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The operating curves including pressure-temperature limita tions are calculated in accordance with 10 CFR 50, A ppendix G and ASME Code, Section III, Appendix G, and WCAP-14040 (Reference 1) requirements. The results of the material surveillance program described in Section 5.3.1.6 will be used to verify that the RTNDT predicted from the effects of the fl uence and copper content curve is appropriate and to make any changes necessary to correct the fluence and copper curves if RTNDT determined from the surveillance program is greater than the predicted RTNDT. Temperature limits for preservice hydrotests and inservice leak and hydrotests will be calculated in accordance with Appendix G of t he ASME Code, Section III. Compliance with Regulatory Gui de 1.99 is discussed in Appendix 3A. 5.3.2.2Operating Procedures The transient conditions that are considered in the design of t he reactor vessel are presented in Section 3.9(N).1.1. These transients are representative of the operating conditions that should prudent ly be considered to occur du ring plant operation. The transients selected form a conservative basis for evaluation of the RCS to insure the integrity of the RCS equipment. Those transients listed as upset condition transients are given in Table 3.9(N)-1. None of these transients will result in pressure-temperature ch anges which exceed the heatup and cooldown limitations , as described in Section 5.3.2.1 and in the PTLR. 5.3.3REACTOR VESSEL INTEGRITY 5.3.3.1Design The reactor vessel is cyli ndrical with a welded hemispher ical bottom head and a removable, bolted, flanged, and gasketed hemispherical upper head. The reactor vessel flange and head are sealed by two hollow metallic O-rings. Seal le akage is detected by CALLAWAY - SP5.3-10Rev. OL-21 5/15means of two leakoff connections: one between the inner and outer ring and one outside the outer O-ring. The vessel contains the core, core suppor t structures, control rods, and other parts directly associat ed with the core. The reactor vessel closure head contains head adaptors. These head adaptors are tubular members, attached by partial penetration welds to the underside of the closure head. The upper end of th ese adaptors are welded to the lower end of a CRDM latch housing or instrumentation port head adapter flange. Inlet and outle t nozzles are located symmetr ically around the vessel. | The operating curves including pressure-temperature limita tions are calculated in accordance with 10 CFR 50, A ppendix G and ASME Code, Section III, Appendix G, and WCAP-14040 (Reference 1) requirements. The results of the material surveillance program described in Section 5.3.1.6 will be used to verify that the RTNDT predicted from the effects of the fl uence and copper content curve is appropriate and to make any changes necessary to correct the fluence and copper curves if RTNDT determined from the surveillance program is greater than the predicted RTNDT. Temperature limits for preservice hydrotests and inservice leak and hydrotests will be calculated in accordance with Appendix G of t he ASME Code, Section III. Compliance with Regulatory Gui de 1.99 is discussed in Appendix 3A. 5.3.2.2Operating Procedures The transient conditions that are considered in the design of t he reactor vessel are presented in Section 3.9(N).1.1. These transients are representative of the operating conditions that should prudent ly be considered to occur du ring plant operation. The transients selected form a conservative basis for evaluation of the RCS to insure the integrity of the RCS equipment. Those transients listed as upset condition transients are given in Table 3.9(N)-1. None of these transients will result in pressure-temperature ch anges which exceed the heatup and cooldown limitations , as described in Section 5.3.2.1 and in the PTLR. 5.3.3REACTOR VESSEL INTEGRITY 5.3.3.1Design The reactor vessel is cyli ndrical with a welded hemispher ical bottom head and a removable, bolted, flanged, and gasketed hemispherical upper head. The reactor vessel flange and head are sealed by two hollow metallic O-rings. Seal le akage is detected by CALLAWAY - SP5.3-10Rev. OL-21 5/15means of two leakoff connections: one between the inner and outer ring and one outside the outer O-ring. The vessel contains the core, core suppor t structures, control rods, and other parts directly associat ed with the core. The reactor vessel closure head contains head adaptors. These head adaptors are tubular members, attached by partial penetration welds to the underside of the closure head. The upper end of th ese adaptors are welded to the lower end of a CRDM latch housing or instrumentation port head adapter flange. Inlet and outle t nozzles are located symmetr ically around the vessel. | ||
Outlet nozzles are arranged on the vessel to facilitate opt imum layout of the RCS equipment. The inlet nozzles are tapered from the coolant loop vessel interfaces to the vessel inside wall to redu ce loop pressure drop. | Outlet nozzles are arranged on the vessel to facilitate opt imum layout of the RCS equipment. The inlet nozzles are tapered from the coolant loop vessel interfaces to the vessel inside wall to redu ce loop pressure drop. | ||
The bottom head of the vessel contains penetration nozzles for connection and entry of the nuclear incore instrumentation. Each nozzle consists of a tubular me mber made of either an Inconel or an Inconel-stainless steel composite tube. Each tube is attached to the inside of the bottom head by a partia l penetration weld. Internal surfaces of the vessel which are in contact with primary coolant are weld overlay with 0.125 inch minimum of stai nless steel or Inconel except for an area approximately 1.5 inches by 0.75 at approximate location 302.94 o from vessel "0" and 384.89 inches down from the flange surface and an area approximately 0.53 in ches by 0.3 inches at approximate location 185o from vessel "0" and 385 inches down from the flange surface. | The bottom head of the vessel contains penetration nozzles for connection and entry of the nuclear incore instrumentation. Each nozzle consists of a tubular me mber made of either an Inconel or an Inconel-stainless steel composite tube. Each tube is attached to the inside of the bottom head by a partia l penetration weld. Internal surfaces of the vessel which are in contact with primary coolant are weld overlay with 0.125 inch minimum of stai nless steel or Inconel except for an area approximately | ||
===1.5 inches=== | |||
by 0.75 at approximate location 302.94 o from vessel "0" and 384.89 inches down from the flange surface and an area approximately 0.53 in ches by 0.3 inches at approximate location 185o from vessel "0" and 385 inches down from the flange surface. | |||
The existence of these areas has been evaluated as acceptable.The reactor vessel is designed and fabricated in accordance with the requirements of the ASME Code, Section III. Principal design para meters of the reactor vessel are given in Table 5.3-2. The reactor vessel is shown in Figure 5.3-1. There are no special design features which would prohibit the in-s itu annealing of the vessel. If the unlikely nee d for an annealing operation wa s required to restore the properties of the vessel material opposite the reactor core because of neutron irradiation damage, a metal temperature greater than 650°F for a peri od of 168 hours maximum would be applied. Various modes of heating may be used, depending on the temperature required. | The existence of these areas has been evaluated as acceptable.The reactor vessel is designed and fabricated in accordance with the requirements of the ASME Code, Section III. Principal design para meters of the reactor vessel are given in Table 5.3-2. The reactor vessel is shown in Figure 5.3-1. There are no special design features which would prohibit the in-s itu annealing of the vessel. If the unlikely nee d for an annealing operation wa s required to restore the properties of the vessel material opposite the reactor core because of neutron irradiation damage, a metal temperature greater than 650°F for a peri od of 168 hours maximum would be applied. Various modes of heating may be used, depending on the temperature required. | ||
The reactor vessel materials surveillanc e program is adequate to accommodate the annealing of the reactor vessel. Sufficient specimens are available to evaluate the effects of the anneal ing treatment. | The reactor vessel materials surveillanc e program is adequate to accommodate the annealing of the reactor vessel. Sufficient specimens are available to evaluate the effects of the anneal ing treatment. | ||
Revision as of 04:27, 9 October 2018
| ML17074A236 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 03/15/2017 |
| From: | Klos L J Plant Licensing Branch IV |
| To: | Diya F Union Electric Co |
| Klos L J | |
| References | |
| Download: ML17074A236 (204) | |
Text
CALLAWAY - SP5.0-iTABLE OF CONTENTSCHAPTER 5.0REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS Section Page5.1
SUMMARY
DESCRIPTION..........
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.....................5.1-15.1.1DESIGN BASES...........
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.....................5.1-35.1.4SYSTEM PERFORMANCE CHARACTERISTICS
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.....................5.1-55.2INTEGRITY OF REACTOR COOLANT PRESSURE BOUNDARY................5.2-15.2.1COMPLIANCE WITH CODES AND CODE CASES.............
.....................5.2-15.2.1.1Compliance with 10 CFR 50.55a.............
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................5.2-15.2.1.2Applicable Code Cases......
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.....................5.2-15.2.2OVERPRESSURE PROTECTION.
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.....................5.2-25.2.2.2Design Evaluation............
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.....................5.2-35.2.2.3Piping and Instrumentation Diagrams............
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.......................5.2-45.2.2.5Mounting of Pressure-Relief Devices...........
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.......................5.2-45.2.2.6Applicable Codes and Classification.......
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................5.2-75.2.2.7Material Specifications.....
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.....................5.2-85.2.2.9System Reliability...........
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.....................5.2-85.2.2.10RCS Pressure Control During Low Temperature Operation................5.2-85.2.2.11Testing and Inspection.....
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.....................5.2-125.2.3MATERIALS SELECTION, FABRICAT ION, AND PROCESSING..........5.2-135.2.3.1Material Specifications...
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.....................5.2-135.2.3.2Compatibility With Reactor Coolant.......
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.................5.2-135.2.3.3Fabrication and Processing of Ferritic Materials......
..........................5.2-165.2.3.4Fabrication and Proce ssing of Austenitic Stainless Steel..................5.2-17 CALLAWAY - SPTABLE OF CONTENTS (Continued)
Section Page5.0-ii5.2.4INSERVICE INSPECTION AND TESTING OF THE REACTOR COOLANT PRESSURE BOUNDARY....
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.....................5.2-235.2.4.1Inspection of Class I Components..............
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........................5.2-245.2.4.2Arrangement and Accessibility................
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..........................5.2-245.2.4.3Examination Techniques and Procedures.............
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.................5.2-275.2.4.4Inspection Intervals........
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.....................5.2-295.2.4.5Examination Categories and Requirements.............
..........................5.2-295.2.4.6Evaluation of Examination Results.............
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........................5.2-305.2.4.7System Leakage and Hydrostatic Tests.........
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.....................5.2-305.2.5REACTOR COOLANT PRESSURE BOUNDARY LEAKAGE DETECTION SYSTEMS...........
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.....................5.2-305.2.5.1Design Bases.......
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.....................5.2-315.2.5.3Safety Evaluation...........
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.....................5.2-405.2.5.4Tests and Inspections....
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.....................5.2-405.2.5.5Instrumentation Applications...................
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..........................5.2-415.
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.....................5.2-415.3REACTOR VESSEL...........
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.....................5.3-15.3.1REACTOR VESSEL MATERIALS.
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...................5.3-15.3.1.1Material Specifications.....
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.....................5.3-15.3.1.2Special Processes Used for Manufacturing and Fabrication................5.3-15.3.1.3Special Methods for Nondestructive Examination......
..........................5.3-25.3.1.4Special Controls for Ferritic and Aust enitic Stainless Steels
................5.3-35.3.1.5Fracture Toughness...........
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.....................5.3-45.3.1.6Material Surveillance........
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.....................5.3-45.3.1.7Reactor Vessel Fasteners.....
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.....................5.3-85.3.2PRESSURE - TEMPERATURE LIMITS............
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.......................5.3-85.3.2.1Limit Curves.......
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.....................5.3-95.3.3REACTOR VESSEL INTEGRITY.
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.....................5.3-9 CALLAWAY - SPTABLE OF CONTENTS (Continued)
Section Page 5.0-iii5.3.3.2Materials of Construction
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.....................5.3-115.3.3.3Fabrication Methods.......
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.....................5.3-115.3.3.4Inspection Requirements....
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.....................5.3-115.3.3.5Shipment and Installation.............
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.....................5.3-115.3.3.6Operating Conditions......
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.....................5.3-135.
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.....................5.3-155.4COMPONENT AND SUBSYSTEM DESIGN...............
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.....................5.4-15.4.1REACTOR COOLANT PUMPS...
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................5.4-15.4.1.1Design Bases.......
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.....................5.4-85.4.1.5Pump Flywheels......
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.....................5.4-105.4.2.1Design Bases.......
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.....................5.4-115.4.2.3Steam Generator Materials.................
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...................5.4-135.4.2.4Steam Generator Inservice Inspection..................
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.....................5.4-175.4.2.6Quality Assurance..........
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.....................5.4-195.4.3REACTOR COOLANT PIPING...
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.....................5.4-215.4.3.3Design Evaluation..........
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.....................5.4-245.4.3.4Tests and Inspections....
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.....................5.4-26 CALLAWAY - SPTABLE OF CONTENTS (Continued)
Section Page 5.0-iv5.4.5MAIN STEAM LINE ISOLATION SYSTEM.........
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.....................5.4-265.4.6REACTOR CORE ISOLATION COOLING SYSTEM.........
.....................5.4-265.4.7RESIDUAL HEAT REMOVAL SYSTEM..............
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...................5.4-375.4.7.4Preoperational Testing.....
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.....................5.4-375.4.7.5Gas Management......
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........................5.4-375.4.8REACTOR WATER CLEANUP SYSTEM..............
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...................5.4-385.4.9MAIN STEAM LINE AND FEED WATER PIPING..................
.................5.4-385.4.10PRESSURIZER.........
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.....................5.4-415.4.10.4Tests and Inspections....
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.....................5.4-425.4.11PRESSURIZER RELIEF DISCHARGE SYSTEM...........
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.....................5.4-455.4.11.4Instrumentation Requirements..............
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.....................5.4-465.4.12VALVES............
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.....................5.4-475.4.13SAFETY AND RELIEF VALVES.................
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.....................5.4-47 CALLAWAY - SPTABLE OF CONTENTS (Continued)
Section Page 5.0-v5.4.13.3Design Evaluation..........
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.....................5.4-485.4.13.4Tests and Inspections....
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.....................5.4-485.4.14COMPONENT SUPPORTS.....
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.....................5.4-525.4ASAFE SHUTDOWN.........
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.....................5.4A-15.4A.1INTRODUCTION.......
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.....................5.4A-15.4A.3SAFE SHUTDOWN SCENARIO..
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..........................5.4A-25.4A.3.1MaintainaHotStandbyCondition...........
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..........................5.4A-25.4A.3.2Achieve and Maintain Cold Shutdown........
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........................5.4A-5 CALLAWAY - SP5.0-viRev. OL-1412/04LIST OF TABLES NumberTitle5.1-1System Design and Operating Parameters 5.2-1Applicable Code Adde nda for Reactor Coolant System Components5.2-2Class 1 Primary Components Material Specifications5.2-3Class 1 and 2 Auxiliary Comp onents Material Specifications5.2-4Reactor Vessel Internals for Emergency Core Cooling Systems5.2-5Recommended Reactor Coolant Water Chemistry Limits 5.2-6Design Comparison with Regulatory Guide 1.45, Dated May 1973, Titled Reactor Coolant Pressure Bo undary Leakage Detection Systems5.3-1Reactor Vessel Quality Assurance Program5.3-2Reactor Vessel Design Parameters5.3-3Deleted.
5.3-4Callaway Unit 1 Reacto r Vessel Material Properties5.3-5Deleted.5.3-6Deleted.
5.3-7Callaway Unit 1 Reactor Vessel Closu re Head Bolting Material Properties5.3-8Deleted.5.3-9Callaway Reactor Vessel Values fo r Analysis of Potential Pressurized Thermal Shock Events 10CFR50.615.3-10Reactor Vessel Material Survei llance Program - Withdrawal Schedule5.3-11Neutron Dosimetry Reactions of Interest5.4-1Reactor Coolant Pu mp Design Parameters5.4-2Reactor Coolant Pump Quality Assurance Program CALLAWAY - SPLIST OF TABLES (Continued)
NumberTitle5.0-viiRev. OL-1412/045.4-3Steam Generator Design Data5.4-4Steam Generator Q uality Assurance Program5.4-5Reactor Coolant Pi ping Design Parameters5.4-6Reactor Coolant Piping Quality Assurance Program5.4-7Design Bases for Residual Heat Removal S ystem Operation5.4-8Residual Heat Removal System Component Data5.4-9Failure Modes and Effects Analysis - Residual Heat Removal System Active Components - Plant Cooldown Operation5.4-10Pressurizer Design Data5.4-11Reactor Coolant System Design Pressure Settings5.4-12Pressurizer Quality Assurance Program5.4-13Pressurizer Relief Tank Design Data5.4-14Relief Valve Discharge to the Pressurizer Relief Tank5.4-15Reactor Coolant Syst em Valve Design Parameters5.4-16Reactor Coolant System Valves Nondestructive Examination Program5.4-17Pressurizer Valves Design Parameters 5.4A-1Design Comparison to Regulatory Positions of Regulatory Guide 1.139 Rev 1, Draft 2 Dated February 25, 1980 Titled "Guid ance for Residual Heat Removal to Achieve and Maintain Cold Shutdown5.4A-2Design Comparison of Table 1 of BTP RSB 5-1 fo r Possible Solutions for Full Compliance5.4A-3ResidualHeatRemoval-Safe ty Related Cold Shutdown Operations-FailureModesandEffectsAnalysis(FMEA)
CALLAWAY - SP5.0-viiiRev. OL-15 5/06LIST OF FIGURES NumberTitle5.1-1(4 Sheets)Reactor Coolant System 5.1-2Reactor Coolant Syst em Process Flow Diagram5.2-1Installation Detail for the Main Steam Pressure Relief Devices5.2-2Primary Coolant Leak Detection Response Time5.3-1Reactor Vessel5.4-1Reactor Coolant Controlled Leakage Pump5.4-2Reactor Coolant Pump Estima ted Performance Characteristic5.4-3Areva Model 73/1 9T Steam Generator5.4-4Deleted 5.4-5Deleted5.4-6Deleted5.4-7Residual Heat Removal System 5.4-8Residual Heat Removal S ystem Process Flow Diagram5.4-9Normal Residual Heat Removal Cooldown5.4-10Single Residual Heat Removal Train Cooldown5.4-11Pressurizer5.4-12Pressurizer Relief Tank5.4-13Reactor Vessel Supports 5.4-14Steam Gener ator Supports5.4-15Reactor Cool ant Pump Supports5.4-16Reactor Building Inte rnals Pressurizer Supports CALLAWAY - SP LIST OF FIGURES (Continued)
NumberTitle5.0-ixRev. OL-15 5/065.4-17Pressurizer Supports5.4-18Crossover Leg Restraint 5.4-19Deleted5.4-20Hot Leg Restraint5.4-21Hot and Cold Leg Lateral Restraints CALLAWAY - SP5.1-1Rev. OL-15 5/06CHAPTER 5.0REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS5.1
SUMMARY
DESCRIPTION5.1.1DESIGN BASES The performance and safety design bases of the reactor coolant system (RCS) and its major components are interrelated. T hese design bases ar e listed below: a.The RCS has the capabi lity to transfer to the steam and power conversion system the heat produced du ring power operation a nd when the reactor is subcritical, including the init ial phase of plant cooldown. b.The RCS has the capability to transfer to the residual heat removal system the heat produced during the subsequent phase of plant cooldown and cold shutdown. c.The RCS heat removal capabili ty under power operation and normal operational transients, including the transition from forced to natural circulation, assures no fuel dam age within the operat ing bounds permitted by the reactor control and protection systems. d.The RCS provides the water used as the core neutron moderator and reflector and as a solvent for chemical shim control. e.The RCS maintains the homogeneity of the soluble neutron poison concentration and the rate of change of the cool ant temperature, so that uncontrolled reactivity changes do not occur. f.The RCS pressure boundary is capable of a ccommodating the temperatures and pressure s associated with operational transients. g.The reactor vessel supports the r eactor core and co ntrol rod drive mechanisms. h.The pressurizer maintains the system pressure during operation and limits pressure transients. During the reduc tion or increase of plant load, the pressurizer accommodates volume changes in the reactor coolant. i.The reactor coolant pumps supply the coolant flow necessary to remove heat from the reactor co re and transfer it to the steam generators.
CALLAWAY - SP5.1-2Rev. OL-15 5/06j.The steam generators provide high qualit y steam to the turbine. The tube and tubesheet boundary ar e designed to preven t the transfer of radioactivity generated wi thin the core to the secondary system. k.The RCS piping contains the c oolant under operating temperature and pressure conditions and limits leakage (and activi ty release) to the containment atmosphere. The RCS piping contains demineralized borated water which is circulated at the flow rate and temp erature consistent with achieving the reactor core ther mal and hydraulic performance. l.The RCS is monitored for loose parts, as described in Section 4.4.6. 5.1.2DESIGN DESCRIPTION The RCS, shown in Figure 5.1-1, consists of four similar heat transfer loops connected in parallel to the reactor pressu re vessel. Each loop cont ains a reactor coolant pump, steam generator, and asso ciated piping and valves. In addition, the system includes a pressurizer, pressurizer relief and sa fety valves, interconnecting piping, and instrumentation necessary for operational control. All the above components are located in the containment building.
During operation, the RCS transfers the heat generated in the co re to the steam generators where steam is produced to drive the turbine generator. Borated demineralized water is circulated in the RCS at a flow rate and temperature consistent with achieving the reactor core thermal-hydraulic performance. The water also acts as a neutron moderator and reflecto r and as a solvent for t he neutron absorber used in chemical shim control.
The RCS pressure boundary is a barrier against the releas e of radioactivity generated within the reactor, and is designed to ensure a high degree of integrity throughout the life of the plant.
RCS pressure is controlled by the use of the pressurize r where water and steam are maintained in equilibrium by electrical heaters and water sprays. Steam can be formed (by the heaters) or condensed (by the pressurizer spray) to minimize pressure variations due to contraction and expansion of the r eactor coolant. Spring-loaded safety valves and power-operated relief valves from the pressurizer prov ide for steam discharge from the RCS. Discharged st eam is piped to the pressurizer relief tank, wher e the steam is condensed and cooled by mixing with water.
The extent of the RCS is defined as: a.The reactor vessel, including co ntrol rod drive mechanism housingsb.The portion of the steam generators containi ng reactor coolant CALLAWAY - SP5.1-3Rev. OL-15 5/06c.Reactor coolant pumpsd.The pressurizere.Safety and relief valvesf.The interconnecting piping, valves, and fittings between the principal components listed aboveg.The piping, fittings, and valves l eading to connecting auxiliary or support systems up to and including the seco nd isolation valve (from the high pressure side) on each line The RCS is shown schematically in Figure 5.1-2. Included on this figure is a tabulation of principal pressures and temperratures and th e flow rate of t he system under normal steady state full power operating conditions. These parameters ar e based on the best estimate flow at the pump discharge.
RCS volume under t he above conditions is presented in Table 5.1-1. A piping and instrumentation diag ram of the RCS is shown in Figure 5.1-1 (Sheets 1 through 4). The diagrams show the extent of the systems located within the containment and the points of separati on between the RCS and the seco ndary (heat utilization) system. Figures 1.2-9 through 1.2-18 provide plan and elevati on views of the reactor building. These figures show principal dimensions of reactor coolant system components in relationship with supporting and surrou nding steel and concrete structures and demonstrate the pr otection provided to the reactor coolant system by its physical layout. 5.1.3SYSTEM COMPONENTS The major components of the RCS are as follows: a.Reactor vessel The reactor vessel is cyl indrical and has a wel ded, hemispherical bottom head and a removable, flanged, hemis pherical upper head. The vessel contains the core, core-supporting structures, control rods, and other parts directly associated with the core.
The vessel has inlet and outlet nozzles located in a horizontal plane just below the reactor vessel flange but above the top of the core. Coolant enters the vessel through the inlet nozzles and flows down the core barrel-vessel wall annulus, turns at the bottom, and fl ows up through the core to the outlet nozzles. b.Steam generators CALLAWAY - SP5.1-4Rev. OL-15 5/06 The steam generators are vertical shell and U-tube evaporators with integral moisture separating equipment. The reactor coolant flows through the inverted U-tubes, entering and leav ing through the nozzles located in the hemispherical botto m head of the steam generator. Steam is generated on the shell si de and flows upward through the moisture separators to the outlet nozzle at the top of the vessel. The steam generator design is designated by Areva as Model 73/19T. c.Reactor coolant pumps The reactor coolant pumps are single speed centrifugal units driven by air-cooled, three-phase induction motors. Heat from the air-cooling system is rejected to the component cooling water. The shaft is vertical with the motor mounted above the pu mp. A flywheel on the shaft above the motor provides additional inertia to extend pump coastdown.
The flow inlet is at the bottom of the pump, and the discharge is on the side. d.Piping The reactor coolant piping is seamless stainless steel piping.
The hot leg is defined as the piping between the reactor vesse l outlet nozzle and the steam generator. The cold leg is defined as the piping between the reactor coolant pump outlet and the reactor vesse
- l. The crossover leg is defined as the piping between the steam generator and the reactor coolant pump inlet. e.Pressurizer The pressurizer is a vertical, cylindrical vessel wit h hemispherical top and bottom heads. Electrical heaters are installed thr ough the bottom head of the vessel while the spray nozzle and relief and sa fety valve connections are located in the top head of the vessel. f.Safety and relief valves The pressurizer safety valves are of the totally enclosed pop-type. The valves are spring loaded and self activated with backpressure compensation. The power-operated relief valves ha ve electric solenoid actuators. They are ope rated automatically based on RCS pressure or by remote manual control.
Remotely operated valves ar e provided to isolate the inlet to the power-operat ed relief valves if exce ssive leakage occurs. Steam from the pressurizer safety and relief valves is discharged into the pressurizer relief tank through a sparger pipe under the water level. This condenses and cools the steam by mixing it with water that is near ambient temperature.
CALLAWAY - SP5.1-5Rev. OL-15 5/065.1.4SYSTEM PERFORMANCE CHARACTERISTICS Design and performance characteristics of t he RCS are provided in Table 5.1-1. a.Reactor coolant flow The reactor coolant flow, a major parameter in the desi gn of the system and its components, is established with a detailed design procedure supported by operating plant performance data, by pump model tests and analysis, and by pressure drop tests and analyses of the reactor vessel and fuel assemblies. Data from all operating plants have indicated that the actual flow has been well above the flow specified for the thermal design of the plant. By applying the design procedure described below, it is possible to specify the expect ed operating flow with reasonable accuracy.
Three reactor coolant flow rates are identified for the va rious plant design considerations. The de finitions of these flow s are presented in the following paragraphs. b.Best estimate flow The best estimate flow is the most likel y value for the actual plant operating condition. This flow is based on the best estimate of the flow resistances in the reactor vessel, steam generator, and piping and on the best estimate of the reactor coolant pump head-flow capacity, with no uncertainties assigned to either the system flow resistance or the pump head. System pressure drops, based on best estimate flow with the replacement steam generator, are presented in Table 5.1-1. Although the best estimate flow is the most likely value to be expected in operation, more conserva tive flow rates are ap plied in the thermal and mechanical designs. c.Thermal design flow Thermal design flow is the flow rate used as a basis for the reactor core thermal performance, t he steam generator ther mal performance, and the nominal plant parameters used throughout the design. The thermal design flow accounts for the uncertainties in flow resistances (reactor vessel, steam generator, and piping), reactor coolant pump head, and the methods used to measure flow rate. The thermal design fl ow is approximately 9.6 percent less than the best estimate flow with 5% st eam generator tube plugging (SGTP). The ther mal design flow is confir med when the plant is placed in operation. Tabulations of import ant design and performance CALLAWAY - SP5.1-6Rev. OL-15 5/06 characteristics of the RCS, as provided in Table 5.1-1 , are based on the thermal design flow. d.Mechanical design flow Mechanical design flow is a conservatively high flow used in the mechanical design of the reactor vessel internals and fuel assemblies. The mechanical design flow is based on a reduced system resistance and on increased pump head capability. The mechanical design flow was increased to 109,200 gpm per loop with the RSGs and is approximately 3.8 percent greater than the best estimate flow at 0% SGTP.
Pump overspeed due to a turbine gen erator overspeed of 20 percent results in a peak reacto r coolant flow of 120 per cent of the mechanical design flow. The overspeed condit ion is applicable on ly to operating conditions when the reac tor and turbine generator are at power.
CALLAWAY - SP Rev. OL-21 5/15TABLE 5.1-1 SYSTEM DESIGN AND OPERATING PARAMETERSPlant design life, years40Nominal operating pressure, psig2,235 Nominal total system volume, including pressurizer and surge line, ft 3**This nominal volume w ill change dependent on SG tube pluggi ng (assumes 3% thermal expansion)13,903 (+/-100 ft 3 at a nominal T avg of 557°F)Nominal system liquid volume, including pressurizer water at maximum guaranteed power, ft 3 13,061 Pressurizer spray rate, maximum, gpm900Pressurizer heater capacity, kW1,800System Thermal and Hydraulic Data4 PumpsRunningNSSS power, MWt 3,579Reactor power, MWt 3,565 Thermal design flows, gpm Per loop Reactor 93,600 374,400Total reactor flow, 10 6 lb/hr 139.4**Based on T in = 556.8°F and pres surizer pressure = 2250 psia. See Tables 1.3-1 , 4.1-1 , and 4.4-1.**Temperatures,°F Reactor vessel outlet
Reactor vessel inletSteam generator outletSteam generator steam
Feedwater**Temperatures based on a 0% tube plugging level in the steam generators, thermal design flow (non-RTDP).
620.0***
556.8*** 556.6 547.2****
390.0 to 446.0***Operation with RCS T avg reduced as low as 570.7°F has been evaluated to meet all criteria for acceptable plant operation with the RSGs installed.
CALLAWAY - SPTABLE 5.1-1 (Sheet 2)
Rev. OL-21 5/15****528.3°F based on T avg of 570.7°F and 0% equiva lent SG tube plugging.System Thermal and Hydraulic Data4 PumpsRunning TAVG=588.4°F4 Pumps Running TAVG=570.7°F****Steam pressure, psia* 1022** 872**Total steam flow, 10 6 lb/hr* 15.96*** 15.85***Best estimate flows, gpm*
Per loop Reactor 104,300 417,200 105,200 420,800 Mechanical desi gn flows, gpm Per loop Reactor 109,200 436,800 109,200 436,800System Pressure Drops*
Reactor vessel P, psiSteam generator P, psi Hot leg piping P, psi Crossover leg piping P, psi Cold leg piping P, psiPump head, ft 52.6 31.9 1.3 3.4 3.6 289 52.1 33.1 1.4 3.6 3.8 285*Based on 0% equivalent steam generator tube plugging.**1016 psia with 5% SGTP at Tavg = 588.4°F; 867 psia with 5% SGTP at T avg = 570.7°F ***15.95 x 10 6 lb/hr with 5% SGTP at T avg = 588.4°F; 15.84 x 10 6 lb/hr with 5% SGTP at T avg = 570.7°F****Flows and pressure drops at 570.7°F assumes thimbl e plugs are removed from core to yield maximum results CALLAWAY - SP5.2-1Rev. OL-21 5/155.2INTEGRITY OF REACTOR COOLANT PRESSURE BOUNDARY This section discusses the measures employed to provide and maintain the integrity of the reactor coolant pressure boundary (RCPB) for the plant design lifetime. Section 50.2 of 10 CFR 50 defines the RCPB as extending to the outermost containment isolation valve in system piping which penetrates the containment and is connected to the RCS. This section is limited to a description of the component s of the RCS as defined in Section 5.1 , unless otherwise noted. Components which are part of the RCPB (as defined in 10 CFR 50) but are not described in this section are described in the following sections:
a.Section 6.3 - RCPB components which are part of the emergency core cooling system.
b.Section 9.3.4 - RCPB components which are part of the chemical and volume control system.
c.Section 3.9(N).1 - Design loadings, stress limits, and analyses applied to the RCS and ASME Code Class 1 components.
d.Section 3.9(N).3 - Design loadings, stress limits, and analyses applied to ASME Code Class 2 and 3 components.
The phrase RCS, as used in this section, is as defined in Section 5.1. When the term RCPB is used in this section, its definition is that of Section 50.2 of 10 CFR 50. 5.2.1COMPLIANCE WITH CO DES AND CODE CASES 5.2.1.1Compliance with 10 CFR 50.55aRCS components are desi gned and fabricated in accordan ce with 10 CFR 50, Section 50.55a, "Codes and Standards." The addenda of the ASME Code applied in the design of each component are listed in Table 5.2-1. 5.2.1.2Applicable Code Cases Regulatory Guides 1.84, 1.
85 and 1.147 are discussed in Appendix 3A. Code Case 1528 (SA-508, Class 2a) material has been us ed in the manufac ture of the SNUPPS pressurizers. Re gulatory Guide 1.85 presently reflects a conditional NRC approval of Code Case 1528. Westinghouse has conducte d a test program which demonstrates the adequa cy of Code Case 1528 material. The results of the test program are documented in Refe rence 1. Reference 1 and a request for approval (Ref.
- 2) of the use of Code Case 1528 have been subm itted to the NRC.
CALLAWAY - SP5.2-2Rev. OL-21 5/15 The specific code cases used were: 5.2.2OVERPRESSURE PROTECTION RCS overpressure protection is accomplished by the utiliz ation of pressurizer safety valves along with the reactor prot ection system and as sociated equipment.
Combinations of these systems provide compliance with the overpressure requirements of the ASME Boiler and Pressure Vessel Code, Sect ion III, Paragraphs NB-7300 and NC-7300, for pressurized wa ter reactor systems. Auxiliary or emergency systems connected to the RCS are not utilized for the prevention of RCS overpressuriza tion protection. 5.2.2.1Design Bases Overpressure protection is pr ovided for the RCS by the pre ssurizer safety valves which discharge to the pressurizer relief tank by means of a common header. The transient which establishes the design requirements for the primary system overpressure protection is a complete loss of steam flow to the turbi ne with operation of the steam generator safety valves and ma intenance of main feedwater flow. However, for the sizing of the pressurizer safety valves, no credit is taken for reactor trip nor the operation of the following: a.Pressurizer power-operated relief valvesb.Steam line relief valvec.Steam dump systemd.Reactor control system e.Pressurizer level control system f.Pressurizer spray valveFor this transient, the peak RCS and peak steam system pressure must be limited to 110 percent of their respec tive design values. Steam Generator: 2142, 2143, N-20-3, and N-474-1Pressurizer: 1528 Piping: 1423-2 Valves: 1649, 1769, and N-3-10 CALLAWAY - SP5.2-3Rev. OL-21 5/15Assumptions for the overpressure analysis include: 1) the plant is operating at the power level corresponding to the engineered safeguards design rating and 2) the RCS average temperature and pressure are at their maximum values. These are the most limiting assumptions with respect to system overpressure.
Overpressure protection for the steam system is provided by steam generator safety valves. The steam system safety valve capa city is based on providing enough relief to remove 105 percent of the engi neered safeguards design steam flow. This must be done by limiting the maximum st eam system pressure to less than 110 percent of the steam generator shell si de design pressure. Blowdown and heat dissipati on systems of the NSSS conn ected to the discharge of these pressure relieving devices are discussed in Section 5.4.11. Steam generator blowdown systems for the balance-of-plant are discussed in Section 10.4.8. Postulated events and transients on which the design requirements of the overpressure protection system are based and discuss ed in Reference 3 and Reference 3a. 5.2.2.2Design EvaluationThe relief capacities of the pressurizer and steam gener ator safety valves are determined from the postulated overpressure transient conditions in conjunction with the action of the reactor protecti on system. An evaluation of t he functional design of the system and an analysis of the capability of the system to perform its function are presented in Reference 3 and Reference 3a.
The report describes in detail the types and number of pressure relief devices employed, relief devic e description, locations in the systems, reliability history, and the details of the methods used for relief device sizing based on typical worst-case transient condi tions and analysis dat a for each transient condition. The description of the analytical model us ed in the anal ysis of the overpressure protection system and the basis for its validity are discussed in Reference4. A description of the pressurizer safety valves performance characteristics along with the design description of the incidents, assumptions made , method of analysis, and conclusions are discussed in Chapter 15.0. 5.2.2.3Piping and Instrumentation Diagrams Overpressure protection for t he RCS is provided by pressuri zer safety valv es shown in Figure 5.1-1 , Sheet 2.
These discharge to the pressurizer relief tank by means of a common header.
CALLAWAY - SP5.2-4Rev. OL-21 5/15The steam system safety valves are discussed in Section 10.3 and are shown on Figure 1.2-155.2.2.4Equipment and Component DescriptionThe operation, significant design parameters, number and types of operating cycles, and environmental conditions of the pressuri zer safety valves are discussed in Sections 5.4.13 , 3.9(N).1 , and 3.11(N). Section 10.3 contains a discussion of the equipment and components of the steam system overpressure system. 5.2.2.5Mounting of Pr essure-Relief Devices5.2.2.5.1Location of Pr essure Relief Devices The design bases for the assu med loads for the primary a nd secondary side pressure relief devices of the steam generator are described in Paragraph 3.9(B).3.3.
Figure 5.2-1 provides design and installation details for the pressure reli ef devices mounted on the secondary side of the steam generator.
Pressure relief devices for the reactor coolant system are three pre ssurizer safety relief valves and two power-operated relief valves. These valves discharge to the pressurizer relief tank via a common header. 5.2.2.5.2Pressurizer Safety Relief Valves The pressurizer safely valve discharge pipi ng system is a closed system in which no sustained reaction force from a free discharging jet of fluid can exist. However, transient hydraulic forces are imposed at various points in the piping system from the time a safety valve begins to open until a stea dy flow is completely dev eloped. Since a water loop seal is applied, transient hydraulic forces caused by the liquid being forced through the safety valve and then acce lerated down the piping system does occur.
The pressurizer relief devices are mounted and installed as follows: a.Each straight leg of t he discharge pipe is supported as necessary to take the valve discharge transient force along that leg. b.The supports at the valve discharge piping are connecte d to the adjacent structure. c.Snubbers are used to restrain the valve discharge transient forces as necessary when thermal movements are of a high magnitude. Thermal hydraulic analysis was performed using the method of characteristics approach to generate fluid parameters as a function of time, includi ng provisions for analysis of valve opening and clos ing situations.
CALLAWAY - SP5.2-5Rev. OL-21 5/15 Unbalanced forces were calculated for each straight segment of pipe from the pressurizer to the relief tank. The time hi stories of thes e forces were used for the subsequent structural anal ysis of the pressurizer safety and relief lines.Hydraulic forcing functions were generated assuming the simultaneous opening of either the safety valves or the relief valves and included water discharge transients when the relief valves were utilized for cold overpressure mitigation.A dynamic analysis using computer program PS + CAEPIPE was performed to verify the design of the support configuration. The results of these analyses are described below: a.For loading combinations see Table 3.9(B)-2. b.Material Typec.Maximum stress points within piping system5.2.2.5.3Main Steam Safety Relief Valves Figure 5.2-1 provides design and installation details. Class I Piping 3" Sc
- h. 160, SA-312, TP-304 6" Sch. 160, SA-312, TP-304B31.1 Piping3" Sch.
80S, SA-312, TP-304 6" Sch. 80S, SA-312, TP-304 12" Sch. 80S, SA-312, TP-304Class I PipingNode point - 4690 Type - reducer Max. primary stress20.9 ksi Allowable primary stress24.2 ksiB31.1 PipingNode point - 5090 Max. primary stress21.8 ksi Allowable primary stress22.6 ksi
Node point - 4610
Max. primary + secondary stress43.3 ksi
Allowable primary +
secondary stress43.4 ksi CALLAWAY - SP5.2-6Rev. OL-21 5/15The steady-state flow condition reached af ter the valve has opened and is exhausting into the stack was considered in the stress analysis of the safety valve installation. With these conditions, the valve moments are balanced due to the split valve discharge design, and the vertical discharge thrust forc e is reacted by the header supports via the header. The discharge force from the vent stack is reacted by an in-line anchor and the supports near the top of the stack. The effects of thermal expansion, pipe weight, seismic anchor movements, seismic occurrence, and relief valve discharge thrust forces were considered in the stress analysis of the vent stack piping. These effects were also considered in the stress analysis of the main steam header piping in addition to the water hammer effects caused by fast valve closur e of the main steam isolation valves.
A 10 percent unbalanced discharge from the two split discharge ports of each safety valve was assumed for the stress analysis of the header piping.
Therefore, one discharge port had an assumed vertical thrust load of 13,574 pounds and the other an assumed thrust load of 12, 227 pounds. These values ar e based on a relief valve discharge from a line pressure of 1,185 psi and a dynamic load factor of 1.2. It was conservatively assumed t hat each valve opened simultaneously, resulting in the following header stresses and support loads:a.For loading combinations see Table 3.9(B)-2 and Table 3.9(B)-14. b.Material type 28-inch OD wall thickness of 1.5 inch, SA 106, Gr C.c.Maximum stress points within system Node point - 45 Node point - 5 d.Support loadsMaximum primary stress12,421 psi Allowable primary stress21,000 psi Maximum secondary stress16,131 psi Allowable secondary stress26,250 psi Node Point Header Support Loads (vertical supports and loads only)521,942 lbs33187,800 lbs CALLAWAY - SP5.2-7Rev. OL-21 5/155.2.2.6Applicable Codes and ClassificationThe requirements of ASME Boiler and Pressure Vessel Code,Section III, Paragraphs NB-7300 (Overpressure Pr otection Report) and NC-7300 (O verpressure Protection Analysis), are followed and co mplied with for pressuriz ed water reactor systems. Piping, valves, and associated equipment used for overpressure protection are classified in accordance with ANS-N18.2, "Nuclear Safe ty Criteria for the Design of Stationary Pressurized Water Reactor Plants." These sa fety class designations are delineated on Table 3.2-1 and shown on Figure 5.1-1. For further information, refer to Section 3.9(N). 5.2.2.7Material Specifications Refer to Section 5.2.3 for a description of material specifications. 83112,700 lbs85166,300 lbs30033,347 lbs294187,800 lbs282112,700 lbs 281166,300 lbs34733,362 lbs341187,800 lbs 329112,700 lbs328166,300 lbs39710,100 lbs 391184,400 lbs380112,800 lbs379166,300 lbs Node Point Header Support Loads (vertical supports and loads only)
CALLAWAY - SP5.2-8Rev. OL-21 5/155.2.2.8Process Instrumentation Each pressurizer safety valve discharge li ne incorporates a cont rol board temperature indicator and alarm to notify the operator of steam dischar ge due to eit her leakage or actual valve operation. Safe ty-related control room posi tive position indication is provided for the PORVs and safety valves. For a further discussion on process instrumentation associated with the system, refer to Chapter 7.0. 5.2.2.9System ReliabilityThe reliability of the pressure relieving devices is discussed in Section 4 of Reference 3. 5.2.2.10RCS Pressure Control During Low Temperature OperationAdministrative procedures are developed to aid the operator in controlling RCS pressure during low temperature operation. However, to provide a back-up to the operator and to minimize the frequency of RCS overpressurization, an automa tic system is provided to maintain pressures within allowable limits.
Analyses have shown that one pressurizer power-operated relief valve (PORV) or one RHR suction relief valve is sufficient to prevent violation of these limits due to anticipated mass and heat input transients. However, redundant pr otection against an overpressurization event is provided through the use of two pressurizer PORVs, two RHR suction relief valves, or one PORV and one RHR suction relief valve to mitigate any potential pressure transients.
The mitigation system is r equired during the Applicability of Technical Specification (TS) LCO 3.4.12, with RCP startup restrictions as governed by TS LCO 3.4.6 and TS LCO 3.4.7, when the PORVs are manually armed and automatically actuat ed and the RHR suct ion relief valves ar e made available by cross-connecting the RCS and RHR systems.5.2.2.10.1System Operation Two pressurizer power-operated relief valves ar e supplied with actuation logic in the cold overpressure mitigation system (COMS) to ensure that a redu ndant and independent RCS pressure control back-up feature is provided fo r the operator during low temperature operations. This system provides the capability for RCS inventory letdown, thereby maintaining RCS pressure within allowable limits. Refer to Sections 5.4.7 , 5.4.10 , 5.4.13 , 7.6.6 , and 9.3.4 for additional information on RCS pressure and inventory control during other m odes of operation.
The basic function of the system logic is to continuously monitor RCS temperature and pressure conditions whenever plant operation is at low te mperatures. An auctioneered low wide range RCS temperatur e will be contin uously converted to an allowable pressure and then compared to the actual wide r ange RCS pressure. The system logic will first annunciate a main control board alarm whenev er the measured pressure approaches within a pre-determined amount of the allowable pressure thereb y indicating CALLAWAY - SP5.2-9Rev. OL-21 5/15 that a pressure transient is occurring. On a further incr ease in measured pressure, an actuation signal is transmitted to open the pressurizer power-operated relief valves (if the COMS actuation logic has been manually armed) when required to mitigate the pressure transient.
When the RCS and RHR systems are cross-connected, th e RHR suction relief valves are available to maintain RCS pr essure within allowable limits.5.2.2.10.2Evaluation of Low Temperature Overpressure Transients The ASME Code (Section III, Appendix G) establishes guidelines and upper limits for RCS pressure primarily for lo w temperature conditions (<350°F). The cold overpressure mitigation system discussed in Section 5.2.2.10.1 addresses these conditions as discussed in the following paragraphs. Transient analyses hav e been performed to determine the maximum pressure for the postulated mass input and heat input events.
The mass input pressure transient which w ould occur most frequently during the course of normal plant operation would involve letdown isolation with charging pumps delivering an input less than or equal to 120 gpm. However, the mass input analysis has been performed assuming one safety-related ECCS charging pump and the non-safety related normal charging pump operating in a configuration produci ng maximum delivery rates.
This configuration is like the case of an inadvertent Safety Injection with flow from both pumps through the bor on injection header and the RCP seals, with a simultaneous loss of letdown.The heat input transient has been performed over the entire RCS shutdown temperature range. This analysis assumes an inadvertent reactor coolant pump startup with a 50°F mismatch between the RCS and the temperature of the ho tter secondary side of the steam generators. Both the heat input and mass input analyses take into account the single failure criterion and therefore, only one pressurizer power-operated relief valve or one RHR suction relief valve was assumed to be available for pressure relief. The above events have been evaluated considering the allowable pressure/temperature limits established by the Appendix G guidelines. The eval uation of the transient results conclude that reactor vessel integrity is not impaired. 5.2.2.10.3Operating Basi s Earthquake Evaluation A fluid systems evaluation has been performed consid ering the potential for overpressure transients following an operating bas is earthquake.
