 1 120 and insulation. 20 3 9 1 60 67 3 15 1 201 29 Install new blowdown piping 122 6 2 1 732 inside containment. 13 6 4 1 78 92 6 \ 5 1 552 177 6 9 1 1,062 30 Install new blowdown piping 193 6 2 1 1,158 outside containment. 128 6 9 1 768 322 6 15 1 1,932
... I I *'
31 Remove electrical inside con- 37 4 2 1 148 tainment so steam generator can 63 4 4 1 252 be removed. 25 4 9 1 150 32 Reinstall electrical inside 90 8 2 1 720 containment. 60 8 9 1 480 75 8 11 1 600 75 8 17. 1 600 33 Instail electrical for new 92 4 2 1 368 blowdown system. 12 4 9 1 48 173 4 15 1 692 32 4 17 1 128
- Refer to Table C-1-5
PALISADES PLANT SGRR TABLE C-1-3 (Sheet 7) MACHINE WELDING OF R.C. PIPE WITH REMOTEI VIEWING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL . )'q_,oCATION 'PERFORMED I MANHOURS COMMENTS 34 Install and remove equipment 30 2 2 l 60 required to monitor position of lQ reactor and steam generators dur- 130 l 9 l 130 ing weldup of reactor coolant 20 2 17 l 40 pipe. 20 2 18 l 40 35 Mobilize, install cages. Remove 1,075 20 l l 21,500 dome and buttress facia, relax tendons, remove tendons, chip concrete, cut rebar and tendon sheathing for containment con-struction opening. 36 Replace opening on containment, 1,275 15 l l 19,125 including the following: Replace tendon sheathing, rebar, con-crete and tendons, stress ten-dons and replace dome and butt-ress facia, demobilize. 37 Construct and remove barge slip. 829 14 1 l 11,606
*Refer to Table C-1-5
PALISADES PLANT SGRR
- TABLE C-1-3 (Sheet 8)
MACHINE WELDING OF R.C. PIPE WITH REMOTE VIEWING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL *LOCATION PERFORMED MANHOURS COMMENTS 38 Foundations for rigging 692 12 \ 1 1 8,304 equipment, including sheet-piling, earthwork, concrete foundations, and removal of foundations (at containment building).
' ... I I *'
39 Mobile heavy lift rigger 100 15 1 1 1,500 40 Assemble 4 crawlers. 150 10 1 1 1,500 41 Assemble jacking frame at barge. 62 12 1 1 744 42 Assemble jacking frame at con- 78 12 1 1 936 tainment. 43 Preassemble equipment for in- 78 12 1 1 936 side containment. 44 Install lifting equipment in- 112 15 2 1 1,680 side containment. 158 15 12 1 2,370 105 15 13 1 1,575
*Refer to Table C-1-5
PALISADES PLANT SGRR TABLE C-1-3 (Sheet 9) MACHINE WELDING OF R.C. PIPE WITH REMOTE VIEWING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL .,,.LOCATION PERFORMED MANHOURS COMMENTS 45 Remove existnig steam generators 23 15 1 2 690 from containment (rigging). 26 15 2* 2 780 44 15 4 2 1,320 19 15 19 2 570 46 Transport and store existing 31 10 19 2 620 steam generators. 47 Receive and ballast barge. 20 *5 1 1 100 48 Offload, store, load, and trans- 158 10 1 2 3,160 port new steam generators. 49 Rerig as required to install. 16 15 1 2 480 50 Install new steam generators. 19 15 1 2 570 21 15 2 2 630 37 15 12 2 1,110 12 15 17 2 360 51 Remove all external rigging 225. 15 1 1 3,375 equipment from site.
