Text

Amendment 28 to Updated Final Safety Analysis Report, Chapter 4, Reactor Coolant System
	ML19298B510
Person / Time
Site:	Browns Ferry
Issue date:	10/04/2019
From:	; Tennessee Valley Authority
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML19298B540	List: ML19298B452; ML19298B455; ML19298B456; ML19298B458; ML19298B461; ML19298B464; ML19298B467; ML19298B469; ML19298B471; ML19298B474; ML19298B477; ML19298B478; ML19298B480; ML19298B485; ML19298B488; ML19298B490; ML19298B497; ML19298B500; ML19298B508; ML19298B510; ML19298B515; ML19298B523; ML19298B525; ML19298B528;
References
	Download: ML19298B510 (146)
	v • d • e

{{#Wiki_filter:BFN-26 4.0-i REACTOR COOLANT SYSTEM TABLE OF CONTENTS 4.0 Reactor Coolant System............................................................................................................ 4.1-1 4.1 Summary Description................................................................................................................ 4.1-1 4.2 Reactor Vessel And Appurtenances Mechanical Design........................................................... 4.2-1 4.2.1 Power Generation Objective...................................................................................... 4.2-1 4.2.2 Power Generation Design Basis................................................................................ 4.2-1 4.2.3 Safety Design Basis................................................................................................... 4.2-1 4.2.4 Description................................................................................................................. 4.2-2 4.2.5 Safety Evaluation....................................................................................................... 4.2-10 4.2.6 Inspection and Testing............................................................................................... 4.2-15 4.3 Reactor Recirculation System................................................................................................... 4.3-1 4.3.1 Power Generation Objective...................................................................................... 4.3-1 4.3.2 Power Generation Design Basis................................................................................ 4.3-1 4.3.3 Safety Design Basis................................................................................................... 4.3-1 4.3.4 Description................................................................................................................. 4.3-1 4.3.5 Safety Evaluation....................................................................................................... 4.3-7 4.3.6 Inspection and Testing............................................................................................... 4.3-8 4.4 Nuclear System Pressure Relief System................................................................................... 4.4-1 4.4.1 Safety Objective......................................................................................................... 4.4-1 4.4.2 Power Generation Objective...................................................................................... 4.4-1 4.4.3 Safety Design Basis................................................................................................... 4.4-1 4.4.4 Power Generation Design Basis................................................................................ 4.4-2 4.4.5 Description................................................................................................................. 4.4-2 4.4.6 Safety Evaluation....................................................................................................... 4.4-9 4.4.7 Inspection and Testing............................................................................................... 4.4-10 4.5 Main Steam Line Flow Restrictor............................................................................................... 4.5-1 4.5.1 Safety Objective......................................................................................................... 4.5-1 4.5.2 Safety Design Basis................................................................................................... 4.5-1 4.5.3 Description................................................................................................................. 4.5-1 4.5.4 Safety Evaluation....................................................................................................... 4.5-2 4.5.5 Inspection and Testing............................................................................................... 4.5-3 4.6 Main Steam Isolation Valves..................................................................................................... 4.6-1 4.6.1 Safety Objectives....................................................................................................... 4.6-1 4.6.2 Safety Design Basis................................................................................................... 4.6-1 4.6.3 Description................................................................................................................. 4.6-2 4.6.4 Safety Evaluation....................................................................................................... 4.6-5 4.6.5 Inspection and Testing............................................................................................... 4.6-8

BFN-26 4.0-ii REACTOR COOLANT SYSTEM TABLE OF CONTENTS (Cont'd) 4.7 Reactor Core Isolation Cooling System..................................................................................... 4.7-1 4.7.1 Power Generation Objective...................................................................................... 4.7-1 4.7.2 [Deleted]..................................................................................................................... 4.7-1 4.7.3 Power Generation Design Basis................................................................................ 4.7-1 4.7.4 Safety Design Basis................................................................................................... 4.7-1 4.7.5 Description................................................................................................................. 4.7-1 4.7.6 Safety Evaluation....................................................................................................... 4.7-4 4.7.7 Inspection and Testing............................................................................................... 4.7-4 4.8 Residual Heat Removal System (RHRS).................................................................................. 4.8-1 4.8.1 Safety Objective......................................................................................................... 4.8-1 4.8.2 Power Generation Objective...................................................................................... 4.8-1 4.8.3 Safety Design Basis................................................................................................... 4.8-1 4.8.4 Power Generation Design Basis................................................................................ 4.8-2 4.8.5 Summary Description................................................................................................. 4.8-2 4.8.6 Description................................................................................................................. 4.8-4 4.8.7 Safety Evaluation....................................................................................................... 4.8-8 4.8.8 Inspection and Testing............................................................................................... 4.8-8 4.9 Reactor Water Cleanup System................................................................................................ 4.9-1 4.9.1 Power Generation Objective...................................................................................... 4.9-1 4.9.2 Power Generation Design Basis................................................................................ 4.9-1 4.9.3 Description (Figures 4.9-1, 4.9-2, 4.9-3, 4.9-5, 4.9-6, 4.9-7, 4.9-8, 4.9-9, and 4.9-10)................................................................................................ 4.9-1 4.9.4 Inspection and Testing............................................................................................... 4.9-3 4.10 Nuclear System Leakage Rate Limits....................................................................................... 4.10-1 4.10.1 Safety Objective......................................................................................................... 4.10-1 4.10.2 Safety Design Basis................................................................................................... 4.10-1 4.10.3 Description................................................................................................................. 4.10-1 4.10.4 Safety Evaluation....................................................................................................... 4.10-7 4.10.5 Inspection and Testing............................................................................................... 4.10-7 4.11 Main Steam Lines, Feedwater Piping, and Drains..................................................................... 4.11-1 4.11.1 Power Generation Objective...................................................................................... 4.11-1 4.11.2 Safety Design Basis................................................................................................... 4.11-1 4.11.3 Power Generation Design Bases............................................................................... 4.11-1 4.11.4 Description................................................................................................................. 4.11-1 4.11.5 Safety Evaluation....................................................................................................... 4.11-3 4.11.6 Inspection and Testing............................................................................................... 4.11-3

BFN-26 4.0-iii REACTOR COOLANT SYSTEM TABLE OF CONTENTS (Cont'd) 4.12 Inservice Inspection And Testing............................................................................................... 4.12-1 4.12.1 Introduction................................................................................................................ 4.12-1 4.12.2 Scope......................................................................................................................... 4.12-2 4.12.3 Responsibility............................................................................................................. 4.12-2 4.12.4 Area and Extent of Examination................................................................................. 4.12-2

BFN-28 4.0-iv REACTOR COOLANT SYSTEM LIST OF TABLES Table Title 4.2-1 Reactor Pressure Vessel Materials 4.2-2 Reactor Vessel Data 4.2-3 Reactor Vessel Attachments 4.3-1 Reactor Recirculation System Design Characteristics (3952 MWt) 4.3-1b (Deleted) 4.4-1 (Deleted) 4.4-1A Nuclear System Main Steam Relief Valves (Units 2 and 3) 4.7-1 Reactor Core Isolation Cooling System Turbine - Pump Design Data 4.8-1 Residual Heat Removal System Equipment Design Data 4.9-1 Reactor Water Cleanup System Equipment Design Data

BFN-26 4.0-v REACTOR COOLANT SYSTEM LIST OF ILLUSTRATIONS Figure Title 4.2-1 Reactor Vessel 4.2-2 Reactor Vessel Nozzles and Penetrations 4.2-3 Reactor Vessel 4.2-4 Reactor Vessel 4.3-1 Recirculation System--Elevation, Isometric 4.3-2a sht 1 Nuclear Boiler Flow Diagram 4.3-2a sht 2 Nuclear Boiler Flow Diagram 4.3-2a sht 3 Nuclear Boiler Flow Diagram 4.3-2b (Deleted) 4.3-3 Jet Pump--Operating Principle 4.3-4 Recirculation System--Core Flooding Capability 4.4-1 2-Stage Safety/Relief Valve Schematic (Closed Position) 4.4-2 2-Stage Safety/Relief Valve Schematic (Open Position) 4.4-3 Safety Valve Sizing Analysis 4.4-4 Deleted by Amendment 13 4.4-5 Deleted by Amendment 13 4.4-6 T-Quencher for Safety/Relief Discharge 4.4-7 Mechanical Main Steam Relief Valve Vent Piping 4.4-8 Mechanical Main Steam Relief Valve Vent Piping 4.5-1 Primary Steam Piping 4.5-2 Primary Steam Piping 4.5-3 Primary Steam Piping 4.6-1 Main Steam Isolation Valve 4.7-1a Reactor Core Isolation Cooling System Flow Diagram 4.7-1b Reactor Core Isolation Cooling System, Mechanical Control Diagram 4.7-1c Reactor Core Isolation Cooling System - Flow Diagram 4.7-1d Reactor Core Isolation Cooling System - Mechanical Control Diagram 4.7-1e Reactor Core Isolation Cooling System - Mechanical Control Diagram 4.7-1f Reactor Core Isolation Cooling System - Flow Diagram 4.7-2a (Deleted) 4.7-2b (Deleted) 4.7-2c (Deleted) 4.7-2d (Deleted) 4.7-2e (Deleted) 4.7-2f (Deleted) 4.7-2g (Deleted) 4.7-2h (Deleted) 4.8-1 Residual Heat Removal System, Unit Cross Connections and Standby Coolant Supply 4.9-1 Reactor Water Cleanup System Flow Diagram 4.9-2 Reactor Water Cleanup Demineralizer Flow Diagram 4.9-3 Reactor Water Cleanup System, Mechanical Control Diagram 4.9-4 (Deleted)

BFN-26 4.0-vi REACTOR COOLANT SYSTEM LIST OF ILLUSTRATIONS (contd) Figure Title 4.9-4a (Deleted) 4.9-4b (Deleted) 4.9-4c (Deleted) 4.9-4d (Deleted) 4.9-5 Reactor Water Cleanup System - Flow Diagram 4.9-6 Reactor Water Cleanup Demineralizer - Flow Diagram 4.9-7 Reactor Water Cleanup System - Mechanical Control Diagram 4.9-8 Reactor Water Cleanup System - Flow Diagram 4.9-9 Reactor Water Cleanup Demineralizer - Flow Diagram 4.9-10 Reactor Water Cleanup System - Mechanical Control Diagram 4.10-1 Drywell Leak Detection System Diagram 4.10-2 Deleted 4.10-3 Axial Through-Wall Crack Data Correlation 4.11-1 Feedwater Piping Arrangement

BFN-16 4.1-1 4.0 REACTOR COOLANT SYSTEM 4.1

SUMMARY

DESCRIPTION The subsections in the "Reactor Coolant System" section describe those systems and components that form the major portions of the nuclear system process barrier. These systems and components contain or transport the fluids coming from or going to the reactor core. The "Reactor Vessel and Appurtenances Mechanical Design" subsection describes the reactor vessel and the various fittings with which other systems are connected to the vessel. The major safety considerations for the reactor vessel are concerned with the ability of the vessel to function as a radioactive material barrier. Various combinations of structural loading are considered in the vessel design. The vessel meets the requirements of various applicable codes and criteria. The possibility of brittle fracture is considered, and suitable limits are established that avoid conditions where brittle fracture is possible. Periodic, cumulative-fatigue usage evaluations are performed for each reactor vessel to verify that the vessel does not approach usage limits. The Reactor Recirculation System pumps coolant through the core. Adjustment of the core coolant flow rate changes reactor power output, thus providing a means of following plant load demand or manually changing reactor output without adjusting control rods. The recirculation system is designed with sufficient fluid and pump inertia that fuel thermal limits will not be exceeded as a result of recirculation system malfunctions. The arrangement of the recirculation system is designed so that a piping failure cannot compromise the integrity of the floodable inner volume of the reactor vessel. The Nuclear System Pressure Relief System is designed to protect the nuclear system process barrier from damage due to overpressure. To accomplish overpressure protection a number of main steam relief valves are provided that can discharge steam from the nuclear system to the primary containment. The Nuclear System Pressure Relief System also acts to automatically depressurize the nuclear system in the event of a loss-of-coolant accidents in which the High Pressure Coolant Injection System (HPCIS) fails to deliver sufficient flow. The depressurization of the nuclear system allows low pressure Emergency Core Cooling Systems to supply enough cooling water to adequately cool the fuel. Six of the main steam relief valves used to provide overpressure protection are arranged to effect automatic depressurization. The main steam line flow restrictors are venturi-type flow devices. One restrictor is installed in each main steam line close to the reactor vessel but downstream of the main steam relief valves. The restrictors are designed to limit the loss of coolant resulting from a main steam line break outside the primary containment. The

BFN-16 4.1-2 coolant loss is limited so that reactor vessel water level remains above the top of the core during the time required for the main steam isolation valves to close. This action protects the fuel barrier. Two main steam isolation valves are installed on each main steam line. One valve in each line is located inside the primary containment, the other outside. These valves act automatically to close off the nuclear system process barrier in the event a pipe break occurs downstream of the valves. This action limits the loss of coolant and the release of radioactive materials from the nuclear system. In the event that a main steam line break occurs inside the primary containment, closure of the isolation valve outside the containment acts to seal the primary containment itself. The Reactor Core Isolation Cooling System (RCICS) includes a turbine-pump driven by reactor vessel steam. Under certain conditions the system automatically starts in time to prevent the core from becoming uncovered without the use of the Core Standby Cooling Systems. The system provides the ability to cool the core during a reactor shutdown in which feedwater flow is not available. The Residual Heat Removal System (RHRS) includes a number of pumps and heat exchangers that can be used to cool the nuclear system under a variety of situations. During normal shutdown and reactor servicing, the RHRS removes residual and decay heat. One operational mode of the RHRS is low pressure coolant injection (LPCI). LPCI operation is an engineered safeguard for use during a loss-of-coolant accident; this operation is described in Section 6.0, "Emergency Core Cooling Systems." Another mode of RHRS operation allows the removal of heat from the primary containment following a loss-of-coolant accident. The Reactor Water Cleanup System functions to maintain the required purity of reactor coolant by circulating coolant through a system of filter/demineralizers. The "Nuclear System Leakage Rate Limits" subsection establishes the limits on nuclear system leakage inside the primary containment so that appropriate action can be taken before the nuclear system process barrier is threatened by a crack large enough to propagate rapidly. The main steam lines, feedwater piping, and their associated drains are attached to the reactor vessel and provide core coolant flow paths external to it. These lines penetrate the primary containment and specified portions of these lines must provide adequate nuclear system process barrier for normal and accident conditions. Four steam lines are utilized between the reactor and the turbine which permit turbine stop valve and main steam isolation valve tests during plant operation with a minimum amount of load reduction. In addition, differential pressures on reactor internals under assumed accident conditions of a broken steam line are limited. Feedwater lines provide water to the reactor vessel entering near the top of the

BFN-16 4.1-3 vessel downcomer annulus. Drains are provided at the low point of each main steam line, at the reactor vessel bottom head, and on each side of the recirculation pumps. The program for preoperational examination and periodic inservice examination of Reactor Coolant System is also defined.

BFN-27 4.2-1 4.2 REACTOR VESSEL AND APPURTENANCES MECHANICAL DESIGN 4.2.1 Power Generation Objective The reactor vessel power generation design objective is to provide a volume in which the core can be submerged in coolant, thereby allowing power operation of the fuel. The reactor vessel and appurtenances design provides the means for the attachment of pipelines to the reactor vessel and the means for the proper installation of vessel internal components. 4.2.2 Power Generation Design Basis

1.

The location and design of the external and internal supports provided as an integral part of the reactor vessel shall be such that stresses in the reactor vessel and supports due to reactions at these supports are within ASME Boiler and Pressure Vessel Code limits.

2.

The reactor vessel design lifetime shall be 40 years. Time Limited Aging Analyses (TLAAs) have been identified and evaluated for the reactor vessel 60 year operating life. Summaries of these evaluations for the reactor vessel life are provided in Appendix O, Section O.3.1 and O.3.2.

3.

The design of the reactor vessel and appurtenances shall allow for the accomplishment of a suitable program of periodic inspection and surveillance. 4.2.3 Safety Design Basis

1.

The reactor vessel and appurtenances shall be designed to withstand adverse combinations of loadings and forces resulting from operation under abnormal and accident conditions.

2.

The reactor vessel shall be designed and fabricated to a high standard of quality to provide assurance of an extremely low probability of failure.

3.

To minimize the possibility of brittle fracture failure of the nuclear system process barrier, the following shall be required: (1) the initial ductile-brittle transition temperature of materials used in the reactor vessel shall be known by reference or established empirically; (2) expected shifts in transition temperature during design service life due to environmental conditions, such as neutron flux, shall be determined and employed in the reactor vessel design; and (3) operation margins to be observed with regard to the transition temperature shall be designated for each mode of operation.

BFN-27 4.2-2

4.

The design shall provide for material surveillance specimens which may be used to verify predicted radiation exposure and to measure the effect of radiation on the vessel material. 4.2.4 Description 4.2.4.1 Reactor Vessel The reactor vessel is a vertical, cylindrical pressure vessel with hemispherical heads of welded construction. The reactor vessel is designed and fabricated for a useful life of 40 years based upon the specified design and operating conditions. TLAAs have been identified and evaluated for the reactor vessel 60 year operating life. Summaries of these evaluations for the reactor vessel life are provided in Appendix O, Sections O.3.1 and O.3.2. The vessel for each unit is designed, fabricated, inspected, and tested in accordance with the ASME Boiler and Pressure Vessel Code, Section III, 1965 edition, Summer 1965 addenda (Unit 3 vessel - Summer 1966 addenda), code cases 1332-1, 1332-2, 1332-3, 1334, 1335-2 (paragraph 4), 1336, 1339, applicable requirements for Class A vessels as defined therein, and additional GE requirements. The reactor vessel and its supports are designed as Class I equipment in accordance with the loading criteria of Appendix C. The materials used in the design and fabrication of the reactor pressure vessel are shown in Table 4.2-1. The Browns Ferry Unit 1 vessel was fabricated by B&W. The Browns Ferry Units 2 and 3 vessels were fabricated by Ishikawajima-Harima Heavy Industries Co. (IHI) in Japan, under a contract between B&W and IHI. IHI had previously manufactured the Fukushima I and II vessels. These vessels are built to the ASME Boiler and Pressure Vessel Code and GE specifications. Reactor vessel data is presented in Table 4.2-2. The cylindrical shell and bottom hemispherical head of the reactor vessel are fabricated of low alloy steel plate which is clad on the interior with weld overlay. The cylindrical shell is clad with stainless steel, and the bottom hemispherical head is clad with Inconel. The plates and forgings are ultrasonically tested and magnetic-particle-tested over 100 percent of their surfaces after forming and heat treatment. Full-penetration welds are used at all joints, including nozzles, throughout the vessel, except for nozzles of less than 3-inch nominal size and the CRD housing-to-stub tube welds. Nozzles of less than 3-inch nominal size are partial-penetration-welded as permitted by ASME Boiler and Pressure Vessel Code, Section III. The electroslag weld process was used on the Browns Ferry pressure vessels. Electroslag welding process variables, quality control procedures and technical details were presented in Appendix F, Dresden 2/3 FSAR, Docket Nos. 50-237 and 50-249.

BFN-27 4.2-3 Although little corrosion of plain carbon or low-alloy steels occurs at temperatures of 500F to 600F, higher corrosion rates occur at temperatures around 140F. The 0.125-inch minimum-thickness cladding provides the necessary corrosion resistance during reactor shutdown and also helps maintain water clarity during refueling operations. Since the vessel head is exposed to a saturated steam environment throughout its operating lifetime, stainless steel cladding is not required over its interior surfaces. Exterior, exposed ferritic surfaces of pressure-containing parts have a minimum corrosion allowance of 1/16 inch. The interior surfaces of the top head and all carbon and low-alloy steel nozzles exposed to the reactor coolant have a corrosion allowance of 1/16 inch. The vessel shape is designed to limit coolant retention pockets and crevices. The nil-ductility transition (NDT) temperature is defined as the temperature below which ferritic steel breaks in a brittle, rather than ductile, manner. The NDT temperature increases as a function of neutron exposure at integrated neutron exposures greater than about 1 x 1017 nvt with neutrons of energies in excess of 1 MeV. Since the material NDT temperature dictates the minimum operating temperature at which the reactor vessel can be pressurized, it is desirable to keep the NDT temperature as low as possible. One way that this is accomplished is by selecting fine-grained steels and by using advanced fabrication techniques to minimize radiation effects. The as-fabricated initial NDT temperature for all carbon and low-alloy steel used in the main closure flanges, closure bolting material, and the shell and head materials connecting to these flanges, including the connecting circumferential weld material, is limited to a maximum of 10F as determined by ASTM E208. For each main closure flange forging, a minimum of 1 tensile, 3 Charpy V-notch, and 2 drop weight test specimens have been tested from each of two locations about 180 apart on the flange. For all other carbon and low-alloy steel pressure-containing materials, including weld materials and the vessel support skirt material, the initial NDT temperature is no higher than 56F for Unit 1, and 40F for Units 2 and 3. A grain size of 5 or finer, as determined by the method in ASTM E112, is maintained. Another way of minimizing any changes (elevating) to the NDT temperature is by reducing the integrated neutron exposure at the inner surface of the reactor vessel. The coolant annulus between the vessel and core shroud and the core location in the vessel limit the integrated neutron exposure of reactor vessel material to less than 1 x 1019 nvt from neutrons with energy levels greater than 1 MeV, within the 40-year design lifetime of the vessel. TLAAs have been identified and evaluated for the reactor vessel 60 year operating life. Summaries of these evaluations for the reactor vessel life are provided in Appendix O, Sections O.3.1 and O.3.2. This is not the expected exposure, nor is it the absolute limit of safe exposure; it is an exposure value that can be demonstrated to be safe and practical to maintain. The maximum calculated exposure for neutrons of 1 MeV or greater is 1.58 X 1018 nvt for Unit 1, per GEH Report No. 0000-0166-0632-R0. The maximum calculated exposure for

BFN-28 4.2-4 neutrons of 1 MeV or greater is less than 1 X 1019 nvt for Unit 2 and 2.23 X 1018 nvt for Unit 3. The vessel top head is secured to the reactor vessel by studs and nuts which are designed to be tightened with a stud tensioner. The vessel flanges are sealed by two concentric metallic seal-rings designed for no detectable leakage through the inner or outer seal at any operating condition, including: (a) cold hydrostatic pressure test at the hydro-pressure specified in the ASME code, and (b) heating to operating pressure and temperature at a maximum rate of 100F/hr. To detect lack of seal integrity, a 1-inch vent tap is provided in the area between the two seal-rings, and a monitor line is attached to the tap to provide an indication of leakage from the inner seal-ring seal (see Subsection 7.8). A 1-inch tap is also provided in the area outside the outer seal-ring for use in monitoring leakage. This tap is used only if the inner seal fails and is piped to an accessible place in the drywell and capped. The head and vessel flanges are low-alloy steel forgings. The sealing surfaces of the reactor vessel head and shell flanges are weld-overlay clad with Inconel 82 (ERNiCr material. The clad thickness is 0.25 inches on both the head flange and shell flange sealing surfaces. All sensitized austenitic stainless steel has been replaced on the Browns Ferry pressure vessels, except the jet pump riser brace pads on all units. These components have been clad with nonfurnace-sensitized stainless steel weld overlay. Austenitic stainless steel used in other component parts of the reactor coolant pressure boundary, including relief and safety valves, is fully annealed to preclude sensitization. Welding processes were limited to 110,000 joules per inch and the interpass temperature limited to 350F to avoid local sensitization of stainless steel. Stainless steel with deliberate additions of nitrogen for enhancing the material strength has not been used. The vessel nozzles (Figure 4.2-2) are low-alloy steel forgings made in accordance with ASTM A508 CL2 as modified by ASME code case 1332-2, paragraph 5. Nozzles of 3-inch nominal size or larger are full-penetration welded to the vessel. Nozzles of less than 3-inch nominal size may be partial-penetration-welded as permitted by ASME Boiler and Pressure Vessel Code, Section III. Nozzles which are partial-penetration welded are nickel-chromium-iron forgings made in accordance with ASME SB166 as modified by code case 1336. The vessel top head nozzles are provided with flanges with small groove facing. The drain nozzle is of the full-penetration weld design and extends 16 inches below the bottom outside surface of the vessel. The recirculation inlet nozzles are located as shown in Figures 4.2-1, 4.2-3, and 4.2-4; feedwater inlet nozzles, core spray inlet

BFN-27 4.2-5 nozzles, and the control rod drive hydraulic system return nozzle have thermal sleeves similar to those shown in the detail of Figure 4.2-2. As a result of cracks discovered in the feedwater nozzle blend and nozzle bore regions of several operating reactors, General Electric and the NRC performed an extensive study of the problem. The program, the solutions, and NRC acceptance of the modifications are fully described in NEDE 21821-A, "Boiling Water Reactor Feedwater Nozzle/Sparger - Final Report," February 1980 (proprietary version). The modifications to the BFNP feedwater nozzles included: (1) removal of the stainless-steel-clad and heat affected zone of the feedwater nozzle bore and nozzle bend radius, and (2) machining the safe end and nozzle bore and inner bend radius to accept the improved double piston ring seal, interference fit spargers with forged tee design, and orificed elbow discharges. Implementing these modifications increased the assurance of maintaining vessel integrity by minimizing the potential for crack initiation due to thermal cycling. The nozzle for the core differential pressure and standby liquid control pipe is designed with a transition so that the stainless steel outer pipe of the differential pressure and standby liquid control line (see Subsection 3.3, "Reactor Vessel Internals Mechanical Design") can be socket-welded to the inner end of the nozzle and so that the inner pipe passes through the nozzle. This design provides an annular region between the nozzle and the inner standby liquid control line to minimize thermal shock effects on the reactor vessel in the event that use of the Standby Liquid Control System is required. The jet pump instrumentation penetration seal is welded directly to the outer end of the jet pump instrumentation nozzle. The stainless steel recirculation loop piping (see Subsection 4.3, "Reactor Recirculation System") is welded to the outer end of the recirculation outlet nozzle. The main steam line piping is welded to the outer end of the steam outlet nozzle. The piping attached to the vessel nozzle is designed, installed, and tested in accordance with the requirements of USAS B31.1.0, 1967 edition and the applicable GE design and procurement specifications, which were implemented in lieu of the outdated B31 Nuclear Code Cases-N2, N7, N9, and N10. Thermocouple pads are located on the exterior of the vessel (see Table 4.2-3). At each thermocouple location, two 3/4-inch-diameter pads are provided: an end pad to hold the end of a 3/16-inch-diameter thermocouple and a clamp pad equipped with a set screw to secure the thermocouple. The reactor vessel is laterally and vertically supported and braced to make it as rigid as possible without impairing the movements required for thermal expansion. Where thermal requirements prohibit the use of rigid supports, spring anchors or hydraulic snubbers are employed to resist earthquake forces, while allowing sufficient flexibility for thermal expansion.

BFN-27 4.2-6 4.2.4.2 Shroud Support The reactor vessel shroud support is a radial, cylindrical shell that surrounds the reactor core assembly and is designed so that stresses due to reactions at the shroud support are within ASME code, Section III, requirements for normal, upset, emergency, and faulted loading conditions. The design of the shroud support also takes into account the restraining effect of the components attached to the support, their weight, and earthquake loadings. The vessel shroud support and other internal attachments (jet pump riser support pads, diffuser brackets, guide rod brackets, steam dryer support brackets, dryer holddown brackets, feedwater sparger brackets, and core spray brackets) are as shown in Figures 4.2-1, 4.2-3, and 4.2-4. 4.2.4.3 Reactor Vessel Support Assembly The reactor vessel support assembly consists of a ring girder, sole plates, and the various bolts, shims, and set screws necessary to position and secure the assembly between the reactor vessel support skirt and the support pedestal. The concrete and steel support pedestal is constructed integrally with the building foundation. Steel anchor bolts are set in the concrete with the threads extending above the surface. The sole plates are set flat and level on the concrete, and the lower flange of the ring girder is set on top of the sole plates and shimmed as necessary to level the ring girder. The anchor bolts extend through both the sole plates and the ring girder bottom flange. High strength bolts are used to bolt the flange of the reactor vessel support skirt to the top flange of the ring girder. The ring girder and sole plates are fabricated of ASTM A36 structural steel according to AISC specifications. 4.2.4.4 Vessel Stabilizers The vessel stabilizers are connected between the reactor vessel and the top of the shield wall surrounding the vessel to provide lateral stability for the upper part of the vessel. Eight stabilizer brackets are attached by full-penetration welds to the reactor vessel at evenly spaced locations around the vessel below the flange. Each vessel stabilizer consists of a stabilizer rod, threaded at the ends, springs, washers, nut, a plate, and a bumper bracket with tapered shims. The stabilizers are attached to each bracket and apply tension in opposite directions. The stabilizers are evenly preloaded with tensioners to the values of the residual loads. The stabilizers are designed to permit radial and axial vessel expansion, to limit horizontal vibration, and to resist seismic and jet reaction forces. 4.2.4.5 Refueling Bellows The refueling bellows form a seal between the reactor vessel and the surrounding primary containment drywell to permit flooding of the space (reactor well) above the vessel during refueling operations. The refueling bellows assembly (see Figures

BFN-27 4.2-7 4.2-1, 4.2-3, and 4.2-4) consists of a Type 304 stainless steel bellows, a backing plate, a spring seal, and a removable guard ring. The backing plate surrounds the outer circumference of the bellows to protect it and is equipped with a tap for testing and for monitoring leakage. The self-energizing spring seal is located in the area between the bellows and the backing plate and is designed to limit water loss in the event of a bellows rupture by yielding to make a tight fit to the backing plate when subjected to full hydrostatic pressure. The guard ring attaches to the assembly and protects the inner circumference of the bellows. The guard ring can be removed from above to inspect the bellows. The assembly is welded to the reactor bellows support skirt and the reactor well seal bulkhead plate. The reactor bellows support skirt is welded to the reactor vessel shell flange, and the reactor well seal bulkhead plate bridges the distance to the primary containment drywell wall. Six watertight, hinged covers are bolted in place for normal refueling operation. For normal operation, these covers are opened and removable air supply ducts and air return ducts permit circulation of ventilation air in the region above the reactor well seal. 4.2.4.6 Control Rod Drive Housings The control rod drive housings are inserted through the control rod drive penetrations in the reactor vessel bottom head and are welded to the stub tubes extending into the reactor vessel1 (Figure 4.2-2). Each housing transmits a number of loads to the bottom head of the reactor. These loads include the weight of a control rod and control rod drive, which are bolted to the housing from below (see Subsection 3.4, "Reactivity Control Mechanical Design"), the weight of a control rod guide tube, one four-lobed fuel support piece, and the four fuel assemblies which rest on the top of the fuel support piece (see Subsection 3.3, "Reactor Vessel Internal Mechanical Design"). The housings are fabricated of Type 304 austenitic stainless steel. 4.2.4.7 Control Rod Drive Housing Supports The control rod drive housing support is designed to prevent a nuclear transient in the unlikely event that there is a control rod drive housing failure. This device consists of a grid structure located below the reactor vessel from which housing supports are suspended. The supports allow only slight movement of the control rod drive or housing in the event of failure. The control rod drive housing support is described in detail in Subsection 3.5, "Control Rod Drive Housing Supports." 1 Kobsa, I. R., and Wetzerl, V. R., "Design and Analysis of Control Rod Drive Reactor Vessel Penetrations," General Electric Co., Atomic Power Equipment Department, November 1968 (APED-5703).

BFN-27 4.2-8 4.2.4.8 In-Core Neutron Flux Monitor Housing The in-core neutron flux monitor housings are inserted up through the in-core penetrations in the bottom head of the reactor vessel and are welded to the inner surface of the bottom head (Figure 4.2-2). An in-core flux monitor guide tube is welded to the top of each housing (see Subsection 3.3, "Reactor Vessel Internals Mechanical Design"), and either a source range monitor/intermediate range monitor (SRM/IRM) drive unit or a local power range monitor (LPRM) is bolted to the seal-ring flange at the bottom of the housing (see Subsection 7.5, "Neutron Monitoring System"). 4.2.4.9 Reactor Vessel Insulation The reactor vessel insulation is an all-metal, reflective insulation having an average maximum heat transfer rate of approximately 80 Btu/hr-ft2 at the operating conditions of 550F for the vessel and 135F for the outside air. The maximum insulation thickness ranges from 4 inches for the upper head to 3-1/2 inches for the cylindrical shell and nozzles and 3 inches for the bottom head. The insulation is designed to permit complete submersion in water without loss of insulating material, contamination from the water, or adverse effect on the insulation efficiency of the insulation assembly after draining and drying. The lower head and cylindrical shell insulation is permanently installed for the 60 year operating life of the vessel. The insulation panels for the cylindrical shell of the vessel are held in place by vessel insulation supports located at two elevations on the vessel. The support brackets for each support are full-penetration-welded to the vessel at 12 evenly spaced locations around the circumference. Provisions are made for removing insulation during inservice inspection. 4.2.4.10 Other Reactor Coolant Pressure Boundary Ferritic Components The fracture or notch-toughness properties and the operating temperature of ferritic materials are controlled to ensure adequate toughness when the system is pressurized to more than 20 percent of the design pressure. Such assurance is provided by maintaining the lowest service metal temperature, when the system pressure exceeds 20 percent of design pressure, at least 60F above the nil-ductility transition temperature (NDTT). The lowest service-metal temperature is the lowest temperature which the metal will experience in service while the plant is in operation. It is established by appropriate calculations considering atmosphere ambient temperatures, the insulation or enclosure provided, and the minimum temperature maintained. Further interpretations and requirements are as follows:

BFN-27 4.2-9 A. Charpy V-notch (American Society for Testing and Material Standard A370 Type A) or drop weight (per ASTM E208) tests have been performed to demonstrate that all materials and weld metal meet brittle fracture requirements at test temperature. Test specimens, for the surveillance capsule pulled in 1994, were prepared and tested with minimum impact energy requirements in accordance with Table N-421 and the general provisions of N-313, N-331, N-332, and N-511 of Section III of the ASME Boiler and Pressure Vessel Code. For the surveillance capsule pulled in 2011, per BWRVIP-271/NP, the Charpy impact tests were conducted in accordance with ASTM Standards E185-82 and E23-02. Prior to the Summer 1972 Addenda of the 1971 ASME Section III Boiler and Pressure Vessel Code, impact testing was not required on materials with a nominal section thickness of 1/2 inch or less. However, this 1/2 inch thickness exclusion was increased to 5/8 inch by the ASME Boiler and Pressure Vessel Code, Section III, 1971 Edition, Summer 1972 Addenda. Therefore, after issuance of the Summer 1972 Addenda, impact testing is not required on materials with a nominal section thickness of 5/8 inch or less. The welding procedures used were qualified by impact testing of weld metal and heat affected zone to the same requirements as the base metal in accordance with N-541. B. Impact tests were not required for the following:

1.

Bolting, including nuts, 1-inch nominal diameter or less,

2.

Bars with a nominal cross-sectional area not exceeding 1 square inch,

3.

Materials with a nominal (section) wall thickness of less than 1/2 inch or 5/8 inch (refer to paragraph 4.2.4.10.A),

4.

Components including pumps, valves, piping, and fittings with a nominal inlet pipe size of 6-inch-diameter and less, regardless of thickness, and

5.

Consumable insert material, austenitic stainless steel, and nonferrous materials. C. Impact testing was not required on components or equipment pressure parts having a minimum service temperature of 250F or more when pressured over 20 percent of the design pressure. Example: Steam line is excluded from brittle fracture test requirement since the steam temperature will be over 250F when the steam line pressure is at the 20 percent design pressure.

BFN-27 4.2-10 D. Impact testing was not required on components or equipment pressure parts whose rupture could not result in a loss of coolant exceeding the capability of normal makeup systems to maintain adequate core cooling for the duration of a reactor shutdown and orderly cooldown. E. These criteria apply to components and equipment pressure parts, including flange bolts of the reactor coolant pressure boundary, and do not apply to related components such as anchors, anchor bolts, hangers, suppressors, and restraints. All components for the Browns Ferry plant were designed and fabricated giving consideration to brittle-fracture control requirements as stated above. However, these specific conditions were not a part of the initial Browns Ferry Units 1 and 2 plant requirements, and due to the status of fabrication on two items, the requirements could not be imposed without scrapping all materials. On Browns Ferry Units 1 and 2 these two items are: (1) feedwater piping through the second containment isolation valve, and (2) the 14-inch HPCI testable check valve (HPCI pump return into feedwater pipe outside the containment). Charpy V-notch impact tests were performed on these items where possible, and results indicate they generally would not meet the conditions under A, above, if they had been imposed. 4.2.5 Safety Evaluation The reactor vessel design pressure of 1250 psig is determined by an analysis of margins required to provide a reasonable range for maneuvering during operation, with additional allowances to accommodate transients above the operating pressure without causing operation of the safety valves. The design temperature for the reactor vessel (575F) is based on the saturation temperature of water corresponding to the design pressure. To withstand external and internal loadings while maintaining a high degree of corrosion resistance, a high-strength, carbon-alloy steel is used as the base metal with an internal cladding applied by weld overlay to the cylindrical shell and bottom head. Use of the ASME Boiler and Pressure Vessel Code, Section III, Class A, pressure vessel code design criteria provides assurance that a vessel designed, built, and operated within its design limits has an extremely low probability of failure due to any known failure mechanism. The reactor vessel is designed for a 40-year life and will not be exposed to more than 1 X 1019 nvt of neutrons with energies exceeding 1 MeV. The reactor vessel is also designed for the transients which could occur during the 40-year life as indicated below. TLAAs have been identified and evaluated for the reactor vessel 60 year operating life. Summaries of these evaluations for the reactor vessel life are provided in Appendix O, Sections O.3.1 and O.3.2.