CALLAWAY - SP5.2-10Rev. OL-21 5/15The Callaway pressurizer power-operated relief valves and RHR suction relief valves have been designed in accordance with the ASME Code and seismically qualified under the Westinghouse valve operability program whic h is discussed in Section 3.9(N).3.2. Therefore, the cold ov erpressure mitigation system will be available to provide pressure relief following an operati ng basis earthquake. 5.2.2.10.4Administr ative Procedures Although the system described in Section 5.2.2.10.1 is installed to maintain RCS pressure within allowable limits, administrative procedures minimize the potential for occurence and the consequences of any transie nt that could actuate the cold over pressure mitigation system. The following discussion highlights these procedural controls, listed in hierarchy of their function in mitigating RCS cold overpressurization transients. Of primary importance is the basic method of operation of the pl ant. Normal plant operating procedures will maxi mize the use of a pressuri zer cushion (steam bubble) during periods of low pressure , low temperature operation. This cushion will dampen the plants' response to potential transient generating inputs, providing easier pressure control with the slow er response rates. An adequate cushion substantiall y reduces the severity of potential pressure transients, such as reactor coolant pump induced heat input , and slows the rate of pressure rise for others. In conjunction with the alarms discussed in Section 7.6.6 , this provides reasonable assurance that most potential transients can be terminated by operator action before the overpressure relief system actuates. 5.2.2.10.4.1Overpressurization TransientHowever, for those modes of operation when water solid operation may still be possible, procedures will further highlight precautions that minimize the severity of, or the potential for, developing an over pressurization transient. The following prec autions or measures are considered in developing the operating procedures: a.The residual heat remova l inlet lines from the reactor coolant loop are normally open when the RCS pressure is less than 400 psi (subject to procedural requirements related to preventing void formation associated with exceeding saturation conditions in the RHR su ction lines from the RWST). This precaution assures that there is a relief path from the reactor coolant loop to the residual heat removal suction line relief valves when the RCS is at low pressure and is water solid. b.Whenever the plant is water solid and the reactor coolant pressure is being maintained by the low pressure letdown control valve, letdown flow CALLAWAY - SP5.2-11Rev. OL-21 5/15normally bypasses the normal letdown or ifices. In addition, all three letdown orifices norma lly remain open. c.If all reactor coolant pumps have stopped for mo re than 5 minutes during plant heatup and the reactor coolant te mperature is greater than the charging and seal injection water temperature, a steam bubble is procedurally required to be formed in the pressurizer prior to restarting a reactor coolant pump. Th is precaution minimizes the pressure transient when the pump is start ed and the cold water prev iously injected by the charging pumps is circulated thr ough the warmer reactor coolant components. The steam bubble will accommodate the resultant expansion as the cold water is rapidly warmed. d.If all reactor coolant pumps are stopped and the RCS is being cooled down by the residual heat e xchangers, a nonuniform te mperature distribution may occur in the reactor coolant loops. Prior to restarting a reactor coolant pump, the RCP startup restrictions in TS LCOs 3.4.6 and 3.4.7 shall be met. If RCS temperature has been reduced by more than 20°F by RHR, a steam bubble is procedurally required to be formed in the pressurizer prior to restarting a RCP. e.During plant cooldown, all steam generators wi ll normally be connected to the steam header to assu re a uniform cooldown of the reactor coolant loops. f.At least one reacto r coolant pump will normally remain in service until the reactor coolant temperature is reduced to 160°F.
These special precautions ba ck-up the normal operational m ode of maximizing periods of steam bubble operati on so that cold overpressure transient prevention is continued during periods of transitional operations. These precautions do not apply to reactor coolant system hydrostatic testing.5.2.2.10.4.2Cold Overpressurization Transients The specific plant configurations of emergency core cool ing system testing and alignment will also hi ghlight procedural re commendations to prev ent developing cold overpressurization transients.
During these lim ited periods of pl ant operation, the following precautions/measures are considered in developing the operating procedures: a.To preclude inadvertent emergency core cooling system actuation during heatup and cooldown, procedures requ ire blocking t he pressurizer pressure and low steam line pressure safety injection signal actuation logic at approximately 1,900 psig (below the P-11 in terlock setpoint).
CALLAWAY - SP5.2-12Rev. OL-21 5/15b.During further cooldown, the accumulator isolatio n valves are closed and power is disconnected at the MCC af ter the RCS has been depressurized below 1,000 psig, providing additional back-up to Item a above. One ECCS centrifugal chargi ng pump and both safety injection pumps are rendered incapable of injecting into the RCS prior to entering TS 3.4.12 Applicability.However, two ECCS centrifugal charging pumps may be capable of injecting for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for pump swap operations throughout TS 3.4.12 Applicability. With t he reactor coolant water le vel below the top of the reactor vessel flange and with the vesse l head on in MO DES 5 and 6, the Safety Injection pumps may be operated to mitigate the effects of a loss of decay heat removal during pa rtially drained conditions.c.The recommended procedure for periodic emergency core cooling system pump performance testing will be to test the pumps during normal power operation or at hot shutdow n conditions. This pr ecludes any potential for developing a cold overpr essurization transient. Should cold shutdown (CSD) testing of the pumps be desired, the test will be done when the vessel is open to atmosphere, again precluding overpressurization potential.
If CSD testing with the vessel closed is necessary, the procedures require emergency core cooling system pumps discharge valve closure and RHRS alignment to isolate potential emer gency core coolin g system pump input and to provide back-up benefit of the RHRS relief valves. d.SIS circuitry testing, if done during CSD, requir es RHRS alignment and steps to preclude RCS injection from one ECCS c entrifugal charging pump and both safety injection pumps to preclude de veloping cold overpressurization transients.
The above procedural precautions covering normal operations with a steam bubble, transitional operations where potentially water solid, and s pecific testing operations provide in-depth cold overpressure preventi ons or reductions, augmenting the installed overpressure relief system. Refer to Technical Specification 3.4.12.5.2.2.11Testing and InspectionTesting and inspection of t he overpressure protection components are discussed in Section 5.4.13.4
.
CALLAWAY - SP5.2-13Rev. OL-21 5/155.2.3MATERIALS SELECTION, FABRICATION, AND PROCESSING 5.2.3.1Material Specifications Material specifications used for the principal pressure retaining applications in components of the RCP B are listed in Table 5.2-2 for ASME Class 1 primary components and Table 5.2-3 for ASME Class 1 and 2 auxiliary components. Tables 5.2-2 and 5.2-3 also include the mate rial specifications of unstabilized austenitic stainless steel used for components in systems required for reactor shutdown and for emergency core cooling.
The material specifications of unstabilized austenitic stainless steel used for reactor vessel internals which are essential for emer gency core cooling and for core structural support are listed in Table 5.2-4. Table 5.2-3 is not totally inclusive of the mate rial specifications used in the listed applications. However, the listed sp ecifications are representative.
The materials utilized conform to the applicable ASME Code rules. The welding materials used for joining the ferritic base materials of the RCPB conform to or are equivalent to ASME Material Specifications SFA 5.1, 5.2, 5.5, 5.17, 5.18, and 5.20.
They are qualified to the requirements of the ASME Code,Section III.
The welding materials used for joining the austenitic stainless steel base materials of the RCPB conform to ASME Material Specifications SFA 5.4, 5.9, and 5.22.
They are qualified to the requirements of the ASME Code,Section III.
The welding materials used for joining nickel-chromium-iron alloy in similar base material combination and in dissimi lar ferritic or austenitic base material combination conform to ASME Material Specifications SFA 5.11 and 5.14. They are qualified to the requirements of the ASME Code,Section III. 5.2.3.2Compatibility With Reactor Coolant5.2.3.2.1Chemistry of Reactor Coolant The RCS chemistry specificati ons are listed in Chemistry Department Procedures, and are based on industr y guidelines (such as the EPRI PWR Primary Water Chemistry Guidelines and the Westinghouse Guidelines for Primary Water Chemistry), Technical Specifications, Table 5.2-5 and FSAR Ch. 16.
The RCS water chemistry is selected to minimize corrosion. Routinely scheduled analyses of the coolant chemical composition are performed to verify that the reactor coolant chemistry meets the specifications.
CALLAWAY - SP5.2-14Rev. OL-21 5/15 The chemical and volume control and RHR systems provide a means for adding chemicals to the RCS wh ich perform the following function s: 1) control the pH of the coolant during pre-startup testing and subs equent operation, 2) scavenge oxygen from the coolant during h eatup, and 3) control radiolysis reactions involving hydrogen, oxygen, and nitrogen during all power operations subsequent to startup. The normal limits for chemical additives and reactor coolant impurities for power operation are shown in Table 5.2-5. The pH control chemical uti lized is lithium hydroxide monohydrate, enriched in the lithium-7 isotope to 99.9 percent. This chemical is chosen for its compat ibility with the materials and water chemistry of borated water/stainl ess steel/zirconium/inconel systems. In addition, lithium-7 is produced in solution from the neutr on irradiation of the dissolved boron in the coolant. The lithium-7 hydroxide may be introduced into the RCS via the charging flow, or via alternate suitable flowpaths. The solution is prepared in the laboratory and is added via various insta lled plant systems or components. Reactor makeup water, or other suitable liquid, may then be used to flush the solution into the RCS. The concentration of li thium-7 hydroxide in the RCS is maintained in the range specified for pH control.
If the concentrati on exceeds this r ange, the cation bed demineralizer is employed in the letdown line in series operati on with the mixed bed demineralizer. During reactor startup from the cold condition, and at other times as necessary, hydrazine is employed as an oxygen scavengi ng agent. The hydr azine solution may be introduced into the RCS in the same manner as described above for the pH control agent. The reactor coolant is treat ed with dissolved hydrogen to control the net decomposition of water by radiolysis in th e core region. The hydrogen also reacts with oxygen and nitrogen introduced into the RCS as impurities under the impetus of co re radiation. Sufficient partial pressure of hydrogen is maintai ned in the volume control tank so that the specified equilibrium concentration of hydrogen is maintained in the reactor coolant. A self-contained pressure control valve maintains a mini mum pressure in the vapor space of the volume control tank. This can be adjusted to provide the correct equilibrium hydrogen concentration.
Boron, in the chemical form of boric acid, is adde d to the RCS for l ong-term reactivity control of the core.
Suspended solids (corrosion product particulates) are minimi zed in the reactor coolant by controlling makeup water and by the use of small micron filters. Other impurity concentrations are maintained below specified limits by controlling the chemical quality of makeup water and chemical additives and by purification of the reac tor coolant through the chemical and volume control system mixed bed demineralizer.
A soluble zinc compound may be added to the reactor cool ant as a means to reduce radiation fields within the pr imary system. The zinc used ma y be either natural zinc or CALLAWAY - SP5.2-15Rev. OL-21 5/15 zinc depleted of 64Zn. When used, the ta rget system zinc conc entration is normally maintained to a concentr ation no greater than 40 ppb.Zinc may be added to the reactor coolant system to reduce radiation fields and may later be added to reduce primary water stress corros ion cracking of Inconel-600 components.5.2.3.2.2Compatibility of Construc tion Materials with Reactor Coolant All of the ferritic low allo y and carbon steels which are used in principal pressure retaining applications have corrosion resist ant cladding on all surf aces that are exposed to the reactor coolant except for an area approximately 1.5 inches by 0.75 inches at approximate location 302.94 from vessel "0" and 384.89 inches down from the flange surface and an area approximately 0.53 inches by 0.3 inches at approximate location 185o from vessel "0" and 385 in ches down from the flange su rface. The existence of these areas has been evaluated as acceptable. The corrosion resistance of the cladding material is at least equivalent to the corrosion resistance of Types 304 and 316 austenitic stainless steel alloys or nickel-chromium-iron alloy, martensitic stainless steel, and precipitation hardened stainless steel. The cladding of ferritic type base materials receives a post-weld he at treatment, as requir ed by the ASME Code. Ferritic low alloy and carbon steel nozzles have safe ends of either stainless steel forged or wrought materials, stainless steel weld metal analysis A-7 (designated A-8 in the 1974 Edition of the ASME Code), or nickel-chromium-iron alloy weld metal F-Number 43. The latter buttering material requ ires further safe ending with austenitic stainless steel base material after completion of the post-weld heat treatment when the nozzle is larger than a 4-inch nominal inside di ameter and/or the wall thickness is greater than 0.531 inches. All of the austenitic stainle ss steel and nickel-chromium-ir on alloy base materials with primary pressure retainin g applications are used in the solution anneal heat treat condition. These heat treatments are as required by the material specifications.
During subsequent fabrication, these materials are not heated above 800°F other than locally by welding operations. The solution annealed surge line material is subsequently formed by hot bending foll owed by a re-solution annealing heat treatment. Components with stainless steel sensitized in the manner ex pected during component fabrication and installation will operate sati sfactorily under nor mal plant chemistry conditions in pressurized water reactor systems because chlorides, fluorides, and oxygen are controlled to very low levels.
CALLAWAY - SP5.2-16Rev. OL-21 5/155.2.3.2.3Compatibility with External Insulation and Environmental AtmosphereIn general, all of the materials listed in Tables 5.2-2 and 5.2-3 which are used in principal pressure-retaining applications and which are subject to elevated temperature during system operation are in contact with thermal insulation that covers their outer surfaces.
The thermal insulation used on the RCPB is eit her the reflective st ainless steel type or made of compounded material s which yield low leachabl e chloride and/or fluoride concentrations. The co mpounded materials in the form of blocks, boards, cloths, tapes, adhesives, cements, etc., are silicated to provide protection of austenitic stainless steels against stress corrosion which may result from accidental we tting of the insulation by spillage, minor leakage, or other contamination from the environmental atmosphere.
Appendix 3A includes a discussion which indicates the degree of co nformance with Regulatory Guide 1.36, "Nonmetallic Thermal Insulation for Austenitic Stainless Steel."
In the event of cool ant leakage, the ferritic materi als will show in creased general corrosion rates. Where minor leakage is anticipated from service experience, such as valve packing, pump seals, etc., only materials which are compatible with the coolant are used. These are as shown in Tables 5.2-2 and 5.2-3. Ferritic materials exposed to coolant leakage can be readily observed as part of t he inservice visual and/or nondestructive inspection program to assu re the integrity of the component for subsequent service. 5.2.3.3Fabrication and Proce ssing of Ferritic Materials5.2.3.3.1Fracture Toughness The fracture toughness properties of the RCPB components m eet the requirements of the ASME Code,Section III, Paragraph s NB, NC, and ND-2300 as appropriate.
The fracture toughness properti es of the reactor vessel materials are discussed in Section 5.3. Limiting steam generator and pressurizer RTNDT temperatures are guaranteed at 60°F for the base materials and the weldments. These materials will meet the 50 ft-lb absorbed energy and 35 mils lateral expansion requirements of the ASME Code,Section III at 120°F. The actual results of these tests are provided in the ASME material data reports which are supplied for each component and submitted to the owner at the time of shipment of the component. Calibration of temperature instruments and Charpy impact test machines are performed to meet the requirements of the ASME Code, Sect ion III, Paragr aph NB-2360. Westinghouse has conducted a test program to determine the fractu re toughness of low alloy ferritic mate rials with specified minimum yield strengths greater than 50,000 psi to CALLAWAY - SP5.2-17Rev. OL-21 5/15 demonstrate compliance with A ppendix G of the ASME Code,Section III. In this program, fracture toughness properties were determined and shown to be adequate for base metal plates and forgings, weld metal, and heat affected zone metal for higher strength ferritic materials used for components of the RCPB. The results of the program are documented in Refe rence 1, which has been submitted to the NRC for review. 5.2.3.3.2Control of Welding
All welding is conducted utilizi ng procedures qualified accordin g to the rules of Sections III and IX of the ASME Code. Control of welding variab les, as well as examination and testing during proc edure qualification and production welding, is performed in accordance with ASME Code requirements.
Appendix 3A includes discussions which indicate the degree of conf ormance of the ferritic materials components of the RCPB with Regulatory Guides 1.34, "Control of Electroslag Weld Properties," 1.43, "Control of Stainless Steel Weld Cladding of Low-Alloy Steel Components," 1.50, "Control of Preheat Temperature for Welding of Low-Alloy Steel," and 1.71, "Welder Qualificat ion for Areas of Limited Accessibility." 5.2.3.4Fabrication and Processing of Austenitic Stainless SteelSections 5.2.3.4.1 through 5.2.3.4.5 address Regulatory Guide 1.44, "Control of the Use of Sensitized Stainless Stee l," and present the methods and controls utilized by Westinghouse to avoid sensit ization and prevent intergranular attack of austenitic stainless steel components. Also, Appendix 3A includes a discussion which indicates the degree of conf ormance with Regulat ory Guide 1.44. 5.2.3.4.1Cleaning and Contami nation Protection Procedures It is required that all austenitic stainless steel mate rials used in t he fabrication, installation, and testing of nuclear steam supply components and systems be handled, protected, stored, and clea ned according to recognized and accepted methods which are designed to minimize contamination which could lead to stress corrosion cracking. The rules covering these controls are stipulated in Westinghouse process specifications.
As applicable, these process specifications supplement the equipm ent specifications and purchase order requirements of every indi vidual austenitic stainless steel component or system which Westinghouse procures for the SNUPPS nu clear steam supply systems, regardless of t he ASME Code clas sification.
CALLAWAY - SP5.2-18Rev. OL-21 5/15 The process specifications which define these requirements and which follow the guidance of the American National Standards Institute N-45 Committee specifications are as follows:
Appendix 3A includes a discussion which indicates the degree of co nformance of the austenitic stainless steel components of the RCPB with Re gulatory Guide 1.37, "Quality Assurance Requirements for Cl eaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants." 5.2.3.4.2Solution Heat Treatment Requirements The austenitic stainless steels listed in Tables 5.2-2 , 5.2-3 , and 5.2-4 are utilized in the final heat treated conditi on required by the res pective ASME Code, Se ction II materials specification for the particular type of grade of alloy. 5.2.3.4.3Material Testing Program Westinghouse practice is that austenitic stai nless steel materials of product forms with simple shapes need not be corrosion tested prov ided that the soluti on heat treatment is followed by water quenching.
Simple shapes are defined as all plates, sheets, bars, pipe, and tubes, as well as forgings, fitt ings, and other shaped products which do not Process NumberSpecification82560HMRequirements for Pressure Sensitive T apes for Use on Austenitic Stainless Steels83336KARequirements for T hermal Insulation Used on Austenitic Stainless Steel Piping and Equipment83860LARequirements for Marking of Reactor Plant Components and Piping84350HASite Receiving Inspection and Storage Requirements for Systems, Material, and Equipment84351NLDetermination of Surface Chloride and Fluoride on Austenitic Stainless Steel Materials85310QAPackaging and Preparing Nuclear Components for Shipment and Storage292722Cleaning and Packaging Requirements of Equi pment for Use in the NSSS597756Pressurized Water Reactor Auxiliary Tanks Cleaning Procedures597760Cleanliness Requirements During Storage Construction, Erection and Start-Up Activities of Nuclear Power System CALLAWAY - SP5.2-19Rev. OL-21 5/15 have inaccessible cavities or chambers that would preclude rapid cooling when water quenched. When testing is required, the tests are perform ed in accordance with ASTM A 262, Practice A or E, as amended by Westinghouse Process Specification 84201MW. 5.2.3.4.4Prevention of Intergranular Attack of Unstabilized Austenitic StainlessSteelsUnstabilized austenitic stainless steels are subject to intergranular attack (IGA) provided that three conditions are present simultaneously. These are: a.An aggressive environment, e.g., an acidic aqueous medium containing chlorides or oxygenb.A sensitized steelc.A high temperatureIf any one of the three conditions described above is not present, intergranular attack will not occur. Since high temper atures cannot be avoided in all components in the NSSS, reliance is placed on the elimination of conditions a and b to prevent intergranular attack on wrought stainless steel components. This is accomplished by: a.Control of primary water chemistr y to ensure a benign environment. b.Utilization of materials in the final heat treated condition and the prohibition of subsequent heat treatments in the 800 and 1,500°F temperature range. c.Control of welding processes and procedures to avoid heat affected zone sensitization. d.Confirmation that the welding procedures us ed for the manufacture of components in the primary pressure boundary and of reactor internals do not result in the sensitization of heat affected zones. Further information on each of these steps is provided in the following paragraphs:
The water chemistry in the RCS is controlled in Chapter 16 and plant procedures to prevent the intrusion of aggre ssive species. Reference 5 describes the precautions taken to prevent the intrusion of chlorides into the system during fabrication, shipping, and storage. The use of hydr ogen over pressure precludes the presence of oxygen during operation. The effectiveness of these controls has been demonstrated by laboratory tests and operating experience. The long-time exposure of severely sensitized stainless in early Westinghouse pressurized water reactors to reactor coolant environments has not resulted in any sign of intergranular attack. Reference 5 describes CALLAWAY - SP5.2-20Rev. OL-21 5/15the laboratory experimental findings and r eactor operating experi ence. The additional years of operations since the issuance of Reference 5 have provided further confirmation of the earlier conclusions that severely sensitized stainl ess steels do not undergo any intergranular attack in West inghouse pressurized water reactor coolant environments. In spite of the fact that there never has been any evidence that pressurized water reactor coolant water attacks sensitized stainless steels, Westinghouse considers it good metallurgical practice to avoid the use of sensitized stainless steels in the nuclear steam supply system components. Accordingly, measures are taken to prohi bit the purchase of sensitized stainless steels and to prevent sensitization dur ing component fabrication. Wrought austenitic stainless steel stock used for components that are part of: 1) the RCPB, 2) systems required for reactor shutdown, 3) systems required for emergency core cooling, and 4) reactor vessel internals (relied upon to permit adequate core cooling for normal operation or under postulated accident conditi ons) is utilized in one of the following conditions: a.Solution annealed and water quenched, orb.Solution annealed and cooled through t he sensitization te mperature range within less than approx imately 5 minutesIt is generally accepted that these practices will prevent sensitization. Westinghouse has verified this by performing corrosion tests on as-rec eived wrought material. The heat-affected zones of welded components must, of necessity, be heated into the sensitization temperature range, 800 to 1,500°F. However, se vere sensitization, i.e., continuous grain boundary precipitates of chromium ca rbide, with adjacent chromium depletion, can be avoid ed by controlling welding parameters and welding processes.
The heat input* and asso ciated cooling rate through the carbide precipitation range are of primary importance. We stinghouse has demonstrated this by corrosion testing a number of weldments.
Of 25 production and qualification weldments tested, r epresenting all major welding processes, and a variety of components, and incorporating base metal thicknesses from 0.10 to 4.0 inches, only portions of two were severely sensitized. Of these, one involved a heat input of 120,00 0 joules, and the other involved a heavy socket weld in relatively *Heat input is calculated according to the formula:Where: H = joules/in.
E = volts I = amperes S = travel speed, in./min.
H EI60S---------------------------
=
CALLAWAY - SP5.2-21Rev. OL-21 5/15 thin walled material. In bot h cases, sensitization was c aused primarily by high heat inputs relative to the section thickness. In only the socket weld did the sensitized condition exist at the surface, where the material is exposed to the environment. The component has been redes igned, and a material change has been m ade to eliminate this condition.
The heat input in a ll austenitic pressure boundary weldments has been controlled by: a.Prohibiting the use of block welding b.Limiting the maximum interpass temperature to 350°Fc.Westinghouse exercising approval rights on all welding procedures5.2.3.4.5Retesting Unstabilized Austenitic Stainless Steels Exposed to Sensitization Temperatures As described in the prev ious section, it is not normal Westinghouse practice to expose unstabilized austenitic stainless steels to the sensitization range of 800 to 1,500°F during fabrication into components. If, during the course of fabrication, the steel is inadvertently exposed to the sensitization temperature range, 800 to 1,500°F, the material may be tested in accordance with ASTM A 262, as amended by Westinghouse Process Specification 84201MW, to verify that it is not susceptible to intergranular attack, except that testing is not required for: a.Cast metal or weld metal with a fe rrite content of 5 percent or more, b.Material with a carbon c ontent of 0.03 percent or le ss that is subjected to temperatures in the range of 800 to 1,500°F for less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, c.Material exposed to special processi ng provided the processing is properly controlled to develop a uniform product and provided that adequate documentation exists of service experience and/or test data to demonstrate that the processing will not result in increased susceptibility to intergranular stress corrosion. If it is not verified that such material is not susceptible to intergranular attack, the material will be resolution annealed and wa ter quenched or rejected. 5.2.3.4.6Control of Welding The following paragraphs address Regulatory Guide 1.
31, "Control of Ferrite Content in Stainless Steel Weld Metal," and present the methods used, and the verification of these methods, for austenitic stai nless steel welding.
CALLAWAY - SP5.2-22Rev. OL-21 5/15The welding of austenitic stai nless steel is controlled to mitigate the occurrence of microfissuring or hot cracking in the weld. Although published data and experience have not confirmed that fissuring is detrimental to the qu ality of the weld, it is recognized that such fissuring is undesirable in a general sense. Also, it has been well documented in the technical literature that the presence of delta ferrite is one of the mechanisms for reducing the susceptibility of stainless steel welds to ho t cracking. However, there is insufficient data to specify a minimum delta ferrite level below which the material will be prone to hot cracking. It is assumed that such a minimum lies so mewhere between 0- and 3-percent delta ferrite.
The scope of these controls discussed herein encompasses welding processes used to join stainless steel parts in components design ed, fabricated, or st amped in accordance with the ASME Code,Section III, Class 1, 2, and core support components. Delta ferrite control is appropriate for the above welding requirements, except where no filler metal is used or for other reasons such control is not applicable. Thes e exceptions include electron beam welding, autogenous gas shie lded tungsten arc we lding, explosive welding, and welding using fully austenitic weldin g materials. The fabrication and installa tion specifications require welding procedure and welder qualification in accordance with Section III, and include the delta fe rrite determinations for the austenitic stainl ess steel welding materials that are used for welding qualification testing and for production processing. Specifically, the undiluted weld deposits of the "starting" welding materials are required to contain a minimum of 5-percent delta ferrite*
as determined by chemical anal ysis and calculation, using the appropriate weld metal constitution diagrams in Section III. When new welding procedure qualification tests are evaluated for these applications, including repa ir welding of raw materials, they are performed in accordance with the requirements of Se ction III and Section IX. The results of all the destructive and nondestructive tests are reported in the procedure qualification record in addi tion to the inform ation required by Section III. The "starting" welding materials used for fabrication and instal lation welds of austenitic stainless steel materials and components meet the requirements of Section III. The austenitic stainless steel weld ing material confo rms to ASME weld metal analysis A-7 (designated A-8 in the 1974 Edition of the ASME Code), Type 308 or 308L for all applications. Bare weld filler metal, including consumable inserts, used in inert gas welding processes conform to ASME SFA 5.9, and are procured to contain not less than 5-percent delta ferrite according to Section III. Weld filler metal materials used in flux shielded welding processes conform to ASME SFA 5.4 or 5.9 and are procured in a wire-flux combination to be capable of providing not less than 5-percent delta ferrite in the deposit according to Section III. Welding materials are tested, using the welding energy inputs to be employed in production welding. *The equivalent ferrite number may be substituted for percent delta ferrite.
CALLAWAY - SP5.2-23Rev. OL-21 5/15 Combinations of approved heat and lots of "starting" weldi ng materials are used for all welding processes. The welding quality a ssurance program includes identification and control of welding material by lots and heats as appropriate. All of the weld processing is monitored according to approved inspection pr ograms which include review of "starting" materials, qualification records and welding parameters. Welding systems are also subj ect to quality assurance aud it including ca libration of gages and instruments; identification of "starting" and comple ted materials; welder and procedure qualificatio ns; availability and use of a pproved welding and heat treating procedures; and documentary evidence of co mpliance with materials, welding parameters, and inspection requirements. Fabrication and installation welds are inspected using nondestructive examination me thods according to Se ction III rules. To assure the reliability of these controls, Westinghouse has completed a delta ferrite verification program, descri bed in Reference 6, whic h has been approved as a valid approach to verify the Westinghouse hypothesis and is considered an acceptable alternative for conformance wi th the NRC Interim Position on Regulatory Guide 1.31. The Regulatory Staff's acceptance letter and topical report evaluati on were received on December 30, 1974. The program results, which do support the hypothesis presented in Reference 6, are summariz ed in Reference 7.
Appendix 3A includes discussions which indicate the degree of conf ormance of the austenitic stainless steel components of the RCPB with Regulatory Guides 1.34, "Control of Electroslag Properties," and 1.71, "Welder Qualific ation for Areas of Limited Accessibility." 5.2.4INSERVICE INSPECTION AND TESTING OF THE REACTOR COOLANT PRESSURE BOUNDARY Inservice inspection and testing of pressure-retaining components, such as vessels, piping, pumps, va lves, and bolting and supports within the reactor coolant pressure boundary, complies with the ASME Code, includi ng addenda, per 10 CFR 50.55a(f) for inservice testing and 10 CF R 50.55a(g) for inservice in spection. Exceptions to compliance with ASME Code is obtained when specific written relief is granted by the NRC per 10 CFR 50.55a(a)
(3), 10 CFR 50.55a(f)(6), or 10 CFR 50.55a(g)(6), or when Code Cases are incorporat ed per 10 CFR 50.55a(b)(5) or 10 CFR 50.55a(b)(6). The limitations and modiifcati ons that the NRC places on the ASM E Code in 10 CFR 50.55a(b) are adhered to. The inservice testing of pumps a nd valves per the requirements of the ASME OM Code are discussed in Section 3.9(B).6. Callaway initially submitted separate pr eservice and inservice inspection program documents, for pumps and va lves, which complied with "NRC Staff Guidance for Complying with Certain Provision of 10 CFR 50.55a(g)--Inservice Inspection Requirements." Subsequent in service inspection program documents are prepared in accordance with the 10-year update requirements in 10 CFR 50.55a and submitted to the NRC for initial approval.
The inspection program documents identify the applicable CALLAWAY - SP5.2-24Rev. OL-21 5/15 Section XI edition and addenda and provide the details for the areas subject to examination, method of ex amination, extent and frequ ency of examination, and applicable Code Cases. 'Relief Requests' seeking relief from applicable code requirements are submitted to the NRC and become part of the inservice inspection program upon approval by the NRC. The repair and replacem ent program identifies the applicable Section XI edition and addenda, applicable Code Cases and relief requests, and provides the administrative controls for performing repairs and replacements.Since the plant will be required to meet the requirements of future editions of Section XI, insofar as practicable, an attempt was made during design to allow access for inspections and coverages anticipated to be requ ired by later editions of the Code. The result of this effo rt has increased the areas on the RPV available to mechanized inservice inspection. Callawa y has attempted to create an inservice inspection program and plant design that are consistent with the 10 CFR 50 philosophy of upgrading inspections. 5.2.4.1Inspection of Class I ComponentsThe system boundary subject to inspection includes all piping and components in quality Group A (ASME Boiler and Pressure Vessel Code, Se ction III, Class I). The reactor pressure vessel (RPV), pressurizer, Class 1 portion of the steam generators, and all Class 1 piping, pumps, and valves are examined exc ept for those areas where relief has been r equested and granted.The scope of examinations, inspections, and acceptance criteria meets the requirements outlined in Section XI of the ASME Boiler and Pressure Vessel Code, "Rules for Inservice Inspection of Nuclear Power Plant Components."
In addition, the ultrasonic examination of ferritic, austenitic, and dissimilar metal components will be performed in accordance with IWA-2232.
The extent of selection of piping welds for examination is de termined by the risk-informed ISI program (RI-ISI) implemented in accordance wi th ASME Section XI and 10 CFR 50.55a.
5.2.4.2Arrangement and Accessibility5.2.4.2.1General
Access for the purpose of inserv ice inspection is defined as t he design of the plant with the proper clearances for exam ination personnel and/
or equipment to perform inservice examinations during a nuclear unit shutdown. During system and component arrangement design, careful attention was given to ph ysical clearances to allow personnel and equipment to perform required inse rvice examinations. Access CALLAWAY - SP5.2-25Rev. OL-21 5/15requirements of the Code have been considered in the design of components, weld joint configuration, and system arrangement. An inservice inspection program design review was undertaken to identify any exceptions to the access requirements of the code with subsequent design modification s and/or inspection techni que development to ensure Code compliance, as required. Additi onal exceptions may be identified and reported to the NRC after plant operations, as specified in 10 CFR 50.55a(g) (5)(iv). Space has been provided to handle and store insulation, st ructural members, shielding, calibration blocks, and similar material related to the inspection. Suitable hoists and other handling equipment are also provided. Lighting, sources of power, and services for the inspection equipment are provided at appropriate locations. Access is provided for volumetric examination of the pressure-containing welds from the external surfaces of components and piping by means of removable insulation and removable shielding. Provisions for suitable a ccess for inservice inspection examinations will minimize the time requir ed for these inspection s to be performed.
Therefore, they will reduc e the amount of radiation exposure to both plant and examination personnel. Working platforms hav e been provided at strategic locations in the plant to permit ready access to those areas of the reactor coolant pressure boundary which are designated as inspection points in the inservice inspection program. Areas without permanent platform s will be provided with temporary platforms and/or scaffolding, as required.5.2.4.2.2Access to Reactor Pressure Vessel Access for inspection of the RPV will be as follows: a.Access to the inner surface of the RPV will be available during refueling outages when the vessel co re barrel is removed.
A remotely operated examination device designed to perform ultrasonic examinations from the inner surface of the ve ssel will be used to examine the vessel-to-flange weld, nozzle-to-shell we lds, and the l ongitudinal circum ferential, and meridional welds of the vessel. However, ve ssel welds below the 2,011-foot-6-inch cavity shelf elevat ion may be examined from the outer surface of the vessel.
Selected areas of reacto r cladding and the internal support attachments welded to the vessel wall will be accessible for remote visual examination when the core barrel is remo ved at the end of the 10-year inspection interval. A camera capable of remote positioning will be inserted into the RPV.b.If required, examinatio n personnel will be able to install tracks for the examination of the nozzl e and piping welds, the l ongitudinal welds in the upper shell course, and the flange-to-shell weld.
These tracks can be lowered from the 2,021-foot-7-3/4-inch containment elev ation through the opening between the vessel fl ange and the insulation.
The mechanized equipment may be installed in a similar manner.
CALLAWAY - SP5.2-26Rev. OL-21 5/15c.The vessel flange seal su rface will be accessible during refueling outages when the closure head is removed. The vessel-to-flange weld can be examined manually or mechanically fr om the flange seal surface, using ultrasonic techniques. The inside surface of the RPV will be available for a mechanized examinatio n of the vessel-to-flange we ld from the vessel side during refueling outages when the core barrel is remo ved. If examination of the vessel-to-flange weld is required when the co re barrel has not been removed, the weld can be examined from the exterior surface of the vessel.d.Access to the exterior surface of the RPV below the 2,011-foot-6-inch cavity shelf elevation for augmented inservice ins pection is available since an annular space has been provided bet ween the vessel exterior surface and the insulation interior surface. This permits the inse rtion of remotely operated inspection devices between the insulation and the reactor vessel.
Examination pers onnel can enter the area bel ow the RPV through one approximately 3-foot-square access port in the insulation to install the pole track remote examination device. Th e bottom head insu lation is designed to allow an examiner to walk on the insulation while installing the examination device. Access to the window is provided through the in-core instrumentation tunnel. A 3-foot annular space between the exterior surf ace of the RPV and the interior surface of the insulation has been provided from the vessel closure flange elevation to the cavity shelf elevation. T he clearance area provides sufficient access for examination pe rsonnel and equipment to perform any augmented inservice examinations on the exterior surfaces of the nozzle-to-shell, safe end, pipe-to-elbow, flange-to-shell, and longitudinal welds in the upper shell course of the vessel.e.The closure head will be dry stored during refueling, which will facilitate direct manual examination. Removable insulation will allow examination of the outside surf aces of the closure head. All reactor vessel nuts and washers will be removed to dry st orage during refueling and may be examined at that time. The reactor vessel studs capable of being removed may be dry stored in ra cks. Required inservic e examinations may be performed in the stud racks or reactor vessel st uds may be removed from the stud racks fo r examination.5.2.4.2.3Pressurizer
The external surface is accessible for visual and volume tric inspection by removing the external insulation. Manways are provided to allow access for internal visual inspection.
The insulation around the pressurizer heaters is provided with a means to identify component leakages during system hydrostatic and pressure testing.
CALLAWAY - SP5.2-27Rev. OL-21 5/155.2.4.2.4Heat Exchangers and Steam Generators The external surface is accessible for volumetric and vi sual inspection by removing portions of the vessel insulation. Manways in the steam generator channel head provide access for internal visual examinations and eddy current tests of steam generator tubes. 5.2.4.2.5Piping Pressure Boundary
The physical arrangement of piping, pumps, and valves has been desi gned to allow personnel access to welds requiring inservice inspection. Modifications to the initial plant design have been incorporated where practical to provide proper inspection access.
Removable insulation has been provided w here required by the Code on those piping systems requiring ultrason ic and/or surface examinations.
In addition, the placement of pipe hangers and supports with respect to these welds has been reviewed and modified where necessary to reduce the amount of plant s upport required in t hese areas during inspection. Working platforms are provided in areas required to facilitate the servicing of pumps and valves. Temporary or permanent plat forms and ladders will be provided, as necessary, to gain access to piping welds. A conscientious effort has been made to minimize the number of fitting-to-fitting welds within the inspection boundary. Welds requi ring inspection have been located to permit ultrasonic examinations from at least one side, but, where component geometries permit, access from both sides of t he weld is provided. The surfaces of the welds requi ring ultrasonic examination by the Code have been prepared to permit effectiv e examination. 5.2.4.2.6Pump Pres sure BoundariesThe internal pressure-retaining surfaces of the pumps are a ccessible for visual inspection by removing the pum p internals. External surf aces of the pump casing are accessible for visual and volumetric examin ation by removing comp onent insulation.
Examination will be performed when the pumps are di sassembled for maintenance purposes. 5.2.4.2.7Valve Pressure Boundaries Class 1 valves over 4-inch nominal size are accessible for disassembly for visual examination of internal pressure boundary surfaces. 5.2.4.3Examination Te chniques and ProceduresTechniques and procedures, including any special technique and pr ocedure for visual, surface, and volumetric examinations, will be writ ten in accordance with the requirements of Subarticle IWA-2200 and Tabl e IWB-2500-1 of Sect ion XI of the ASME Code, applicable year and addenda.
The liquid penetrant or magnetic particle methods CALLAWAY - SP5.2-28Rev. OL-21 5/15 will be utilized for surface examinations, radiographic (RT), and/or ultrasonic (UT) methods (either automated or manual) for volu metric examinations. 5.2.4.3.1Equipment for Inservice InspectionProcedures governing the use of the following examination devices will be qualified prior to examinations in the plant. 5.2.4.3.1.1Ultrasonic Examination Equipment Although the permanent tracks described in this section are still installed, they have never been used, due to advances in technology. The descr iption of the permanently installed tracks is left in this section to document their existence for future use, if deemed necessary.The remotely operated device for inservice inspection of the vesse l and connected piping from their inner surfaces, as required, may be attached to the RPV at the flange surface. The device is capable of moving the transduc ers over the surface of the components in any direction.
The equipment for augemented inservice ins pection of the reactor pressure vessel circumferential and vertical welds below the 2,011-foot-6-inch cavity elevation consists of remotely operated devices wh ich can travel over the ve ssel shell or on permanently installed tracks between the vessel surface and the insulation. Tracks are located such that the devices requiring them are capable of movi ng ultrasonic transdu cers over all required lengths of the sh ell welds. Remote ultr asonic scanning equipment for examination of nozzle-to-vessel welds, sa fe ends, and pipe-to-el bow welds from the outer surface of the co mponent can be mounted on the pipe or elbow. The nozzle-to-vessel weld examination equipment will provide radial and circumferential motion to the ultrasonic transducer while rotating about the nozzl
- e. The pipe weld examination device will provi de longitudinal and circumferent ial motion to the ultrasonic transducer while rota ting about the pipe.
An electronic system with a receiver or data channel fo r each ultrasonic transducer will be used for acquiring and storing data w hen using remote automated examination equipment. Reflected signa ls may be transmitted through an ultrasonic instrument, gated, and multiplexed to initiate a digital recording.
Scanning position will be indicated by encoders and subsequently logged by the data acqu isition system. The key parameters of each reflector recorded in clude location, maximu m signal amplitude, depth below the scanning surface, and length of reflector. However, similar or compatible systems of data acquisition may be utilized. 5.2.4.3.1.2Surface Ex amination Equipment Mechanized surface examinatio n techniques will provide results which are at least equivalent to those obta inable by manual surface examination techniques.
CALLAWAY - SP5.2-29Rev. OL-21 5/155.2.4.3.1.3Visual Ex amination Equipment Remote visual examinat ion techniques will provide a resolution cap ability which is at least equivalent to that obtainable by direct visual observation. 5.2.4.3.2Coordination of Inspecti on Equipment with Access ProvisionsAccess to areas of the plant requiring inservice inspection is provided to allow the use of existing equipment, wher ever practicable. 5.2.4.3.3Manual ExaminationIn areas where manual ultrasonic examination is performed, all reportable indications will be mapped and records made of maximum signal amplitude, depth below the scanning surface, and length of the reflector. The data compilation format will be such as to provide for comparison of data from subsequent examinations. Radiographic techniques may be used where ultrasonic techniques ar e not applicable. In areas where manual surface or direct vi sual examinations are performed, all reportable indications will be mapped with respect to size and location in a manner to allow comparison of data from subsequent examinations. 5.2.4.4Inspection Intervals The inspection interval, as defined in Subarticle IWA-2400 of Section XI, is a 10 year interval of service. These inspection intervals represent calendar years after the reactor facility has been placed into commercial service. The interval may be extended by as much as one year to permit inspections to be concurrent with plant outages. The inspection schedule shall be in accordance with IWB-2400.
Inservice examinations are performed during normal plant outages, such as refueling shutdowns or maintenance shutdowns occurring during the inspection inte rval. No examinations will be performed which require draining of th e reactor vessel further than just below the nozzles or removal of the core solely for the purpose of accomp lishing the examinations. 5.2.4.5Examination Categories and Requirements The extent of the ex aminations performed and the examination me thods utilized are in accordance with Table IWB-250 0-1 of ASME Section XI. In lieu of the above for class 1, 2, and 3 piping welds, a risk-informed ISI program (RI-ISI) was implemented in accordance with ASM E Section XI and 10 CFR 50.55a.
In addition, preser vice inspections comply with IWB-2200.
CALLAWAY - SP5.2-30Rev. OL-21 5/155.2.4.6Evaluation of Examination ResultsEvaluation of examination results for Class 1 components are conducted in accordance with the requirements of Ar ticle IWB-3000 of t he ASME Code,Section XI, 1998 Edition with 2000 Addenda. In addition, the recording and evaluation of examinations results for the reactor pressure vessel (RPV) are done as per Regulatory Gui de 1.150, Revision 1.5.2.4.7System Leakage and Hydrostatic TestsThe hydrostatic test for th e reactor pressure vessel and reactor cool ant pressure boundary will be conducted in a ccordance with the requirements of Articles IWA-5000 and IWB-5000. System leakage tests will be conducted prior to startup following each reactor refueling outage, in ac cordance with Paragraph IWB-5221, as required by Table IWB-2500-1. A system hydrosta tic test will be co nducted at or near the end of each inspection interval in accordance with Paragraph IWB-5222, as required by Table IWB-2500-1. Examinati ons performed during these tests will be co nducted without the removal of insulation. 5.2.5REACTOR COOLANT PRESSURE BOUNDARY LEAKAGE DETECTION SYSTEMS5.2.5.1Design Bases5.2.5.1.1Safety Design Bases There is no safety design basis for the reactor coolant pre ssure boundary leakage detection system.5.2.5.1.2Power Gener ation Design BasesPOWER GENERATION DESIGN BASIS ONE - For leaks of 1 gpm or greater, other than identified leakage sources, the reactor coolant boundary leakage detec tion systems are designed to detect leaks and determine the leakage rate (in accordance with Regulatory Guide 1.45 and 10 CFR 50, Appendix A, General Design Criterion 30). A comparison with the Regulatory Guide requirements is provided in Table 5.2-6
.POWER GENERATION DESIGN BASIS TWO - The leakage detection equipment is designed to continuously monitor the environmental conditions within the containment so that a background level is identif ied which is indicative of the normal level of leakage from primary systems and components. Significant upw ard deviation from normal containment environmental conditi ons provides positive indicati on in the control room of increases in leakage rates.
CALLAWAY - SP5.2-31Rev. OL-21 5/155.2.5.2System Description5.2.5.2.1General DescriptionIDENTIFIED LEAKAGE DETECTION - Certain components of t he reactor coolant pressure boundary may have small amounts of leakage and cannot, from a practical standpoint, be made leaktight.
These identified sources of leakage are piped to the reactor coolant drain tank or the Pressurizer Relief Tank w hose levels are indicated and alarmed in the control room. The annular gap between the O-rings in the reactor vessel head flange is tapped and piped to a temperature indicator and then to the reactor coolant drain tank. Reactor coolant le akage gives a high temperature indication and alarm. Additionally, the controlled leakage shaft seal system for the reactor coolant pumps is monitored by reactor coolant drain tank le vel indication and alarm.
UNIDENTIFIED LEAKAGE DETE CTION - The reactor c oolant pressure boundary leakage detection system consists of the sump level and flow monitoring system, the containment air particulate monitoring system, and the contai nment cooler condensate measuring system. The sump level and flow monitoring syst em indicates leakage by monitoring increases in sump level. The containment cooler condensate measuring system detects leakage from the release of steam or water to the containment atmosphere, The air particulate monitoring system detects leakage fr om the release of radioactive materials to the containment atmosphere. The containment gaseous radioactivity monitors could provide additional indi cation of leakage if significant reactor coolant gaseous activity is present from fuel cladding defects. The containment humidity measuring system is also av ailable as an indirect in dication of leakage to the containment.