*Refer to Table C-1-5
PALISADES PLANT SGRR TABLE C-1-3 (Sheet 10) MACHINE WELDING OF R.C. PIPE WITH REMOTE VIEWING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL 1'LOCATION PERFORMED MANHOURS COMMENTS 52 Remove all rigging equipment 75 15 2 1 1,125 from containment. 105 15 12 1 1,575 70 15 13 1 1,050 53 Remove internal rigging equipment 63 15 1 1 945 from site. 54 Miscellaneous rigging. 150 5 1 1 750 55 Stearn generator storage building. 406 10 1 1 4,060 56 Cut and remove top support of 8 5 10 1 40 steam generator. 5 5 2 1 25 3 5 9 1 15 57 Remove shims other steam generator 6 4 10 1 24 top support. '( 4 4 2 1 16 2 4 9 1 8 58 Remove hydraulic snub be rs. 4 5 10 8 160 3 5 2 8 120 1 5 9 8 40 59 Unbolt existing steam generator. 4 3 4 2 24 3 3 2 2 18 1 3 9 2 6
*Refer to Table C-1-5
PALISADES PLANT SGRR
- TABLE C-1-3 (Sheet 11)
MACHINE WELDING OF R.C. PIPE WITH REMOTE VIEWING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL *LOCATION PERFORMED MANHOURS COMMENTS 60 Remove shims at steam generator 15 3 4 2 90 base. 9 3 2 2 54 6 3 9 2 36 I 61 Bolt down steam generators 10 3 5 2 60 6 3 2 2 36 4 3 9 2 24 62 Reshim bottom steam generator 50 3 5 2 300 (sliding base). 30 3 2 2 180 20 3 9 2 120 63 Replace top steam generator 120 3 11 2 720 support. 72 3 2 2 432 48 3 9 2 288 64 Reshim top steam generator 40 3 11 2 240 supports. 24 3 2 2 144 16 3 9 2 96 65 Reinstall steam generator hydraulic 30 4 11 8 960 snubbers. 18 4 2 8 576 12 4 9 8 384
*Refer to Table C-1-5
* ~*
PALISADES PLANT SGRR TABLE C-1-3 (Sheet 12) MACHINE WELDING OF R.C. PIPE WITH REMOTE VIEWING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL *LOCATION PERFORMED MANHOURS COMMENTS 66 Miscellaneous pipe operations 1 25,058 Welders tests, (welders tests, material 2 6,993 training, material handling, scaffolding, training, 4 1,064 handling and fabri-hangers and supports, line 5 1,954 cation of tents are testing, cleanup, tests). 7 217 in Location 1. Re-8 20 mainder of manhours 9 8,358 were allocated on 10 543 the basis of piping 11 1,107 manhours in each 12 434 location. 15 933 17 87 67 Distributables (startup, 1 11,400 cleanup, scaffolding, welders 2 13 ,385 tests other than pipe fitters, 3 1,003 miscellaneous). 4 1,526 5 3,052 6 131 7 1,744 8
*Refer to Table C-1-5
PALISADES PLANT SGRR TABLE C-1-3 (Sheet 13) MACHINE WELDING OF R.C. PIPE WITH REMOTE VIEWING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL *LOCATION PERFORMED MANHOURS COMMENTS 67 (Continued) 9 11,466 10 567 11 2 ,921 14 2,616 13 1,090 14 218 15 1,526 16 785 17 1,090 19 480 68 Nonmanual 1 99,160 Office personnel +50% 2 6,431 of engineers and 3 53 supervision manhours 4 1,191 are in Location 1. 5 1,648 Remainder of manhours 6 16 by discipline were 7 238 allocated to the 8 61 proper location based 9 6,753 on direct manhours 10 420 expended in that 11 1,191 location by discipline. 12 . 911
*Refer to Table C-1-5
PALISADES PLANT SGRR TABLE C-1-3 (Sheet 14) MACHINE WELDING OF R.C. PIPE WITH REMOTE VIEWING NUMBER AVERAGE OF NUMBER TIMES TASK DURATION OF TASK NUMBER DESCRIPTION (HOURS) PERSONNEL 1'LOCATION PERFORMED MANHOURS COMMENTS 68 (Continued) 13 149 For example: The 14 11 manhours for elec-15 1,158 trical engineers 16 141 and supts were al-17 401 located based on the 18 electrical direct 19 67 hours expended on each task at each location.
*Refer to Table C-1-5
TAll!.E C-1-*4 (Sheet 1)
SUMMARY
Of MANllOURS FOR ALL TASKS BY LOCATlON WEI.DING OF rrnACTOR COOLANT PIPE BY MACllINI\ WlTll RF.MOTE VIEWING TASK NO. 1 2 3 4 5 6 8 9 10 11 12 13 14 15 16 17 18 19 1 498 1,080 2 600 1,398 3 1,680 812 1,890 4 1108 360 360 288 5 1,260 2,088 828 6 1,020 1,680 660 7 l, 7 53 935 32 12 3,276 8 2110 8 400 9 2 ,1100 1,280 611 16 4,464 JO 11/18 16 720 ll %0 1,056 211 1,05.6 12 720 960 24 240 l3 12 3 14 8 15 24 48 21, 24 16 36 72 1114 72 17 24 72 18 36 144 19 304 256 200 256 20 1,4110 960 1,200 l,200 21 120 72 120 72 22 1,560 1,662 240 888 888 23 160 100 260 24 MB 428 100 980 25 80 180
TABLE C-1-4 (sh;~
SUMMARY
OF MANllOUHS FOR ALL TASKS BY LOCATION WEI.DING 01' RF.ACTOR COOLANT PIPE llY MACllINJ*; WITH REMOTE VIEWING TASK NO. 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 1B 19 26 211() 200 160 200 27 300 200 500 28 120 60 201 29 732 78 552 1,062 30 1, 158 768 1,932 31 1411 252 150 32 720 480 600 600 33 368 48 692 128 34 60 130 40 1,0 35 21 ,500 36 19,125 37 11,606 38 8,104 39 1,500 40 1,500 41 744 42 936 43 936 44 1,680 45 690 780 1,320 2,370 1,57 5 570 46 620 47 100 48 3,160 49 480 so 570 630 ] , 110 360 51 3,375 52 1,125 l, 575 1,050 53 945