BFN-27 4.2-11 No. of Type of Cycle Cycles Boltup 123 Design hydrostatic test at 1250 psig 130 Startup (100F/hr heatup rate) 120 Daily reduction to 75 percent power 10,000 Weekly reduction to 50 percent power 2,000 Control rod worth test 400 Loss of feedwater heaters (80 cycles total) Turbine trip at 25 percent power 10 Feedwater heater bypass 70 Scram (200 cycles total) Loss of feedwater pumps, isolation valves close 10 Turbine trip, feedwater on, isolation valves stay open 40 Reactor overpressure with delayed scram, feedwater stays on, isolation valves stay open 1 Single safety relief valve blowdown 2 All other scrams 147 Improper start of cold recirculation loop 5 Sudden start of pump in cold recirculation loop 5 Shutdown (100F/hr cooldown rate) 118 Hydrostatic test at 1563 psig 3 Unbolt 123 Stress analysis and load combinations for the reactor vessel are evaluated for the cycles listed above, with the conclusion that ASME code limits are satisfied. The details of assumed loading combinations are described in Appendix C for Class 1 equipment. It is possible that the specified number of cycles for some of the events listed above may be exceeded over the life of the plant. A plant procedure has been implemented at Browns Ferry to maintain surveillance on the number of cycles which have occurred and the resulting fatigue usage factors. When the fatigue usage factor reaches a value of 0.7, the procedure requires a reevaluation to be completed in a timely manner to assure that the allowable fatigue usage factor of 1.0 is not exceeded. Operating limits on pressure and temperature during inservice hydrostatic testing were established using as a guide Appendix G to the ASME Boiler and Pressure Vessel Code, Section III, 1971, which was first added to the code in the summer 1972 addenda. The intent of Appendix G is to set criteria based on fracture toughness to provide a margin of safety against a nonductile failure. The resulting operating limits ensure that a large postulated surface flaw, having a depth of one-quarter of the material thickness and a length of one and one-half of the material thickness, can be safely accommodated in regions of the reactor vessel shell remote from discontinuities. Operating limits on temperature and pressure

BFN-28 4.2-12 when the core is critical were established by using 10 CFR 50, Appendix G, "Fracture Toughness Requirements," paragraph IV.A.2.C. The 1998 Edition of the ASME Section XI Boiler and Pressure Vessel Code including 2000 Addenda was used in the development of the Unit 1 P-T curves. The P-T curve methodology includes the following: 1) the use of K1C from Figure 4200-1 of Appendix A to Section XI and 2) the use of the Mm calculation in the ASME Code paragraph G.2214 of Appendix G to Section XI for a postulated defect normal to the direction of maximum stress. An exemption from specific requirements of 10 CFR Part 50, Appendix G is taken by use of ASME Code Case N-640 for Unit 2 and Unit 3. ASME Code Case N-640 permits the use of an alternative reference fracture curve K1c for RPV materials for use in determining the PT limits. The PT limit curves based on the K1c fracture toughness curve enhance overall plant safety by minimizing challenges to operators since requirements for maintaining a high vessel temperature during pressure testing are lessened. ASME Code Case N-588 methodology was also used as a basis for the PT curves. This code case permits the use of an alternative procedure for calculating applied stress intensity factors during normal operation and pressure test conditions due to pressure and thermal gradients for axial flaws. This methodology is incorporated into the ASME, Section XI Code, 1995 Edition, 1996 Addenda, which is the current code of record for the Unit 2 inservice inspection program. Since Unit 3 uses an earlier code of record for the inservice inspection program, Unit 3 implements the requirements of only the 1995 Edition, 1996 Addenda of ASME Section XI, Appendix G to allow the use of the ASME Code Case N-588 methodology for PT curves. The operating limits are provided in the technical specifications for Browns Ferry. For the purpose of setting these operating limits, the initial RTNDT (nil-ductility reference temperature) was determined from the impact test data taken in accordance with the requirements of the code to which the reactor vessels were designed and manufactured. The maximum NDT temperature allowed by the vessel specifications was 40F. Although test data on beltline base material show lower NDT temperatures, an assumed RTNDT of 40F was used in the vessel beltline area, as well as the areas remote from the beltline because the generally accepted NDT temperature for electroslag welds used in the beltline longitudinal seams is 40F. The current operating limits on the pressure/temperature (P/T) curves in the technical specifications are based on the following (RTNDT) values. Unit 1 has used 23.1F for the (RTNDT) value, Unit 2 has used 23.1F for the (RTNDT) value, and Unit 3 has used 23.1F for the (RTNDT) value. For the current P/T curves, fluences were conservatively calculated for licensed operating periods of 38 EFPY for Unit 1, 48 EFPY for Unit 2, and 54 for Unit 3. These periods reflect 60-year reactor pressure vessel operating life and a conservative period of plant operation at 3952 MWt power level. The higher fluence was used to evaluate the vessel against the requirements of 10 CFR 50, Appendix G in accordance with Regulatory Guide 1.99, Revision 2. The end-of-life shelf

BFN-27 4.2-13 energy was evaluated by an equivalent margin analysis (EMA). The results of these evaluations indicated that: (a) The results of the upper shelf energy EMA for limiting welds and plates for the three vessels remain less than the acceptance criterion in all cases. (b) The effective full power year (EFPY) shift is slightly increased and, consequently requires a change in the adjusted reference temperature (ART), which is the initial RTNDT plus the shift. The beltline material ART will remain within the 200 degree screening criterion. In addition to the minimum requirements of the ASME Boiler and Pressure Vessel Code, the following precautions were taken and tests made either to ensure that the initial NDT temperature of the reactor vessel material is low or to reduce the sensitivity of the material to irradiation effects.

a.

The material was selected and fabrication procedures were controlled to produce as fine a grain size as practical. It is an objective in fabrication to maintain a grain size of 5 or finer.

b.

Drop weight impact tests were performed on each heat and heat treatment charge of all low-alloy steel-plate material in its "as-fabricated" condition.

c.

Drop weight impact tests were made on the weld metal, the heat-affected zone of the base metal, and the base metal of the weld test plates simulating seams. If different welding procedures were used for nozzle welds, drop weight tests of similarly prepared coupons were made. The NDT temperature test criteria for the weld and heat-affected zone of the base material are the same as for the unaffected base metal.

d.

The actual NDT temperature of the plates opposite the center of the reactor core was determined. In other areas it was sufficient to demonstrate that the two drop weight test specimens did not break at 10F above the design NDT temperatures. The area of the vessel located opposite the core was fabricated entirely of plate and was not penetrated by nozzles, nor were there any other structural discontinuities in this area which would act as stress risers. The reactor assembly is designed such that the average annular distance from the outermost fuel assemblies to the inner surface of the reactor vessel is approximately 80 centimeters. This annular volume, which contains the core shroud, the jet pump assemblies, and reactor coolant, serves to attenuate the fast neutron flux incident upon the reactor vessel wall. Using assumptions of plant operation at 3440 Mw(t), 100 percent plant availability, and 40-year plant life, the neutron fluence at the inner

BFN-27 4.2-14 urface of the vessel was calculated to be 3.8 X 1017 nvt for neutrons having energies greater than 1 MeV. The results of the analyses of the vessel wall neutron dosimeters which were removed from the Browns Ferry reactor vessels at the end of the first core cycle indicated that the neutron fluence at the inner surfaces of the vessels at the end of 40-year plant life would be 1.56 x 1018, 1.34 x 1018, and 1.31 x 1018 nvt for Units 1, 2, and 3, respectively. These results ranged from 3-1/2 to 4 times the calculated fluence of 3.8 x 1017 nvt. Thus, additional analyses were required to predict the shifts in RTNDT based on fluence obtained from the analyses of the vessel wall neutron dosimeters. The procedures in Regulatory Guide 1.99, "Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," Revision 1, April 1977 were used to predict the RTNDT shifts. Response to Generic Letter 92-01 provides updated fluence data. TLAAs have been identified and evaluated for the reactor vessel 60 year operating life. Summaries of these evaluations for the reactor vessel life are provided in Appendix O, Sections O.3.1 and O.3.2. Quality control methods were used during the fabrication and assembly of the reactor vessel and appurtenances to ensure that the design specifications were met. The fabrication test program was carried out by the reactor vessel vendors on material representative of the formed, heat-treated, and fully fabricated vessel. Tests of base metal and welded joint were performed, and the results were reported during the early stages of vessel construction. Tensile specimens (of 0.505 inch in diameter) from the shell plate material were prepared for various thickness levels of the plate material. These specimens were tested at various temperatures per ASTM Specifications E8 and E21 to determine tensile strength, yield strength, elongation, and reduction of area. Tensile specimens whose gauge diameter is at least 80 percent of the reactor vessel wall thickness were prepared from base metal and weld material. These specimens were tested at room temperature per ASTM Specification E8 to provide stress-strain curves, tensile strength, yield strength, elongation, reduction of area, and macrophotographs of the breaks. Charpy V-notch impact specimens were prepared from base metal and tested per ASTM Specification E23, Type A, to establish curves for determining the transition temperature at which 30 ft-lb of absorbed energy result in ductile fracture for various thickness levels of the plate material. The Reactor Coolant System was cleaned and flushed before fuel was loaded initially. During the preoperational test program, the reactor vessel and Reactor Coolant System were given a hydrostatic test in accordance with code requirements at 125 percent of design pressure. The vessel temperature is maintained at a minimum of 60F above the NDT temperature prior to pressurizing the vessel for a hydrostatic test. A hydrostatic test at a pressure not to exceed system operating pressure is made following each removal and replacement of the reactor vessel head. Other preoperational tests included calibrating and testing the reactor vessel

BFN-27 4.2-15 flange seal-ring leakage detection instrumentation, adjusting reactor vessel stabilizers, checking all vessel thermocouples, and checking the operation of the vessel flange stud tensioner. During the startup test program, the reactor vessel temperatures were monitored during vessel heatup and cooldown to assure that thermal stress on the reactor vessel was not excessive during startup and/or shutdown. The average rate of reactor coolant temperature change during normal heatup and cooldown is limited to 100F in any 1-hour period. Only during some postulated events, or in local areas, would this rate of fluid temperature change be exceeded as a result of rapid blowdown, valve operation, or rupture accident. 4.2.6 Inspection and Testing The inservice inspection and testing program for the reactor vessel and appurtenances is outlined and detailed in Subsection 4.12. Extent and areas of examination, inspection methods, and frequency of examination are established therein. The surveillance test program provides for the preparation of a series of Charpy V-notch impact specimens and tensile specimens from the base metal of the reactor vessel, weld heat-affected zone metal, and weld metal from a reactor steel joint which simulates a welded joint in the reactor vessel. The reactor vessel material surveillance program is described in report NEDO-10115, Mechanical Property Surveillance of General Electric BWR Vessels, by J. P. Higgins and F. A. Brant. It describes the specimens, specimen inventory, capsule design, associated equipment, material selection and instructions for handling the specimens. All the requirements of paragraphs 3.1 through 3.3 of ASTM E-185-66 are met. All the requirements of paragraphs 4, 5, 6, 7, and 8 of ASTM E-185-66 are met, except that thermal control specimens discussed in paragraph 4.3 are not used. NEDO-10115, paragraph 5.7 states, "Because the BWR is a constant-temperature device, no special temperature monitoring devices are required." It is felt paragraph 4.3 of E-185-66 is a recommendation rather than a requirement. The vessel surveillance samples were prepared in accordance with GE purchase specification 21A1111, Rev. No. 9, Attachment B.

BFN-27 4.2-16 The NDT temperatures for the three core region plates were as follows. Heat No. Plate No. NDTT (F) C2884-2 6-139-19 0 C2868-2 6-139-20 0 C2753-1 6-139-21 -20 The two test plates furnished by Babcock & Wilcox under the requirements of paragraph 3.1.1 of attachment B to specification 21A1111 were fabricated from Heat No. C2884-2 and C2868-2. The two plates were electroslag-welded (B&W Weld Procedure WR-12-4) and heat-treated the same as the core region plates. Tensile and Charpy impact specimen samples were removed as indicated in Figures 3, 4, 5, 6, and 7 of attachment B to 21A1111. (See FSAR Appendices J, K, and L.) The surveillance test plate 610-0127 was 139 in. long and 60 in. wide, and all excess material is under TVA control in the event that additional material is needed. It is estimated that enough extra material is available for several hundred additional Charpy specimens. No weak direction specimens were included in the reactor vessel material surveillance program. All Charpy V-notch specimens were taken parallel to the direction of rolling. The majority of developmental work on radiation effects has been with longitudinal specimens. This is considered the best specimen to be used for determination of changes in transition temperature. At the low neutron fluence levels of BWR plants, no change in transverse shelf level is expected and transition temperature changes are minimal. The specimens and neutron monitor wires were placed near core midheight adjacent to the reactor vessel wall where the neutron exposure is similar to that of the vessel wall (see Subsection 3.3). The specimens were installed at startup or just prior to full-power operation. For Units 1, 2, and 3, Integrated Surveillance Program (ISP) implementation and surveillance specimen schedule withdrawal and testing is governed and controlled by BWRVIP-86 Revision 1-A, the BWRVIP responses to NRC RAIs dated May 30, 2001, December 22, 2001, and January 11, 2005, and the NRCs Safety Evaluation dated February 1, 2002. (NOTE: WRVIP-86, Revision 1-A, was approved by NRC and issues in October 2012, superseding both BWRVIP-86-A and BWRVIP-116.) Surveillance and chemistry data for all representative materials in the BWRVIP ISP have been consolidated into BWRVIP-135 {Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations.} A test specimen surveillance capsule (the second set of Unit 2 test specimens located at Azimuth 120º) was withdrawn in accordance with the ISP in 2011 during Unit 2 Refueling Outage 16 (U2R16) at approximately 23 EFPY of operation. An additional test specimen surveillance capsule is scheduled for withdrawal during the license

BFN-28 4.2-17 renewal period, this being the third set of Unit 2 test specimens located at Azimuth 300º, which are currently scheduled for removal in the refueling outage closest to without exceeding 40 EFPY of operation. At the present time, this would correspond to Unit 2 Refueling Outage 24 (U2R24) in 2027. Presently, there are no plans to withdraw any capsules from either Unit 1 or Unit 3, as per BWRVIP-135, the BFN Unit 2 capsules provide the best representative plate material for all three units and the best representative weld material for Units 2 and 3. Supplemental Surveillance Program (SSP) Capsules A, B, D, G, E, and I provide the best representative weld material for Unit 1. Test results will provide the necessary data to monitor embrittlement for Units 1, 2, and 3. Since the predicted transition temperature shift of the reactor vessel beltline steel is less than 100F at end-of-life, the use of the capsules per the ISP meets the requirements of 10 CFR 50, Appendix H, and ASTM E185-82. Revisions to fluence calculations using data obtained from the surveillance capsule specimens will use an NRC approved methodology that meets Regulatory Guide 1.190. {By letter dated August 14, 2008 (EDMS Number L44 080828 014), NRC issued License Amendment 273 for BFN Unit 1, and by letter dated January 28, 2003 (EDMS Number L44 030204 001), NRC issued License Amendment Numbers 279 and 238, for BFN Units 2 and 3 respectively, authorizing adoption of the BWRVIP Integrated Surveillance Program to address the requirements of Appendix H to 10 CFR Part 50.]

BFN-26 TABLE 4.2-1 REACTOR PRESSURE VESSEL MATERIALS Component Form Material

Spec. (ASTM/ASME)

Heads, Shell...................... rolled plate low-allow steel SA-302 B cc 1339 Closure Flange.................. forged rings low-alloy steel A-508 CL 2 cc 1332-2 Cladding............................ weld overlay austenitic SA-371 type ER309-type stainless steel ER308 (and carbon content -inconel <0.08 w/o)-Inconel 82 and 182 Nozzles.............................. forged shapes low-alloy steel A-508-CL2 cc 1332-2 Control Rod Drive.............. forged tubes Inconel SB-166 cc 1336 Stub Tubes Control Rod Drive.............. pipe austenitic Housings stainless steel In-Core Housings.............. pipe austenitic stainless steel Vessel Supports-............... rolled plate low-alloy steel SA-302 Gr.B External Shroud Support-................ forging Inconel SB-168 Annealed cc 1336 Internal Nozzle Safe Thermal......... pipe austenitic SA-312 TP.304 Sleeves stainless steel Nozzle Safe Ends.............. forging austenitic SA-336-F8/F8M stainless steel and some low-carbon SA-105-2 cc 1332-1 steel Nozzle for Instrument........ forging Inconel SB-166 cc 1336 para. 1 Penetrations

cc - Code Case that modifies/augments the material specification.

BFN-26 Table 4.2-2 Table 4.2-3 REACTOR VESSEL DATA REACTOR VESSEL ATTACHMENTS Reactor Vessel Qty. Inside Diameter, in. (min.).............................. 251 3/8 in. Internal Attachments Inside Length............................................ 73 ft 11-1/2 in. Guide Rod Bracket......................................... 2 Design Pressure and Temperature, Steam Dryer Support Bracket......................... 4 psig @ °F................................................... 1250 @ 575 Dryer Holddown Bracket................................. 4 Feedwater Sparger Bracket.......................... 12 Vessel Nozzles (number and size) Jet Pump Riser Support Pads....................... 20 Recirculation Outlet................................ 2-36 in. to 28 in. Jet Pump Diffuser Bracket............................ 20 Steam Outlet....................................................... 4-26 in. Core Spray Bracket........................................ 4 Recirculation Inlet............................................. 10-12 in. Feedwater Inlet................................................... 6-12 in. External Attachments Core Spray Inlet................................................. 2-10 in. Stabilizer Bracket............................................ 8 Instrument (one of these is Head Spray)**......... 2-6 in. Top Head Lifting Lug....................................... 4 Control Rod Drive............................................. 185-6 in. Insulation Supports......................................... 2 Jet Pump Instrumentation..................................... 2-4 in. Insulation Support Brackets...... 12 ea; 2 places Vent...................................................................... 1-4 in. Thermocouple Pad....................................... 36 Instrumentation..................................................... 6-2 in. Control Rod Drive Hydraulic System Return *...... 1-4 in. Core Differential Pressure and Liquid Control...... 1-2 in. Drain..................................................................... 1-2 in. In-Core Flux Instrumentation.............................. 55-2 in. Head Seal Leak Detection.................................... 2-1 in. Approximate Weights (in pounds) Bottom Head...................................................... 207,500 Vessel Shell........................................................ 842,000 Vessel Flange..................................................... 106,000 Support Skirt......................................................... 28,000 Other Vessel Components................................... 65,500 Total Vessel without Top Head........................ 1,249,000 Top Head1.......................................................... 252,000 Total Vessel..................................................... 1,501,000 1This weight includes 60,095 lbs. which is the weight of the reactor pressure vessel studs, nuts and washers.

CRD Return Line nozzle is capped.
- Unit 1, 2, and 3 head spray nozzle line is capped.

I l

u I

~

?! l I

Ii') z C: 0 lJ l:l mar Ni l ' ... ~*

i..

r iril!IIA~ SLEi~E \\ ~£** Al,£ S~ALL ~OT . [tCcE021KJ GPII n,s, IP UAVE 5QIIARE r ElrO fOII fRIIUllhG / IN flflll I -~EDIATEII NOZ~£ ~TAD\\. FLO* _uwio co ,,rf* \\ ,- ~A~ ~,,:NLESS StEEL f'lllYIOE TRANSITION 10 TYi'£ lOUTAlllUUTEEL FOR lltELDIIIG 8Y DTll!IS

.r;.tilfJ =k t~\\' _;=.;; _--_. **--

1;'.E=s,~

!! 11111 z.a

>i ,..z IC"' mm ....,a -<,a zZ >c rO -< r-t11ffl tll ::a

a m"'II

,,r-o> ,a :z: .... -4 ~ ~ 2,. SCSHTA':1[~~ nm I Rtfl ~- ,IJUL CORE OIFf PRESSURE PIEPARlD fOA?'" -!000 lb SOCJIETWELD FITTING COltE lllFFER£NTIAl PR[~RE AND u,,iti10 col{'rliol flilzZ((""" ~ -~V!Offl L[AVf" < _____.L 3 EWALLY SPACE.!2..J C£1i1Elllu AID SUPPORr PADS R£ClltCULATIDII IKL[T NOZZLE IIIZZL£ --*~*--,.---+- l'ELD B't OTIIEIIS L3£QUALlY SPACED ZE.RO LEAIIAGE CENTERIIIG A,NO 51Jf'PORT REQUIRED PADS CJIR~ SP~~YJ!OZZL E LEAVE SQUARE DD FOIi TRIIIIING II FIELD ~BIiii ..1*~ IICIINEL I"! ;- ! 10111 1911 II~ AINllLAR 1' MIii R.O* GM'

~
CL*D0111t:. 'l'Ti 1 CLAOOIN:ll --2DDll+O~ I -0001 T~ERIIAL SLEEVES CON* ff
fSSEL BORE s1sr111G or 3 CONCtNTR~C

I

TP l04 SST SLEEIIES
ILl':ft:.* OIA 1Rfrl
, L,_ .JOOO l 98~-<' 110i DIA IIW:F 1 THERMAL SLt!Wt AS REQUIRED ---, TO P80TECT VESSEL JIOUlE HD INTERNAL PIPE CGN*ECHIII HQIIRLY SPACED CE~ TERlll(I AIIII SUPl'Ollf PAllS CONl'IIOL ftQ_I) DRIV! HY1!!!,W!.IC ~TEii RETURN NOZZLE ~IOINIIIG Pd'E, d I ' J l I l' NOZZLE
-T-*
l UA~
c:i.,..,,
"w.,N*+ t ~- ...,.~t.J \\ '\\.:_ VESIEL WALL RECIRCULA RON 'lUfLE ~ IIIIZlLt_ UIUU!l!!H IN INCHES _CJIIIT!IOL ll!)D DRIVE tEJ!WJ6!11!1 I~~'!!, _ Pf/lETI!All!ft
BROWNS FERRY NUCLEAR PLANT FIN... LS... FETYANALYSISREPORT
BROWNS FERRY NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT REACTOR VESSEL FIGURE 4.2-4
I ~ "II a:11 I
l!Q fo c,iz
~ > Cl' l "ll.... mm I Q 2 -<,a C 0
x, i m
zZ ,..c ~ l ,-n <<r-N Coll m itm/t.1f:.fE.!u SUPPORT PAO$ RlCll!tlll~ llllh lttU:T HOULE 7 ~ :io i
a "'D ffl,..
111 ~~ f ! ~... - LEAVE 5QUARE r ENO fOlftl!IIUI.J!IC IN flELII I l'a 11,MIN IIOUU -*~*---r---t-l[LO BV OT11915 Ls EQUALl 'I' SPACED ZERO LEAKAGE CEUtRlllG MIi S\\JPPQRl lf.QUll(D p ADS CJ)~ IP'!~f..!OUL E
~..col!l.
.l'f!IETIW!!N_ lEAYESQU.al EIIOFOlllRlallO IIIFIELO T8EIIMAI. :11.EEV~ A~ REOIIIR£0 ---1 TQ PIIOTEtT ¥ESSf.L lll!UU IID INTERNAL fl!'£ l:GWIU:10! i --~J t_ -~:j*~~ l, Ki.1n'-f ~~.'>,.~ ~;., w-l \\ '-. 'ff.Slit!. WALL RWffCIILA TillN 'lllfLl 1'. IIIIULC_
C D E r J ,._ 900 "" Bl.IND Fi.ANG£. FOR MK. NO SEE. TVA DWG 47W452-8 I I I, r.,,
--~-'*'-------;,,--*- ~~~--.. !J.1-.... ~----~-
' \\ \\ rr \\ \\--~,,,-- ~ T'"'" ~-- 11:9 s REACTOR .- C:.* I!. ~~U~Er~ l~J~!i'm. I :...~a:.. ~~ol~l;\\.I tli/~L~~H~~~~ * ~ FUEL [N THE cu*ot:nT C<IIE L!)Jl). HI. FOR ADDITIONAL JET PLIIP DETAILS REFEREICE THE L-FQ.LOWING DRAIINQS PER GE CONTRACT ODOD1704: 124D1567-SERIES... SLIP JOINT CLAMPS 124D1243-SERIES,,,SLIP JOINT CLAMl'S 124D2088-SERIES... SLIP JOINT CLAlll'S 124D1566-SERIES... AUXJLIARY IEOOES 177D4693,,,,,,,,,,AUXILIARY IEOOES 23'1C7181-SERIES,,,INLET-UIXER MAIN WEDGE 234C5141.......... INLET-MIXER IAAIN WEDGE 177D4867,,,,,,,,,,INLET-UIXER MAIN WEDGE 124D1029,........, INLET-MIXER MAIN WEDGE ALSO SEE 0-730EBB3. TAII..E 2. FOR JET PUMP IAODIFICATION COMPLETION STATUS. 14" ii1 ~~'fi.Jl~~I:. ~~Esi~~ Lms1:D~i~4) PU~S 1.2,3,4,5,11,12.13,14 AIII 15. Hi. UNIT 3 STEAM DRYER REPLACEt.ENT JS PERFORt,ED l PER GE HITAQU BROINS FERRY REPLACEMENT STEAM DRYER INSTALLATION SPECIFICATION 002N2886 AIII DRAWINGS 105E5413 Alll 105E5414 SHEETS 1 THROUGH 10, CONTRACT ND. 8871. L-L- AMENDMENT 28 .r l...'.'..U:.:.NI:..T:.......:::3:__ ______________ --i" BROWNS FERRY NUCLEAR PLANT L-FINAL SAFETY ANALYSIS REPORT
SOOll C 0 E 2 91 01.lTe:R ---..........._ 0' RING ( INNER
o* RING "O" RING GASKET (SEC owe 0-122879C) 1MO.JIE m ~~~~ vr-cMo.
G'-900 LB-BLIND flANQE, IDi,ND. 4714~2-921 llJMI0.'11. n,.,g;i:,s W,T* OOMfl. OLS*>o,$', ~ "'U:~,:,:,M",:olO;P. ~"""" - """""""" ~"'-E..,., ,43al.OIO- ~."JS'"""'l'.P P'!CMlll/,~O<-QF 1111.,--:,u. n"""" ~ ~~ ~ ~= ~ = =* = RETAINER RING CL IP I rr*-;,i,,u-.fW~'* ~* ~~* ~.. ./ \\ \\-* SEE NOTE 19-~~ 0 ,o 18 II Ii ~ 1a ~r-,e~- "W'S 00<L~ "' -.,_.,,,. ='-~"""" ~<> u<S~~~ ""- lft~T "!.-.<;,;- ""'~'TIO" ~ ~*Jl<3 ~ti.U.TIIJ.>;UI ooi >&\\IT, 90 \\ ---j Cut 'THiRPW. !i....t.Ni. '10 Frt f-.<<l.O l"aP PU,U,.T D-'O ,i.11)2:32. - ~~ o ~ a ~ UIC) 0, s.n.va. n.CM m 1z.*sJo;.,,* "1"f9 WPL ~ ~ ~ SECT C.*C i i I i 1----,u i !il--*!11
I I
~d fl ~L-~EF --~ ~ii """'!UL ~ = ~~ 1!; ~ " ~\\>'~i I ~ ~~ 0 CF = ~
I
~ --- 0 ~ - ~,~~ t ~ - ~ L!...o... 1,._ *t) i =rJON A-Ar.1..-r..,,....,1 !-f04R935.. IS 17 19 ,o J; ~ " ""m~ .. ~ - ~ N
g
~ N ~~ N:JT[ 1:----..,.._:_~-, NOTE 13 182 PLOO ALL QYPASS HCLOS 1,IIC Cl~. lltf. ( 129 Pi.CS ro, 111. l - -- I = Iii DET >t:1, \\ 0 u~ .12 R Ml:111, JL_ " .rrr**-
l
!YP 5S ~L ~:n: s L-L REa\\C~O~ '/ES>;~- SvPfOR'"*Rl:F SEE =~ 73CC844 srn:£s ~~ tF.1/irm,l
10 J
l-!04R935 = Cllll.1111111.t.. t EM,>:BGED 1/IEW E rsvu) -VP 24 "~ ENIJtSGEC> V1ffl TYP 185 PL TT RV:SiEQ VIEW I(~... ~~ f'l'P 43 PL [Srl 2, SECT J-D) 40 aon
I - --
NOTE'S' "'" ASSEMBLY REACTOR !-104R935 Clllf.811 111.t ** I
a. roll :>Eo.C."tQI>.,....,~.... '-**no,1. >>v,.~,..co.,.
-=~ ~e:...,_ c;-,-,::,,,u,1.*.i<>1
z........_ ~J.!O.'I:' -~~
~...,,;: " ~o.E~,,_ I".!: 1.-1 >A '-!,TI:()... ~1 l. 00 l~ M"'l,T~ ~1',fl.l,a. L*'l>f- \\. --.oos,,.;1. ~ s~f't'OIQ" *~acs "'IO._,-!~ "'* s!Y\\>. ""'.~.. ~lOi<T =- ;,t< T_,..._...,.. I,.....-T co,<.r. W """" (c.>1,..,....,,,1: ~~" or..._,. ~u~rLC.!.~- ~ 0.U. D;CE.'l.5 LVes!IC.\\NT IWTUI. r=,.,>;.G. ..... !, """'-=="'--=ii.'!.. "WO ,n<. ~-"""'.,,~ -~....-.......an.,.,. O'T>,£;::,.,,;. 11.. - '"SU::. d,,-... ~rtK. '5"!~ <;J~t.........,.... ~-~... 6,. "'"' ""°B.~,.,.... c..._.;. '!'Q !&,-~-".. "T. "'4i'!,W~;'l,~~~,_.,..~s:=i... It,. :fE.AC"!OII' Y!tJ "'5Tli' Dall; 5e; Oil"/,.,:-,1-rTli
9. IE,l,C~Sfl,f"TCST "'*STI<:;ING SE£ W'I,. Z*l-~