Primary-to-secondary re actor coolant leakage, if it occurs, is detected by the following radioactivity monitors: the main condenser evacuation, th e steam generator liquid, the steam generator blowdown proc essing, and the steam gener ator blowdown discharge (Section 11.5.2). Reactor coolant pressure b oundary leakage is also indica ted by increasing charging pump flow rate compared wi th reactor coolant system inventory changes and by unscheduled increases in re actor makeup water usage.INTERSYSTEM LEAKAGE - Lea kage to any significant degree into the auxiliary systems connected to the RCPB is not expected to occur.
Design and administrative provisions which serve to li mit leakage include is olation valves designed for low seat leakage, periodic testing of RCPB check valves (see Section 6.3.4.2
), and inservice inspection (see Section 6.6
). Leakage will be detected by increasing the auxiliary system level, temperature, a nd pressure indications or lif ting of the relief valves accompanied by increasing values of monitored parameters in the relief valve discharge path. These systems are isolated from the RCS by normally closed valves and/or check valves.
CALLAWAY - SP5.2-32Rev. OL-21 5/15a.Residual Heat Removal System (S uction Side) - The RHR system is isolated from the RCS on the suct ion side by motor-operated valves 8701A/B and 8702A/B. Leakage past these valves w ill be detected by lifting of relief valves 8708A or 8708B, accompanied by increasing pressurizer relief tank le vel, pressure, and temp erature indications and alarms on the main control board. b.Safety Injection System/Accumulato rs - The accumulators are isolated from the RCS by check valves 8948A/B/C/D and 8956A/B
/C/D. Leakage past these valves and into the accumulator subsystem will be detected by redundant control room accumulator pr essure and level indications and alarms. c.Safety Injection System/RHR Discharge Subsystem - The RHR pump portion of the safety inje ction system is isolated from the RCS by check valves 8948A/B/C/D, 8818A/B/C/D, 8949B/C, 8841A/B, and normally closed motor-operated valve 8840. Leakage past these valves will eventually pressurize the RHR discharge header and result in lifting of the relief valves 8856A and 885 6B or 8842. Relief valve lifting will be accompanied by control r oom indication a nd alarms of increasing boron recycle holdup tank levels. d.Safety Injection System/SI Pump S ubsystem - The safety injection pump portion of the safety inje ction system is isolated from the RCS by check valves 8948A/B/C/D; EP-V010, V 020, V030, V040; 8949A/B/C/D; EM-V001, V002, V0 03, V004; and normally closed motor-operated valves 8802A/B. Leakage past these valves will pressuri ze the safety injection pump discharge header, result ing in control room i ndication of increasing pressure and eventually lifting of relief valve 8851 or 8853A/B. Relief valve lifting will be accompanied by cont rol room indication and alarms of increasing boron recycle holdup tank levels. e.Safety Injection System/ECCS Ch arging Pump Subsystem - The ECCS charging pump subsystem is isolated from the RCS by check valves BB-V001, V022, V040, V059, and EM-8815, and normally closed motor-operated valves EM-8801A/B. Leakage past th ese valves will pressurize the header between EM-8801A/B and normally closed motor-operated valves EM-8803A/B. If valves EM-8803A
/B leak, the ECCS charging pump discharge piping could be pressurized. If an ECCS charging pump were not ru nning, check valves BG
-8481A/B would protect the suction side of the pumps. All piping and valves in this pathway from the discharge of the ECCS charging pumps to the RCS pressure boundary are safety-related and rated for RCS pressure. Thus, for a pathway to low pressure components, ther e would need to be a le akage through at least three check valves and two normally closed motor-operated gate valves. Any leakage through this pathway woul d be extremely unlikely and should CALLAWAY - SP5.2-33Rev. OL-21 5/15 be very small. Any si gnificant leakage to the suction of the ECCS centrifugal charging pumps could be detected by a change in charging flow rate.f.Waste Processing System - The wast e processing system is isolated from the RCS by manua l valves BB-V008, V028, V047, V066 and BB-V009, V029, V048, V067. Leakage past these valves wil l result in increasing the control room indication of reactor coolant drain tank level and reactor coolant drain tank pump flow. g.Head Gasket Monitoring Connections - Leakage pa st the reactor vessel head gasket(s) will result in temperature indication and alarm in the control room. h.Component Cooling Water - Leakage from the reac tor coolant system to the component cooling water system, which services all components of the reactor coolant pressure boundary that require cooling, is detected by the component cooling water r adioactivity monitoring system and/or increasing surge tank level. (Section 11.5.2
). Leakage to the containment at mosphere from the reactor coolant pressure boundary would cause a change in the containment airborne radioactivity which would be detected by the air particulate monitors. If the reactor is operating with a known rate of leakage, at a constant power level, with a constant reactor coolant activity and a constant purge rate, both the gross particulate and gross noble gas activities would reach an equilibrium level. Under these conditions, an abnormal in crease in monitored ac tivity would be the result of increased l eakage. Such leakage is classified as unidentified until its source is determined.
During the expected modes of oper ation, the reactor coolant ac tivity level fluctuates due to power variations and variations in letdown flow rate. However, significant increases in leakage can be detected.Leakage detection systems have been designed to aid operating personnel, to the extent possible, in differentiating between possi ble sources of detecte d leakage within the containment and ident ifying the physical lo cation of the leak.The containment atmosphere particulate monitoring system provides the primary means of remotely determining the presence of reactor coolant leakage within the containment. Increases in containment airbor ne activity levels detected by either of t he monitors indicate the reactor coolant pressure boundary as the source of leakage. Conversely, if the condensate measuring system detects increased containment moisture without a corresponding increase in airbor ne activity level, the indica ted source of leakage would be judged to be a nonradioacti ve system, except during ti mes when reactor coolant activity may be low.
CALLAWAY - SP5.2-34Rev. OL-21 5/15 Less sensitive methods of l eakage detection, such as unexpl ained increases in reactor plant makeup requirements to maintain pressurizer level, also provide indication of the reactor coolant pressure bounda ry as a potential leakage s ource. Increases in the frequency of a particular containment sump pump operation or increases in the level in a particular sump facilitate lo calization of the source to components whose leakage would drain to that sump. Leakage rates of the magnitude necessary to be detectable by these latter methods are expected to be noted first by the more sensitive radiation and moisture detection equipment.Normally, unidentified l eakage from the reactor coolant pressure boundary is essentially zero. The reactor c oolant system is an all welded system , with the exce ption of the connections on the pressurizer safety valves , reactor vessel head, and the pressurizer and steam generator manways, which are flanged. All connections to the reactor coolant system are welded. All isolation or check valves between the reactor coolant system and other systems have been designed for low seat leakage, and reactor coolant pressure boundary check valve backleakage is checked periodically.
In general, va lves in the reactor coolant system 2 inches and under are of the packless type.
All valves larger than 2 inches have either dual packing with a leak-off conne ction to the reactor coolant drain tank between the two packings or a carbon spacer and 5 rings of packing.The plant containment has the capability for a continuous purg e of 4,000 cfm.
The time to recirculate one containment free air volume through the containment air coolers is 4.57 minutes. The component operation for various le ak detection systems, as discussed in Section 5.2.5.2.3, is based on this containm ent purge and recirculation time.MAXIMUM ALLOWABLE TOTAL LEAKAGE - The limits for the reactor coolant pressure boundary leakage are:
identified, 10 gpm and unidentified, 1 gpm. When leakage has been identified, it will be evaluated by the operating staf f to determine if operation can safely continue. Under these conditions, an allowable total leakage from known sources of 10 gpm has been est ablished. Conti nued operation of the reacto r with identified or unidentified leakage shall be in accordance with the Technical Specifications.
For normal chemical and volume control system operation, the chargi ng pump flow rate is 87 gpm, which consists of 55 gpm being charged through the normal charging line and 32 gpm being supplied to the reactor coolant pump seals. Tota l flow leaving the reactor coolant system via the normal letdown pa th is 75 gpm, with the remaining 12 gpm returning via the seal water re turn line to the chemical an d volume contro l system. The design flow rate of the normal charging pump is 130 gpm at the design head of 5900 feet, which, during normal oper ation, leaves adequate pump capacity to make up for reactor coolant pressure boundary leakage. The two 150-gpm ECCS centrifugal charging pumps are also used in charging service. Maximum letdown flow is 120 gpm with an additional 12 gpm leaving via t he seal water return line. A centrifugal charging pump with 150-gpm rated capacity (the NCP can deliver this flow rate at less than the design value of total develo ped head) provides an CALLAWAY - SP5.2-35Rev. OL-21 5/15adequate reserve capacity to make up for leakage. Thus, under normal or maximum letdown flow conditions, a 10-gpm maximum limit on reactor coolant pressure boundary leakage can be accomm odated by operation of one charging pump.
The reactor coolant pressu re boundary leakage detecti on system provides ample protection to assure that, in the unlikely event of a failure of the reactor coolant pressure boundary, small cracks wil l be detected prior to becoming la rge leaks. In particular:a.The sensitivity of the detection eq uipment is such that leaks can be identified when small, an d the plant can be shut down. The limit on continued operation for uni dentified leakage is 1 gpm.
This is well within the detection capability of the reactor coolant pressure boundary leakage detection system. b.The time span for a crack to go from detectable size to critical size varies from 5 to more than 40 y ears. This assures adequate safety from a major loss-of-coolant accident. The above methods are supplemented by visual and ultrasonic inspections of the reactor coolant pressure boundary during plant shutdown periods, in a ccordance with the inservice inspection program (Section 5.2.4). 5.2.5.2.2Component DescriptionCONTAINMENT AIR PARTICULATE MONITOR - An air sample is drawn outside the containment into a closed system by a sa mple pump and is then consecutively passed through a particulate fi lter with detectors, an iodine filter with detector, and a gaseous monitor chamber with detector.
The sample transport system includes:a.A pump to obtain the air sample b.A flow control valve to provide flow adjustmentc.A flow meter to i ndicate the flow rated.A flow alarm assembly to prov ide high and low flow alarm signalsThe particulate filter is cont inuously monitored by a scintillation crystal with a photo multiplier tube which provides an output signal propor tional to the activi ty collected on the filter. The particulat e monitor has a range of 10
-12 to 10-7 Ci/cc and a minimum detectable concentration of 10-11Ci/cc. The containment air particulate monitoring system is capable of performing its radioactive monitoring functions follow ing an SSE. More details concerning the partic ulate monitors can be found in Section 11.5.2.3.2.2
.
CALLAWAY - SP5.2-36Rev. OL-21 5/15CONTAINMENT GASEOUS RADIOACTIVITY MONITOR - The containment gaseous radioactivity monitor determines gaseous radioactivity in the containment by monitoring continuous air samples from the containment atmosphere. After passing through the gas monitor, the sample is returned via the cl osed system to the containment atmosphere.
Each sample is continuously mi xed in a fixed, shielded volume where its activity is monitored. The moni tor has a range of 10
-7 to 10-2Ci/cc and a minimum detectable concentration of 2 x 10
-7 Ci/cc.The containment gaseous radioactivity monitors are fu lly described in Section 11.5.2.3.2.2
.The containment gaseous radioac tivity monitoring system is capable of performing its radioactivity monitoring functions following an SSE.CONTAINMENT PURGE MONITORS - The containment pur ge system radioactivity monitors (Section 11.5.2.3.2.3) serve as a backup to the containment air particulate and gaseous airborne radioactivity monitoring system while the pur ge is in operation.CONTAINMENT COOLER CONDENSATE MONITORING SYSTEM - The condensate monitoring system permits measurements of the liquid runoff from the containment cooler units. It consists of a containm ent cooler drain collection header, a vertical standpipe, valving, and standpipe level instrumentation for each cooler.The condensation from the containment cool ers flows via the collection header to the vertical standpipe. A different ial pressure transmitter provides standpipe level signals.
The system provides measurements of low leakages by monitoring standpipe level increase versus time.CONTAINMENT HUMIDITY MONITORING SYSTEM - The cont ainment humidity monitoring system, utilizing tem perature compensated humidity detectors, is provided to determine the water v apor content of the containment atmosphere.
An increase in the humidity of the containment atmosphere indicates release of water within the containment. The range of the containment humidity measuring system is 10-to 98-percent relative humidity at 80°F with a temperature range of 40 to 120°F.CONTAINMENT SUMP LEVEL AND FLOW MONITORING SYSTEM -
Since a leak in the primary system would result in reactor coolant flowing into the containment normal or instrument tunnel sumps, leakage would be i ndicated by a level increase in the sumps. Indication of increasing sump level is transmitted from the sump to the control room level indicator by means of a sump level transmitter. The system provides measurements of low leakages by monitoring level increase versus time.
CALLAWAY - SP5.2-37Rev. OL-21 5/15CHARGING PUMP OPERATION - Du ring normal operation, ei ther the normal charging pump or an ECCS centrifugal c harging pump will be in operat ion. If a gross loss of reactor coolant occurs which is not detected by the method s previously described, the flow rate of the normal chargi ng pump, if operating, would indicate t he leakage from the reactor coolant system. This leakage must be suffic ient to cause a decrease in pressurizer or volume control tank level that is within the sensitivity range of the level indicators. The flow from the normal (or ECCS) chargi ng pump would automatically increase to try to maintain pressurizer level. Charging pump discharge flow indication is provided in the control room.SUMP PUMP OPERATION - Sinc e a leak in the primary syst em may result in reactor coolant flowing into the cont ainment normal or instrument tunnel sumps, gross leakage can be indicated by an incr ease in the frequency of operation of the containment normal or the containment instrument tunnel sump pumps. Pu mp operation can be monitored from the control room.LIQUID INVENTORY - Larger l eaks may also be det ected by unschedu led increases in the amount of reactor coolant makeup water which is required to maintain the normal level in the pressurizer. Pressurizer level can be monitored in the control room. Total makeup water flow is also available from the plant computer.5.2.5.2.3Component Operation CONTAINMENT AIR PARTICULATE MONITOR - Particulate acti vity is determined from the containment free volume and the coolant fission and corrosion product particulate activity concentrations. The sensitivity of the containment air particulate monitors for primary coolant leakage detection is dependent on both the primary coolant activity level and the background radi ation level in cont ainment which is dep endent upon the power level, percent failed fuel, crud bursts, iodine spiking, and natur al radioactivity brought in by the containment purge. Any increase of more than two standard deviations above the count rate for background would indicate a possible leak. The total particulate activity concentration above back ground, due to an abnormal leak and natural decay, increases almost linearly with time for the first several hours after th e beginning of a leak. As shown in Figure 5.2-2 , with 0.1-percent failed fuel, containment background airborne particulate radioactivity equivalent to 10
-4 percent/day, and a partition factor equal to 0.2, a 1-gpm leak would be detected in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Larger lea ks would be detected in proportionately shorter times (exclusive of sample transport ti me, which remains constant). The detection capabilities and res ponse times are shown on Figure 5.2-2. Shortly after startup and also during steady operation with low levels of fuel defects, the level of radioactivity in the reactor coolant is lower than what was assumed in the original design bases calculation. Us ing a reactor coolant source term based on representative real-time data, with no fuel defects, it was determined that the containment air particulate monitors are capable of detect ing a one gpm leak in one hour.
CALLAWAY - SP5.2-38Rev. OL-21 5/15 The leakage flow rate can be determined from the count rate when the specific background radioactivity present before the leakage begins is known. This method is limited by the fact that large uncertainties are possible when dete rmining the associated leak rate by calculation. Therefore, in the event of an alarm or in creasing trend on these monitors, a water inventory bal ance is performed to determ ine the equivalent RCS leak rate.CONTAINMENT GASEOUS RADIOACTIVITY MONITOR - The containment atmosphere gaseous radioactivity monitor is less sensitive than the containment air particulate monitor but provides a positive indication of leak age in the event that reactor coolant gaseous activity exists as a result of fuel-cladding defects.
Gaseous radioactivity is determined from the containment free vo lume and the gaseous ac tivity concentration of the reactor coolant. Any increase more than two standar d deviations above the count rate for background woul d indicate a possible leak. The total gase ous activity level above background (after 1 year of normal operation) increases almost linearly for the first several hours after the beginning of the leak. As specified in Figure 5.2-2, with 0.1-percent failed fuel, containment background air borne gaseous radioactivity equivalent to 0.1 percent/day, and a partition fact or equal to 1 (NUREG-0017 assumptions), a 1-gpm leak would be detected within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Larger leaks would be detected in proportionately shor ter times (exclusive of the sample transport time which remains constant). The detection capabilities and re sponse times are shown on Figure 5.2-2.Analyses have shown that the pre-existing containment radioactive gaseous background levels for which reliable detec tion is possible is dependent upon the reactor power level, percent failed fuel, and nat ural radioactivity brought in by the containment purge. With primary coolant concentration s less than equilibrium levels, such as during reactor startup and operation with no fuel defects, the increase in detector count rate due to leakage will be partially masked by the statistical variat ion of the mi nimum detector background count rate, rendering reliable detection of a 1 gpm leak uncertain. The containment atmosphere gaseous radioactivi ty monitors were designed in accordance with the sensitivities specif ied in Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems," its alarm setpoint is set to indicate a 1 gpm RCS leakage based on Regulatory Guide 1.45 assumptions, it is fully functioning in accordance with its design requirements, however, it has been removed as part of the reactor coolant pressure bounda ry leakage detection system due to its inability to promptly detect a 1 gp m RCS leakage within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> with reduc ed radioactivity levels in the reactor coolant system. (Reference 8)CONTAINMENT PURGE MONITORS - The containment purge monitors function the same as the containment air pa rticulate and gaseous radioacti vity monitors, except that the purge monitors sample from the containment purge exhaust line.CONTAINMENT COOLER CONDENSATE MONITORING SYSTEM - The condensate flow rate is a function of containment humidity, essential service water temperature leaving the coolers, and containment purge rate. The water vapor dispersed by a 1 gpm CALLAWAY - SP5.2-39Rev. OL-21 5/15 leak is much greater than the water vapor brought in with the outside air.
Air brought in from the outside is heated to 50°F before it enters the containment.After the air enters the contai nment, it is heated to 100-120°F so th at the relative humidity drops. The water vapor brought in with the outside air does not build up in the containment since it is continually purged. The most important fact or in condensing the water vapor is the temperatur e of the essential service wa ter which is provided to the containment coolers. This water can vary between 38 - 100°F on the ou tlet of the coolers, depending on seasonal conditions.
Level changes of as little as 0.25 inches in the cooler condensate standpipes can be detected. Increases in the condensation rates over normal background are monitored by the Plant computer based upo n level checks each minute in order to de termine the unidentified leakage. Figure 5.2-2 shows the detection capab ilities of the system for various seasonal conditions with no airborne identified leakage. The system is capable of detecting a 1 gpm leak within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after water vapor has reached th e coolers and started to condense. Normal background leakage will increase containment humidity to the point where condensation will more readily occur and, thereby, will improve the detection capabilitie s of this system.
The rate of leakage can be determined when the precise essential service water, outside air, and containment air temperatures and the outsi de relative humidity are known by use of psychrometric charts.
A check of other instrumentation would be required to eliminate possib le leakage from non-radioactiv e systems as a cause of an increase in condensate flow.CONTAINMENT HUMIDITY MONITORING SYSTEM - Since the humidity level is influenced by several factors, a quantitative evaluation of an indica ted leakage rate may be questionable and should be compared to obser ved increases in liquid flow from sumps and condensate flow from air coolers. Humidity le vel monitoring is used as an indirect indicating devic e to alert the operator to a potential problem.
The accuracy of the humidity detectors is +/-3 percent. A rapid increase of humidity over the background level by more than 10 percent can be taken as a probable indication of a leak.CONTAINMENT SUMP LEVEL AND FLOW MONITORING SYSTEM - The detection capabilities of the containment normal sump and instru ment tunnel sump are shown in Figure 5.2-2 , assuming that the water from the leak has reached the sump.The minimum detectable change in the containment normal su mp level is 5 gallons and in the instrument tunnel sump level is 15 gallons. When the Instrument Tunnel Sump is completely dry, the initial minimum detectable level change is 25 gallons. The levels are scanned by the Plant computer once per minute, and the normal background rate of increase in sump level is subtract ed to determine t he leakage rate.
CALLAWAY - SP5.2-40Rev. OL-21 5/15 The actual reactor coolant leakage rate can be established from the increase above the normal rate of change of sump level after consideration of 35 percent of the high temperature leakage which initially ev aporates but may be condensed by the containment coolers an d then is routed to the sump. A check of other instrumentation would be required to eliminate possible leakage from nonradioactive systems as a cause of an increase in sump level.CHARGING PUMP OPERATION - The normal c harging pump normally operates at 132 gpm. For 75 gpm letdown the pump will send 87 gpm to the RCS and seal s and 45 gpm in recirculation. For 120 gpm letdown the pump will send 132 to the RCS and seals. Any significant increase in t he flow rate is a possib le indication of a leak.During some transient modes of operation it may be desirable to operate with only the 45 gpm letdown orifice in operation. During these conditions the normal charging pump will send 57 gpm to the RCS and seals and 45 gpm will be in recirculati on. Any significant increase in the flow rate is a possible indication of a leak.
The leakage rate can be deter mined by the amount that the charging pump rate to the RCS increases above either 87 or 132 gpm depending on letdown flow to maintain constant pressurizer level.SUMP PUMP OPERATION - U nder normal conditions, the containment normal and instrument tunnel sump pumps will operate very infrequently. Gross leakage can be surmised from unusual freq uency of pump operation. Su mp level and pump running indication are provided in the cont rol room to aler t the operators.
The leakage rate can be determined from sump volume s and frequency of sump pump operation.LIQUID INVENTORY - The oper ators can surmise gross le akage from changes in the reactor coolant inventory. No ticeable decreases in the pressurizer level not associated with known changes in operat ion will be invest igated. Likewise, makeup water usage information which is availabl e from the plant co mputer will be c hecked frequently for unusual makeup rates not d ue to plant operations.5.2.5.3Safety Evaluation Inasmuch as this system has no safety desi gn basis, no safety eval uation is provided. Criteria for the selection of safety design bases are stated in Section 1.1.7
.5.2.5.4Tests and Inspections Periodic testing of leakage det ection systems is conducted to verify the operability and sensitivity of detector equipment. These tests include installation calibrations and alignments, periodic channel ca librations, functional test s, and channel checks. A CALLAWAY - SP5.2-41Rev. OL-21 5/15 description of calibration and maintenanc e procedures and frequencies for the containment radioactivity moni toring system is presented in Section 11.5.2
.The humidity detector and con densate measuring system are al so periodically tested to ensure proper operation.
The condensate measuring system is also tested to verify sensitivity.
Inservice inspection criteria , the equipment used, procedure s involved, the frequency of testing, inspection, surveillanc e, and examination of the structural and leaktight integrity of reactor coolant pressure boundary components are described in detail in Section 5.2.4.5.2.5.5Instrumentation Applications The following indications are provided in the control room to allow operating personnel to monitor for leakage:a.Containment air particulate monitor - air particulate activityb.Containment gaseous activity monitor - gaseous activityc.Containment cooler condensate monitoring system - standpipe leveld.Containment normal sump level and instrument tunnel sump levele.Containment humidity measuring system - containment humidityf.Gross leakage detection methods - Charging pump flow rate, let-down flow rate, pressurizer level and reactor cool ant temperatures are available for the charging pump flow method. Contai nment sump levels and pump operation are available fo r the sump pump operation method. Totalized makeup water flow is available from the plant computer for liquid inventory.5.
2.6REFERENCES
1.Logsdon, W. A., Begley, J. A., and Gottshall, C. L., "Dynamic Fracture Toughness of ASME SA508 Class 2a and ASME SA533 Grade A Class 2 Base and Heat Affected Zone Material and Applicable Weld Metals," WCAP-9292, March 1978. 2.Letter NS-CE-1730, dated Marc h 17, 1978, C. Eicheldinger (Westinghouse) to J. F. Stolz (NRC). 3.Cooper, L., Miselis, V. and Starek, R. M., "Overpressure Protection for Westinghouse Pressurized Water Reactors
," WCAP-7769, Revision 1, June, 1972 (also letter NS-CE-622, dated Ap ril 16, 1975, C.
Eicheldinger CALLAWAY - SP5.2-42Rev. OL-21 5/15(Westinghouse) to D. B. Vassallo (NRC), additional informat ion on WCAP-7769, Revision l). 3a.J. E. Fontes, "Overpressure Protection Report for the Union Electric Co. Callaway Plant," Revision 3, dated 8/
- 94. (SCP 94-143) Prepared for Amendment 128 (OL-1186) MSSV Setpoint Tolerance Change.4.Burnett, T. W. T., et al., "LOFTRAN Code Descr iption," WCAP-7907, October 1972. 5.Golik, M. A., "Sensitized Stainless Steel in Westinghouse PWR Nuclear Steam Supply Systems," WCAP-7477-L (Proprietary), Ma rch, 1970 and WCAP-7735 (Non-Proprietary), August 1971. 6.Enrietto, J. F., "Control of Delta Ferrite in Austenitic Stainless Steel Weldments," WCAP-8324-A, June 1975. 7.Enrietto, J. F., "Delta Ferrite in Production Austenitic Stainless Steel Weldments," WCAP-8693, January 1976.8.NRC Letter, "Callaway Plant, Unit 1 - License Amendment Request to the Reactor Coolant System Leakage Detection Instrumentation Methodology (TAC NO.
MC8220), May 16, 2006.
CALLAWAY - SP Rev. OL-21 5/15TABLE 5.2-1 APPLICABLE CODE ADDENDA FOR REACTOR COOLANT SYSTEM COMPONENTSReactor vesselASME III, 1971 Ed ition through Winter 1972 Reactor vessel closure headASME II I, 2001 Edition th rough 2003 AddendaSteam generatorASME III, 1989 Edition ASME XI 1995 Edit ion through 1996 AddendaPressurizerASME III, 1974 Ed ition through Summer 1974CRDM housingASME III, 2001 Ed ition through 2003 Addenda CRDM head adapterASME III, 2001 Edition thr ough 2003 AddendaReactor coolant pumpASME III, 1971 Edition thr ough Summer 1973Reactor coolant pipeASME III, 1974 Edition through Winter 1975Surge linesASME III, 1974 Ed ition through Winter 1975ValvesPressurizer safetyASME III, 1974 Edition thr ough Summer 1975 Motor operatedASME III, 1974 Edition through Summer 1975 Manual (3 inch and larger) ASME III, 1974 Editi on through Summer 1975 ControlASME III, 1974 Edit ion through Summer 1975 CALLAWAY - SP Rev. OL-21 5/15TABLE 5.2-2 CLASS 1 PRIMARY COMPONENTS MATERIAL SPECIFICATIONSReactor Vessel Components Shell plates (other than core region)SA-533, Grade A, B or C, Class 1 or 2 (vacuum treated)Shell plates (core region)SA
-533, Grade A or B, Class 1 (vacuum treated)Closure head forgingSA-508, Grade 3, Class 1 Shell, flange and nozzle forgings, nozzle safe ends SA-508, Class 2 or 3; SA-182, Grade F304 or F316 CRDM and/or ECCS appurtenances, upper head SB-167 (UNS N06690) and SA-479, Type 304/304LInstrumentation tube appurtenances, lower headSB-166 or SB-167 and SA-182, Grade F304, F304L or F316Closure studs, nuts, washers, inserts, and adaptors SA-540, Class 3, Gr ade B23 or B24 (as modified by Code Case
1605)Core support padsSB-166 with carbon less than 0.10 percentMonitor tubes and vent pipeSA-312 or SA-376, Grade TP304 or TP316 or SB-166 or SB-167 or
SA-182, Grade F316Vessel supports, seal ledge, and heat lifting lugs SA-516, Grade 70 (quenched and tempered) or SA-533, Grade A, B or C, Class 1 or 2 (vessel supports may be of weld metal buildup of equiva lent strength of the nozzle material)Cladding and butteringStainless Steel Weld Metal Analysis A-8 and Ni-Cr-Fe Weld Metal F-Number 43Steam Generator ComponentsPressure forgings (including nozzles and tube sheet)
SA-508, Class 3a CALLAWAY - SPTABLE 5.2-2 (Sheet 2)
Rev. OL-21 5/15Nozzle safe endsSA-182F 316LNChannel headsSA-508 Class 3a TubesSB-163 (Ni-Cr-F e thermally treated) Cladding and butteringStainless Steel Weld Metal Analysis A-8 and Ni-Cr-Fe Weld Metal F-Number 43Closure studs or boltsnuts washers SA-193, Grade B16 SA-194, Grade 7 ASTM F436, Thru hardenedPressurizer ComponentsPressure platesSA-533, Grade A, Class 2Pressure forgingsSA-508, Class 2aNozzle safe endsSA-182, Grade F316L Cladding and butteringStainless Steel Weld Metal Analysis A-8 and Ni-Cr-Fe Weld Metal
F-Number 43Closure studs or bolts nuts washersSA-193, Grade B7 SA-194, Grade 7 ASTM F436, Thru hardened Reactor Coolant PumpPressure forgingsSA-182, Grade F304, F316, F347 or F348Pressure castingSA-351, Grade CF8, CF8A or CF8MTube and pipeSA-213; SA-376 or SA-312, Seamless, Grade TP304 or TP316Pressure platesSA-240, Type 304 or 316Bar materialSA-479, Type 304 or 316 Closure boltingSA-193; SA-320; SA-540, SA-453, Grade 660, or Inconel 718, SB-637FlywheelSA-533, Gr ade B, Class 1 CALLAWAY - SPTABLE 5.2-2 (Sheet 3)
Rev. OL-21 5/15 Reactor Coolant PipingReactor coolant pipeSA-351 , Grade CF8A Centrifugal Casting Reactor coolant fittings, branch nozzles SA-351, Grade CF8A and SA-182, (Code Case 1423-2) Grade
316NSurge lineSA-376, Grade TP304, TP316 or F304N Auxiliary piping 1/2 through 12 inch and wall schedules 40S
through 80S (ahead of second isolation valve)ANSI B36.19 All other auxiliary piping (ahead of second isolation valve)ANSIB36.10Socket weld fittingsANSI B16.11Piping flangesAN SI B16.5 Full Length CRDMLatch housingSA-182M, Grade F304 Rod travel housingSA-182M, Grade F304
Welding materialsStainless Steel Weld Metal Analysis A-8 CALLAWAY - SP Rev. OL-21 5/15TABLE 5.2-3 CLASS 1 AND 2 AUXILIARY COMPONENTS MATERIAL SPECIFICATIONSValves BodiesSA-182, Grade F316 or SA-351, Grade CF8 or CF8M BonnetsSA-182, Grade F316 or SA-351, Grade CF8 or CF8M SA-479, Type XM-19 (solenoid-operated head vent valves only)DiscsSA-182, Grade F316 or SA-564, Grade 630, or SA-351, Grade CF8 or CF8M StemsSA-182, Grade F316 or SA-564, Grade 630 300 Series SST (solenoid-operated head vent valve rods only)Pressure-retaining boltingSA-453, Grade 660 Pressure-retaining nutsSA-453 , Grade 660 or SA-194 Grade 6 Auxiliary Heat Exchangers HeadsSA-240, Type 304 Nozzle necksSA-182, Grade F304 TubesSA-213, Grade TP304 Tube SheetsSA-182, Grade F304 ShellsSA-240 and SA-312, Grade TP304 Auxiliary Pressure Vessels, Tanks, Filters, etc. Shells and headsSA-240, Type 304 or SA-264 (consisting of SA-537, Class 1 with Stainless Steel Weld Metal Analysis A-8 Cladding)
CALLAWAY - SPTABLE 5.2-3 (Sheet 2)
Rev. OL-21 5/15Flanges and nozzlesSA-182 , Grade F304 and SA-105 or SA-350, Grade LF2 or LF3 with Stainless Steel Weld Metal Analysis A-8 CladdingPipingSA-312 and SA
-240, Grade TP304 or TP316 Seamless Pipe fittingsSA-403, Grade WP304 Seamless Closure bolting and nutsSA-193, Grade B7 and SA-194, Grade 2H Auxiliary Pumps Pump casing and headsSA-351 , Grade CF8 or CF8M; SA-182, Grade F304 or F316 Flanges and nozzlesSA-182, Grade F304 or F316; SA-403, Grade WP316L Seamless PipingSA-312, Grade TP304 or TP316 Seamless Stuffing or packing box coverSA-351, Grade CF8 or CF8M; SA-240, Type 304 or 304L or 316 Pipe fittingsSA-403, Grade WP316L Seamless Closure bolting and nutsSA-193 , Grade B6, B7 or B8M; SA-194, Grade 2H or 8M; SA-453 Grade 660, and Nuts, SA-194, Grade 2H, 6 and 8 M CALLAWAY - SP Rev. OL-13 5/03TABLE 5.2-4 REACTOR VESSEL INTERNALS FOR EMERGENCY CORE COOLING SYSTEMSForgingsSA-182, Grade F304PlatesSA-240, Type 304 PipesSA-312, Grade TP304 Seamless or SA-376, Grade TP304TubesSA-213, Grade TP304 BarsSA-479, Type 304 and 410 CastingsSA-351, Grade CF8 and CF8A BoltingSA-193, Grade B8M (65 MYS/90 MTS)
Code Case 1618 Inconel-750; SA-461, Grade 688NutsSA-193, Grade B8 Locking devicesSA-479, Type 304 CALLAWAY - SP Rev. OL-1610/07TABLE 5.2-5 RECOMMENDED REACTOR COOLANT WATER CHEMISTRY LIMITS (c )Hydrogen is maintained in the reactor coolant during plant oper ation per the EPRI PWR Primary Water Chemistry Guidelines. Twenty four hours prior to a scheduled reactor shutdown and cooldow n, hydrogen may be reduced to 15 cc (STP)/kg water.
Electrical conductivityDetermined by the concentration of boric acid and alkali present. Solution pHDetermined by t he concentration of boric acid and alkali present.
Oxygen (a)NOTES:(a)Oxygen concentraton shoul d be controlled by scavenging with hydrazine to less than 0.1 ppm in the reactor coolant prior to exceeding a temperature of 250°F.
During power operation with the specified hydrogen concentration maintained in the coolant, the resi dual oxygen concentration do es not exceed 0.005 ppm.
0.005 ppm, maximumChloride0.15 ppm, maximumFluoride0.15 ppm, maximum Hydrogen(c) 25 to 50 cc (STP)/kg H 2 O pH control agent (Li 7 OH)Control Limits and Action Guidelines are listed in pl ant procedures.Boric acidVariable from 0 to ~4000 ppm as BSilica1.0 ppm, maximum Aluminum (b)(b)Aluminum, Calcium and Magnesium analyse s are performed on major reactor makeup water sources such as the demineralized water storage tank which feeds the reactor makeup water storage tank and the boric acid storage tanks.
0.05 ppm, maximum Calcium + Magnesium (b)0.05 ppm, maximum Magnesium (b)0.025 ppm, maximumZincPer Chemistry Program (maximum 40 ppb steady-state)
CALLAWAY - SP Rev. OL-16 10/07TABLE 5.2-6 DESIGN COMPARISON WITH REGULATORY GUIDE 1.45, DATED MAY 1973, TITLED REACTOR COOLANT PRESSURE BOUNDARY LEAKAGE DETECTION SYSTEMS Regulatory Guide 1.45 PositionUnion ElectricC. REGULATORY POSITION The source of reactor coolant leakage should be identifiable to the extent practical. Reactor c oolant pressure boundary leakage detection and collection systems should be se lected and designed to include t he following:1.Leakage to the primary reactor containment from identified sources should be collected or otherwise isolated so that:a.the flow rates are monitored separately from unidentified leakage, andb.the total flow rate can be established and monitored.1.Complies. Flow to the RCDT and the PRT can be established, is monitored, and is separated from unidentified leakage.2.Leakage to the primary reactor containment from unidentified sources should be collected and th e flow rate monitored with an accuracy of one gallon per minute (gpm) or better.2.Complies. The instrumentation provided is such that over a period of time (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or more), the collected flow rate can be determined with an a ccuracy of better than 1 gallon per minute.
CALLAWAY - SPTABLE 5.2-6 (Sheet 2)
Rev. OL-16 10/073.At least three separate detection methods should be employed and two of these me thods should be (1) sump level and flow monitoring and (2) airborne particulate radioactivity monitoring. The third method may be selected from the following:a.monitoring of condensate flow rate from air coolers,b.monitoring of airborne gaseous radioactivity.Humidity, temperature, or pressure monitoring of the containment atmosphere should be c onsidered as alarms or indirect indication of leakage to the containment.3.Complies. The methods provided are sump-level and flow (level versus time) monitoring, airborne particulate radioactivity monitoring, and containment cooler condensate monitoring.Containment atmosphere humidity monitoring is also av ailable as an indirect indication of leakage to the containment. As such, periodic testing of the sensitivity of the humidity monitori ng system is not required.4.Provisions should be made to monitor systems connected to the RCPB for signs of intersystem leakage. Methods should
include radioactivity monitoring and indicators to show abnormal water levels or flow in the affected area.4.Complies. Refer to Sections 5.2.5.2.1, 9.3.3 , and 11.55.The sensitivity and response time of each leakage detection system in regulatory posit ion 3. above employed for unidentified leakage should be adequate to detect a leakage rate, or its equivalent, of one gpm in less than one hour.5.Complies, as described in Section 5.2.5.2.3 and as shown on Figure 5.2-2
.Regulatory Guide 1.45 PositionUnion Electric CALLAWAY - SPTABLE 5.2-6 (Sheet 3)
Rev. OL-16 10/076.The leakage detection systems should be capable of performing their functions following seismic events that do not require plant shutdown.
The airborne particulate radioactivity monitoring syst em should remain functional when subjecte d to the SSE.6.Complies. The airborne particulate radioactivity system is designed to remain functional when subjec ted to the SSE.
Refer to Sections 11.5.2.3.2.2 and 11.5.2.3.2.3. The remaining leakage detection systems ca n reasonably be expected to remain functional following seismic events of lesser severity than the SSE. However, no special qualification program is used to assure operability under such conditions.7.Indicators and alarms for each leakage detection system should be provided in the main control room. Procedures for converting various indications to a common leakage equivalent should be availabl e to the operators. The calibration of the indicato rs should acco unt for needed independent variables.7.Complies, as described in Sections 5.2.5.2.3 and 5.2.5.5.8.The leakage detection system s should be equipped with provisions to readily permi t testing for operability and calibration during plant operation.8.Complies. Refer to Section 5.2.5.4
.9.The technical specifications should include the limiting conditions for identified and unidentified leakage and address the availability of various types of instruments to assure adequate cover age at all times.9.Complies. Refer to the Technical Specifications Regulatory Guide 1.45 PositionUnion Electric CALLAWAY - SP5.3-1Rev. OL-21 5/155.3REACTOR VESSEL5.3.1REACTOR VESSEL MATERIALS 5.3.1.1Material Specifications Material specifications ar e in accordance with the ASME Code requirements and are given in Section 5.2.3. The ferritic material s of the reactor vessel beltline are restricted to the following maximum limits of copper, phosphorous, and vanadium to reduce sensitivity to irradiation embrittlement in service:5.3.1.2Special Processes Used for Manufacturing and Fabricationa.The vessel is Safety Class 1. Desi gn and fabrication of the reactor vessel is carried out in strict accordance with ASME C ode,Section III, Class 1 requirements. The cl osure head and nozzles are manufactured as forgings. The cylindrical portion of the vessel is ma de up of formed plates joined by full penetration longitudinal and girth we ld seams. The bottom hemispherical head is made from dished plates. The reactor vessel parts are joined by welding, us ing the single or multiple wire submerged arc and the shielded metal arc processes. b.The use of severely sensitized st ainless steel as a pressure boundary material has been prohibited and has be en eliminated by ei ther choice of material or prog ramming the method of assembly. c.The surfaces of the guide studs ar e chrome plated to prevent possible galling of the mated parts. d.At all locations in the reactor vessel where stainless st eel and Inconel are joined, the final joining weld beads are Inconel weld metal in order to prevent cracking.
ElementBase Metal(percent) As Deposited Weld Metal (percent)Copper0.10 (Ladle) 0.12 (Check) 0.10Phosphorous0.012 (Ladle) 0.017 (Check) 0.015Vanadium0.05 (Check)0.05 (as residual)
CALLAWAY - SP5.3-2Rev. OL-21 5/15e.The location of full penetration we ld seams in the upper closure head and vessel bottom head are restricted to areas that permit acce ssibility during inservice inspection. f.The stainless steel clad surfaces ar e sampled to assure that material composition requirements are met. g.Freedom from underclad cracking is assu red by special ev aluation of the procedure qualification for cladding ap plied on low allo y steel (SA-508, Class 2). h.Minimum preheat requirements have been established for pressure boundary welds, using low alloy material. The preheat is maintained until either an intermediate or full post-weld heat treatment is completed or until the completion of welding. 5.3.1.3Special Methods for Nondestructive Examination The nondestructive examinat ion of the reactor vessel and its appurtenances is conducted in accordance with ASME Code,Section III requirements; also numerous examinations are performed in addition to ASME Code,Section III requirements.
Nondestructive examinat ion of the vessel is discussed in the following paragraphs and the reactor vessel quality assu rance program is given in Table 5.3-1. 5.3.1.3.1Ultrasonic Examinationa.In addition to the required ASME Code stra ight beam ultrasonic examination, angle beam inspection over 100 percent of one major surface of plate material is performed during fabrication to detect discontinuities that may be undetected by the straight beam examination. b.In addition to the ASME Code,Section III nondestr uctive examination, all full penetration ferritic pr essure boundary welds in the reactor vessel are ultrasonically examined during fabricati on. This test is performed upon completion of the welding and intermediate heat treat ment but prior to the final post-weld heat treatment. c.After hydrotesting, all full penetrati on ferritic pressure boundary welds in the reactor vessel, as well as th e nozzle to safe end welds, are ultrasonically examined. These inspections are also per formed in addition to the ASME Code, Se ction III nondestructi ve examinations. 5.3.1.3.2Penetrant ExaminationsThe partial penetration welds for the contro l rod drive mechanism head adaptors and the bottom instrumentation tubes ar e inspected by dye penetrant after the root pass, in CALLAWAY - SP5.3-3Rev. OL-21 5/15addition to code requirements. Core support block attachme nt welds are inspected by dye penetrant after the first layer of weld metal and after each 1/2 inch of weld metal. All clad surfaces and other vesse l and head internal surfac es are inspected by dye penetrant after the hydrostatic test. 5.3.1.3.3Magnetic Pa rticle ExaminationThe magnetic particle examination requirements below ar e in addition to the magnetic particle examination requirements of Section III of the ASME Code. All magnetic particle examinations of materials and weld s are performed in accordance with the following: a.Prior to the final post-wel d heat treatment -
Only by the prod, coil, or direct contact method.b.After the final post-wel d heat treatment - Only by the yoke method.
The following surfaces and welds are examined by magnetic particle methods. The acceptance standards are in accordance wit h Section III of the ASME Code.
Surface Examinationsa.Magnetic particle exami ne all exterior vessel and head surfaces after the hydrostatic test. b.Magnetic particle exami ne all exterior closure st ud surfaces and all nut surfaces after final machining or rolling. Conti nuous circular and longitudinal magneti zation is used. c.Magnetic particle exami ne all inside diameter su rfaces of carbon and low alloy steel products that have thei r properties enhanced by accelerated cooling. This inspection is performed after forming and machining (if performed) and prior to cladding. Weld ExaminationMagnetic particle examination of the weld metal build-up for vessel support welds, the closure head lifting lugs, and the refueling seal ledge to the reactor vessel after the first layer and each 1/2 inch of weld metal is deposit ed. All pressure boundary welds are examined after back chipping or back grinding operations. 5.3.1.4Special Controls for Ferritic and Austenitic Stainless SteelsWelding of ferrite steels and austenitic stainle ss steels is discussed in Section 5.2.3. Section 5.2.3 includes discussions which indica te the degree of c onformance with CALLAWAY - SP5.3-4Rev. OL-21 5/15 Regulatory Guide 1.44.
Appendix 3A discusses the degree of conformance with Regulatory Guides 1.43, 1.50, 1.71, and 1.99. 5.3.1.5Fracture Toughness Assurance of adequate fracture toughness of ferritic materi als in the reactor coolant pressure boundary (ASME Code,Section III, Class 1 components) is provided by compliance with the requirements for fracture toughness te sting included in NB-2300 to Section III of the ASME Code and Appendix G of 10 CFR 50.
The initial Charpy V-notch minimum upper s helf fracture energy le vels for the reactor vessel beltline (including weld s) are 75 foot-pounds, as r equired per Appen dix G of 10 CFR 50. Materials having a section thickness greater than 10 inches with an upper shelf of less than 75 foot-pounds are evaluated with regard to effects of chemistry (especially copper content), initial upper shelf energy, and fluence to assure that a 50-foot-pound shelf energy, as required by Appendix G of 10 CFR 50 is maintained throug hout the life of the vessel. The s pecimens are oriented as required by NB-2300 of Sect ion III of the ASME Code. The vessel fracture toughness data is provided in Table 5.3-4 for Callaway Plant. 5.3.1.6Material SurveillancePrior to Refuel 20 at Callaway, the reac tor vessel surveillance program was directed toward evaluation of the effect of radiation on the fracture toughness of reactor vessel steels, based on the transit ion temperature ap proach and the fracture mechanics approach. The program conformed to ASTM E-185 "Recomm ended Practice for Surveillance Tests for Nuclear Reactor Vessels," and 10 CFR 50, Appendix H.The program used six specimen capsules that we re put in place prior to plant start-up. The capsules were located in guide baskets welded to the outside of the neutron shield pads and positioned directly oppos ite the center portion of the core. The capsules were able to be removed or replaced only when the vessel head and upp er internals were removed. The six capsules contained reactor vessel steel specimens oriented both parallel and normal (longitudina l and transverse) to the principal rolling direction of the limiting base material locat ed in the core region of the reactor vessel and associated weld metal and weld heat-affected zone metal. In total, the six capsules contained 54 tensile specimens, 360 Charpy specimens, and 72 compact tension (CT) specimens.