TABLE C-1-4 (Sheet 3) SIJMMARY OF MANllOllRS FOR ALL TASKS BY LOCATION WELDING OF REACTOR COOLANT PIPE llY MACll I NI\ WI 'I'll REMOTE VIEWING TASK NO. 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 54 750 55 4,060 56 25 15 40 57 8 24 16 58 120 40 160 59 18 24 6 60 511 90 36 61 36 60 24 62 180 300 120 63 /i)2 2BB 720 64 11,4 96 240 65 576 3B4 960 Subtotal 82 ,021 25,209 2,478 2,664 5,479 360 4,056 207 19,453 820 6,028 5,943 2,625 500 2,825 1,906 2,528 110 1,190 66 25,058 6,993 - 1,064 1, 954 217 20 8,358 543 l, 107 1134 933 117 Total Di rec ts 107 ,079 32 ,202 2,47B 3,728 7 ,1133 360 11,273 227 27 ,811 1,363 7, 135 6,377 2,625 500 3,758 1,906 2,615 1,0 l,190 67 11,400 13 ,385 1,003 1,526 3 ,052 131 1,744 - 11,466 567 2 ,921 2,616 1,090 21B 1,526 7B5 1,090 480 68 99. 160 6,1131 53 1,191 l,64B 16 238 61 6,753 420 1,, ~? 1 911 149 11 l,15B 141 1101 67 Total 217 ,639 52 ,018 3,534 6,445 12'133 507 6,255 288 46 ,030 2,350 11,247 9,904 3 ,86/1 729 6,422 2,832 4 '106 /10 1 '7 3 7
PALISAllES PLANT SGRR TABLE C-1-5 (SHEET I) MAN-HF.M ESTIMATE (2) (1) Estimated Manhours in RRdlat1on FJeld Area Mnn-Hem Dose Average Radiation Field (Hnnual Welding) (Mnchlne Welding) (Mrrn-Rem) Work Area (LocatJon) (rem/hr) Al A2 (Manual Welding) (Machine Welding)
- 1. Outside of power plant buildlng but wi.thin -5 security fence. .5 x 10 217659 217639 1.1 1..1
- 2. Checking in and out through security and heal th physics as well as time spent s1Ji ting up, cleaning up and * 'J moving to and from work nreas for personnel working in radioactive areas. .0025 52143 52018 130.3 130
- 3. Inside containment near new construct ion opening. .001 3534 3.5 3.5
- 4. Within 6' of outslde of reactor coolant pipe or bottom of steam generator prlor tn removal of steam generators. .030 - .050 6276 188.3 - 313.8 193.3 - 322.3
- 5. WI.thin 6' outslde of reaclor coolant pipe after stenm generator's removal. .010 - .030 19678 12133 196.8 - 590.3 121.3 - 364.0
- 6. Within 6' of outside of reactor coolant pipe or bottom of steam generators with partial exposure to inside of reactor coolant pipe prior to stea~ ge11erator's re111oval. . 050 - 0.100 507 507 25.4 -* so. 7 25.11 - so. 7
- 7. Within 6' of outside of reactor coolant plpe with partlal exposure to inside of reactor coolant pipe after steam generator's removal. .050 - .100 5658 6255 283 312.B - 625.5
PALISADES PLANT SGRR TABLE C-1-5 (SHEET 2) MAN-REM 1'8TIMATE
'J (2)
(1) Estimated Manhours i.n Hadiation Field Area Man-Rem Dose Average Radiation Field (Manual Welding) (Machine Welding) (Man-Rem) Work Area (Locat:J.on) (rem/hr) Al A2 (Manual Welding) (Machine Welding)
- 8. a) Inside of reactor coolant p:J.pe before decontamination. 9.0 - 12.0 4.3 4.2 38.7 - 51.6 37.8 - 50.4 b) Outside of pipe w/partial exposure inside before decontamination. 1.0 - 2.0 136 134 136 - 272.0 134 - 268.0 c) Inside of reactor coolant pipe after decontamlnation. .035 148 150 5.2 5.3
- 9. Low radiation area within containment .005 39198 46030 196 230.2
., I
- 10. Withjn 6' of top half of original steam generators (installed in place) . .005 2365 2350 11.8 11.8
- 11. Within 6' of top half of new steam generators (installed in place). .001 11213 1124 7 11. 2 l l. 3
- 12. Operating floor or contalnment. .005 - .010 9867 9904 49.3 - 98.7 49.5 - 99.0
- 13. Inside containment, at polar crane elevation. .001 3867 3864 3.9 3.9
- 14. Auxiliary building near clean resin tank and cooling water tank. .001 729 729 0.7 0.7
- 15. Auxiliary build.ing near blowdown tank. .001 6408 6442 6.4 6 *
- 16. Spent fuel pool floor. .005 2832 2832 14.2 14. 2
- 17. Within 6' of the bottom.half of new steam generators On place). .OJO - .030 liJ06 4106 41.l - 123.2 l1l. l - 123.2
- PALISADES PLANT SGRR TAlll.E C-1-5 (SHEET J)
MAN*-1\EM l\STIM/\TE
" l (2)
(1) Estimated Mnnhours in Rndintlon Field Area Man-Rem Dose Average Radiati.on Field (Hnnual Welding) (Maddne Welding) (Mnn-Rem) Work Aren (J.ocatl.on) (rem/hr) Al A2 (Manual Weldfog) (Machtne Welding)
- 18. Within 6' of the outside of the reactor vessel. :100 40 40 4.0
- 19. Next to the exlsting steam generators outsl<le of the con-tainment .020 - .030 1737 1737 34.7 - 52.1 Jl1.7 - 52.l
- 20. Installation of sh.Leldlng and locnl decontmn.inat I.on. (3) (3) (3) 164.7 - -
285.3
- 2807.6 1537 - 2662.9 NOTE:
(1) Reduct1011 factors attributed to shieldi.ng and/or decontaminnti.on have been incorporated into these field estimates and wi.11 not be presented here as a separate column. (2) Further reduction in area man-rem dose could occur as work pnckages and ALARA studies contJnne. The estimates are based on conservat.1.ve assnmpt:l.ons. (3) The manhonr estimates for placement of shielding ancl locn) decontamination are tentative due to the continuation of ALARA analysis and work package development. Tl1e 1111mbers presented here are an est1mnte and n>present a percentage of the total man-rem.