10. PUIDS TD E LOCATED Al CIIE l'OSlTICl!NS IIS WUCATID 24->> I IJO' 11UU! UICAJICII IIASlD al ll£AC1tll VEISEl. AZIIIJT15J, n-4! 1270' lllL£ LacATI<IIII! BASED OIi IIUIC'RI!
VESIIEL AZDIIIHSJ. 11
JtT 111.11P !IISft'l.l<<l!T SUI. SAF[ El!II AS PER DIIAl'DGS l-47E408-3, -4.
12-fir D.EVA.f!OI SIIJ!III IS nil nlE Ifl!Tlid.. FLIIEI. IIIIDLU. -'C'lllM. ti# t1.EVI.TJOllc MY VM'f OEl'DIDIIC O!i! TtlE T'l'I'£ OF Ftlll II! n£ CURRHT I.I. IIS:lElilLlltltiN I.. ET A Olffl.£T SAi"£ SIDS AS l'P DltUllll; 1-47£~.-c.-7.
14. J£T Pl.WP SIEXSIIIC Ult£ Cl.MI'S IDIG 11010M)
AIIE tNlitAU.EO IW '!ti£ ISEN&DIG LIIEI a, ~ f'IIIPS 1. 2. 5. 4, ;i;, 11, 12, 15, H NID 11. Hi. DRYER LUG MODIFICATIONS ARE PERFORMED IN ACCORDANCE 11TH GE INSTALLATION SPECIFICATION, 28A8891 AND DRAWING 2J4CS566.
18. JET PUMP AUXILIARY SPRING WEDGES ARE INSTALLED AT LOCATIONS JP-02 VS, JP-OJ VS, JP-04 VS, JP-08 VS, JP-D7 VS, JP-10 SS, JP-12 VS AND JP-14 VS (JP-JET PUMP, VS-VESSEL SIDE. SS-SHROOD SIDE) IN ACCORDANCE WITH GE INSTALLATION SPECIFICATION 28A8483 AIII DRAWING 12401566 C
0 E (SHEETS 1 TI<<OUGH 4). REcnt.lENDATIONS FOR J' 1/18~/H BE:FORE BENDING 1 ~AFTER r-**-- ~--+-'" """"" o:::-=.:._k-""1,c;,11----J---" CONTROL ROC DRiV!:,.;yo ~ SYSTEM Pl?'NG ARRAiliGEMtNT SC:E DWG 729E:55 SERfES (REF) INSPECTION AFTER OPERATION ARE PROVIDED IN GE Doet.llENT GE-NE-0000-0038-8778.
17. CORE SPRAY T-BOX MODIFICATIONS ARE PERFORIIE:D IN ACCalDANCE Willi GE INSTALLATION SPECIFICATION 26A69J4 AND DRAWINGS 124D2158 AIII 124D2157 (SHEETS 1 TI<<OUGH 5). REcnt.lENDATIONS FOR INSPECTION AFTER OPERATION ARE PROVIDED IN GE DOCUl.:NT GE-NE-OOOO-OOSJ-8742, 18, UNIT 1 STEAM DRYER REPLACEMENT JS PERFORIIE:D PER GE HITACHI BROWNS FERRY REPLACEliENT STEAM DRYER INSTALLATION SPECIFICATION OD2N2886 AND DRAWINGS 105E541J AND 105E5414 SHEETS 1 lliROUGH 10.
catTRACT NO. 6871.
19. UNIT 1 STEAM DRYER MODIFICATIONS ARE PERFORIIE:D IN ACCalDANCE Willi AREVA INSTALLATION SPECIFICATION 9D888J28-000 AND DRAWINGS 602908JC, 8D29D84C, 8029085C-1 A -2 AND 80290920-1 A -2 (CONTRACT NCI, 00020538, SUP, NCI, 10).
~'7'.:~*-"' AME NOME NT 28 PROJECT BFNP DISCIPLINE MECH CONTRACT l!QZH UNIT 1 DESC. 6~~y B~M. !t,!l;;L~6B aQIL~B DIC/DOC NO. l-l!HB9J:~ SHEET ___.l__Of'--REV !l!l5 DCN DATE REACTOR BLDG UNIT 1 BROWNS FERRY FINAL SAFETY NUCLEAR PLANT ANALYSIS REPORT REACTOR VESSEL FIGURE 4.2-4 H
BFN-28 4.3-1 4.3 REACTOR RECIRCULATION SYSTEM 4.3.1 Power Generation Objective The objective of the Reactor Recirculation System is to provide forced cooling of the core and a variable moderator (coolant) flow to the reactor core for adjusting reactor power level. 4.3.2 Power Generation Design Basis
a.
The Reactor Recirculation System shall provide sufficient subcooled water to the core during normal power operation to maintain normal operating temperatures.
b.
The Reactor Recirculation System shall operate over a flow control range of 20 percent to 105 percent flow to allow power variation.
c.
The Reactor Recirculation System shall be designed to minimize maintenance situations that would require core disassembly and fuel removal. 4.3.3 Safety Design Basis
a.
The Reactor Recirculation System, including the recirculation pump trip (RPT) feature, shall be designed so that adequate fuel barrier thermal margin is assured following recirculation pump system malfunctions and postulated transients (such as turbine-generator trip or load rejection).
b.
The Reactor Recirculation System shall be designed so that failure of piping integrity does not compromise the ability of the reactor vessel internals to provide a refloodable volume. 4.3.4 Description The Reactor Recirculation System consists of the two recirculation pump loops external to the reactor vessel which provide the driving flow of water to the reactor vessel jet pumps (see Figures 4.3-1 and 4.3-2a sheets 1, 2, and 3). Each external loop contains one high-capacity motor-driven recirculation pump and two motor-operated gate valves for pump maintenance. Each pump discharge line contains a venturi-type flowmeter nozzle. The recirculation loops are a part of the nuclear system process barrier and are located inside the drywell containment structure. The jet pumps are reactor vessel internals and their location and mechanical design are discussed in Subsection 3.3, "Reactor Vessel Internals Mechanical Design." However, certain operational characteristics of the jet pumps are discussed in this subsection. A summary of the characteristics of the Reactor Recirculation System is presented in Table 4.3-1.
BFN-26 4.3-2 The recirculated coolant consists of saturated water from the steam separators and dryers which has been subcooled by incoming feedwater. This water passes down the annulus between the reactor vessel wall and the core shroud. A portion of the coolant exits from the vessel and passes through the two external recirculation loops to become the driving flow for the jet pumps. The two external recirculation loops each discharge high pressure flow into an external manifold from which individual recirculation inlet lines are routed to the jet pump risers within the reactor vessel. The remaining portion of the coolant mixture in the annulus becomes the driven flow for the jet pumps. This flow enters the jet pumps at the suction inlet and is accelerated by the driving flow. The driving and driven flows are mixed in the jet pump throat section resulting in partial pressure recovery. The balance of recovery is obtained in the jet pump diffusing section (see Figure 4.3-3). The adequacy of the total flow to the core is discussed in Subsection 3.7, "Thermal and Hydraulic Design." Tests have been conducted and documented1 to show that the jet pump design is sound and that jet pump operation is stable and predictable. The pump is started at slow speed with the discharge valve closed. Pump speed is not increased until after the discharge valve has been opened utilizing the jogging circuit that opens the valve in steps. There is actually a very low probability that a recirculation loop that has been allowed to cool would need to be placed in service again with the nuclear system hot. A valid reason for closing both the pump discharge valve and the suction valve is to prevent leakage out of that portion of the recirculation loop between the valves, e.g., excessive leakage through the pump mechanical seal. A leak of this nature cannot be repaired without shutting the plant down to permit access to the drywell; the nuclear system would in all probability have been cooled prior to repairing the leak. Since the removal of Reactor Recirculation System valve internals requires unloading of the nuclear fuel, the valves are provided with high-quality back seats and trim to facilitate stem packing renewal without draining the vessel and to provide adequate leak tightness during normal operation. The Reactor Recirculation System valves are designed and constructed to meet the requirements of USAS B31.1.0, 1967 edition, with added GE requirements which were implemented in lieu of the outdated B31 Nuclear Code Cases-N2, N7, N9, and N10. The valves are designed to operate under maximum prevailing operating conditions and postulated accident conditions in the drywell. 1 "Design and Performance of G.E. BWR Jet Pumps," General Electric Company, Atomic Power Equipment Department, Sept. 1968 (APED-5460).
BFN-28 4.3-3 The Units 1 and 3 Reactor Recirculation Header equalizer valves were removed during the respective unit recirculation piping replacements in order to reduce the number of welds and, therefore, minimize susceptibility to Intergranular Stress Corrosion Cracking (IGSCC). Under all operating conditions (Unit 2 only), one equalizer valve in the line between the two pump discharge lines shall be open and the other valve shall be closed (both valves having motive power removed). This is to prevent pressure buildup due to ambient and conduction heating of the water between the equalizer line valves. The idle pump loop is not completely valved off if it is desired to return the idle loop to service prior to the next reactor cooldown (such as VFD repair for Units 1, 2, and 3). The recirculation pump casing allowable heatup rate is 100F per hour, the same as the reactor vessel. It is possible to keep the idle loop hot with the equalizer line valved off (Unit 2 only) and the idle loop valves left open, permitting the pressure head created by reverse flow through the idle jet pumps to cause reverse flow through the idle loop. However, it is first necessary to stop the pump rotation by closing either the pump suction or discharge valve until pump rotation stops. Once the oil film is squeezed out of the pump thrust bearing, the pump will not rotate even with both the suction and discharge valves open. Following one recirculation pump operation, an operational restriction is applied such that the discharge valve of the low speed pump may not be opened unless the speed of the faster pump is less than 50 percent of its rated speed. This limitation provides assurance when going from one-to-two pump operation that excessive vibration of the jet pump risers will not occur. The feedwater flowing into the reactor vessel annulus during operation provides subcooling for the fluid passing to the recirculation pumps, thus determining the additional net positive suction head (NPSH) available beyond that provided by the pump location below the reactor vessel water level. If feedwater flow is below 17 percent, the recirculation pump speed is automatically limited. The recirculation pumps can be operated during nuclear system heatup for hydrostatic tests. At this time, they act in conjunction with any contribution from reactor core decay heat to raise nuclear system temperature above the limit imposed on the reactor vessel by nil-ductility transition temperature (NDTT) considerations so that the hydrostatic test can be conducted. A decontamination connection is provided in the piping on the suction side of the pump to permit flushing and decontamination of the pump and adjacent piping. This connection is arranged for convenient and rapid connection of temporary piping.
BFN-26 4.3-4 The piping low point drain is used during flushing or decontamination to conduct crud away from the piping low point and is also designed for connection of temporary piping. Each recirculation pump is a single-stage, variable-speed, vertical, centrifugal pump equipped with mechanical shaft seal assemblies. The pump is capable of stable and satisfactory performance while operating continuously at any speed corresponding to a power supply frequency range of 11.5 to 57.5 Hz. For loop startup, each pump operates at a speed corresponding to a power supply frequency of 11.5 Hz with the discharge gate valve closed. The recirculation pump shaft seal assembly consists of two seals built into a cartridge which can be readily replaced without removing the motor from the pump. The seal assembly is designed to require minimum maintenance over a long period of time, regardless of whether the pump is stopped or operating, and seal over a wide range of pressures and temperatures. The original seal design for Units 1, 2, and 3 has been changed to a seal assembly with an extended design life. Each individual seal in the cartridge is capable of sealing against pump design pressure so that any one seal can adequately limit leakage in the event that the other seal should fail. A breakdown annulus is provided along the pump shaft to reduce leakage in the event of a gross failure of both shaft seals. Provision is made for monitoring the pressure drop across each individual seal as well as the cavity temperature of each seal. Provision is also made for piping the seal leakage to a flow measuring device. Various control room alarms indicate improper seal performance. The Reactor Building Closed Cooling Water System and the Control Rod Drive Hydraulic System provide cooling to the recirculation pump seals. If either one of these systems is operating, recirculation pump operation without the second cooling system may continue with no harm to the seals. If both seal cooling systems are inoperable (e.g., due to a loss of AC power), the pump seals will overheat approximately 7 minutes after the total loss of cooling and seal deterioration may begin. Based on fluid loss analysis of extremely degraded seals, the leakage is less than 70 gpm. This amount of leakage will not lead to a safety concern but may degrade the seals such that they would have to be repaired prior to resuming operation. Each recirculation pump motor is a variable-speed AC, electric motor which can drive the pump over a controlled range of 20 percent to 102 percent of rated pump speed. The motor is designed to operate continuously at any speed within the power supply frequency range of 11.5 Hz to 57.5 Hz. Recirculation pump motors are designed, constructed, and tested in accordance with the applicable sections of the NEMA Standards.
BFN-26 4.3-5 For Units 1, 2, and 3, a variable frequency drive unit located outside the drywell supplies power to each recirculation pump motor. Minimum speed corresponds to a frequency of 11.5 Hz. For Units 1, 2, and 3, the combined rotating inertia of the recirculation pump and motor are modeled consistent with a coastdown of flow following loss of power to the drive motors, so that the core is adequately cooled during the loss-of-power transient. The effective inertia of these devices are specified in the following form, which takes into account the torque and speed conditions on each rotating shaft. J = Time gTo where J = inertia (lb-ft2)
= rated speed (rad/sec) g = gravitational constant (32.3 ft/sec2)
To = torque at rated speed (ft-lb). From this equation, the required inertia (J) is calculated. The recirculation pumps are Classified as machinery, and, as such, are specifically exempt from the jurisdiction of any section of the ASME Boiler and Pressure Vessel Code or of the USA Standard Code for Pressure Piping. The standards of the Hydraulic Institute are the only standards which are applicable; however, they are more pertinent to the testing and performance of the pump and consequently provide little or no guidance in the areas of casing quality and structural integrity. To assure that the pump casing can withstand a pressure equivalent to that inside the reactor vessel, the pump casing is designed in accordance with the ASME Boiler and Pressure Vessel Code, Section III, Class C, as far as this code can be applied. The requirements of Section III of the ASME Boiler and Pressure Vessel Code for Class C vessels (1965 edition) are used as a guide in calculating the thickness of pressure-retaining parts of the recirculation pumps. The casings and forgings are fabricated from austenitic stainless steel. Class C is used because the pump casing does not experience temperature transients as severe as those that portions of the reactor vessel and certain piping connections experience; therefore, it is not necessary to make the cyclic analysis required for Class A equipment.
BFN-26 4.3-6 The design objective for the recirculation pump casing is a useful life of 40 years, accounting for corrosion, erosion, and material fatigue. For the 60 year operating life, the recirculation pump aging effects will be managed using the ASME Section XI Subsections IWB, IWC, and IWD Inservice Inspection Program, Chemistry Control Program, and BWR Stress Corrosion Cracking Program described in Appendix O, Section O.1.4. O.1.5, and O.1.10. Material fatigue for the 60 year operating life has been evaluated as a Time Limited Aging Analysis (TLAA). The summary of this evaluation is provided in Appendix O, Sections O.3.2.3 and O.3.2.4. The pump-drive motor, impeller, and wear rings are designed for as long a life as is practical. The design objective is to provide a unit which will not require removal from the system for rework or overhaul at intervals of less than 5 years. The recirculation system piping is of stainless steel construction and is designed and constructed to meet the requirements of the USA Standard Code for Pressure Piping, Power Piping, USAS B31.1.0, 1967 edition, and the additional requirements of GE design and procurement specifications which were implemented in lieu of the outdated B31 Nuclear Code Cases-N2, N7, N9, and N10. The suction and discharge pipes are welded to the pump casing. The coolant in the nuclear process system is at high pressure and contains a large amount of energy. Substantial failure of the nuclear process system could result in a rapid loss of coolant. Although loss of the moderator (coolant) would render the reactor core subcritical, lack of cooling could cause overheating of the reactor core from residual and decay heat, leading to fuel damage and fission product release. The Core Standby Cooling Systems (which adequately cool the reactor core following a design basis loss-of-coolant accident) and the primary containment and containment cooling systems (which control the release of fission products and absorb the energy released by the accident) are not intended to diminish the overall design objective of the entire nuclear system (to design and construct a nuclear system which will not fail). The intent of using Section III of the ASME Boiler and Pressure Vessel Code and USAS B31.1.0, with added GE requirements for the recirculation system, is to design piping systems of high quality. The Reactor Recirculation System, except for the VFDs on Units 1, 2, and 3, is designed as Class I equipment (see Appendix C) to resist sufficiently the response motion at the installed location within the supporting structure for the Design Basis Earthquake, with the pump assumed filled with water for the analysis. Vibration snubbers located at the top of the motor and at the bottom of the pump casing are designed to resist the horizontal reactions. The recirculation piping, valves, and pumps are supported by constant support hangers and by sway braces to avoid the use of piping expansion loops which would be required if the pumps were anchored. In addition, the recirculation loops are provided with a system of restraints designed to limit pipe motion so that reaction
BFN-28 4.3-7 forces associated with any split or circumferential break do not jeopardize containment integrity. This restraint system provides adequate clearance for normal thermal expansion movement of the loop. The spacing between limit stops is set on the basis that a split pipe retains its structural load-resisting characteristics. Impact loading is negligible on limit stops, since possible pipe movement is limited to slightly more than the clearance required for thermal expansion movement. The recirculation system piping, valves, and pump casings are covered with all-metal, reflective, thermal insulation having an average maximum heat transfer rate of 80 Btu/hr-ft2 with the system at rated operating conditions. The insulation is prefabricated into components for field installation. Removable insulation is provided at various locations to allow for periodic inspection of the insulated equipment. 4.3.5 Safety Evaluation Reactor Recirculation System pump malfunctions that pose threats of damage to the fuel barrier are described and evaluated in Chapter 14.0, "Plant Safety Analysis." There it is shown that none of the malfunctions results in fuel damage; thus, the recirculation system has sufficient flow coastdown characteristics to maintain fuel thermal margins during abnormal operational transients. In addition, in order to achieve a more rapid core reactivity reduction in the event of a turbine or generator trip (thereby limiting the magnitude of the fuel thermal transient), a recirculation pump trip (RPT) feature has been added. By utilizing the recirculation pump trip in response to a turbine-generator trip or load rejection, the MCPR margin is reduced, allowing normal operation at higher power than without the RPT feature. This satisfies safety design basis a. The RPT feature is described in Subsection 7.9. The core-flooding capability which is provided by a jet pump design is pictured in Figure 4.3-4. There is no postulated recirculation line break which can prevent reflooding of the core to the level of the jet pump suction inlet. The core-flooding capability of a jet pump design is discussed in detail in the Core Standby Cooling Systems document filed with the AEC as a GE Topical Report.2 This satisfies safety design basis b. The Reactor Recirculation System piping and pump design pressures (see Table 4.3-1) are based on peak steam pressure in the reactor dome, plus the static head above the lowest point in the recirculation loop, plus dynamic head due to system operation. Piping and related equipment pressure parts are chosen and analyzed in accordance with applicable codes. Use of the listed code design 2 Ianni, P.W., "Core Standby Cooling Systems for Boiling Water Reactors," General Electric Company, Atomic Power Equipment
BFN-26 4.3-8 criteria provides assurance that a system designed, built, and operated within design limits has an extremely low probability of failure due to any known failure mechanism. 4.3.6 Inspection and Testing Quality control methods were used during the fabrication and assembly of the Reactor Recirculation System to assure that the design specifications were met. Inspection and testing were carried out in accordance with USAS B31.1.0. The reactor coolant system was thoroughly cleaned and flushed before fuel was loaded initially. During the preoperational test program, the Reactor Recirculation System was given a hydrostatic test at 125 percent of reactor vessel design pressure. A hydrostatic test at a pressure not to exceed system operational pressure is made following each removal and replacement of the reactor vessel head. Other preoperational tests on the Reactor Recirculation System included operating valves and verifying that seal leakage was small enough to permit pump maintenance work, operating pumps and variable frequency drive (Units 1, 2, and 3), and checking flow control transient operation. During heatup in the startup test program, the horizontal and vertical motions of the Reactor Recirculation System piping and equipment were observed and adjustments of supports were made, as necessary, to assure that components were free to move as designed. Nuclear system responses to recirculation pump trips at rated temperatures and pressure were evaluated during the startup tests, and the plant power response to recirculation flow control was determined. Inservice inspection is considered in the design of the Reactor Recirculation System to assure adequate working space and access for inspection of selected components. The criteria for selecting the components and locations to be inspected are based on the probability of a defect occurring or enlarging at a given location, including areas of known stress concentrations and locations where cyclic strain or thermal stress might occur. The recirculation pump casing and valve bodies can be examined when the pump or valve is disassembled for normal maintenance. The piping connection welds can be examined to the extent practical within the limitations of design, geometry, and materials of construction of the components. The inservice inspection and testing program for the recirculation system is detailed in Subsection 4.12.
BFN-28 Table 4.3-1 REACTOR RECIRCULATION SYSTEM DESIGN CHARACTERISTICS (3952 MWt) External Loops Number of Loops............................................................................................................................................... 2 Pipe Sizes (nominal o.d.) Pump Suction,in.................................................................................................................................. 28 Pump Discharge,in.............................................................................................................................. 28 Discharge Manifold,in........................................................................... 22 (Units 1 & 2), 12 & 22 (Unit 3) Recirculation Inlet Line,in.................................................................................................................... 12 Equalizer Line,in (Unit 2 only)............................................................................................................. 22 Design Pressure (psig)/Design Temperature(F) Suction Piping.......................................................................................................................... 1148/562 Discharge Piping...................................................................................................................... 1326/562 Pumps..................................................................................................................................... 1500/575 Operation at 3952 MWt (100% Core Flow) Recirculation Pump Flow gpm (approximate).............................................................................................................. 47,400 Flow,lb/hr.............................................................................................................................. 17.95 X 106 Total Developed Head,ft................................................................................................................... 643 Suction Pressure (static),psia........................................................................................................ 1,056 Available NPSH*(min.),ft................................................................................................................... 558 Water Temperature (max.),F........................................................................................................... 529 Pump Brake HP (min.),hp.............................................................................................................. 6,734 Flow Velocity at Pump Suction, fps (approximate)........................................................................... 30.5 Variable Frequency Drives (Units 1, 2, and 3) and Power Supply Frequency (operating range),Hz.............................................................................................. 11.5-57.5 Total Required Power to Variable Frequency Drive (Units 1, 2, and 3) HP/set............................................................................................................................................. 7,194 HP total........................................................................................................................................ 14,388 Jet Pumps Number............................................................................................................................................... 20 Total Jet pump flow,lb/hr....................................................................................................... 102.5 X 106 Throat I.D.,in.................................................................................................................................... 8.18 Diffuser I.D.,in.................................................................................................................................. 19.0 Nozzle I.D.,in. (representative)......................................................................................................... 3.14 Diffuser Exit Velocity,ft/sec............................................................................................................... 15.3
Includes velocity head.
BFN-28 Table 4.3-1b (Deleted)
DRIVING FLOW [_======C;:\\=- RECIRCUlATION PUMP ...,, CD )>
a Zo
~ ~~ ~ UJ~ z JTI ~ ):. \\II 0 -ri -n D Ci) ~ ;o ~ f'TI 3:: ):, 0 -< ;;o C
! C

o rr,
o 0~ )>-<
z [Tl Z-1 z en o 'l> ~ --1 ~ o:z ~ 0 (.;J s::: <.n U> r ELEVATION RECIRCULATION OUTLET OR PUMP SUCTION ----, REACTOR PRESSURE VESSEL DISCHARGE RISER OR JET PUMP RISER -._ __ MANIFOLD OR RING HEADER .__ __ RECIRCULATION INLET OR PUMP DISCHARGE ISOMETRIC (UNITS 1 & 2) RECIRCULATION OUTLET ~ ~ tn ~ O') RECIRCULATION PUMP "\\2 ;! ~ ISOMETRIC i l ___ rd~~~--J _______________ ~(~UN:l:T~3~)-----------------------------~
o
-I~
8-t,Qij l-Ll83Lt-l V'l L9 PRIMARY CONTAINMENT ,,~x"-17 tJ I RPV HEAD VENT q:J 4" -! J 'r X-234D X-235D 1 Bo" X-234E I (rw\\ 64-1.§.1 E X-235E X-234F 1 " WATER SEAL BETWEEN RPV AND PRIMARY _,,A CONTAINMENT~_,, N 2.. I 10-so1 N 1-----"-t~.. -----* .+....J'--------,--t,-----, CRW I 10-sao l&--1-------ill CONDENSING POT N7~ ,I. ( T - -!al m -ra, PCV -17 1 " X-30A J I 1-=-2+9A I TO FT 1 13,PDIS 1 13A,8,C,D I ICDNT 1-47E610-1-1,G5,H5 X-235C TW \\ 64-!.§.1D TW 64-t§.2D TW 64-162C TW TW 64-!.§.1F -235F 64-1§2[ TW 64-!.§2F ~ IX-234G ws+-1~ 1-- 2" V u T3 f'i:s\\9 BA .l. 1 0-502 ,..l, 1 \\tl,a~ 112" ~ T L!J f-------Cl--t:ix:J~+:J---;..----+-' < 1-+7[803-5 ~ 1/4" =:'\\ F-1 'TESTABLE 73-~:~ \\.~~; CHECK 14" ~ 16" FCV 3-!,-888 ~, LOW PRESSUREF----------+---J SEAL LEAKOFF 12" HPCI CONT 1 47E812 l,E6 TEST. J/4" ":l-7-3-l-s.,s::o6t--7-3-.. 51C5>151-',"'"'-'"'""....1 THERMAL CRW.... f-"'-"-i<!*-.... o...
---....1
( N4A 12.. I 21-~~.;.*..:.r~*4* [IJ----4---oJ SLEEVE - 1 " VENT 3-562 3-561 7 TEST 73-558 7 3-557 [~~~~~~=====Jf----------~J--C:::=!:~2~4~"------.!F:!E;,!E~D!!W!ATl;E;!R~Lc;l!N,;E.;A~----;::,,..jtir--.L--t:Jb----_;2~4~"-L-..l1l2;"~-71 1-47E803-1, GS N4C 3 5581 ~ 71-531 1 /4" 6" ~TESTABLE CHECK (rev,,'/ '4".. TEST~ 71-40{_ 71-536 I /71-535 X-9A Ll.J 1" - 1 " ~ I,:::;.:.....+----H-~- CRW 1' ~ 3-559 3-560 ~ N12A IRCIC CONT 1-47E813 1.F6 71-534 3/ 4 " TEST 71-533 4" 85-577 ~ 12" icRD CONT 1-47E820-2.H9 ~ 85-576 3/4" TEST [ 69-581 ( N4F 85-575 85-574 T ,1r1 12* I ~ 21----~rN4E X-98 24" ~ y 24" 12" 1" VENT 3_-57_ 5 RWCU CONT ON 1-47E810-1,E6 I 69_579 69-629 69-580 [ ~~~~~~::::::::::::::::::i-~~~~--;:;:::::-~~;f-~fl:~~;:-__:;,._~~~~~~~!F~E~ED~W~A~T~E~R~L~I~N~E~Bi...~';:;;;';~~~.J.~.c,l(J-~~~~.:;,:--.i.~~~ij'~~N4D 1-47E803-1, F6 3-572 ~ THERMAL 1./ rr'-~X--36 ,1, 1" _ SLEEVE-/ ~ ~ 2~/4" TEST ..-1/2" .. L_...,.... _--t,.. _f---,~CRW 15-626 75-627 I\\ 17'1 3-573 3-574 N9 SHV-75 ~ 3 -53B tr5~ ~'-!. 3/4" 2 ~;5-6258 2" 2., N1 SA r:-:-1 V N11A CD SHV VTV-75-53 r X-168 TESTABLE~ 75::,.55 10" ~ N5B r------~------=-~ ~-----------~-- 1-47E814-1, G6 DRV-75-53 SHV-75-SJA FCV POI 1/2~ 1 2" CKV-75-56 RTV-75-56A 1.1 " / 12" [ 1/2" 3/4" CHECK If CKV 54 LO ~ TEST 3/4" TEST r TV-75 SHV-75 SHV-75 TV-75 1 0" I :E'-/k2 *H.,..1"'3~1/.:!.*4_"..,.-.... -, [B---( 1/2" 75-625A 75-624A VTV-SHV-75-25 75-56 7!-25 ~-----~.E-II04HI04t--...:i-_ 1-47E814-1, F6 ~-----~DRV_ rs 25 75-258...,_ 1 2"
I
-54A 1h~**.:5-"4A'--~-5'-'4.:B ___
=.54c,Bc_ __ _,
X 27C 111 X 16A I 12" 3/4" TESTABLE CHECK CKV-26 I 3/4" r -. ~NSA 1 O" (Pois SHV-75-25A ~ 28 1 /2" ECKV-75-28 1 " *I RTV-75-28A l.. I I
I 1"
'I ... 1..
J
... 1"
l 1 "
I I -1 - l " I I 1"
J 1"

I TEST (TYP 2)...
1 " 1;2~ TV-75 SHV-75 SHV-75 TV-75 - 1/2" X-270 ,I, -26A -26A -26B -26B ,pl-------------------='--~=---='---=~---~ X-30F -, ~ X 30E ~ X-498 -, ~ X 49A X 34F ~ X 490 -, X 34E ~ - ~ X-49C FCV 28" RECIRCULATION LOOP A 43:;.P 3/4" 1;2*.., 3/4" TEST ..3. ....1 12" NIA ..... ~---tk>llf--i>I. rt'C~x-1-4:.;'.,;'~/:_:4:._" ___ -:;::=~l+-- 4~-5~9 43-0125 43-1057 314.. 4 43-815A 43-8158 I ~,. 1-.l....-,>_(_j---------------e-;2~4~"---, lsAMPLING CONT 1-47E610-43-1, A7 ~ 69 ~ 3;4°68-6602 X-138 CKV-74-68 24" 68-6 03 .--e---,IN2K ( N2J 12" M M 1>1-3 28" 12" VENT ~7 74-6348 74-6338 ~ ~-* 24" RHR CONT 1 47E811 1,F4
.. I,..
IM 3 4" 74-6388 M ~TEST 74-6398 I,s~- 3/4" 74-7948 74-7958 3 '4" ~.-* ~ 74-6368 74.-6378-\\ -II04t-+;:4--I)!--:. -r, \\rTEST
l 0
74-663 74-664 -~; __ J L J LU FE 8-5; - ~ ijj r N2H 1-:-:.. ::---+~-:1,::: DECON TAP-- -;;/Il-7-!4"'_6"6.J.2-7!'4"_,661-1-.-fil---"'C.:!..- IQ~ If.. 74-755 74-756 ' - FLOOR DR, r,n., NOTE 12 .i..:....:::,I ""' SEAL FLANGE COVER DR CRW 1 " FL.:OOR DRAUi 1!l! 68-801 1 /2" m N 0 w ~~ ~~ 00 ~c rW zw o~ um 74-658 74-655 FCV 74-47 ,..,,. *~--1a11---1~ ~ FCV 74j8 68-~! 3/ - ri>rri~ 4" '-1/2" 1;:; lJ f...--'-C~**>frTI\\,li<!L..;,~:-,,- CRW-I).Q... I ,,l5s~s2l I ~ ~9 r68~55~ 7 l L3/4" ~z.1/2" SEAL is---rl,:J---='-"'-,,-, ~ B~AKAGE 1-1/2" l >--:"'"3/4" I CRW J / / 74 654 ""'20".. X 12 J RHR OONT ON 1 47E811 1, ES ,r.3/4" FCV 69-648 649 l DRIVER MOUN ' - 'il'I© I w DRAIN / ~ TW FL DOR 6 8-21-'-"'-' DRAIN RECIRCULATION 'L_.::r ,.*_06*57--7-14*-06*56--' 69 _505 3(4. ~9-551 69-55_3 _!Sl-J-6" RWCU CONT ON 1 47E810-1,G7 I X-14 PUMP A []r- ~ a 69 583 ~ 2 ~------' I-1 .~ LO~ ./ ~ m L-II04P'/cc2:.'3 '-o, 69-584 1 j~6 7 ~ 69-634 DO "!!::-==----- ~ TO PS 68-93 CONT 1-47E610-S8-1,J5 TO PS 68-94 CONT 1-47E610-68-1,J5 I I I I ~'---.... OIII---.C:+-1:a N (, ' ECKV-68-9:_:3~i:~t;'!..':'..' _x_-3-jl}E~::::)----j'h 69-503 69-504 ~a':~ 6~-22"sA 1 ,pl-~ 16~8--2~3~7-~6~8-~2~3~8---I ECKV-68-94 1.. I X-31 F,I, 1 " ,-J /.I J ,.. -r J, 68-226A 'I' I - ~_E_c_K_v_-C6l8-J_!~5-A-{)::(]---~1~*~*a::::::x'.::J-33A I>, -~-6_B_-2_,_9 __ 6_B_-_2_4_o---'rrl=7=---' TO PDT 68-65 CONT 1-47E610-S8-1.J5 I 1 1* I '--------~---~ 68-213A '--t,*----1~ 1" X-SOE Ec.--2...:' "{t:;::J--:3: 68-249 68-250 2" TW ~ 68-6 ~ DO -t*' *;:::7 DECONTN CONN CRW - ECKV-68-64 l.. IX-31C *I* TO Pl 68-648 & PT 68-64 ,J 1-------------------------- CONT 1-47E610-68-1, J3 68 217A 11' TO PI 68-638 & PT 68-63 ECKV 68:.§3 1" I X-31 A h -1 CONT 1-47E610-68-1. J3 ECKV-68-=-sse s8-2f6A 1" I' TO PDT 68-65 L..:..: ___ <d----;;,t';;!t;--...2.~c~lx=-~3,:3~8,I, CONT 1-47E610-68-1,J5 f 68-214A
I 11'1--------------------------~
a=-----------=a:
ECKV-68:_?A 1 " I X-32A -1 h r z w > 0 ~ N ' m w. ~ N ' m m ~ N m w ~ ~ N ' ::: 1--~ N.. N.. m N m m TO FT 68-SA, 58, SC, & 50 CONT 1-47E610-68-1 ECKV-68-=;::8~6~'!f.c-~2~d-r(D~A---~'-i" :I::~~lx=-~3=2="1.:: I 1----<C sa-201 A
J
'l't---------,-/-4-. _T_E_S_T_E:-'------.. --IC-,.-.. -------.. -1<-,.-.. ---------~-~-,-,--s-'----~ X-37C __ 3/4" 1" 1" - PUMP SEAL INJECTION 3/4" 68 565 68 564 _ 314*
J N '
m m CONT 1 47E820 2,E1 68-552 3/4" TEST-1" (TYP) 12 LINES 68-567 68-566 68-508 68-507 TO NOZZLE NBA~ 68-550 <,-I m,........l, u,........l, o,........l, w,........l, ~ rl <,-I m,........l, u,........l, o,........l, w,........l, ~ I I..l.......:.. I..L.......:.. I..l.......:.. I...t........:. I...I I I..l.......:.. I J..........:. I..l.......:.. I...t........:. I L=====-< CD CD CD CD CD CD O 0--0--0--0--ci-O 0--0--0--0--0
~* ;1;:!: ~* ~* :i* :;*
~:!: !* ~* ~* :::.:: 3/4" VENT L!J <O <D <D <D <D u:, r,.. r,.. r,.. r,.. r,.. N N N N N N N N 1.N N N ~ ~,~ 1~ ~ ~ ~ ~1~ ~ 10 ION N N N IO m..- ~ - - ~ Ji J, J, ~ _. I _._ ~ ~ ~ <1QCD<co u:, "'ID --coCD<1 LO co IOLO,o co I u:, I en I NI LO ID LOLOa:, en I IQ I NN INN I N N N?; N 2: I..-..- ID 2: 2: N ,I, m 7 ~ ~ ~ 6 2" REACTOR 8 1 NlS X-2340 ~ ~161C W 1 " X 34A 90° I 2" 64-1628 ,~ ~ F~V 1-250A " ~---,=---=, .., 1-14 5 1,..;,.;;.26;."~L.....;.;:;,,,-;i,..~...:::;,,,-~~-..L. '~.;;;,,..L. ,-~.;;,,;.,,-L...i-~ ",..L.....;.~-, FE }---!j~:1---~~~2~6~" ~X:;;.:,;7A""(:::;::::]3-~~~~~"'.""~-f-----j~~~~~~~~~~~ N3AT:' MAIN STEAM CONT 1 47E801 1,87 I-f.AAIN STEAM LINE A l._ ___ *1-~-1----"""."l'::Dl":"D'-"--. ~ - s+-1s2G::.,--.... ~ x-2J X-235G W 64-162A 64-162H -{I] 1 O" 64-1618 64-161H 3/4" TEST 1-505 1-504 ~,--,X-234H 73-225A X-32F X-2348 I CONT 1 47E610 73 1,D2 --...,.6.:.*~-1'!6"'-"'--<-~ X-235A I X-235H X-234A TO POIS 73 1A,PDI5 73 18 t----[] N12B f.-l- ~ X-32 V 73~22<A ~rf1cv 73-3 N118 4- ~ N168 L ~ 73-583
~~,' FE,<--,r--.,c;'3----------~-...::r:.. ___ ~~~X'.:_!l2l{:~:I:e-:-"'.".".~~--..2.':"~~-~~~:iii~::::~~:::;:~i!i~~~~::J 73-lf HPCI CONT 1 47E812 1.G6 TORUS TEMPERATURE MONITORING SYSTEM PCV
n-
-18 PCV 1-19 ~ ~, ? ct]'~- ~ 0 MAIN STEAM LINE 8

1.~

PCV 1-22 PCV 1-23 N f~-~,i'i-" 26" I 1 O" I 3" 3(4" T ST KJ-+:++++::J-<t, N 73-585 73-584 '-1" x-3DB~~~==~~~~~=~";:r~==°9_~!i.'._~~~~:l.::~£!!~'.'.J t--S~Y-S_T,-EM~P-R-E~S~S~.-~T-E+M-P~D~A~T--,A TO FT 1-25,PDIS 1-25A,8,C,D X_348 CONT 1-47E610-1-1,G6,H6 LINE DESIGN DESIGN 1-251A 1" MK NO. PRESSURE (PSIG) TEMP(°F) I I ,.:2s2A <1:; 1 1 HG ss2 6 2 150 300 V MAIN STEAM DRAIN 3 1375 562 6 FCV CONT 1-47E801-1,D7 4 1375 376 1-27 ~ 5 1250 376 3/4" TEST 6 1148 562 , FE }-;-t;J<:i---~~.:X~7:!;B~~~3-'.'/.k::J*--2':::-:.s~1 *:....2'.:::.5!:19:..__~M~AlI ~N~sf,T~E~A~*C::--:-~:----1,_.,1_-+----,'-a'.,26~--1---a-s,,.62=--< N - X-8 SYSTEM 1 PCV-1-4 PCV-1-5 PCV-1-18 -1 *" PCV-1-22 PCV-1-23 PCV-1-30 PCV-1-31 SYSTEM 10 2-1/2" 10" 10-506 10-521 10-507 ~- 10-508 ~ 10-510 10-511 10-512 10-513,. 10-526 10-528 -~ ~~ L..J -... 3{4* = CONT 1-+7E801-1.C7 B 625 500 TST---t-f--..._"' 9 1250 575 3/4" TEST 10 150 212 I I 26" N38i,-~~~~--=~~.:..~~~--I~ 1 O" -1 PCV-1 PCV 1 PCV 1 PCV 1 42 179 180 10 515 10 516 10 519 10 520 I, O" 1 O" 0 0 N 26" l---{]101 -.. PCV -34 0 1 " I 1.. 1 " 1 " 1 " X-10 X 51f I I ' ;:(@1,-513 1-514 ~ FCV
g r-- 71-3
~ 1/4"- 71-563 3;. I I RCIC ~--, CONT 1-47E813-1,GS TO PS 71-1A & 1C CONT 1-47E610-71-1,F7 I TO POIS 71 1A, FT 71 1A CONT 1-47E610-71-1,F7 II 71-216A I ~X::;5~1c!E{:::::::If------t:XX::f------_l_--J TO PS 71 -1 8 & 1 D CONT 1-47E610-71-1,F7 71-217A !X-:;3e,3!!fz~=[(3-------l:XX:J---------jl TO PD IS 71 -1 8, FT 71 -1 8 CONT 1-47E610-71-1,F7 X 33E I I _ 71 -222A ~~{::::::::r:1-----t:XX::i-:-.:__::.::::::_==_--J TO POIS 71-18, FT 71-18 ~ 3 * @ 71-221A tW CONT 1-47E610-71-1.F7 Y FE FE "i\\,,,,-[>'(3---1-' 7 ,- 71-1',A---171-18 r _3/4" TEST _ ~ PC -m _..... ~,
:i:. ~~CV
~r.-l-:-~-f;.kl_,...~~...!...!..!..~~~2-6.,*~---7{c=1:::::~~-:}_-1,-~s~3~3~==~*~;s1,!.2-_+,.::}~~b1.<_1---j ~~~~ i~~~~801-1, E7 to 1-41 8~
1-42 y L 1" X-30C 1" l "' 1 81 J r I TO POIS 1-36A,8,C,D & L-,--'--c-'c ~ 1 __ !! 10" N in* _ l" I X 34c 1-255A ~----< FT 1-36 CONT 1-47E610-1-1, _'.___c:,::r-~:....!1~"--:--J?:~_JI ~A_6,~B_6 _________ ~ N3D ', ~,- ~,- ~T ~T F";\\E 26'" f'"'"--"'--""-'""'-.._""..a"---"---f 0)----l~:l--*---:-------------"'."'.1 I 1-256A MAIN STEAM LINE D \\!-5;? 1" X-34D 1" 1 10" I ~ 1-254A 1:::=;~~=11~TO~F~T~1:!i5t0i, P~Dfl~S':i1J50rA~,JBt,~C~.~D >--191 (Hcv ,-l-~t--'-3'-/4:.."+*,i*1--i>l
TEST I~------------
1" l-{::::;::::lxJ-G3e-O-D_l1_::"_.cx,(J-__JJ CONT 1-47E610-1-1,85,A5 N1 o I ~ ~~ 2 63-538 63-539 63-526 1 _253A f_~v 1 " PRESS. BELOW CORE PLATE X-70 9 I - - VENT-+:4----tOIIH1 1" PRESS.BELOW CORE PLATE 68-6608 68 6607 I L..-{::;:::::J:r-ji':~!n""'ss'.'."::4'i""::1;,k::J----j MAIN STEAM 63-540 63 541 CONT 1-47E801-1. E7 .... )-1--'-1 /'-2;_" __ - *-------{=l~x1-,_4.:2_[ 1:~-a.. _..,_111_--~ N '-1_)!-.... ~-~::;;;::::.;;;;~:::::::c~.-T-E_S_T-~ Nl ~12" N2CI,-_......., N2D l 12" 12" N2C "J I 12
N28 l:',J-i--*~
28" 24" 1" PRESS.ABOVE CORE PLATE 74-794A 74-795A DECON / TAP_,./ en! ~ m N 1-540 3/ 4 " 1-541 iii X 278

1,1, 111 X 52F

1,1, I I

3/4" CHECK VAL VE TEST
(
63-528 63-529 ST AND BY LIOU ID CONTROL I CONT 1-47E854-1, E7 63-525 ~5 m.. m,
lhl-!;X-~2;!7!A(~'~r,-,
R~EC_I_R_C_U-LA_T_I_O_N_L_OO_P_B_ I --2-.-.-tl, I I X-13A TO FT 68-158,18,19,21,22 68-203A 1" 258,26,28,29,30, PT 68-95 CKV-74-54 __.. -.r - 637A 74-636A 74-638A 74-639°A. ~...:.--,--jPDT 68-52.PDT 3-51, CONT 1-47E610-3-1,EB 1-47E610-68-1 B8,C7,D7,D8,E8,F8,F7,A6,A4 ECKV-68-1588 11 1146 575 12 1500 575 TABLE 8 13 1750 150 SEE DETAILS 1A1 AND 181 TABLE A NOTES: 1.
2.

3.

4.

5.

6.

7.

8.

9.
1 D. ALL VALVES ARE THE SAME SIZE AS PIPING UNLESS OTHERWISE NOTED. ALL VALVE BODY DRAINS, BONNET VENTS AND PACKING LEAKOFFS ARE 1/2" UNLESS OTHERWISE NOTED. ALL VALVE NUMBERS SHALL BE PREFIXED WITH THE UNIT NUMBER AND SYSTEM NUMBER I.E. 1-1-5. ALL DRAINS SHOWN~, INDICATES CLOSED SYSTEM. CRW DRAINS ARE CONNECTED TO THE REACTOR BUILDING EQUIPMENT DRAIN SUMP. [I)ETC, DENOTES DESIGN PRESSURE AND TEMPERATURE AS GIVEN IN TABLE (A) SYSTEM PRESS.-TEMP DATA THIS DRAWING. HYDROSTATIC TESTING SHALL BE IN ACCORDANCE WITH THE APPLICABLE CODES. THE DESIGN PRESSURE AND TEMP OF ALL DRAIN AND VENT LINES THROUGH THE LAST ISOLATION VALVE SHALL BE THE SAME AS THE PROCESS LINE. VENT, DRAIN AND TEST CONNECTIONS 1-1/2" AND BELOW CAN BE PROVIDED WITH PIPE CAPS OR HOSE CON~ECTION FITTINGS WHERE REQUIRED BY PLANT PERSONNEL. THIS CONFIGURATION IS SUPPORTED BY ENGINEERING CALCULATION CD-Q0999-923399. ALL SENSING LINES WITH EXCESS FLOW CHECK VALVES EXCEPT X-40A-A THROUGH X-40D-F HAVE AN ORIFICE COUPLING, (REFER 10 DETAIL P20 ON 0-47W600-20).
11. UNIDS ON DRAWING ARE FOR REFERENCE ONLY AND ARE ABBREVIATED TO MEET SPACE CONSTRAINTS. REFER TO MEL FOR COMPLETE UNIDS.