The evaluation of radiation damage was based on pre-irradiation an d post-irradiation testing of Charpy V-no tch and tensile specimens.
CALLAWAY - SP5.3-5Rev. OL-21 5/15Each of the six capsules cont ains the following specimens:After Refuel 20, the CT specimens, capsules, and material left from Charpy V-notch and tensile testing wil l be stored by the analyst to support future testing, reconstitution, or reinsertion, unless gi ven NRC approval to discard. Ar chive material, which has never been irradiated and is sufficient for two additional capsules, will be retained beyond Refuel 20 as well. A descrip tion and location of the archived material is provided in WCAP-15151 (Reference 1A).As part of the reactor vessel surveillance program prior to Refuel 20, dosimeters were placed in filler blocks drilled to contain them. The dosimeters permitted evaluation of the flux seen by the specim ens and the vessel wall. In addition, ther mal monitors made of low melting point alloys were included to monitor the ma ximum temperature of the specimens. The specimens were enclosed in a tight-fitting stainless steel sheath to prevent corrosion and ensure good thermal conductivity. The complete capsule was helium leak tested. As pa rt of the surveill ance program, a report of the residual elements in weight percent to the nearest 0.
01 percent was made for surveillance material and as-deposited weld metal.
The fast neutron exposure of the specimens occu rred at a faster rate than that experienced by the vessel wa ll, with the specimen s being located between the core and the vessel. Since these sp ecimens experienced accelerat ed exposure and were actual samples from the materials used in the vessel, the transition temperature shift measurements are representative of the vessel at a later time in life. Data from CT fracture toughness specimens provide additi onal information for use in determining allowable stresses for irradiated material.
Correlations between the calculations and measurements of the irradiated samples in the capsules, assuming the same neutron spectrum at the samples and the vessel inner Material Number of Charpys Number ofTensiles Number ofCTs Limiting base material**Specimens oriented in the major rolling or working direction. 1534 Limiting base material****Specimens oriented normal to the ma jor rolling or working direction. 1534Weld metal******Weld metal to be selected per ASTM E-185.1534Heat-affected zone15--
CALLAWAY - SP5.3-6Rev. OL-21 5/15 wall, are described in Section 5.3.1.6.1. The anticipated degree to which the specimens perturbed the fast neutron flux and energy distri bution is considered in the evaluation of the surveillance specimen data. Verification and possible readjustment of the calculated wall exposure is made by the use of data on all capsules withdrawn. The schedule for removal of the capsules fo r post-irradiation testing co nformed to ASTM E-185 and Appendix H of 10 CFR 50 and is included as Table 5.3-10
.The material surveillance program for radi ation damage evaluation af ter Refuel 20 is described in Section 5.3.1.6.3.
The following dosimeters and thermal monitors were included in each of the six capsules:
Dosimeters Iron Copper Nickel Cobalt-aluminum (0
.15 percent Co) Cobalt-aluminum (cadmium shielded)
Np-237 (cadmium shielded)
Thermal Monitors 97.5 percent Pb, 2.
5 percent Ag (579F melting point) 97.5 percent Pb, 1.75 percent Ag, 0.75 percent Sn (590F melting point)
The material surveillance program for radiation damage evaluation after Refuel 20 is described in Section 5.3.1.6.3. 5.3.1.6.1Measurement of Integrated Fast Neutron (E > 1.0 MeV) Flux at the Irradiation Samples Methodology for the measurement of Integrated Fast Neutron (E>1.0 MeV) Flux at the Irradiation Samples is described in WCAP 14040 (Reference 1) and was approved by the NRC.
CALLAWAY - SP5.3-7Rev. OL-21 5/155.3.1.6.2Section Deleted5.3.1.6.3Ex-Vessel Neutron DosimetryThe Code of Federal Regulations, Title 10, Part 50, Appendi x H, requires that neutron dosimetry be present to monitor the reactor vessel throughout plant life and that material specimens be used to measure damage associat ed with the end-of-life fast neutron exposure of the reactor vessel. The Ex-Vessel Neutron Dosimetry (EVND) Program at Callaway is designed to provid e a verification of fast neut ron exposure distributions within the reactor vessel wall and to establish a me chanism to enable long-term monitoring of those portions of the reactor vessel and vessel support structure that could experience significant radiati on induced increases in reference nil ductility transition temperature (RTNDT) over the service lifetime of the plant. When us ed in conjunction with dosimetry from internal surveillance capsules (discussed above) and with the results of neutron transport calculations, the reactor cavity neutron measurements allow the projection of embrittlement gradients through the reactor vessel wall with a minimum uncertainty.Technical Descri ptionTo achieve the goals of t he EVND Program, two types of measurements are made. Comprehensive sensor sets, in cluding radiometric monitors (RMs), are employed at discrete locations within the r eactor cavity to characterize the neutron energy spectrum variations axially and azimutha lly over the beltline region of the reactor vessel. In addition, stainless steel gradien t chains are used in conjunction with the encapsulated dosimeters to complete the mapping of the neutron envir onment between the discrete locations chosen for s pectrum determinations.In choosing sensor set locations for the EVND Program, advantage is taken of the octant symmetry typical of pressurized water reactors. That is, subject to access limitations, spectrum measurements are concentrated to obtain azimuthal flux distributions in a single forty-five degree sector. Placement of the descrete sensor sets is such that spectrum determinations ar e made at various locations (5, 15, 30, and 40 degrees) on the midplane of the active co re to measure the spectrum changes ca used by the vaying amounts of water located between the core and the reactor vessel. The varied thickness of water is due to the stai r step shape of the reactor core periphery relative to the cylidrical geometry of the reactor internals and vessel and to the local nature of the neutron pads. The remaining sensor sets may be posit ioned opposite the top and bottom of the active core or opposite key reactor vessel welds at particular azimuthal angles of interest. The intent is to measure axial variations in neutron spectrum over the core height, particularly near the top of the fuel where ba ck-scattering of neutrons from primary loop nozzles and reac tor vessel support structures can produce significant differences. At each of the azimuthal locations selected for spectrum measurements, stainless steel gradient chai ns extend over the full hei ght of the active fuel.
CALLAWAY - SP5.3-8Rev. OL-21 5/15 Sensor Sets The EVND Program employs advanced sensor sets that are recommended by and are designed to the latest ASTM neutron dosimetry standards. The sensor sets consist of the encapsulated dosimeters and gradient chains shown on Table 5.3-11, which also lists the neutron reactions t hat are of interest.5.3.1.7Reactor Vessel Fasteners The reactor vessel closure studs, nuts, and washers are designed and f abricated in accordance with the requirements of the ASME Code,Section III. The closure studs are fabricated of SA-540, Class 3, Grade B24.
The closure stud material meets the fracture toughness requirements of the ASME Code,Section III and 10 CFR 50, Appendix G.
Compliance with Regulatory Gui de 1.65, "Materials and Inspections for Reactor Vessel Closure Studs," is discussed in Appendix 3A. Nondestructive examinations are performed in accordance with the ASME Code,Section III.
Bolting materials fracture toughness data is provided in Table 5.3-7
.Refueling procedures require that the studs, nuts, and wa shers be removed from the refueling cavity and st ored at convenient locations prior to refueling ca vity flooding. Studs which cannot be removed are protect ed by enclosures and/
or other means to prevent corrosive damage.
Therefore, the removed reac tor closure studs are never exposed to the borated refueling cavity water and remain ing studs are protected.
Additional protection against the possibility of incurring corrosion effects is assured by the use of a manganese base phos phate surfacing treatment.
The stud holes in the reac tor flange are sealed with special plugs before removing the reactor closure, thus prevent ing leakage of the borated re fueling water into the stud holes. 5.3.2PRESSURE - TEMPERATURE LIMITS 5.3.2.1Limit CurvesStartup and shutdown operating limitations will be based on the properties of the reactor pressure vessel beltline material
- s. Actual material property test data will be used. The methods outlined in Appendix G to Section III of the ASME Code will be employed for the shell regions in the analysis of protection against nonductile failure. The initial operating curves are calculated, assumi ng a period of reactor operat ion such that the beltline material will be limiting. The heatup and cooldown curves are given in the Pressure and Temperature Limits Report (PTLR).
CALLAWAY - SP5.3-9Rev. OL-21 5/15 Beltline material properties degrade with r adiation exposure, and this degradation is measured in terms of the adjusted reference nil-ductility temperat ure, which includes a reference nil-ductility temperature shift (RTNDT).Predicted RTNDT values are derived using two curves: the effect of fluence and copper content on the shift of RTNDT for the reactor vessel steels exposed to 550 F temperature curve and the maximum fluence at 1/4 T (thickness) and 3/4 T location (tips of the code reference flaw when flaw is assumed at inside diameter and outside diameter locations, respectively) curve. For a select ed time of operation, this shift is assigned a sufficient magnitude so that no unirradiated ferr itic materials in other components of the reactor coolant system (RCS) will be limiting in the analysis.
The operating curves including pressure-temperature limita tions are calculated in accordance with 10 CFR 50, A ppendix G and ASME Code,Section III, Appendix G, and WCAP-14040 (Reference 1) requirements. The results of the material surveillance program described in Section 5.3.1.6 will be used to verify that the RTNDT predicted from the effects of the fl uence and copper content curve is appropriate and to make any changes necessary to correct the fluence and copper curves if RTNDT determined from the surveillance program is greater than the predicted RTNDT. Temperature limits for preservice hydrotests and inservice leak and hydrotests will be calculated in accordance with Appendix G of t he ASME Code,Section III. Compliance with Regulatory Gui de 1.99 is discussed in Appendix 3A. 5.3.2.2Operating Procedures The transient conditions that are considered in the design of t he reactor vessel are presented in Section 3.9(N).1.1. These transients are representative of the operating conditions that should prudent ly be considered to occur du ring plant operation. The transients selected form a conservative basis for evaluation of the RCS to insure the integrity of the RCS equipment. Those transients listed as upset condition transients are given in Table 3.9(N)-1. None of these transients will result in pressure-temperature ch anges which exceed the heatup and cooldown limitations , as described in Section 5.3.2.1 and in the PTLR. 5.3.3REACTOR VESSEL INTEGRITY 5.3.3.1Design The reactor vessel is cyli ndrical with a welded hemispher ical bottom head and a removable, bolted, flanged, and gasketed hemispherical upper head. The reactor vessel flange and head are sealed by two hollow metallic O-rings. Seal le akage is detected by CALLAWAY - SP5.3-10Rev. OL-21 5/15means of two leakoff connections: one between the inner and outer ring and one outside the outer O-ring. The vessel contains the core, core suppor t structures, control rods, and other parts directly associat ed with the core. The reactor vessel closure head contains head adaptors. These head adaptors are tubular members, attached by partial penetration welds to the underside of the closure head. The upper end of th ese adaptors are welded to the lower end of a CRDM latch housing or instrumentation port head adapter flange. Inlet and outle t nozzles are located symmetr ically around the vessel.
Outlet nozzles are arranged on the vessel to facilitate opt imum layout of the RCS equipment. The inlet nozzles are tapered from the coolant loop vessel interfaces to the vessel inside wall to redu ce loop pressure drop.
The bottom head of the vessel contains penetration nozzles for connection and entry of the nuclear incore instrumentation. Each nozzle consists of a tubular me mber made of either an Inconel or an Inconel-stainless steel composite tube. Each tube is attached to the inside of the bottom head by a partia l penetration weld. Internal surfaces of the vessel which are in contact with primary coolant are weld overlay with 0.125 inch minimum of stai nless steel or Inconel except for an area approximately
1.5 inches
by 0.75 at approximate location 302.94 o from vessel "0" and 384.89 inches down from the flange surface and an area approximately 0.53 in ches by 0.3 inches at approximate location 185o from vessel "0" and 385 inches down from the flange surface.
The existence of these areas has been evaluated as acceptable.The reactor vessel is designed and fabricated in accordance with the requirements of the ASME Code,Section III. Principal design para meters of the reactor vessel are given in Table 5.3-2. The reactor vessel is shown in Figure 5.3-1. There are no special design features which would prohibit the in-s itu annealing of the vessel. If the unlikely nee d for an annealing operation wa s required to restore the properties of the vessel material opposite the reactor core because of neutron irradiation damage, a metal temperature greater than 650°F for a peri od of 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> maximum would be applied. Various modes of heating may be used, depending on the temperature required.
The reactor vessel materials surveillanc e program is adequate to accommodate the annealing of the reactor vessel. Sufficient specimens are available to evaluate the effects of the anneal ing treatment.
Cyclic loads are introduced by normal power changes, reactor trips, and startup and shutdown operations. These design base cycles are selected for fatigue evaluation and constitute a conservative de sign envelope for the projected plant life. Vessel analysis results in a usage factor that is less than 1.
The design specifications requ ire analysis to prov e that the vessel is in compliance with the fatigue and stress limits of the ASME Code,Section III. The loadings and transients specified for the analysis ar e based on the most severe conditions ex pected during CALLAWAY - SP5.3-11Rev. OL-21 5/15 service. The heatup a nd cooldown rates imposed by the PTLR are 100 F in any one hour, except for cooldown of the pressurizer which is limited to 200F in any one hour. In practice, these operations will occur more slowly. These rates are reflected in the vessel design specifications. 5.3.3.2Materials of ConstructionThe materials used in the fabrication of the reactor vesse l are discussed in Section 5.2.3. 5.3.3.3Fabrication Methods The SNUPPS reactor vessel manufacturer is Combustion Enginee ring Corporation.
The replacement reactor vessel closure head manufacturer is AREVA.
The fabrication methods used in the construc tion of the reactor ve ssel are discussed in Section 5.3.1.2
.5.3.3.4Inspection Requirements The nondestructive examinati ons performed on the reacto r vessel are described in Section 5.3.1.3. 5.3.3.5Shipment and Installation The reactor vessel is shipped in a horizontal position on a shipping sled with a vessel-lifting truss assembly. All vessel openi ngs are sealed to prevent the entrance of moisture, and an adequate quantity of desiccant bags is placed inside the vessel. These are usually placed in a wire mesh basket attached to the vessel cover. All carbon steel surfaces, except for the vessel support surfaces and the top surface of the external seal ring, are painted with a heat-resistant paint before shipment.
The closure head is also shipped with a sh ipping cover and skid. The shipping cover encloses and protects the co ntrol rod mechanism housings. The shippi ng cover is sealed and pressurized with nitrogen to prevent the entranc e of moisture and oxygen, and an adequate quantity of desiccant bags is placed inside. A lifting frame is provided for handling the vessel head. 5.3.3.6Operating ConditionsOperating limitations for the reactor vessel are presented in Section 5.3.2, as well as in the PTLR.
In addition to the analysis of primary components discussed in Section 3.9(N).1.4 , the reactor vessel is further qualified to ensure against unstabl e crack growth under faulted conditions. Actuation of th e emergency core cooling system (ECCS) following a CALLAWAY - SP5.3-12Rev. OL-21 5/15 loss-of-coolant accident produces relatively high thermal st resses in regions of the reactor vessel, which come into contact with ECCS water. Primary consideration is given to these areas, includi ng the reactor vessel beltli ne region and the reactor vessel primary coolant nozzle, to ens ure the integrity of the reactor vessel under this severe postulated transient.
For the beltline region, significant developments have recent ly occurred in order to address Pressurized Thermal Shock (PTS) events. On the basis of recent deterministic and probabilistic studies, taking U.S. PWR operating experi ence into account, the NRC staff concluded that conserva tively calculated screening crit erion values of less than 270 F for plate material and axial welds, and less than 300 F for circumferential welds, present an acceptably low risk of vessel failure from PTS events. These values were chosen as the screening criterion in the PTS Rule 10CFR50.61 for operating plants.
Conservative equations chosen by the NRC staff for the calculation of RTPTS for the purpose of comparison with the screening criterion are presented in par agraph (b) (2) of 10CFR50.61. Details of th e analysis method and the basis for the PTS Rule can be found in SECY-82-465.
The reactor vessel beltline materials are specified in Section 5.3.1. The design basis fluence of 3.29 x 10 19 n/cm 2, which is the design basis fluence at the vessel inner radius after 32 EFPY for the peak azimuthal location, was used for calculating the RTPTS values. RTPTS is the reference temper ature as calculated by the method presented in paragraph (b) (2) of 10CFR50.61. The PTS Rule states that this method of calculating RTPTS should be used in reporting values used to be compar ed to the abov e screening criterion. The screening criter ion will not be exce eded using the met hod of calculation prescribed by the PTS Rule for the vessel design lifetime. The material properties, initial RTNDT and end-of-life RTPTS values are listed in Table 5.3-9. The materials identified in Table 5.3-9 are those materials that are exposed to high fluenc e levels at the beltline region of the reactor vessel an d are, therefore, the subject of the PTS Rule. These materials, therefore, are a subset of the material s identified in Section 5.2.3. The plant-specific calculated flu ence at 35 EFPY for this peak azimuthal location is 2.02 x 10 19 n/cm 2.Note: Based on the Callaway Plant performance, the plant is predicted to reach 35 EFPY at the end of its operating li fe. The principles and procedures of linear elastic fracture mechanics (LEFM) are used to evaluate thermal effects in the regions of interest. The LEFM approach to the design against failure is basically a stress intensity consideration in which criteria are established for fracture instability in the presence of a crack. Consequently, a basic assumption employed in LEFM is that a crack or crack-like defect exists in the structure.
The essence of the approach is to relate the stress field developed in the vicinity of the crack tip to the appli ed stress on the structure, the material properties, and t he size of defect necessary to cause failure.
The elastic stress field at the crack tip in any cracked body can be described by a single parameter designated as the stress intensity factor, K. The magn itude of the stress CALLAWAY - SP5.3-13Rev. OL-21 5/15 intensity factor K is a function of the geometry of the body containing the crack, the size and location of the crack, and the magni tude and distribution of the stress.
The criterion for failure in the presence of a crack is that failure will occur whenever the stress intensity factor exceeds some critical value. Fo r the opening m ode of loading (stresses perpendicular to the major plane of the crack), t he stress intensity factor is designated as K I and the critical stress inte nsity factor is designated K IC. Commonly called the fracture toughness, K IC is an inherent material pr operty which is a function of temperature and strain rate.
Any combination of applied load, structur al configuration, crack geometry, and size which yiel ds a stress intensity factor K IC for the material will result in crack instability.
The criterion of the applicabilit y of LEFM is based on plasti city considerations at the postulated crack tip. Strict applicability (as defined by ASTM) of LEFM to large structures where plane strain conditions prevail requires t hat the plastic zone developed at the tip of the crack does not exceed 2.25 percent of the crack depth. In the present analysis, the plastic zone at th e tip of the postulated crack can reach 20 percent of the crack depth. However, LEFM has been successfully used to provide conservative brittle fracture prevention evaluations, even in cases where strict ap plicability of the theory is not permitted due to excessive plasticity. Recently, experimental results from the Heavy Section Steel Technology (HSST) Program intermediate pressure vessel tests have shown that LEFM can be appli ed conservatively as long as the pressure component of the stress does not exceed t he yield strength of the material. T he addition of the elastically calculated thermal stresses, which results in tota l stresses in excess of the yield strength, does not affect the conservatism of the results, provided that these thermal stresses are included in the evaluation of the stress intensity factors. Therefore, for faulted conditions analyses, LEFM is considered applicabl e for the evaluation of the vessel inlet nozzle an d beltline region.
In addition, it has been well established that the crack propagation of existing flaws in a structure subjected to cyclic loading can be defined in terms of fracture mechanics parameters. Thus, the principles of LEFM are also applicable to fatigue growth of a postulated flaw at the vessel inlet nozzle and bel tline region. Additional details on this meth od of analysis of reactor vessels under severe transients are given in Reference 2.5.3.3.7Inservice Surveillance The internal and external su rfaces of the reactor vesse l are accessible for periodic inspection. Visual and/or nondestructive te chniques are used. Du ring refueling, the vessel cladding is capable of being inspected in certain areas betw een the closure flange and the primary cool ant inlet nozzles, and, if deemed necessary, the core barrel is capable of being removed, making the enti re inside vessel surface accessible.
CALLAWAY - SP5.3-14Rev. OL-21 5/15 The closure head is examined visually during each refueling. Opti cal devices permit a selective inspection of the cladding, control r od drive mechanism nozzles, and the gasket seating surface. The closure studs and nuts can be inspect ed periodically using visual, magnetic particle, and ul trasonic techniques. The closure studs, nuts, washers, and the vess el flange seal surface, as well as the full penetration welds in the following areas of the installed reac tor vessel, are available for nondestructive examination: a.Vessel shell - from the inside and outside surfacesb.Primary coolant nozzles - from the inside and outside surfaces*c.Closure head - from the inside and outside surfaces. Bottom head - from the inside and outside surfacesd.Field welds between the reactor ve ssel nozzle safe end s and the main coolant piping - from the inside and outside surfaces The design considerations which have been incorporated into the system design to permit the above inspec tion are as follows: a.All reactor internals ar e completely removable.
The tool s and storage space required to permit these inspections are provided. b.The closure head is stored dry on the reactor operating deck during refueling to facilitate di rect visual inspection. c.Reactor vessel studs, nuts, and wa shers can be remov ed to dry storage during refueling. Studs which cannot be removed are independently evaluated.d.Access is provided to the reactor ve ssel nozzle safe ends. The insulation covering the nozzle-to-pipe welds may be removed. e.Reactor cavity is designed to allow access to the outside surface of the vessel. Tracks are instal led to allow mechanical equipment to inspect the vessel surface. The reactor vessel presents access problems because of the radiation levels and remote underwater accessibility to this component.
Because of these limitations on access to the reactor vessel, several steps have been incorpor ated into the design and *Only partial outside diam eter coverage is provided.
CALLAWAY - SP5.3-15Rev. OL-21 5/15manufacturing procedures in preparation for the periodic nondestructive tests, which are required by the ASME in service inspection code. These are: a.Shop ultrasonic examinatio ns are performed on all internally clad surfaces to an acceptance and repair standard to assure an adequate cladding bond to allow later ultrasonic testing of the base metal from inside surface. The size of cladding bond defect allowed is 1/4 inch by 3/4 inch with the greater direction parallel to the weld in the region bounded by 2T (T = wall thickness) on both sides of each full penetration pressure boundary weld. Unbounded areas exceeding 0.442 square inches (3/4 inch diameter) in all other regions are rejected. b.The design of the reacto r vessel shell is an unclu ttered cylindrical surface to permit future positioni ng of the test equipment without obstruction. c.The weld deposited clad surface on both sides of the welds to be inspected is specifically prepared to assure meaningful ultrasonic examinations. d.During fabrication, a ll full penetration ferritic pr essure boundary welds are ultrasonically examined in additi on to Code examinations. e.After the shop hydrostatic testing, all full penetration ferritic pressure boundary welds, as well as the nozzle to safe end welds, are ultrasonically examined from both the inside and outside diameter s in addition to ASME Code,Section III requirements.
The vessel design and constr uction enables inspection in accordance with the ASME Code,Section XI. The reactor vessel inservice inspection program is detailed in Chapter 16 , the Inservice Inspection Program, and the PTLR. 5.
3.4REFERENCES
1."Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," WCAP-14040 NP-A.1a.Terek, E., Senkewitz, T.S., and Singer, L.R., "Wes tinghouse Archived Reactor Vessel Materials," WCAP-15151, December 1998.2.Buchalet, C., Bamford, W. H., and Chirigos, J. N., "Method for Fracture Mechanics Analysis of Nuclear Reactor Vessels Under Severe Thermal Transients," WCAP-8510, December 1975.3.Lott, R.G., Vanichko, S.E., Locante, J., Schmertz, J.C., "Analysis of Capsule U From the Union Electric Company Callaway Plant Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP-11374, Rev. 1, June 1987.
CALLAWAY - SP5.3-16Rev. OL-21 5/154.Terck, E., Anderson, S.L., Madeyski, A., "Analysis of Capsul e Y From the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP-12946, June 1991.5.Torek, E., Perock, J. D., Williams, J. F., "Analysis of Capsule V from the Union Electric Company Unit 1 Reactor Ve ssel Radiation Surveillance Program,"
WCAP-14895, July 1997.6.Laubham, T. J., Roberts, G.
K. Harik, E., "Analysis of Capsule X from AmerenUE Callaway Unit 1 Reactor Vessel Radiat ion Surveillance Progr am," WCAP-15400, June 2000.
CALLAWAY - SP Rev. OL-21 5/15TABLE 5.3-1 REACTOR VESSEL QUALITY ASSURANCE PROGRAM RT*UT*PT*MT*Forgings FlangesYesYes Studs and nutsYesYes CRD latch housingYesYes CRD head adapter tubeYesYes Instrumentation tubeYesYes Main nozzlesYesYes Nozzle safe endsYesYes PlatesYesYes Weldments Main seamYesYesYes
CRD head adapter to closure head connectionYes Instrumentation tube to bottom head connectionYes Main nozzleYesYesYes CladdingYesYesNozzle to safe endsYesYesYes
CRD latch housing to CRD head adapter tubeYesYes All full penetration ferritic pressure boundary welds accessible after hydrotestYesYes Full penetration nonferritic pressure boundary welds accessible after hydrotest (Nozzle to safe ends)YesYes
Seal ledgeYes Head lift lugsYes Core pad weldsYes CALLAWAY - SPTABLE 5.3-1 (Sheet 2)
Rev. OL-21 5/15 NOTE: Base metal weld repairs as a result of UT, MT, RT, and/or PT indications shall be cleared by the same NDE technique/ procedure by which the indications were found. The repair shall meet all Section III requirements.
In addition, UT examinati on per the in-process/post-hydro UT requirements shall be performed on the following: 1.Base metal repairs in the core region. 2.Base metal repairs in the ISI zone (1/2 T). *RT - Radiographic UT - Ultrasonic PT - Dye Penetrant MT - Magnetic Particle CALLAWAY - SP Rev. OL-21 5/15TABLE 5.3-2 REACTOR VESSEL DESIGN PARAMETERS Design/operating pressure, psig2,485/2,235Design temperature, F650 Overall height of vessel and cl osure head, bottom head outside diameter to top of control rod mechanism latch housing, ft-in.47-11.8 Thickness of RPV head insulation, minimum, in.4.3 Number of reactor closure head studs **54 Closure head studs are provided. 53 Studs are required to be tensioned for oper-ation in modes 1 thru
- 4. (See ULNRC-1663 fo r supporting evaluation).
54 Diameter of reactor closure head/studs, minimum shank, in.6-13/16Outside diameter of flange, in.205
Inside diameter of flange, in.167Outside diameter at shell, in.190-1/2 Inside diameter at shell, in.173 Inlet nozzle inside diameter, in.27-1/2Outlet nozzle inside diameter, in.29 Clad thickness, minimum, in.1/8 Lower head thickness, minimum, in.5-3/8 Vessel beltline thickness, minimum, in.8-5/8 Closure head thickness, minimum, in.7Nominal water volume, ft 3 3,700 CALLAWAY - SP Rev. OL-13 5/03TABLE 5.3-3 Deleted.
CALLAWAY - SP Rev. OL-21 5/15TABLE 5.3-4 CALLAWAY UNIT 1 REACTOR VESSEL MATERIAL PROPERTIES*Major working direction**Normal to major working direction Avg.UpperShelfCOMPONENTCODENO.MATERIALSPEC.NO.Cu(%)P(%)T NDT (F)RT NDT (F)NMWD**(FT-LB)MWD*(FT-LB)Closure Head10DW96-1A508, G3, CL. 10.030.006-40-40207-
Vessel FlangeR2701-1A508 CL. 2 -0.010 40 40123 -
Inlet NozzleR2702-1A508 CL. 2 -0.013 10 10138 -
Inlet NozzleR2702-2A508 CL. 2 -0.011 10 10141 -
Inlet NozzleR2702-3A508 CL. 2 -0.009-10-10139 -
Inlet NozzleR2702-4A508 CL. 2 -0.010-10-10134 -
Outlet NozzleR2703-1A508 CL. 2 -0.010-10-10130 -
Outlet NozzleR2703-2A508 CL. 2 -0.009 10 10108 -
Outlet NozzleR2703-3A508 CL. 2 -0.004 10 10126 -
Outlet NozzleR2703-4A508 CL. 2 -0.006 0 0122 -
Nozzle ShellR2706-1A533B, CL. 10.050.010 10 20103 -
Nozzle ShellR2706-2A533B, CL. 10.060.009 0 30 88 -
Nozzle ShellR2706-3A533B, CL. 10.080.011 0 30101 -
Inter. ShellR2707-1A533B, CL. 10.040.008-40 40 78 99 Inter. ShellR2707-2A533B, CL. 10.050.008-50 10100 121 Inter. ShellR2707-3A533B, CL. 10.060.010-40-10 99 122 Lower ShellR2708-1A533B, CL. 10.070.006 0 50 82 95 Lower ShellR2708-2A533B, CL. 10.050.007-30 10105 130 Lower ShellR2708-3A533B, CL. 10.070.006-10 20101 122 Bottom Head TorusR2714-1A533B, CL. 10.150.010-20-20139 -
Bottom Head DomeR2715-1A533B, CL. 10.170.011-40-40152 -
Inter. and lower shellG2.03SAW0.040.008-60-60143 -
long. weld seams Inter. to lower shellE3.14SAW0.040.006-60-60112 -
girth weld seam Weld HAZ - - - --80-70144 -
CALLAWAY - SP Rev. OL-13 5/03TABLE 5.3-5 Deleted.
CALLAWAY - SP Rev. OL-13 5/03TABLE 5.3-6 Deleted.
CALLAWAY - SP Rev. OL-14 12/04TABLE 5.3-7 CALLAWAY UNIT 1 REACTOR VESSEL CLOSURE HEAD BOLTING MATERIAL PROPERTIESClosureHeadStudsHeatNo.MaterialSpec. No.Bar No.0.2% YieldStrength (Ksi)UltimateTensile Strength (Ksi)Elongation
(%)Reductionin Area (%)Energyat 10 F(FT LB)LateralExpansion (MILS)BHN84299SA540, B24437143.5158.516.048.152, 51, 5131, 27, 29331 84299SA540, B24437-1138.0155.017.051.451, 51, 5329, 31, 30341 84299SA540, B24439141.5156.516.552.553, 54, 5329, 30, 32331 84299SA540, B24439-1140.5154.516.049.555, 53, 5233, 32, 29331 83320SA540, B24443139.0156.017.553.351, 50, 5233, 29, 30341 83320SA540, B24443-1141.0157.017.554.149, 49, 4929, 32, 30331 83320SA540, B24447140.5156.017.053.849, 50, 4932, 30, 2934183320SA540, B24447-1144.0159.017.552.149, 49, 4932, 31, 3033183320SA540, B24451141.0155.017.053.652, 51, 5132, 30, 29341 83320SA540, B24451-1141.0156.017.552.251, 50, 5130, 30, 31341 83320SA540, B24456141.0157.017.051.748, 50, 4831, 31, 30331 83320SA540, B24456-1139.8154.017.053.549, 47, 4733, 27, 30331ClosureHeadNuts&Washers63182SA540, B24132148.0162.017.557.351, 52, 5131, 32, 3033163182SA540, B24132-1148.7162.017.054.749, 48, 4929, 26, 29331 63182SA540, B24133147.2161.017.055.252, 50, 5131, 30, 30321 63182SA540, B24133-1149.2162.517.554.751, 51, 4929, 31, 27331 63182SA540, B24135147.6161.017.053.049, 49, 5128, 29, 30321 63182SA540, B24135-1143.2157.017.555.255, 54, 5233, 32, 31321 63182SA540, B24137145.0159.016.554.854, 54, 5333, 33, 29331 63182SA540, B24137-1147.0160.017.055.754, 55, 5434, 36, 33321 63182SA540, B24143145.0159.018.058.155, 54, 5433, 32, 32331 63182SA540, B24143-1147.0160.017.057.354, 50, 5233, 29, 30321 63182SA540, B24145145.0159.017.056.054, 54, 5534, 35, 34321 63182SA540, B24145-1146.2159.717.057.056, 55, 5436, 35, 36331 63182SA540, B24148144.0157.517.556.556, 55, 5533, 34, 34331 63182SA540, B24148-1148.6162.017.055.652, 51, 5233, 28, 30321 63182SA540, B24150144.7158.017.555.755, 55, 5433, 30, 31331 63182SA540, B24150-1145.7160.017.056.553, 50, 5233, 30, 31331 CALLAWAY - SP Rev. OL-13 5/03TABLE 5.3-8 Deleted.
CALLAWAY - SP Rev. OL-1610/07TABLE 5.3-9 CALLAWAY REACTOR VESSEL VALUES FOR ANALYSIS OF POTE NTIAL PRESSURIZED THERMAL SHOCK EVENTS 10CFR50.61 Note: See also Table 5.3-4
.* Indicates numbers were calculat ed using surveillance capsule data.
- Average material/weld pr operties from WCAP 14894.Material Code No.Cu%C-EAnal Ave**
Ni%C-EAnal Ave**
Initial RTNDT(°F) End of Life RTPTS (°F)Intermediate ShellR2707- 1.04.05.57.58 40111- 2.05.06.59.61 10 88- 3.06.06.61.62- 10 68Lower ShellR2708- 1.07.07.59.58 50 97*- 2.05.06.57.57 10 88- 3.07.08.59.62 20115Inter. and Lower Shell Long. Weld SeamsG2.03.04.06- 60 15.48* Inter. to Lower Shell Girth Weld SeamsE3.14.04.04- 60 15.88* Ave. for All Beltline Weld Metal-.04-.06--
CALLAWAY - SP Rev. OL-21 5/15TABLE 5.3-10 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE For additional information, s ee References 3, 4, 5, and 6.CAPSULE NUMBERVESSEL LOCATION LEAD FACTOR (a)(a)Updated in Section 6 of Reference 6.WITHDRAWAL TIME (EFPY)(b)(b)Effective Full Power Years (EFPY) from plant startup. Capsule Number U, Y, V, and X EFPYs referenced from WCAP-15400, Capsule Number W, Z EFPY generated by Callaway Reactor Engineering.U58.5°4.421.05 (Removed)Y241°3.854.6 (Removed)
V61°3.979.85 (Removed)X238.5°4.3412.4 (c) (Removed)(c)At 12.4 EFPY, the fluence at Capsule X is approximat ely equal to the calculated peak reactor vessel surface fluence at 51 EFPY.W121.5°4.2925.8 (Removed)Z301.5°4.2916.53 (Removed)(d)(d)Capsule Z removed during Refuel 13. Curr ently in long-term st orage in spent fuel pool.
CALLAWAY - SP Rev. OL-21 5/15TABLE 5.3-11 NEUTRON DOSIMETRY REACTIONS OF INTERESTMATERIALREACTION OF INTERESTNEUTRON ENERGY RESPONSES (a)(a)Energies between which 90% of activity is produced (235U fission spectrum).
PRODUCT HALF-LIFE DOSIMETER CAPSULE POSITION (b)(b)"B" denotes bare iron and cobalt radiometric monitors, while "Cd" denotes cadium-shielded radiometric monitors. Cadium-shielded foils include foils of the following metals: copper, titanium, iron, nickel, niobium, and cobalt-aluminum. Cadium-shielded fast fission reactions include 238 U and 237Np in vanadium oxide encapsulated detectors.GRADIENT C HAIN(c)(c)These stainless steel bead chains connect and support the dosimeter capsules containing the radiometric monitors. The segmented chains provide iron, nickel, and cobalt reactions that are used to complete the determination of the axial and azimuthal gradients. The high purity iron, nickel, and cobalt-aluminum foils contained in the multiple foil sensor sets provide a direct correlation with the measured reaction rates from these gradiant chains. Cross-comparisons permit the use of the gradient measurements to derive neutron flux distributions in the reactor cavity with a high level of confidence.Copper 63Cu(n,)60Co4.53-11.0 MeV5.271 y2-CdNoTitanium 46Ti(n,p)46Sc3.70-9.43 MeV83.79 d2-CdNoIron 54Fe(n,p)54Mn2.27-7.54 MeV312.3 d1-B & 2-CdYesNickel 58Ni(n,p)58Co1.98-7.51 MeV70.82 d2-CdYes 238 U(d)238 U(n,f)137Cs1.44-6.69 MeV30.07 y3-CdNo 93 Nb 93 Nb(n,n')93mNb0.95-5.79 MeV16.13 y3-CdNo 237 Np (d)(d)Vanadium-encapsulated 238U and 237Np fission monitors are currently unavailable.
237 Np(n,f)137Cs0.68-5.61 MeV30.07 y3-CdNoCobalt-Al 59Co(n,Y)60CoThermal5.271 y1-B & 2-CdYes CALLAWAY - SP5.4-1Rev. OL-21c 1/165.4COMPONENT AND SUBSYSTEM DESIGN5.4.1REACTOR COOLANT PUMPS5.4.1.1Design Bases The reactor coolant pump prov ides an adequate core cooling flow rate for heat transfer to maintain a departure from nucleate boilin g ratio (DNBR) greate r than 1.17 within the parameters of operation. T he required net positive suctio n head is by conservative pump design always less than that available by system design and operation. Sufficient pump rotation inertia is provided by a flywheel, in conjunction with the impeller and motor assembly, to provide adequate flow during coastdown. This forced flow following an assumed loss of pump power, and the subsequent natural circulation effect provides the core with adequate cooling flow.
The reactor coolant pump mo tor is tested, with out mechanical damage, at overspeeds up to and including 125 percent of normal speed. The retention of integrity of the flywheel during a LOCA is dem onstrated in Reference 1. Steam/water tests planned jointly by Westinghouse, Framat one, and the French Atomic Energy Commission (CEA) are di scussed in Reference 2. The ultimate use of the data from this testing w ill be to develop an empirical two-phase flow pum p performance model. It is expected that this new model will confirm that the present pump model conservatively predicts performance in all LOCA conditions and thus increase the safety margin available in the emergency core c ooling system (ECCS) and reactor coolant pump overspeed analyses.
The pump/motor system is designed for the SSE at the site. 5.4.1.2Pump Description5.4.1.2.1Design Description The reactor coolant pump is shown in Figure 5.4-1. The reactor cool ant pump design parameters are given in Table 5.4-1. Code and material requirements are provided in Section 5.2. The reactor coolant pump is a vertical, single stage, controlled leakage, centrifugal pump designed to operate at high tem peratures and pressures. The pump consists of three ma jor sections. They are the hydraulics, the seals, and the motor. a.The hydraulic section consists of the casing, thermal barrier, flange, impeller/diffuser, and diffuser adapter.
CALLAWAY - SP5.4-2Rev. OL-21c 1/16b.The shaft seal section consistes of three primary devices. They are number 1 controlled-leakage, film-riding fa ce seal, and the number 2 and number 3 rubbing face seals. These seals are contained within the thermal barrier heat exchanger asse mbly and seal housing. Collectively, they provide a pressure breakdown from the reactor coolant system (RCS) pressure to ambient conditions. A fourth sealing device called a shutdown seal is housed within the number 1 seal area and is passively actuated by high temperature if s eal cooling is lost.c.The motor is a drip-proof squirrel cage induction mo tor with a vert ical solid shaft, an oil lubricated double-acting Ki ngsbury type thrust bearing, upper and lower oil lubricated radial gui de bearings, and a flywheel. Additional components of the pump are the shaft, pump r adial bearing, thermal barrier heat exchanger, coupling, spool piece, and motor stand. 5.4.1.2.2Descripti on of Operation Reactor coolant enters the su ction nozzle, is directed to the impeller by the diffuser adapter, is pumped through the diffuser, and exits through the discharge nozzle. Seal injection flow, under slight ly higher pressure than the reactor coolant, enters the pump through a connection of the thermal barrier flange and is directed into the plenum between the thermal barrier housing and the shaft. The flow splits with a portion flowing down the shaft through the radi al bearing and into the RCS; the remainder flows up the shaft through the seals.
Component cooling water is pr ovided to the thermal barrier heat exchanger. During normal operation, the thermal barrier limits the heat transfer from hot reactor coolant to the radial bearing and to the seals. In addition, if a loss of seal injection flow should occur, the thermal barrier he at exchanger cools reactor coolant to an acceptable level before it enters the beari ng and seal area.
The reactor coolant pump shaf t sealing system can operate acceptably with either seal water injection or co mponent cooling water al one for an unlimited ti me. As described in Sections 9.2.2 and 9.3.4 the component cooling water and the seal water injection paths provide diverse cooling means which precludes seal failures due to any single failure or due to the effects of an SSE.
The reactor coolant pump mo tor bearings are of conventi onal design.
The radial bearings are the segmented pad type, and the thrust bearing is a double-acting Kingsbury type. All are oil lubricated. Component coolin g water is supplied to the external upper bearing oil cooler and to the integral lower bearing oil cooler. The reactor coolant pump motor bearings ar e qualified for 10 minutes operation without component cooling water with no resultant damage.
CALLAWAY - SP5.4-3Rev. OL-21c 1/16 The motor is a water/air cooled, Class F t hermalastic epoxy insulated, squirrel cage induction motor. The rotor and stator are of standard construction and are cooled by air. Six resistance temperature detectors are imbedded in the stator windings to sense stator temperature. The top of the motor consists of a flywheel and an antireverse rotation device. The internal parts of the motor are cooled by air. Integral vanes on each end of the rotor draw air in through cooling slot s in the motor frame. This air passes thro ugh the motor with particular emphasis on the stator end turns. It is then routed to the external water/air heat exchangers, which are supplied with component cooling water. Each motor has two such coolers, mounted diametrically opposed to each other. In passing through the coolers, the air is cooled to bel ow 122°F so that little heat is rejected to the containment from the motors.
Each of the reactor coolant pumps is equipped fo r continuous monitoring of reactor coolant pump shaft and frame vibration levels. Shaft vibr ation is measured by two relative shaft probes mounted on top of the pump seal housin g; the probes are located 90 degrees apart in the same horizontal plane and mounted near the pump shaft. Frame vibration is measured by two velocity se ismoprobes located 90 degrees apart in the same horizontal plane and mounted at the top of the motor support st and. Proximeters and converters linearize the pr obe output, which is display ed on a monitor in the control room. The monitor displays t he vibration levels for both relative probes and both seismoprobes for each reactor coolant pump; manual selecti on allows the monitoring of gap voltages of individual probes. Indicator lights display alert and danger limits of vibration.A removable shaft segment, the spool piece, is locat ed between the motor coupling flange and the pump c oupling flange; the spool piece allows removal of the pump seals with the motor in place. The pump internals, motor, and motor stand can be removed from the casing without disturbing the reactor c oolant piping. The fl ywheel is available for inspection by removing the cover. All parts of the pump in contact with the reactor coolant are austenitic stainless steel, except for seals, bearings, and special parts. 5.4.1.3Design Evaluation5.4.1.3.1Pump PerformanceThe reactor coolant pumps are sized to deliver flow at ra tes which equal or exceed the flow rates required for core cooling. Initial RCS tests confirm the total delivery capability.
Thus, assurance of adequate forc ed circulation coolant flow is provided prior to initial plant operation.
CALLAWAY - SP5.4-4Rev. OL-21c 1/16The estimated performance characteristic is shown in Figure 5.4-2. The "knee" at about 45-percent design flow introduces no operational restrictions, since the pumps operate at full flow.
The reactor trip system assu res that pump operation and core cooling capability are within the assumptions used for loss of flow analyses (See Chapter 15.0
). Long-term tests have been conduc ted on less than full scale pr ototype seals, as well as on full size seals. Operating plants continue to dem onstrate the satisfactory performance of the controlled leakage shaft seal pump design. The support of the stati onary member of the num ber 1 seal ("seal ri ng") is such as to allow large deflections, both axial and tilting, while still maintaining its controlled gap relative to the seal runner. Even if all the graphite were removed from the pump bearing, the shaft could not def lect far enough to cause opening of the controlled leakage gap.
The "spring-rate" of the hydraulic forces associated with the main tenance of the gap is high enough to ensure that the ring follows the runner under very rapid shaft deflections. Testing of pumps with the number 1 seal entirely bypassed (full system pressure on the number 2 seal) shows that sm all (approximately 4 to 12 gpm) leakage rates would be maintained for a period of time sufficient to secure the pump. Even if the number 1 seal were to fail entirely during nor mal operation, the number 2 seal would maintain these small leakage rates if the proper action is taken by the operator. An increase in number 1 seal leakoff rate will warn the plant operator of number 1 seal damage. Following warning of excessive seal l eakage conditions, the plant operator will take corrective actions. Gross leakage from the pump does not occur if these procedures are followed. Loss of offsite power causes loss of power to the pum p and causes a temporary stoppage in the supply of seal injection flow to the pum p and also of the component cooling water flow to the pump and motor. The emergency diesel generators are started automatically due to loss of offsite power so t hat seal injection flow is provided by the ECCS charging pumps. Component cooli ng water flow is subsequently restored automatically, within 2 minut es. Load shedding and sequ encing is discussed in Section 8.3. In the event of a loss of all AC power and/
or loss of all seal cooling, the shutdown seal (SDS) will actuate on high seal cooling temperature to limit leakage from the RCP seal package. Leakage is limited when a thermal actuator retracts and causes the SDS piston ring and polymer ring to clamp down around the pump shaft.5.4.1.3.2Coastdown Capability It is important to reactor prot ection that the reactor coolant flow is maintained for a short time after a pump trip in order to remove heat stored in the fuel elements of the core. In order to provide this flow after interruption of power to the pump, each r eactor coolant pump is provided with a flywheel. The rotating inertia of the pump, motor, and flywheel is CALLAWAY - SP5.4-5Rev. OL-21c 1/16employed during the coastdown period to continue the reactor coolant flow. An inadvertent early actuation of the SDS on the pump shaft, with the shaft still rotating, will not adversely impact RCP coastdown. The coastdown flow transients are provided in the figures in Section 15.3. The coastdown capability of the pumps is maintained even under the most adverse case of a blackout coincident with the SSE. Core flow transients and figures are provided in Section 15.3.1 5.4.1.3.3Bearing IntegrityThe design requirements for t he reactor coolant pump bearin gs are primarily aimed at ensuring a long life with negligible wear, so as to give accurate a lignment and smooth operation over long periods of time. The surf ace-bearing stresses are held at a very low value, and even under the most severe seismic transients remain bel ow stress values that can be adequately carried for short periods of time. Because there are no established criteria for short-time stre ss-related failures in such bearings, it is not possible to make a meaningful quantification of such parameters as margins to failure, safety factors, etc. A qualitative analysis of the bearing design, embodying such considerations, gives assu rance of the adequacy of the bearing to operate without failure.