SGRR C-2
- C-2 PALISADES PLANT SGRR Describe the designated contamination control envelopes and your plan to maintain occupational exposure within these envelopes "ALARA. 11 Include also dose rates, exposure times, and numbers of workers involved in the tasks (Section 4.3.3).
RESPONSE
The contamination control envelopes will be used for the cutting of reactor coolant piping (Task 4). Although the design of the envelopes has not been finalized, the envelopes will include - a high efficienGy filtration system. The flow of air within the envelopes will preclude the escape of contaminants through the tent openings used for entering and exiting the area. Each of the 12 cuts on primary coolant piping will require two workers approximately 15 manhours. It is estimated that 720 man-hours will be spent within the control envelopes, with an average radiation field of 30-50 mr/hr. This results in an estimated. 21-36 man-rem of occupational exposure.
- As described in the Repair Report, Section 4.3-1, personnel in-volved in work within areas with a high level of contamination will wear two sets of protective clothing. Respiratory protection will be required in accordance with Palisades health physics procedures. Sheet lead, lead wool blanket, or other shielding will be used where possible in accordance with ALARA guidelines *
- C-2-1 Revision 3 July 1979
SGRR C-3 PALISADES PLANT SGRR
- C-3 Provide a diagram showing the radiation surveys around the steam generator replacement/repair activities. Include similar radiation surveys for Figures 4.2-4 through 4.2-7.
Include a table showing the whole body dose received during the inspection and plugging of the degraded steam generator tubes for 1976, 1977, and 1978.
RESPONSE
- 1. Figures 4.3-3, 4.3-4, and 4.3-5, which are included in the Steam Generator Repair Report, show radiation surveys around the steam generators at various times after shutdown.
The fields specified are representative of those expected for the replacement repair activiti~s. It should be noted that the fields specified in Figure 4.3-5 are expected to decrease significantly at the time that pipe cutting begins, considered to be 42-140 days post-shutdown for study purposes. The decrease will follow the general radiation field near the steam generator piping shown in Figure 4.3-7. Figures 4.2-4 through 4.2-7 have been modified to include general field information in areas where replacement/repair activities will occur and are now designated as Figures C-3-1 through C-3-4.
- 2. As requested, Tables C-3-1 and C-3-2 show the whole body
- dose received for inspection and plugging steam generator tubes during 1976 and 1978, respectively.
TABLE C-3-1 PALISADES PLANT - RADIATION DOSE
SUMMARY
STEAM GENERATOR WORK 1976 Activity 1976 Exposure
- 1. ECT Personnel Inside steam generator (without shielding) 16.5R Outside steam generator 19.3R Total (received over a period of 21 days) 35.8R
- 2. Insert and remove shielding Inside steam generator 17.8R Outside steam generator 2.0R Total 19.8R
- 3. Insert templates Inside steam generator 25.3R Outside steam generator 2.8R total 28.lR
- 4. Brushing and rolling Inside steam generator 37.6R Outside steam generator 4.3R
- Total C-3-1 41.9R Revision 3 July 1979
SGRR C-3-2 PALISADES PLANT SGRR Activity 1976 Exposure
- 5. Insert plugs Inside steam generator 28.8R outside steam generator 2.9R Total 31.7R
- 6. Weld plugs Inside steam generator 54.4R Outside steam generator 15.3R Total 69.7R
- 7. QC inspection of above operation Inside steam generator 23.6R Outside sleam generator l.6R Total 25.2R
- 8. Engineers support of above operations 10.7R
- Exposure accumulated inside steam generator 204.0R Exposure accumulated outside steam
. generator 48.2R TOTAL ACCUMULATED EXPOSURE 262.9R NOTE: The exposure designated inside and outside steam generator are only estimates (although the total for
- each operation is accurate) , since inside steam generator data is extracted from high radiation dose
- summary sheets which also include some outside steam generator work. The net result is that the inside steam generator data r~ads slightly higher and the outside steam generator data reads slightly lower than actually occurred. Radiation exposure to engineers is not included in this breakdown.