12. SEAL FLANGE COVER IS VENTED THROUGH AN OPEN-ENDED *TEE" CONNECTED THROUGH A DRIP LOOP.
REFERENCE DRAWINGS: 0-47E800-1.......... FLOW DIAGRAM-GENERAL PLANT SYSTEMS 0-47E800-2.......... MECHANICAL SYMBOLS & FLOW DIAGRAM DRAWING INDEX 1-47E811-1.......... FLOW DIAGRAM-RHR SYSTEM 1-47ES14-1.......... FLOW DIAGRAM-CORE SPRAY SYSTEM 1-47E610-75-1....... CONTROL DIAGRAM-CORE SPRAY SYSTEM 1-47E610-3-1........ CONTROL DIAGRAM-REACTOR FEEDWATER SYSTEM 1-47E812-1.......... FLOW DIAGRAM-HPCI SYSTEM 1-47E803-1.......... FLOW DIAGRAM-REACTOR FEEDWATER SYSTEM 1-47E813-1.......... FLOW DIAGRAM-RCIC SYSTEM 1-47E820-2.......... FLOW DIAGRAM-CONTROL ROD DRIVE SYSTEM 1-47E810-1.......... FLOW DIAGRAM-REACTOR WATER CLEANUP SYSTEM 1-47E610-68-1....... CONTROL DIAGRAM-REACTOR WATER RECIRCN SYSTEM 1-47E610-43-1....... CONTROL DIAGRAM-SAMPLING & WATER QUALITY SYSTEM 1-47E801-1.......... FLOW DIAGRAM-MAIN STEAM 1-47ES01-2.......... FLOW DIAGRAM-MAIN STEAM 1-47E610-1-SERIES... CONTROL DIAGRAM-MAIN STEAM 1-47E610-74-SERIES.. CONTROL DIAGRAM-RHR SYSTEM 1-47E610-73-1....... CONTROL DIAGRAM-HPCI SYSTEM 1-47E610-71-1....... CONTROL DIAGRAM-RCIC SYSTEM 1-47E610-69-1....... CONTROL DIAGRAM-REACTOR WATER CLEANUP SYSTEM 1-47ES54-1.......... FLOW DIAGRAM-STANDBY LIQUID CONTROL SYSTEM 1-47E852-2.......... FLOW DIAGRAM-CLEAN RADWASTE SYSTEM 1-47E610-64-1,2..... CONTROL DIAGRAM-PRIMARY CONTAINMENT SYSTEM 1-47E822-1.......... FLOW DIAGRAM-CLOSED COOLING WATER SYSTEM MEL................. INSTRUMENT TABULATIONS MEL................. TABULATION OF VALVE MARKER TAGS I 12
1 " TEST-3/4"
-68 6605 TYPICAL~OR 13 RELIEF VALVES ~~~~~~~~~~~~~1-47E652-1.......... FLOW DIAGRAM-DIRTY RADWASTE 1" TO FT 68-78 *8*10 *11
13 *388

31000215C-8, -9..... FLOW DIAGRAM-VFD COOLANT N2A 1,--e--
~ 68-6604-68-202A C:,,-'---,----j39,40,42,43, PT 68-96 N88 D-3/4":.. ftft ftft~ft in! --\\FLOOR DRAIN ~ SEE NOTE 12 ECKV-66-78 ~g~~3~Et:~;:g;~:s~9g4,C3, 94
GENERAL ELECTRIC co DRAWINGS:
'

::::::::::::::::::::::::::::::::~ 021-043090....... SPECIAL TEMPERATURE FE s8:..51 28" POIS 75-28, 1-47E610-75-1,D2 6~/~Q2 L_..::C,rs~...:3/+,4:}"Lll.'.."_J----1.~ CRW 68-204A 1" PD PDIS 75-56, 1-47E610-75-1,C2 r
68-PDT 85-17.1-47E820-2.H4 fFCV\\ ECKV-68-52.._... PDT 85-18.1-47E820-2.H5 SEAL FLG 68-803 SEAL --.,_--"3,:1*4:,_'_s' ='-.!',t:,H'" FIS " 3/4" COVER OR .1/2" bfi~KAGE FLOOR I B-74 *r* =10 CRW DRAIN r, 1" L1;2~ ~ 3/4" .11 /'r 1s\\ 11
3/4" CRW 74-53 Pl 85-19, 1-47E820-2,H5 FCV 68-79
,1 LJ/ ~SEE TABLE (B) FOR VACUUM lh4-633A 74-634A RHR CONT I "BREAKER IDENTIFICATION 1-47E811-1,F7 FOR PCV'S 5,179,23,34,180 AND 42, ~ SEE DET 1A1 FOR VACUUM BREAKER CONFIGURATION
_-J (
) "68-68>>1 3/ 4~ 1-1/2" ~ Ll/2" ~ DRIVER MOUNT J_ [ID-------1, '----', DR FL DOR / DRAIN TW..--+' 'i2' RECIRCULATION 68-78 11..!..=..1 PUMP 8 ~ 1----..-----, --,--------,./ 1 FOR PCV'S 4,18,19,22,30,31 AND 41 SEE DET 181 FOR VACUUM BREAKER CONFIGURATION 2 112* ~ w m N o N ' * :l 28" ...,._~--,,...-I TW 2 " 6@.-5~0 68-83 I*~ -' >,), m CRW 4" DECONTN CONN FCV 68j7 L J SEE NOTE 10 -" 50F I I X-33C I 68-22DA
B-J 68-219A 1 O"
, I 12" I -,I-- PRESSURE SUPPRESSION CHAMBER - ~,,,,.... DETAIL 1 A1 I TO PDT 68-82 E!:,C.. KV---5-0-_-82-A-, CONT 1-47E610-68-1,J6 1 "*I* X 310
t,-------jl TO PT 68-76, Pl 68-768 Y'--------------------,,1, 1
ECKV-S8_76 CONT 1-47E610-68-1, J7 I COMPENSATOR SEE TABLE {B) FOR VACUUM \\"' BREAKER lD~NiIF ICATION 2 1/2':. *- 10". ~ ' I 12" I Ii I - PRESSU;- SUPPRESSION CHAMBER ~..,, " DETAIL 1B1 1 " 1h~~X~3:!_18~=:I) IB-_:6:85-0 2~18:'.A:_-0:~:::-::--:::-"i TO PT 68-75. Pl 68-758 .I' [* CONT 1-47E610-68-1, J7 I 1 1 - ECKV-68-75 1 "*lo X 330 68 221A -._,,_ __

jl TO POT 68-82

'------1------------------l,1, 1 '-:::CO::N::T::::1=-4::7::E::6::10::-=6::8=-::' ::' J::6~ AMENDMENT 26 I I ECKV-68-828,... 1 "1h~~X:;3"'2~C:.-c:::::::r:~6~8D2:x2~3:::Ai.~~:-::-:~ ~-----------1-------------------<,1, [* TO FT 68-81A, 818, 81C, lr-----------------------------------1 11 ECKV-68-818 & BID POWERHOUSE 1 ",I, X-320 68 224A CONT 1-47E610-68-1 '----------------------------1,1, I UN IT 1 .-*111---1... 1-':J I ECKV-68-BlA ~--------~~-=~:....:._:_ ________________________ -I rn...J-Ji""'iJ 68_561 3/4" TEST ---~3L/*4+/-.::*==-b, I~ =:.J--1,,)6~*:-:5r60~::.~~,i,,,!J*~*1 ::i-~*c.:*~X:::-3~*~c:.c~~r1*.::**-c,-1-t*,,,.i:3~/;4~*~-:1:::--{~~:::i:;~~~~~::::JI PUMP SEAL INJECTION r r 1 .1.,~ 1 CONT 1-47E820-2.G1 ,--1" (TYP/ 12 LINES 68-522 68-523 6 S 3,_ 3/4" TEST 68-555 68-557 I TO NOZ~N88 L_ 68-5~i6; CD 7 ~ ~ )LL-r:c-T ~ ~ ).w LL-BROWNS FERRY NUCLEAR PLANT > ~
. '/..L.....'.. '/..L.....'.. '1'...L...'.. ' -i ' ',___'._ 'l'...L...'.. '/..L.....'.. '/..L.....'.. '
r--UI U U (.) U U QI C Q Q Q Q -o 0--0--0--0- O 0--0--0--0--0 FINAL SAFETY ANALYSIS REPORT 68-556 I 7 < I < I
ii I
I I < I < I., I I I 7 ~ I x ~x ~x r--x ~x lilx
x
~x ~x
x

2x IQx lo N
N N N N N N N N N
,,N N
~ *,-, ~ ~ ~ h ~ ~ l'"I ~ r:s I..-~ ~ I J.,- '-,._ I
g I
J. -'- '7,- - l'"I m' ~
0-Ji I

g
~< IQID<Ji <D
0-1')
I I CD<co O I 2: >I I') I all <ID IQIQID O I en I 12: 'i'2:.... ""' 1 12: IQ~ I::.:::
2:P"'>r..1N 1
..,,.,I .... 2: ,.,2: LU 3/4" VENT PRIMARY CONTAINMENT COMPANION DRAWINGS: 2-47[817-1 3-47E817-1 NUCLEAR BOILER FLOW DIAGRAM I I 2: m I R m u I l<DID :> I I 2: I ~ I u o IQCO_ o IQ~ coo aa 11 IQIQ-IQ o IQ o ID w ID w ID If 8 ID w ID ID UJ ID UJ ID,..,P"'> 8 ID ID 8 <D UJ <D UJ IL0:>11:> llll::: Ill<: 1~ I~ 1~111 I~ 1-
g,.!,G:l:G :g8

gf;l :l8 :l8 :lu:l:~ :l8 :lf:il I
UJ UJ UJ U
g 8 L:================:'.J FIGURE 4_3-2a SH 1 I-I-I-I-
I-I- I-1-1 I-I-I- I- .!=. ~ ~ ..!:=. ..!:=. .... ~~ ~ ..!:=. 1 1 I-I- I-I- I-1... .... ~ ..!:=. ..!:=. ~ ~ ~ 8 7 6 5 t 4 3 H G F E D C B A
9!.Qij l-Ll83Lt-Z ~ L9 X-234D X-235D TW X-234E I TW\\ 64-161E -235C 64-161D TW X-235E X-234F ( TW \\'-.. 64-161 f,... -235F 2" 1" ',. 2" 1 0-501 ~1" 10-sOo N WATER SEAL BETWEEN RPV AND PRIMARY.,.,,,,, CONTAINMENT-_,, N 2" CONDENSING PRIMARY CONTAINMENT RPV HEAD VENT
j 4"
N7 \\ SYSTEM PRESS.- TEMP DATA LINE DESIGN DESIGN MK NO. PRESSURE (PSIG) TEMP(°F) 2 3 4 5 6 7 1146 150 1148 1326 150 1500 1750 562 300 562 562 212 575 150 NOTES:
1. ALL VALVES ARE THE SAME SIZE AS PIPING UNLESS OTHERWISE NOTED AND ARE SHOWN IN THEIR NORMAL OPERATING POSITION.

2. COt.iPONENT UNIDS SHALL BE PREFIXED WITH THE UNIT NUMBER "2-" UNLESS OTHERWISE NOTED.
UNIDS ON DRAWINGS ARE FOR REFERENCE ONLY AND ARE
3. t~jRfi~ltJ§0AJ~ ~6~JE~ftffEr8°~~lR~J~l¥oRR§ifEol~cMfaur~~E~~Mbkfl~ M"J~~-

4. [I] ETC, DENOTES DESIGN PRESSURE AND TEMPERATURE AS GIVEN IN TABLE (A),
SYSTEM PRESS-TEMP DATA THIS DRAWING.
5. HYDROSTATIC TESTING SHALL BE IN ACCORDANCE WITH THE APPLICABLE CODES.

6. THE DESIGN PRESSURE AND TEMP OF ALL DRAIN AND VENT LINES THROUGH THE LAST ISOLATION VALVE SHALL BE THE SAME AS THE PROCESS LINE.

7. DRAWING 2-+7E817-2 REPRESENTS THE REACTOR WATER RECIRCULATION M-G OIL SET SYSTEM ONLY.

8. DELETED

9. (H) -
IS HIGH PRESSURE AND (L) - IS LOW PRESSURE SIDE.
10. VENT, DRAIN AND TEST CONNECTION 1-1 /2" AND BELOW WERE PROVIDED WITH PIPE CAPS OR HOSE CONNECTION, BY PLANT PERSONNEL.
THIS CONFIGURATION IS SUPPORTED BY ENGINEERING CALCULATION CD-00999-923399. H 6+-162D TW 64-162[ - y N ~
11. VESSEL "D" (VESSEL INVERT) AT 578'-3" BUILDING ELEVATION. SEE CALCULATION I--
CD-00303-940391. 90' TW TW 10-517 ~ 64-2_!i2C X-234C (_1_*.1 4-161C 64-162F ~ ( }.'N_. ) X-23+G 6+-161G/ - 2"" - '\\._ VENT LINE RPV DRAIN FOR TABLE A
12. ALL SENSING LINES WITH EXCESS FLOW CHECK VALVES EXCEPT X-40A-A THRU x-+OD-F HAVE AN ORIFICE COUPLING (REFER TO DETAIL P20 ON o-+7W600-20).

13. SEAL FLANGE COVER DRAIN IS VENTED THROUGH AN OPEN-ENDED *TEE" CONNECTED TO A DRIP-LCXJP.
-2358 TW ~ TW 64-1628 64-162G (TW\\ (TW\\ ~ 64-162A 64-162H & 64-1618 64-161H < 2-47E801-1, F6 1-1/2" 2-47E803-5, D5 10-502 -235G CONT ON (2-47E803-5,H5) SYSTEM 1 PCV-1-4 PCV-1-5 SYSTEM 10 2-1/2" 10-506 10-507 10" 10-521 10-522
14. OPERATIONS MAY CLOSE VALVE (2-SHV-10-100) AS NECESSARY TO REDUCE IN-LEAKAGE.
REFERENCE DRAWINGS: MEL.................. VALVE MARKER TAG TABULATIONS MEL.................. INSTRUMENT TABULATIONS FOR REACTOR WATER RECIRCN SYSTEM 47E600-SERIES........ MECHANICAL INSTRUMENTS & CONTROLS (TW\\ 64-161A X-2348 X-23+H ,. g FROM PCV-1-18 2-47E801-1,F8 1 O" PCV-1-18 PCV-1-19 10-508 10-509 10-523 10-524 2-47E610-3-1......... CONTROL DIAGRAM-REACTOR FEEDWATER SYSTEM ' o-1 0" PCV-1-22 2-47E610-43-1........ CONTROL DIAGRAM-SAMPLING & WATER QUALITY SYSTEM 2-47E610-64-1,-2..... CONTROL DIAGRAM-PRIMARY CONTAINMENT SYSTEM 10-510 10-525 X-235A I X-235H 10* FROM PCV-1-30 2-47E801-1,FB/>-----~... ----- 2-47E801-1,F7 FROM PCV-1-19 PCV-1-23 PCV-1-30 10-511 10-512 10-526 10-527 2-47E610-68-1........ CONTROL DIAGRAM-REACTOR WATER RECIRCN SYSTEM 0-47E800-1........... FLOW DIAGRAM-GENERAL PLANT SYSTEMS PT X-234A TORUS TEMPERATURE MONITORING SYSTEM
205A
-_ 207A - 209A r ?_11~
212A
_ 21 DA - 208A -- 206A 1
X 30F 1

II X-30E 1

X-498 1 "
X 49A 1 " X-34F 1 " I I X 49D 1 " I I X 34E 1 "