Low oil levels in the lube oil sumps signal alarms in the control room. Each motor bearing contains embedded temperat ure detectors, and so initiation of failure, separate from loss of oil, is indica ted and alarmed in the contro l room as a high bearing temperature. This requires pump shutdown. If these indications are ignored, and the bearing proceeded to failure, the low melting point of Babbitt metal on the pad surfaces ensures that sudden seizure of the shaft will not occur. In this event, the motor continues to operate, as it has sufficient reserve capacity to drive the pump under such conditions. However, the high torque required to drive the pump will require high current which will lead to the motor being shutdown by the electric al protection systems. 5.4.1.3.4Locked Rotor It may be hypothesized that the pump impeller might severely rub on a stationary member and then seize. This constitutes a loss-of-coolant flow in the loop. Analysis has shown that under such conditions, assuming instantaneous seizure of the impeller, the pump shaft fails in torsion ju st below the coupling to the motor, thus disengaging the flywheel and motor from the shaft. Following such a postulated seiz ure, the motor continues to run without any overspeed, and the flywheel maintains its integrity, as it is still supported on a shaft with two bearings. Flow transients are provided in the figures in Section 15.3.3 for the assumed locked rotor.
There are no credible sources of shaft seizure other t han impeller rubs. A sudden seizure of the pump bearing is precluded by graphite in the bearing.
Any seizure in the seals results in a shearing of the antirotation pin in the seal ring. Further, an inadvertent actuation of the shutdown seal on the shaft has no significant effect on pump/shaft CALLAWAY - SP5.4-6Rev. OL-21c 1/16rotation and will not interrupt core cooling flow provided by the RCP. The motor has adequate power to continue pump operation even after the above occurrences.
Indications of pump malfunctio n in these conditions are init ially by high temperature signals from the bearing water temperature detector, and excessive number 1 seal leakoff indications, and offscale number 1 seal leakoff indications respectively. Following these signals, pump vibration levels are checked. Exce ssive vibration, excessive number 1 seal leak-off, and high bearing inlet temperature indi cate mechanical trouble.
Administrative procedures provide for shutting down the affected pu mp or tripping the reactor based upon pre-established criterion. 5.4.1.3.5Critical SpeedThe reactor coolant pump shaft is designed so that its operating speed is below its first critical speed. This shaft design, even under the most severe postulated transient, gives low values of actual stress. 5.4.1.3.6Missile GenerationPrecautionary measures taken to preclude missile formati on from primar y coolant pump components assure that the pumps will not produce missiles under any anticipated accident condition. Each component of the primary pump motors has been analyzed for missile generation. Any fragments of the motor rotor would be contai ned by the heavy stator. The same conclusion applies to t he pump impeller because the small fragments that might be ejected would be contained in the heavy casi ng. Further discussion and analysis of missile generation is contained in Reference 1. 5.4.1.3.7Pump Cavitation
The minimum net positive suct ion head required by the reactor coolant pump at running speed is approximately a 192-foot head (approximately 85 psi). In order for the controlled leakage seal to operate correctly, it is necessary to require a minimum differential pressure of approximately 200 psi across the number 1 seal. This corresponds to a primar y loop pressure at which the mini mum net positive suction head is exceeded, and no limitation on pump operation occurs.5.4.1.3.8Pump Over speed ConsiderationsFor most turbine trips actuated by either the reactor trip syst em or the turbine protection system, the generator and reactor coolant pumps are maintained connected to the external network fo r 30 seconds to prevent any pum p overspeed condition. The exceptions to this are turbine trips actuated by the turb ine protection system for low bearing oil pressure, thrust bearing wear, turbine high vibratio n, high condenser backpressure, and manual trip pushbutton. These turbine trips have no time delay added to the trip circuit.
CALLAWAY - SP5.4-7Rev. OL-21c 1/16 An electrical fault requiring im mediate trip of the generator (with resulting turbine trip) could result in an overspeed condition. However, the turbine control system and the turbine overspeed protection system will limit the overspeed to less than 120 percent. As additional backup, the turbine protection system has a mechanical overspeed protection trip, usually set at about 110 percent (of turbine speed). In the cases wh ere a turbine trip results in an immediat e generator trip or a gen erator trip due to an electrical fault that de-energizes the pump busses, the reactor coolant pump motors will be tr ansferred to offsite power within 6 to 10 cycles.
Further discussion of pump overspeed considerations is cont ained in Reference 1. 5.4.1.3.9Antireverse Rotation Device Each of the reactor coolant pumps is provided wi th an antireverse rotation device in the motor. This antireverse mechanism consists of pawls mounted on the outside diameter of the flywheel, a serrated ratc het plate mounted on the motor frame, a spring return for the ratchet plate, and two shock absorbers.
At an approximate forward speed of 70 rpm, the pa wls drop and bounce across the ratchet plate; as the motor continues to slow, the pawls drag across the ratchet plate. After the motor has slowed and come to a stop, the dropped pawls engage the ratchet plate and, as the motor tends to rotate in the opposite direction, the ratchet plate also rotates until it is stopped by the shock absorbers. The rotor remains in this position until the motor is energized again. When the motor is started, the ratchet plate is returned to its original position by the spring return. As the motor begins to rotate, the pawls drag over the ratchet plate. When the motor reaches sufficient speed, the pawls are bounced into an elevated position and are held in that position by friction resulting from centrifugal forces acting upon the pawls. Wh ile the motor is running at speed, there is no contact between the pawls and ratchet plate.
Considerable plant experience with the design of the antireverse rotation device has shown high reliability of operation. 5.4.1.3.10Shaft Seal Leakage
During normal operation, leakage along the reactor coolant pump shaft is controlled by three shaft seals arranged in series so that reactor coolant leakage to the containment is essentially zero. Injection flow is directed to each reactor coolant pump via a seal water injection filter. It enters the pumps through a connection of the thermal barrier flange and flows to an annulus around the shaft inside the thermal barrier. Here the flow splits: a portion flows down the shaft to cool the bearing and enters the RCS; the remainder flows up the shaft through the seals.
This flow provides a backpressure on the number 1 seal and a controlled flow through the seal. Above the seal, most of the flow leaves the pump via the number 1 seal discharge line. Minor flow passe s through the number 2 seal and leakoff line. A back flush injection from a head tank fl ows into the number 3 seal between its "double dam" seal area. At this point, the flow divide s with half flushing through one side of th e seal and out the number 2 seal leakoff wh ile the remaining half CALLAWAY - SP5.4-8Rev. OL-21c 1/16 flushes through the ot her side and out of the number 3 seal leakoff.
This arrangement assures essentially zero leakage of reactor coolant or trapped gases from the pump.
In the event of a loss of all AC power and/
or loss of all seal c ooling, reactor coolant begins to travel along the RCP shaft and displace the cooler seal injection water. The shutdown seal (SDS) actuates once the number 1 seal package temperature reaches the SDS actuation temperature. SDS actuation controls shaft seal leakage and limits the loss of reactor c oolant through the RCP seal package.5.4.1.3.11Seal Discharge Piping The number 1 seal reduces the coolant pressure to that of the volume control tank. Water from each pump number 1 seal is piped to a common manifold, through the seal water return filter, and thr ough the seal water heat exchanger where the te mperature is reduced to that of the volume control tank. The number 2 and number 3 leakoff lines dump number 2 and 3 seal leakage to the reactor coolant drain tank and the containment sump, respectively.5.4.1.4Tests and InspectionsThe reactor coolant pumps can be inspected in accor dance with the ASME Code,Section XI, for inservice inspection of nuclear reactor coolant systems.
The pump casing is cast in one pi ece, eliminating we lds in the casing.
Support feet are cast integral with the casing to eliminat e a weld region.
The design enables disassembly and removal of the pump internals for usual access to the internal surfaces of the pump casing.
The reactor coolant pu mp quality assurance pr ogram is given in Table 5.4-2. 5.4.1.5Pump Flywheels5.4.1.5.1Pump Flyw heel IntegrityThe integrity of the reactor coolant pump flywheel is assured on the basis of the following design and quality a ssurance procedures. 5.4.1.5.2Design Basis
The calculated stresses at operating speed are based on stresses due to centrifugal forces. The stress resulting from the interference fit of the flywheel on the shaft is less than 2,000 psi at zero speed , but this stress becomes ze ro at approximately 600 rpm because of radial expansion of the hub. The prim ary coolant pumps run at approximately 1,190 rpm and ma y operate briefly at over speeds up to 109 percent (1,295 rpm) during lo ss of load. For conservatism, however, 125 percent of operating CALLAWAY - SP5.4-9Rev. OL-21c 1/16 speed was selected as the design speed for the primary coolant pumps. The flywheels are given a preoperational test of 125 percent of the maximum synchronous speed of the motor. 5.4.1.5.2.1Fabrication and InspectionThe flywheel consists of two thick plates bolted together. The flywheel material is produced by a process that mini mizes flaws in the material and improves its fracture toughness properties, such as vacuum degassing, vacuum melting, or electroslag remelting. Each plate is fabricated from SA
-533, Grade B, Class 1 steel. Supplier certification reports ar e available for all plat es and demonstrate the acceptability of the flywheel material on the basis of the requirements of Regulatory Guide 1.14. Flywheel blanks are flame-cut from the SA-533, Grade B, Class 1 plates with at least 1/2 inch of stock left on the outer and bore radii for machining to final dimensions. The flywheel plates, both before and after assembly , are subjected to magnetic particle or liquid penetrant examinat ion. Included in this examinat ion are all surf aces within a minimum radial distance of 4 inches beyond the final machined bore. This includes the bore surface and the keyways.
The finished flywheels, as we ll as the flywheel material (rolled plate), are subjecte d to 100-percent volumetric ultrasonic ins pection, using procedures and acceptance standards specif ied in Section III of the ASME Code. 5.4.1.5.2.2Material Acceptance Criteria The reactor coolant pump moto r flywheel conforms to the following material acceptance criteria:a.The nil-ductility transition temperature (NDTT) of the flywheel material is obtained by two drop weight test s (DWT) which exhibit "no-break" performance at 20°F in accordance with ASTM E-208. The above drop weight tests demonstrate that the ND TT of the flywheel material is no higher than 10°F. b.A minimum of three Charpy V-notch impact spec imens from each plate shall be tested at ambient (70°F) temperature in accordance with the specification ASME SA-370.
The Charpy V-notch (C V) energy in both the parallel and normal orientat ion with respect to the rolling direction of the flywheel material is at least 50 foot pounds at 70°F, and, therefore, and RTNDT of 10°F can be assumed. An evaluation of flywheel overspeed has been performed which concludes that fl ywheel integrity will be maintained (Ref. 1). Thus, it is concluded that flywheel plate materials are suitable for use on the bases of the suppliers' certification data.
The degree of compliance with Regulatory Guide 1.14 is further discussed in Appendix 3A.
CALLAWAY - SP5.4-10Rev. OL-21c 1/165.4.1.5.2.3AccessabilityThe reactor coolant pump motors are designed so that, by removing the cover to provide access, the flywheel is avail able to allow an inservice ins pection program in accordance with requirements of Sect ion XI of the ASM E Code and the re commendations of Regulatory Guide 1.14. 5.4.1.5.2.4Spin Testing Each flywheel assembly is spin tested at the design speed of t he flywheel, i.e., 125 percent of the maximum synchronous speed of the motor. 5.4.1.5.3Preservice Inspection Post spin testing of reactor cool ant pump flywheels is discussed in Appendix 3A under the response to Regul atory Guide 1.14. 5.4.1.5.4Inservice Inspection
The reactor coolant pump fl ywheels will be inspected inse rvice in accordance with the recommendations given in Regulatory Guide 1.14, "Reactor Coolant Pump Flywheel Integrity," Revision 1, August 1975. A description of th e inspections is included in Technical Specification 5.5.7.5.4.2STEAM GENERATORS5.4.2.1Design BasesSteam generator design data are given in Table 5.4-3. Code classifications of the steam generator components are given in Section 3.2. Although the ASME classification for the secondary side is specified to be Class 2, all pressure-retaining parts of the steam generator, and thus both the primary and secondary pressu re boundaries, are designed to satisfy the criteria specif ied in Section III of the ASME Code for Class 1 components. The design stress limits, tr ansient conditions, and combined loading conditions applicable to the steam gener ator are discussed in Section 3.9(N).1. Estimates of radioactivity levels anticipated in the se condary side of the steam generators during normal operation and t he bases for the esti mates are given in Chapter 11.0. The accident analysis of a steam generator tube r upture is discussed in Chapter 15.0. The internal moisture separation equipmen t is designed to ens ure that moisture carryover does not exceed 0.
1 percent by weight under the following conditions: a.Steady state operation up to 100 percent of full load steam flow, with water at the normal oper ating level.
CALLAWAY - SP5.4-11Rev. OL-21c 1/16b.Loading or unloading at a ra te of 5 percent of fu ll power steam flow per minute in the range from 15 to 100 percent of full load steam flow. c.A step load change of 10 per cent of full power in the range from 15 to 100 percent full load steam flow.
The water chemistry on the reactor side is selected to provid e the necessary boron content for reactivity contro l and to minimize corrosion of RCS surfaces. The water chemistry of the steam side and its effectiveness in corros ion control are discussed in Chapter 10.0. Compatibility of steam generator tubing with both primary and secondary coolants is discussed further in Section 5.4.2.3.2. The steam generator is designed to prevent unacceptable damage from mechanical or flow-induced vibration. Tube support adequacy is discussed in Section 5.4.2.5.3. The tubes and tube sheet are analyzed and confirmed to withstand the maximum accident loading conditions as they are defined in Section 3.9(N).1. Further consideration is given in Section 5.4.2.5.4 to the effect of tube wall thi nning on accident condi tion stresses.
Access is provided to the prim ary side channel heads of the steam generat or in order to permit inservice inspection a nd tube plugging/sle eving, when required. Access is provided to the shell side of the steam generator in the region of the tube sheet and flow distribution baffle in order to permit inservice inspecti on and removal of accumulated sludge. 5.4.2.2Design Description The steam generator is an Areva Model 73/19T, vertical shell and U-tube evaporator, with integral moisture separating equipment.
Figure 5.4-3 illustrates the design, indicating several of its design features whic h are described in the following paragraphs.On the primary side, the reactor coolant flows through the inverted U-tubes, entering and leaving through nozzles lo cated in the hemispherical bottom head of the steam generator. The head is divided in to inlet and outlet chambers by a vertical divider plate extending from the apex of the head to the tube sheet.Steam is generated on the shell side, flows upward, and exit s through the outlet nozzle at the top of the vessel. Feedw ater enters the steam generator at an elevation above the top of the U-tubes, throu gh the feedwater nozzle.
The water is distributed circumferentially around the steam generator by means of a feedwater ring and then flows downward through an annul us between the tube wr apper and shell. The distribution of feedwater is of fset, with a greater volume of feedwate r supplied to the other side of the tube bundle. The feedwater enters the ring via an anti-stratification helix inlet and it leaves through inve rted "J-nozzle" tubes located t hat flow holes, which are at the top of the ring.
These features are desig ned to prevent a condit ion which can result in water hammer occurrences in the feedwater piping. The steam water mixture from the tube bundle rises into the stea m drum section, where 16 indi vidual centrifugal moisture CALLAWAY - SP5.4-12Rev. OL-21c 1/16separators remove most of the entrained water from the steam. The steam continues to the secondary separators, whic h remove most of the rema ining moisture and provide a quality of at least 99.9 percent. The separated water is combined with entering feedwater to flow back down the annulus between the wrapper and the shell for recirculation through the steam generator. The dry steam exists from the steam generator through the outlet no zzle which is provided with a steam flow restriction.
A major design feature is the use of improved material for steam g enerator tubing. The tubing material is Inconel 690 Thermally Treated (TT) tubing. Improved fabrication processes were also used to increase tube reliability. The tube heat transfer surface area is approximately 80,000 ft 2.The holes in the tube support plates of the Model 73/19T generator have three-lobe shape that provides three lands to support the tube laterally. The holes are fabricated by drilling, followed by broaching.
The tubes are seal welded to the tube sheet cladding. Fu sion welds are performed in compliance with Section III and IX of the ASME Code and are dye penetrant inspected and leakproof tested. After we lding, each tube is hydraulically expanded for the full depth of the tube sheet to the secondary surface to eliminat e crevices between the tube and tube sheet.
The steam generators are fabric ated from all forged components. The reactor coolant portion of the steam generator consists of a low alloy steel forged channel head with two integral nozzles connecting it to the reactor coolant system. The primary channel head is also self-draining which el iminates a channel head drai n line and electro-polished to enhance decontamination efforts during inspectons. The secondary portion of the steam generator consists of a boili ng region and a stea m drum, also fabr icated from full forgings. The steam drum consists of two cyclindrical shells, includi ng the feedwater nozzle and two 16" diam eter Manway openings.
The Model 73/19T steam generator has a mass circulation ratio of 4.0. The circulation ratio is the total mass flow through the tube bundle divided by the steam outlet mass flow. The benefits of a high circulation ratio include 1) a reduced precipitate deposit as a result of reduced quality and increased liquid velocity, 2) increas ed water level control during transients due to less void and lower volume displace ment, and 3) an improved blowdown efficiency as the reci rculated downcomer fluid is less diluted by relatively clean feedwater.
The Model 73/19T also has an integrated fo reign object capture system and a grooved-ring nozzle dam retention system.
The foreign object capture system acts as a filter for foreign objects in feedwater before the feedwater reaches the tubes. This system is used to preclude tube damage.
The grooved-ring retenti on system is located in the primary channel head at the top of the entrance and ex it nozzles. This ring provides for quick installation of nozzle dams to facilitate tube inspections.
CALLAWAY - SP5.4-13Rev. OL-21c 1/165.4.2.3Steam Generator Materials5.4.2.3.1Selection and Fabr ication of MaterialsAll pressure boundary materials used in the steam generator are selected and fabricated in accordance with the requirements of Section III of t he ASME Code. A general discussion of materials spec ifications is given in Section 5.2.3 , with types of materials listed in Tables 5.2-2 and 5.2-3. Fabrication of reacto r coolant pre ssure boundary materials is also discussed in Section 5.2.3, particularly in Section 5.2.3.3 and 5.2.3.4. The steam generator materials are carbon steel, except for the tubes, tube sleeves, tube support plates, flow distribution baffle, antivibration bars , and the channel head divider plate. The interior surfac es of the reactor coolant channel head, nozzles, and manways are clad with austenitic stainle ss steel. The primary side of the tube sheet is weld clad with Inconel (ASME SFA-5.14). The tubes are Inconel-690, a nickel-chromium-iron alloy (ASME SB-163). The ch annel head divider plate is Inconel (SB-168). Tube support plates are ferritic stainless steel (Type 410).
Code cases used in material selection are discussed in Section 5.2.1. The extent of conformance with Regulatory Guides 1.84 and 1.85 is discussed in Appendix 3A. During manufacture, cleaning is performed on the primary and secondary sides for the steam generator, in accordance with written procedures which foll ow the guidance of Regulatory Guide 1.37 and the ANSI Standard N45.2.1-1973, "Cleaning of Fluid Systems and Associated Components for Nuclear Power Plants." Onsite cleaning and cleanliness control also follow the guidance of Regulatory Guide 1.37, as discussed in Appendix 3A. Cleaning process specif ications are discussed in Section 5.2.3.4. The fracture toughness of the materials is discussed in Section 5.2.3.3. Adequate fracture toughness of ferritic materials in the reactor coolant pressure boundary is provided by compliance with Appendix G of 10 CFR 50 a nd with Paragraph NB-2300 of Section III of the ASME Code. As discussed in Section 5.4.2.1, consideration of fracture toughness is only necessary for materials in Class 1 components. 5.4.2.3.2Compatibility of Steam Generator Tubing with Primary and SecondaryCoolantsAs mentioned in Section 5.4.2.3.1, corrosion tests, which subjected the steam generator tubing material, Inconel-690 (ASME SB-163), to simula ted steam generator water chemistry, have indica ted that the loss due to general corrosion over the 40-year plant life is insignificant, compared to the tube wall thickness. Testing to investigate the susceptibility of heat exchanger construction materials to stress corrosion in caustic and chloride aqueous solutions has indicated that Inconel-690 has excellent resistance to general and pitting type corrosion in severe operating water conditions. Many reactor years of successful operati on have shown the same low g eneral corrosion rates as indicated by the laboratory tests.
CALLAWAY - SP5.4-14Rev. OL-21c 1/16 Adoption of the all volatile treatment (AVT) chemistry control pr ogram eliminates the possibility for recurrence of the tube wall thinning phenomenon related to phosphate chemistry control. Successful AVT operation requires maintenance of low concentration of impurities in the steam generator water, thus reducing the potential for formation of hi ghly concentrated solutions in low flow zones, which is the precursor of corrosion. By restriction of the total alkalinity in the steam generator and prohibition of extended operation with free alkalinity, the AVT control program mini mizes the possibility for o ccurrence of intergranular corrosion in localized ar eas due to excessive leve ls of free caustic. Additional extensive operating data are presently being accumulated with the conversion to AVT chemistry. A comp rehensive program of steam generator inspections, in accordance with the Technical Specification, will ensure detection and correction of any unanticipated degradation that might occur in the st eam generator tubing. Another corrosion-related phenomenon, termed tube denting, was first discovered during the April 1975 steam generator insp ection at the Surry Unit No. 2 plant. This discovery was evidenced by eddy curr ent signals resembling those produced by scanning dents and by difficulty in passing the standard eddy current probe thr ough the tubes at the intersections with the support plates. Subsequent to the initial finding, steam generator inspections at other operating plants revealed indications of denting to various degrees. An intensive program of investigations, which has included removal of dented tubes and tube/support plate samples from affected steam generators and laboratory tests of heated crevices and m odel boilers, has reve aled that the source of tube denting is corrosion of the car bon steel tube suppo rt plate (TSP) in the crevices between the tube and TSP. The corrosion rate in these locations is apparently accelerated by deposition of impurities from the se condary fluid, caused by low flow velocity and superheated fluid in the crevice. The corrosion product has a larger volume then the base metal. The results are simultaneous reduct ion of the tube diameter, dilation of the hole in the TSP, and secondary effects (e.g., TSP distortions) relat ed to dilation of the TSP holes. Denting has been most pronounced in plants having a history of chloride contamination resulting from condenser leakage. The presence of acid chloride has been found to be a common factor in tube dent ing produced in laboratory tests. Measures to inhibit denting concentrate on providin g a more corrosion resistant TSP material and on eliminating conditions conducive to corrosion at the tube support locations (e.g., chemical impurities in the secondary fluid and localized superheat). The tube support plates used in the Model 73/19T steam generator are Type 410 ferritic stainless steel which has been shown in laboratory tests to be resistant to corrosion in the AVT environment.
When corrosion of ferritic stainless steel does occur, the volume of the corrosion products is equivalent to the volume of the parent material. Thus, substitution of Type 410 ferritic stainless stee l for carbon steel used in previous steam generators substantially reduces the potential for tube denting.
CALLAWAY - SP5.4-15Rev. OL-21c 1/16 Other features of the Mode l 73/19T generator further r educe the potential for tube denting. The trifoil geometry of the tube support plates is less susceptible to the accumulation of corrosion products which caus e tube denting. The trifoil geometry also results in a reduced flui d pressure drop across t he tube support plates and, therefore, a higher recirculation rati o and higher fluid velocities in the tube bundle.
Operating experience, verified in numerous steam generator inspections, indicates that the tube degradation associ ated with phosphate water treatment is not occurring where only AVT has been utilized. Adherence to the AVT chemical specifications and close monitoring of the condenser integrity will assure the co ntinued good performance of the steam generator tubing. 5.4.2.3.3Control of Se condary-Side ImpuritiesSeveral provisions exist in the SNUPPS plants to limit the accumulations of impurities in the steam generator, either by limiting ingress or by facilitating removal. The materials of construction of the secondary system are such as to minimize the formation of corrosion products. The materials include stainless steel tubing in all feedwater heaters and Corten tubing in the moisture-separator-reheaters. A full-flow condensate demineralizer system is provided. A piping connection is provided from the feedwater heater, ahead of the steam generators, to the condenser hot well. During startup, this connection is used to circulate secondary system water through the condensate de mineralizers. The flow circulation removes suspended corrosion products that may have accumulated during extended shutdowns.
For removal of impurities, the blowdown system has a capacity slightly in excess of 1 percent of full-load feedwater flow. As described in Section 5.4.2.2 and 5.4.2.3.2 , the design of the Model 73/19T st eam generator is expected to result in an increased efficiency of impurity remova l by the blowdown system.
The feedwater system mate rials are discussed in Section 10.4.7 , the steam generator blowdown system is discussed in Section 10.4.8 , and the condensat e demineralizer system is discussed in Section 10.4.6. Instrumentation to monitor secondary side water chemistry is described in Section 9.3.2. During shutdowns, sludge lancin g may be used to remove accumulated material. In sludge lancing, a hydraulic jet is inserted through an access opening (handhole) to loosen sludge deposits, which are remov ed by means of a suction pump. 5.4.2.4Steam Generator Inservice InspectionThe steam generator and associated insulation is designed to permit inspection of Class 1 and 2 parts, including indivi dual tubes. The design includ es a number of openings to provide access to both the primary and secondary sides of the steam generator, and the inspection program followed complies with Sectio n XI of the ASME Code, including addenda per 10 CFR 50.55a (g) with certain except ions whenever specif ic written relief CALLAWAY - SP5.4-16Rev. OL-21c 1/16 is granted by the NRC per 10 CFR 50.55a (g) (6).
These openings include four manways, two for access to both chambers of the reactor coolant channel head inlet and outlet sides and two in the steam drum for inspection and maintenance of the moisture separators, six 6-inch handhol es above the top of the tube sheet, and 12 2-inch inspection ports located between each tube supp ort plate elevation. Access to the tube U-bend is provided through the deck plates and wrapper canopy. For proper functioning of the steam generator, some of the deck plate openings ar e covered with welded, but removable, hatch plates. Inspection/access to the primary side is provided by two 16-inch manways located in the channel head. The insulation in the area of circumferential welds, includ ing tube-sheet-to-head or shell welds, primary nozzle-to-vessel head welds and nozzle-to-head inside radiused sections; primary nozzle-to-safe end welds; integrally welded vessel supports, circumferential butt welds, and nozzle-to-vessel welds on the secondary side is removable. The pressure-ret aining bolting can be removed for examination. Manways in the primary hea d allow direct visual examination of the head cladding.
The manways allow sufficient access for the installati on of the remotely operated eddy current equipment capable of perform ing inservice inspections. 5.4.2.4.1Compliance with Sect ion XI of the ASME Code Eddy current examinations of steam gene rator tubing/sleeves are performed in accordance with Appendix IV to Section XI of the ASME Boiler and Pressure Vessel Code. Other Class 1 and Class 2 components of the steam generators are examined in accordance with the requirements of the ASME Boiler and Pressure Vessel Code,Section XI. The inservice inspection pr ogram of Class 1 components of the steam generators is described in Section 5.2.4. The inservice inspection of Class 2 components of the steam genera tors is discussed in Section 6.6. 5.4.2.4.2Program for Inservice Inspection of Steam Generator TubingSteam generator tubing is inspected in accordance with Technical Specifications. This guide covers the in spection equipment, basel ine inspections, tube selection, sampling and frequency of inspection, methods of recording, an d required actions based on findings. The design of the steam generators permits inservice inspection, plugging and/
or sleeving, if required, of each tube. Regulatory Guide 1.121 and the Technical Specifications provide recommendations concerning tube plugging. The remotely operated equipment is capable of examining the entire length of the tubes. All original examination data, results, and reports are stored in a fireproof facility and in an atmosphere controlled to minimize deterioration. The data is stored in a limited-access facility and retained for the operati ng life of the plant.
CALLAWAY - SP5.4-17Rev. OL-21c 1/16Standards consisting of sim ilar as-manufactured steam generator tubing with known imperfections are used to establish sensitivity and to ca librate the equipment. Where practical, these sta ndards include reference flaws that simulate the length, depth, and shape of actual imper fections that are characteristic of past experience. Personnel engaged in taking or interpreting data are test ed and qualified in accordance with ANSI/ASNT CP-189 and supplements des ignated by the Ed ition and Addenda of Section XI used during the examination. Procedures governing the above examinations are qualified prior to ex amination in the plant.
All of the tubes or tube sleev es in the steam gen erators shall be inspected by eddy current prior to service to establish a baseline co ndition of the tubing. The sample selection and testing of tubes, the inspection intervals, and the actions to be taken if defects are identified will follow the Steam Generator Program in the Technical Specifications.5.4.2.5Design Evaluation Seismic and LOCA loads are discussed in Section 3.9(N)
.5.4.2.5.1Forced Convec tion of Reactor Coolant The limiting case for heat transfer capability is the "nominal 100-percent design" case. The steam generator effective heat transfer coefficient is based on the coolant conditions of temperature and flow for th is case. The best estimate for the heat transfer coefficient applied in steam generator design calculations and plant pa rameters selection is 1406 Btu/hr-ft 2-°F. The coefficient incorporates a specified fouling factor resistance of 0.00005 hr-ft 2-°F/Btu, which is the value selected to account for the differences in the measured and calculated heat transfer performance as well as provide the margin indicated above.
Although margin for tube f ouling is available, operati ng experience to date has not indicated that steam generat or performance decreases ov er a long-tim e period.
Adequate tube area is selected to ensure that the full design heat removal rate is achieved. 5.4.2.5.2Natural Circula tion of Reactor Coolant The driving head created by the change in coolant density as it is heated in the core and rises to the outlet nozzle initiates convection circulation. This circulation is enhanced by the fact that the steam generators, which prov ide a heat sink, are at a higher elevation than the reactor core, wh ich is the heat source. Natural circulation is sufficient for the removal of decay heat during hot shutdown and cooldown in the event of a loss of forced circulation.
CALLAWAY - SP5.4-18Rev. OL-21c 1/165.4.2.5.3Mechanical and Flow-Induced Vibration Under Normal Operation The possibility of vibratory failure of tubes due to either mechanical or flow-induced excitation has been thoroughly evaluated. This evaluation includes detailed analysis of the tube support systems as well as an extensive research program with tube vibration model tests.
In evaluating possible fa ilure due to vibration, considerat ion is given to such sources of excitation as those generated by the primary fluid flowing within the tubes. The effects of these as well as any other mechanicall y induced vibrations are considered to be negligible and should caus e little concern.
Another source of possibl e vibratory failure in heat exchanger components is hydrodynamic excitation by the secondary fluid on the outside of the tubes.
Consideration of secondary flow-induced vibration involves two types of flow, parallel and cross, and it is eval uated in three regions:a.At the entrance of t he downcomer feed to the tube bundle (cross flow)b.Along the straight sections of the tube (p arallel flow)c.In the curved tubed section of the U-bend (cross flow)For the case of parallel flow , analysis is done to determine the vibratory deflections in order to verify that the flow velocities are sufficiently bel ow those required for damaging fatigue or impacting vibratory amplitude. Thus, the support system is deemed adequate to preclude parallel flow excitation.
For the case of cross-flow excitation, several possible mechanisms of tube vibration exist. For the Model 73/19T steam generato r design and conditions, only two of these mechanisms are deemed significa nt enough to merit extensiv e consideration: 1) turbulence and 2) fluidelastic vibration. The st eam generator is analyz ed to ensure that the turbulence flow velocity is acceptable and that unst able fluidelastic vibration does not exist. In order to achieve this, adequate tube supports must be provided. An evaluation using the specific parameters for the M odel 73/19T steam generator confirms the integrity of the support system.
Assurance against damaging flow induced tube vibration ha s been accomplished by a combination of analysis and testing. Cross and parallel fl ow velocities were calculated from thermal-hydraulic analysis of the secondary flow.
Three possible vibrational mechanisms, vortex shedding, fluidelastic excitation, and turbul ence were studied.Tests carried out on the tube bundle geometry showed no vortex s hedding for small pitch ratios.
CALLAWAY - SP5.4-19Rev. OL-21c 1/16For fluidelastic excitation, the ratios of the effective cross fl ow velocity to the critical velocity were calculated. The results indicate that no critical excitation will occur during steady state conditions as well as operational transients. Turbulence responses are low a nd therefore risk of fatigue due to turbulent flow is excluded. 5.4.2.5.4Allowable Tube Wall Th inning Under Acci dent Conditions An evaluation is performed to determine the ext ent of tube wall th inning that can be tolerated under accident condi tions. The worst-case loadi ng conditions are assumed to be imposed upon uniformly thinned tubes, at the most critical loca tion in the steam generator. Under such a postul ated design basis accident, vibration is of short enough duration that there is no endur ance problem to be consider ed. The steam generator tubes, existing original ly at their minimum wall thickness and reduc ed by a conservative general corrosion and erosion loss, can be shown to provide an adequate safety margin, that is, sufficient wall thickness, in ad dition to the minimum required for a maximum stress less than the allowable stress limit, as it is defined by the ASME Code. 5.4.2.6Quality Assurance The steam generator nondest ructive examination pr ogram is given in Table 5.4-4. Radiographic inspection and acceptance stan dards are in acco rdance with the requirements of Section III of the ASME Code. Liquid penetrant inspection is performed on weld deposited tube sheet cladding, channel head cladding, divider plate to tube sheet and to channel head weldments, tube-to-tube sheet weldments, and weld deposit cladding.
Liquid penetrant inspection and acceptance standards are in accordance with the requirements of Section III of the ASME Code. Magnetic particle inspection is performed on the tube shee t forging, channel head casting, nozzle forgings, and the following weldments: a.Nozzle to shell b.Support bracketsc.Instrument connection (secondary)d.Temporary attachments after removal e.All accessible pressure retaining welds after hydrostatic test CALLAWAY - SP5.4-20Rev. OL-21c 1/16Magnetic particle inspection and acceptance standards are in accordance with the requirements of Section III of the ASME Code. Ultrasonic tests are performed on the tube sheet forging, tube sheet cladding, secondary shell and head plate, and nozzle forgings.
The heat transfer tubing is subject ed to eddy current testing. Hydrostatic tests are performed in accordanc e with Section III of the ASME Code. 5.4.3REACTOR COOLANT PIPING5.4.3.1Design Bases The RCS piping is designed and fabricated to accommodat e the system pressures and temperatures attained under all expected modes of plant operation or anticipated system interactions. Stresses are mainta ined within the limits of Section III of the ASME Code.
Code and material requirements are provided in Section 5.2
.Materials of construction ar e specified to minimize co rrosion/ erosion and ensure compatibility with the operati ng environment. The piping in the RCS is Safety Class 1 and is designed and f abricated in accordance with ASM E Code,Section III, Class 1 requirements. Stainless steel pipe conforms to ANSI B36.19 for sizes 1/2 inch through 12 inches and wall thickness Schedules 40S through 80S. Stainless steel pipe outside of the scope of ANSI B36.19 conforms to ANSI B36.10.
The minimum wall thicknesses of the l oop pipe and fittings ar e no less than those calculated using the ASME Code,Section III, Class 1 formula of Paragraph NB-3641.1(3) with an allowable stress value of 17,550 psi.
The pipe wall thickness for the pressurizer surge line is Schedule 160. The minimum pipe bend radius is 5 nominal pipe diameters, and ovality does not exceed 6 percent.
All butt welds, bran ch connection nozzle welds, and boss welds are of a full penetration design. Full structural weld overlays (FSWOLs) have been installed on the dissimilar metal welds and adjacent stainless steel weld s of the pressurizer surge, spray, safety and relief nozzles.Processing and minimization of sensitization are discussed in Section 5.2.3 Flanges conform to ANSI B16.5.
Socket weld fittings and socket joints conform to ANSI B16.11.
CALLAWAY - SP5.4-21Rev. OL-21c 1/16 Inservice inspection is discussed in Section 5.2.4. 5.4.3.2Design Description The RCS piping includes thos e sections of piping interc onnecting the r eactor vessel, steam generator, and reactor coolant pump. It also includes the following: a.Charging line and alternate charging line from the system is olation valve up to the branch conn ections on the reactor coolant loopb.Letdown line and excess letdown line from the branch conn ections on the reactor coolant loop to t he system isolation valvec.Pressurizer spray lines from the reac tor coolant cold l egs to the spray nozzle on the pre ssurizer vesseld.Residual heat removal lines to or from the reactor coolant loops up to the designated check valve or isolation valvee.Safety injection lines from the designated check valv e to the reactor coolant loopsf.Accumulator lines from the designate d check valve to the reactor coolant loopsg.Loop fill, loop drain, sample*, and instrument* lines to or from the designated isolation valve to or from the reactor coolant loopsh.Pressurizer surge line from one r eactor coolant loop hot leg to the pressurizer vessel inlet nozzlei.Resistance temperature detector scoop element, pressurizer spray scoop, sample connection* with scoop, reac tor coolant tem perature element installation boss, and the temperature element thermowell itself j.All branch connection nozzles attached to reactor coolant loopsk.Pressure relief lines from nozzles on top of the pressurizer vessel up to and through the power operated pressurizer relief valves and pressurizer safety valves*Lines with a 3/8-inch (liquid service) or less flow restricting orifice qualify as Safety Class 2. In the event of a break in one of these Safety Class 2 lines, the normal makeup system is capable of providing makeup flow while maintaining pressuri zer water level.
CALLAWAY - SP5.4-22Rev. OL-21c 1/16l.Seal injection water lines to the reactor coolant pump to the designated check valve (injection line)m.Auxiliary spray line from the isolation valve to t he pressurizer spray line headern.Sample lines* from pressuri zer to the isolation valveo.Reactor vessel head vent li nes* to the isolation valvesPrincipal design data for the reac tor coolant piping are given in Table 5.4-5
.Details of the materi als of construction and codes used in the fabrication of reactor coolant piping and fittings are discussed in Section 5.2
.The reactor coolant pi ping and fittings which make up the loops are austenitic stainless steel. Pipe and fittings are cast, seamless without longitudinal or electroslag welds, and comply with the requirements of the ASME Code,Section II (Parts A and C),Section III, and Section IX. All smaller piping which is part of the RC S, such as the pressurizer surge line, spray and relief line, loop drains and connecting lines to other systems, are also austenitic stainless steel.
The nitrogen supply line for the pressurizer relief tank is carbon steel. All joints and connections are welded, except for the pressurizer code safety valves, where flanged joints are used. A thermal sleeve is installed on the pressurizer spray line nozzle.
All piping connections from auxiliary system are above the horizontal cent erline of the reactor coolant piping, with the exception of:a.Residual heat removal pump suction lines, which are 45 degrees down from the horizontal centerli ne. This enables the water level in the RCS to be lowered in the reactor coolant pipe while continuing to operate the residual heat removal syst em, should this be requi red for maintenance. b.Loop drain lines.c.The differential pressure taps for flow measurement, which are downstream of the stea m generators of the first 90-degree elbow. d.The pressurizer surge line, which is attached at the horizontal centerline. e.Two of the three scoops in each resistance temperatur e detector hot leg connection. f.The hot leg sample co nnections, the loop 3 ther mowell, and the loop 4 boron injection con nection, all located on the horizontal centerline.
CALLAWAY - SP5.4-23Rev. OL-21c 1/16g.The connections for measurement of water level in the RCS during refueling and maintenance operation.
Penetrations into the coolant flow path are limited to the following: a.The spray line inlet connections extend in to the cold leg piping in the form of a scoop so that the velocity head of the reactor coolant loop flow adds to the spray driving force. b.The reactor coolant sample system taps protrude in to the main stream to obtain a representative sample of the reactor coolant. c.The hot leg connections to the resistance temp erature detectors have scoops which extend into the reactor coolant to collect a representative temperature sample for the individual hot leg resistance temperature detectors. d.The wide range temper ature detectors are located in resistance temperature detector wells that extend into both the hot an d cold legs of the reactor coolant pipes.One hot leg and one cold leg temperature reading are provided from each coolant loop to use for protection. Narro w range, thermowell-mounted Resistance Temperature Detectors (RTDs) are provided for each coolant loop. In the hot legs, sampling scoops are used because the flow is stratified. That is, the fluid temperature is not uniform over a cross section of the hot leg. One dual element RTD is mounted in a thermowell in each of the three sampling scoops a ssociated with each hot leg. The scoops extend into the flow stream at locations 120° apart in the cross sectional plane. Each scoop has five orifices which sample the hot leg flow along the leading edge of the scoop. Outlet ports are provided in the scoops to direct the sampled fluid past the sensing element of the RTDs. One of each of the RTD's dual elements is used while the other is an installed spare. Three readings from each hot leg are aver aged to provide a hot leg reading for that loop.One dual element RTD is mounte d in a thermowell associated with each cold leg. No flow sampling is needed becaus e coolant flow is well mi xed by the reactor coolant pumps. One RTD element is used while the other is an installed spare.The thermowells are pressure boundary parts wh ich completely enclose the RTD. They have been shop hydrotested to 1.25 times the RCS design pr essure. The external design pressure and temperat ure are the RCS design tem perature and pressure. The RTD is not part of the pressure boundary. The scoop, ther mowell, and thermowell/scoop assembly have been analyzed to the ASME Boiler and Pressure Vessel Code,Section III, Class 1. The effects of seismic and flow-induced l oads were considered in the design.
CALLAWAY - SP5.4-24Rev. OL-21c 1/16 Signals from the temperatur e detectors are used to co mpute the reac tor coolant T (temperature of the hot leg, T hot , minus the temperature of the cold leg, T cold) and an average reactor coolant temperature (T avg). The T avg for each loop is indicated on the main control board. 5.4.3.3Design Evaluation Piping load and stress evaluat ion for normal operating loads, seismic loads, blowdown loads, and combined normal, blowdown, and seismic loads is discussed in Section 3.9(N).5.4.3.3.1Material Corros ion/Erosion EvaluationThe water chemistry is selected to minimize corrosion. A periodic analysis of the coolant chemical composition is perfo rmed to verify that the reactor coolant quality meets the specifications (see Section 5.2.3
). Periodic analysis of the c oolant chemical composition is performed to monitor the adherence of the system to desired reac tor coolant water quality listed in Table 5.2-5. Maintenance of the water qualit y to minimize corrosion is accomplished, using the chemical and volume control system and sampling system which are described in Chapter 9.0
.The design and construction ar e in compliance with the ASME Code,Section XI. Pursuant to this, all pressure containing welds out to the second valve that delineates the RCS boundary are accessible for examination and are fitted with removable insulation. 5.4.3.3.2Sensitized Stainless SteelSensitized stainless steel is discussed in Section 5.2.3
.5.4.3.3.3Contaminant Control Contamination of stainless steel and Inconel by copper, low melting temperature alloys, mercury, and lead is prohibited. Thread lubricants are app roved in accordance with applicable procedures. Pr ior to application of thermal insu lation, the austenitic stainless steel surfaces are cleaned and analyzed to halogen limits as defined by Westinghouse Process Specifications.5.4.3.4Tests and Inspections The RCS piping quality assu rance program is given in Table 5.4-6. Volumetric examination is per formed throughout 100 percent of the wall volume of each pipe and fitting in accordance with the applicable requirements of Section III of the ASME CALLAWAY - SP5.4-25Rev. OL-21c 1/16 Code for all pipe 27-1/2 inches and larger. All unacceptable defect s are eliminated in accordance with the requirements of the same section of the code.
A liquid penetrant examinatio n is performed on both the entire outside and inside surfaces of each finished fitti ng, in accordance with the criteria of the ASME Code,Section III. Acceptance standards are in accordance with the applicable requirements of the ASME Code,Section III.
The pressurizer surge line conforms to SA-376, Grade 304, 30 4N, or 316 with supplementary requirements S2 (transverse tension tests) and S6 (ultrasonic test). The S2 requirement applies to each length of pipe. The S6 require ment applies to 100 percent of the piping wall volume.
The end of pipe sectio ns, branch ends, and fi ttings are machined back to provide a smooth weld transition adjacent to the weld path. 5.4.4MAIN STEAM LINE FLOW RESTRICTOR5.4.4.1Design Basis The outlet nozzle of th e steam generator is provided with a flow restrictor designed to limit steam flow in the unlikely event of a break in the main steam line.
A large increase in steam flow will create a backpressure which limits further increase in flow. The flow restrictor performs the follow ing functions: Rapid rise in containment pressure is prevented, the rate of heat re moval from the reactor coolant is such as to keep the cooldown rate within acceptable limits, thru st forces on the main steam line piping are reduced, and stresses on internal steam generator components, particularly the tube sheet and tubes, are limi ted. The restrictor is confi gured to minimize the unrecovered pressure loss across the restri ctor during normal operation.5.4.4.2Design DescriptionThe flow restrictor consists of seven Inco nel (ASME SB-166) venturi inserts which are installed in holes in an integral low alloy steel forging. The inserts are arranged with one venturi at the centerline of the outlet nozzle and the other six equally spaced around it. After insertion into the low alloy steel forg ing holes, the Inconel venturi nozzles are welded to the Inconel cl adding on the inner surfac e of the forging. 5.4.4.3Design Evaluation The flow restriction design has been analyzed to assure its structural adequacy. The equivalent throat diameter of the steam generator outlet is 16 inches, and the resultant pressure drop through the restrictor at 100-percent steam flow is approximately 3.8 psig.