TABLE C-3-2 PALISADES PLANT - RADIATION DOSE
SUMMARY
STEAM GENERATOR WORK 1978
*1978 Exposure Organization/Activity (Level 9 R/hr 3. 5 R/hr)
- 1. Consumers Power company Repairmen
- a. Manway cover removal/vacuuming 5.6
- b. Dam installation 45.0
- c. Surge line shielding + miscellaneous 5.5
- d. Flooring/RSS hot legs 8.5
- e. Flooring, struts, tracks, bridge A cold leg 14.3 B cold leg 22.5
- C-3-2 Revision 3 July 1979
SGRR C-3-3
- f.
PALISADES PLANT SGRR Tracks, bridge Activity *1978 Exposure A hot leg 6.1 B hot leg 6.3
- g. General maintenance 1.9
- h. Tube plugging A cold leg 3.1 A hot leg 1. 6 B cold leg 3.9 B hot leg 1. 8
- i. Equipment removal/leg cleanup 6.8 j . Dam removal 4.6
- k. Manway cover replacement 4.9
- 1. Cont*cleanup 2.6 145.0
- 2. HP coverage 6.0 3* NDT lab/contractors 44.3
- 4. CE - Setup 2.4 Deplugging A cold leg 7.4 A hot leg 4.3 Sleeving 14.9 29.0 Total exposure *224.3 Man-rem NOTE: *These exposures are based on dosimeters and TLDs.
Source is containment entry logs. These numbers should only be considered close approximations *
- C-3-3 Revision 3 July 1979
STM. 6EN. E-SOA
,* ( ES0*.8 IS TYP.; 13Ui OPPOSITE General Field WAND) 20-30 mr/hr CUT POINT---
- //
42 l.D. PIPING~(..,__~- 35 mr r General Field CUT POINT NEARSIDE' FARS/DE. f 30 "I. l). PIP/#6 ____,,,r_ NEARSJD£$ ~- FAR.SIDE -~ CUT POINT NEARSIDE J FAR.SID£ ELEVATION PALISADES PLANT STEAM GENERATOR REPAIR REPORT PRIMARY COOLANT _ J?Il?JNG_~UT . roINTS Figure C-3-1 Revision 3 July 1979
FUTURE LOCATION:~
---- . *""-"-----~
PR£5£ NT L OCAT/ON General Field CUT POINT-__,.-+--1 s-10 mr/hr CUT POINT i I
~ 36" M/l!N_ STEAM
~ STM. GEN.
. E-SOA (E-508 15 T.Yl?_J BUT OPPOSITE HAND) "
ELEVATION PALISADES PLANT STEAM GENERATOR REPAIR REPORT MAIN STEAM PIPING CUT POINTS Figure C-3-2 Revision*3 July 1979
r:---- 4 5T/V!. GEN. E- SOA T ( E-50 B IS TY?.) BlJT OPPOS/7£ HAND) General Field 5-10 mr/hr
/
*/
CUT POINT
- Cl.JT POINT II
/8. FEEDWAIE~
PLAN PALISADES PLANT STEAM GENERATOR REPAIR REPORT FEEDWATER PIPING CUT POINTS Figure C-3-3 Revision 3 July 1979
General Field CUT POINT 5-10 rnr/hr 2" 51.JRFAC E
':t::== <f /3L OW DOWN LINE General Field 5-10 rnr/hr
~ SrM. GEN £*50A
- ELEVATION STM. G/\/. E-50 8 TYP. J BUT OPPOSITE HANC IS rCUT POINT I 2" !30TIOM
<[ BLOWDOWN LIN~
35 rnr/hr General Field with Manways Covered ELEVATION ROTATCD .30" CLOCKW!Sc PALISADES PLANT STEAM GENERATOR REPAIR REPORT
- BLOWDOWN PIPING CUT_:;>OINTS Figure C-3-4 Revision 3 July 1979
SGRR C-4
- PALISADES PLANT SGRR C-4 Discuss briefly how you would avoid imbalance of the permanent ventilation systems due to the additional construction -related ventilation equipment (portable fans, hoods and filters, etc).
RESPONSE
Additional construction - related.ventilation equipment will be used to supplement the permanent ventilation system and will remove fumes associated with welding and cutting operation as well as ventilating temporary enclosures. Significant imbalance of the permanent ventilation system is not expected since the construction related ventilation equipment will exhaust inside the containment after filtration, and/or have relatively low flowrate. However, should imbalance of the permanent ventilation system occur, balance can be restored by controlling the dampers on the permanent ventilation system .