X 49C 1 "
1 " 1 " 1 " 1 " 1 " 1 " 1.. ~ 0 CONT ON 20.. (2-47E811-1,FS) ,-=-e--J RHR SUPPLY 1 E-- N9 28" RECIRCULATION LOOP A NI A SAMPLING <2-47ES10-43-1,G4t-- FLOOR ! DRAIN~: I CRW z
.i,., *<
n o ~ ~ FIS 68-5 3/4" 6602 3/4" 6603 / CONT ON 24" (2-47E811-1,FSJ ,--<~ RHR RETURN 12* .,_....,_-,fN2F (FCv\\, I') ( ~~ ~,.... 1a.2',o' _.,...N2G 12*c..._ V -,_.,;. ~FE " I ..,...._-frl:,joo.... '""'N2H
,, +I L..,;. ~ -28" V N
Ll.J .JLl,.J 12" I N2J ,__-I---~- r:3/4" ~ w V> 0 w~ ~~ ~o 12" r N2K ..... -t N~ REACTOR RPV-68-1000 (\\yp 16) 1 O" FROM PCV-1-31 2-47E801-1,F7 1 O" FROM PCV-1-3+ 2-47E801-1,ES/>-----,-, 1 0" FROM PCV-1-42 2-47E801-1,ES 1 O" FROM PCV-1-41 2-47E801-1,E7 />-""'"-e,'I FROM PCV-1-180 2-47E801-1.E7 m 0 U z -m 0 1 0" 10" ... -""----< 2-47E801-1,G7 FROM PCV-1-22 1 O" ...._2 47E801 1,GS FROM PCV-1-23 1 O" ......a.-- 2-47E801-1,H7 FROM PCV-1-4 1 O" 1 O" ...._2-47E801-1,H6 FROM PCV-1-5 2-47E801-1.HS FROM PCV-1-179 1 O" TYPICAL FOR 13 RELIEF VALVES---~--. PCV-1-31 PCV-1-34 PCV-1-41 PCV-1-42 PCV-1-179 PCV-1-180 10-513 10-514 10-515 10-516 10-519 10-520 TABLE B 10-528 10-529 10-530 10-531 10-532 10-533 SEE DETAILS 2A1 AND 281 TO PDT-3-51 2-47E803-5,E7 f-(PT\\ 68-96 0-47E80D-2........... MECHANICAL-SYMBOLS & FLOW DIAGRAM DRAWING INDEX 2-47E801-1........... FLOW DIAGRAM-MAIN STEAM 2-47E803-5........... FLOW DIAGRAM-REACTOR FEEDWATER SYSTEM 2-47E810-1........... FLOW DIAGRAM-REACTOR WATER CLEANUP SYSTEM 2-47E811-1........... FLOW DIAGRAM-RHR SYSTEM 2-47E812-1........... FLOW DIAGRAM-HPCI SYSTEM 2-47E813-1........... FLOW DIAGRAM-RCIC SYSTEM 2-47E814-1........... FLOW DIAGRAM-CORE SPRAY SYSTEM 2-47E820-2........... FLOW DIAGRAM-CONTROL ROD DRIVE SYSTEM 2-47E844-2........... FLOW DIAGRAM-RAW COOLING WATER SYSTEM 2-47E852-1........... FLOW DIAGRAM-FLOOR AND DIRTY RADWASTE DRAINAGE 2-47E852-2........... FLOW DIAGRAM-CLEAN RADWASTE SYSTEM 2-47E854-1........... FLOW DIAGRAM-STANDBY LIQUID CONTROL SYSTEM 2-47E865-3,-13....... FLOW DIAGRAM-RB HEATING & VENT. AIR FLOW 2-47E867-3........... FLOW DIAGRAM-SAMPLING & WATER QUALITY SYSTEM 2-47E2865-SERIES..... FLOW DIAGRAM-RB HEATING & VENT. AIR FLOW GENERAL ELECTRIC CO DRAWINGS: 78-CD-3131 AND 3140.... SCHEMATIC PIPING DIAGRAM-LH & RH COOLER CONN 531E261 AND 531E267.... OUTLINE (PUMP DRIVE SYSTEM) 117C1461-1............. RECIRCULATION PUMP P & ID 719E597-3.............. NUCLEAR BOILER P & ID 104R930-1.............. CONTROL ROD DRIVE HYDRAULIC SYSTEM P & ID 719E532-SERIES......... PRIMARY CONTAINMENT PENETRATIONS SY~BOLS, K JET PUMP I CONT ON THIS SHEET FT-68-158 (84), FT-68-18 1 " PRESS. BELOW CORE PLATE ,l,:r----X~-~2::7::Bc:::;:n--"-1-" --t2xi0c:J3A~ ~,__..J.. __ _:__,1 FT-68-19 ( 84). FT-68-21 '----------,-1---------------------1.,, ( I FT-68-22 (B4), FT-68-259 1" 1" PRESS. BELOW CORE PLATE X-52F 1 1" ECKV-68-159 FT-68-26 (84). Fl-68-28 (84) rn~; (B4) ( 83) ~SEE TABLE (B) FOR VACUUM ~AKER IDENTIFICATION I NIB 'rai,2:' N2El,,'----I 12" ~ N2Dl-,----I '..,,-12" N2C 2B"J 12" N28i,-+----t 12" N2A i.-+---t 41 -N8A 231 28" 232 1 "PRESS. ABOVE CORE PLATE RECIRCULATION LOOP 8 24", CONT ON l-..... e"--\\1(2-47E811-1,F6) CONTROLLED SEAL BLEED OH 3/4" 1" .875 1 " FIS.875 68-'(* ~ 3/4" 3/4" 6605 3/4" 6606 r31** 15 r CRW r 3/4" ,i.._ __ ~::~::::::n-r-"-----"l::'::J:--< FT 29 ( 84), FT 30 ,1r
I*
X 27A 1 1" 202A ECKV-68-78 2 112,\\_ ' I 28" SEAL LEAKAGE 1.. OR FIS ~-:§8 1 " ~ SEAL FLANGE f 204A ' ' ECKV-68-52 FT-68-7B FT-68-10 FT-68-13 FT-68-39 FT-68-42 POT '-,-~-~68-52 ~ (86), !861. 85. 87, (87). TO PDIS 75-56 >----<2-47E814-1,G6 TO POIS 75-28 ~~~-;2-47E814-1.GS ~-----< 2-47E820-2,E4 TO PI85-19 FT-68-8 FT-68-11 FT-68-389 FT-68-40 FT-68-43 ( 86) ~iii ( 86) ' I 1-SEE TABLE CB) FOR VACUUM ~KER IDcNTIFICATION 10* l 2-1/2~
I
' I , 1-~ 12" - ,1-~ 12" 12" I\\. I\\. I\\. ll... al'l1111 DETAIL 2B1 SEAL FLANGE COVER DR 1 " 3/4"...... ( 1.;-. l ~o ~w zw o~ um .,t-JET Pll.n ~ IL VESSEL~'O" FE 68~1 ct 3/4" 1 "./ COVER DR 12".- DRIVER MOUNT DR 1-1/2" 3/4" CRW FLOOR DRAIN (SEE NOTE 13) ro / -eJ RECIRCULATION \\,S.7 PUMP A [§]_,. TO FLOOR PMP-68-SOA DRAINS 3/4" m PS ECKV-68-93 1" X-31~.\\~'LErE-N-O-TE_1_2_(_T_Y_P_1_6_)+--N-I mv./ 504
m N
~ 68-93 ,1r ~ - 225A I 237 1.. 238 PS \\.-:=:::'°~-J:-:---t:O::J---'!..'~' :i:::::::::=X-)-3-'_F.!1_:",I 8-94 111 ~.,,.-t:11<:J,--;,
g ECKV-6i='94 226A 1 "
Li ,
in PDT'--,:;( 1 L) 1 "
X-33A 1 ",I,........ 2'C19f-'-.... 24CIO~- 4 + 3 ' * :g s~5= L~ 213A ,1:.~...... ,....:;:..:.._, -......:;:..:.._='--' DECON =-~rn.* ~ I ECKV-68-530 I ~ ECKV-68-~SA X-SOE 249 250 2 ( c, \\
i:...._,,!;c:-H~f' _::"e:::;;;r;:.!'.::"c:.~:I~~i==------!:C02JN!!!N!....::=:.=::'..cCsJR!!WL...J 68_648 PANEL 25-57._
y ECKV-68.=_64 215A 1 " I IX 31c 1",1, 260 261 ,1:,i-==--...c.;.;... __________________ _ 2" 1--~*r ~ N ~ z w '" 0 n N m N N
..~ J n
CRW N.. N n.. N NOTE 11 I \\\\ CALIBRATE~D CALIBRATED (TYP 2) (TYP 2 ) 3 N15 2" ~ ':il '
O V-3/4" TYP 2 LINES 3 / 4 " __________.
3/4" -------- 3/< * -~---1rTI TYP 2 LINES LP CONN :::::----J TYP 8 LINES L p CONN -=:----.JU TYP 2 LINES HP CONN -- 28" m N **~ I U ~.. N 0 z w " m N = N 2" 5" 28" T TW 68-83 ~ { FOR PCV'S 5,22,34,42,179 AND 180, SEE DETAIL 2A1 FOR VACUUM BREAKER CONFIGURATION FOR PCV'S 4, 18, 19,23,30,31 AND 41, SEE DETAIL 281 FOR VACUUM BREAKER CONFIGURATION. - ECKV-68-82A L..,;1:,."----j;i(cxL:J-c!)\\);:,~.PDT 68-82;---~ ~ 220A -~i 2 f'
,;D /".EC.,ON Lnc-,..;
ECKV-68-531 7 L....!C::!O~N!!N!._ __.1._:2;:6:_:2:...__:2:::63:._....:x:.:::!5!:!0F'.....-{~::Jf1.::"-1~1-o!-'>_J PANEL 25-57 PI 68-769 1 " IIL.,;X::::;3~1 !cD--,c:;l;::1&!1-"_2_2_2_A""i:~~EC~K~V~-~6'.:8-=7::6~ PT L.,.JL..--~--------------1,ii 68-76 ~ CRW DETAIL 2A1 PT 68-63 s~4 I IIX-31A1",h,1:,t------------------------- ECKV-68-658 ( H) 216A 1
11 X-33B I "
ECKV-68-63 217 A 1 " N N I HP CONN 3/4" TYP 8 LINES LP CONN r3/4" II 219A .UU ~ 1",Ii X-318 1" PT PI ~~-----------------1 I I Eeyv SB 75 ses 68-759 ~ 1 /2" _...;=:$~~~--~-~c;:;;c.i.::,*1:*1-----------------' 214A 111 N PI 68-639 ~ A A F;\\IEcKV-68-5A CL> 1" IIX-32A1" *h ,:..~J,:..~J S~A 200A ,,:,t-----------------------------1--- - FT I ( H) l.. 11 X-32B 1 " *I* 6~-5JB I - "'1Ql), ,,:,t---------------------3-,-.4-.. ----------' I 1/2" TEST--,~-to----to--=~ m-+--rn ECKV-68-SB X-37C 565 56-4 I.. ;:-;: 3/4* _ 1" 1
3/4" PUt.tP SEAL INJECTION 3/4 552 550 1/2' TEST 567 VENT_____.-.
1/2" X 3/4* 2-47E820*2, E8 I CONT ON E2 ( THIS SHEET) < _..1._~----------,,,----------~ 8 I 7 I N N CD Cl CONT ON E2 ~ '° (THIS SHEET) N w
,i m
7 ~ ~ T ~ ' N 8 V 6 I TYP 2 LINES LP CONN /Fcv\\ 68-;s ru /Fcv\\ 68-33 5 l N N ~ N 1" IIL.,;X::-~3~3:CD--,c:;l;::l&-1_::"_2_1_8_A~(~H~)~~,n~~=:=E=C=KV=-=6=8=:82~8=~==;=:::!, '------------------------1,ii I 1" X 32C 11 1" (H) 22" ~ ,..l.._ -~----------------------------t'h -!_FI. 1 _( F_,: J FT I I II ECK\\1'=6B-81A68-81A 68-81C 68-81 1" I' 1,. 223A (L) '-T-"' .. :i--:X::::;3~2:0D"""1:::::::i:,.!..::_---;j;~f-;l;~~:--: l~-,-!,FT 3 '4" 1 ;2 *
ii
\\ 224A EcKv-sa-a1 e I 68-.fil B 3 TEST L---~-~ I_/ 3/4" _5so 561 x 38C L.:;:.::__~~-~;;;i,.:_,:~iii~=i~'~ 1,4~"~i,.l.' *:...~!!!!::~:;=B-~l>l--t;13L/!4.::"------,-t~l PUMP SEAL INJECTION 522 523 = 555. 5s7- -~L'" 563 Z - n 1/2" TEST VENT\\~ ~ w c u m ~ w c u m -*1 I I I I I I I I I I I ~ ~ ~ r-m -m--m--a,--m--m- <0 --;!a--< 0 --< 0 o 1---0 o o o o_ o o_ V V V V V V V V V V V V I 10 I V I n I N I I O ~ O> I ~ I ~ I CD I I{) I x:;:;x:;:;x:;:;x:;:;x:;:;x:;:; ~x~x~x~x~x I..--\\ Y 1 ~ " I a:, ~ I ;1i Y I 'i', N "1 o, I <O I Y1 ~ I 'i', ~ >N I,o r,..;_N >N I CD t:>N ~I ~I ~I ~I I~' >N ~I N
~
ua::, uao uao ~ 1 ua::, 1/2" 0 1 OJi ~J, uio OJi OJi L,JCD WCD WCD U<O L,JCD L,J:g WCD WCD W\\O WCD I WCD WW ~ N
, 0
>~ ~' u~ w~ FT FT FT FT F!) {Jr FT FT FT FT F~ {fr 6r 6r 6r 6r 6 ~ 6r 6r 6r 6r 6r 6 ~ 6r ~ PRIW\\RY CONTONE2 _._ (THIS SHEET) CONT ON E2 CONT ON E2 (THIS SHEET) (THIS SHEET) m N t 4 I 3 CONTAINMENT 1/2" X 3/4" 2-47E820-2, DB>-- COMPANION DRAWINGS: 1-47E817-1 J-47E817-1 2-47E817-2 AMENDMENT 26 POWERHOUSE UNIT 2 BROWNS FINAL FERRY SAFETY NUCLEAR ANALYSIS PLANT REPORT NUCLEAR BOILER FLOW DIAGRAM FIGURE 4.3-2a SH 2 G F E D C B A
L+O~ l-Ll83Lt-£ ~ L9 ~1-~ RPV HEAD VENT 2* rr, ~ -, N 1-1/2" g 1 2 2" 10-501 = 1 0-500 N WATER SEAL BETWEEN RPV AND PRIMARY / CONTAINMENT-_,, N -, VENT LINE L---t<J4-,3~ FOR RPV DRAIN CONDENSING POT7 ,--/ 2" . ~---1\\.J 10~ TO MAIN STEAM N 2" 3-47E803-1, ES I <..:3~47'.!E:!8~0~1:_1!._:,~E~J __ r,*~~';'---' 10-502 < 3-47E803-5, F3 HIGH PRESSURE ~---------,1 SEAL LEA.KOFF 3 +7E803 5, FJ (Fcv\\
05) J~~A 112* 3~
9 1/2" N ~ m w ~ ~ (Fcv\\ 3~8 1.. I LOW PRESSURE ---t---J SEAL LEA.KOFF - L - L -- L - L - 1.. 1
1..
1.. 1.. 1.. 1
1..
I I I ) 1 I I I X-30F -, X 300... I X-498 -. X 49A -. X 34F -, 1 X 490 -, 1 "34E -. X-49C -. ~ CRW... f--'--~:,--,..,t-----' ~ -~ +-~----IN9 28" RECIRCULATION LOOP A NIA TO SAMPLING < 3-47E610-43-1, H4+-""""-~ FROM RHR SYSTEM --------=2~4-"~ CONT ON l'""... --(~N2F 3-47E811-1,F4 1 ~ 513 '1 2" & ~ 3;E4" _1-... --i/-,IN2G FLOOR DRAIN 3/4" m N 8 28" 12* I Ji L J Ji ~,::5/)--1.-+1-2-.--r,I N2H 6602 3/4" SEE 6603 NOTE 16 S1 FLOOR ORAIN SEAL FLANGE COVER DR71.. 314 " CRW{,;;*~rs l TO RHR SYSTEM, SEAL,.."; / 68-55 CONT ON ',-~---ILEAKAG~ 3-47E811-1.F5 DRAIN 1-1/2" "'\\. ~-j J.DRIVER *oUNT I >--:-"3/4" °TORAIN '\\J ,... r:;-i a...::=-i FLOOR / ~~W DRAIN 8-2 RECIRCULATION III PUMP A ./ I-~~~---~ ifil,di ~ , ECKV-68-93 1.-...l. \\.:Q-to::.Jl------...2.' :_" (l,c;:x)-,..'-' E-l,1o
1* I
'f-37 1.. ~38 504 314: f m N (_e_s_ "\\ '---" 225A "1--,C:O:l---.!,l_:"(l,c:=X)--3_1 F-II ~ ECKV-68-94 226A
1* I.
1 " ~ /,.p--;;--T,, ECKV-68-65A~ (L) 1" X-33A 1h '-,2.,3>191--t2o40... r:\\:J r---( 68-65 J ,1-T'" I....:;_....,,,,,,....:;__-='-.J 213A I PI E'---_!l_:"~t:X=*}5_:0..:E __,3 249 rECKV-68-64 X-31C 250 2" TW 1-,.... _. 68-6
L1.J 1 r ~....
~ ~,j:; ~ ~ _j
If)
DECONTN ~ CONN 2 CRW 0 ~ N I PI 68-6JB Y 68-648 ~ J ~ PT \\._L,..i, '-i~t----___J1_:"£C::)---j':I 68-64 1 11 1t--------------------------.L..-...j
,-, T 217A I
I ECKV 68 63----....... 1" X-31 A iii I t--------------------------~:_ 68-63 216A ECKV-68-65Bi'j,""'\\ (HI 1 * 'I' I X-338 I It--------------------------' CRW ~ N ~ N 214A ~ECKV-68-5A I X-32A
p
~~~L---ci:f-./P. _((LL~):i--....'..' ~ .. J,t:::>----1':1*1-------------------------~--' ' -F; I F; I ;~ ',---Y...1._-cq,-t(xH,:l]----2-0_0_, ___ ,'...*-1* ::t,;x::->-'-,_"--1 I , Y68-50 68T-5~ 6~5~ 201 A ."1-------;,:;;.~-:--;T~E;ST~~ _=::::.;;;.==;.:==:;--c[z]---+--Ll]~=i==c,Il1 _ __. '-ECKV-68-_5B_ X-37CPUMP SEAL INJECTION 3/4* 565 564 I - t-... --V-,t N2J 12
I
._... _-,IN2K rn-1 0. -~- w onw U,,- wo >z 3 I FROM CRD lr-3--4-7_E_8-20---2-.-F,-"")>-~ 552 550 TEST 50; 507 1" (TYP) 12 LINES TO NOZZLE NBA~ ,\\ mCu...d.o.r::iw J::2~J~ I I I I I ~-~-~-~-gi Ol..-1Nll'll l ~ X ~ X ~ X ~ X ~ X 3/4" VENT REACTOR RPV-68-1000 ,,, j\\\\ N15 N V ~ 0 ~ ' 0 ~ al 0 1 PRIMARY CONTAINMENT FROM TW 1-179 I 3-47E801-1, 83 FROM TW 1-5 I 3 47[801 1, B+ FROM TW 1-4 lr:,-4-:7-:E"ao,..,,....,., -. -=e-:,---,, FROM TW 1-23 1r-,-_-41_E_8_0_,-_,-.-c-4""""'>----------, FROM TW 1-22 I 3-47[801-1, CJ FROM TW 1-19 I 3-47[801-1, D2 FROM TW 1-18 1r:,-_4-:1-:E"80,..,,.._.,.1-. -=0.,.1---,':,-------, FROM TW 1-30 I 3-47[801-1. E1 FROM TW 1-31 I 3-47[801 1, E2 FROM TW 1-34 1r:,-4-:7-:E"8o,..,,....,.,-, -:Fe:,---,, FROM TW 1-42 lr-,-_-4 7_E_8_0_1--1-.-F-4""""'>----, FROM TW 1-41 I 3-47E801-1, F2 FROM TW 1-180 I 3-47E801-1. FJ TO PDT 3-51 N1 0 i---rn t:=:'.:;}::;--,-----\\' STANDBY LIQUID CONTROL SYSTEM Ji 'CONT ON 3-47E854-1,87 1 3-47E803-5, F2 ~TYPICAL FOR ~~T CONT ON THIS DWG 13 RELIEF y68-95 68-96 VALVES ~?~~868 FT-68-158(A6), FT-68-18(A6) ,11,o._:X::::2~7~8:c:::::::r::1~:::it=ii:,,.J...-L._ __ -;---,---,jFT-68-19(A6), FT-68-21(A5)

111r l

FT-68-22(A6). FT-68-25B(A7) __Jlo x-52F 1 1 2o3A FT-68-26CA6l. FT-68-28CA6) r FT-68-29(A6). FT-68-30(A7) I' I 202A ECKV 68 78 1 " 297 296 1" PRESS. BELOW CORE PLATE I 1" PRESS. BELOW CORE PLATE 1.. I 231 232 1" PRESS. ABOVE CORE PLATE CONT ON THIS owe -::-::-ilof-_.!_X~2:.'.;7 A~~<Jj~~-;;j~I N1 B RECIRCULATION LOOP B 28" I' 204A ECKV !"".... ---------"""""""'"'""""""'""'""""'"-----------~~--... ~~~ 52 FT-68-7B!A4). FT-68-10 M), '---I FT-68-13 AS). FT-68-8(.6.4) FT-68-11(A4) FT-68-388(A4) FT-68-40('3) FT-68-43(A4) 1-CTJ 12" N2Ei,-_........, '1 12"j N20~-...... -t ~ 12" N2C '\\J I 12" N281~'-.J-----I I N2AL2' 28" 24", FROM RHR SYSTEM ._..a.a._., CONT ON '3-47E811-1,F6 CONTROLLED SEAL BLEEDOFF \\,_ FLOOR DRAIN 3/4" 6605 3/4" 6606 \\.. FIS .--~---<68-74!---~- 5---,1 ,12.J \\~EE \\NOTE 16

1i.... FLOOR

~DRAIN 1 * .--- SEAL FLANGE COVER DR ~ SEAL LEAKAGE OR PDT) 68-52 FT-68-39(A3), FT-68-42(A3), TO POIS 75-56 f-------i3-47E814-1, G6 TO POIS 75-28 '---,3-47E814-1, G6 TO PDT 85-17 AND -18 3-47E820-2,H3 & G5 TO PI 85-19 3-47E820-2, F5 FE 68-j1 1,,....__., 3/4~-;{ ( ~3/4" ) 6~8 } CRW 1-1/2" CRW (Fcv\\ 68-79 ~* r DRIVER MOUNT DR IB---1~ / FLOOR DRAIN 10-506 THRU 10-516 10-519 AND 10-520 J/4" ~,A L..:2~*:.:":...i:::!cc+d*"'~-.....J,......,, 'TW\\ I III RECIRCULATION .1 ~*,u- .1 ~*-"""cl,_ a 1 2-1/2", 1 O" m.. N w.. N VENT m N N 28" 2" Ji_. CRW 520 DECONTN CONN TW 68-83 LU I 10-521 THRU, 10-533 =- 10 X-33C ( , I 12
PRESSURE SUPPRESSION CHAMBER 1" (L)
.(POT L-'--[2:X2a,:AJ---E-CK-v-C_!:6.. 8--8-2-A-----i 68~ 2
.::5::0:.,F C'.:'.JG---3
(;;'\\ B (TYP) I J ECKV SEE NOTE 13 c l ~6E -@, 1"
IL __.:,X~3~1~0c::::o---:i~q-_:6~80-~7~6=-( PT B
l---------------'--l*11 11 r 219A ECKV 68-76 1" ,h )( 318 68 75 / PT
I I
1"
I
)(330 II ( 218A (H) ECKV rn -~8 820* 1
8-75 X-2340 180 X-234E X-235D X-235E TW
./---II--......... TW 64-161E (rw\\ X-234F X-235C 64-1610 TW (rw\\ 64-161 F -235F X-234C 90 6'4-1620 64-162[ w 64-161C w 64-1628 w ~4-161G 64-162G X-234G X-2358 0w\\ 0w\\ (nv\\ 64-162A 64-16 H (fii"' ~1618 64-16~ ":iJSG X-2348 (ni'\\ 64-161A X-234H X-235A X-235H X-234A TORUS TEMPERATURE MONITORING SYSTEM SYSTEM PRESS.- TEMP DATA LINE DESIGN DESIGN MK NO. PRESSURE (PSIG) TEMP(°F) 2 3 4 5 6 7 8 1146 150 1148 1326 150 1500 1750 1250 562 300 562 562 212 575 150 575 TABLE A NOTES:
1.
ALL VALVES ARE THE SAME SIZE AS PIPING UNLESS OTHERWISE NOTED.
2.
ALL VALVE BODY DRAINS, BONNET VENTS AND PACKING LEAKOFFS ARE 1/2" UNLESS OTHERWISE NOTED.
3.
DELETED
4.
ALL DRAINS SHOWN I ~... INOICATES CLOSED SYSTEM.
5.
CRW DRAINS ARE CONNECTED TO THE REACTOR BUILDING EQUIPMENT DRAIN SUMP.
6.
0JETC, DENOTES DESIGN PRESSURE AND TEMPERATURE AS GIVEN IN TABLE (A) SYSTEM PRESS-TEMP DATA THIS DRAWING.
7.
HYDROSTATIC TESTING SHALL BE IN ACCORDANCE WITH THE APPLICABLE CODES.
8.

9.

10.

11.

12.

13.

14.

15.
THE DESIGN PRESSURE AND TEMP OF ALL DRAIN AND VENT LINES THROUGH THE LAST ISOLATION VALVE SHALL BE THE SAME AS THE PROCESS LINE. ALL VALVES ARE PREFIXED "3-68-" AND ALL INSTRUMENTS ARE PREFIXED "3-" UNLESS OTHERWISE NOTED. (H) - IS HIGH PRESSURE AND (L) - IS LOI PRESSURE SIDE. VESSEL "O" (VESSEL INVERT AT 578'-3" BUILOING ELEVATION SEE CALCULATION CD-Q0303-940391. UNIDS ON DRAWING ARE FOR REFERENCE ONLY AND ARE ABBREVIATED TO MEET SPACE CONSTRAINTS, REFER TO MEL FOR COMPLETE UNIDS. ALL SENSING LINES WITH EXCESS FLOW CHECK VALVES EXCEPT X-40A-a THROUGH X-40D-f HAVE AN ORIFICE COUPLING. (REFER TO DETAIL P20 ON 0-47W600-20). VENT, DRAIN, AND TEST CONNECTIONS 1-1/2" AND BELOW CAN BE PROVIDED WITH PIPE CAPS OR HOSE CONN(CTION FITTINGS WHERE REQUIRED BY PLANT PERSONNEL. THIS CONFIGURATION IS SUPPORTED BY ENGINEERING CALCULATION CD-00999-923399. OPERATIONS MAY CLOSE VALVE (3-SHV-10-100) AS NECESSARY TO REDUCE INLEAKAGE.
16. SEAL FLANGE COVER DRAIN IS VENTED THROUGH AN OPEN-ENDED "TEE" CONNECTED TO A DRIP-LOOP.
REFERENCE DRAWINGS: 0-47E800-1.......... FLOW DIAGRAM-GENERAL PLANT SYSTEMS 0-47E800-2.......... MECHANICAL SYMBOLS & FLOW DIAGRAM DRAWING INDEX 3-47E811-1.......... FLOW DIAGRAM-RHR SYSTEM 3-47E814-1.......... FLOW DIAGRAM-CORE SPRAY SYSTEM 3-47E610-75-1....... MECHANICAL CONTROL DIAGRAM-CORE SPRAY SYSTEM 3-47E610-3-1,....... MECHANICAL CONTROL DIAGRAM-REACTOR FEEDWATER SYSTEM 3-47E812-1.......... FLOW DIAGRAM-HPCI SYSTEM 3-47E803-1.......... FLOW DIAGRAM-REACTOR FEEDWATER SYSTEM 3-47E813-1.......... FLOW DIAGRAM-RCIC SYSTEM 3-47E820-2.......... FLOW DIAGRAM-CRD HYDRAULIC SYSTEM 3-47E810-1.......... FLOW DIAGRAM-REACTOR WATER CLEANUP SYSTEM 3-47E610-68-1....... MECHANICAL CONTROL DIAGRAM-REACTOR WATER RECIRN SYSTEM 3-47E610-43-1....... MECHANICAL CONTROL DIAGRAM-SAMPLING & WATER QUALITY SYSTEM 3*47[610-85*1....... MECHANICAL CONTROL DIAGRAM-CONTROL ROD DRIVE SYSTEM 3-47E801-1.-2....... FLOW DIAGRAM-MAIN STEAM 3-47E610-1-SERIES... MECHANICAL CONTROL DIAGRAM-MAIN STEAM 3-47E610-74-SERIES.. MECHANICAL CONTROL DIAGRAM-RHR SYSTEM 3-47E610-73-1....... MECHANICAL CONTROL DIAGRAM-HPCI SYSTEM 3-47E610-71-1....... MECHANICAL CONTROL DIAGRAM-RCIC SYSTEM 3-47E610-69-1....... MECHANICAL CONTROL DIAGRAM-REACTOR WATER CLEANUP SYSTEM 3-47E854-1..,....... FLOW DIAGRAM-STANDBY LIQUID CONTROL SYSTEM 3-47E610-63-1....... MECHANICAL CONTROL DIAGRAM STANDBY LIQUID CONTROL SYSTEM 3-47E822-1.......... FLOW DIAGRAM-CLOSED COOLING WATER SYSTEM MEL................. INSTRUMENT & VALVE MARKER TAG TABULATIONS 3-47E852-2.......... FLOW DIAGRAM - CRW DRAINAGE 3-47E803-5...,.,.... MECHANICAL RPV LEVEL SENSING LINES INSTRUMENTS AND CONTROLS MEL................. INSTRUMENT TABULATION FOR SYSTEM 3 3-47E867-3.......... FLOW DIAGRAM-SAMPLING AND WATER QUALITY SYSTEM AMENDMENT 27 1 "
1.._ __ !x:::a~2:!:cC~1~1,G-:!<~H~ll--2_2_'.... A0E~~;~v~8~1~e!l(r FT
@T,@Tl FT
1r I
223A(Ll ECKV ~8-8~A 68-818 s-s1c 68-81D 1 N ,1L __ ~x:::,~2~a:c:::::::o---:i::c,::i----l-:i";:"-:::"~1c.:Ac... ..,.....1... __ .1... __ 1.._ _ _J L--'------------------------r~~~~-~--~~~i~-;,"j.~~3--l.,11 POWERHOUSE UNIT 3 ( 1" (TYP) 12 LINES 3 7 -3/4" TEST 224" FROM CRO TO NOZZLE N88 _..J---~-l:o-W:J--...l.-j*""'~5.:_60T'°"::::56~1~::::~'~4'... -=--~X:*=38~C:c;r I I *~ .:l"~~r---<"~3::'.4~7~E:'.8~2~0.::!2:., _'.G~1 _ _j PUMP SEAL INJECTION I ~ 557 522 523 3/4* TEST 555 562 563 - 9----?--?--? ?.\\J 8 §-5!- 0 g +-: ...-1 C<\\11 l'"J1 +I IOI IOI CIOX ODX COX COX IX>X COX N N N <'I N N 556 3/4" VENT ""--- PRIMARY CONTAINMENT r N N~ BROWNS FERRY FINAL SAFETY NUCLEAR ANALYSIS PLANT REPORT NUCLEAR BOILER 1 /2" TO P.A.S.S. (REACTOR 3_43_129 RECIRC SAMPLE) ~* ,m mW ~,:. ,.. 1, COMPANION DRAWINGS: 1-47E817-1 2-47E817-1 & -2 3-47E817-1-ISI FLOW DIAGRAM 3 .!:!:. w TO LT 62 & LT 62A r CONT ON 3-47E803-5,810 3-47E867-3, C1 FIGURE 4.3-2a SH CONT ON THIS DWG CD2 8 I 7 I 6 I 5 t 4 I 3 H G F E D C B A
BFN-16 Figure 4.3-2b Deleted by Amendment 11.
SUCTION~ I DRIVING FLOW~--L _ _ _ I FLOI Yi/} I -,.. I V, I I I 1 l I I 1 ~--~----~~~----~--~~ I l ' l D~NG I FLT I' I FPO DRtVING FLOI t.t.P I r r I SUCTION FLOW AP AMENDMENT 16 BROWMS FERRY HUCLUR Pl.MfT FINAL SA'F£TY ~HAL '!"SIS REPORT J111 Pump...()pe,1ting Principia FIGURE 4.3-3
NORMAL STEAM SEPARATORS WATER '-i----roi-i.-~ST""'=E!"""A~M--t LEVEL SEPARATION r-----~ r-1 I I I 0 : '---i- _) ,*:V DISTRIBUTION
PLENUM WATER
--, _A1LEVEL ~ ,J./'A AFTER

~ I BREAK

). . :~ h"'i ~ ACTIVE CORE )i... _f2~~;:;~~~~~. i -~ 1...
.*_. ::r:

.*.:::-.*.:::_=.*_*'..**
tt':t:~!!lti~f/2:. AMENDMENT 16 BROWNS FERRY NUCLEAR PLANT FIN AL SAF ETV ANALYSIS REPORT &ecirculation System - Core Flooding Capability Figure 4.3-4
BFN-26 4.4-1 4.4 NUCLEAR SYSTEM PRESSURE RELIEF SYSTEM 4.4.1 Safety Objective The safety objective of the Nuclear System Pressure Relief System is to prevent overpressurization of the nuclear system; this protects the nuclear system process barrier from failure which could result in the uncontrolled release of fission products. In addition, the automatic depressurization feature of the Nuclear System Pressure Relief System acts in conjunction with the Emergency Core Cooling Systems for reflooding the core following breaks in the nuclear system process barrier; this protects the reactor fuel barrier (UO2 sealed in cladding) from failure due to overheating, which would result in the uncontrolled release of fission products from the reactor fuel barrier. 4.4.2 Power Generation Objective The power generation objective of the Nuclear System Pressure Relief System is to relieve normal overpressure transients occurring during normal plant isolations and load rejections. 4.4.3 Safety Design Basis
1.
The Nuclear System Pressure Relief System shall prevent overpressurization of the nuclear system in order to prevent failure of the nuclear system process barrier.
2.
The Nuclear System Pressure Relief System shall provide automatic nuclear system depressurization, if needed, for breaks in the nuclear system so that the Low Pressure Coolant Injection (LPCI) and the Core Spray Systems can operate to protect the fuel barrier. This depressurization is permissive on: (1) concurrent high drywell pressure and low reactor water level, or (2) sustained reactor low water level, and (3) availability of one of the RHR pumps in the LPCI mode or two of the appropriate core spray pumps.
3.
The main steam relief valve (MSRV) discharge piping shall be designed to accommodate forces resulting from relief action and shall be supported for reactions due to flow at maximum MSRV discharge capacity so that system integrity is maintained. The MSRV discharge piping shall be routed to the pressure suppression pool.
4.
The Nuclear System Pressure Relief System shall be designed for testing prior to nuclear system operation and for periodic verification of the operability of the Nuclear System Pressure Relief System.
BFN-26 4.4-2 4.4.4 Power Generation Design Basis
1.
The nuclear system main steam relief valves shall not discharge to the primary containment drywell.
2.
The main steam relief valves shall properly reclose following a plant isolation or load rejection, so that normal operation can be resumed as soon as possible.
3.
The capacity of the main steam relief valves shall be sufficient to prevent reactor pressure from exceeding the allowable overpressure of ASME Boiler and Pressure Vessel Code, Section III, during an isolation transient with indirect scram. 4.4.5 Description The Nuclear System Pressure Relief System includes 13 main steam relief valves, all of which are located on the main steam lines within the drywell between the reactor vessel and the flow restrictors. The main steam relief valves provide three main protection functions:
1.
Overpressure relief operation. All 13 main steam relief valves can be opened manually from the main control room or are self-actuated to limit the pressure rise.
2.
Overpressure safety operation. The valves are opened (self-actuated) to prevent exceeding the design allowable stress limits on the reactor vessel and associated piping.
3.
Depressurization operation. Six of the 13 valves are available to be opened automatically as part of the Emergency Core Cooling System (ECCS). The main steam lines, in which the main steam relief valves are installed, are designed, installed, and tested in accordance with USAS B31.1.0, 1967 edition, and the applicable GE design and procurement specifications, which were implemented in lieu of the outdated B31 Nuclear Code Cases-N2, N7, N9, and N10. The main steam relief valves are distributed among the four main steam lines so that an accident cannot completely disable a safety, relief, or automatic depressurization function. (See Figure 4.3-2a sheet 1 of Subsection 4.3 and Figures 11.1-1a, 11.1-1c, and 11.1-1e of Subsection 11.1 for schematic location, and Figures 4.5-1, 4.5-2, and 4.5-3 of Subsection 4.5 for layout details of the valves and piping.)
BFN-26 4.4-3 The design and installation of the main steam relief valves include the following:
a.
Clearance of at least 6 in. is provided between valves and other equipment (excluding MSRV pilot solenoid valves),
b.
Space is provided between all welds on the header for inspection greater than 2t + 2" (where t is minimum wall thickness),
c.
Clearance is provided between header and bottom of flange for bolt removal when valve is installed,
d.
A flange rating of 1500 lb. was provided for structural stability instead of a 900 lb.-rated flange required for pressure-temperature rating,
e.
An inlet pipe Schedule 160 was used for structural stability instead of Schedule 80 required for pressure-temperature rating, and
f.
The discharge piping provides for equalization of discharge thrust forces. For analysis, the special loadings listed below are considered in addition to the usual design loads such as weight, pressure, temperature, and earthquake:
1.
The jet force exerted on the main steam relief valves during the first millisecond when the valve is open and steady-state flow has not yet been established. (With steady-state flow, the dynamic flow reaction forces will be self-equilibrated by the discharge piping.)
2.
The dynamic effects of the kinetic energy of the piston disc assembly when it impacts on the internals of the valve. All code-allowable stresses are met with these special loads acting concurrently with other design loads. The highest stress is at the branch connection to the header. The results of this analysis are contained in Appendix C, Table C.4-2. The main steam relief valves are designed, constructed, and marked with data in accordance with the ASME Boiler and Pressure Vessel Code, Section III, 1968 edition and addenda through summer 1970 for the two-stage valves. Setpoint tolerance (pressure at which valve "pops" wide open) is in accordance with ASME Boiler and Pressure Vessel Code, Section I, paragraph PG-72(c). Pressure-containing parts of the valve body are fabricated of ASTM A216, Grade WCB. The main steam relief valve is designed for operation with saturated steam containing less than 1 percent moisture. The relieving pressures for overpressure
BFN-26 4.4-4 relief and safety operating modes are adjustable between 1025 and 1190 psig, with a maximum backpressure of 40 percent of the set pressure. The lowest MSRV setpoint has been raised to 1135 psig. This serves to alleviate the "simmering" problems that contribute to valve failures. Also, the bore size of the valves has been increased slightly (from 4.94 or 5.03 inches to 5.125 inches) to accommodate more relief capacity. The delay time (maximum elapsed time between overpressure signal and actual valve motion) and the response time (maximum valve stroke time) are less than 0.5 second total. Each valve is self-actuating at the set relieving pressure, but may also be actuated by remotely-operated devices to permit remote-manual or automatic opening at lower pressures. The remote air actuators are controlled by DC powered solenoid valves. The power actuated device is capable of opening the valve at any steam pressure above 50 psig and is capable of holding the valve open until the steam pressure decreased to about 20 psig. The solenoid valves are normally closed, fail-closed valves, and a power or valve malfunction will prevent the main steam relief valve from operating for Automatic Depressurization System (ADS). Abnormal solenoid-valve operation would be detected during the operational tests of the main steam relief valve. A complete rupture of the solenoid valve would result in a low air pressure/accumulator alarm. Each of the six main steam relief valves provided for automatic depressurization is equipped with an air accumulator and check valve arrangement. These accumulators are provided to assure that the valves can be held open following failure of the air supply to the accumulators, and they are sized to contain sufficient air for a minimum of five valve operations. To ensure an emergency supply of air is available to provide for the five valve operations under accident conditions, an accumulator leak test is performed once per operating cycle. The first and second actuations are assumed to occur with drywell pressure at 35 psig and subsequent actuations with the drywell at 0 psig. Redundant sources of pneumatic pressure are provided by the Drywell Control Air (DCA) and Containment Atmospheric Dilution (CAD) systems. Accumulators are not required for the main steam relief valves not used for automatic depressurization. The main steam relief valves which are a part of the ADS normally receive their motive air from the drywell control air system. The air pressure in each accumulator is continuously monitored by a pressure switch which annunciates in the control room on low air pressure. The pressure switch, in order to ensure operability, is calibrated and functionally tested once per operating cycle. The drywell control air system is also continuously monitored for low air pressure by means of a pressure switch located in the system downstream of the receivers and which annunciates in the control room. A manual transfer can also be made to the plant control air system as another backup for control air. The main steam relief valves are designed to operate under maximum prevailing operating conditions and postulated accident conditions in the drywell. In addition, the ADS
BFN-26 4.4-5 accumulators and piping up to and including the isolation check valves are seismically qualified and capable of performing their functions during and following an accident. The automatic depressurization feature of the Nuclear System Pressure Relief System serves as a backup to the High Pressure Coolant Injection (HPCI) System under loss-of-coolant accident conditions. If high drywell pressure and low water level persist and one of the low pressure coolant injection (LPCI) pumps or two of the appropriate core spray pumps are available, the nuclear system is depressurized sufficiently to permit the LPCI and Core Spray Systems to operate to protect the fuel barrier. Depressurization is accomplished through automatic opening of some of the main steam relief valves to vent steam to the pressure suppression pool. For small line breaks, if the HPCI system fails, the nuclear system is depressurized in sufficient time to allow the Core Spray or LPCI Systems to provide core cooling to prevent excessive fuel clad temperatures. When HPCI is considered to be the single failure, six ADS valves are required to meet the requirements for ADS. As shown in Table 6.5-3, dependent upon the recognized single failure, between four and six valves remain available and the results of LOCA analyses confirm that the requirements for ADS continue to be met. For large breaks, the vessel depressurizes rapidly through the break without assistance from ADS. Discharge pressure indication of one LPCI pump or two core spray pumps combined with one of the following initiation paths will cause the main steam relief valves to open: (l) reactor vessel low water level and primary containment (drywell) high pressure in conjunction with a 120 seconds timer timed out; or (2) sustained reactor low water level for 360 seconds. Further descriptions of the operation of the automatic depressurization feature are found in Section 6.0, "Emergency Core Cooling Systems," and Subsection 7.4, "Emergency Core Cooling System Control and Instrumentation." The Automatic Depressurization System is designed as seismic Class I equipment in accordance with Appendix C. A manual depressurization of the nuclear system can be effected in the event the main condenser is not available as a heat sink after reactor shutdown. The steam generated by core decay heat is discharged to the pressure suppression pool. The main steam relief valves are operated by remote manual controls from the Main Control Room to control nuclear system pressure. The number, set pressures, and capacities of the main steam relief valves are shown in Table 4.4-1a (it should be noted that the three percent tolerance is for analytical purposes only). Actual MSRV opening setpoints following testing must still be set at nominal values one percent. The original three-stage Target Rock valves (Model 67F) have been changed to two-stage valves (Target Rock MSRV model No. 7567F) to minimize spurious openings and to respond to NUREG 0737, Item II.K.3.16.
BFN-26 4.4-6 Two-Stage Valve Operation The Target Rock pilot-operated main steam relief valve (Model 7567F) consists of two principal assemblies: a pilot stage assembly and the main stage assembly (refer to Figure 4.4-1). These two assemblies are directly coupled to provide a unitized, self-actuated safety/relief valve. The pilot stage assembly is the pressure sensing and control element and the main stage assembly is a hydraulically (system fluid) actuated follower valve which provides the pressure relief function. Self-actuation of the pilot assembly at set pressure vents the main piston chamber, permitting the system pressure to fully open the main assembly. The pilot assembly consists of two relatively small, low-flow, pressure-sensing elements. The spring loaded pilot disc senses the set pressure, and the pressure-loaded stabilizer disc senses the reseat pressure. Spring force (preload force) is applied to the pilot disc by means of the pilot rod. Thus, the adjustment of the spring preload force will determine the set pressure of the valve. The main assembly of the Target Rock main steam relief valve is a reverse-seated, hydraulically-actuated angle globe valve. Actuation of the main assembly permits discharge of fluid from the protected system at the valve's rated flow capacity and provides the system pressure-relief function of the valve. The major components of the main stage are the valve body, disc/piston assembly, and preload spring. A typical sequence of operation for overpressure relief self-actuation can be described as follows (refer to Figures 4.4-1 and 4.4-2).
1.
In its normally closed position, the main stage disc is tightly seated by the combined forces exerted by the system internal pressure acting on the area of the disc and the preload spring. Note that in the closed, no-flow position, the static pressures will be equal in the valve inlet nozzle and in the chamber over the main stage piston. This pressure equalization is made possible by leakage past the piston, via the ring gap and drain and vent grooves.
2.
When system pressure increases to the valve set pressure, pilot stage operation will vent the chamber over the main stage piston to downstream of the valve via internal porting. This venting action creates a differential pressure across the main stage piston in a direction tending to open the valve. The main stage piston is sized such that the resultant opening force is greater than the combined spring preload and hydraulic seating force.
3.
Once the main stage disc starts to open, the hydraulic seating force is reduced, causing a significant increase in opening force and the characteristic full opening or "popping" action.
BFN-26 4.4-7
4.
When system pressure has been reduced sufficiently, the pilot disc reseats and precludes depressurization of the main piston chamber. Leakage of system fluid, past the main stage piston and stabilizer seat, repressurizes the chamber over the piston, canceling the hydraulic opening force and permitting the preload spring and flow forces to close the main stage. Once closed, the additional hydraulic seating force, due to system pressure acting on the main stage disc, seats the main stage tightly and prevents leakage. A remotely-controlled air operator is fitted to the pilot stage assembly to provide selective operation of the valve at system pressure other than set pressure. This is a diaphragm-type, pneumatic actuator which must be actuated to open the valve. It is actuated by means of a solenoid control valve which admits drywell control air to the air-operator piston chamber and strokes the air operator stem, in turn stroking the pilot disc via the pilot rod. The main stage then opens as described in previous paragraphs. Deenergizing the solenoid vents the air operator and permits the pilot disc to reseat. The main stage then reseats as previously described. Main Steam Relief Valve Position Indication The main steam relief valve position is monitored by two systems. A single-train acoustic monitoring system has been installed on all the main steam relief valves to provide unambiguous Main Control Room indication (and alarm) of valve position. The system responds to NRC requirements of NUREG 0578, item 2.1.3.a. The system is qualified as seismic Class I and is powered by a Class 1E power supply. There also exists a temperature sensor in the discharge piping of each valve which can be used to determine individual valve positions. Temperature indications are also provided in the control room. The acoustic monitor satisfies the valve position alarm and annunciation requirements. Refer to Paragraph 7.4.3.3.4 for additional details. (High-temperature alarm and annunciation is removed for Units 1, 2, and 3) Non Safety Related Alternate Automatic Means of Opening the MSRVs Upon Overpressurization: During inservice pressure transient events in the relief mode, safety grade pressure sensors (found in Section 7.4.3, "Automatic Depressurization System") actuate the MSRVs. This method of automatically opening the MSRV permits application of the full main steam line pressure to break the corrosion bonds that may have developed between the pilot/disc interface. When the relief mode is actuated, the setpoint spring preload is removed from the pilot disc, and a rapidly applied full differential pressure is seen across the pilot disc. This alternate means of actuation is capable of opening the MSRVs. This non-safety related automatic means of opening the MSRV is applicable for Units 1, 2, and 3.