This is based on a design flow rate of 3.99 x 10 6 lb/hr (see Tables 5.1-1 and 10.3-2 for uprated steam flow rates).
Materials of construction and manufacturing of the flow restrictor are in accordance with Section III of t he ASME Code.
CALLAWAY - SP5.4-26Rev. OL-21c 1/165.4.4.4Tests and Inspections Since the restrictor is not a part of the steam system boundary, no tests and inspection beyond those during fabrication are anticipated. 5.4.5MAIN STEAM LINE ISOLATION SYSTEM
The main steam line isolat ion system is discussed in Section 10.3. 5.4.6REACTOR CORE ISOLATION COOLING SYSTEMThis section is not applicable to SNUPPS.
5.4.7RESIDUAL HEAT REMOVAL SYSTEM5.4.7.1Design BasesThe residual heat removal system (RHRS) functions to remove heat from the RCS when RCS pressure and temperature are below approximately 400 psig and 350°F, respectively. Heat is transferred from the RHRS to the component cooling water system.
Portions of the RHRS also serve as portions of the ECCS during the injection and recirculation phases of a LOCA (see Section 6.3
). The RHRS also is used to transfer refueli ng water between the ref ueling cavity and the refueling water storage tank at the beginning and end of the refueling operations. The RHRS is designed to be isolated from the RCS whenever the RCS pressure exceeds the RHRS design pressure. 5.4.7.2Design Description5.4.7.2.1Functional DesignRHRS design parameters are listed in Table 5.4-7. Nuclear plants employing the same RHRS design as the SNUPPS units are given in Section 1.3. During normal approaches to cold shutdow n, the RHRS is placed in operation approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after reactor shutdown when the temperature and pressure of the RCS are below approximately 350°F and 400 psig, respectively. Assuming that two heat exchangers and two pumps are in service and that each heat exchanger is supplied with component cooling water at des ign flow and temperature, the RHRS is designed to reduce the temperature of the reactor coolant from 350°F to 140°F within 14.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> (See Figure 5.4-9). The time required, under these conditions, to reduce reactor coolant temperature from 350°F to 212°F is 2.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> based on a 50
°F/hr cooldown rate. The heat load handled by t he RHRS during the cooldown trans ient includes residual and decay heat from the core and reactor coolant pump heat. The design heat load is based CALLAWAY - SP5.4-27Rev. OL-21c 1/16 on the decay heat fraction that exists at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> following reactor shutdown from an extended run at full power. The design heat load used here and in Section 9.2.5 (UHS) is based on the decay heat gener ation rate of 78.9 x 10 6 Btu/hr. Based on the MODE 4 ECCS standby alignment requirements of Technical Specification 3.5.3 imposed on one RHR train, as well as procedural requirements related to preventing void formation associated with exceeding saturation conditi ons in the RHR suct ion lines from the RWST, the assumption of a two-train cooldown starting at the RHR cut-in conditions does not reflect plant operating practice.
Assuming that only one heat exchanger and pump are in service and that the heat exchanger is supplied with com ponent cooling water at design flow and temperature, the RHRS is capable of reducing th e temperature of the reacto r coolant from 350°F at 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after shutdown to 200°F at 30.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after shutdown (See Figure 5.4-10). The time required under these condi tions to reduce reactor cool ant temperature from 350°F to 212°F is approxi mately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. The RHRS is isolated from the RCS on the suction side by two motor-operated valves in series on each suction line.
Each motor-operated valve is interlocked to prevent its opening if RCS pressure is greater than 360 psig. A control room alarm will actuate if an RHR suction isolation valve is not fully closed and RCS pr essure is greater than the design pressures for RHR system operation. The RHRS is isolated from the RCS on the discharge side by two check valv es in each return line. Al so provided on the discharge side is a normally open, motor-operated valve down stream of each RHRS heat exchanger. (These check valves and motor-operated valves are not considered part of the RHRS. They are shown as part of the ECCS, see Figure 6.3-1
.) Each inlet line to the RHRS is equipped with a pr essure relief valve designed to relieve the combined flow of all the charging pumps at the relief valv e set pressure. These relief valves also protect the syst em from inadvertent overpr essurization during plant cooldown or startup and provide LTOP for the RCS during low tem perature water solid operation. Each discharge line from the RHRS to the RC S is equipped with a pressure relief valve designed to reli eve the maximum possible ba ckleakage through the valves isolating the RHRS from the RCS.
The RHRS is provided for a si ngle nuclear power uni t, and is not s hared among nuclear power units.
The RHRS is designed to be full y operable from the control room for normal operation.
Manual operations required of the operator are:
opening the suction isolation valves, positioning the flow contro l valves downstream of t he RHRS heat exchangers, and starting the residual heat removal pumps. By nature of its redundant two-train design, the RHRS is desi gned to accept all major component single failures with the only effect being an extension in the re quired cooldown time. For tw o low probability electrical system single failures, i.e., failure in the suction isolation valve interlock circuitry or diesel generator failure in conjunction with loss of offsite power, operator action outside the control room is requi red to open the sucti on isolation valves.
Manual actions are CALLAWAY - SP5.4-28Rev. OL-21c 1/16discussed in further detail in Sections 5.4.7.2.7 and 5.4.7.2.8. Spurious operation of a single motor-operated valve can be accepted without lo ss of function, as a result of the redundant two-train design.
Missile protection, protection against dynamic effects asso ciated with the postulated rupture of piping, and seismic design are discussed in Sections 3.5 , 3.6 , and 3.7(B), and 3.7(N) respectively. 5.4.7.2.2Piping and Instrumentation Diagrams The RHRS, as shown in Figures 5.4-7 (piping and instrumentation diagram) and 5.4-8 (process flow diagram), consists of two residual heat e xchangers, two residual heat removal pumps, and the associ ated piping, valves, and instrumentation necessary for operational control. The inle t lines to the RHRS are connected to th e hot legs of two reactor coolant loops, while the return lines are connected to the cold leg of each of the reactor coolant loops. Thes e return lines are also the ECCS low head injection lines (see Figure 6.3-1
). EJ-HV-8716A and B and EJ-H V-8809A and B are maintained open during operating modes 1-3 in order that either RHR pump is ab le to inject to all four RCS cold legs.
The RHRS suction lines are is olated from the RCS by two motor-operated valves in series located inside the containment. Each discharge line is isolated from the RCS by two check valves in series located inside the contai nment and by a normally open motor-operated valve located outside the containment. (The check valves and the motor-operated valve on each discharge line are shown as part of the ECCS, see Figure 6.3-1). During RHRS operation, reactor coolant flows from the RCS to the residual heat removal pumps, through the tube side of the residual heat exch angers, and back to the RCS. The heat is transferred to the component cooling water circul ating through the shell side of the residual heat exchangers.
Coincident with operation of th e RHRS, a portion of the reactor co olant flow may be diverted from downstream of the residual heat e xchangers to the chem ical and volume control system (CVCS) low pressure letdown line for cleanup an d/or pressure control.
By regulating the diverted fl owrate and the charging flow , the RCS pressure may be controlled. Pressure regulation is necessary to maintain the pressure range dictated by the fracture prevention criter ia requirement of t he reactor vessel, by the number 1 seal differential pressure, and by net positive suction head requirements of the reactor coolant pumps.
The RCS cooldown rate is m anually controlled by regulati ng the reactor coolant flow through the tube side of the RHR heat exchangers. The flow control valve in the bypass line around each RHR heat exchanger automatically maintains a constant return flow to the RCS. Instrumentation is provided to monitor system pressure, temperature, and total flow.
CALLAWAY - SP5.4-29Rev. OL-21c 1/16 The RHRS is also used for filling the refueling cavity before refueling. After refueling operations, water is pumped back to the refueling water storage tank until the water level is brought down to the flange of the reactor vessel. The remainder of the water is removed via a drain connection at the bottom of the refueling canal.
When the RHRS is in operation, the water chemistry is the sa me as that of the reactor coolant. Provision is made for the nuclear samp ling system to extract samples from the flow of reactor cool ant downstream of the residual heat exchangers.
A local sampling point is also provided on eac h residual heat re moval train between the pump and heat exchanger.
The RHRS functions in conjunction with t he high head portion of th e ECCS to provide direct injection of bor ated water from the refueling water storage tank into the RCS cold legs during the injection phase following a LOCA. During normal operation, the RHRS is aligned to inject bora ted water upon receipt of a safety injection signal. EJ-HV-8716A and B and EJ-HV-8809A and B are maintained open during operating modes 1-3 in order that either RHR pump is able to inject to all four RCS cold legs.In its capacity as the low head portion of the ECCS, the RHRS also provides long-term recirculation capability for co re cooling following the inject ion phase of a LOCA. This function is accomplished by aligning the RHRS to take fl uid from the containment sump, cool it by circulation thr ough the residual heat exchangers, and supply it to the core directly as well as via the ECCS centrifugal charging pumps and safety injection pumps. The use of the RHRS as part of the ECCS is more completely described in Section 6.3. The RHR pumps, in order to perform their ECCS function, are interlocked to start automatically on receipt of a safety injection signal (see Section 6.3
). The RHR suction isolation valves in each inlet line from the RCS are separately interlocked to prevent both their being opened w hen RCS pressure is greater than 360 psig. A control room alarm will actuate if an RHR suction isolation valve is not fully closed and RCS pre ssure is greater than the des ign pressures for RHR system operation. These interlocks are described in more detail in Sections 5.4.7.2.5 and 7.6.2. The RHR suction isolation valves are also interlocked to prev ent their being opened unless the isolation valves in the following lines are closed: a.Recirculation lines from the residual heat exchanger outlets to the suctions of the safety injection pumps and ECCS centrifugal charging pumpsb.RHR pump suction lines from the refueling water storage tankc.RHR pump suction lines from the containment sump CALLAWAY - SP5.4-30Rev. OL-21c 1/16 The motor-operated valves in the RHR miniflow bypass lines are interlocked to open when the RHR pump di scharge flow is less than appro ximately 816 gpm at 300°F (783 gpm at 68°F) and close when the flow exceeds approximately 1650 gpm at 300°F (1582 gpm at 68°F). 5.4.7.2.3Equipment and Component DescriptionsThe materials used to fabricate RHRS components are in accordance with the applicable code requirements. All parts of the components in cont act with borated water are fabricated or clad with austenitic stainless steel or equivalent corrosion-resistant material. Component pa rameters are given in Table 5.4-8. Residual Heat Removal PumpsTwo pumps are installed in the RHRS. The pumps are sized to deliver r eactor coolant flow through the RHR heat exchangers to meet the plant cooldown requirements. The use of two separate RHR trains assures that cooling capacity is only partially lost should one pump become inoperative. The RHR pumps are prot ected from overheating and loss of discharge flow by miniflow bypass lines. A valve located in each miniflow line is regulated by a signal from the flow indicating switch located in each pump discharge header. The control valves open when the RHR pump discharge flow is less than approximately 816 gpm at 300°F (783 gpm at 68°F) and close when the flow exceeds approximately 1650 gpm at 300°F (1582 gpm at 68°F). A pressure sensor in each pump discharge header provides a signal for an indicator in the control room. A high pre ssure alarm is also actuated by the pressure sensor. The two pumps are vertical, centrifugal units with mechanical seals on the shafts. All pump surfaces in contact with reactor coolant are austenitic stainless steel or equivalent corrosion resistant material. The RHR pumps also function as the low head safety injection pumps in the ECCS (see Section 6.3 for further information and for the residual heat remova l pump performance curves).Residual Heat ExchangersTwo residual heat exchangers are installed in the system. The heat exchanger design is based on heat load and temperature differences betw een reactor coolant and component cooling water existing 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after reactor shutdown when the temperature difference between the two systems is small.
CALLAWAY - SP5.4-31Rev. OL-21c 1/16The installation of two heat exchangers in separate and independent re sidual heat removal trains assures that the heat removal capacity of the system is only partially lost if one train becomes inoperative.
The residual heat exc hangers are of the s hell and U-tube type.
Reactor coolant circulates through the tubes, wh ile component cooling water circulates through the shell.
The tubes are welded to the tube sheet to prevent leak age of reactor coolant.
The residual heat e xchangers also function as part of the ECCS (see Section 6.3
). Residual Heat Removal System ValvesValves that perform a modulating function are equipped with two sets of packings and an intermediate leakoff connection that discharges to the drain header.
Manual and motor-operated valves have backseats to facilitate repacking and to limit stem leakage when the valv es are open. Leakage conne ctions are provided where required by valve size and fluid conditions. EncapsulationThe RHR suction lines from the containment recirculation sumps are each provided with a single motor-operated gate valve outside the containment. This valve, including its operator, is encapsulated in a pressure vessel which is leaktight at containment design pressure. The piping from t he sump to the valve is also encapsulated in a concentric guard pipe which is leaktight.
A leaktight seal is prov ided so that neither the compartment nor the guard pipe is connected directly to the sump or containment atmosphere. Component parameters for the encapsulation tank are given in Table 5.4-8
.The valve provides a barrier outside the containment to prevent loss of sump water should a leak develop in the recirculation loop.
Should a leak develop in the valve body or in the pipe between the valve and the sump, the sump fluid is contained by the leaktight seal and/or by the guard pipe.Each encapsulated gate valve is installed with a pathway from t he valve bonnet to the RHR system piping. This path way ensures that the intern al valve bonnet pressure will never be greater than the RHR system pressure, and t hus preclude the formation of pressure locking conditi ons for these valves.With this system, no single failure of either an active or a passive component will prevent the recirculation phase or adversely affect the integrity of the containment.5.4.7.2.4System OperationReactor Startup CALLAWAY - SP5.4-32Rev. OL-21c 1/16Generally, while at cold shutdown condition, decay heat from the reactor core is being removed by the RHRS. The number of pumps and heat exchangers in service depends upon the heat load at the time.
At initiation of the plant startup, the RCS is filled, and the pressurizer heaters are energized. The RHRS is oper ating and is connected to the CVCS via the low pressure letdown line for purification and/or to control reactor coolant pressure.
During this time, the RHRS acts as an alternate letdown path. The manual valves downstream of the residual heat exchangers leading to the letdown line of the CVCS are opened. The control valve in the line from the RHRS to the letdown line of the CVCS is then manually adjusted in the control room to permit letdown flow. RCS pressure control is maintained via the RHRS and the low pressure letdown line until the pressurizer steam bubble is formed. Indication of steam bubble formation is provided in the control room by t he damping out of the RCS pres sure fluctuations and by pressurizer level indication fo r solid plant operati on, or by saturati on temperature for vacuum fill.After the pressurizer steam bubbl e is formed and the reactor coolant pumps are started, the RHRS is isolated from the RCS. RCS pressure is then controlled by normal letdown and the pressurizer spra y and pressurizer heaters.
Power Generation and Hot Standby Operation During power generation and hot st andby operation, th e RHRS is not in service but is aligned for operation as part of the ECCS. EJ-HV-8716A and B and EJ-HV-8809A and B are maintained open during operating modes 1-3 in order that either RHR pump is able to inject to all four RCS cold legs.
Normal Reactor Cooldown Reactor cooldown is defined as the operation which bri ngs the reactor from no-load temperature and pressure to cold conditions. The inital phase of reactor cooldown is acco mplished by transferrin g heat from the RCS to the steam and power conversion system through the use of the steam generators and dumping steam to the condenser.
When the reactor coolant temperature and pr essure are reduced below approximately 350°F and 400 psig, approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after reactor shutdown, the second phase of cooldown starts with the RHRS being placed in operation. Startup of the RHRS includes a warmup period during which time reactor coolant flow through the heat e xchangers is limited to minimize thermal shock.
The rate of heat removal from the reactor coolan t is manually controlled by regulating the coolant flow CALLAWAY - SP5.4-33Rev. OL-21c 1/16through the residual heat exchangers. By adjusting the control valves downstream of the residual heat exchangers, the mixed mean temperature of the return flows is controlled.
Coincident with the manual adjustment, each heat exchanger bypass valve is automatically regulated to give the required total flow. The reactor cooldown rate is limited by RCS equipment cooling rates based on allowable stress limits, as well as the operating temperature limits of the component cooling water system. As the reactor coolant temperature decreases, the reactor coolant flow through the residual heat excha ngers is increased by adjusting t he control valve in each heat exchanger's tube si de outlet line. As cooldown continues, pressure control is accomplished by regulat ing the charging flow rate and the rate of letdown fr om the RHRS to the CVCS. After the reactor coolant pressure is reduced and the temperature is 140°F or lower, the RCS may be opened fo r refueling or maintenance.
RefuelingResidual heat removal pumps are utilized duri ng refueling to pump borated water from the refueling water storage tank to the refueling cavity. During this operation, the RHRS isolation valve(s) in the su ction line(s) from the RCS are closed, and the isolation valve(s) to the refueling water storage tank are opened. The reactor vessel head is lifted off the reactor vessel. Th e refueling water is then pumped into the reactor vessel through the normal RHRS return lines and into the refueling cavity through the open reactor vessel. After the water level reaches the normal refueling level, the RHRS suct ion isolation valve(s) for the RCS are opened, the refueling water storage tank supply valve(s) is(are) closed, and residual heat removal is resumed.
During refueling, the RHRS is maintained in service with the number of pumps and heat exchangers in operation as r equired by the heat load.
Following refueling, the RHR pumps are used to drain the refueling ca vity to the top of the reactor vessel flange by pumping water from the RCS to the ref ueling water storage tank. The vessel head is then replaced and the normal RHRS flowpath re-established. The remainder of the water is removed from the refueling canal via a drain connection in the bottom of the canal. 5.4.7.2.5Control Each inlet line to the RHRS is equipped with a pre ssure relief valve si zed to relieve the combined flow of all the charging pumps at the reli ef valve set pressu re. These relief CALLAWAY - SP5.4-34Rev. OL-21c 1/16 valves also protect the syst em from inadvertent overpr essurization during plant cooldown or startup and provide LTOP for the RCS during low tem perature water solid operation. Each valve has a relief flow capacity of 986 gpm at a set pressure of 450 psig.
Each discharge line from the R HRS to the RCS is equipped with a pressure relief valve to relieve any backleakage through the valves separating t he RHRS from the RCS. Each valve has a relief flow capacity of 20 gpm at a set pressure of 600 psig. These relief valves are loca ted in the ECCS (see Figure 6.3-1). The fluid discharged by the suct ion side relief valves is collec ted in the pressurizer relief tank. The fluid discharged by th e discharge side relief valves is collected in the recycle holdup tank of the boron recycle system.
The design of the RHRS includes two motor-operated gate isol ation valves in series on each inlet line between the high pressure RCS and the lower pressure RHRS. They are closed during normal operations, and are only opened for residual heat removal during a plant cooldown after the RCS pressure is reduced below approximately 400 psig and RCS temperature is reduced to approximately 350°F. During a plant startup, the inlet isolation valves are shut after drawing a bubble in the pressurizer and prior to increasing RCS pressure above 600 psig. These isolat ion valves are provided with "prevent-open" interlocks which are designed to prevent possible exposure of the RHRS to normal RCS operating pressure. The inlet isolation valves in each subsystem are separately and independently interlocked with pressure signals to prevent their being opene d whenever the RCS pressure is greater than 360 psig. A control room alarm will actuate if an RHR suction isolation valve is not fully closed and RC S pressure is greater than the design pressures for RHR system operation.
The use of two independently pow ered, motor-operated valves in each of the two inlet lines, along with two independent pressure interlock signals for each function, assures a design which meets applicable single failure criteria. These protective interlock designs, in combination with plant op erating procedures, provide diverse means of accomplishing the protective functi on. For further information on the instrumentation and control features, see Section 7.6.2. The RHR inlet isolation valves are provided with re d-green position indicator lights on the main control board.
Isolation of the low pressure RHRS from the high pressure RC S is provided on the discharge side by two check valves in series. These check valves are located in the ECCS, and their testin g is described in Section 6.3.4.2. 5.4.7.2.6Applicable Codes and Classifications
The entire RHRS is des igned as Safety Class 2, with the exception of the suction isolation valves, which are Safety Class 1.
Class 1 discharge valves are discussed in Section 6.3. Component codes and classifications are given in Section 3.2.
CALLAWAY - SP5.4-35Rev. OL-21c 1/165.4.7.2.7System Reliabi lity Considerations General Design Criterion 34 requires that a system to remove residual heat be provided.
The safety function of this required system is to transfe r fission product decay heat and other residual heat from the core at a rate sufficient to prevent fuel or pressure boundary design limits from being e xceeded. Safety grade system s are provided in the plant design, both nuclear steam supply system (NSSS) scope and balance-of-plant (BOP) scope, to perform this func tion. The NSSS scope safety grade systems which perform this function for all plant conditions, except a LOCA are: the RCS and steam generators, which operate in conjunction with the auxiliary feedwater system and the steam generator safety and power-operated relief valves; and the RHRS, which operates in conjunction with the component cooling wa ter and service water systems. The BOP scope safety grade systems which perform this function for all plant conditions, except a LOCA, are: the auxili ary feedwater system; the steam generator safety and power-operated relief valves, which operate in conjunction with the RCS and the steam generators; and the component cooling water and service water systems, which operate in conjunction with the RHRS.
For LOCA conditions, the safety grade system which performs the function of removing residual heat from the reactor core is the ECCS, which operates in conjunction with the component cooling wa ter system and the essential service water system.
The auxiliary feedwater system, along with the st eam generator safety and power-operated relief valves , provides a completely separate, independent, and diverse means of performing the safety function of removing resi dual heat, which is normally performed by the RHRS when RCS temperature is less than 350°F. The auxiliary feedwater system is capable of performing this fu nction for an extended period of time following plant shutdown.
The RHRS is provided with tw o residual heat removal pumps and heat exchangers arranged in two separate, independent flow paths. To assure reliability, each residual heat removal pump is connected to a different vital bus. Each train is isolated from the RCS on the suction si de by two motor-operated valves in series with each valve receiving power via a separate motor control center and from a different vital bus. Each suction isolation valve is also interlocked and alarmed to prevent exposure of the RHRS to the normal operating pressure of the RCS (see Section 5.4.7.2.5). RHR system piping and components have the potentia l to develop voids and pockets of entrained gases. Preventing and managing gas intrusion and accumulation in the pump suction and pump discharge piping, however, supports proper operation of the RHR system and may also prevent water hammer, pump cavitation, and pumping of no ncondensible gas into the reactor vessel.RHRS operation for normal conditions and for major failures is acco mplished completely from the control room. Th e redundancy in the RHRS design provides the system with the capability to maintain its cooling func tion even with a major singl e failure, such as CALLAWAY - SP5.4-36Rev. OL-21c 1/16 failure of a residual heat re moval pump, valve, or heat exchanger without impact on the redundant train's continue d heat removal.
Although such major system fail ures are within the system des ign basis, there are other less significant failures whic h can prevent opening of the re sidual heat re moval suction isolation valves from the co ntrol room. Since these fail ures are of a minor nature, improbable to occur, and easily corrected outside the control room, with ample time to do so, they have been realisti cally excluded from the engineering design basis. Such failures are not likely to occu r during the limited time peri od in which they can have any effect (i.e., when opening the suction isolation valves to initiate residual heat removal operation). However, even if they should occur, they have no adverse safety impact and can be readily corrected. In such a situation, the aux iliary feedwater system and the steam generator power-operated relief valves can be used to perform the safety function of removing residual heat and, in fact, can be used to continue the plant cooldown below 350°F, until the RHRS is made available. One example of this type of a failure is the interlock circuitry which is designed to prevent exposure of the RHRS to the normal operating pressure of the RCS (see Section 5.4.7.2.5). In the event of such a failure, RHRS operation can be initiated by defeating the "prevent-open" interlock through corrective action at the solid state protection system cabinet or at the individual af fected motor control centers.
The other type of failure which can prevent opening the residual heat removal suction isolation valves from the contro l room is a failure of an elec trical power train. Such a failure is extremely unlikely to occur during the few minutes out of a year's operating time during which it can have any consequence. If such an unlikely event should occur, several alternatives are available. The most realistic approach would be to obtain restoration of offsite power, which can be expected to occur in less than 1/2 hour. Other alternatives are to restore th e emergency diesel generator to operation or to bring in an alternative power source. The only impact of either of the above types of failures is some delay in initiating residual heat removal operation, while action is taken to open the residual heat removal suction isolation valves. This delay has no adverse safety impact bec ause of the capability of the auxiliary fe edwater system and steam generator power-operated relief valves to continue to remove residual heat, and, in fact, to c ontinue plant cooldown. A failure mode and effects analysis of the RHRS for normal plant cooldown is provided as Table 5.4-9. 5.4.7.2.8Manual Actions
The RHRS is designed to be full y operable from the control room for normal operation.
Manual operations required of the operator are:
opening the suction isolation valves, positioning the flow contro l valves downstream of t he RHRS heat exchangers, and starting the residual heat removal pumps.
CALLAWAY - SP5.4-37Rev. OL-21c 1/16Manual actions required outside the control r oom, under conditions of single failure, are discussed in Section 5.4.7.2.7. 5.4.7.3Performance Evaluation The performance of the RHRS in reducing reac tor coolant temperature is evaluated through the use of heat balance calculations on the RC S, and the component cooling water system at stepwise intervals following the initiation of RHR operation. Heat removal through the RHR and component cooling water heat exchangers is calculated at each interval by use of standard wate r-to-water heat ex changer performance correlations. The resultant fluid temper atures for the RHRS a nd component cooling water system are calculated and used as input to the next interv al's heat balance calculation.
Assumptions utilized in the se ries of the heat balance calculations describing plant RHR cooldown are as follows: a.RHR operation is initiated 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after reactor shutdown (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after shutdown for single train cooldown).b.RHR operation begins at a reactor coolant temperature of 350°F.c.Thermal equilibrium is mainta ined throughout t he RCS during the cooldown. d.Component cooling water temperature at the CCW heat exchanger outlet during cooldown is limited to a maximum of 120°F. e.Expected cooldown rates of 50° F per hour are not exceeded.
Cooldown curves calculated usin g this method are provided for the case of all residual heat removal components operable (Figure 5.4-9) and for the case of a single train residual heat removal cooldown (Figure 5.4-10). 5.4.7.4Preoperational Testing Preoperational testing of t he RHRS is addressed in Chapter 14.0. 5.4.7.5Gas Management The RHR system is operable when it is sufficiently filled with water. The Technical Specifications include Surveillance Requirements for verifying systems are sufficiently full of water. Voiding may occur, however, due to the accumulation of entrained gas; acceptance criteria are estab lished for the volume of accu mulated gas at susceptible locations. If accumulated gas is discovered that exceeds the acceptance criterion for the susceptible location (or if t he volume of accumulated gas at one or more susceptible CALLAWAY - SP5.4-38Rev. OL-21c 1/16locations exceeds an acceptance criterion for gas volume at the suction or discharge of a pump), the Technical Specification Surveillance Requirement is not met and past operability reviews are initiate
- d. If it is determined by subsequent evaluation that the RHR system was not rendered inoperable by the accumulated gas (i.e., the system was sufficiently filled with wa ter), the Surv eillance Requirement ma y be declared met.
Accumulated gas should be eliminated or brought within the acceptance criteria limits.
RHR system locations susceptible to gas accumulation are monitored and, if gas is found, the gas volume is compared to the acceptance cr iteria for the location.
Susceptible locations in the same system flow path that are subject to the same gas intrusion mechanisms may be verified by monitoring a representative subset of susceptible locations. Monitoring may not be practical for locations that are inaccessible due to radiological or environmental condi tions, the plant configuration, or personnel safety. For these lo cations, alternative methods (e.g., operating parameters, remote monitoring) may be used to moni tor the susceptible location. Monitoring is not required for susceptible locations w here the maximum potential a ccumulated gas void volume has been evaluated and determi ned to not challenge system operability. The accuracy of the method used for monitori ng the susceptible locations and trending of the results must be sufficient to assu re system operability between surveillance performances.5.4.8REACTOR WATER CLEANUP SYSTEM This section is not applicable to SNUPPS. 5.4.9MAIN STEAM LINE AND FEED WATER PIPINGDiscussion pertaining to the ma in steam line and feedwater piping are contained in the following sections:a.Main Steam Line Piping -
Section 10.3. b.Main Feedwater Piping -
Section 10.4.7. c.Auxiliary Feedwater Piping -
Section 10.4.9. d.Inservice Inspecti on of a, b, and c -
Section 6.6. 5.4.10PRESSURIZER5.4.10.1Design Bases The pressurizer provides a point in the RC S where liquid and vapor are maintained in equilibrium under saturated condit ions for control of pressu re of the RCS during steady state operations and transients.
CALLAWAY - SP5.4-39Rev. OL-21c 1/16The volume of the pressurizer is equal to, or greater than, the minimum volume of steam, water, or total of the two which satisf ies all of the following requirements: a.The combined saturated water volume and steam expansion volume is sufficient to provide the desired pr essure response to system volume changes. b.The water volume is sufficient to prevent the heaters from being uncovered during a step load increase of 10 percent at full power. c.The steam volume is large enough to accommodate the surge resulting from a 50-percent reduction of full load with automa tic reactor control and a 40-percent steam dump wi thout the water level reaching the high level reactor trip point. d.The steam volume is large enough to prevent water re lief through the safety valves following a loss of load wi th the high water level initiating a reactor trip, without reacto r control or steam dump. e.The pressurizer will not empty following reactor tr ip and turbine trip. f.The emergency core coolin g will not be acti vated because of a reactor trip and turbine trip.
The surge line is sized to mi nimize, to an accept able value, the pr essure drop between the RCS and the safety valves with maximum discharge flow from the safety valves.
The surge line and the thermal sleeves are designed to withstand the thermal stresses resulting from volume surges of water of different temperatures, which occur during operation. 5.4.10.2Design Description5.4.10.2.1Pressuri zer and Surge Line The pressurizer is a vertical , cylindrical vessel with hemis pherical top an d bottom heads constructed of carbon steel, with austenitic stainless steel clad ding on all internal surfaces exposed to the reactor coolant. Stai nless steel is used on all surfaces in contact with the reactor coolant.
The general configuration of t he pressurizer is shown in Figure 5.4-11. The design data of the pressurizer are given in Table 5.4-10. Codes and material requirements are provided in Section 5.2. The pressurizer surge line connects the pressurizer to one reactor hot leg, thus enabling continuous coolant volume pressure adjustments between the RCS and the pressurizer.
CALLAWAY - SP5.4-40Rev. OL-21c 1/16 The surge line nozzle and remo vable electric heaters are lo cated in the bottom of the pressurizer. The heaters are removabl e for maintenance or replacement.
The pressurizer surge line no zzle diameter is given in Table 5.4-10 , and the pressurizer surge line dimensi ons are shown in Figure 5.1-1 , Sheet 2.
A thermal sleeve is provided in the surge line nozzle to minimize thermal stresses. A retaining screen is located above the nozzle to prevent forei gn matter from entering the RCS. Baffles in the lower se ction of the pressu rizer prevent an insurge of cold water from flowing directly to the steam/water interfac e and assist in mixing.Spray line nozzles, relief and safety valve c onnections are located in the top head of the pressurizer vessel. Spray flow is modulat ed by automatically controlled air-operated valves. The spray valves also can be operated manually by a switch in the control room.
A small continuous spray flow is provided through a manual bypass valve around the power-operated spray valves to assure that the boron concentration in the pressurizer is not dissimilar from that in t he reactor coolant and to prev ent excessive cooling of the spray piping. During an outsurge of water from the pressurizer, flashing of water to steam and generation of steam by automatic actuation of the heaters keep the pressure above the minimum allowable limit. During an insurge from the RCS, the sp ray system, which is fed from two cold legs, condenses steam in the vessel to prev ent the pressurizer pressure from reaching the setpoint of the power-operated relief valves for normal design transients. Heaters are energized on hi gh water level during insurge to heat the subcooled surge water that en ters the pressurizer from the reactor coolant loop.
Material specifications are provided in Table 5.2-2 for the pressurizer, pressurizer relief tank, and the surge line. Design transients for the components of the RCS are discussed in Section 3.9(N).1. Additional details on the pressuri zer design cycle analysis are given in Section 5.4.10.3
.5.4.10.2.2Pressurizer Instrumentation Refer to Chapter 7.0 for details of the instrumentat ion associated with pressurizer pressure, level, and temperature. Temperatures in the spray li nes from the cold legs of two loops are measured and indicated. Alarms from thes e signals are actuated to wa rn the operator of low spray water temperature or indicate insuff icient flow in the spray lines. Temperatures in the pressurizer safety an d relief valve discharge lines are measured and indicated. An increase in a discharge line temperature is an indication of leakage or relief through the as sociated valve.
CALLAWAY - SP5.4-41Rev. OL-21c 1/165.4.10.3Design Evaluation5.4.10.3.1System Pressure Whenever a steam volume is present within the pressurizer, t he RCS pressure is governed by conditions in the pressurizer. A design basis safety limit is that RCS pressure does not exceed the maximum transient value allowed under the AS ME Code,Section III.
Evaluation of plant conditions of operation, which follow, indi cate that this safety limit is not reached. During startup and shutdown, the rate of temperature change in the RCS is controlled by the operator. H eatup rate is controlled by energy i nput from the reactor coolant pumps and by the pressurizer electrical heating capacity. This heatup rate takes into account the continuous spray flow provided to the pressurizer. When the reac tor core is in cold shutdown, the heaters are de-energized. If the pressurizer is filled water solid during system heatup or near the end of the second phase of plant cooldown, RCS pressure is maintained by the letdown flow rate via the RHRS. 5.4.10.3.2Pressurizer Performance
The normal operating water volume at full load conditions is given in Table 5.4-10. 5.4.10.3.3Pressure Setpoints The RCS design and operating pressure, together with the safety, power relief, and pressurizer spray valves setpoints and the protection system pressure setpoints, are listed in Table 5.4-11. The design pressure allows for operating transient pressure changes. The selected design margin consi ders core thermal lag, coolant transport times and pressure drops, instrumentation and control res ponse characteri stics, and system relief valve characteristics. 5.4.10.3.4Pressurizer SprayTwo separate, automatically controlled spra y valves with remote manual overrides are used to initiate pressurizer spray. In parallel with each spra y valves is a manual throttle valve which permits a small continuous flow through both spray lines to reduce thermal stresses and thermal shock when the spray valves open and to help maintain uniform water chemistry and temperature in the pressurizer. Te mperature sensors with low alarms are provided in each spray line to alert the operator to insufficient bypass flow.
The layout of the common spra y line piping routed to the pressurizer forms a water seal which prevents the steam buildup back to the control valves.
The spray rate is selected CALLAWAY - SP5.4-42Rev. OL-21c 1/16 to prevent the pressu rizer pressure from reaching the operating setpoint of the power relief valves during a step reduction in power level of 10 percent of full load.
The pressurizer spray lines and valves are large enough to provi de the required spray flow rate under the driving force of the differential pre ssure between the surge line connection in the hot leg and the spray line connection in the cold leg. The spray line inlet connections extend into the cold leg piping in the form of a scoop in order to utilize the velocity head of the reactor coolant loop flow to add to the spray driving force. The spray valves and spray line connections are arranged so that the sp ray will operate when one reactor coolant pump is not operating. The line may al so be used to assist in equalizing the boron concentra tion between the reactor coolant loops and the pressurizer. A flow path from the CVCS to the pressurizer spray line is also provided. This path provides auxiliary spray to the vapor space of the pressurizer during cooldown when the reactor coolant pumps are not operating. The thermal sleeve s on the pres surizer spray connection and the spray piping are designed to withstand the thermal stresses resulting from the introduction of cold spray water. 5.4.10.4Tests and Inspections The pressurizer is designed and construct ed in accordance wi th the ASME Code,Section III. To implement the requirements of the ASME Code,Section XI the following welds are designed and constructed to pr esent a smooth transition surface between the parent metal and the weld metal. Th e weld surface is gr ound smooth for ultrasonic inspection. a.Support skirt to the pressurizer lower headb.Surge nozzle to the lower head c.Nozzles to the safety, relief, and spray lines d.Nozzle to safe end attachment weldse.All girth and longitudinal full penetration weldsf.Manway attachment weldsThe liner within the safe end nozzle region extends beyond the weld region to maintain a uniform geometry for ul trasonic inspection.
Peripheral support rings are furnished for the remo vable insulati on modules.
The pressurizer quality assur ance program is given in Table 5.4-12.
CALLAWAY - SP5.4-43Rev. OL-21c 1/165.4.11PRESSURIZER RELIEF DISCHARGE SYSTEM5.4.11.1Design Bases The pressurizer relief discharge system collects, cools, and directs for processing the steam and water discharged from safety and relief valves in the containment. The system consists of the pressurizer relief tank, the safety and relief valve discharge piping, the relief tank internal sp ray header and associated piping, the tank nitrogen supply, the vent to containment, and the drain to the wast e processing system.
The system design is based on the require ment to absorb a discharge of steam equivalent to 110 percent of the full power pressurizer steam volume. The steam volume requirement is approximately that which would be experienced if the plant were to suffer a complete loss of load accompanied by a turbine trip but without the resulting reactor trip. A delayed reactor tr ip is considered in th e design of the system.
The minimum volume of water in the pressurizer relief tank is determined by the energy content of the steam to be condensed and cooled, by the assu med initial temperature of the water, and by the desired final temperatur e of the water volume. The initial water temperature is assumed to be 120°F, which correspond s to the design maximum expected containment temperatur e for normal conditions. Pr ovision is made to permit cooling the tank should the water temperature rise above 120°F during plant operation. The design final temperature is 200°F, which allows the content of the tank to be drained directly to the waste proce ssing system without cooling. A safety-related flowpath downstream of the excess letdown heat exchanger is provided to direct a cooled flow to the PRT. This flow path may be used if the normal and excess letdown paths are unavail able or if it is desired to cont ain the reactor c oolant inside the containment. Another flowpath is provided for the controll ed release of fluid from the PRT to the containment normal sump. The vessel saddle supports and anchor bolt arrangement are designed to withstand the loadings resulting fr om a combination of nozzle loadi ngs acting simult aneously with the vessel seismic and static loadings. 5.4.11.2System DescriptionThe piping and instrumentation diagram for the pressurizer relie f discharge system is given in Figure 5.1-1 , Sheet 2.
Codes and materials of the pressurizer relief tank and a ssociated piping are given in Section 5.2. Design data for the tank are given in Table 5.4-13. The steam and water discharged from the va rious safety and relief valves inside the containment is routed to the pressurizer relief tank if the discharged fluid is of reactor CALLAWAY - SP5.4-44Rev. OL-21c 1/16grade quality. Table 5.4-14 provides an itemized list of valves discharging to the tank, together with references to the correspondi ng piping and instrumentation diagrams. The tank normally contains water and a predominantly nitrogen atmosphere. In order to obtain effective condensing and cooling of the discharged steam, the tank is installed horizontally with the steam discharged through a sparger pipe located near the tank bottom and under the water level. The sparger holes are designed to ensure a resultant steam velocity close to sonic. The water in the tank may be discharged to allow increased capacity for RC letdown via the excess letdown path. In this mode, the water is cooled before it enters the tank. The tank is also equipped with an internal spray and a drain which are used to cool the water following a discharge. Cold water is drawn from the reactor makeup water system, or the contents of the tank are circulated through the reactor coolant drain tank heat exchanger of the waste processing system and back into the spray header. The nitrogen gas blanket is used to control the atmosphere in the tank and to allow room for the expansion of the original water pl us the condensed steam discharge. The tank gas volume is calculated, using a final pressure based on an arbitrary design pressure of 100 psig. The design discharge raises the worst case initial conditions to 50 psig, a pressure low enough to prevent fatigue of the rupture discs. Provision is made to permit the gas in the tank to be periodically analyzed to moni tor the concentrati on of hydrogen and/or oxygen. The contents of the tank can be drained to the waste holdup tank in t he waste processing system or the recycle holdup ta nk in the boron recycle system via the reactor coolant drain tank pumps in the waste processing system. Un der emergency conditions, the tank contents can be drained to the containment normal sump. 5.4.11.2.1Pressurizer Relief Tank
The general configuration of the pressurizer relief tank is shown in Figure 5.4-12. The tank is a horizontal, cylindrical vessel wi th elliptical dished heads. The vessel is constructed of austenitic stainless steel, and is overpressure protected in accordance with the ASME Code, Secti on VIII, Division 1, by mean s of two safety heads with stainless steel rupture discs. The PRT saddle supports are designed to withstand the loadings resulting fr om a combination of nozzle loadi ngs acting simult aneously with the vessel seismic and static loadings.
A flange nozzle is provided on the tank for the pressurize r discharge line connection to the sparger pipe. The tank is also equipped with an internal spray connect ed to a cold water inlet and with a bottom drain, which are used to cool the tank following a discharge.
CALLAWAY - SP5.4-45Rev. OL-21c 1/165.4.11.3Design Evaluation The pressurizer relief discharge system does not constitute part of t he reactor coolant pressure boundary per 10 CF R 50, Section 50.2, since all of its components are downstream of the RCS safety and relief valves. Thus, G eneral Design Criteria 14 and 15 are not applicable. Furthermore, complete failure of the auxiliary systems serving the pressurizer relief tank will not impair the capability for safe plant shutdown.
The design of the system piping layout and piping restraints is consistent with the hazards protection requirements indicated in Appendix 3.B. The safety and relief valve discharge piping is restrained so that the integrity and operability of the valves are maintained in the event of a rupture. Regulatory Guide 1.
67 is not applicab le, since the system is not an open discharge system.
The pressurizer relief discharge system is capable of handl ing the design discharge of steam without exceeding the desig n pressure and temperature of the pressurizer relief tank. The volume of water in the pressurizer relief tank is cap able of absorbing the heat from the assumed discharge, maintain ing the water temperature below 200°F. If a discharge exceeding the design basis should occur, the relief device on the tank would pass the discharge through the tank to the containment. The rupture discs on the relief tank have a relief capacity equal to or greater than the combined capacity of the pressurizer safety valves. The t ank design pressure is twice the calculated pressure resulting from the design basis safety valve discharge described in Section 5.4.11.1. The tank and rupture discs holders are also designed for full vacuum to prevent tank collapse if the content cools following a discharge without nitrogren being added. The discharge piping from the pre ssurizer safety and relief valves to the relief tank is sufficiently large to prevent backpressure at the safety valves from exceeding 20 percent of the setpoint pressure at full flow. 5.4.11.4Instrumentation RequirementsThe pressurizer relief tank pre ssure transmitter provides an in dication of pressure relief tank pressure. An alarm is provided to indicate high tank pressure. The pressurizer relief tank le vel transmitter supplies a sign al for an indicator with high and low level alarms. The temp erature of the water in the pressurizer relief tank is indicated, and an alarm actuat ed by high temperature inform s the operator that cooling of the tank contents is required.
CALLAWAY - SP5.4-46Rev. OL-21c 1/165.4.11.5Tests and InspectionsThe system components are subject to nondestructive and hydrostatical testing during construction, in accordance with Section VIII, Division 1 of the ASME Code.
During plant operation, periodic visual inspections and preventive maintenance are conducted on the system components accord ing to normal indus trial practice. 5.4.12VALVES 5.4.12.1Design BasesAs noted in Section 5.2, all valves out to and including the second valve normally closed or capable of automatic or remote closure, larger than 3/4 inch, are ANS Safety Class 1, and ASME III, Code Class 1 valves.
All 3/4-inch or smaller va lves in lines connected to the RCS are Class 2, since the interface with the Class 1 piping is provided with suitable orificing for such valves. Design dat a for the RCS valves are given in Table 5.4-15. For a check valve to qualify as part of the RCS, it must be located inside the containment system. When the second of tw o normally open check valves is considered part of the RCS (as defined in Section 5.1
), means are provided to per iodically assess back-flow leakage of the first valve when closed. To ensure that the valves wil l meet the design objectives, the materials of construction minimize corrosion/erosion and ensure compatibility with the environment. Leakage is minimized to the extent practicable by design. 5.4.12.2Design Description All manual, motor oper ated, and throttling control valv es are provided with either double-packed stuffing boxes and intermediate lantern ring leakoff connections which are piped to a closed collection system when so equipped, or a four or five ring packing configuration and carbon spacer (if required). Both packing configurations minimize leakage to atmosphere to the ex tent practicable by design.
Gate valves at the engineer ed safety features interf ace are wedge design and are essentially straight through.
The wedges are flex-wedge or so lid. All gate valves have backseats.
Globe valves are "T" and "Y" styl es. Check valves are swing type for sizes 2-1/2 inches and larger. All check valves which contain radioactive fluid are stainless steel, and do not have body penetrati ons other than the inlet, outlet, and bonnet. The check hinge is serviced through the bonnet. All operating parts are contained within the valve body. The disc has limited rotation to provide a change of seating surface and alignment af ter each valve opening.
CALLAWAY - SP5.4-47Rev. OL-21c 1/165.4.12.3Design EvaluationThe design requirements for Class 1 valves, as discussed in Section 5.2 , limit stresses to levels which ensure the struct ural integrity of the valves. In addition, the testing programs described in Section 3.9(N) demonstrate the ability of the valves to operate, as required, during anticipated and postulated plant conditions. Reactor coolant chemistry parameters are s pecified in the design specifications to assure the compatibility of valve construction materials with the reactor coolant. To ensure that the reacto r coolant continues to meet these parameters, the chemical composition of the coolant will be analyzed periodically. The above requirements and pr ocedures, coupled with the pr eviously described design features for minimizing leak age, ensure that the valves will perform their intended functions, as required du ring plant operation. 5.4.12.4Tests and InspectionsThe tests and inspections discussed in Section 3.9(N) are performed to ensure the operability of the active valves. There are no full-penetration welds within the valve body walls. Valves are accessible for disassembly and internal visual inspection, to the extent practi cal. Plant layout configurations determine the degree of inspectability.
The valve nondestructive examination program is given in Table 5.4-16. Inservice inspection is discussed in Section 5.2.4. 5.4.13SAFETY AND RELIEF VALVES5.4.13.1Design Bases Given that the steam generator safety valv es open when steam pr essure reaches the steam side safety setting, the combined capacity of the pr essurizer safety valves can accommodate the maximum pressurizer surge resulting from comp lete loss of load, without reactor trip or any operator action.