- C-4-1 Revision 3 July 1979
A SGRR C-5
- PALISADES PLANT SGRR C-5 Your description of compliance with Regulatory Guide 8.8, "Information Relevant to Ensuring that Occupational Ex-posure at Nuclear Power Stations Will Be As Low As Reasonably Achievable", Revision 3, June 1978, states that the follow-ing considerations were not implemented:
- 1. Radiation zones in the containment work areas, identifying the exposure levels in each work zone. C.2.(a)
- 2. Streaming or scattering of radiation from installed shielding, such as plugs in open ended pipe lines following cutting. C.2.b.(4)
- 3. Outleakage of airborne contamination from the contain-ment due to steam generator replacement/repair ~ctivities when the equipment hatch is open.
- 4. Operating experiences should be recorded, evaluated, and reflected in the selection of replacement fnstrumen-ta t ion. C. 2. ( c) ( 3 )
- 5. Provision to be implemented to minimize exposure of station personnel in performing code inspection, such as removable insulation, smooth welds, etc. (C.2.(i)(ll)
- 6. An adequate emergency lighting system can reduce potential exposures of station personnel by permitting prompt egress from high radiation areas if the station lighting system fails.
- 7. A staff member who is a specialist in radiation protection assigned the responsibility for contributing to and coor-dinating ALARA efforts in support of operation that could result in substantial individual and collective dose levels.
- 8. * ** Station work areas to limit the average concentration of radioactive material in air to levels below *** in Appendix B, Table 1, Column of 10 CFR Part 20. C.2.d Provide justification for not implementing these provisions of Regulatory Guide 8.8, and demonstrate that alternative precautions you have taken will provide comparable levels of protection *
- C-5-1 Revision 3 July 1979
SGRR C-5-2
- RESPONSE:
PALISADES PLANT SGRR The philosophy of the radiation protection group is to maintain the occupational dose to all personnel as low as is reasonably achievable (ALARA). This has been stated as a policy in the Palisades Plant Radiation Protection Manual. In order to ensure that the various provisions of Regulatory Guide 8.8 are evaluated adequately, a third party ALARA review of the repair effort has teen utilized. The reviewers' responsibilities are to develop ALARA "checklists" to be used in conjunction with the work pack-age descriptions for each of the repair tasks. The "checklists" will evaluate and recommend any measure found appropriate to maintaining the personnel exposure ~LARA as defined in Regulatory Guide 8.8 (Revision 3). Although the detailed work packages are presently in the develop-ment stage, each of the f0110wing specific provisions will neces-sarily be given complete considerat~on for use during any re-placement/repair activity:
- 1. The identification of exposure levels in any work zone
- is presently implemented per Palisades Plant health physics procedures. The Repair Report Section 4.3~5.4 describes the low background radiation waiting ar~as that will be used near each containment work area.
As specified, special signs, tape, or rope-off areas will be utilized to designate these zones. Inter-mediate zones could be utilized, if found ALARA, as individual work packages are developed.
- 2. The effective streaming or scattering of radiation from installed shielding, such as plugs in open ended pipe lines, can be minimized through the use of local decontaminating of pipe stubs, use of temporary shielding, or exposure control by controlling ingress or egress to work areas. (See Repair Report Sections 4.3.5.2 and 4.3.5.3). All of these ALARA techniques are being evaluated for use with the individual work packages per third party ALARA review described above *
- C-5-2 Revision 3 July 1979
,. SGRR C-5-3 PALISADES PLANT SGRR
- 3. In addition to the temporary enclosure on the construc-tion opening, airborne radioactivity inside containment during the steam generator repair effort will be con-trolled, monitored, and ultimately released via the plant vent stack. The air will be drawn through the hatches and construction opening and exhausted by the purge system via the plant ventilation stack, thus precluding airborne radioactive particles or gases from leaving containment openings utilized for construc-tion activities. (See Repair Report Section 4.3.3).
- 4. Continuous air monitors, area radiation monitors, and portable survey instruments will be used in ac-cordance with Palisades Plant health physics procedures.
Daily and weekly operational checks, calibration, and response settings will be implemented and recorded as required per Palisades Plant health physics procedures.
- 5. Appropriate provisions will be implemented to minimize exposure of station personnel in performing code in-spect"ion, such as removable insulation, smooth wel*ds, etc. *Design features to improve maintenance and in-spection are discussed in the Repair Report Section 2.2.2 .
- 6. An emergency lighting system will be available for the steam generator replacement activities.
- 7. A staff member who is a specialist in radiation protection will be assigned to the responsibility for coordinating ALARA efforts.
- 8. Special measures will be implemented to minimize and control the average concentration of radioactive material in air to below those specified in Appendix B, Table 1, Column 1 of 10 CFR Part 29. In addition to J-temporary enclosures in areas where cutting will occur, containment air will be conditioned for the removal of airborne radioactivity by use of filters .
- C-5-3 Revision 3 July 1979
SGRR C-6
- C-6 PALISADES PLANT SGRR Explain what steps you plan to take to help main-tain doses ALARA in this project. Indicate what use will be made of contaminations tents, lead wool blankets, gloveboxes, remote cutting and welding equipment, temporary shielding and venti-lation systems. Also indicate what equipment will be mocked-up for training purposes.