BFN-26 4.4-8 Main Steam Relief Valve (MSRV) Discharge The main steam relief valves are installed so that each valve discharge is piped through its own uniform-diameter discharge line to a point below the minimum water level in the primary containment pressure suppression pool to permit the steam to condense in the pool. Thermal mixing in the pool during main steam relief valve blowdown is enhanced by T-quencher discharge devices at the pressure suppression pool end of the main steam relief valve discharge lines. Water in the line above pressure suppression pool water level would cause excessive pressure at the valve discharge when it is again opened. For this reason, one small check valve and one large check valve venting to the drywell are provided on each main steam relief valve discharge line to prevent drawing water up into the line, due to steam condensation, following termination of main steam relief valve operation. The main steam relief valves are located on the main steamline piping, rather than on the reactor vessel top head, primarily to simplify the discharge piping to the pool and to avoid the necessity for removing sections of this piping when the reactor head is removed for refueling. In addition, the main steam relief valves are more accessible during a quick shutdown to correct possible valve malfunctions when located on the steam lines. The discharge piping has been modified as part of the torus integrity program. This modification has been described in a letter from L. M. Mills to Harold R. Denton dated May 22, 1981. A submittal by GE (NEDO-21888, "Mark I Containment Program Load Definition Report," December, 1980), on behalf of TVA, describes the reassessment of the torus design to include pressure suppression pool hydrodynamic loads due to MSRV discharge and pressure suppression pool response. The reassessment of Mark I containments was precipitated from the large-scale testing of the Mark III containment system. Pressure suppression pool hydrodynamic loads resulting from the effect of drywell air and steam being rapidly forced into the pressure suppression pool during a postulated LOCA and/or MSRV discharge were identified, which had not been considered in the original design. The Mark I Owners Group, of which TVA is a member, and GE responded by submitting the Mark I Containment Program Load Definition Report (described above) and the Mark I Containment Program Structural Acceptance Criteria Plant Unique Analysis Application Guide (NEDO-24538-1). These reports describe the generic pressure suppression-pool hydrodynamic-load definition and assessment procedures for use in plant-unique pressure suppression-chamber design analyses. TVA has applied the load definitions and approved structural acceptance criteria to the entire torus, torus internals, MSRV piping, and attached piping of Browns Ferry Nuclear Plant.
BFN-28 4.4-9 A plant-unique main steam relief valve discharge test was performed as part of the Browns Ferry Nuclear Plant Unit 2 unique analysis, as requested by the NRC in NUREG-0661. This test did confirm the methods used to calculate containment loads from the various MSRV discharge cases. For the results of this test to be completely acceptable, all modifications which significantly influence torus motion had to be in their final configuration. The magnitude of the MSRV discharge related loads is a function of the type of discharge device used. The device found to substantially reduce the hydrodynamic discharge loads, compared to other devices, is the T-quencher developed specifically for the Mark I torus (see Figures 4.4-6, 4.4-7, and 4.4-8). The devices have been added to the MSRV discharge lines at Browns Ferry Nuclear Plant. Discharge piping and relief valves were analyzed for deadweight, thermal, seismic and relief valve blowdown loads. The support locations, orientation, and design loads satisfy ASME Boiler and Pressure Vessel Code, Section III, Class 2, equations and stress allowables. One main steam safety valve (1-501 from line B) and one safety valve (1-537 from line C) were removed and the connections blanked off with blind flanges and a relief valve was added to Main Steam Line A and Main Steam Line D. Additionally, the valve throat diameters were increased from 5" nominal size to 5.125" nominal size. These modifications increased the installed relief capacity per valve to 870,000 lbm/hr at 1090 psig. The addition of these MSRVs do not adversely affect the stresses imposed on the headers to which the valves are attached. The torus shell is also adequate for the larger discharge loads. 4.4.6 Safety Evaluation The ASME Boiler and Pressure Vessel Code requires that each vessel designed to meet Section III be protected from pressure in excess of the vessel design pressure. A peak-allowable pressure of 110 percent of the vessel design pressure is allowed by the code. The main steam relief valves are set to open by self-actuation (overpressure safety mode) in the range from 1135 to 1155 psig. This satisfies the ASME code specifications for safety valves, since the lowest-set valve opens below the 1250 psig nuclear system design pressure, and the highest-set valve opens below 1313 psig (105 percent of nuclear system design pressure). The setpoints are also set high enough to avoid MSRV simmering problems. Main steam relief valve capacity to support operation at the original licensed thermal power level of 3293 MWt was determined by analyzing the pressure rise accompanying the main steam flow stoppage resulting from a 3-second main steam isolation valve closure initiated from turbine-generator design operating conditions. The analysis hypothetically assumed the reactor is shut down by an indirect scram. The analysis indicated that the main steam
BFN-28 4.4-10 relief valve capacities provide sufficient flow to maintain an adequate margin below the peak ASME code-allowable pressure in the nuclear system (1375 psig). Figure 4.4-3 is representative of the nuclear system response which might be expected during such a transient. The required main steam relief valve capacity is currently determined for each core reload by analyzing the pressure rise accompanying the main steam flow stoppage resulting from a three second main steam isolation valve closure initiated at an initial dome pressure of 1055 psig (corresponding to the Improved Technical Specifications LCO value of 1050 psig plus 5 psi margin). The analysis hypothetically assumed the reactor is shutdown by a high neutron flux scram signal (i.e., failure of the MSIV position direct scram signal). For the analysis, the self-actuated setpoints of the 12 main steam relief valves were assumed to be as shown in Table 4.4-1a (one MSRV with the lowest opening setpoint is assumed inoperable). The analysis indicated that the main steam relief valve capacities shown in Table 4.4-1a provide sufficient flow to maintain an adequate margin below the peak ASME code-allowable pressure in the nuclear system (1375 psig). Additional discussion and results of this overpressurization analysis are documented in Chapter 14. The results of the specific analysis for each unit can be found in the current reload licensing analysis for that unit (Appendix N). The sequence of events assumed in this analysis was investigated only to meet code requirements for pressure-relief-system evaluation Evaluations of the automatic depressurization capability of the Nuclear System Pressure Relief System are presented in Section 6.0, "Emergency Core Cooling Systems" and Subsection 7.4, "Emergency Core Cooling System Controls and Instrumentation." The piping attached to the main steam relief valve discharges was initially designed, installed, and tested in accordance with USAS B31.1.0, 1967 edition and the applicable GE design and procurement specifications, which were implemented in lieu of the outdated B31 Nuclear Code Cases-N2, N7, N9, and N10. New analyses of the main steam system and MSRV discharge piping have been performed in accordance with ANSI B31.1, 1973 edition, with Addenda up to Summer 1975. This analysis included deadweight, thermal, seismic, and main steam relief valve blowdown loadings. Snubbers have been added to reduce stresses in the main steam and main steam relief valve piping.
BFN-26 4.4-11 4.4.7 Inspection and Testing The main steam relief valves were tested in accordance with the manufacturer's quality control procedures to detect defects and prove operability prior to installation. The following final tests were witnessed by a representative of the purchaser:
a.
Test at USAS-specified hydrotest pressure using nitrogen, and
b.
Nitrogen leakage test at design pressure with a maximum permitted leakage of 2cc per inch of seat diameter per hour. The main steam relief valves were installed as received from the factory. The setpoints were adjusted, verified, and indicated on the valves by the vendor prior to shipment. Proper manual and automatic actuation of the main steam relief valves was verified during the preoperational test program. It is recognized that it is not feasible to test the main steam relief valve setpoints while the valves are in place or during normal plant operation. The valves are mounted on 6-inch-diameter, 1500-pound, primary service rating flanges so that they may be removed for maintenance or bench checks and reinstalled during normal plant shutdowns. The external surface and seating surface of all main steam relief valves are 100 percent visually inspected when the valves are removed for maintenance or bench checks. Operational tests of the main steam relief valves are performed once per operating cycle by means of an automatic actuation of the ADS valve logic circuitry from a simulated or actual initiation signal and by means of a manual actuation of all the main steam relief valves until thermocouples or acoustic monitors downstream of the valves indicate steam is flowing from the valve. This can also be demonstrated by the response of the turbine control valves or bypass valves, by a change in measured steam flow, or by any other method suitable to verify steam flow. Main steam relief valves are removed and bench-tested following each operating cycle. The testing procedures include criteria for set pressure and seat leakage to determine valve acceptability. Monitoring and recording of valve stroke time, disc lift, and blowdown reseat pressure are included in the test to determine proper valve operation. Bench-testing is also required following any activity that will affect valve operability or set pressure prior to installing the valve. During unit operation, discharge tailpipe temperatures and acoustic monitors are monitored and evaluated to determine if the valves are leaking excessively.
BFN-26 4.4-12 In response to NUREG-0578, Item 2.1.2, "Performance Testing for Relief and Safety Valves," TVA elected to participate in the BWR Owners Group Test Program of the safety/relief valves. The test program addressed those conditions that could result in single-phase liquid or two-phase flow through the safety/relief valves at low-pressure conditions. The results of the tests are summarized in the BWR Owners Group S/RV Test Program Final Report, entitled "Analysis of Generic BWR Safety/Relief Valve Operability Test Results," NEDO-24988, submitted to D. G. Eisenhut by T. J. Dente, September 25, 1981. The tested valves satisfy the acceptance criteria for operability; and therefore, the operational adequacy of the Browns Ferry MSRVs has been demonstrated.
BFN-22 TABLE 4.4-1 (Deleted by Amendment 22)
BFN-22 TABLE 4.4-1A NUCLEAR SYSTEM MAIN STEAM RELIEF VALVES Units 1, 2, and 3 Number Set of Pressure Capacity at Set Valves (psig) Pressure (each),(lb/hr) Main Steam Relief 4 1135 (+3%) 905,000 Valves 4 1145 (+3%) 913,000 5 1155 (+3%) 921,000
OUTLET AIR OPERA TOR ASSEMBLY SOLENOID VALVE BFN-17 ASSEMBLY~--.---..__ ___ ~-* PILOT VALVE DISCHARGE PORT MAIN PISTON(OISC PRELOAD SPR NG SETPOINT SPRING ASSEMBLY --~.,l,,;l,...i,=..--plLOT VALVE DISC (SEATED} + * ~ * ~ - -, i' I 4- ..f ~ p f ~ I f
If
.* *.*.*.*.~-* -*.*-* AMENDMENT 17 BROWNS FERRY NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT 2-STAGE SAFETY/RELIEF VALVES SCHEMATIC (CLOSED POSITION) FIGURE 4.4-1
I OUTLET AIR OPERATOR ASSEMBLY BFN-17 SOLENOID VALVE ASSEMBLY~~----.------...._~ MAIN PISTON/DISC PRELOAD SPRING~ ~... ~..... ' 1 ,4. 9 ~ ~ * ,I
~*-*.*~*.*-*.*.*-*.*
AMENDMENT 17 BROWNS FERRY NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT 2-STAGE SAFETY/RELIEF VALVES SCHEMATIC (OPEN POSITION) FIGURE +.+-2
NOTE: This figure is representative of ~he nuclear sy&tem response. See current reload amendment for ~p-to-date system response. AMENDMENT 16 I I I BROWNS FERRY NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT S.ft'lY ValYI Sizing Analysit FIGURE 4.4-3
BFN-16 Figure 4.4-4 Deleted by Amendment 13.
BFN-16 Figure 4.4-5 Deleted by Amendment 13.
AMENDMENT 17 =~=1£7:-~-r,~-;~oo,;,....=~~~~~---, ,_,____ BROWNS FERRY NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT T-OUENCHER FOR SAFETY/RELIEF DISCHARGE FIGURE 4.4-6
AMENDMENT 17 BROWNS FERRY NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT MECHANICA.LMA!NSTEAM RELIEF VALVE VENT PIPING FIGURE4.4-7
AMENDMENT 25 BROWNS FERRY ~~;t~~~/~~~6RT FINAL SAFETY MECHANICAL ~:c~ES~~~¥ ~i~i~b FIGURE 4.4-8
BFN-21 4.5-1 4.5 MAIN STEAM LINE FLOW RESTRICTOR 4.5.1 Safety Objective To protect the fuel barrier, the main steam line flow restrictors limit the loss of water from the reactor vessel before main steam isolation valve closure in case of a main steam line rupture outside the primary containment. 4.5.2 Safety Design Basis
1.
The main steam line flow restrictor shall be designed to limit the loss of coolant from the reactor vessel following a steam line rupture outside the primary containment to the extent that the reactor vessel water level does not fall below the top of the core within the time required to close the main steam isolation valves.
2.
The main steam line flow restrictor shall be designed to withstand the maximum pressure difference expected across the restrictor following complete severance of a main steam line. 4.5.3 Description A main steam line flow restrictor is provided for each of the four main steam lines. The restrictor is a complete assembly welded into a vertical section of the main steam line between reactor vessel and the first main steam isolation valve, and downstream of the main steam relief valves. The restrictor limits the coolant flow rate from the reactor vessel in the event of a main steam line break outside the primary containment to the maximum (choke) flow. The restrictor assembly consists of a venturi-type nozzle insert welded into a carbon steel pipe. The venturi-type nozzle insert is constructed utilizing all austenitic stainless steel and is held in place with a full circumferential fillet weld. The restrictor assembly is self draining (low point pockets are internally drained to steam line). The flow restrictor is designed and fabricated in accordance with USAS B31.1.0, 1967 edition and the applicable GE design and procurement specifications, which were implemented in lieu of the outdated B31 Nuclear Code Cases-N2, N7, N9, and N10. Preinstallation inspection and testing are in accordance with the ASME Boiler and Pressure Vessel Code, Sections I, III, and IX, 1965 edition. The container pipe is also designed and fabricated in accordance with USAS B31.1.0, 1967 edition, the applicable GE design and procurement specifications, and with the ASME Boiler and Pressure Vessel Code, Sections I, III, and IX, 1965 edition. The flow restrictor has no moving parts, and the mechanical structure of the restrictor is capable of withstanding the velocities and forces under main steam line break conditions where maximum differential pressure is approximately 1375 psi.
BFN-21 4.5-2 The ratio of the venturi throat diameter to a steam line diameter is approximately 0.6. This results in less than a 9 psi pressure difference at rated flow. This design limits the steam flow in a severed line to about 200 percent of its rated flow, yet it results in negligible increase in steam moisture content during normal operation. The restrictor is also used in the measurement of steam flow to provide indication in the control room, to provide input to the feedwater level control system, and to initiate closure of the main steam isolation valves when steam flow exceeds preselected operational limits. 4.5.4 Safety Evaluation In the event of a main steam line break outside the primary containment, steam flow rate is restricted in the venturi throat by a two-phase mechanism similar to the critical flow phenomenon in gas dynamics. This limits the steam quantity flow rate, thereby reducing the reactor vessel coolant blowdown and the fuel clad temperature increase subsequent to the blowdown. The probability of fuel failure and its consequences are therefore decreased. Analysis of the steam line rupture accident (see Section 14.0, "Plant Safety Analysis") shows that the core remains covered and that the amount of radioactive materials released to the environs through the main steam line break does not exceed the values of 10 CFR 50.67. Pressure surges caused by a two-phase mixture impinging on the flow restrictor result in stresses which do not exceed code-allowable limits. There is adequate margin in the code for withstanding the pressure load due to impact pressure from the possible oncoming two-phase mixture predicted during main steam line break accident conditions. Tests were conducted on a scale model to determine final design and performance characteristics of the flow restrictor, including maximum flow rate of the restrictor corresponding to the accident conditions, irreversible losses under normal plant operating conditions, and discharge moisture level. The tests showed that the flow restrictor operation at critical throat velocities is stable and predictable. Unrecovered differential pressure across scale model restrictor is consistently about 10 percent of the total nozzle pressure differentials, and the restrictor performance is in agreement with existing ASME correlation. Full size restrictors have slightly different hydraulic shape and a differential pressure loss of approximately 15 percent.
BFN-21 4.5-3 4.5.5 Inspection and Testing Because the flow restrictor forms a permanent part of the main steam line piping and has no moving components, no testing program is planned. Only very slow erosion will occur with time, and such a slight enlargement will not have safety significance.
BFN-21 4.5-1 4.5 MAIN STEAM LINE FLOW RESTRICTOR 4.5.1 Safety Objective To protect the fuel barrier, the main steam line flow restrictors limit the loss of water from the reactor vessel before main steam isolation valve closure in case of a main steam line rupture outside the primary containment. 4.5.2 Safety Design Basis
1.
The main steam line flow restrictor shall be designed to limit the loss of coolant from the reactor vessel following a steam line rupture outside the primary containment to the extent that the reactor vessel water level does not fall below the top of the core within the time required to close the main steam isolation valves.
2.
The main steam line flow restrictor shall be designed to withstand the maximum pressure difference expected across the restrictor following complete severance of a main steam line. 4.5.3 Description A main steam line flow restrictor is provided for each of the four main steam lines. The restrictor is a complete assembly welded into a vertical section of the main steam line between reactor vessel and the first main steam isolation valve, and downstream of the main steam relief valves. The restrictor limits the coolant flow rate from the reactor vessel in the event of a main steam line break outside the primary containment to the maximum (choke) flow. The restrictor assembly consists of a venturi-type nozzle insert welded into a carbon steel pipe. The venturi-type nozzle insert is constructed utilizing all austenitic stainless steel and is held in place with a full circumferential fillet weld. The restrictor assembly is self draining (low point pockets are internally drained to steam line). The flow restrictor is designed and fabricated in accordance with USAS B31.1.0, 1967 edition and the applicable GE design and procurement specifications, which were implemented in lieu of the outdated B31 Nuclear Code Cases-N2, N7, N9, and N10. Preinstallation inspection and testing are in accordance with the ASME Boiler and Pressure Vessel Code, Sections I, III, and IX, 1965 edition. The container pipe is also designed and fabricated in accordance with USAS B31.1.0, 1967 edition, the applicable GE design and procurement specifications, and with the ASME Boiler and Pressure Vessel Code, Sections I, III, and IX, 1965 edition. The flow restrictor has no moving parts, and the mechanical structure of the restrictor is capable of withstanding the velocities and forces under main steam line break conditions where maximum differential pressure is approximately 1375 psi.
BFN-28 4.5-2 The ratio of the venturi throat diameter to a steam line diameter is approximately 0.6. This results in less than approximately 12 psi pressure difference at rated flow. This design limits the steam flow in a severed line to about 200 percent of its rated flow, yet it results in negligible increase in steam moisture content during normal operation. The restrictor is also used in the measurement of steam flow to provide indication in the control room, to provide input to the feedwater level control system, and to initiate closure of the main steam isolation valves when steam flow exceeds preselected operational limits. 4.5.4 Safety Evaluation In the event of a main steam line break outside the primary containment, steam flow rate is restricted in the venturi throat by a two-phase mechanism similar to the critical flow phenomenon in gas dynamics. This limits the steam quantity flow rate, thereby reducing the reactor vessel coolant blowdown and the fuel clad temperature increase subsequent to the blowdown. The probability of fuel failure and its consequences are therefore decreased. Analysis of the steam line rupture accident (see Section 14.0, "Plant Safety Analysis") shows that the core remains covered and that the amount of radioactive materials released to the environs through the main steam line break does not exceed the values of 10 CFR 50.67. Pressure surges caused by a two-phase mixture impinging on the flow restrictor result in stresses which do not exceed code-allowable limits. There is adequate margin in the code for withstanding the pressure load due to impact pressure from the possible oncoming two-phase mixture predicted during main steam line break accident conditions. Tests were conducted on a scale model to determine final design and performance characteristics of the flow restrictor, including maximum flow rate of the restrictor corresponding to the accident conditions, irreversible losses under normal plant operating conditions, and discharge moisture level. The tests showed that the flow restrictor operation at critical throat velocities is stable and predictable. Unrecovered differential pressure across scale model restrictor is consistently about 10 percent of the total nozzle pressure differentials, and the restrictor performance is in agreement with existing ASME correlation. Full size restrictors have slightly different hydraulic shape and a differential pressure loss of approximately 15 percent.
BFN-21 4.5-3 4.5.5 Inspection and Testing Because the flow restrictor forms a permanent part of the main steam line piping and has no moving components, no testing program is planned. Only very slow erosion will occur with time, and such a slight enlargement will not have safety significance.
BFN-27 4.6-1 4.6 MAIN STEAM ISOLATION VALVES 4.6.1 Safety Objectives Two main steam isolation valves (MSIVs), one on each side of the primary containment barrier, in each of the main steam lines close automatically to:
a.
Prevent damage to the fuel barrier by limiting the loss of reactor coolant water in case of a major leak from the steam piping outside the primary containment, and
b.
Limit release of radioactive materials by closing the primary containment barrier in case of a major leak from the nuclear system inside the primary containment. 4.6.2 Safety Design Basis The main steam isolation valves, individually or collectively, shall:
a.
Close the steam lines within the time established by design basis accidents to limit the release of reactor coolant or radioactive materials,
b.
Close the steam lines at a speed slow enough so that simultaneous (inadvertent) closure of all steam lines will not induce a more severe transient on the nuclear system than closure of the turbine stop valves while the bypass valves remain closed,
c.
Close the steam lines when required despite single failure in either valve or the attached controls, to provide a high level of reliability for the safety function,
d.
Use separate energy sources, as the motive force, to independently close the redundant main steam isolation valves in an individual steam line,
e.
Use local stored energy (compressed air and springs) to close at least one main steam isolation valve in each steam line without relying on continuity of any variety of electrical power for the motive force to achieve closure,
f.
Be able to close the steam lines during or after seismic loadings to ensure isolation if the nuclear system is breached by an earthquake, and
g.
Be testable during normal operating conditions, to demonstrate that the valves will function.
BFN-28 4.6-2 4.6.3 Description Two main steam isolation valves (MSIVs) are welded in a horizontal run of each of the four main steam lines, with one valve inside the primary containment barrier and the other as close as practical to the outside of the primary containment barrier (see Figures 4.5-1, 4.5-2, and 4.5-3 of Subsection 4.5). The valves, when closed, form part of the nuclear system process barrier for openings outside the primary containment, and part of the primary containment barrier for nuclear system breaks inside the containment. The description and testing of the controls for the main steam isolation valves are included in Subsection 7.3, "Primary Containment and Reactor Vessel Isolation Control Systems." A drawing of a main steam isolation valve is shown in Figure 4.6-1. Each valve is a "Y"-pattern, 26-inch globe valve connected to matching 26-inch, schedule 80 (nominal I.D. 23.647 in.) pipe. A nominal rate of steam flow for Extended Power Uprate (3952 MWt) is 4.1 x 106 lb/hr at 1050 psia RPV dome pressure. The main disc or poppet is attached to the lower end of the stem and moves in guides at a 45-degree angle from the inlet pipe. Normal steam flow tends to close the valve and higher inlet pressure tends to hold the valve closed. The bottom end of the valve stem closes a small pressure-balancing hole in the poppet; when open, it acts as a pilot valve to relieve differential pressure forces on the poppet. The valve stroke for a 26-inch valve has approximately a 14-inch stem travel; the main poppet travels approximately 13 inches with approximately the last inch of valve travel closing the pilot hole. A helical spring between the stem and the poppet keeps the pilot hole open when the poppet is off its seat, but failure of the spring will not prevent closure of the valve. The air cylinder can open the poppet with a maximum of 200 psi differential pressure across the isolation valve in a direction tending to hold the valve closed. The diameter of the poppet seat is approximately the same size as the inside diameter of the pipe, and the 45-degree angle permits stream lining of the inlet and outlet passage to minimize pressure drop during normal steam flow and to avoid blockage by debris. The valve stem penetrates the valve bonnet through a stuffing box having a set of replaceable packing. The poppet backseats on the bonnet cover in the fully open position, and leakage is prevented by the stem packing. The bonnet has provisions for seal welding in case leaks develop after the valve has extensive service.
BFN-27 4.6-3 The upper end of the stem is attached to a combination air cylinder and hydraulic dashpot that are used for opening and closing the valve and for speed control, respectively. Speed is adjusted by a valve in the hydraulic return line alongside the dashpot; the valve closing time is adjustable to meet the required Technical Specification limits (a minimum of three seconds and a maximum of five seconds). The cylinder is supported on large shafts screwed and pinned into the valve bonnet. The shafts are also used as guides for the helical springs used to assist the valve to close. The springs exert downward force on the spring seat member which is attached to the stem. Spring guides prevent scoring in normal operation and prevent binding if a spring breaks. The spring seat member is also closely guided on the support shafts and rigidly attached to the stem to control any eccentric force in case of a broken spring. On each MSIV, switches located at approximately 90 percent open, 85 percent open, and up to 7 percent open positions are actuated by the motion of the spring seat member. On each MSIV, the 90 percent open switches initiate reactor scram if several MSIVs close simultaneously (see Subsection 7.2, Reactor Protection System), the 85 percent open switch turns on the closed lights valve position, and the switch set at up to 7 percent open position indicates the valve is closed. The MSIV is operated by pneumatic pressure and action of compressed springs. The control unit is attached to the air cylinder, and contains the pneumatic, AC and DC control valves for opening, closing, and slow-speed exercising of the main valve. The control power available is 120-V AC at 60 cycles and 250-V DC. Both the AC and DC control valve solenoids use approximately 0.5 amps of control power for each solenoid. Remote manual switches in the control room enable the operator to operate or close each valve either at fast speed for primary containment isolation or at the slow speed (approximately 45 to 60 seconds) for exercising or testing. MSIV operating air is supplied at approximately 81 psig to 105 psig for the outboard valves and approximately 90 to 105 psig for the inboard valves from the various plant air systems through a check valve. An air accumulator between the control valve and the check valve provides a source of backup operating air. This accumulator is designed to provide for one closing actuation following loss of air supply. Once closed, the valve is held closed by the springs. The valve is designed for saturated steam flow at 1250 psig and 575F, with a moisture content of approximately 0.23 percent. In the event that the main steam line should rupture downstream from the valve, the steam flow quickly increases to no more than 200 percent of rated, flow being limited from further increase by the venturi flow restrictor upstream of the valves.
BFN-27 4.6-4 During valve closure, the MSIV initially has little effect in reducing flow because the flow is choked by the venturi restrictor upstream from the valves. After the main valve poppet enters the flow stream, flow is reduced as a function of the MSIV cross-sectional flow area versus travel characteristic. The design objective for the valve is a minimum of 40 years of service at the specified operating conditions. The estimated operating cycles per year is 100 cycles during the first year and 50 cycles per year thereafter. In addition to minimum wall thickness required by applicable codes, a corrosion allowance of 0.120-inch minimum is added to provide for 40 years of service. For the 60 year operating life, the Technical Specification Surveillance Requirements will assure the MSIVs are capable of performing their design functions and the MSIV aging effects will be managed using the ASME Section XI Subsections IWB, IWC, and IWD Inservice Inspection Program, Chemistry Control Program, BWR Stress Corrosion Cracking Program and One-Time Inspection Program described in Appendix O, Sections O.1.4, O.1.5, O.1.10, and O.1.26. Design specification normal and maximum ambient operating conditions for the MSIVs are tabulated in drawing 47E225-110 for each unit. See FSAR, Appendix M, Subsection M.8. However, the inside valves are not exposed to maximum conditions continuously, particularly during reactor shutdown, and the valves outside the primary containment and shielding are in much less severe ambient conditions. The main steam isolation valve installations are designed as seismic Class I equipment to resist sufficiently the response motion at the installed location within the reactor building from the Design Basis Earthquake (see Appendix C). The valve assembly is manufactured to withstand the design basis seismic forces. The stresses caused by horizontal and vertical seismic forces are considered to act simultaneously and are added directly. The seismic coefficients are specified as 0.73g horizontal and 0.07g vertical. The stresses in the actuator supports caused by seismic loads are combined with the stresses caused by other live and dead loads including the operating loads. The allowable stress for this combination of loads is based on the ordinary allowable stress as set forth in the applicable codes. The parts of the main steam isolation valves which constitute a process fluid boundary are designed, fabricated, inspected, and tested as required by USAS B31.1.0, 1967 edition and the applicable GE design and procurement specifications, which were implemented in lieu of the outdated B31 Nuclear Code Cases-N2, N7, N9, and N10. The control valves and other equipment provided in the valve assembly were designed, manufactured, and shop-tested in accordance with the then-current revision of the following codes and standards, where applicable: USA Standards Institute B31.1 and B16.5, American Society for Testing and Materials (ASTM), American Society of Mechanical Engineers (ASME),
BFN-28 4.6-5 ASME Boiler and Pressure Vessel Code, Sections I, III, and VIII, American Institute of Electrical Engineers, Pipe Fabrication Institute, and National Electrical Manufacturers Association. 4.6.4 Safety Evaluation In a direct cycle nuclear power plant, the reactor steam goes to the turbine and other equipment outside the reactor containments. The analysis of a complete sudden steam line break outside the primary containment is described in Section 14.0, "Plant Safety Analysis." It shows that the fuel barrier is protected against loss of cooling if main steam isolation valve closure takes as long as 5.5 seconds (includes up to 0.5 seconds for the instrumentation to initiate valve closure after the break and the maximum allowable valve stroke time). For the LOCA inside of containment, the inboard main steam isolation valve closure can take as long as 2 minutes, which is before any radiation releases occur as described in Section 14.6. The calculated radiological effects of the radioactive material assumed released with the steam are shown to be well within the guideline values for such an accident. Thus, safety design basis "a" is shown to be satisfied with considerable margin. The shortest closing time (approximately 3 seconds) of the main steam isolation valves is also shown to be satisfactory by Chapter 14.0, "Plant Safety Analysis." The switches on the valves initiate reactor scram when several valves are 90 percent open. The pressure rise in the system, from stored and decay heat, may cause the nuclear system main steam relief valves to open briefly, but the rise in fuel cladding temperature will be insignificant. The transient is less than that from sudden closure of the turbine stop valves (in approximately 0.1 second), coincident with postulated failure of the turbine bypass valves to open. No fuel damage results. Thus, safety design basis "b" is shown to be satisfied with considerable margin. The ability of this 45, Y-design globe valve to close in a few seconds after a steam line break, under conditions of high pressure differentials and fluid flows, with fluid mixtures ranging from mostly steam to mostly water, has been demonstrated in a series of tests in dynamic test facilities. Dynamic tests with a 1-inch valve show that the analytical method is valid. A large size, 20-inch valve was tested in a range of steam/water blowdown conditions simulating postulated accident conditions. The following specified hydrostatic, leakage, and stroking tests, as a minimum, were performed by the valve manufacturer in shop tests. E. Van Zylstra, W. Sutherland, and D. Rockwell, "Design and Performance of GE BWR Main Steam Isolation Valves," General Electric Co., Atomic Power Equipment Department, March 1969 (APED-5750).