The pressurizer power-operated relief valves are designed to limit pressurizer pressure to a value below the fixed high pressure reactor trip setpoint. They are designed to fail to the closed position on loss of power. 5.4.13.2Design DescriptionThe pressurizer safety valves are of the pop type. The valves are spring loaded, open by direct fluid pressure action, and have backpressure compensation features.
CALLAWAY - SP5.4-48Rev. OL-21c 1/16The pipe connecting each pressurizer nozzle to its safety valve is shaped in the form of a loop seal. Conden sate resulting fr om normal heat losses accumu lates in the loop. The water prevents any leakage of hydrogen gas or steam through the safety valve seats. If the pressurizer pressure exceeds the set pressure of the safety valves, they start lifting, and the water from the seal dischar ges during the actuation period.
For any increases made to core differential pressure or pressurization rates from core reload or plant modifica tions, the loop seal pur ge time for the pressurizer safety valves will be re-examined. A complete discussion of the loop seal purge ti me can be found in Reference 4.
The pressurizer power-operated relief valves are sol enoid actuated valves which respond to a signal from a pressure s ensing system or to manual control.
Motor-operated valves are provided to is olate the power-operat ed relief valves if excessive leakage develops or if the PORV fails to close. Temperatures in the pressurizer safety an d relief valve discharge lines are measured and indicated. An increase in a discharge line temperature is an indication of leakage or relief through the as sociated valve.
The power-operated relief valv es provide the safety-related means for reactor coolant system depressurization to achieve cold shutdown. Design parameters for the pressurizer safety and power relief valves are given in Table 5.4-17. 5.4.13.3Design Evaluation The pressurizer safety valves prevent RCS pressure from exceeding 110 percent of system design pressure, in compliance with the ASME Code,Section III.
The pressurizer power relief valves prevent ac tuation of the fixed r eactor high pressure trip for all design transient s up to and including the desi gn step load decreases with steam dump. The relief valves also limit undesirable opening of the spring loaded safety valves. 5.4.13.4Tests and Inspections All safety and relief valves are subjected to hydrostatic tests, seat leakage tests, operational tests, and inspections, as required. For safety valves that are required to function during a fault ed condition, additional tests are performed. These tests are described in Section 3.9(N). There are no full penetration welds within the valve body walls. Valves are accessible for disasse mbly and internal vi sual inspection.
CALLAWAY - SP5.4-49Rev. OL-21c 1/165.4.14COMPONENT SUPPORTS5.4.14.1Design BasesComponent supports allow unres trained lateral ther mal movement of the loop during plant operation and provide restraint to the loops and components during accident and seismic conditions. The loadi ng combinations and design stress limits are discussed in Section 3.9(N).1.4. Support design is in accordance with the ASME Code,Section III, Subsection NF. The design maintains the integrity of the RCS boundary for normal, seismic, and accident conditi ons and satisfies the requirements of the piping code. Results of piping and supports stress evaluation are presented in Section 3.9(N). 5.4.14.2Design Description The support structures are weld ed structural steel sections. Linear type structures (tension and compression struts, columns, and beams) are used in all cases, except for the reactor vessel supports, which are plate-type structures. Attachments to the supported equipment ar e nonintegral type that are bolted to or bear against the components. The supports-to-concrete attachments are either anchor bolts or embedded fabricated assemblies. The supports permit unrestrained thermal grow th of the supported systems but restrain vertical, lateral, and rotational movement resulting from seismic and pipe break loadings.
This is accomplished using spherical bushi ngs in the columns for vertical support and girders, bumper pedestals, and tie-rods for lateral support. To compensate for manufactur ing and construction tolerances, adjustment in the support structures is provided to ensure proper erection alignment and fit-up. This is accomplished by shimming or grouting at the supports-to-concrete interface and by shimming at the supports-t o-equipment interface. The supports for the various components are described in the following paragraphs. 5.4.14.2.1Reactor Pressure VesselSupports for the reactor vessel (Figure 5.4-13) are individual air cooled rectangular box structures beneath the vessel no zzles bolted to the primary sh ield wall concrete. Each box structure consists of a horizontal top plate that receives loads from the reactor vessel shoe, a horizontal bottom plat e which transfers the loads to the primary shield wall concrete, and connecting vertical plates wh ich bear against an em bedded support. The supports are air cooled to maintain the supporting concrete te mperature within acceptable levels.
CALLAWAY - SP5.4-50Rev. OL-21c 1/165.4.14.2.2Steam Generator As shown in Figure 5.4-14 , the steam generator supports consist of the following elements: a.Vertical supportFour individual columns provide vertical support for each steam generator. These are bolted at the top to the steam generator and at the bottom to the concrete structure. Spherical ball bushings at the top and bottom of each column allow unrestrained lateral movement of the steam generator during heatup and cooldown. The column base design permits both horizontal and vertical adjustment of the steam generator for er ection and adjustment of the system. b.Lower lateral support Lateral support is provided at the generator tube s heet by fabr icated steel girders and struts. These are bolted to the compartment walls and include bumpers that bear against the steam generator but permit unrestrained movement of the steam generator during changes in system temperature. Stresses in the beams caused by wall displacement during compartment pressurizaton are consi dered in the design. c.Upper lateral supportThe upper lateral support of the steam generator is provided by a ring band at the operating deck. One-way acting limit stops restrain sudden seismic or blowdown induced moti on, but permit t he normal thermal movement of the steam generator. Movement perpendicular to the thermal growth direction of the steam generator is preven ted by shear keys. 5.4.14.2.3Reactor Coolant Pump
Three individual column s, similar to those used for the steam generator, provide the vertical support for each pump. Lateral support for seis mic and blowdown loading is provided by three lateral tension tie bars. The pump supports are shown in Figure 5.4-15. 5.4.14.2.4PressurizerThe supports for the pressurizer, as shown in Figures 5.4-16 and 5.4-17, consist of: a.A steel ring between the pressurizer skirt and the supporting concrete slab.
The ring serves as a leveling and adjusting member for the pressurizer, CALLAWAY - SP5.4-51Rev. OL-21c 1/16 and may also be used as a template for po sitioning the concrete anchor bolts. b.The upper lateral support consists of struts cantilevered off the compartment walls that bear agains t the "seismic lugs" provided on the pressurizer. 5.4.14.2.5Pipe Restraintsa.Crossover leg Restraint at each elbow of the reactor coolant pi pe between th e pump and the steam generator (crossover leg) is not r equired for postulated breaks due to implementatio n of leak-before-break methodology. The horizontal and vertical shim plates from the elbow of the s addle block were removed to deactivate the two cr ossover leg elbow restraints as shown in Figure 5.4-18. The RCS vertical crossover leg pipe whip restraint was deactivated by removal of the ti e rod and associated pins as shown in Figure 5.4-19. b.Hot leg The hot leg restraint located near the 50 degree elbow in the reactor coolant system hot leg wa s also removed due to implementation of leak-before-break methodology. It was deac tivated by remo val of the pipe saddle and shim pa cks as shown on Figure 5.4-20. c.Hot leg and cold leg lateral restraints A restraint on each reactor coolant system hot le g and cold leg is located near the reactor vessel safe-end to reactor coolant system piping weld with the reactor vessel primary shield wall to prevent excessive displacement of either the hot leg or the cold leg following a postul ated guillotine break at the reactor vessel safe-end to piping weld. These restraints are shown in Figure 5.4-21. 5.4.14.3Design EvaluationDetailed evaluation ensures the design adequacy and structural integr ity of the reactor coolant loop and the primary equipment supports system. This detailed evaluation is made by comparing the analytical results with established criteria for acceptability. Structural analyses are performed to demonstrate de sign adequacy for safety and reliability of the plant in case of a larg e or small seismic di sturbance and/or LOCA conditions. Loads whic h the system is expected to encounter often during its lifetime (thermal, weight, and pressure) are applied, and stresses are comp ared to allowable values as described in Section 3.9(N).1.4.
CALLAWAY - SP5.4-52Rev. OL-21c 1/16The safe shutdown earthqua ke and design basis LOCA , resulting in a rapid depressurization of the the syst em, are required design condi tions for public health and safety. The methods used for the analysis of the safe shutdown earthquake and LOCA conditions are given in Sections 3.9(N).1.4. 5.4.14.4Tests and Inspections Nondestructive examinations are performed in accordance with the procedures of the ASME Code,Section V, except as modified by the ASME C ode,Section III, Subsection NF. 5.4.15REFERENCES 1."Reactor Coolant Pump Integrity in LOCA," WCAP-8163, September, 1973. 2.Eggleston, F. T., "Safet y-Related Research and Development for Westinghouse Pressurized Water Reactor, Program Su mmaries - Winter 1 977 - Summer 1978," WCAP-8768, Revision 2, October, 1978. 3.Deleted. 4.Barrett, G. O., et al., "Pressurizer Safety Valve Set Pressure Shift," WCAP-12910, March 1991.
CALLAWAY - SP Rev. OL-13 5/03TABLE 5.4-1 REAC TOR COOLANT PUMP DESIGN PARAMETERSUnit design pressure, psig2,485Unit design temperature,°F650 (a)
Unit overall height, ft26.93Seal water injection, gpm8Seal water return, gpm3 Cooling water flow, gpm366 Maximum continuous cooling water inlet temperature 105 PumpCapacity, gpm100,200 +/- 2000 (hot)Developed head, ft 290 +/- 10 (hot)NPSH required, ft Figure 5.4-2Suction temperature,°F556.6 Pump discharge nozzle, inside diameter, in.
27-1/2 Pump suction nozzle, inside diameter, in.
31Speed, rpm1,187Water volume (casing), ft 3 78.6Weight total (including pump casing, motor, and motor supports), dry, lb 204,035 (b)MotorTypeDrip proof, squirrel cage induction, water/
air cooledPower, hp7,000 Voltage, Volts13,200Phase3Frequency, Hz60 Insulation classClass F, ther malastic epoxy insulation CALLAWAY - SPTABLE 5.4-1 (Sheet 2)
Rev. OL-13 5/03StartingCurrent1,600 amp @ 13,200 VoltsInput, hot reactor coolant265 +/- 5 ampInput, cold reactor coolant336 +/- 7 amp Pump moment of inertia, maximum (lb-ft 2)Flywheel64,000Shaft 745Impeller 1,980 Rotor core27,700Runner 675Coupling 190(a)Design temperature of pressure-retaining parts of the pump assembly exposed to the reactor coolant and injection water on the high pressure side of the controlled leakage seal shall be that temperature determined for the parts for the primary loop temperature of 650°F.(b)Approximately 206,000 lbs. if studs and nuts ar e utilized in the main flange joint.
CALLAWAY - SP Rev. OL-13 5/03TABLE 5.4-2 REAC TOR COOLANT PUMP QUALITY ASSURANCE PROGRAM RT**RT - Radiographic UT - Ultrasonic
PT - Dye penetrant MT - Magnetic particle UT*PT*MT*CastingsYesYes ForgingsMain shaftYesYesMain studsYesYes Flywheel (rolled plate)YesWeldmentsCircumferentialYesYesInstrumentant connectionsYes CALLAWAY - SP Rev. OL-15 5/06TABLE 5.4-3 STEAM GENERATOR DESIGN DATADesign pressure, reactor coolant side, psig2,485Design pressure, steam side, psig1,185 Design pressure, primary to secondary, psi1,600 Design temperature, reactor coolant side,°F 650 Design temperature, steam side,°F 600 Design temperature, primary to secondary,°F 650Total heat transfer surface area, ft 2 78,946Maximum moisture carryover, wt percent 0.10Overall height, ft-in.68-4 Number of U-tubes5,872 U-tube nominal diameter, in.0.75 Tube wall nominal thickness, in.0.0429Number of manways 4
Inside diameter of manways, in. 16 Number of handholes 6Design fouling factor, ft 2-hr-F/Btu 0.00005Steam flow (per unit), lb/hr 3.99 x 10 6Nominal primary side water volume, ft 3No loadFull load 1,345.9 1,345.9Nominal secondary side water volume, ft 3No loadFull load 3,322 2,358 CALLAWAY - SP Rev. OL-15 5/06TABLE 5.4-4 STEAM GENERATOR QUALITY ASSURANCE PROGRAM RT (a)UT (a)PT (a)MT (a)ET (a)Tube SheetForgingYesYes
CladdingYes(b)Yes Channel Head (if fabricated)
FabricationYes (c)Yes(d)YesCladdingYes Secondary Shell and Head (forgings)YesTubesYesYes Nozzles (Forgings)YesYes Weldments Shell, circumferentialYesYesCladding (channel headtube sheet joint cladding restoration)YesPrimary nozzles to fab headYesYes Manways to fab headYesYes Steam and feedwater nozzle to shellYesYesSupport bracketsYes Tube to tube sheetYesInstrument connections (primary and secondary)YesTemporary attachments after removalYes CALLAWAY - SPTABLE 5.4-4 (Sheet 2)
Rev. OL-13 5/03After hydrostatic test (all major presssure boundary welds and complete cast channel head - where accessible)YesNozzle safe ends (if weld deposit)YesYes(a)RT - Radiographic UT - Ultrasonic PT - Dye penetrantMT - Magnetic particleET - Eddy Current(b)Flat surfaces only(c)Weld deposit(d)Base material only RT (a)UT (a)PT (a)MT (a)ET (a)
CALLAWAY - SP Rev. OL-13 5/03TABLE 5.4-5 REACTOR COOLANT PIPING DESIGN PARAMETERS Reactor inlet piping, inside diameter, in.27-1/2 Reactor inlet piping, nominal wall thickness, in.2.32 Reactor outlet piping, inside diameter, in. 29 Reactor outlet piping, nominal wall thickness, in.2.45 Coolant pump suction piping, inside diameter, in.31 Coolant pump suction piping, nominal wall thickness, in.2.60 Pressurizer surge line piping , nominal pipe size, in.14 Pressurizer surge line piping, nominal wall thickness, in.1.406 Nominal water volume, all four loops including surge line, ft 3 1,225 Reactor Coolant Loop PipingDesign/operating pressure, psig Design temperature, °F 2,485/2,235 650 Pressurizer Surge LineDesign pressure, psig Design temperature,°F 2,485 680Pressurizer Safety Valve Inlet LineDesign pressure, psig Design temperature,°F 2,485 680Pressurizer (Power-Operated) Relief Valve Inlet LineDesign pressure, psig Design temperature,°F 2,485 680 CALLAWAY - SP Rev. OL-13 5/03TABLE 5.4-6 REACTOR COOLANT PIPING QUALITY ASSURANCE PROGRAM RT**RT - Radiographic UT - Ultrasonic
PT - Dye Penetrant UT*PT*Fittings and Pipe (Castings)YesYesFittings and Pipe (Forgings)YesYes WeldmentsCircumferentialYesYes
Nozzle to runpipe (except no RT for nozzles less than 6 inches) YesYesInstrument connectionsYesCastingsYesYes (after finishing)ForgingsYesYes (after finishing)
CALLAWAY - SP Rev. OL-21 5/15TABLE 5.4-7 DESIGN BASES FOR RESIDU AL HEAT REMOVAL SYSTEM OPERATION(1)Maximum temperature at the CCW heat exchanger outlet at 18.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> after plant shutdown.(2)Refer to Section 5.4.7.2.1. (3)Refer to Section 5.4.7.2.1 and Figure 5.4-10 for single train cooldown which will take the plant to cold shutdown (200°F) within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after shutdown (actual value is 30.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after shutdown).Residual heat removal system startup, hours after reactor shutdown
~4 3Reactor coolant system initial pressure, psig~400 Reactor coolant system initial temperature,°F~350 Component cooling water design temperature,°F 105 1Cooldown time, hours after initiation of residual heat remova l system operation 14.9 3 Reactor coolant system temperat ure at end of cooldown,°F 140 3 Decay heat generation at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> after reactor shutdown, Btu/hr 78.9 x 10 6 2 CALLAWAY - SP Rev. OL-21 5/15TABLE 5.4-8 RESIDUAL HEAT REMOVAL SYSTEM COMPONENT DATA*Maximum temperatures at 18.
9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> after plant shutdown.Residual Heat Removal PumpsNumber2Design pressure, psig600Design temperature,°F400 Design flow, gpm3,800Design head, ft350NPSH required at 3,800 gpm, ft17 Power, hp500 Residual Heat ExchangersNumber2Design heat removal capacity, Btu/hr 39.1 x 10 6Estimated UA, Btu/hr F LMTD 2.3 x 10 6TubeSideShellSideDesign pressure, psig600150Design temperature, F400200Design flow, lb/hr 1.9 x 10 6 3.8 x 10 6Inlet temperature, F*140105Outlet temperature, F*120116MaterialAustenitic stainless steel Carbon steelFluidReactor coolant Component cooling waterRHR Isolation Valve Encapsulation Tank (TEJ01A & B)ManufacturerRichmond Eng.Quantity2Height ft-in.12-6 Diameter ft-in.5-6Design Pressure, psig75Design Temperature,°F400 MaterialAustenitic stainless steelCodes and StandardsASME Section III, Class 2Seismic CategoryI CALLAWAY - SP Rev. OL-13 5/03TABLE 5.4-9 FAILURE MODES AND EFFECTS ANALYSIS - RESIDUAL HEAT REMOVAL SYSTEM ACTIVE COMPONENTS - PLANT COOLDOWN OPERATIONComponentFailureModeEffectonSystemOperation*FailureDetectionMethod**Remarks1.Motor-operated gate valve EJ-HV-8701A (EJ-HV-8701B
analogous)Fails to open on demand (open manual mode CB switch selection)Failure blocks reactor coolant flow from hot leg of RC loop 1through train "A" of RHRS. Fault reduces redundancy of RHR coolant trains provided. No effect on safety for system operation. Plant cooldown requirements will be met by reactor coolant flow from hot leg of RC loop 4 flowing through train "B" of RHRS. However, time required to reduce RCS temperature will be extended.Valve position indication (closed to open position change) at CB; RHR train "A" discharge flow indication (EJ-FI-618) and low flow alarm at CB; and RHR pump discharge pressure indication (EJ-PI-614) at CB.1.Valve is electrically interlocked with the containment sump isolation valve EJ-HV-8811A and the RWST isolation valve BN-HV-8812A, with RHR to charging pump suction line isolation valve EJ-HV-8804A and with a "prevent-open" pressure interlock (BB-PB-405A) off the seal table. The valve cannot be opened remotely from the CB if one of the indicated isolation valves is open or if RC loop pressure exceeds 360 psig. The valve can be manually opened. See Section 5.4.7.2.7.2.If both trains of RHRS are unavailable for plant cooldown due to multiple component failures, the auxiliary feedwater system and SG power-operated relief valves can be used to perform the safety function of removing residual heat.2.Motor-operated gate valve BB-PV-8702A (BB-PV-8702B analogous)Same failure modes as those stated for item 1.Same effect on system operation as that stated for item 1.Same methods of detection as those stated for item 1.Same remarks as those stated for item 1, except for pressure interlock (BB-PB-403A) control.*See list at end of table for definition of acronyms and abbreviations used.**As part of plant operation; periodic tests, surveillance inspections, and instrument calibrations are made to monitor equipment and performance. Failures may be detected during such monitoring of equipment, in addition to detection methods noted.
CALLAWAY - SPTABLE 5.4-9 (Sheet 2)
Rev. OL-13 5/033.RHR pump 1 (RHR pump 2 analogous)Fails to deliver working fluid.Failure results in loss of reactor coolant flow from hot leg of RC loop 1 through train "A" of RHRS. Fault reduces redundancy of RHR coolant trains provided. No effect on safety for system operation. Plant cooldown requirements will be met by reactor coolant flow from hot leg of RC loop 4 flowing through train "B" of RHRS. However, time required to reduce RCS temperature will be extended.Open pump switchgear circuit breaker indication at CB; circuit breaker close position monitor light for group monitoring of components at CB; common breaker trip alarm at CB; RHR train "A" discharge flow indication (EJ-FI-618) and low flow alarm at CB; and pump discharge pressure indication (EJ-PI-614) at CB.The RHRS shares components with the ECCS. Pumps are tested as part of the ECCS testing program (see Section 6.3.4
). Pump failure may also be detected during ECCS testing.4.Motor-operated gate valve EJ-FCV-610 (EJ-FCV-611 analogous)a.Fails to open on demand (open
manual mode CB switch selection).Failure blocks miniflow line to suction of RHR pump "A" during cooldown operation or during checking boron concentration level of coolant in train "A" of RHRS via EJ-HV-14. No effect on safety for system operation. Operator may establish miniflow for RHR pump "A" operation by opening of CVCS letdown control valve BG-HCV-128 and manual valve EJ-V001 to allow flow to CVCS. If RHR train "A" is degraded, plant cooldown requirements will be met by redundant RHR train "B".
However, time required to reduce RCS temperature will be extended.Valve position indication (closed to open position change) at CB.Valve is automatically controlled to open when pump discharge is less than 816 gpm at 300°F (783 gpm at 68°F) and close when the discharge exceeds 1,650 gpm at 300°F (1,582 gpm at 68°F). The valve protects the pump against overheating and loss of discharge flow from the the pump. CB switch set to "Auto" position for automatic control of valve positioning.b.Fails to close on demand ("Auto" mode CB switch selection).Failure allows for a portion of RHR heat exchanger "A" discharge flow to be bypassed to suction of RHR pump "A."
RHRS train "A" is degraded for the regulation of coolant temperature by RHR heat exchanger "A." No effect on safety for system operation. Cooldown of RCS within established sp ecification cooldown rate may be accomplished through operator action of throttling flow control valve EJ-HCV-606 and controlling cool-down with redundant RHRS train "B."Valve position indication (open to closed position change) and RHRS train "A" discharge flow indication (EJ-FI-618) at
CB.ComponentFailureModeEffectonSystemOperation*FailureDetectionMethod**Remarks CALLAWAY - SPTABLE 5.4-9 (Sheet 3)
Rev. OL-13 5/035.Air diaphragm-operated butterfly valve EJ-FCV-618 (EJ-FCV-619 analogous)a.Fails to open on demand ("Auto" mode CB switch selection)Failure prevents coolant discharged from RHR pump "A" from bypassing RHR heat exchanger "A" resulting in mixed mean temperature of coolant flow to RCS being low. RHRS train "A" is degraded for the regulation of controlling temperature of coolant. No effect on safety for system operation. Cooldown of RCS within established specification rate may be accomplished through operator action of throttling flow control valve EJ-HCV-606 and controlling cooldown with redundant RHRS train "B".RHR pump "A" discharge flow temperature and RHRS train "A" discharge to RCS cold leg flow temperature recording (EJ-TR-612) at CB; and RHRS train "A" discharge to RCS cold leg flow indication (EJ-FI-618) at CB.Valve is designed to fail "closed" and is electrically wired so that electrical solenoid of the air diaphragm operator is energized to open the valve. Valve is normally "closed" to align RHRS for ECCS operation during plant power operation and load follow.b.Fails to close on demand ("Auto" mode CB switch selection).Failure allows coolant discharged From RHR pump "A" to bypass RHR heat exchanger "A", resulting in mixed mean temperature of coolant flow to RCS being high. RHRS train "A" is degraded for the regulation of controlling temperature of coolant. No effect on safety for system operation. Cooldown of RCS within established specification rate may be accomplished through operator action of throttling flow control valve EJ-HCV-606 and controlling cooldown with redundant RHRS train "B." However, cooldown time will be extended. Same methods of detection as those stated for item 5.a.6.Air diaphragm-operated butterfly valve EJ-HCV-606 (EJ-HCV-607 analogous)a.Fails to close on demand for flow reduction.Failure prevents control of coolant discharge flow from RHR heat exchanger "A," resulting in loss of mixed mean temperature coolant flow adjustment to RCS. No effect on safety for system operation. Cooldown of RCS within established specification rate may be accomplished by operator action of controlling cooldown with redundant RHRS
train "B."Same methods of detection as those stated for item 5.a. In addition, monitor light and alarm (valve closed) for group monitoring of components at CB. Valve is designed to fail "open". Valve is normally "open" to align RHRS for ECCS operation during plant power operation and load follow.ComponentFailureModeEffectonSystemOperation*FailureDetectionMethod**Remarks CALLAWAY - SPTABLE 5.4-9 (Sheet 4)
Rev. OL-13 5/03b.Fails to open on demand for increased flow.Same effect on system operation as that stated for item 6.a.Same methods of detection as those stated for item 6.a.7.Manual globe valve EJ-V001 (EJ-V002 analogous)Fails closed.Failure blocks flow from train "A" of RHRS to CVCS letdown heat exchanger. Fault prevents (during the initial phase of plant cooldown) the adjustment of boron concentration level of coolant in lines of RHRS train "A" so that it equals the concentration level in the RCS, using the RHR cleanup line to CVCS. No effect on safety for system operation. Operator can balance boron concentration levels by cracking open flow control valve EJ-HCV-606 to permit flow to cold leg of loop 1 of RCS in order to balance levels using normal CVCS letdown flow.CVCS letdown flow indication (BG-FI-132) at CB.Valve is normally "closed" to align RHRS for ECCS operation during plant power operation and load follow.8.Air diaphragm-operated globe valve BG-HCV-128Fails to open on demand.Failure blocks flow from trains "A" and "B" of RHRS to CVCS letdown heat exchanger. Fault prevents use of RHR cleanup line to CVCS for balancing boron concentration levels of RHR trains "A" and "B" with RCS during initial cooldown operation and later in plant cooldown for letdown flow. No effect on safety for system operation. Operator can balance boron concentration levels with similar
actions, using pertinent flow control valve EJ-HCV-606 (EJ-HCV-607), as stated for item 7. Normal CVCS letdown flow can be used for purification if RHRS cleanup line is not available. Valve position indication (degree of openings) at CB and CVCS letdown flow indication1.Same remark as that stated above for item 7. (BG-FI-132) at CB. 2.Valve is a component of the CVCS that performs an RHR function during plant cooldown operation.ComponentFailureModeEffectonSystemOperation*FailureDetectionMethod**Remarks CALLAWAY - SPTABLE 5.4-9 (Sheet 5)
Rev. OL-13 5/03List of acronyms and abbreviationsAuto - AutomaticCB - Control boardCVCS - Chemical and volume control system ECCS - Emergency core cooling systemRC - Reactor coolantRCS - Reactor coolant system RHR - Residual heat removalRHRS - Residual heat removal systemRWST - Refueling water storage tank SG - Steam generator9.Motor-operated gate valve BN-HV-8812A (BN-HV-8812B
analogous)Fails to close on demand.Failure reduces the redundancy of isolation valves provided to isolate RHRS train "A" from RWST. No effect on safety for system operation. Check valve EJ-8958A in series with motor-operated valve provides the primary isolation against the bypass of RCS coolant flow from the suction of RHR pump "A" to RWST.Valve position indication (open to closed position change) at CB and valve (closed) monitor light and alarm at CB.Valve is a component of the ECCS that performs an RHR function during plant cooldown. Valve is normally "open" to align RHRS for ECCS operation during plant power operation and load follow.ComponentFailureModeEffectonSystemOperation*FailureDetectionMethod**Remarks CALLAWAY - SP Rev. OL-13 5/03TABLE 5.4-10 PRESSURIZER DESIGN DATADesign pressure, psig2,485Design temperature, °F680Surge line nozzle diameter, in.14
Heatup rate of pre ssurizer using heaters only, °F/hr55Internal volume, ft 3 1,800 Normal conditions at 100% rated loadSteam volume, ft 3Water volume, ft 3 720 1,080 CALLAWAY - SP Rev. OL-15 5/06TABLE 5.4-11 REACTOR COOLANT SYSTEM DESIGN PRESSURE SETTINGS PsigHydrostatic test pressure3,107Design pressure2,485 Safety valves (begin to open)2,460 High pressure reactor trip2,385 High pressure alarm2,310 Power-operated relief valves2,335**At 2,335 psig, a pressure signal initiates actuation (opening) of these valves. Remote manual control is also provided.
Pressurizer spray valves (full open)2,310 Pressurizer spray valves (begin to open)2,260 Proportional heaters (begin to operate)2,250Operating pressure2,235
Proportional heater (full operation)2,220Backup heaters on2,210 Low pressure alarm2,210 Low pressure reactor trip1,885 CALLAWAY - SP Rev. OL-13 5/03TABLE 5.4-12 PRESSURIZER QUALITY ASSURANCE PROGRAM RT (a)(a)RT - Radiographic UT - Ultrasonic PT - Dye Penetrant MT - Magnetic Particle UT(a)PT (a)MT (a)HeadsPlatesYesCladdingYes ShellPlatesYesCladdingYes HeatersTubing (b)(b)Or a UT and ETYesYesCentering of elementYesNozzle (Forgings)YesYes (c)(c)MT or PTYes (c)WeldmentsShell, longitudinalYesYes Shell, circumferentialYesYesCladdingYes Nozzle safe endYesYesInstrument connectionYes Support skirt, longitudinal seamYesYesSupport skirt to lower headYesYesTemporary attachments (after removal)Yes All external pressure boundary welds after shop hydrostatic testYes CALLAWAY - SP Rev. OL-13 5/03TABLE 5.4-13 PRESSURIZER RELIEF TANK DESIGN DATADesign pressure, psig100 Normal operating pressure, psig3 Final operating pressure, psig50 Rupture disc release pressure, psig Nominal91 Range86 to 100Normal water volume, ft 3 1,350 Normal gas volume, ft 3 450 Design temperature, °F340 Initial operating water temperature, °F120 Final operating water temperature, °F200 Total rupture disc relief capacity at 100 psig, lb/hr 1.6 x 10 6 Cooling time required fo llowing maximum discharge approximately, hr Spray feed and bleed1 Utilizing RCDT heat exchanger8 CALLAWAY - SP Rev. OL-13 5/03TABLE 5.4-14 RELIEF VALVE DISCHARGE TO THE PRESSURIZER RELIEF TANK Reactor Coolant System 3 Pressurizer safety valvesFigure 5.1-1, Sheet 2 2 Pressurizer power-operated relief valvesFigure 5.1-1, Sheet 2 Residual Heat Removal System 2 Residual heat remova l pump suction line from the reactor coolant system hot legsFigure 5.5-7Chemical and Volume Control System 1 Seal water return line Figure 9.3-8, Sheet 11 Letdown line Figure 9.3-8, Sheet 1 CALLAWAY - SP Rev. OL-13 5/03TABLE 5.4-15 REACTOR COOLANT SYSTEM VALVE DE SIGN PARAMETERSDesign/normal operating pressure, psig2,485/2,235 Preoperational plant hydrotest, psig3,107Design temperature, °F650 CALLAWAY - SP Rev. OL-13 5/03TABLE 5.4-16 REACTOR COOLANT SYSTEM VALVES NONDESTRUCTIVE EXAM INATION PROGRAM RT (a)(a)RT - Radiographic UT - Ultrasonic PT - Dye Penetrant UT (a)PT(a)Boundary Valves, PressurizerRelief and Safety ValvesCastings(larger than 4 inches)YesYes (2 inches to 4 inches)Yes(b)(b)Weld ends onlyYes Forgings(larger than 4 inches)(c)(c)Either RT or UT (c)Yes (2 inches to 4 inches)Yes CALLAWAY - SP Rev. OL-13 5/03TABLE 5.4-17 PRESSURIZER VALVES DESIGN PARAMETERS Pressurizer Safety Valves Number3Minimum relieving capacity at 2,485 psig, ASME rated flow,lb/hr 420,000Set pressure, psig2,460Design temperature, °F650 FluidSaturated steamTransient condition, °F(S uperheated steam) 680 Backpressure Normal, psig3 to 5Expected during discharge, psig500Environmental conditions Ambient temperature (°F)50 to 120Relative humidity (%)0 to 100 Pressurizer Power-Operated Relief Valves Number2Design pressure, psig2,485Design temperature, °F650Relieving capacity at 2,335 psig, per valve, lb/hr 210,000FluidSaturated steamTransient condition, °F(S uperheated steam) 680 CALLAWAY - SP5.4A-1Rev. OL-21 5/15APPENDIX 5.4A SAFE SHUTDOWN5.4A.1INTRODUCTION The SNUPPS powerblock has been designed to enable the plant to be placed in a safe shutdown (hot standby or cold shutdown) cond ition, using only safe ty-related systems. The intent of Regulatory Gu ide 1.139 and BTP RSB 5-1 for achieving cold shutdown are met. Clarifications and specific exceptions to these guides are discussed in Tables 5.4A-1 and 5.4A-2, respectively.Appendix 3B and Section 9.5.1 provide the results of integrated hazards analyses which demonstrate that the SNUPPS units have been designed to withstand postulated events.
Items considered include tornados, floods, missiles, pipe breaks, fires, and seismic events. The single failure criteria utilized in the design are discussed in Section 3.1
.The safe shutdown functions described in this Appendix are controlled and monitored from the control room. For a discussion of safe shutdown using controls and indications entirely outside of the control room, see Section 7.4.3. 5.4A.2SYSTEMS REQUIRED TO GO FROM HOT STANDBY TO COLD SHUTDOWNIn order to safely shutdo wn the plant, the following f unctions must be performed:a.Circulation of the reactor coolantb.Heat removal (short term and long term) c.Borationd.DepressurizationTable 3.11(B)-3 provides a detailed listing of every component that is required to achieve and maintain a safe shutdown. The systems have redundancy/diversity, and no single failure will compromise safety functions.
All power supplies and control functions for required portions of these systems are Class 1E, as described in Chapters 7.0 and 8.0 , except as described for the bo ric acid transfer system (Section 9.3.4
). As discussed in Section 3.2, all components meet the requirements of Regulatory Guides 1.26 and 1.29.
The following are the major syst ems that are employed to achieve and maintain a safe shutdown:a.Reactor Coolant System (See Chapter 5.0
)b.Main Steam System (See Section 10.3
)c.Auxiliary Feedwater System (See Section 10.4.9
)
CALLAWAY - SP5.4A-2Rev. OL-21 5/15d.Chemical and Volume Control System (See Section9.3.4
)e.Borated Refueling Water System (See Section6.3
)f.Residual Heat Removal System (See Section5.4.7
)g.Component Cooling Water System (See Section 9.2.2
)h.Essential Service Water System (See Section9.2.1.2
)i.Supportive HVAC Systems (See Section 9.4
)j.Emergency Diesel Generators (See Sections9.5.4 through 9.5.8)k.Spent Fuel Pool Co oling System (See Section9.1.3
)l.Supportive Portions of Instrument Air System (See Section9.3.1
)Each of the system descripti ons identifies the integral ro le that the system plays in achieving and maintaining a safe shutdown. Instrumentation applications for safe shutdown are described in this section and in Chapter 7.0
.5.4A.3SAFE SHUTDOWN SCENARIO The plant is designed with a number of systems which wil l be used, if available under normal or emergency conditions, to safely shut down the plant. The following shutdown scenario demonstrates that the plant can be taken to a cold shutdown condition, using only safety-related equipment.
Although the use of cert ain nonsafety-related items would be preferable in most situations, this scenario does not take credit for nonsafety-related items because of the assumptions stated in Section 3.1 for the single failure criteria.The safe shutdown licensing basis is hot standby and the safe shutdown design basis is cold shutdown. Should an event occur which would plac e the plant under a Limiting Condition of Operation, or if recovery from the event will cause the plant to be shut down for an extended period of time, the plant may be taken from a hot standby condition to a cold shutdown condition. T he RHR system has a lower design pressure than the RCS.
Therefore, cooldown from hot standby to cold shutdown requires a two-step process.
During the first step, transfer of decay heat and t he stored thermal energy of the reactor coolant system after reactor shutdown, will be via the steam gener ators. During the second step, the RHR system will be ut ilized as a means of heat transfer. 5.4A.3.1MaintainaHotStandbyCondition In the physical layout of t he reactor coolant system, the r eactor core is at a lower elevation with respect to the steam generators; consequently, the higher temperature CALLAWAY - SP5.4A-3Rev. OL-21 5/15 heat source is below the heat sink. This configuration ens ures heat wi ll be transported from the reactor core to th e steam generators vi a the free convecti on flow phenomena.
Imbalance of forces is needed to initiate a convective flow. A thin layer of fluid near the heat transfer surfaces in the reactor core is heated, generating a gradient in temperature and density. When a particle of heated fluid is displaced from near the heat transfer surface, it enters a region of greater average density and is , therefore, subject to a buoyant force. The buoyancy force is opposed by viscous drag and by heat diffusion. Convection begins when buoyancy overcomes the dissipative effect of viscous drag and heat diffusion. As the temperat ure of the fluid in the reactor core is increased relative to the temperature of the fluid in the steam generators, a convective flow will be maintained throughout the reacto r coolant system. The auxiliary feedwater system, in conjunction with the safety-related portion of the main steam system, is initially relied upon to transfer residual core heat from the RCS, via the steam generators, to the atmosphere. This is accomplished by releasing steam from the secondary side of the steam generators, while maintain ing steam generator pressure. Steam is released via the power-operated atmospheric steam dump (ASD) valves. The auxiliary feedwater system is used to maintain a level in the steam generators during this period of time.Water is provided to the auxiliary feedwater pumps (AFP) from the nonsafety-related condensate storage tank (CST); however, this tank is not se ismic Category I and, in the case of an SSE or tornado haz ard, the unprotect ed CST may be unavailable and is not credited for accident mitigation.
In this case, redundant pr essure transmitters in the suction lines of the auxiliary feedwater pumps (AFP) will detect loss of AFP suction pressure and isolate the AFP suction header from the CST. Concurrently, with CST isolation, the essential service water (ESW) pumps are star ted, and the valves in the ESW headers are opened to adm it ESW to the AFPs.The CST isolation valves are in series with check valv es to further preclude the short-circuiting of ESW flow to the nonsafety-related CST. The ESW auxiliary feedwater supply valves are segregated by train relationships to t he motor-driven AFPs with the turbine-driven AFP being fed by both train A and B ESW headers. T herefore, even with a single failure, ESW will be aligned to a minimum of one motor-driven AFP and the turbine-driven AFP.
The motor-operated AFP discharge valves ar e segregated by train relationship to the motor-driven AFPs, each pump feeding two steam generators. The turbine-driven AFP has two air-operated disc harge valves of one train and two of the other, so as to have redundant and opposite train segregat ion to the motor-operated valves associated with the motor-driven AFPs. A safety-related gas supply is provi ded for these valves. In all cases, adequate auxiliary feedwater is supplied to the steam generators for residual heat removal.
CALLAWAY - SP5.4A-4Rev. OL-21 5/15The steam generator ASDs are air-operated valves segregated by train relationship with the steam generators, such that adequate relief capability exists at all times to accomplish residual heat removal. The ASDs are remotely controlled valves, which can either automatically maintain a preset pr essure in the main steam piping or can be manually controlled from eit her the main control board, the auxiliary shutdown panel, or locally for AB-PV-2 and 3. A safety-related gas supply is provided for these valves.In order to maintain an extended hot standby (greater than 24hours), additional negative reactivity must be added to the RCS. Th is is accomplished by borating the RCS while relying on natural circulation in the RCS to ensure adequate mixing of the injected boric acid within the reactor coolant.The design boration condition is based on adding sufficient boric acid to bring the reactor to a xenon-free cold shutdown condition from t he hot full-power peak xenon condition.
The addition of approximately 2 700 gallons of minimum 7000 ppm boric acid is required within 25hours after reactor shutdown from hot full power equilibrium xenon to maintain the reactor in a hot standby condition. This is equivalent to appr oximately 8,500 gallons of water from the refueling water storage tank (RWST) (at 2350 ppm boron). In terms of boron concentration in the RCS, this corresponds to approximately 300 ppm boron, assuming zero boron conc entration in the RCS initially. A total of 13,450 gallons of makeup is required to maintain the RCS in a hot standby condition.Boration may be accomplished by using the boric acid transfer pumps (BATPs) and boric acid tanks or by using the RWST and the EC CS centrifugal charging pumps. At least one BATP will be available under most plant conditions. Ea ch pump is pow ered from a redundant separation grou p of the onsite emergency pow er distribution system. The supply circuit breakers are s hunt tripped only upon the occurrence of an SIS. However, operation of the BATPs cannot be assured following a seismic event or upon occurrence of an SIS. When this is the case, the RWST will be used as the source of borated makeup to the RCS. The BA T system is available for all events following which the RWST is assumed to be unavai lable. Redundant level indication for the BATs and RWST is provided on the MCB.
These level indications are used to determine that sufficient boron concentration has been attained for safe shutdown.If the normal charging path is unavailable, boron will be added thr ough one of two diverse flow paths in the charging system (reactor coolant pump seals or the boron injection path). Each path is capable of delivering a controlled flow of borated water, using jog control switches, from the RWST or BAT system which can be matched to the letdown rate in order to maintain pressurizer level.
The emergency safety-related letdown path diverts cooled letdown (after the excess letdown heat ex changer) to the PRT. In addition, letdown from the RCS may also be accompli shed, utilizing the pressurizer PORVs. These valves are powered by redundant power trains.
CALLAWAY - SP5.4A-5Rev. OL-21 5/15The PRT has a total volume capacity of 13,500 gallons. Prior to initiating letdown through the excess letdown heat exchanger, the 10,000 gallons of relatively clean water in the PRT can be discharged to the containment normal sump at a controlled rate. This will make the PRT available to contain the cooled letdown from the excess letdown heat exchanger and, thereby, mi nimize release of airborne radioactivity to the containment. During the hot standby condition, the reacto r coolant pump seals require cooling by either seal injection or component cooling water. Normally, the operator will ensure that a continuous source of component cooling water is provided. Subsequent seal injection via the ECCS chargi ng pump should only be allowed based on RCS boration/inventory consideration. RCS leakage pa st the seal, with no seal injection, will go to the PRT via the seal return line relief valve, and the loss is considered in the RCS inventory.5.4A.3.2Achieve and Maintain Cold Shutdown With the RCS in a hot standby condition, cold shutdown procedur es may be initiated.
The essential functions which must be conti nued or initiated to achieve cold shutdown are:a.Continued residual heat removal vi a the steam gener ators, utilizing auxiliary feedwater and the atmo spheric steam dump valves.b.Letdown and boration to cold shutdown boric acid concentrations.c.Continued circulation of the coolant in the RCS.d.RCS depres surization.e.Initiation of the re sidual heat removal (RHR) system when the RCS temperature and pressu re are reduced below approximately 350°F and 400psig.Following boration to the cold shutdown concentration, cool down is accomplished by increasing the steam dump from the steam generator ASDs to attain a primary side cooling of approximately 50°F/hr. In conjuncti on with this portion of the cooldown, the ECCS charging pumps are used to deliver refueling water to makeup for primary contraction due to cooling. Letdown and boration to achieve cold shutdown boric acid concentrations are identical to procedures described above for the hot standby conditi on. The completion of this step requires that the RCS boro n concentration be increas ed to approximately 1400 ppm boron at the beginning of an oper ating fuel cycle and to appro ximately 1000 ppm boron at the end of the cycle. These concentrations range from approximately 200 to 700 ppm higher than the boron concent rations at hot full power equilibrium xenon and about 500 to 1000 ppm highe r than concentrations at hot full power peak xenon. The specific CALLAWAY - SP5.4A-6Rev. OL-21 5/15required cold shutdown concen tration for any time during any fuel cycle and for the actual xenon condition may be calculated by the plant operator using a written procedure.
Boration of the RCS to the cold shutdown concentration is required by procedure to be accomplished prior to any si gnificant cooldown of the RCS. In order to maintain pressurizer level within the defined operating band, boration is a combined charging and letdown process. Typical volumes of wa ter to be charged and letdown range from 33,500 to 40,000 gallons when the RWST is th e source of borated water, depending on the fuel cycle burnup.The RCS cold shutdown concentra tion is ensured by process c ontrol, i.e., knowledge of initial RCS boron concentrations and knowledge of amounts and concentrations of injected fluid (either RWST or BAT fluids) ensures that the cold shutdown concentration is obtained.
Continued circulation of RCS is accomplished by natural circulati on resulting from heat removal via the steam generators.
Depressurization of the RCS is achieved through the use of the power-operated pressurizer relief valves. As previously stated, each valve has an independent safety-related powe r actuation train.
Prior to reducing RCS pressure below 1000 psig, it is nec essary to ensure that all accumulator tank isolation valves are in the closed position to avoid their discharge to the RCS. For two of the four valves (those powered by the assumed oper ational diesel generator), this is accomplished from the associated MCC. If power is not available for the remaining two valves, the operator can vent cover gas from the affected accumulator.
When the RCS has been cooled and depressurized below appr oximately 350°F and 400 psig, the residual heat removal (RHR) system is put in service. This is done by establishing component cooling water flow through the RHR heat exchanger by opening the associated motor-operated valve and by closing the moto r-operated isolation valves to the RCS cold legs and to the RHR pump suction from the RWST.
The next operation requires th at the RCS/RHR isolation va lves be opened. With a loss of offsite power in conjunction with the fa ilure of one diesel generator, one of the two isolation valves in each RHR suction line cannot be opened from the main control board.
In order to initiate system o peration, the motor control cent er (MCC) of t he failed diesel generator train must be energized by providing a temporary ca ble intertie from the MCC located in the opposite elec trical penetration room powered by the operational diesel generator.After opening the RCS/RHR isolation valves, the RHR pump is manually started to circulate flow through the mini flow line. Either of the miniflow bypass valves will CALLAWAY - SP5.4A-7Rev. OL-21 5/15automatically open to maintain minimum flow based on the signal received from the flow indicating switch in the outlet piping of the pump.At this point, it is possible to obtain a manual sample of the RHR loop fluid to ensure that the boron concentration is gr eater than or equal to the required cold shutdown RCS concentration. This sample can be obtained at any of several drain and vent connections, the local sample connection, or via t he direct connecti on to the nuclear sampling system, if available.