RESPONSE
An independent ALARA review of the steam generator repair effort has been utilized to make recommendations for maintaining doses ALARA (see question C-5) . The ALARA recommendations made will be incorporated into the various work packages. Each of the following items will be used where appropriate to maintaining doses ALARA.
- 1. Contamination Tents - The cutting of primary coolant piping will be contained within specially designed contamination control envelopes. The envelopes will be provided with high efficiency filtration.
- 2. Temporary shielding in the form of; lead wool blankets, lead shield plugs and sheet lead will be used where effective to maintaining doses ALARA.
Experience has shown that lead wool blankets can effectively reduce streaming around shield plug and sheet lead fittings.
- 3. Gloveboxes - As yet, there are no repair/replace-ment activities described, which can effectively use glovebox enclosures for maintaining d95es ALARA. As work procedures develop, glovebox techniques will remain a viable option.
- 4. Remote cutting and welding equipment - Automatic welding machlnes with remote viewing will be used for welding operations made to interior located stainless steel cladding. Measuring equipment for determining the location of reactor vessel and steam generator during weld-up and cutting opera-tions will utilize remote indicators.
Although the cutting techniques have not been finall.zed, the ALARA considerations will be one of the determining factors for final selection . C-6-1 Revision 3 July 1979
SGRR C-6-2
- PALISAGES PLANT SGRR
- 5. Ventilation Systems - Local construction related ventilation equipment will be used to supplement the permanent ventilation system and will remove fumes associated with welding and cutting operation as well as controlling the airborne contamination existing in the temporary enclosures.
- 6. Mock-Ups of reactor coolant pipe, and the steam generator primary head as well as the actual cutting and welding equipment will be utilized for training and work planning .
- c-6-2 Revision 3 July 1979
SGRR C-7 PALISADES PLANT SGRR C-7 Provide a table showing the occupational col-lective whole body dose estimates for the fol-lowing phases of the steam generator replacement repair acitivities: (1) preparation, (2) removal, (3) installation and (4) storage. Discuss briefly your procedure for calculating these doses, taking into account the dose reduc-tion measures proposed to maintain doses as low as reasonably achievable (ALARA) , including local decontamination, temporary lead shielding, pre-job planning, pre-job training and use of remote tools where practicable.
RESPONSE
Table C-7-1 provides the requested information. The table groups the various tasks into preparation, removal, installation, and storage phases of the replacement/repair effort. The man-rem is the product of (man-hours) X (average radiation field) for each task. It should be noted that each task consists of man-hours accumulated in several
- locations, and each location has a corresponding radiation field. A description of locations and average radiation fields is presented in Table c-1-s, credit taken for decon-tamination and temporary shielding is incorporated into the radiation field estimates. Reduction factors for shielding and/or decontamination are presented in the repair report, Table 4.3.2.
C-7-1 Revision 3
- July 1979
PALISADES PLANT SGRR TABLE C-7-l MAN-REM ESTIMATE (Preparation, Tnstallatlon, Removal, and Storage) Preparation Removal Installation Stor~----- Task No. }fils Task No. Mils Han-Rem Task No. Man-Rem Task No.
----- Man-Rem
- --- - -- - - -Mils
---- - --- - -Mils
--- - Man-Rem l 1,578 2.32 1 1,416 31.26 2 1, 993 2.90 46 620 12 .11 3 4,382 11.58 lJ 15 27 .36 5 11, 176 111.69 55 4,060 0.02 34 140 2.60 15 120 26.52 6 3,360 ' 89 .85 35 21, 500 0.11 16 324 79.JB 7 6,008 32 .13 37 11 ,606 0.06 19 1,016 l0.72 8 6118 2.88 38 8 ,JOl1 0.01 21 115 3.02 9 8, 2211 114 .88 39 1;500 0.01 22 476 .912 10 1. J 811 5.28 40 l,500 0.01 24 577 l. 9 l 11 3,096 19.08 41 7114 Cl.01 26 240 4.03 12 l, 9114 13.44 42 936 0.01 27 300 1.00 14 12 .112 43 936 0.01 31 550 8.68 l7 % .78
' \
44 5,625 17.62 45 3,360 52.95 18 180 1.53 47 100 0.01 56 80 . 311 20 4,800 9.60 48 J,160 0.01 57 48 .*1.1. 21 269 1.00 49 1180 0.01 58 320 l. 30 22 4,762 14. 27 59 48 .79 23 520 1.16 (1) (l) 165.28 60 180 J.02 24 1,579 3.33 25 260 16 .18 26 560 5.37 27 700 1.25 28 381 .80 29 2,11211 15.00 30 3,858 8.67 32 2, 1100 10.8 33 I, 236 3.13 34 130 2.6 36 19,125 0.09 50 2,670 10.72 51 3,375 .01 52 3,750 11. 73 53 945 0.01 54 750 0.01 61 120 0.81 62 600 1, .05 63 1,440 3 .21. 64 1180 1.08
----------*------ 65 l, 920 4.32 S11btotal 62 ,1191 199.66 25 l .l1J --'89;980 *----------*
11, 680 12.112 9,185 1154. 09 66 25 058 75.82 66 __ ____!_, 584 _______:!!.:_67 66 20 126 __ --=.=,: 60.911 --------------- Total Directs 87 ,5119 275.48 10,769 258 .10 110,106 515.03 11,680 12.112 DlstrlbutabJes 67 22,597 1211 2, 779 15.28 28,416 156.3 1,208 6. 6li Nonmnnual 68 49,301 73.86 6 ,.Q_~-----~:.2~ 62,000 92.33 ~§35 _ _...l.:2.L _____ Total 159.4117 1173.311 19,612 282.36 200', 522 763.66 8,523 23.01 (l) llecontamination/Shielding Installation
..,, SGRR C-8 PALISADES PLANT SGRR C-8 Discuss your cutting and welding operations and cleanup of surface contamination in respect to "ALARA" guide-lines (Section 4.3.3).