During this period of time, the operator has determined the status of the RHR system controls and is prepar ed to put the RHR system into operation. The next step is to establish flow from the RCS hot leg to the RCS cold legs via the RHR pump and heat exchanger. To do this, the RHR pump is stopped while the RHR/RCS cold leg re turn valve is opened.The RHR pump is restarted to initiate the final co oldown phase. At th is point, since the air-operated flow control valv es may not be functi onal, administrative control is required to avoid excessive heat loads (and resulting excessive duty) on the com ponent cooling water system. Two methods of achieving this control are: 1) only one RHR pump may be operated or two RHR pumps can be started/stopped, over an extended period of time, to limit the total heat load on the RHR heat e xchangers, or 2) throttlin g of the CCW flow to the RHR heat exchanger can be accomplished. This will result in less flow, though at a higher temperature, back to the CCW heat exchangers.
Continued operation in this m ode will decrease the RCS temperature to cold shutdown conditions.The capability of the RHRS M to accommodate a single component failur e and still perform a safety grade co oldown is demonstrated in the failure mode and effects analysis of the RHRS for safety-related cold shutdown operations provided as Table 5.4A-3.
CALLAWAY - SP Rev. OL-21 5/15TABLE 5.4A-1 DESIGN COMPARISON TO REGULATORY POSITIONS OF REGULATORY GUIDE 1.139 REV 1, DRAFT 2 DATED FEBRUARY 25, 1980 TITLED "GUIDANCE FOR RESIDUAL HEAT REMOVAL TO ACHIEVE AND MAINTAIN COLD SHUTDOWN:A complete discussion of the SNUPPS plant cold shutdown capability is provided in Appendix5.4A
.REGULATORYPOSITIONUNIONELECTRIC1.FUNCTIONALThe methodutilized to take the reactor from normal operating conditions to cold shutdown should satisfy the functional guidance presented below.a.The design should be such that the reactor can be taken from normal operating conditions to cold shutdown using only safety-related systems that satisfy General Design Criteria1 through 5.1a.The reactor coolant system, in conjunction with several supporting systems, can be brought to a cold shutdown condition following any given hazard (GDCs 2, 3, and 4) using safety-related systems (design in compliance with GDC 1).b.These safety-related systems should have suitable redundancy in components and features and suitable interconnection, leak detection and containment, and isolation capabilities to ensure that, for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available), the system safety function can be accomplished assuming a single failure.1b.Complies.
Section 3.1.2 provides the single failure criteria thatis used, including the bases for operator action outside the control room. Table 5.4A-3 provides a safety related cold shutdown (CSD) FMEA.In demonstrating that the method can be utilizedto perform its function assuming a single failure, limited operator action outside the control room would be acceptable if suitably justified. Necessary operatoractions to maintain hot shutdown or proceed from that plant conditiontocoldshutdownshouldbe planned nosooner thanonehourfromthetimewhenshutdown is commenced. This limitedoperator action should notresult in an exposure beyond the allowed limitsassuming highradio- activity inthereactorcoolant or containment building environment.c.The method should be capable of bringing the reactor to a hot shutdown condition, whereRHRcoolingmaybeinitiated , within approximately 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> following shutdown with only offsite power or onsite power available, assuming the most limiting single failure.1c.Complies. See Section 5.4.7.2.1
.d.Instrumentationandcontrolsincludingprotectivemeasuresandinterlocksassociatdwiththesafety-relatedsystemsrequiredto achieveormaintain cold shutdown should meettherequirementsof IEEE Standards 279-1971,323,384, and 344 and theguidanceprovidedinRegulatoryGuides1.89,1.75,and1.100.1d.Except for the boric acid transfer system controls and the pressurizer heaters, the instrumentation and controls are designed in accordance with applicable Regulatory Guides and IEEE standards. The highly reliable design of the pressurizer heaters and the BAT system (both of which are capable of being manually loaded on the diesels) are described in Sections 5.4 , 7.4 , 8.3, and 9.3.4.e.Thesafety-relatedsystemsshouldbeclassifiedasSeismicCategoryI and meet the guidanceprovidedinRegulatoryGuide1.29.1e. Except as discussed in 1d, all components and systems comply.
CALLAWAY - SPTABLE 5.4A-1 (Sheet 2)
Rev. OL-21 5/15 2.REACTIVITYCONTROL2.Complies.Asafety-relatedsystemshouldmeetGDC1-5,26,and27andbecapableof controlling andmonitoringboronconcentrationinordertoensurereactor subcriticality fromoperatingconditionsthroughcoldshutdown.
3.HEATREMOVALTOREDUCETHERCSFROMPLANTOPERATING CONDITIONS TORHRSYSTEMOPERATINGCONDITIONS a.PWRPlants (1)AuxiliaryFeedwaterSystem3.a.(1)The auxiliary feedwater system complies with these requirements,as discussed in 10.4.9. The essential service water system (Section 9.2.1.2) which provides the ultimate water supply has adequate inventory to supply short-term and long-term requirements.Safety-related (Class 1E) indication of the AFW flow to each generator is provided in the control room. Safety-related steam generator level indication provides a backup means of determining the AFW flow.Asafety-relatedauxiliaryfeedwatersystemshouldbedesigned and constructed to provideareliablesourceofcoolingwater atPWR plants in accordancewithGDC1-5,44,45,and46. The safety-related water supply for the auxiliary feedwater system for a PWR should have sufficient inventory to permit operation at hot-standby conditions for at least 4hours followed by cooldown to the conditions permitting operation of the RHR system. The inventory needed for cooldown should be based on the longest cooldown time needed with either only onsite or only offsite power available with an assumed single failure. Thecapability should exist for providing cooling water from the ultimate heat sinkpriorto exhaustion of the safety-related watersupply. Automatic initiationshouldbe provided for the auxiliary feedwater system. Theautomatic initiationsignals and circuits shouldbe safety-related and bedesignedsothat a single failure will notresult in the lossofAFWSfunction.Testabilityoftheinitiating signals andcircuits should beafeatureofthedesign.Manual initiation capability fromthe control roomshouldbe safety-relatedandbe designed so thata single failure will notresultinthelossofsystem function. Thea-c motor-drivenpumpsandvalvesin theAFWS should be included in the automatic actuation(simultaneousand/orsequential)of theloads to the emergency buses.Theautomaticinitiating signalsandcircuits should be designed sothattheirfailure willnotresult in the loss of manual capability to initiate theAFWSfromthecontrolroom.A safety-related redundant systemshouldbeprovidedforindicationinthecontrolroom of auxiliary feedwaterflowtoeachsteamgenerator.
(1)SteamRelief3.a.(2)Complies, as discussed in Section 10.3
.Asafety-relatedredundantatmosphericsecondarysidesteam relief system shouldbedesignedtoprovideforreductionof the RCS temperature toRHRsystemoperatingconditions.
(2)SteamGeneratorInventory3.a.(3)Complies, as discussed in Section 10.4.7
.Eachsteamgeneratorshouldbeequippedwithasafety-related redundant water levelindicationandalarmsystem.REGULATORYPOSITIONUNIONELECTRIC CALLAWAY - SPTABLE 5.4A-1 (Sheet 3)
Rev. OL-21 5/15 b.BWRPlants3.b.Not applicable to SNUPPS.
(1)SteamReliefAsafety-relatedredundantsteamreliefsystemshouldbe designed to allow forcontrolledreactorcoolantsystem depressurization by steam relief tothepressure suppressionpool.
(2)ReactorVesselInventoryThereactorvesselshouldbeequippedwithasafety-related redundant water level indication andalarmsystem.
4.RESIDUALHEATREMOVAL4.The RHR system meets the applicable GDCs, as described in Section 5.4.7
.TheRHRsystemshouldmeetGDC1-5and34withatleasttworedundant trains of pumps andheatexchangers.Beginning4hoursafterreactor shutdown, each train should havesufficientheatremovalcapability (a)for maintaining the RCS at hot shutdown (RHRsysteminitialoperating conditions)atthattimeincorelifewhen the greatest amountofdecay andresidualheatispresent,and (b)to provide for cooldown oftheRCS fromhotshutdowntocoldshutdownconditions.a.RHR System Isolation(1)Isolation of the suction side of each RHRsystem trainfrom directRCS pressure shouldbe provided by at least two power-operated valves in series, with valve position indicated in the control room. Alarms in the control room should be provided to alert theoperator if either valve is open when the RCSpressure exceeds the RHR system design pressure. The isolationvalvesystem should havetwoormore independent interlocks to prevent the valves from being opened unless the RCS pressure is below the RHR system design pressure. Upon loss of actuating power,isolationvalvesshouldnotchange positionunless movement istoapositionthatprovidesgreater safety.The isolation valve systemshouldhavetwoormore independent protective measures to close any open valve in the eventof an increase in the RCS pressure above theRHR system design pressure. All isolation valves onthe discharge and suctionsidesoftheRHR systemshouldbe classified ASME OM Code Subsection ISTC,CategoryA, and be leak tested at each refueling outage.4.a.1Complies, except that automatic closure in the event of an increase in RCS pressure is not provided. Instead, a control room alarm will alert the operator if a valve is open when RCS pressure exceeds RHR system design pressure. Operating procedures will verify that isolation valves are closed prior to increasing RCS pressure above RHR system design pressures. See Section 5.4.7.(2)One of the following should be provided on the discharge side of the RHR system toisolate it from the RCS:4.a.2Complies. Meets Paragraph C.(a)The valves, position indicators, alarms, and interlocks described in item (1).REGULATORYPOSITIONUNIONELECTRIC CALLAWAY - SPTABLE 5.4A-1 (Sheet 4)
Rev. OL-21 5/15(b)One or more check valves in series with a normally closed power-operated valve. Thepositionofthepower-operated valve should beindicatedinthecontrolroom. If the RHR system discharge line is used for an ECCS function, the power-operated valve should be opened upon receipt of a safety-injection signal once the reactor coolant pressure has decreased below the ECCS design pressure.(c)Two check valves in series.b.RHR System Pressure Relief4.b.Complies, as described in Sections 5.2.2.10 and 5.4.7.2.5.To protect the RHR system against accidental overpressurization when it is in operation (not isolated from the RCS), pressure relief in the RHR system should be provided with relieving capacity in accordance with the ASME Boiler and Pressure Vessel Code. The most limiting pressure transient during the plant operating condition when the RHR system is not isolated from the RCS should be considered when selecting the pressure relieving capacity of the RHR system. For example, during shutdown cooling in a PWR with no steam bubble in the pressurizer, inadvertent operation of an additional charging pump inthenormal charging modeorahighheadECCSpump(for thoseplantsatwhichthehigh head pumps serveadualfunction) should be considered in selecting the design bases.Fluid discharge through the RHR system pressure relief valves should be collected and contained so that a relief valve that is stuck in the open position will not:(1)Result in flooding of any safety-related equipment.
(2)Reduce the capability of the ECCS below that needed to mitigate the consequences of a postulated LOCA.(3)Result in a non-isolatable situation in which the water provided to the RCS to maintain the core in a safe condition is discharged outside the containment.If interlocks are provided to automatically close the isolation valves when the RCS pressure exceeds the RHR design pressure, relief capacity should be provided during the time that the valves are closing suchastoprevent the RHR design pressure frombeingexceeded.c.RHR System Pump Protection4.c.Complies. See Section 5.4.7
.The design and operating procedures of the RHR system andplant operating procedures should be such that no single failure or single operator error can result in loss of the RHR function duetodamageoftheRHR system pumps including overheating, cavitation,orlossofadequatepumpsuctionhead.REGULATORYPOSITIONUNIONELECTRIC CALLAWAY - SPTABLE 5.4A-1 (Sheet 5)
Rev. OL-21 5/15d.RHR System Testing4.d.Complies. See Chapters 7.0 and 8.0 for IEEE testing. See the responses to Regulatory Guides 1.22 and 1.68 for testing.For the RHR system, the isolation valve operability and interlock circuits should be designed to permit on-line testing when operating in the RHR mode. System testing should meet the requirements of IEEE Standard 338 and the guidance of Regulatory Guide 1.118.The preoperational and initial startup test program should be in conformance with Regulatory Guide 1.68. Inaddition, the programs for pressurized water reactors should include tests with supporting analysis to confirm (a) that adequate mixing of borated water added tothereactorcoolantsystem prior to or during cooldown can be achieved under natural circulation conditions and permit estimation of the times required to achieve such mixing and (b) that the cooldown under natural circulation conditions can be achieved within the guidelines specified in the emergency operating procedures.TheRHRsystemshouldbedesignedtopermiton-linepressureand functional testing toassure (1)thestructuralandleaktightintegrityofitscomponents, (2)theoperabilityandperformanceoftheactivecomponentsof the system, and (3)theoperabilityofthesystemasawholeand,underconditions asclose to design aspractical,thetransferbetweennormal andemergency power sources, and theoperationofthe associatedcoolingwatersystem.
e.RHRSystemOperationalIndication4.e.Complies.Indicationofisolationvalveposition,systempressureandflow, and pump operating statusshouldbeavailableinthecontrolroom.
f.RHRSystemIntegrity4.f.Complies.TheRHRsystemshouldbedesignedandconstructedtohavethe capability to remove heatfromthereactorcoolantduringnormal andfollowingaccident conditions. Sincethereactorcoolantmay behighlyradioactive following accident conditions, theRHRsystem integrityshouldbesuchthat radioactivityis not released to theenvironmentbeyond acceptedlimits. The design should include features to preventunacceptable degradationof long-term heat removal capability and leakage resultingfromadegradedcore condition or the containment post-accident environment.In addition,the system should be designed sothat the operator canassessthe status, isolate, maintain and repairtheRHR system, as needed. Specifically, the RHR system integrity should meet the following criteria:REGULATORYPOSITIONUNIONELECTRIC CALLAWAY - SPTABLE 5.4A-1 (Sheet 6)
Rev. OL-21 5/15 (1)Leakagefromthesystemsuchasfromvalvesandpumpseals should be monitored andcontrolled.Theleakagelimitsat whichanRHRtrain is to be declaredinoperableandisolated shouldbestatedinthe Plant Technical Specifications. Indicationoftheamount ofleakage, suchas sump level indication,radiationlevelsandsystemisolationshouldbe available locallyandinthecontrolroom.Valvelineupand isolation capability shouldbesuchastoprecludethe possibilitythathighly radioactive sumpwatercanbe automatically transferred to the radwaste processing system.4.f.1Complies, leakage detection is discussed in Section 9.3.3
.(2)Shieldingshouldbeprovidedtomaintainpersonnelexposure as low as is reasonably achievable(ALARA).Shielding protectionshouldalso be provided forinstruments,components, orotheritemswhich mightbe adversely affectedbyhigh radiationfields.Provisionsshould be made for access to, andminorrepairof,equipmentoutsidecontainment which may fail during apost-incidentrecoveryperiod.4.f.2Complies, except that area temperature monitoring is not provided for the SNUPPS project. High temperature alarm is provided in the MCR for the RHR pump rooms. Compliance with ALARA requirements are discussed in
Section 12.3.1
.Provisionsshouldbemadefortie-inofadditionalequipment or systems in theeventthatmajorrepairisnecessary. Areatemperaturemonitoring and controlshouldbeprovided fortheRHRsystemenvironmentwith indication andcontrol inthecontrolroom.
(3)TheRHRsystemincludingtheleakagecollectionsumpshould be located in a closedareawhichisequippedwithan engineeredsafetyfeature filtration system(asgivenin RegulatoryGuide1.52)andradiation monitors. Theseareas shouldbemaintainedatasufficient negative pressure (typically,atleast-1/8inch,watergauge)withrespect tothe ambient atmospheretopreventexfiltrationof activitywhich could bypass the ESFfiltersystem.4.f.3The emergency exhaust filtration system which serves this function following an LOCA is discussed in Sections 9.4.2 and 9.4.3.g.RHRCoolingWaterSupplySystem4.gComplies, except that the radiation monitors are not located on the outlet of the RHR heat exchanger. Instead, each train of component cooling water is provided with radiation monitors within the system. See Section 9.2.2
.Thesafety-relatedsystemshouldbedesignedandconstructedwith atleasttwo independent subsystemsortrainssuchthateachhas thecapacitytoadequately remove heatfromthereactorcoolant inaccordancewithGDC1,2,3, 4, 5, 44, 45 and 46. Cooling waterradioactivityshouldbemonitoredattheoutputoftheRHR heat exchangerswithindicationandanalarminthecontrolroom.REGULATORYPOSITIONUNIONELECTRIC CALLAWAY - SPTABLE 5.4A-1 (Sheet 7)
Rev. OL-21 5/15 5.NATURALCIRCULATIONCOOLINGFORPWRPLANTS5.Complies. See response to Regulatory Guides 1.22 and 1.68. The natural circulation test was performed at Diablo Canyon and verified by a partial test at Callaway. The unit is provided with two groups of backup pressurizer heaters. The heater groups and their associated controls are powered from a diesel-backed bus through qualified isolation devices that shed their load only upon an SIS or emergency bus undervoltage signal. If desired, these devices can be manually reclosed from the control room, following reset of the initiating trip signals. The emergency diesel generators are sized in excess of that required to carry all connected pressurizer heaters concurrent with the loads required for a LOCA. They are provided with a full complement of status indication in the control room.The pressurizer is provided with two Class 1E power-operated relief valves (PORV) and two Class 1E power-operated relief valve isolation valves (PORVIV). These valves are powered from the onsite emergency power supply, with redundant Class 1E power supplying the two valves associated with each flow path.Three loops of the pressurizer level instrumentation are powered from Class 1E power supplies. In addition, a fourth nonsafety grade instrumentation loop is provided.Toensurethecapabilitytoachieveandmaintainnaturalcirculation within the primary system, redundantemergencypower, whichmeets GeneralDesign Criteria17 and 18, should beprovidedtoeachofthefollowing:
a.Theminimumnumberofpressurizerheatersrequiredtomaintain natural circulation conditions.
b.Thecontrolandmotivepowersystemsforthepower-operatedrelief valvesand associated blockvalves,and c.Thepressurizerlevelindicationinstrumentchannels.6.REACTORCOOLANTSYSTEMINVENTORY a.PWRPlants6.aComplies. The chemical and volume control system is described in Section 9.3.4.Asafety-relatedsystemshouldbedesignedand constructedto meet GDC 1-5 and 33andcapable ofprovidingreactorcoolantmakeupandletdown control with a sufficientwatersupplytoaccount forcooldownshrinkage, required letdown for boration, andtechnicalspecificationallowedleakage from operating conditions to cold shutdown.
b.BWRPlants6.bNot applicable to SNUPPS.Asafety-relatedsystem(orsystems)shouldmeet GDC1-5and33 andbe designed andconstructedto provideareliablesourceofmakeupwatertothe reactor coolantinventory.Asafety-relatedwater supplyshouldhave sufficient inventory topermit maintenanceofplantoperatingconditionsforat least4hours followed bycooldowntoRHRsystem operatingconditions.REGULATORYPOSITIONUNIONELECTRIC CALLAWAY - SPTABLE 5.4A-1 (Sheet 8)
Rev. OL-21 5/15 7.OPERATIONALPROCEDURES7.Complies.The operational procedures for bringing the plant from normal operating power to cold shutdown should be in conformance with Regulatory Guide 1.33. For pressurized water reactors, the operational procedures should include specific procedures and information required for cooldown under natural circulation conditions. In addition, plant proceduresforallactivities should provide instruction in suchamannerthatwill notleadtoalossoftheRHRsystem.Emergencyproceduresshouldaddresscooldownduring orafteranaccident, including naturalcirculation cooldowninthecaseofPWRplants.Theseemergency procedures shouldincludeguidanceonsafeshutdown tocoldconditions in the event of failureofnon-safety-relatedequipmentandsinglefailuresofsafety-related equipment. Othercaseswhichtheemergency procedures should address are RHR heat exchangertube leak,highradioactivityinthereactorcoolant,and high airborne radioactivity intheRHRsystemroom.Emergencyproceduresshouldbepreparedtoaddressthe transferofthe pressurizer heaters totheemergency powersourceintheeventthatthisactionisnecessary. The method andtimerequiredtoaccomplishthetransfer ofthe preselected pressurizer heaterstotheemergency busesshould be described in written approved procedures andbeconsistentwiththetimelyinitiationand maintenance of natural circulation.REGULATORYPOSITIONUNIONELECTRIC CALLAWAY - SP Rev. OL-19 5/12TABLE 5.4A-2 DESIGN COMPARISON OF TABLE 1 OF BTP RSB 5-1 FOR POSSIBLE SOLUTIONS FOR FULL COMPLIANCEDesign Requirements of BTP RSB 5-1 Process and (System or ComponentPossible Solution forFull Compliance Union ElectricI.Functional Requirement for Taking to Cold Shutdowna.Capability using only safety grade systemsb.Capability with either only onsite or only offsite power and with
single failure (limited action outside CR to meet SF)c.Reasonable time for cooldown, assuming most limiting SF and only offsite or only onsite powerLong-term cooling (RHR drop line)
Provide double drop line (or valves in parallel) to prevent single valve failure from
stopping RHR cooling function. (Note: This requirement in conjunction with meeting effects
of single failure for long-term cooling and isolation requirements involve increased number of independent power supplies and possibly more than four valves.)
Series power-operated valves are provided in both RHR/RCSshutdown lines. Design can withstand a single failure, as discussed in Section 5.4.7.
CALLAWAY - SPTABLE 5.4A-2 (Sheet 2)
Rev. OL-19 5/12 Heat removal and RCS circulat ion during cooldown to cold shutdown. (Note: Need SG cooling to maintain RCS circulation even after RHRS in operation when under
natural circulation) (steam dump valves.)
Provide safety-grade dump valves, operators, and power supply, etc. so that manual action should not be required after SSE, except to meet
single failure.
Complies.Depressurization (Pressurizer auxiliary
spray or power-operated relief valves)
Provide upgrading and additional valves to ensure operation of auxiliary pressurizer spray, using only safety-grade subsystem meeting single failure. Possible alternative may involve using pressurizer power-operated
relief valves which have been upgraded. Meet SSE and single failure without manual operation inside containment.
Complies. Fully qualified Class 1E pressurizer
power-operated relief valves are provided Design Requirements of BTP RSB 5-1 Process and (System or ComponentPossible Solution forFull Compliance Union Electric CALLAWAY - SPTABLE 5.4A-2 (Sheet 3)
Rev. OL-19 5/12 Boration for cold shutdown (CVCS and
boron sampling)
Provide proc edure and upgrading where necessary, such that boration to cold shutdown concentration meets the requirements of I. Solution could range from (1) upgrading and adding valves to have both letdown and charging paths safety grade and meet single failure to (2) use of backup procedures involving less cost.
For example, boration without letdown may be acceptable and eliminate need for upgrading letdown path. Use of ECCS for injection of borated water may also be acceptable. Need
surveillance of boron concentration (boronometer and/or sampling). Limited operator action inside or outside of containment if justified.
The excore detector will alert the operator of any criticality potential. Charging to and letdown from the RCS are controlled quantities. No boron sampling is required.II.RHR IsolationRHR SystemCo mply with one of allowable arrangements given.II.Complies. See Section 5.4.7
.III.RHR Pressure ReliefDesign Requirements of BTP RSB 5-1 Process and (System or ComponentPossible Solution forFull Compliance Union Electric CALLAWAY - SPTABLE 5.4A-2 (Sheet 4)
Rev. OL-19 5/12b.Collect and contain relief dischargeRHR SystemDetermine piping, etc., needed to meet requirement and
provide in design.III.Complies. See Section 5.4.7
.V.Test Requirementb.Meet R.G. 1.68. For PWRs, test plus analysis for cooldown
under natural circulation to confirm adequate mixing and cooldown within limits specified in EOP.Run tests and confirming analysis to meet requirement.V.Complies. See R.G. 1.68 response. The
natural circulation test was performed at Diablo Canyon and verified by a partial test at Callaway.VI.Operational Procedurea.Meet R.G. 1.33. For PWRs, include specific procedures and information for
cooldown under natural circulation.
Develop procedures and information from test and analysis.VI.Complies. VII.Auxiliary Feedwater Supply Emergency Feedwater SupplyDesign Requirements of BTP RSB 5-1 Process and (System or ComponentPossible Solution forFull Compliance Union Electric CALLAWAY - SPTABLE 5.4A-2 (Sheet 5)
Rev. OL-19 5/12a.Seismic Category I supply for auxiliary FW for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at hot shutdown plus cooldown to RHR
cut-in based on longest time for only onsite or only offsite
power and assumed single failure.From tests and analysis obtain conservative estimate of
auxiliary FW supply to meet requirement and provide seismic Category I supply.VII.Preoperational and startup test will establish the amount of makeup water required.
Essential service water is the seismic Category I supply.Hot shutdown is within the RHR cut-in temperature; therefore, there is no ensuing
cooldown required.Design Requirements of BTP RSB 5-1 Process and (System or ComponentPossible Solution forFull Compliance Union Electric CALLAWAY - SP Rev. OL-21 5/15TABLE 5.4A-3 RESIDUALHEATREMOVAL-SAFETY RELATED COLD SHUTDOWN OPERATIONS-FAILUREMODESANDEFFECTSANALYSIS(FMEA)ComponentFailureModeFunctionEffectonSystemOperation*FailureDetection Method**Remarks1.Motor-operated gate valve EJ-HV-8701A (EJ-HV-8701B analogous)Fails to open on demand.Provides isolation of fluid flow from the RCS to RHR pump 1 (pump 2).Failure blocks RC flow from hot leg of RC loop 1 through train "A" of RHRS. Fault reduces redundancy of RHR coolant trains provided. No effect on safety for system operation.
Plant cooldown requirements will be met by RC flow from hot leg of RC loop 4 flowing through train "B" of RHRS.
However, time required to reduce RCS temperature will be extended.Valve position indication (closed to open position change) at CB; RHR train "A" discharge flow indication (EJ-FI-618) and low flow alarm at CB; and RHR pump dicharge pressure indication (EJ-PI-614) at CB.1.Valve is electrically interlocked with the containment sump isolation valve EJ-HV-8811A and the RWST isolation valve BN-HV-8812A, with RHR to charging pump suction line isolation valve EJ-HV-8804A and with a "prevent-open" pressure interlock (BB-PB-405A) off the seal table. The valve cannot be opened remotely from the CB if one of the indicated isolation valves is open or if RC loop pressure exceeds 360 psig. The valve can be manually opened. See Section 5.4.7.2.7
.2.If both trains of RHRS are unavailable for plant cooldown due to multiple component failures, the auxiliary feedwater system and SG power-operated relief valves can be used to perform the safety function of removing residual heat. 2.Motor-operated gate valve BB-PV-8702A (BB-PV-8702B analogous)Same failure modes as those stated for item 1.Same function as that stated for item
1.Same effect on system operation as that stated for item
1.Same methods of detection as those stated for item 1.Same remarks as those stated for item 1., except for pressure interlock (BB-PB-403A) control. *See list at end of table for definition of acronyms and abbreviations used. **As part of plant operation, periodic tests, surveillance inspections, and instrument calibrations are made to monitor equipment and performance. Failures may be detected during such monitoring of equipment, in addition to detection methods noted.
CALLAWAY - SPTABLE 5.4A-3 (Sheet 2)
Rev. OL-21 5/153.RHR pump 1 (RHR pump 2 analogous)Fails to deliver working fluid.Provides fluid flow of RC through RHR heat exchanger 1 (heat exchanger 2) to reduce RCS temperature during cooldown operation.Failure results in loss of RC flow from hot leg of RC loop 1 through train "A" of RHRS.
Fault reduces redundancy of RHR coolant trains provided. No effect on safety for system operation. Plant cooldown requirements will be met by RC flow from hot leg of RC loop 4 flowing through train "B" or RHRS. However, time required to reduce RCS temperature will be extended.
Open pump switchgear circuit breaker indication at CB; circuit breaker close position monitor light for group monitoring of components at CB; common breaker trip alarm at CB; RHR train "A" discharge flow indication (EJ-FI-618) and low flow alarm at CB; and pump discharge pressure indication (EJ-PI-614) at
CB.The RHRS shares components with the ECCS. Pumps are tested as part of the ECCS testing program (see Section 6.3.4). Pump failure may also be detected during ECCS testing.4.Motor-operated gate valve EJ-FCV-610 (EJ-FCV-611 analogous)a.Fails to open on demand.Provides regulation of fluid flow through miniflow bypass line to suction of RHR pump 1 (pump 2) to protect against overheating of the pump and loss of discharge flow from the
pump.Failure blocks miniflow line to suction of RHR pump "A". RHR train "A" is degraded for the protection of RHR pump A. No effect on safety for system operation. Plant cooldown requirements will be met by operator action of controlling cooldown with redundant RHR train "B". However, time required to reduce RCS temperature will be extended.Valve position indication (closed to open position change) at CB.Valve is automatically controlled to open when pump discharge is less than 816 gpm at 300°F (783 gpm at 68°F) and close when the discharge exceeds 1,650 gpm at 300°F (1582 gpm at 68°F). CB switch set to "Auto" position for automatic control of valve positioning.ComponentFailureModeFunctionEffectonSystemOperation*FailureDetection Method**Remarks CALLAWAY - SPTABLE 5.4A-3 (Sheet 3)
Rev. OL-21 5/15b.Fails to close on demand.
Same function as that stated for item 4.a.Failure allows for a portion of RHR heat exchanger "A" discharge flow to be bypassed to suction of RHR pump "A". RHRS train "A" is degraded for the regulation of coolant temperature by RHR heat exchanger "A." No effect on safety for system operation.
Cooldown of RCS within established specification cooldown rate may be accomplished through operator action of throttling flow control valve EJ-HCV-606 and controlling cooldown with redundant RHRS train "B." Valve position indication (open to closed position change) and RHRS train "A" discharge flow indication (EJ-FI-618) at CB. 5.Air diaphragm-operated butterfly valve EJ-FCV-618 (EJ-FCV-619
analogous)a.Fails to open on demand.Controls rate of fluid flow bypassed around RHR heat exchanger 1 (heat exchanger 2) during cooldown operation.Failure prevents coolant discharged from RHR pump "A" from bypassing RHR heat exchanger "A" resulting in mixed mean temperature of coolant flow to RCS being low. RHRS train "A" is degraded for the regulation of controlling temperature of coolant. No effect on safety for system operation. Cooldown of RCS within established specification rate may be accomplished through operator action of throttling flow control valve EJ-HCV-606 and controlling cooldown with redundant RHRS train "B."RHR pump "A" discharge flow temperature and RHRS train "A" discharge to RCS cold leg flow temperature recording (EJ-TR-612) at CB; and RHRS train "A" discharge to RCS cold leg flow indication (EJ-FI-618) at CB. 1. Valve is designed to fail "closed" and is electrically wired so that electrical solenoid of the air diaphragm operator is energized to open the valve. Valve is normally "closed" to align RHRS for ECCS operation during plant power operation and load follow.2.Valve operation is not required for safety grade cold shutdown operations.ComponentFailureModeFunctionEffectonSystemOperation*FailureDetection Method**Remarks CALLAWAY - SPTABLE 5.4A-3 (Sheet 4)
Rev. OL-21 5/15b.Fails to close on demand.
Same function as that stated for item 5.a.Failure allows coolant discharged from RHR pump "A" to bypass RHR heat exchanger "A", resulting in mixed mean temperature of coolant flow to RCS being high. RHRS train "A" is degraded for the regulation of controlling temperature of coolant. No effect on safety for system operation. Cooldown of RCS within established specification rate may be accomplished through operator action of throttling flow control valve EJ-HCV-606 and controlling cooldown with redundant RHRS train "B." However, cooldown time will be extended. Same methods of detection as those stated for item 5.a.6.Air diaphragm-operated butterfly valve EJ-HCV-606 (EJ-HCV-607
analogous)a.Fails to close on demand for flow reduction.Controls rate of fluid flow through RHR heat exchanger 1 (heat exchanger 2) during cooldown operation.Failure prevents control of coolant discharge flow from RHR heat exchanger "A", resulting in loss of mixed mean temperature coolant flow adjustment to RCS. No effect on safety for system operation.
Cooldown of RCS within established specification rate may be accomplished by operator action of controlling cooldown with redundant RHRS train "B."Same methods of detection as those stated for item 5.a. In addition, monitor light and alarm (valve closed) for group monitoring of components at CB. 1.Valve is designed to fail "open". Valve is normally "open" to align RHRS for ECCS operation during plant power operation and load follow.2.Valve operation is not required for safety grade cold shutdown operations.b.Fails to open on demand for increased flow.
Same function as that stated for item
6.a.Same effect on system operation as that stated for item 6.a. Same methods of detection as those stated for item 6.a. 7.No entryComponentFailureModeFunctionEffectonSystemOperation*FailureDetection Method**Remarks CALLAWAY - SPTABLE 5.4A-3 (Sheet 5)
Rev. OL-21 5/158.No entry9.Motor-operated gate valve BN-HV-8812A (BN-HV-8812B analogous)Fails to close on demand.RWST to RHR suction isolationFailure prevents isolation of RWST from RHR pump 1 (pump 2). Negligible effect on safety for system operation.
Alternate RHR train is available by isolating RWST to RHR pump 2 (pump 1) via isolation valve BN-HV-8812B (BN-HV-8812A). Only effect is an increase in time required to reduce RCS temperature. Valve position indication (open to closed position change) at CB. Valve closed position monitor light and alarm for group monitoring of components.1.Valve is normally open during plant operation (for alignment of ECCS). Valve interlocked so it must be closed before valves EJ-HV-8701A and BB-PV-8702A (EJ-HV-8701B and BB-PV-8702B) can be opened.2.See item 3 "Effect on System Operation".10.Solenoid-operated globe valve BG-HV-8154A (BG-HV-8154B analogous)a.Fails to open on demand. Provides isolation of fluid flow from the RCS to the PRT via the excess letdown heat exchanger. Failure reduces redundancy of providing flow from the RCS to the PRT. Negligible effect on safety for system operation. Letdown flow provided by parallel letdown path through alternate isolation valve BG-HV-8154B (BG-HV-8154A). Valve open/close position indication at CB; and letdown high temperature indication and alarm at CB.The letdown path to the PRT provides fluid flow out of the RCS to accommodate boration makeup flow into the RCS.b.Fails to close on demand.
Same function as that stated for item
10.a.Failure reduces redundancy of isolating flow from the RCS to the PRT. Negligible effect on safety for system operation. RCS letdown flow isolation provided by alternate series isolation valve BG-HV-8153A (BG-HV-8153B). Same methods of detection as those stated for item 10.a.11.Solenoid-operated globe valve BG-HV-8153A (BG-HV-8153B analogous)a.Fails to open on demand. Same function as that stated for item
10.a.Same effect on system operation as that stated for item 10.a, except for alternate isolation valve BG-HV-8153B (BG-HV-8153A).Same methods of detection as those stated for item 10.a.Same remarks as those stated for item 10.a.ComponentFailureModeFunctionEffectonSystemOperation*FailureDetection Method**Remarks CALLAWAY - SPTABLE 5.4A-3 (Sheet 6)
Rev. OL-21 5/15b.Fails to close on demand.
Same function as that stated for item 10.a.Same effect on system operation as that stated for item 10.b, except for alternate series isolation valve BG-HV-8154A (BG-HV-8154B).Same methods of detection as those stated for item 10.a.12.Solenoid-operated globe valve BB-HV-8157A (BB-HV-8157B analogous)Fails to open on demand.Same function as that stated for item 10.a.Same effect on system operation as that stated for item 10.a, except for alternate parallel isolation valve BB-HV-8157B (BB-HV-8157A).Same methods of detection as those stated stated for item 10.a.Same remarks as those stated for item 10.a.13.Solenoid-operated power-operated relief valve BB-PCV-456A (BB-PCV-455A analogous)a.Fails to open on demand.Provides relief to and isolation of fluid flow from pressurizer to PRT. Failure reduces redundancy of providing flow from pressurizer to PRT. Negligible effect on safety for system operation.
Pressurizer vent flow provided by a parallel pressurizer vent path through alternate relief valve BB-PCV-455A. Valve open/closed position indication at CB; pressurizer power-operated relief valve outlet temperature indication at CB.Pressurizer vent path to the PRT provides fluid flow out of the RCS to permit RCS depressurization to RHRS initiation conditions.b.Fails to close on demand.
Same function as that stated for item 13.a.Failure reduces redundancy of isolating flow from the pressurizer to the PRT. Negligible effect on safety. for system operation. Pressurizer vent flow isolation provided by series isolation valve BB-HV-8000B (BB-HV-8000A). Same methods of detection as those stated for item 13.a.14.Motor-operated gate valve BB-HV-8000A (BB-HV-8000B analogous)Fails to close on demand.Provides isolation of fluid flow from pressurizer to PRT.Same effect on system operation as that stated for item 13.b, except pressurizer vent flow isolation provided by series relief valve BB-PCV-455A (BB-PCV-456A) if the RCS pressure is below the PORV setpoint.Same methods of detection as those stated for item 13.a.Same remarks as those stated for item 13.a.ComponentFailureModeFunctionEffectonSystemOperation*FailureDetection Method**Remarks CALLAWAY - SPTABLE 5.4A-3 (Sheet 7)
Rev. OL-21 5/1515.Motor-operated gate valve EP-HV-8808A (EP-HV-8808B, EP-HV-8808C EP-HV-8808D analogous)Fails to close on demand.Provides isolation of fluid flow from accumulator 1 (accumulator 2, accumulator 3, accumulator 4) to the RCS. Failure prevents isolation of accumulator 1 (accumulator 2, accumulator 3, accumulator 4) from the RCS. Negligible effect on safety for system operation. Accumulator 1 (accumulator 2, accumulator 3, accumulator 4) is depressurized by opening vent isolation valve EP-HV-8950A (EP-HV-8950B or C, EP-HV-8950D or E, EP-HV-8950F). Valve open/closed position indication at CB; valve (closed) monitor light and alarm at CB; and accumulator pressure indication and low alarm at CB.Accumulators are isolated or vented during plant cooldown to not affect RCS depressurization to RHRS initiation
conditions.16.Solenoid-operated globe valve EP-HV-8950A (EP-HV-8950F analogous)Fails to open on demand. Provides venting of nitrogen gas from accumulator 1 (accumulator 4) to containment. Failure prevents venting of accumulator 1 (accumulator 4) to containment. Negligible effect on safety for system operation. Accumulator 1 (accumulator 4) is isolated from RCS by closing isolation valve EP-HV-8808A (EP-HV-8808D). Valve open/closed position indication at CB and accumulator pressure indication and low alarm at CB.Same remarks as those stated for item 15.17.Solenoid-operated globe valveEP-HV-8950B/
8950D (EP-HV-8950C/8950E analogous)Fails to open on demand.Provides venting of nitrogen gas from accumulator 2/
accumulator 3. Failure reduces redundancy in venting accumulator 2/accumulator 3. Negligible effect on safety for system operation. Accumulator 2/accumulator 3 venting capability provided by valves EP-HV-8950C/8950E if accumulator isolation valves EP-HV-8808B/8808C cannot be closed. Same methods of detection as those stated for item 16.Same remarks as those stated for item 15.ComponentFailureModeFunctionEffectonSystemOperation*FailureDetection Method**Remarks CALLAWAY - SPTABLE 5.4A-3 (Sheet 8)
Rev. OL-21 5/1518.Centrifugal charging pump PBG05A (PBG05B
analogous) Fails to deliver working fluid. Provides fluid flow of borated water from the RWST to the RCS and RCP seal injection.Failure reduces redundancy of providing borated water to the RCS at high RCS pressures.
Fluid flow from charging pump PBG05A (PBG05B) will be lost. Minimum flow requirements for boration, makeup, and seal injection will be met by PBG05B (PBG05A).Charging pump discharge header pressure and flow indication at CB. Open/
closed pump switchgear circuit breaker indication on CB. Circuit breaker closed position monitor light for group monitoring of component at CB.
Common breaker trip alarm at CB. The ECCS charging pumps provide boration, seal injection, and makeup flow to the RCS during safety grade cold shut down operations.19.Motor-operated gate valve BG-LCV-112C (BG-LCV-112B
analogous)Fails to close on demand. Provides isolation of fluid discharge from the VCT to the suction of charging pumps. Failure reduces redundancyof providing VCT discharge isolation. Negligible effect on safety for systemoperation.
Alternate isolation valve BG-LCV-112B (BG-LCV-112C) provides back-up tank discharge isolation. Valve open/closed position indication at CB and valve (closed) monitor light and alarm at CB.The ECCS charging pumps' suction is isolated from the VCT and aligned to the RWST (for boration/makeup) during safety grade cold shutdown operations.20.Motor-operated gate valve BG-HV-8105 (BG-HV-8106
analogous)Fails to close on demand. Provides isolation of fluid flow from the charging pump discharge header to the CVCS normal charging line to the RCS. Failure reduces redundancy of providing isolation of charging pump discharge to normal charging line of CVCS.
Negligible effect on safety for system operation. Alternate isolation valve BG-HV-8106 (BG-HV-8105) provides backup normal CVCS charging line isolation. Valve position indication (open to closed position change) at CB. Valve closed position monitor light and alarm for group monitoring of components at CB.Normal charging line is isolated during safety grade cold shutdown operations. Boration and makeup flow provided to RCS through redundant ECCS headers to the RCS cold legs.21.Motor-operated gate valve BN-LCV-112E (BN-LCV-112D analogous)Fails to open on demand. Provides isolation of fluid discharge from the RWST to the suction of charging pumps. Failure reduces redundancy of providing fluid flow from RWST to suction of PBG05B. Negligible effect on safety for system operation. Alternate isolation valve BN-LCV-112D (BN-LCV-112E) opens to provide backup flow path to suction of PBG05A. Valve open/closed position indication at CB and valve (open) monitor light and alarm at CB. The ECCS charging pumps' suction is aligned to the RWST for makeup/boration to the RCS during safety grade cold shutdown operations.ComponentFailureModeFunctionEffectonSystemOperation*FailureDetection Method**Remarks CALLAWAY - SPTABLE 5.4A-3 (Sheet 9)
Rev. OL-21 5/1522.Motor-operated globe valve BG-HV-8110 (BG-HV-8111 analogous)Fails to open on demand. Provides isolation of ECCS charging pump mini-flow
line. Failure reduces redundancy of providing boration/makeup flow from the RWST to the RCS under low flow throttled conditions where ECCS charging pump minimum flow requirements cannot be met without mini-flow. Negligible effect on safety for system operation.
PBG05B (PBG05A) minimum flow requirements will be met utilizing mini-flow isolation valve BG-HV-8111 (BG-HV-8110). Boration/
makeup flow requirements are satisfied by the redundant alternate train. Valve position indication (open to closed position change) at CB. Valve closed position monitor light and alarm for group monitoring of components at CB.1Valve aligned to close upon receipt of an SIS coincident with ECCS charging pump flow 258.9 gpm.2.Normally open valve.23.Motor-operated globe valve BG-HV-8357A (BG-HV-8357B analogous)Fails to open on demand.Provides safety grade seal injection flow path. Failure reduces redundancy of providing seal injectionflow to the RCP seals. Negligible effect on safety for system operation. Alternate valve BG-HV-8357B (BG-HV-8357A) opens to provide a seal injection flow path to the RCPs.
Seal injection flow requirements are satisfied by the redundant alternate path. Valve open/closed position indication at CB.24.Motor-operated gate valve EJ-HV-8716A (EJ-HV-8716B analogous)Fails to close on demand. Provides separation between the two RHR trains during cooldown operation. Failure reduces redundancy for isolating RHR trains during cooldown. Negligible effect on system operation. Isolation valve EJ-HV-8716B (EJ-HV-8716A) provides backup isolation between the two RHR trains. Valve open/closed position indication at CB and valve (closed) monitor light and alarm at CB.ComponentFailureModeFunctionEffectonSystemOperation*FailureDetection Method**Remarks CALLAWAY - SPTABLE 5.4A-3 (Sheet 10)
Rev. OL-21 5/15List of acronyms and abbreviationsCB-Control boardCVCS-Chemical and volume control systemECCS-Emergency core cooling system RC-Reactor coolantRCS-Reactor coolant systemRHR-Residual heat removal RHRS-Residual heat removal systemRWST-Refueling water storage tankSG-Steam generator RCP-Reactor coolant pump25.Motor-operated globe valve EM-HV-8803A (EM-HV-8803B analogous)Fails to open on demand. Provides flow control from PBG05A (PBG05B) to RCS for boration/
makeup. Failure reduces redundancy of providing boration/makeup flow to RCS from PBG05A (PBG05B). Negligible effect on safety for system operation. Alternate control valve EM-HV-8803B (EM-HV-8803A) controls flow from PBG05B (PBG05A). Valve open/closed position indication at CB. Valve open position monitor light and alarm for group monitoring of components. Path utilized for boration/makeup flow to RCS for safety grade cold shutdown operation. (Also ECCS injection/
recirculation)26.Motor-operated gate valve BB-HV-8037A (BB-HV-8037B analogous)Fails to open on demand.PRT to containment sump isolation.
Boron injection discharge to RCS.Failure reduces redundancy of providing flow from the PRT to containment sump. Negligible effect on safety for system operation. Letdown flow provided by parallel path through alternate isolation valve BB-HV-8037B (BB-HV-8037A). Valve position indication (closed to open position change) at CB. Valve
open position monitor light and alarm for group monitoring of components. Letdown path to containment sump provides flow out of PRT to accommodate flow out of RCS during shutdown operations. 27.Motor-operated gate valve EM-HV-8801A (EM-HV-8801B analogous)Fails to open on demand.Boron injection discharge to RCS.Failure reduces redundancy of providing flow via the boron injection header to RCS. Negligible effect on safety for system operation. Flow path provided by parallel isolation valve EM-HV-8801B (EM-HV-8801A).Valve position indication (closed to open position change) at CB. Valve open position monitor light and alarm for group
monitoring of components.Path utilized for boration/makeup flow to RCS for safety grade cold shutdown operation. (Also ECCS injection/recirculation)ComponentFailureModeFunctionEffectonSystemOperation*FailureDetection Method**Remarks CALLAWAY - SPTABLE 5.4A-3 (Sheet 11)
Rev. OL-21 5/15VCT-Volume control tankPRT-Pressurizer relief tank