RESPONSE
We note that a definite schedule has not been set for the cutting and welding, that experience is being gained by way of similar operations at other plants, and that detailed work plans have not been completed. The final operations will reflect applicable experience, and ALARA consider-ations. The following is an outline of operations under consideration at this time.
- 1. Cutting of Reactor Coolant Pipe Since the Palisades reactor coolant pipe is carbon steel w/stainless steel cladding, a plan is to utilize a track mounted oxygen-acetylene torch to cut the pipe.
Consideration is also being given to mechanically cutting the pipe in order to minimize the pipe lost during the cutting process and to facilitate the machining operation. The cutting operation will be accomplished in an enclosure to* limit spread of contamination.
- 2. Handling of Pipe to Decontamination Area After the pipe has been cut, temporary shield plugs will be secured to each open end of the reactor coolant pipe and the short pieces of pipe moved to a decontami-nation area. The temporary shield plugs will be mechanically attached to the pipe in order to reduce the number of welding and cutting operations.
- 3. Field Machining of the Re Pipe Attached to Reactor and Reactor Coolant Pumps Temporary shield plugs will be inserted a short distance into the pipe. The pipe between the shield plug and the end of the pipe will then be decontaminated to the extent practical. The pipe weld preparations would then be field machined.
C-8-1 Revision 3 July 1979
1 SGRR C-8-2 PALISADES PLANT SGRR
- 4. Field Machining of the Short Pieces of Reactor Coolant Pipe After the new steam generator has been placed, the dimensions between the existing reactor coolant pipe and the new steam generator nozzles will be transferred to the short pieces of reactor coolant pipe and the pipe weld preparations machined.
- 5. Welding of RC Pipe After set-up of the reactor coolant pipe the joints may be welded by one of the following methods:
- 1. Manually weld the carbon steel portion of the reactor coolant pipe and utilize an automatic welding machine with remote viewing for welding the stainless steel interior cladding.
- 2. Utilize an automatic welding machine .with remote viewing to weld both the carbon steel and stain-less steel interior cladding.
- 6. Clean-up of Surface Contamination Loose surface contamination will be removed manually
*from the outside of reactor co6lant pipe pieces, prior to cutting operations and again, prior to removal from contamination control envelopes. Plastic or other impervious sheeting will be used to cover pipe pieces before relocating to decontamination area.
C-8-2 Revision 3
'July 1979
l SGRR C-9
- C-9 PALISADES PLANT SGRR Your estimated dose range of 1,200 to 5,000 man-rem for the steam generator replacement/
repair activities is too wide to assure that occupational exposure will be as low as reasonably practicable. Justify the high end of the range as being ALARA. Experiences with other designs indicate the feasibility of performing such work with substantially lower total doses than the high end of range you have predicted.
RESPONSE
As presented in the task description (Question C-1), th~re are two welding techniques being considered for reactor coolant piping. One technique utilizes manual welding of reactor coolant carbon steel pipe with a machine weld-up of the stainless steel cladding. This technique would result in an estimated 1,547-2,808 man-rem for the repair. The second technique utilizes a machine weld-up of both carbon steel pipe and stainless steel cladding. This alternative results in an estimated 1~537-2,663 man-rem for the repair effort
- The high end of the range presented in Table 4.3.2, 5,000 man-rem, resulted from a manual weld-up of carbon steel pipe and cladding entirely from the inside. Due to the technical feasibility of the two welding techniques described above, this third alternative is no longer con-sidered ALARA and has since been eliminated.
C-9-1 Revision 3 July 1979
\ '
SGRR C-10 PALISADES PLANT SGRR C-10 Provide a rough breakdown of the activities, person-hour occupancies, and projected dose rates which are used in deriving the esti-mated total of 40,250 man-rems per unit for retubing in place (Section 8.7).
RESPONSE
Refer to Appendix A of the SGRR response to Question A-2, as transmitted to the USNRC by the Consumer Power Company letter dated June 11, 1979. That analysis is currently under reevaluation. c-10-1 Revision 3 July 1979}}