BFN-27 4.6-6
a.
Each valve was tested at rated pressure (1,000 psig) and no flow to verify capability to close between 3 and 10 seconds. The valve was stroked several times and the closing time recorded. The valve is closed by the air cylinder and springs, and may also be closed by the springs only. The closing time is usually slightly greater when closed by springs only.
b.
At least the first valve of each size was tested to demonstrate that the valve will close at rated pressure and no flow in the specified time after the valve had been held open (energized) for 1 week.
c.
Leakage with the valve seated and backseated was measured. Seat leakage was measured by pressurizing the upstream side of the valve to 1250 psig. The specified maximum seat leakage, using cold water at design pressure, was 2cc per hour per inch of seat diameter. In addition, an air seat leakage test was conducted using 50 psi pressure upstream. Maximum permissible leak was 1/10 SCFH per inch of seat diameter. No visible leakage from the stem packing at design pressure was allowed. The valve stem was operated a minimum of three times from the closed to open position, and the packing leakage was verified to still be zero by visual examination.
d.
Each valve was hydrostatically tested at USAS B16.5-specified test pressure (2,380 psig) with cold water.
e.
During valve fabrication, extensive nondestructive tests and examinations were made, including radiographic, liquid penetrant, or magnetic particle examinations of castings, forgings, welds, hardfacings, and bolts. The spring guides, the guiding of the spring seat member on the support shafts, and rigid attachment of the seat member ensure proper alignment of actuating components. Binding of the valve poppet in the internal guides is prevented by making the poppet in the form of a cylinder longer than its diameter, and by applying the stem force near the bottom of the poppet. Clearance is provided between the poppet and its guides so that some cocking of the poppet or warpage of the seat can be tolerated and still achieve a seal. After the MSIVs were installed in the nuclear system, each valve was tested several times in accordance with the extensive "Preoperational Test Procedures," and "Startup Test Procedures." The startup tests were performed at several reactor operating conditions. During the initial plant startup tests, the MSIV leak tightness was determined. When nuclear system pressure had reached approximately 800 psig, the leak tightness was checked by closing the MSIVs, evacuating the steam lines downstream and the turbine steam chest to the condenser, closing the steam chest valves, and recording the steam chest pressure. No pressure rise meant the valves were tight. If leakage
BFN-28 4.6-7 is indicated, each valve may be checked individually by opening the other valve in the same steam line with all other MSIVs closed, evacuating and closing the steam chest, and checking for pressure rise. Redundancy is provided by two MSIVs in each steam line so that either can perform the isolation function, and either can be tested for leakage after closing the other. The inside valve, the outside valve and their control systems are physically separated. Considering the redundancy, the mechanical strength, the closing forces, and the leakage tests discussed above, the main steam isolation valves satisfy safety design bases "c", "d", and "e" to limit the release of reactor coolant or radioactive materials, within the margins evaluated in Section 14.0, "Plant Safety Analysis." The MSIVs and their installation are designed as seismic Class I equipment for inclusion of seismic loadings, as delineated in Appendix C. The design of the MSIVs for seismic loadings is discussed in paragraph 4.6.3 above. These loads are small compared with the pressure and operating loads the valve components are designed to withstand. The cantilevered support of the air cylinder, hydraulic cylinder, springs, and controls is the key area. The increase in loading at the joints between the support shafts and the valve bonnet caused by the specified earthquake loading is negligible. Therefore, the seismic loading requirement of design basis "f" is met. Electrical equipment, associated with the MSIVs, that operates in an accident environment is limited to the wiring, solenoid valves, and position switches on the MSIVs. The design and purchase specifications for the wiring, solenoid valves, and position switches for accident environmental conditions are contained in the BFN 10 CFR 50.49 program. Under the accident conditions, ambient pressure and temperature increase to approximately 50 psig and 337F; each valve is required to close within a 2 minute exposure to these conditions. The valve closing is completed during this two minute time frame. Operation of the valves in the normal operating conditions and postulated accident environments is ensured by the requirements of the purchase specifications, review and approval of equipment design and vendor drawings, extensive control of materials, fabrication procedures, fabrication tests, nondestructive examinations, shop tests, preoperational and startup tests of the installed valves, and prescribed periodic inspections and tests during the plant life. Safety design basis "g" is met, as described in paragraph 4.6.5.
BFN-27 4.6-8 4.6.5 Inspection and Testing The main steam isolation valves may be tested during plant operation, and tested and inspected during refueling outages. The test operations are listed below. The main steam isolation valves may be tested and exercised individually to the 85-percent-open position in the following manner. A minimum amount of load reduction may be required during testing.
a.
Press the test pushbutton until the closed light goes on (85-percent-open position). The valve moves at the slow speed.
b.
Release the test pushbutton and the valve will automatically reopen, turning off the closed light.
c.
Repeat the test on each MSIV. The main steam isolation valves may be tested and exercised individually to the fully closed position in the following manner.
a.
Reduce reactor power to approximately 75 percent of full power.
b.
Turn the MSIV control switch to the closed position, observing the time interval between switch closure and the open light going off. The closing time should be within the established Technical Specification limits.
c.
Return the MSIV control switch to the open position.
d.
Repeat the test on each MSIV.
e.
After all the MSIVs have been tested, reactor power may be returned to the normal level. During reactor shutdowns for refueling, the main steam isolation valves are tested and visually inspected as necessary. Leakage from the valve stem packing may become suspect, during reactor operation, from measurements of leakage into the primary containment or from observations or similar measurements in the secondary containment. During shutdown, while the nuclear system is pressurized, the leak rate through the packing can be observed by visual inspection.
BFN-27 4.6-9 The leak rate through the MSIV seats (pilot and poppet seats) can be measured accurately during shutdown by pressurizing between the closed valves with compressed gas. During pre-startup tests following a refueling outage or MSIV disassembly, the valves will receive the same hydrostatic or inservice leakage tests which are imposed on the primary system. This test and leakage measurement program will ensure that the valves are operating properly, and that a leakage trend is detected.
I HELICAL SPRINGS~ ** I,, I CLEARANCE PILOT SPRING AIR CYLINDER HYDRAULIC DASH POT "' "' SPRING GUlOE -~- "'SPEED CONTROL VALVE ACTUATOR SUPPORT AND SPRING GUIDE SHAFT LEAK OFF CONNECTION CAPPED On' BONNET BOLTS BODY POPPET (PLUG, MAIN DISK) '-\\. MAIN VALVE SEAT AMENDMENT 16 BROWNS FERRY NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT Main Steaa l**laciou Valve Figun 4.6-1
BFN-28 4.7-1 4.7 REACTOR CORE ISOLATION COOLING SYSTEM 4.7.1 Power Generation Objective The Reactor Core Isolation Cooling System (RCICS) provides makeup water to the reactor vessel during shutdown and isolation from the main heat sink to supplement or replace the normal makeup sources and operates automatically in time to obviate any requirement for the Core Standby Cooling Systems (see Chapter 6, Emergency Core Cooling Systems). 4.7.2 [Deleted] 4.7.3 Power Generation Design Basis
1.
The system shall operate automatically in time to maintain sufficient coolant in the reactor vessel so that the Core Standby Cooling Systems are not required.
2.
Provision shall be made for remote-manual operation of the system by an operator.
3.
The power supply for the system shall be provided by immediately-available energy sources of high reliability in order to provide a high degree of assurance that the system shall operate when necessary.
4.
Provision shall be made so that periodic testing can be performed during plant operation, in order to provide a high degree of assurance that the system shall operate when necessary. 4.7.4 Safety Design Basis Piping and equipment, including support structures, shall be designed to withstand the effects of an earthquake without a failure which could lead to a release of radioactivity in excess of the guideline values given in 10 CFR 50.67. 4.7.5 Description The RCICS consists of a steam-driven, turbine-pump unit and associated valves and piping capable of delivering makeup water to the reactor vessel. A summary of the design requirements of the turbine-pump unit is shown on Table 4.7-1. The transient analyses are based on a RCIC flow rate of 600 GPM. A system diagram is shown in Figures 4.7-1a, 4.7-1c, and 4.7-1e.
BFN-28 4.7-2 The steam supply to the turbine comes from the main steam line from the reactor vessel. The steam exhaust from the turbine dumps to the pressure suppression pool. The pump takes suction from the condensate header, or from the pressure suppression pool header, via a core spray pump supply header. The pump discharges either to the feedwater line or to a full-flow return test line to the condensate storage tanks. A minimum-flow bypass line to the pressure suppression pool is provided for pump protection. The makeup water is delivered into the reactor vessel through a connection to the feedwater line and is distributed within the reactor vessel through the feedwater sparger. The connection to the feedwater line is provided with a thermal sleeve. Cooling water for the RCICS turbine lube-oil cooler and gland-seal condenser is supplied from the discharge of the pump (see Figures 4.7-1b, 4.7-1d, and 4.7-1f). Whenever RCIC is lined up to take suction from the condensate storage tank, the discharge piping of the RCIC is periodically vented from the high point of the system and water flow observed in accordance with Technical Specifications surveillance frequency requirements for system operability. Following any reactor shutdown, steam generation continues due to heat produced by the radioactive decay of fission products. Initially the rate of steam generation can be as much as approximately 6 percent of rated flow, and is augmented during the first few seconds by delayed neutrons and some of the residual energy stored in the fuel. The steam normally flows to the main condenser through the turbine bypass or, if the condenser is isolated, through the main steam relief valves to the pressure suppression pool. The fluid removed from the reactor vessel can be entirely made up by the feedwater pumps if the main steam line isolation valves are open or partially made up from the Control Rod Drive System which is supplied by the control rod drive feed pumps. If makeup water is required to supplement these sources of water, the RCICS turbine-pump unit either starts automatically upon receipt of a Reactor Vessel Water Level - Low Low, Level 2 signal or is started by the operator from the control room by remote-manual controls. The RCICS delivers its design flow within 30 seconds after actuation. To limit the amount of fluid leaving the reactor vessel, a Reactor Vessel Water Level - Low Low Low, Level 1 signal also actuates the closure of the main steam isolation valves. For events other than pipe breaks, RCICS has a makeup capacity sufficient to prevent the reactor vessel water level from decreasing to the level where the core is uncovered without the use of Core Standby Cooling Systems (see Section 14.0, "Plant Safety Analysis"). The pump suction is normally lined up to the condensate storage tanks through the condensate supply header. Other systems which use the same tanks for condensate, and could jeopardize the availability of this reserve quantity, are restricted by a standpipe to the use of water in the upper portion of the tanks. About 135,000 gallons are below the standpipe in each condensate tank. This quantity represents the conservatively calculated amount of water required to maintain reactor vessel level for at least 5.5 hours in hot shutdown conditions (MODE 3).
BFN-28 4.7-3 The backup supply of cooling water for the RCICS is the pressure suppression pool Ring Header. The turbine-pump assembly is located below the level of the condensate storage tank and below the minimum water level in the pressure suppression pool to ensure positive suction head to the pump. Pump NPSH requirements are met by providing adequate suction head and adequate suction line size. All components normally required for initiating operation of the RCICS are completely independent of auxiliary AC power, plant service air, and external cooling water systems, requiring only DC power from a unit battery to operate the valves, vacuum pump, and condensate pumps. The power source for the turbine-pump unit is the steam generated in the reactor pressure vessel by the decay heat in the core. The steam is piped directly to the turbine, and the turbine exhaust is piped to the pressure suppression pool. If for any reason the reactor vessel is isolated from the main condenser, pressure in the reactor vessel increases but is limited by automatic or remote-manual actuation of the main steam relief valves. Main steam relief valve discharge is piped to the pressure suppression pool. Throughout the period of RCICS operation, the exhaust from the RCICS turbine and main steam relief valve discharge being condensed in the pressure suppression pool results in a temperature rise in the pool. During this period RHR heat exchangers are used to control pool water temperature, if normal AC power is available for operation of the RHR system. The results of a conservative isolation scenario at an initial power level of 102% of 3952 MWt, where
1) it is assumed that only a single RHR pump and RHR heat exchanger are available for pool cooling, 2) pool cooling is delayed for 10-minutes following the isolation and 3) the RCICS suction is from the condensate storage tank, show the maximum pool temperature of (approximately) 184ºF would be reached at about 4 hours into the event.
The RCICS turbine-pump unit is located in a shielded area to ensure that personnel access areas are not restricted during RCICS operation. An analysis of the possibility of the failure of the RCIC turbine has been performed. Stresses in the turbines are sufficiently low, such that wheel failure is not predicted, even at the theoretical run-away condition of twice rated speed. Even though similar results were obtained for the analysis of the HPCI turbine, the HPCI and RCIC turbines are located in separate concrete rooms within the Reactor Building. An assumed failure of either turbine could not cause sufficient damage to prevent safe shutdown of the plant. The turbine controls provide for automatic trip of the RCICS turbine upon receiving any of the following signals:
a.
Turbine overspeed--to prevent damage to the turbine and turbine casing,
b.
Pump low-suction pressure--to prevent damage to the turbine-pump unit due to loss of cooling water,
BFN-27 4.7-4
c.
Turbine high-exhaust pressure--indicating turbine or turbine control malfunction, and
d.
Automatic isolation signal-indicating RCIC steam line rupture. Since the steam supply line to the RCICS turbine is a primary containment boundary, certain signals automatically isolate this line causing shutdown of the RCICS turbine. Automatic shutdown of the steam supply is described in Subsection 7.3, "Primary Containment and Reactor Vessel Isolation Control System. The turbine control system is positioned by the demand signal from a flow controller, and satisfies a twofold purpose:
a.
Limit the turbine pump speed to its maximum normal operating value, and
b.
Position the turbine governor valve(s) as required to maintain constant pump discharge flow over the pressure range of system operation. The RCICS piping within the drywell up to and including the outer isolation valve is designed in accordance with the USA Standard Code for Pressure Piping, USAS B31.1.0, 1967 edition, and the applicable GE design and procurement specifications, which were implemented in lieu of the out dated B31 Nuclear Code Cases-N2, N7, N9, and N10, plus ASME Boiler and Pressure Vessel Code, Section I, 1965 edition. Other piping is designed in accordance with the USAS B31.1.0, 1967 edition, as applicable. The thermal sleeve (liner) in the feedwater line is designed as a nonpressure-containing liner and is provided to protect the pressure-containing piping tee from excessive thermal stress. 4.7.6 Safety Evaluation The safety design basis is satisfied by design of the RCICS containment function to seismic Class I specifications (see Appendix C). 4.7.7 Inspection and Testing A design flow functional test of the RCICS is performed during plant operation by taking suction from the condensate header and discharging through the full flow test return line back to the condensate storage tank. The discharge valve to the feed line remains closed during the test and reactor operation is undisturbed. Testing of the pump discharge valve is accomplished in accordance with Subsection 4.12, Inservice Inspection and Testing. Control system design provides automatic return from test to operating mode if system initiation is required during testing. Periodic inspection and maintenance of the turbine-pump unit are based on manufacturer's recommendations and sound maintenance practices. Valve position indication, as well as instrumentation alarms, is displayed in the control room.
BFN-28 Table 4.7-1 REACTOR CORE ISOLATION COOLING SYSTEM TURBINE - PUMP DESIGN DATA PUMP Number required - 1 Design Temperature - 40F to 140F Capacity - 100% Design Pressure - 1500 psig NPSH - 20 ft (minimum) Developed Head - 2930 ft @ 1189 psia reactor pressure 525 ft @ 165 psia reactor pressure Flow Rate Injection Flow 600 gpm Cooling Water Flow 16 gpm Total Pump Discharge 616 gpm TURBINE Number required - 1 Capacity - 100% Steam Inlet Pressure range (psig) 150 to 1174 (saturated) Steam Exhaust Pressure (psia) 25 (Unit 1) 32 (Unit 2), 29 (Unit 3)
090~ ~-t ~83Lt-Z ~ Ls I I 1 REACTOR a* ..--< 2-47E818-1,C7 r--,:--'PS I ~ 71-=!A ~PDT'-'-~~--" PS PRIMARY CONTAINMENT\\ l.. FROM CRD I 71.=..!A 71.:J.B CONDENSING NOTE 14 (TYP /; ~ ~ CHAMBER SEE NOTE 16 ---,, 3/4" RTV-lAL ECKV-71-1AA (TYP 2) --;r:=CP::jO"-Ti:-7_:1_-1_:A:_--11::F-:.!.1 ;;," ~~,.:X::;-~51~F--~~L~---t-71~~A 71~~0 MAIN STEAt.4 11 ~---1*:* 1 " X 51 E RTV~1AH -LECKV-71-1AB 2 47E801 1, FS 1h 1 " X-33F L FT 71,:18 ~ 71.:.!B FLG-71-1 I' ~ SHV-71-62 : 5 12 FCVtJ
--l(
71-19 r---t, CPOT-71-18 RTV-lBL L_ECKV-71-188 ",1, 31"-X:,3e,3.s.E __ -t:,Hx:)---- / ~ I' 1" RTV-1BH en ~fE \\~ @(TYP 2) FCV ~ 3* 7..1_;.!/ ~-3 I I ~ HPCI SYSTEM ,,. ~ ~ ECKV-71-1BA I 0-47E820-1.C4 SYSTEM PRESS. - TEii' DATA LINE DESIGN DESIGN MK NO. PRESSURE (PSIG) TEMP( 9'") 1 1146 562 2 150 325 3 137 5 545 4 1500 200 5 150 200 6 1375 200 7 1500 40 TO 140 8 1250 575 9 165 366 10 27D 415 11 400 450 12 25D 150 13 1 DO 103 14 2DD 200 ~ w ~ ' ~ ~ ~ LOV-565 114,, T ~ AUX BLR SYS 1/4" 3" 2-47E815-4,G6>----------I~ f-----f',J-'-~ lsHV-S CKV-564 l---m65 PRE-STARTUP _/ STEAM SUPPLY X 10 r---"'"'<} 2-47E812-1 G6 u ( r'E J\\--,---i> -1c-:H,~\\'l-.;3_:_,---c:,i.*:3-----l<l-----------4;.'.,' __ _;,1 ______________________________ ;,_-_-_-_-_-_-_-_-_-_.. _________ _;M;:A;;I~N:_.;:S.:,TE;a:A:;:M::.,;S;:U:;_P;,.P:;_L Y;._--------------------------------------------*-*-.. 71-18 PT 71-4 ~ w w r N z -
i
~. - STEAM DRAIN --""' TEST N "' ~ TEST (Fsv\\ 7~8 1/2" IO ~4".:.:-1-.1--1 ~ DRV-556 SHV-556 U ~ ~ _,........--PUMP TEST RETURN TO ,Ill""' CONDENSATE STORAGE N S 1/2" 1 " \\/ENT 1 " VT\\/-530 3/4" LOW POINT DRAIN 1 /2. 1 " HPCI DRAIN _,....l..._ VTV-529 ~--<2-47[812-1, G7 I f.Fg_\\ '-----'---' 71..:;.~1/2" TELL-TALE 0~ rW - v, C V,Z ~ v,w w <C ~ ~z.. ~o, IDU V ~ TO CRW [.___[2-47E852-2,E5 CKV-40 ~ 6" µ- m C.. I ~, '7 1---11-- 1 /2. TEST ~ N ~ ' r 41 NOTE 15 ~ N ~ ' r SHV-553 DRV-554 -;-;;-1,5375" PR0-38 fFc0i 7~ -37 6" LOV-37 -~ ~ m N ~ ~ N N ~ i C L~V-18 112* . ~ 6" RELIEF VALVE SET o 150 PSIG RFV-19 '° 71-208~=--, 1/2" PT LL--Do:J--f 71 -20 RTV-20 --~: 2-47[852-2,85 ~ (Ps\\ J/2" 113A)r-lRDTv"'-:12:..1--11-_1_*., re?\\_ 71-:!18 r3/4" VENT m, RCIC PUMP (616 GPM) w V SPOOL PIECE SPPC-71-19 (NOTE 13) r1 /2" PRESS. ~ 3/4" TEST ~ PMP-71-19 r '-I-Pl-----,1"""1/v TV-19 N ' 7 SHV-19D = fi FE.!---',
-~...,...... _.... TW 6" :
M[z] 71-36 = 71-33 m m -,~ @-J~_. -' r m r m m ~ ~ r ~ C ~ ~ 11*i.:-. """:.,~ w - m < z PT 71-35 ~ ~..... ~ C ~ C ~ 71-358 ~ -:-F' 0.375* PR0-547 I!; -g---i--._i_j;'c . 1-...... -1>1' CPLG-9 I TURBINE TRIP THROTTLE VALVE-~ FC~~ ~-~ 1/1" B TURBINE GOVERNING VALVE ---lf--~,1.FCV:-.,;:;,._D:l,~:J:l'x:r.:~------~~--3._"-+::i-tl\\4-*-*----1 1 " RTV-4 1 " RTV-5H PI 71-4B ~ (t.tOUNTED TO TURBINE CASING) \\ 1 l 1/2" 7\\.::1.V- ~ .--"--iM+------------1---=;__..1 u2..:. PI 1 /f" 71-55 ~. N..... DRIVE TURBINE TRB-9,.__ L.ru I
1 "
~ T HP PACKING DRAIN PI 71-56 ~ ~ - 0.185" 0.185"~~..I' PR0-9AA-PR0-98 7 ~~ 2 ABOVE AND BELOW SEAT DRAINS --.":::--....f---J 1 " SHV-570 SHV-618 STEAM SUPPLY DRAIN POT CPOT-5 1 " RTV-5L D STEAM TRAP TRP-10 PISTON REMOVED SHV-617 SHV-571 1 * ~1-------~----------~--~~--~/~----{~- r-w' TEST N TEST PUMP MINI~UM ~ FLOW BYPASS---/" ~ ' ~ DRW .___,2-47E852-2,85 1 " fFc0i CKV-3-572 (LG\\ (n\\ 7 -46 71-509 T X-98...J THIS DWG,C+ THERf.tAL SLEEVE N \\_REACTOR FEEDWATER LINE 8 N ~ ~ w ~.. N V 1 4f--+-{:;4£]-, N ~.. ~ ' = i ~ ~.. ~ r N ~ VENT TO A~M-"J ~.: ~ ~~=1~ ~ ~--: I 1 " [ VENT TO ATM - RPD-118 z - ~ C i'l AUX BLR SYS 2-47E815-4,G6 >---, 71-68 1.. Y CKV-574 IX ffi-+--EJ
x*
~ LS 1-5 ~..... 1 -1" <J-1 TW '[' 1-1i2" ~ 14---------< THIS DWG,C4 [ ~o,..,__.__~. ~ ~ al 1-fRPD-11A z < N N -' TE"\\ 71-47 PMP-19 RCIC PIM' 1 " I I RCIC C: ~.. m, co -~ 0~ I I ., /R." .~ !§~ ~.,. 3/4* 00 -~ TURBINE TRB-9 ca.. 71-507 112.. TW "'1r17-~1 l:~2::-' ~-1._'_'-"i=,-~~~,6.l.i.ii.,f8 "11, '--1 /2" TE 71-45' 3/4" OIL ~/A" I... 71-24 ~ PUMP O ~/2" IT! 1 1 PMP-45 -.-1;2* in 71-45 ~ Ill:: --.J,- \\.. 1" VENT~ s,.. ~ DRV-45 c.:: ,.-1/4" .I \\I ~ -~..... -<~ ~u Pl 71-501 - TI 7~3 1 " -3 TW 71 = 3/4"
t ry
~ CKV-598. N SUMP :l-'-- ~ ~ PS 7~4 1 /f" 1 3/4" RCIC LUBE OIL PIPING DIAGRAM (C4) N CKV-600 F CVIJc*'-'1 lc,:2;_"-,3, 71-~'f, ~ 1/2" GASKET TEST 2" ci<v-597 CKV-599 TURB EXH VAC RELIEF.....J SHV -520 2" TEST < 7,G\\ "> 71-14 ~ ~ LOV-1+ l/r r m ~ ' ~ ~ ~ N m ~ ' 1 /2". TEsT::i--=-'~ 2" 7~2 N CKV-580
0. 625" --
PRO 551T 2" 1 O" TURBINE EXHAUST 3/f" LOW POINT DRAIN SHV-551 DRV-552 < 2 47E852 2.85 C'\\jo..c. ~a,: 1/2" TEST__.. ~ 10" TEST N..... / 1* PRESSURE VALVE CONTROL r- (PCV\\i N ~ N 0 ~ 71-22 ~ ~' 1/2" LOW POINT DRAIN "' 1----1-..--..,.f---, SHV-538 1 " TV-5+o DRV-539 N ' N ~ ~ w :; N z C - r ~ 'la ~ ~ z -
i I /2" p T l-,-t,C(]---f 71-12 RTV-12
~ 71-128 ~ EXHAUST DRAIN POT-.....** ~ TNK-642 ............ ~~~~..:..~~~~-' STEAM TRAP TRP-9 1 * (Pl) PRESSURE RELIEF SET o 10 PSIG 2* RFV-27 J_ ill VALVE l-=-3l.: 4._"+:>4--.... "4--C,f1 !..!a/2 0TEST ~ m ~ 1 *. CKV-592 1 " .;.Fo.+1s* PR0-540 N C r 71-27: 8 ~ 1/2" CAP ----~1/4" X-218--....__ * / N 7 SHV-602 TV-603 N - X 221--...... X-212 N N 3/4" l<l----l~t"-'"--1<'4--I ~ /'... TV-610 SHV-609 g L_l/2" TEST ' t )ffi N 2 18" /X-21DA 14 18" FROM RESIDUAL HEAT REMOVAL SYSTEM 2-47E811-1, E7
t_
r ~ "'.. m ~ ' r TEST N SUPPRESSION POOL ...,..-- STRAINERS ~PRIMARY CONTAINMENT SUPPRESSION POOL RING HDR 7 ~ fFc0i 71-16 1... 1-:,,1 6 LOV-17 1/4" 1/2" SUPPRESSION POOL SUCTION 6* 6"" ~------,-------~,~---t.* --~ .~-------~w~----1..... ~J--------.;..... 0?i[s] ~---~---~-"----\\' CONT ON 2-47E814-1, 87 C ~ ~ ~ ~ ' .7 ~ ~ CKV-508 TV-541 N w IE ~ ITHIS DWG,E7)>----~... BAROMETRIC CONDENSER CND-27 ~-=':t'::'t'::i' (:' 'f' '-'-;:: r:::-,.PS::::--::::::-~lr'VACUUM TANK r _ _::::;_J..250" TNK-27 w (rriiHc,sscoiiwii:c::.,E,1J.I ----~ 314.. LUBE OIL COOLER 16 GPM 140"f MAX CLR-25 11-201-;=!:1 5 ~ 1 /2" RTV-27H PRESSURE RELIEF VALVE SET o 75 PSIG RfV-543 ~ ~ 2-47E852-2,85 2" 1 " w C z 8 -~T . '~ ~-N P ( TYP 4)---.. 2" LS 7...!_:39 ~ MECH LINK 1/2" f---to:J-'--ICXX::r---!PI 71-~02 RTV-502 ISIV-502 SEE NOTE 9 1 /2. 8* PIPE I) '-,tc,..-~. PRESS.
.,.., i,

*.,;;
- TEST ""-"""'""""---r....., TV-26 2" 2-47E852-2,85 > SHV-26 DRV-26 1-1/4.~ r l!) ""'- CKV-589 CKV-588C~LN-D-EN_S_A,,TE PUMP 43 GPM x PMP-29 ~ 71-3.U"' ~. BYPASS ~ CONTROL n VALVE 1 /2" TEST 3/4" ~ ~/"I...., TV-591 SHV-590 x~ ~ < 2-47E852-2,B5 f-- 2".1-1/4" LJ \\!) VACUUM PUMP 17 CfM PMP-31 2" VACUUM PUMP DISCHARGE TO SUPPRESSION VESSEL TEST 6 5 4 CALCULATION
REFERENCES:
MDQODD07120120D0031: RCIC PUMP NPSHa AND SYSTEM HYDRAULICS 3 COMPANION DRAWINGS: 1-.J-47EB13-1 N - ~ w ~.. ' N r TEST NOTES:
1. CRW DRAINS ARE CONNECTED TO REACTOR BUILDING EQUIPMENT DRAIN SUMP.

2. DRW DRAINS ARE CONNECTED TO REACTOR BUILDING FLOOR DRAIN SUMP.

3. ALL VALVES ARE THE SAME SIZE AS PIPING, UNLESS OTHERWISE NOTED.

4. ALL VALVE DRAINS AND VENTS ARE PIPED TO CRW.

5. UNIDS ON DRAWINGS ARE FOR REFERENCE ONLY AND ARE ABBREVIATED AS SHOWN IN THE EXAMPLE TO MEET SPACE CONSTRAINTS. REFER TO MEL FOR COMPLETE UNIDS. ALL UNIDS ARE IN UNIT 2, SYS 71 UNLESS OTHERWISE NOTED.
LEADING ZEROES SHOWN IN MEL AS PART Of THE UNID ARE NOT DEPICTED. FOR ADDITIONAL GUIDANCE, REFER TO NEDP-4. EXAMPLE: MEL UNID DRAWING UNID BfN-0-SHV-018-0502 SHV-502 BfN-2-PT-003-0204C PT-204C
6. [I] ETC, DENOTES DESIGN PRESSURE AND TE~PERATURE AS GIVEN IN TABLE, THIS DRAWING.

7. DELETED

8. THE DESIGN PRESSURE AND TEMPERATURE Of' ALL DRAIN AND VENT LINES THROUGH THE LAST ISOLATION VALVE SHALL BE THE SAME AS THE PROCESS LINE.

9. CONDENSATE PUMP AND VACUUM PUMP SUCTION LINES ARE INTERNAL TO TANK AND SHOWN ONLY FOR CLARITY.

10.

12.

13.

14.
1 5. 1 6. HYDROTESTING Of THERMOWELLS AND FITTINGS SHALL BE PERFORMED IN SHOP AT 30 PSIG. SYSTEM LEAK TEST PER ANSI 831.10 POWER PIPING CODE. FOR DESIGN PRESSURE AND DESIGN TEMPERATURE SEE TERRY STEAM TURBINE MANUAL. CONTROL# 0467 (CONTRACT# 90744). VENT, DRAIN AND TEST CONNECTIONS 1-1/2" AND BELOW CAN BE PROVIDED WITH PIPE CAPS OR HOSE CONNECTION FITTINGS WHERE REQUIRED BY PLANT PERSONNEL. THIS CONFIGURATION IS SUPPORTED BY ENGINEERING CALCULATION CD-00999-923399. SPOOL PIECE MAY BE REPLACED BY STRAINER DURING MAINTENANCE. THE SPOOL PIECE MUST BE INSTALLED PRIOR TO RETURNING THE SYSTEM TO SERVICE. SEE 0-47E456-1, -2 FOR STRAINER DETAIL. ORIFICE COUPLING, SEE 0-471600-20, DETAIL P20. VALVE 2-fCV-071-0039 WILL SEAL LEAK TIGHT ONLY IN THE DIRECTION Of fEEDWATER TO RCIC SYSTEM DUE TO A PRESSURE LOCKING MODIFICATION WHICH DRILLED A 1/4* HOLE IN THE fEEDWATER SIDE FACE Of THE VALVE. (DCN 141301) THE PIPING COMPONENT IDENTIFIED IS A LINE STOP FITTING WITH AN INTERNAL PLUG AND BLIND FLANGE. NORMAL CONFIGURATION DOES NOT ISOLATE FLOW AND FITTING, PLUG AND BLIND FLANGE ARE CONSIDERED PRESSURE RETAINING. REFERENCE DRAWINGS: MEL................... VALVE MARKER TAG TABULATION 2-47E456-408.......... MECHANICAL-RCIC OIL PIPING INSTRUMENTS 2-47E852-1,-2......... FLOW DIAGRAM - DRAINAGE INSTALLATION SECTIONS AND DETAILS 0-47E820-1............ FLOW DIAGRAM - CONTROL ROD DRIVE 0-47E800-1............ FLOW DIAGRAM - GENERAL PLANT SYSTEMS 0-47E800-2............ MECHANICAL - SYMBOLS & FLOW DIAGRAM DRAWING INDEX 2-47E815-4............ FLOW DIAGRAM AUXILIARY BOILER SYSTEM 2-47E801-1............ FLOW DIAGRAM - MAIN STEAM 2-47E811-1............ FLOW DIAGRAM - RHR SYSTEM 2-47E812-1............ FLOW DIAGRAM - HPCI SYSTEM 0-47E815-1............ FLOW DIAGRAM - AUX BOILER SYSTEM 2-47E818-1............ FLOW DIAGRAM - CNDS STORAGE & SUPPLY SYSTEM MEL................... INSTRUMENT TABULATION 2-47E610-71-1,2....... CONTROL DIAGRAM - RCIC SYSTEM 2-47E803-1............ FLOW DIAGRAM - REACTOR fEEDWATER 2-47E810-1............ FLOW DIAGRAM - REACTOR WATER CLEANUP 2-47E814-1............ FLOW DIAGRAM - CORE SPRAY AMENDMENT 26 POWERHOUSE UNIT 2 BROWNS FINAL FERRY SAFETY NUCLEAR ANALYSIS PLANT REPORT REACTOR COOLING CORE ISOLATION SYSTEM FLOW DIAGRAM FIGURE 4.7-1a H G F E D C B A
FROM TR-69-29 9-3 I 2-47E610-69-1,Gt>- @ r--r--1--r-- RIRIRIRI 1 I --r-,--T--r----- RIRIRIRI --,---r--1--r---- RIRIRIRI 71-2A 71-2B 71-2C 71-20 I Y2-45E626-1 I I I I 71-2E 71-2F 71-2G 71-2H 71-2J 71-2K 71-2L 71-2M TO TR-69-29 /~ -------------------------------------------, I Q-47E610-69-1.G2 ~:::-+------------:-L-----------:-.,
I

I REACTOR WATER I
A A I CLEANUP -- ~ t----j 2-45E779-12 I ( TE l ( TE l R, 71-41A I 71-418 71-41C 71-41D m < u w - r z ' M 0 ~ " w ~ 0... ~ ' ~ N THERMAL SLEEVE r-------., 1 9-31 ~--- r-0 H 2-45E714-4 I [crssJ I LOG I ~~ 1 -~,1,--a...i ~-3 1 m .11-, CL. ~ ~ X w cc 4 ~. ~ml--1-~ r-J,z: r-~ ~ M1M1C 9-3 9-3~~----t l CISS I -@-I '---' ~ '* )Q_ )&_ II (HS\\ A ~ XS --~- --!--L~ M 0 m N X ' ~ 0 '$5-31 ~ --1 !2-45E626-4 f--1 I MIMIC I 7~8 I 71-3 CISS I 7~C r3El: 71~~G CL 8 11 fG{ ~ co:N~D-EENLS_I N)G-P[::OaT :J--, d-3 W,)-18 'ff=TA 11-1 r I I FCV ~ /'9{- /'y-- ~ 71-3 >r-- LOG L""""'--.J / _JM~ 71-3 0 9-19 V/ 9-19 9-3 /FM\\.. rm 25-7A 71-18 ~~ FT ----- ')::' _ Y 71-18 r-_J 25-7' n 25-7A I PS PS 25-7Aj.!-82 9-81 25-7A 7YA 1yc FT ~ -- PDT f----+--12-45E626-2 I 71-1A 71-18 71-1A 71-1A A5-7A.,....L... T T ( PS\\ ( PS\\25-7A 71-18 71-1D 9-18 I I
-31
@=.: _J @@]--1..L@:u--ill)(D CONDENSATE TO HPCI TEST LINE AND CNDS & DEMINERALIZED WATER STORAGE SYS ~ 12-451670-21 ~J L-j2-45E670-15 I 2-47E610-73-1.E5 I I I I /HS\\ /HS\\
* ~ r-r--i 71-58 71-57 I
GD9-3 *
I
~ I MCC 71-388 >~.( ).!:,.(_ I I _),;,(' y riis\\*-3 @9-3 ~ ~ FSV FSV ..,rl ;&, I r"iT:3s'A I LOG ~ 71-58 71-57 I 1 _,......L...,. _J A A71033 s s
~_38 --{-~1x_s3~ '>fw" 25-31 I I
I I y~,, (ex"\\"_ _j ~ I I l2-45E714-6f-J L _ _{Hs'\\ ~-...l lEiillHB-47 r_J 11-3sc L ( PT\\25-58 TR I FG 1 71-35 7 ~13 I DISK 71n5 8 CRW 9-3@-A I 25-58 I I POSITION -~ I I ~ ~, Z!_::; 25-58 PI R. FIS --- I 71 358 TE I RW 71-36 I 71-33 9-3 25-58 /PI\\. PT 9-3 IPA'\\ I'" 'tf::2fl, I 25-58 7T=""2DA -,. 71 -20,__, __ _, I ~ I--..-!. PS ,..-J-....25-31 25-58 7 l.:!J B ( PX' PI 71-20 71-208 25-58 _j PS 71-21A ----r-- I ~-3 I ~A I I I I J._ (Lr\\ 71-509 L ~ ~ w ~ ~ ' 0 THIS DWG, CS 2s-sa 2s-sa I s-3 (Pi) PT _J_-@ 71-48 71-4 71-4 ~ J¥MCC I 1--.,., ~~~8 +#1 W ~,~I ~ ---+-- _ _{Hs'\\ I I 7..!.j!8~ I ~G R I XS HS in 11-8 I 1uc c~~5 ~ ~ 1t~8 S@z ZI-71 ~ .._...9_3 -10 r-3&: I TH Is owe. 01 >, I ~ FROM XS-71-368 _j G 9-3 I, ' I j FC SC,),-,w-.J 71-10 71-10 ~ - uo M u... ~z - e>m 0~ ~~ ~r THIS DWG,C5 r r;\\_ ~-3 LOG I r-m w. ~ ' ~ m N w ~ N 71-42A~ 1-~A Rcic TURBINE A5156 I I RCIC PUMP PMP-19 TRB-9 ~I~ T ~7T""~: I /~25-32 DRAIN POT CPOT-5 D LS 71-5 (#$-3 'f' I I I I J STEAM TRAP TRP-71-10 PISTON REt,AQVED I I X Lj_ ___ 1 I._!'!.::.3.JZI-71 I ~ .... ~~~.,,,,,,,,.,,,,...~~..._~--(>i<Je--~....,,,,..,,,,,,,,_,_ FE }---~~~a....~..,..... ~~~-'. TW,--~~~--,..-~~~~~~~ ~-.-.,....~ ~-.-.,....:@8, 11->6 1 1u, r-A--* ---------- ~25~t!J8 l,1\\r1--,I ! TT X 11 ~5-425 I 7-:.:;!5 """l, 7-!36 /~~ 9-b I CKV-71-4D ~ ' 0 z-ow w r~ Zv o, UN ..J 9-3 G ..J 9-3 G ,>-<_ 25-58 I ~* r.:::,.../, ( FT ' r £.l!Ll I r (.l!a..\\ I 71-36 I w FCV 71-39A I w FCV 71-37A I I _/px\\5-311
=,-----1
/~ 7-!?9 Y3*t /~ 1-!?1 Y*1 r 71~~6A I wr-FCV I MCC XS I MCC XS I I I R 71-34 I AU' :@t, AU7 ~ I I MCC Ir YR1'©1 (HS ' HS +-~ (HS ' HS I I I l~~:'..Lt 71-398 71-39C I ~ 71-378 71-37C I 1 ~ I I~ 1 ym---'YR1' ____, ~ c@____,<'iJ ____, :@8 t-~ t-___ _j_ __ _[HS\\ I (Ti\\ NOTE 11 ~ _J 4 TS I TE 71-46 71-45 TE .r---73-54-11 L-7b8 11-4 frr'\\ ITI\\ 11-41 fBR~?~f OI~uMP r _J I 71-46 71-45 THIS DWG. C5 I ~
.@S41 r ;;:58-1 L~ I-~Px 25_;-*1
~5-32
I
~-1-73~-12 : I PI' I PX' y6 I 71-128 A5152 71-12 ZI-71-6At-42-45E626-41 I SEE NOTE ,'C:_X----:i r1_,X. ___.J 1 -1.. 11.::;10 1 TO SC-71-10 r~ _l_ _/xs"'\\_ I t-T fxs\\25->'.__~- /rrc\\9 _ _: ____ _fxs\\25->1 +/- rl 7U4 11 I r----_J Y _25-58 1 '-@*-3 i y;:,( 1 1 r--+---, ~ -1.fPT).._4 ___ L PI I r;S(@-3 I (mis DWG. F3 -, 1::36~ -tr:m 71 _368 1 ~ J i '---' SEE NOTE 10 JT ! ~-3 ~ I @illf-11 25_32 11 ~A 71-34C L ____..JFIC\\_ _____________ _J /@z (R)"(G) I = ,~" '---' 1~0 -L--:'.I'~'T-_J 11 TW )----..._--r-------.,,,.~- 71-23 0i'I 71-23 TE)-,-+~ PI 71-24 71-501 25-425~ 9-47 ( TS \\..,/ TR 1y 4 73-54-10 1~2--.-,,-6-26--3~~ -t ~y 25-31...J-... LOG 'r,:{_ ___ J XS' A71024 )DC 71-24 25-32 '(------i ~ XI-71-24 ~-J - ~ TA ~ 71-2 ~ u 0 z w w "' 0 r w 0 M cc r ~5-58~3 ~5-58 71-12 1-12A I I HS '):::(._ -t ( PS\\ PA ( PS\\ 1 -- 1;:= VENTS..._.... I ~ 71-6 fi-cz~1T-68 71-138 'Q_j. 71-13A 25-58 25-581 ~---.....J PS PS I _J 7YA 7y8 I 3 L--------------- '-----1 +----+--.!. PA I 11-1 ..--L...25-58 ~5-58 en (P1' ( Ps\\ ( rs, N (/)~ (Pr\\ 71-27 71-11C 71-11D J, iE~11~io2 0 (TI) ~ LOGIC B 9-3 9-3
E~!._~
(xs'\\ - RCIC AUTO ISOLATION ~~ ~ ~ LOGIC A RESET ~ ..._,,, L ~IL-71-51 A IL-71-51 B 9-3 )CZ /xs\\ RCIC INITIATION ~TsEAL':i:N &RESET-LUBE --.-~~,,,...~~--10IL COOLER CLR-25 71T-"2s II 7J71----------N---~ '---' IPA\\ 9-37f'=2'g 1--..-1 ~ y 1-<io!-I BAROMETRIC 71-30 CONDENSER CND-27 .._...~ IL-71-52 w z M ~ "' ~ r u.. - a u ~ "' - 0 " r r PS ii 11-28;-..--;..;,L __ ~ __ J---{:::b(J--:1 ~ w-3 ~ LS FROM RCIC TURBINE ~ 71-29 OIL PUMP
';'T"'
THIS OWG, El FG 12-45E714-6 rf 71~7~1 (G)',,.-.._(Rj' ~9-3 >~-{ "f'(.~~.)'.l"' HS I VACUUM -TANK - TNK-27 I 7YCI I yAI 7!$18 T,, L_1-._j_J_.!.-L-- ~ \\ ~~3~ r* ~I ! 9-3 RCIC f'Hs\\ AUTO INITIATION "7T=sr MANUAL TSOLATION r:-:-,. 25-32 I <A \\ 71-498 T rSEE NOTE 11 'rc{_ ~*---'J-* 1-~PMP-29 I /HS\\ I :;; 9-3 MCC )!::,,(:....,,,.. ' 71-318 i r Zl-71-32, {5HV\\ I.!) ..--~*-.. ~ I N fzs'LTURBINE TRIPPED _ __r-;:{5~31 lzA\\9 - 3 71-49 71-24 ~A /, 71-32 -~ \\!) ~-~ r-~~~~~~~~-t:*:J,-~~~--1,.,"f-~-t--;-~~~~~~~~~~~~----,rL__.,, 1'-3:©:~ X -::::r-, X VACUUM,.J-.,,. PUMP (XS\\... __ PMP-31 71-31 25-32 SEE NOTE 11 Xl-71-44 TURBINE BRG OIL l... 71-44 PRTsSUREToi"'____ -- 9-3 ~~-3 I HS I 11-5 13©:1 ¢ 5-32 s 5 I I I I I I I I 1-...J I ~ 9-3 ~ I ~ 9-:(HS\\ I I
r--,
'r' I I R G ~-3 I~--::~ '-----====---' L / -::...._/ ' (A'( 25-32 (A'( 25-32 rc_:1-12:~---~*'\\ XI-71-48 I ~Al 12-45E714-5 ~l 1~1'8A I '-r'\\Ll;;{R I I I ~ FCV 71-7A I I _h ECCS-ANALOG TRIP POWER 12-47E610-71-2.F5>-~-[Iz)(j) I, I _/xs\\_ _j_ __/xs\\_ I I j'Z.1.;.17--r- -r--'Z.';,1Bj I ~ Ir~
, I
~ CRW ~~~ 71-31C 25-425 ~5-31 (rs\\,HP BRG OIL XS (m9-3 71-47 TEMPHThH-r 71-4 -r-71-47 k 11 I I -425 I I SEE NOTE I (HS\\_ J I (HS\\_ j L _fiis\\ I L _/HS\\ I ~ I I ' \\z2._:.17c I \\z.!;_178 'Z.1.;.(80 I --\\z.~.:/sc I 8§z I,.~ FC ~.J ~ L _____ J.. __ 1 11-11 11-18 I 6 "*;~: 1-- ~:& --, __ T L- ~c' MC~ _ __ J. _____ _J I 2-47E610-71-2.F4>--ITE)--ETI) ,,--.._TA 9-3 f'....8fil'..Q\\b._J L_ I TA\\ 71-48 TEMP HIGH -'.771-'\\4;.!8 COMPANION DRAWINGS: 1-47E610-71-1 3-47E610-71-1 2-47E610-71-2 8 7 6 5 4 3 NOTES:
1.
THE REACTOR CORE ISOLATION CCXJLING SYSTEM IS DESIGNED TO COOL THE CORE WHEN THE REACTOR VESSEL IS ISOLATED AND REACTOR FEEDWATER IS UNAVAILABLE. WHEN THE WATER LEVEL IN THE REACTOR VESSEL REACHES SETPOINT, THE RCIC SYSTEM IS AUTOMATICALLY STARTED BY ACTIVATION OF REACTOR LEVEL SWITCHES LIS-3-58A, LIS-3-588, LIS-3-58C AND LIS-3-58D (2-47E610-3-1). ARRANGED IN ONE-OUT-OF-TWO-TWICE LOGIC TO OPEN VALVE FCV-71-8. THE SAME SIGNAL CLOSES DRAIN LINE VALVES FCV-71-6A AND FCV-71-68.
2.
A MECHANICAL SPEED GOVERNOR PREVENTS TURBINE OVERSPEED BY TRIPPING THROTTLE TRIP VALVE FCV-71-9 WHEN A TURBINE OVERSPEED CCCURS. A CONTROL GOVERNOR, FIC-71-36, EQUIPPED WITH AUTOMATIC AND MANUAL MEANS FOR SPEED SET POINT CONTROL REGULATES TURBINE SPEED VIA "TURBINE SPEED CHANGER* TO CONTROL RCIC MAIN PUMP DISCHARGE FLOW. 3A. A RCIC STANDBY MODE SIGNAL SHALL BE INITIATED AT REACTOR VESSEL WATER LEVEL 8 (HIGH LEVEL). THE REACTOR VESSEL WATER LEVEL 8 SIGNAL SHALL BE INITIATED BY TWO LEVEL SWITCHES ARRANGED IN A TWO-OUT-OF-TWO LOGIC CONFIGURATION. THE REACTOR VESSEL WATER LEVEL 8 SIGNAL SHALL: (A) CLOSE THE STEAM ADMISSION VALVE FCV-71-8 AND (B) CLOSE THE MINI-FLOW ISOLATION VALVE FCV-71-34. IN THE EVENT AN INITIATION SIGNAL IS RECEIVED AFTER THE SYSTEM HAS BEEN SHUTDOWN ON HIGH RV LEVEL, THE SYSTEM SHALL BE CAPABLE OF AUTo.4ATIC RESTART.
38. THE RCIC TURBINE TRIP VALVE IS TRIPPED BY:
A. TURBINE OVERSPEED VIA THE SPEED GOVERNOR. C. LOW RCIC PUMP SUCTION PRESSURE SENSED BY PS-71-21. D. HIGH RCIC TURBINE EXHAUST PRESSURE SENSED BY EITHER PS-71-13A OR PS-71-138. E. MANUALLY FROM THE MAIN CONTROL ROCM BY HS-71-9A OR BY HS-71-9C ON PANEL 25-32. A TURBINE TRIP: A. CLOSES FCV-71-9.
8. CLOSES FCV-71-34.
C. ANNUNCIATES ON ZA-71-49A OR 498. D. ACTIVATES THE ALARM ASSOCIATED WITH THE CONDITION INITIATING THE TRIP.
4.
THE TURBINE CASING IS PROTECTED FROM OVERPRESSURE BY TWO RUPTURE DISCS. PS-71-11A THRU D DETECTS DISC RUPTURES AND ANNUNCIATES ON PA-71-11 IN THE MAIN CONTROL RCXJM. PROVIDES 1 OUT OF 2 TWICE LOGIC TO INITIATE AUTO ISOLATION OF TURBINE (SEE NOTE 5).
5.
THE RCIC TURBINE WILL BE AUTOMATICALLY ISOLATED FROM THE STEAM SUPPLY BY CLOSING ISOLATION VALVES FCV-71-2 AND FCV-71-3 IF:
6.
A. A HIGH PRESSURE DROP OCCURS ACROSS THE RCIC ELBOW TYPE FLOW ELEMENT FE-73-1A & 18 WHEN THE STEAM LINE RUPTURES. THE HIGH FLOW IS SENSED BY EITHER PDT-71-1A OR PDT-71-18 AND ANNUNCIATED ON POA-71-1.
8. HIGH AREA TEMPERATURES INDICATING A STEAM LINE BREAK ARE DETECTED BY ANY OF FOUR SETS OF TEMPERATURE SWITCHES (TS-71-2A THROUGH TS-71-2S). EACH SET CONNECTED IN ONE-OUT-OF-TWO-TWICE LOGIC. FOUR INDEPENDENT TEMPERATURE SENSING ELEMENTS (TE-71-41A THROUGH TE-71-41Dl ACTIVATE THE MAIN CONTROL RCXJM ANNUNCIATOR TA-71-41.
C. LOW REACTOR PRESSURE IS DETECTED BY PS-71-1A THROUGH PS-71-1D ARRANGED IN ONE-OUT-OF-TWO-TWICE LOGIC. D. HIGH PRESSURE BETWEEN THE TURBINE EXHAUST RUPTURE DISCS (SEE NOTE 4). NORMALLY MAKEUP WATER FOR THE RCIC PUMP SUCTION IS SUPPLIED BY THE CONDENSATE STORAGE TANK THROUGH FLOW CONTROL VALVE FCV-71-19. AN ALTERNATE SOURCE OF WATER IS OBTAINED FROM THE SUPPRESSION POOL THROUGH ISOLATION VALVES FCV-71-17 AND FCV-71-18. PUMP DISCHARGE FLOW ROUTES ARE AS FOLLOWS: A. NORMALLY VIA THE FEEDWATER LINE TO THE REACTOR THROUGH ISOLATION VALVES FCV-71-37, FCV-71-39, AND CHECK VALVE FCV-71-40.
8. DURING TESTING VIA TEST LINE TO CONDENSATE STORAGE THROUGH FLOW CONTROL VALVE FCV-71-38.
C. MINIMUM FLOW BYPASS TO THE SUPRESSJON POOL THROUGH FLOW CONTROL VALVE FCV-71-34.
7.
IF RCIC SYSTEM OPERATION IS REQUIRED WHEN IT IS IN THE TEST MODE, THE CONTROL SYSTEM IS DESIGNED TO AUTOMATICALLY RETURN THE SYSTEM TO THE OPERATING MODE.
8.
COOLING WATER FOR THE PUMP, TURBINE, AND LUBE OIL COOLER IS SUPPLIED FROM THE PUMP DISCHARGE LINE THROUGH PRESSURE CONTROL VALVE PCV-71-22 SET AT 75 PSIA.
9.
TRANSFER ANO CONTROL SWITCHES ASSOCIATED WITH BACKUP CONTROL WILL BE ON THE DESIGNATED PANELS OR LOCATED ON THE APPROPRIATE 250V DC t.tO\\I BOARDS.
10. XS-71-368 JS SHOWN TWICE TO SHOW HOW THIS TRANSFER SWITCH DISCONNECTS FIC-71-36A AND CONNECTS FIC-71-368 LOCATED AT THE BACKUP CONTROL CENTER.

11. XS-71-24, PS-71-44, TS-71-47 & TS-71-48 ARE SHOWN TWICE TO SHOW HOW THESE SWITCHES TRANSFER INDICATION FROM THE MAIN CONTROL PANEL 9-3 TO THE BACKUP CONTROL PANEL 25-32.

12. UNIDS ON DRAWINGS ARE FOR REFERENCE ONLY AND ARE ABBREVIATED AS SHOWN IN THE EXAMPLE TO MEET SPACE CONSTRAINTS.
REFER TO MEL FOR COMPLETE UNIDS. ALL UNIDS ARE IN UNIT 2 SYSTEM 71, UNLESS OTHERWISE NOTED. LEADING ZEROES SHOWN IN MEL AS PART OF THE UNID ARE NOT DEPICTED. FOR ADDITIONAL GUIDANCE, REFER TO NEOP-4. EXAMPLE: MEL UNJD DRAWING UNID BFN-0-SHV-018-0502 SHV-18-502 BFN-1-HS-031-0160A HS-31-160A BFN-2-PT-003-0204C PT-3-204C
13. COMPLETE LOGIC FOR REACTOR CORE ISOLATION SYSTEM IS SHOWN ON MECHANICAL LOGIC DIAGRAM LISTED IN REFERENCE DRAWINGS.
REFERENCE DRAWINGS: 0-47E800-1.................. FLOW DIAGRAM-GENERAL PLANT SYSTEMS 0-47E800-2.................. MECHANICAL SYMBOLS AND FLOW DIAG DWG INDEX 2-47E2847-5................. FLOW DIAGRAM-CONTROL AIR SUPPLY 2-47E813-1.................. FLOW DIAGRAM-RCIC 2-47E61D-73-1............... CONTROL DIAGRAM-HPCI 2-47E610-3-1................ CONTROL DIAGRAM-REACTOR FEEDWATER 2-47E610-1-1................ CONTROL DIAGRAM-MAIN STEAM 2-47E61D-69-1....,.......... CONTROL DIAGRAM-REACTOR WATER CLEANUP 0-47E610-2-2................ CONTROL DIAGRAM-CONDENSATE SYSTEM 0-47E610-12-1............... CONTROL DIAGRAM-AUXILIARY BOILER SYSTEM 0-47E456-1,-2............... RCIC SYSTEM MECHANICAL DRAWINGS 0-47W456-3,-4............... RCIC SYSTEM MECHANICAL DRAWINGS MEL......................... INSTRUMENT TABULATIONS FOR SYSTEM 71 2-45E670-SERJES............. WIRING DJAGRAM-ECCS 2-45E626-1,-2,-3,-4......... WIRING DIAG REACTOR CORE ISOLATION COOLING 2-451670-21................. WIRING DJAG EMERG CORE COOLING SYS DIV I 2-45E714-5,-6............... WIRING DIA 250V REAC MDV BD 2C 2-45E779-5,-12.............. WIRING DIAGRAM 480V SHUTDOWN AUX POWER SCH DIA 2-47E611-64-SERIES.......... MECHANICAL LOGIC DIAGRAM PRIMARY CONTAINMENT ISOLATION SYSTEM 2-47E611-71-SERIES.......... MECHANICAL LOGIC DIAGRAM REACTOR CORE ISOLATION COOLING SYSTEM GE DRAWINGS: 729E652-1.............. FUNCTIONAL CONTROL DIAGRAMS-RCIC 2-729E652-4............. FUNCTIONAL CONTROL DIAGRAMS-RCIC TERRY TURBINE CO.: C-883-X............... RCIC TURBINE GOVERNOR CONTROL SYSTEM NASH ENGR CO. : 0-14-1365............. RCIC BAROMETRIC CONDENSER SYMBOLS: ~ QUICK DISCONNECT WITH SINGLE ENO SHUTOFF ~ QUICK DISCONNECT AMENDMENT POWERHOUSE UNIT 2 28 BROWNS FERRY NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT REACTOR CORE ISOLATION COOLING SYSTEM MECHANICAL CONTROL DIAGRAM FIGURE 4.7-lb H G F E D C B A
1:11011 1-tllllLt-£ " u, ' ii I 8 7 6 5 3 AMENDMENT 27 l'OIEIIHOUSE UNtTJ BROWNS FERRY NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT REACTOR CORE l SOLATIOH COOLING SYSTEM FLOW DIAGRAM FIGURE +. 7-l c H G 8 A
(tO!! l-lL-0 l93Lt-£ 11'-l L9 r-r--,--,-- ~ ~ ~ ~ 71-2A 71-28 71-2C 71-2D 1 L-j 0-45E626-1 I I I --,----,--"t--,--- ~~~~ 71-2E 71-2F 71-2G 71-2H ~~~~ 71-2J 71-2K 71-2L 71-2M ~-3 L __ ------,------- +-=r__-------- ----~-------- TO XM-69-29 /
T=
I -~-------------. -------- (3-47E610-69-1~:::-*------+-------'-.!..--+---------: 1 I ~E~~~~~ WATER', ../rr\\.._ ~9-21 I L--j3-45N3654-1 I ~ +.9-21 LOG LOG 71-41A ~A l 71-418 ~ 71-2 0 71-2 CL r-------1

e--------

' I I I ,.-l...-@:-21 (TE\\_ TS 71-41C 1-41 I I

~

CISS A CISS A 1----------4 t------------, r----j 3-45E714-5
-*-i*--,-*i-----14 3-45E779-121 r----*i--1--i*-~it:-*-1-1
.f.rc /HS\\ 9c3 ~ ___/HS\\ I @-3 4 ../Hs'\\ .fi.irc 11 11 ~_J7~71~ MCC MCC 70A ~\\... 7dC I I ~ I 1,L~3G o I 1c~ss1 J L 1c1;s1 L -"t--,----,--1 ~~~~ 71-2N 71-2P 71-2R 71-2S I I I I ,.-'--. -4:-21 (TE\\_ TS 71-410 1-41 FROM AUXILIARY BOILER SYSTEM 0-47E610-12-1,D3 ~.~5-31 25-58 25-58 I I 9-3 /PI\\ PT I.... /p0 71-48 71-4 t"" -~ I L - - _,.--,--4-5_E6_2_6_-,-. fit:~ !~9-3 NOTES:
1.
THE REACTOR CORE ISOLATION COJLING SYSTEM IS DESIGNED TO COOL THE CORE WHEN THE REACTOR VESSEL IS ISOLATED AND REACTOR FEEDWATER IS UNAVAILABLE. WHEN THE WATER LEVEL IN THE REACTOR VESSEL REACHES.31.5" BELOW REACTOR FEEDWATER ZERO, THE RCIC SYSTEM IS AUTOMATICALLY STARTED BY ACTIVATION OF REACTOR LEVEL SWITCHES LIS-3-SBA, LIS-3-588, LIS-3-58C, AND LIS-3-58D (SEE 3-47E610-3-1), ARRANGED IN ONE-OUT-OF-TWO-TWICE LOGIC TO OPEN VALVE FCV-71-8. THE SAME SIGNAL CLOSES DRAIN LINE VALVES FCV-71-SA AND FCV-71-68.
2.
3A.
38.
A MECHANICAL SPEED GOVERNOR PREVENTS TURBINE OVERSPEED BY TRIPPING THROTTLE TRIP VALVE FCV-71-9 WHEN A TURBINE OVERSPEED OCCURS. A CONTROL GOVERNOR, FIC-71-36, EQUIPPED WITH AUTOMATIC AND MANUAL MEANS FOR SPEED SET POINT CONTROL REGULATES TURBINE SPEED VIA "TURBINE SPEED CHANGER" TO CONTROL RCIC MAIN PUMP DISCHARGE FLOW. A RCIC STANDBY MODE SIGNAL SHALL BE INITIATED AT REACTOR VESSEL WATER LEVEL 8 (HIGH LEVEL). THE REACTOR VESSEL WATER LEVEL 8 SIGNAL SHALL BE INITIATED BY TWO LEVEL SWITCHES ARRANGED IN A TWO-OUT-OF-TWO LOGIC CONFIGURATION. THE REACTOR VESSEL WATER LEVEL 8 SIGNAL SHALL: (A) CLOSE THE STEAM ADMISSION VALVE FCV-71-8 AND (B) CLOSE THE MINI FLOW ISOLATION VALVE FCV-71-34. IN THE EVENT AN INITIATION SIGNAL IS RECEIVED AFTER THE SYSTEM HAS BEEN SHUTDOWN ON HIGH RV LEVEL, THE SYSTEM SHALL BE CAPABLE OF AUTOMATIC RESTART. THE A.
8.
C. D. E. RCIC TURBINE TRIP VALVE IS TRIPPED BY: TURBINE OVERSPEED VIA THE SPEED GOVERNOR. DELETED. LOW RCIC PUMP SUCTION PRESSURE SENSED BY PS-71-21A. HIGH RCIC TURBINE EXHAUST PRESSURE SENSED BY EITHER PS-71-13A OR PS-71-138. MANUALLY FROM THE MAIN CONTROL ROOM BY HS-71-9A OR BY HS-71-9C ON PANEL 25-32. A TURBINE TRIP:
4.
A.
8.
C. D. CLOSES FCV-71-9. CLOSES FCV-71-34. ANNUNCIATES ON ZA-71-49A OR 498. ACTIVATES THE ALARM ASSOCIATED WITH THE CONDITION INITIATING THE TRIP. POT .-----~ 7~ ','-tL_}--Cxx::JI-0>----';-,;~25-7 A 25-7 A r -l 3-45E 626-3 ,-------------------------------------------r-,--,-, r --+-,------,--,----, r-:-..25 321~<1 1.-1,--------------..... I 9-3 : ~5-32 I~
5.
THE TURBINE CASING IS PROTECTED FROM OVERPRESSURE BY TWO RUPTURE DISCS. PS-71-11A THRU O SET AT 10 PSIG DETECTS DISC RUPTURES AND ANNUNCIATES ON PA-71-11 IN THE MAIN CONTROL ROOM. THE RCIC TURBINE WILL BE AUTOMATICALLY ISOLATED FROM THE STEAM SUPPLY BY CLOSING ISOLATION VALVES FCV-71-2 AND FCV-71-3 IF: PDT FT I
~
71-18 71-18 ------,---, I o-45E626-1 ~1* Y ~ .. ~.. 9-3 f-tL_}--Cxx::J-O>--,-----;l--1 25-7A 25-7A ~ ~ I PS PS 9-3 9_, 8 2s-1A I 25-7A 1yA 1yc r,;'\\ - ,,-;;;<i71F~1 FT I PDT +----+-----f3-45E626-2 ~~ 71-1A~1-1A ~5-7~25_7A I 1--t 9-82 DIS I/PS\\ (PS\\ 3-45E626-3 1 71-18 I 71-18 71-1D CONDENSATE TO HPCI TEST LINE .__-II AND CNDS a:. DEMINERALIZED WATER ~ STORAGE SYS 9 18 I,- 9-3 9-81 3-47E61D-73-1 CB ______ _J PDA DIS I 71-1 I 71-lA ECCS ANALOG TRIP POWER L-..::-:::::~-- --<3-47E610-71-2 I 3-47E610-11-2 _..., y 3~20-3 I 1 __ ~r,--;--y,;---, 3-47E610-1-1, C6 71-58 I I ,, /~ ~ ).~{ I lc~~;31 To **rN STEAM SYSTEM ~ 1 (HS'\\ r";is\\s-3 I 1 *cc 11-300 ,-;':'rs', r, Y ZI-71-408 Y Y
Ir (DISK POSITION) 11 - 58
~.Ji,( I I R G I/~ 5 I~ I A I _. ___ _/xs\\_../Hs'\\ I 71-38 71-38C I II 4 3-+5E714-6 I CRW I .... --j3-45E714-5 I r Hs, - 1;:;<r I 1 1 r--------,,--,---,
: + '@f.
~ ~ I FROM CNOS & DEMINERALIZED 7~c + ',ff I I I CTiiiD._..J/ I I i /x, r-:-..~' 9_3 11 WATER STORAGE SYS A5-32 /-@., ~ 3 I ~ I I G ( HS \\.._.J XS\\.. HS 0-47E610-2-2,G4 ( XS'L HS I I ,1-a 71-19C 71-19 1-19A I 71-9A... 71-9 I I L+*MCC ~ ~ fHS\\... __ -::---___.:r: __ ':'_ jrT<lc,,__, I L-t------0s II 71-198 3-45E626-2 ~ I I ili-.:..-i __ _, _r-----+---- rt 7R'L J 0-45E626-4 ~ 11 I I ~ I o-45E62S-1 f,.l xx 11 ..-j 3-45E714-5 I R G MCC I 3-45E714-6 f-.... 1-9 11 I. . f'iis\\9-3..... 3-45E626-3 I-1 9-3 I IP?\\ I 7i':35', I 25-31 'f----_J ~ LOG ~--+ A71033 UQ[]_ji ~9-47 rl 3-45E626-3 9-3 (';;'\\. 71'=2DA .._, I I ,,-l-.....25-31 ( PX\\ 71-20 25-58 25-58 PI 71-208 r4 3-45E626-3 I: ~-,- 1_9 I CTiiiD._ __ ~ ~
@J::
I I I I~ @-r-:-.. I ~-3 I 0-45E626-1 H TRIP LINKAGE L XS _/ HS ' I r"tf':m I I --i o-45E626-4 I 71 - 0 7.!.."3c L_... 25-::-;; I @9-3 A f "*9-3 FROM XS-71-368 j HS J XS \\25-32 j f'A -<ill t---r-1 1 11-s 11-90 1 c 21-11-10 r 08 25_32 I Y 1 25-672 ~ 25-58 -+ ~25-32 I THIS FC St.4 I 5 ~2 PS HS .J DWG,CS 71-10 71-10 11-21A @--- 11 ~ ~ Y I '-' I 9-3 I ~ J, I ~OG 9-31 I 3-45Es2s-2 ~- ___ _J J~~JA t-- I 3-45Es20-2 6~A~cs ~ I As,ss fsi'\\ : ~ 65 672 ~ I I LG ,), _1 ___...J I '-('- JB 10035 DRAIN POT CPOT-5 3-47E3847-5,F6 n.T.-.,~x I ! FSV ~ (PT' 7~13 11-35 y RCIC PUMP PMP-19 RCIC TURBINE TRB-9 F'---~ SE I 71-10A 71-42A_L, 1-1 B I 7DA I ,----------...J -,-j-o--4-5_E6_2_6--1--, I I 3-45E626-3. -4 I I I I I I I I I...J r s I STEAf.t TRAP TRP-lD PISTON REMOVED 3-45E626-3, -4
6.
A. A HIGH PRESSURE DROP OCCURS ACROSS THE RCIC ELBOW TYPE FLOW ELEMENT FE-71-1A & 18 WHEN THE STEAM LINE RUPTURES. THE HIGH FLOW IS SENSED BY EITHER PD1S-71-1A OR PDlS-71-18 ANO ANNUNCIATED ON PDA-71-1. B. HIGH AREA TEMPERATURES INDICATING A STEAM LINE BREAK ARE DETECTED BY ANY OF FOUR SETS OF TEMPERATURE SWITCHES (TS-71-2A THROUGH TS-71-2S SET AT 194°F). EACH SET CONNECTED IN ONE-OUT-OF-TWO-TWICE LOGIC. FOUR INDEPENDENT TEMPERATURE SENSING ELEMENTS AND ASSOCIATED TEMPERATURE SWITCHES TE-71-41A AND TS-71-41A THROUGH TE-71-41D AND TS-71-41D SET AT 175 F ACTIVATE THE MAIN CONTROL ROOM ANNUNCIATOR TA-71-41. C. LOW REACTOR PRESSURE IS DETECTED BY PS-71-1A THROUGH PS-71-lD EACH SET AT 50 PSIG AND ARRANGED IN ONE-OUT-OF-TWO-TWICE LOGIC. NORMALLY MAKEUP WATER FOR THE RCIC PUMP SUCTION IS SUPPLIED BY THE CONDENSATE STORAGE TANK THROUGH FLOW CONTROL VALVE FCV-71-19. AN ALTERNATE SOURCE OF WATER IS OBTAINED FRc:>>.1 THE SUPPRESSION POOL THROUGH ISOLATION VALVES FCV-71-17 AND FCV-71-18. PUMP DISCHARGE FLOW ROUTES ARE AS FOLLOWS: A.
8.
C. NORMALLY VIA THE FEEDWATER LINE TO THE REACTOR THROUGH ISOLATION VALVES FCV-71-37, FCV-71-39, AND CHECK VALVE FCV-71-40. DURING TESTING VIA TEST LINE TO CONDENSATE STORAGE THROUGH FLOW CONTROL VALVE FCV-71-38. MINIMUM FLOW BYPASS TO THE SUPPRESSION POOL THROUGH FLOW CONTROL VALVE FCV-71-34.
7.
IF RCIC SYSTEM OPERATION IS REQUIRED WHEN IT IS IN THE TEST MODE, THE CONTROL SYSTEM IS DESIGNED TO AUTOMATICALLY RETURN THE SYSTEM TO THE OPERATING MODE.
8.
COOLING WATER FOR THE PUf.tP, TURBINE, AND LUBE OIL COOLER IS SUPPLIED FROM THE PUMP DISCHARGE LINE THROUGH PRESSURE CONTROL VALVE PCV-71-22 SET AT 75 PSIA.
9.
TRANSFER AND CONTROL SWITCHES ASSOCIATED WITH BACKUP CONTROL WILL BE ON THE DESIGNATED PANELS OR LOCATED ON THE APPROPRIATE 250V DC MDV BOARDS.
10. XS-71-368 IS SHOWN THREE TIMES TO SHOW HOW THIS TRANSFER SWITCH DISCONNECTS FIC-71-36A AND CONNECTS FIC-71-368 LOCATED AT THE BACKUP CONTROL CENTER.
~-3 25-58 I ~ 25-58 ~ CRW~ >-<._ 25_58 PI,_ _ _, /FIS'---- l 71-358 TE ~'--_4_5_E6_2_6_-_3~r-I I I (TE\\ SEE NOTE 11 71-48 Y (xM)25-672
71-10 ly I
I .3-47E3847-5,F6
11. XS-71-23, XS-71-45, TS-71-45 a:. -46, XS-71-24 AND PS-71-44 ARE EACH SHOWN TWICE TO SHOW HOW THESE SWITCHES TRANSFER INDICATION FROM THE MAIN CONTROL PANEL 9-3 TO THE BACKUP CONTROL PANEL 25-32.

12. UNIDS ON DRAWINGS ARE FOR REFERENCE ONLY AND ARE ABBREVIATED AS SHOWN IN THE EXAMPLE TO MEET SPACE CONSTRAINTS.
REFER TO MEL FOR COMPLETE UNIDS. ALL UNIDS ARE IN UNIT 3 UNLESS OTHERWISE NOTED. LEADING ZEROES SHOWN IN MEL AS PART OF THE UNID ARE NOT DEPICTED. FOR ADDITIONAL GUIDANCE, REFER TO NEDP-4. 4 71-36 71-33 FE I TW t-11-36 I 11-33*'------..-------- r T, -,-~-,.>-<.25-58 I I I w FCV I I I I I '>:k< FCV I I I / FT l l~71-39'>*<'~I I 1~11-31W~l1~6 25_31 I ___ _ 1 .cc Y ~ >9,1 1
cc Y /~.( ~
1 +-.../PX\\ r 1 +-,--,-, I f'iis\\9-31 I I 19-3f'iis\\ 7~A I I FCV J_,. I I I i I 7i':39'A I
i I
~. I I 11-34,QQ_* ii G I y I G I y [i][J__J I y MCC R G I / l.,.t._1, l.,.t._ ~ I/ I I L ________.... CKV-71-40 t-..... Jf--PRESS. TEST t-----
A I
705 25-425 TS SEE NOTE 111 71 _4 I I I 3-45E626-2 ~-+ _______ J --l 25-31 1 ~~ NOTE 11 '~ -~~~ TS TW 71-56 l,~;6~ 71-45 71-45.._... TE LsEE NOTE 10 71-47 TW SEE 1-45A NOTE r ,--1 ---,11 ,, o-45E626-4 r r-- ------------, ~5-58 ~5-31 I I I I Ir.c~X I s TT -*~-~~:-45E626-1 I~ THERMAL SLEEVE~~ ~-... _..fxs, I (iis\\..__.. ~_fxs, I i L-A-~ I 71-398 I 71-39 I 71-378 1 71-37 I 71-348 I !-~~-PRESS. TEST 1 I ( e, 1 LOG ( PX ~----, I __ _J 71-128 A6152 1y21 3-45E626-3 h I ~5-32 I ~:~ 21-71-6AA I I I I TO SM-71-10 I I I XS I I r !.... 1,-...J ....!__._J _J r r-,--.--..-,--,-, I ' 1--.,~ r-71-34 I WW I 0-45E626-1 t--, is ~ 3-45E714-5 rJ ~I*.... 13---45-E7-1-4--5'P '@'1~ I
;8( ~:
I 3-45E626-3 r-1 !. ~ ~ ~ /~ ,"1:,1,1---ri,1--__J R /+ c+, +, 1 8 ~ ~ 7~C 7t:JC ~ I /,I_,, /~lfez r-:<'-9-. A~~ I r---,.---, A5-58~3 ~-58 (PS\\ PA (PS\\ 71-138 71-1 71-13A 25-58 I 9-3 I : ~ 1--'--'7];2 t_*_@. ____ J: ~~ T ~w U VENT~ 1 I _.Q ~(.? 25-58 25-58t- 00 ::C:ll= I I 1-U 1-C PS PS 25-31 I ~9-3 I ~25-31 "f I ~9-3 ~ ~1~~5["".-f;,x_s2~A ~.,. XS --+- FIC __.,.. __....( XS ~-----------------------...J I ~A 71-34C I - ~ 1-36 1-3SA 71-368 1 1 71-11A 71-118 I Y Y -@9-3 .. ---+--_. PA 71-1 8 I y --., SEE NOTE 10__/'r" ... ~-----------------1---------------,--,----l-------'----(PI\\..,,,P""r ~ (TI\\ L--J 3-47E610-71-1 I t-j 3-45E714-5 i!lL 71-27 ~5-58 ~-58 8 L----_/fic\\_5-3~----.J SEE NOTE 11 71-23A MCC ~ ~~ ~ 7~8 1r~44)-,-...,,<'--'7i!23"A __ _,7 ~~23 SEE ~ 7~:15 - -Lr_--:_~_j-j::.......::::::'..~~~L=~--,7n1-27,.L-----~t----::::::----------' 71-11C 71-11D ~ ~ ----ili-fu ../Hs'\\ 71-59 I I l@-3 XA I 11-5 I '-' L-j3-45E779-12 I 7 ~ NOTE 11~ TE NOTE Pl 703 7..!..:34 11 7~1 LUBO w z - T m w ~ ~. L " u C - u "' ~ -" 0 ~ L t-.... t---1 RHR TEST LINE 3-47E811-1,CS 9-3 Zl-71-32f (sHV\\ 71-.32 V,L WV> ~w ~L w z - T m w ~ ~. L " u C - u "' ~ -" 0 L L 1---,,,._--,,,._--I COOL ER FILTER CLR-25 3-45E620-3 L-, THIS c_ ___ _,r DWG, D3 I FROM RCIC FG 3 !TURBINE 71-507 -~IL PUMP---- LS 71-29 I 3-45E626-2 1-:::. BAROMETRIC CONDENSER CND-27 VACUUM -TANK - -TNK-27 r-,---,--,--,-~--------+ W '@f. 'i< + '@f 'i< I 3-45E714-6 ~j,-PJ....S,'-r---+-', ~ / ' ~' /~ / '~ - 111-28 ~ /HS) f'iis\\9-3 fHS\\...-------+:::: p:-3 TO CRW 71-29C ~A 71-298 \\ r,1:-,1/ '[ ___ '::::-:_-:_ '-' _-:_-:_-:_-:_-.c~/\\_.,....,-..._ 7 ~:DS ! ~ ¥-, _/ PUMP_.,.,1...., 71-318 ~ (!) ,y-14-----------1.,.-"l----------!PMP-29 't)!!.("°C l!l r-r------------~~-----~~--~~-------------------.~P"~P~-3~1 1 1 I,--,--,--,-.,.---j 3-45E626-2
~ ~ L--j3-45E714-5 I I I.@_9-3

ilii ~
I fHs\\ A I J1~11c" 11-11 "l +-~-~~l/ : ~ : ,,,,(9), I I 3-45E714-5 13-45E626-2 \\::~/""1 L /H5L I ZI~;:{sJ ll:!JB --, 9-(R (sHV\\ I ~,~c /, 11-1s IL 11-11 .___-;I: --- 6 r-,-.-,-"f, '>:k< ',ff I I I ~~I II ~-3 1 11 ~A.J_...J_..I ~-, y ,QQ_.)9,1 ~-+~! I,-~----, I R G I I I l _/HS\\ I r 1~s I ~~ I /i:: 71-18 ______ J X X L----+r.. -------~~----* TO CRW I I I I I I I I L
==1 FCV 71-78 I
25-3©'
3-47E3847-5,G7
~:7 L-j~0_-4-5-E6-2-6-_4~ I I 3-45E714-6 .3-47E3847-5,G7 L-~-... ---,...!-, Zl-71-7AA* ~ ~!:©: +---iL_o_-4_5_E_6_2_6-_1~ ~!~IC :&hr---ili_.,_J 5 R I@/ R '1f ZI-71-7BA / '~' 9-3 9-3 MIMIC 7 4 VACUUM A I PUMP (XS\\_,.__ 71-31 ,-T 21-71-S~A,--,1;-w /,59_ ~/' HS 9-3 9-3 1 _6 MIMIC TO MAIN CONDENSER
~
9-3 MIMIC --j 3-45E626-2 r --IL_0-_4_5E_6_26_-_1 -' _J I LOGIC 8 I 9-3 9_3 r--, 3-45E626-2 RESET j ~ I r--~ /v2'\\ ! RCIC AUTO ISOL .... ___);;;< ~ ~-, LOGIC A RESET ~ IL-71-518 ~IL-71-51A 9_3 r-l o-45E626-1 I /v2'\\ l _ RCI£...~I.!]AT!,_()N ~e-rSEAL-IN RESET ..._.... '+ IL-71-52 9-3 ~ ~ RCIC AUTO INITIATION 71-5 ' -MANUAL-ISOLATION ,__, I L 1 3-45ES26-2 9-47 (TE\\_ ~ \\?.2_:.,24 73010 ~ 9-47 (TE\\_ __ @ --- LOG \\z.2.::..47 73011 A71047 1@: 47 l,1',~481 ©--,- (ffj:3 I 25-32 .. --~ XI-71-44 I ~--~ L_____ 3-45E626-2 0-45E626-4 ©--~-~ @:-3 I 25-32 iL..---0' XI-71-23 I /' ~----~ L_____ 3-45E626-2 @s---Ws-~ @:-3 I ~25-32 0-45E626-4 iL--- A Xl-71-45 I / ~---~ L_____ 3-45E626-3 /TS\\_ d-31 _ill::}-3 \\z.2.::..4s - - '-Q:.4s T 'Z.S46 I 25-32 0-45E626-4 COMPANION DRAWINGS: 1-47E610-71-1 2-47E610-71-1 3 iL..--~ Xl-71-46 I ~ L------j o-45E626-4 EXAMPLE: MEL UNID DRAWING UNID BFN-0-SHV-018-0502 SHV-18-502 BFN-1-HS-031-0160A HS-31-160A BFN-2-PT-003-0204C PT-3-204C
13. FLEX HDSE{llf\\J'i~ TO BE INSTALLED FOR TESTING PURPOSES ONLY.
REFERENCE DRAWiNGS: MEL.......................... INSTRUMENT TABULATIONS 3-47ES10-1-1A................ CONTROL DIAGRAM-MAIN SYSTEM 0-47E610-2-2................. CONTROL DIAGRAM-CONDENSATE SYSTEM 3-47E610-3-1................. CONTROL DIAGRAM-REACTOR FEEDWATER 3-47E610-69-1................ CONTROL DIAGRAM-REACTOR WATER CLEANUP 2-47E611-71-4................ LOGIC DIAGRAM-RCIC 0-47E800-1................... FLOW DIAGRAM-GENERAL PLANT SYSTEM 0-47E800-2................... MECHANICAL SYMBOLS~ FLOW DIAGRAM INDEX 3-47E610-73-1................ HPCI SYSTEM 3-47E813-1................... FLOW DIAGRAM-RCIC SYSTEM 3-45E779-12.................. WIRING DIAGRAM-4BOV SHUTDOWN AUXILIARY POWER SCHEMATIC DIAGRAf.t 3-45E714-5,-6................ WIRING DIAGRAM-250V MDV BOARD 1C SCHEMATIC DIAGRAM 0-45E626-1,-4................ WIRING DIAGRAM-RCIC SCHEMATIC DIAGRAM 3-45E626-2,-3................ WIRING DIAGRAM-RCIC SCHEMATIC DIAGRAM 0-47E610-12-1................ CONTROL DIAGRAM-AUXILIARY BOILER 3-45E620-2,-3................ KEY DIAGRAM-ANNUNCIATOR SYSTEM 3-47E811-1................... FLOW DIAGRAM-RHR SYSTEM 3-45N3654-1.................. WIRING DIAGRAM-UNIT CONTROL BOARDS PANEL 9-21, SH 1 3-45W670-SERIES.............. WIRING DIAGRAM ECCS GE DRAWINGS: 729E652, SH 1 THRU 4.... FUNCTIONAL CONTROL TERRY TURBINE CO. DIAGRAMS-RCIC C-883-X................. RCIC TURBINE GOVERNOR NASH ENGR CO. CONTROL SYSTEM D-14-1365............... RCIC BAROMETRIC CONDENSER AMENDMENT 26 POWERHOUSE UNIT 3 BROWNS FERRY FINAL SAFETY NUCLEAR PLANT ANALYSIS REPORT REACTOR CORE ISOLATION COOLING SYSTEM MECHANICAL CONTROL DIAGRAM FIGURE 4.7-ld H G F E D C B A
D1D11l-Ll-OLl3Lt-l,. LI ,.,.,_Ji, 8 7 5
~-H
-~
~
" '"".;rm!f.~*..........,..,..,, _____ _ -:.,o._\\l,,,.............. --- AMENDMENT 27 l"OIEIIHOUSE UNIT 1 BROWNS FERRY NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT B REACt8SL~SEsi~~~QTION A MECHANICAL CONTROL DIAGRAM FIGURE +. 7-le
ltO~ l-£ l83Lt-l 3 L9 I I PRIMARY CONTAINMENT~ CONDENSING-~'-'C,!:P~OT-1 A L CHAMBER r..J'""l.i_j-------.1.1..,"rl_-~-'Xa.-.:5a.1:..F...p~-:":1;J'°'.:"---'. {TYP) J; I 216A ECKV-71-lAA ~ 1
X-51 E H
PS, _, MAIN STEAM \\\\-<=1=-~.,-E~8~0=, _=,->-+-I-ill r-----..L..~l-i:;~.:a.a!a.~2X 17~AI--E-CtKV~-*7-,---,A-B-~S~HV~-~71-1BH rOT-lB 1" X 33F H 1" VTV-71-lBH -© (rcV'i 7~6 10" -y - -© to CONDEN~AT: SUPPLY SYSTEM (1-47E818-1) l"--l FE). 71-1A I ~ I 221A ECKV-71-1BB "-1/2" 1" X-33E L 1". 1/2" ,_+~=~~r-1~\\_~-~~--~~--~~---~ FCV 1-2 222A FCV ECKV-71-18A 1-3 .-----I~ r-X-10 SHV-71-lBL VTV-71-lBL HPCI PUMP TEST l FROM CONDENSATE SUPPLY SYSTEM CONT ON 1-47E818-1 t----<:1--_ FLG-71-1 SEE NOTE 14 3/4" _ SHV-71-62 ~: ~ ~ ~ ~ LOV-565 . T .,,i AUX BLR SYS 1/2" 3* L,:..._...:4.:.7:.E8:..1...:5_-.:.3:..*...:c...:*_,,)>------*'"~----1"-..!-,;..-, C:'5 CKV-564 PRE-STARTUP / STEAM SUPPLY_/
REACTOR STEAM DRAIN __,,..,..
~. :::. i ili 1/2" 8
t_
r--llOII--'.:..:.+*-{ CKV-556 - 1 " ~ z - ~ FE 1-36 N TO CONDENSER A~ CKV-3-572 3/4" TEST HPCI DRAIN (1-47E812-1) 1 X-98 ~ N ~ ~ t I TEST 1" VENT-DRV-556 SHV-556 1 " VTV-530 .--,~--1<1--..,..-..!....! VTV-529 ~TO N0.1 JET PUMP INSTRUMENT PNL 25-51 DRAIN LINE .... 1/2* TELL-TALE 6 -t+/-J CRW
v f--Q_FCV 71 I.,.......--PUMP TEST RETURN TO Lr-"""
CONDENSATE STORAGE /'""3/4" LOW POINT CRW SHV-553 DRV-554-DRAIN 1.5375" PR0-38 ~ (rcV'i L/7r\\_ w .--3/4" CRW 6" p~:" 1-2o RTV-20A. w PS 71-21A RTV-21A PS 71-218 RFV-19 1
RELIEF VALVE SET o 150 PSIG CRW

~3/4" VENT m
j 1 SPOOL PIECE SPPC-71-19 SEE NOTE 13 N ~ ~ 1/2" DRIVE >i'!,,. >_SHY,-619 151 TURBINE I TRB-9 1/2" PRESS. TtST.~ RCIC PUMP PMP-71-19 (616 GPM) ABOVE AND l\\. - T BELOW SEAT L.J::-1.!j DRAINS -~-I PR0-9~CTS.~~::;-J~~::.-..J*-'f,~~ ~ PR0-98 FCV 1-8 '1'" 4" RTV-4A RTV-5H m 1 " . I. 1 " 8~~1t ~g~PL y CPOT-5 1 " RTV-SL rr LS 1 ~,...-NOTE 12 1... 1-;.9 - ~ SHV-19D 71-3-7 ~ ~ ~6 ~ ! CKV-71-40 6" t-j::. r.. --t,1-1"------.... -,-... 4'----T"'-_.,--t~J----r--*~_..--t[7FE,l,-.._~... -..-... --f TW / r--... r TV-19 I r.:;;-.........., ..-' Lwfi-";~~=~=====~===;c~PL[]~~*======~,:~~[ID;a~__::9-=;=-----'""'P--"PAe,c:,K"1,:,NGe...,De,Re,A.c1 ""---~ [zJ ~ u Ill cl:J 00 _.-:..... l m THERMAL SLEEVE L.:. < 1-36 71-3.> m o 0 "II" ~ IO II) ~ ~ 7 ~ r--t-**iJi--,-3/4" ~ ~ ~ ~ ~~I~~WDRAIN ~ ~ 8-1 - ~ g: ~ co u, ~ PT ~ ~ 10 71-35 i::: t i LL\\) Jt-1 :s TEST 7 7
\\_0.375" PR0-547 CKV-3-568 v
N TEST TEST CRW ii 4 PUMP MINIMUM / ._ ~ FLOW BYPASS 1 _____,.,......- , ?
I, s:;:
0 r3/4" 41 m v ~ ' ~ .~ N ~ ~ ~ *~ 7~~55
2m f
<2 ~ 0, ~ 1 a, y DRW ~ C 1-1L4" BRACKET,...L:"~3/4" CASING DRAINS I:. D~AINS DRYWELL TEST PS 71;)38 CRW PS 71-3A @1D 1/~ @1c PS i ~ ~ = ,., £:". t'* -'--,---'-'-Kl-, z - ~ C 1
PI 71-56 0
~~ V> ' m * ~m V> ' STEAM TRAP l TRP-10 PISTON REMOVED ">~ ~~ V> ' 1 *
. {0:1.
§A ~~:_~1. ___.i ... -------------1-.. ---' y* '¥"' ,_..~.,..-1-i>l<*h-N-t:,l<:]--....J ~ V-574 X ~ X ~ 11 1
t.
AUX BLR SYS I 1-47E815-3.C6 )>--t 1/2* GASKET TEST-REACTOR FEEDWATER LINE 8 ~
1 "{f'=
1t 1 -- - RPD-118 N CKV-598 N
2" CKV-597 (1-47E812-1) 2" 2"
I, 3" '-.: CKV-600 cKv-Sss X-218-----.__ z z - z 8
i -
D.. N 0 ~ ' ~ / [3/4" TEST SHV-602 TV-603 m SYSTEM PRESS. - TEMP DATA LINE DESIGN MK NO. PRESSURE (PSIG) 1 2 3 4 5 6 7 8 9 11 14 8 1146 150 1375 1500 150 1500 1500 1250 1 65 400 200 DESIGN TEMPC°F) 562 325 545 200 200 376 40 TO 140 575 366 450 200 2" LO - SHV-s2o TURB EXH VAC RELIEF _J N l N ~~ ~ 7...!.;J. 4 LOV-683 ~. 1/2" v 0 W N V. CKV-580 m ~ ' ~ N m 2* 1 o* TURBINE EXHAUST TEST--.,: ? , i'! 3/4" TEST.
o. 625*
[3/4" LOW POINT DRAIN PR0-55"\\l----t,:,4-.L..--IIOll----l-~CRW ~ (SHVl SHV-551 DRV-552 71-32 T 2" N 14
~ ~
i_: 18" ,/ ~ 18" 3/4" TEST__/ l"'_;;_ __ ..i.; __ ...;. __ _; ___ -,~--FROM RESIDUAL HEAT REMOVAL SYSTEM X-221--.._, .,,,,-x-210A J X-212 N {TI\\ 71-46 ~ (TE\\ r.:::-. 71-48 I TS\\ N CKV-592 N TEST ' m TV-540 TV-541 ~. :;: g ' *~ m,., PRESSURE CONTROL VALVE (1") 1 o* .--1/2" LOW POINT DRAIN f---tc~-~.L.'--to-~-1----~-CRW sHv-538 oRv-539 ,,---0.416" PR0-540 N 3/16" f RPD-11 A PT 1't}J) RTV-12A PI 71-28 EXHAUST m DRAIN POT----......_:""',_ ___ STEAM TRAP TRP-9 m 7 :;:; ,a~ 1
I=
PRESSURE RELIEF VALVE SET o 10 PSIG 2"' ... ji{RFV-27 ~ CRW RTV-27L Tl 71:;J-7,------C:,. LG 71-27
~
71-46 TW I 1-1/Z-TW ~ V> z w 0 BAROMETRIC CONDENSER CND-27 ~ i~.,,,, rm 1-1/4" SUPPRESSION POOL ,-r-STRAINERS a SUPPRESSION POOL RING HOR 7 ~ 7 6" (rcV'i Tl 7.1.,:!5 71-46A~--.pt-.._..._..._.__,71-46 ~ TW 71-45A - .375.ll. J 0 ~ ~ ~ PMP-19,., o ~------------------ - _.:,:...:.-1-------~- 3/4"

- -------=e....:...--1

=~~ f1------,.l!:"3~~8~*~*,-...._3~/~4.*~, PRESSURE RELIEF VALVE SETO 75 PS~ RFV-543.. N 71-25 LUBE OIL fr~~: 8 140*F MAX CLR-25 1/4"- 1-112*~;1;.,...._..2.:1-:!;l/~2::.*---, /re\\ 1e1 ~,I I l:"r'"" j§~ ~ ,........ RCIC a;;;.. ~ 3/4" 7!-2_4 TURBINE ~ ~1-1/2' TRB-9 o 71-507, (rcV'i C.- ~ I 5 L..--1..:.11---....1.--1->t---' - TW 71-45 J" ~
,_ VENT 71.;j.5
.\\. 1" ~! ....._1J'2" ,1---'~=.-------1 . 756 11 OIL 3/8" ~ r-'*-o,1rt>llrlp,..p ~, *112*
PMP 45
'j"'". ~ ~ i _( u m G PI PS 71-501 71-44 1/4" 3/4" N --~===--....... 1 /2" 1/2" (TYP 4)-- 2" ~PRESSURE TEST LS 1,--' 71-29
t. -'
VACUUM TANK ] TNK-27
t. -'
RTV-27H PCV 71-3""--,. BYPASS ~ CONTROL,., VALVE !"' ' :!l ~J 2 I'"'">- I") $ ii) 71-17 'ii' LOV-17 TW 71,:36 3/4", 6" PIPE \\. COLL-26 1-1/4"~~---- I r. '-CORE SPRAY SYS CONT ON 1-47E814-1 1 /2" 3/4" TEST___,,, w S:J CKV-508 ~ 6 0 ~ . 750" TS TW t-.... 71-23 71-23
~
LI.IC SUMP :1--'--' ~ 1/4"
,,-a/4 "'
TI TW 71.::3." 71-23A l RCIC LUBE OIL PIPING DIAGRAM 2" 5 + + ~ ' ' ili 1 /2" LOW '° POINT DR----........., f ~ 6' ',c C W 2" I .PMP-29 / CKV-588 CONDENSATE CKV-589 PUMP FCV 43 GPM
_,c-,,71-7A 3/4" TEST '~:J-*----o4....::!.:.-l.. f---!
TV-591 SHV-590, FCV .*"*'1-78 CRW 2" - ~ r 1-1/4" w (!) 1PMP-31 / VACUUM PUMP 17 CFM VACUUM PUMP DISCHARGE TO SUPPRESSION VESSEL COMPANION DRAWINGS: 2- & 3-47E813-1 CALCULATION
REFERENCES:
MDC0000712012000031: RCIC PUMP NPSHc AND SYSTEM HYDRAULICS 3 2 TEST
NOTES,

1. CRW DRAINS ARE CONNECTED TO REACTOR BUILDING EQUIPMENT DRAIN SUMP.

2. DRW DRAINS ARE CONNECTED TO REACTOR BUILDING FLOOR DRAIN SUMP.
J. ALL VENTS AND DRAINS SHOWNS.0'4! ~ INDICATES CLOSED SYSTEM. VENTS AND DRAINS SHOWNS *=* f INDICATES OPEN SYSTEM.
4. ALL VALVES ARE THE SAME SIZE 1: PIPING, UNLESS OTHERWISE NOTED.

5. ALL PRESSURE AND TEST CONNECTIONS ARE 1/2", UNLESS OTHERWISE NOTED.

6. ALL VALVE DRAINS AND VENTS ARE PIPED TO CRW.

7. UNIDS ON DRAWINGS ARE FOR REFERENCE ONLY AND ARE ABBREVIATED AS SHOWN IN THE EXAMPLE TO MEET SPACE CONSTRAINTS. REFER TO MEL FOR COMPLETE UNIDS.
ALL UNIDS ARE IN UNIT 1 UNLESS OTHERWISE NOTED. LEADING ZEROES SHOWN IN MEL AS PART OF THE UNID ARE NOT DEPICTED. FOR ADDITIONAL GUIDANCE. REFER TO NEOP-4. EXAMPLE: MEL UNID BFN-O-SHV-018-0502 BFN-1-HS-031-0160A BFN-2-PT-003-0204C
8. [] ETC, DENOTES DESIGN PRESSURE AND TEMPERATURE AS GIVEN IN TABLE, THIS DRAWING.

9. HYDROSTATIC TESTING SHALL BE IN ACCORDANCE WITH THE APPLICABLE CODES.

10. THE DESIGN PRESSURE AND TEMPERATURE OF ALL DRAIN AND VENT LINES THROUGH THE LAST ISOLATION VALVE SHALL BE THE SAME AS THE PROCESS LINE.

11. VENT, DRAIN, AND TEST CONNECTIONS 1-1/2* AND BELOW CAN BE PROVIDED WITH PIPE CAPS OR HOSE CONN(CTION FITTINGS WHERE REQUIRED BY PLANT PERSONNEL. THIS CONFIGURATION IS SUPPORTED BY ENGINEERING CALCULATION CD-00999-923399.

12. VALVE FCV-71-39 WILL SEAL LEAK TIGHT ONLY IN THE DIRECTION OF FEEDWATER TO RCIC SYSTEM DUE TO A PRESSURE LOCKING MODIFICATION WHICH DRILLED A 1/4" HOLE IN THE FEEDWATER SIDE FACE OF THE VALVE.

13. SPOOL PIECE MAY BE REPLACED BY STRAINER DURING MAINTENANCE.
THE SPOOL PIECE MUST BE INSTALLED PRIOR TO RETURNING THE SYSTEM TO SERVICE. SEE 0-47E456-1. -2 FOR STRAINER DETAILS.
14. THE PIPING COMPONENT IDENTIFIED IS A LINE STOP FITTING WITH AN INTERNAL PLUG AND BLIND FLANGE. NORMAL CONFIGURATION DOES NOT ISOLATE FLOW AND FITTING. PLUG AND BLIND FLANGE ARE CONSIDERED PRESSURE RETAINING.
REFERENCE DRAWINGS: 0-47E800-1........... FLOW DIAGRAM-GENERAL PLANT SYSTEMS 0-47E800-2........... MECHANICAL-SYMBOLS & FLOW DIAGRAM DRAWING INDEX 1-47E801-1........... FLOW DIAGRAM - MAIN STEAM 1-47E811-1........... FLOW DIAGRAM - RHR SYSTEM 1-47E812-1........... FLOW DIAGRAM - HPCI SYSTEM 3-47E815-5........... FLOW DIAGRAM - AUX BOILER SYSTEM 1-47E818-1........... FLOW DIAGRAM - CNDS STORAGE & SUPPLY SYS 0-47E820-1........... FLOW DIAGRAM - CONTROL ROD DRIVE HYOR SYS MEL................. INSTRUMENT TABULATION 1-47E610-71-1........ CONTROL DIAGRAM - RCIC SYSTEM MEL................. VALVE MARKER TAG TABULATION 1-47E852-1.......... FLOW DIAGRAM - FLOOR AND DIRTY RADWASTE DRAINAGE (ORI) 1-47E852-2.......... FLOW DIAGRAM - CLEAN RADWASTE & DECON DRAINAGE (CRW) AMENDMENT 28 POWERHOUSE UNIT 1 BROWNS FINAL FERRY SAFETY NUCLEAR ANALYSIS PLANT REPORT REACTOR COOLING CORE ISOLATION SYSTEM MECHANICAL FIGURE CONTROL DIAGRAM 4.7-1f H G F E D C B A
BFN-22 Figures 4.7-2a through 4.7-2h (Deleted by Amendment 22)
BFN-26 4.8-1 4.8 RESIDUAL HEAT REMOVAL SYSTEM (RHRS) 4.8.1 Safety Objective The safety objectives of the Residual Heat Removal System (RHRS) are as follows:
a.
To restore and maintain the coolant inventory in the reactor vessel so that the core is adequately cooled after a loss-of-coolant accident. The Residual Heat Removal System also provides cooling for the pressure suppression pool so that condensation of the steam resulting from the blowdown due to the design basis loss-of-coolant accident is ensured.
b.
The Residual Heat Removal System further extends the redundancy of the Core Standby Cooling Systems by providing for containment cooling. 4.8.2 Power Generation Objective The Residual Heat Removal System provides the means to meet the following power generation objectives:
a.
Remove decay heat and residual heat from the nuclear system so that refueling and nuclear system servicing can be performed.
b.
Supplement the Fuel Pool Cooling and Cleanup System capacity when necessary to provide additional pool cooling capacity. 4.8.3 Safety Design Basis
1.
The RHRS shall act automatically (except when in the shutdown cooling mode), in combination with other Core Standby Cooling Systems, to restore and maintain the coolant inventory in the reactor vessel such that the core is adequately cooled to preclude fuel clad temperatures in excess of 2200F following a design basis loss-of-coolant accident.
2.
The RHRS, in conjunction with other Core Standby Cooling Systems, shall have such diversity and redundancy that only a highly improbable combination of events could result in their failure to provide adequate core cooling.
3.
The source of water for restoration of reactor vessel coolant inventory shall be located within the primary containment in such a manner that a closed cooling water path is established.
BFN-26 4.8-2
4.
To provide a high degree of assurance that the RHRS operates satisfactorily during a loss-of-coolant accident, each active component shall be capable of being tested during operation of the nuclear system. The inboard isolation check valve can only be tested during cold shutdown (MODE 4 or MODE 5).
5.
To provide an additional source of water for post-accident containment flooding a crosstie shall be provided between the RHR Service Water System and RHRS. (This long term capability is not credited in the mitigation of design basis accidents and does not perform an active safety related function.) 4.8.4 Power Generation Design Basis
1.
The RHRS shall be designed with enough capacity that the service water outlet temperature can be limited during shutdown conditions to minimize fouling. 4.8.5 Summary Description The RHRS is designed for five modes of operation to satisfy all the objectives and bases:
1.
Shutdown cooling (Units 1, 2, and 3),
2.
Containment spray and pool cooling,
3.
Low pressure coolant injection, and
4.
Standby cooling.
5.
Supplemental fuel pool cooling. To provide clarity to the information presented herein, each mode of operation is defined as a subsystem of the RHRS and is discussed separately. It is shown how each subsystem contributes toward satisfying all the objectives and bases of the RHRS. The major equipment of the RHRS consists of four heat exchangers and four pumps for each unit. There are twelve RHR service water pumps for the plant (see Section 10.9, "RHR Service Water System"), eight of which can be used for RHRSW purposes. The equipment is connected by associated valves and piping, and the controls and instrumentation are provided for proper system operation. A process diagram of the RHRS is shown in Figures 7.4-6a sheets 1, 2, and 3 of Section 7.4.
BFN-26 4.8-3 A description of the controls and instrumentation is presented in Section 7.4, "Core Standby Cooling Control and Instrumentation." A description of how operation of the equipment in the RHRS in conjunction with other Core Standby Cooling Systems protects the core in case of a loss-of-coolant accident is presented in Chapter 6.0, "Core Standby Cooling Systems." The RHRS pumps are sized on the basis of the flow required during the low pressure coolant injection (LPCI) mode of operation, which is the mode requiring the maximum flow rate. In addition, the system pumps are equipped with discharge flow limiting orifice plates to prevent pump operation in "runout" conditions and to prevent any damage that might occur in the case of a recirculation line break. The heat exchangers are sized on the basis of their required duty for the pressure suppression pool cooling function. It is concluded that the power generation design objective is met. A summary of the design requirements of the RHRS pumps and the heat exchangers is presented in Table 4.8-1. See Section 6.5 for system requirements utilized in the Emergency Core Cooling System analysis. Permanent connections with normally closed valves are provided on the shutdown cooling piping circuit for supplying cooling water to the Fuel Pool Cooling and Cleanup System (see Figures 7.4-6a sheets 1, 2, and 3). This permits the RHRS heat exchangers to be used to assist fuel pool cooling when required (see Section 10.5, "Fuel Pool Cooling and Cleanup System"). One of the RHRS loops, consisting of two heat exchangers, two pumps in parallel, and associated piping, is located in one area of the Reactor Building. The other heat exchangers, pumps, and piping, forming a second loop, are located in another area of the Reactor Building to minimize the possibility of a single physical event causing the loss of the entire system. This arrangement satisfies the safety design basis 2. In addition, the pump suction and heat exchanger discharge lines of one loop in Unit 1 (Loop II) are cross-connected to the pump suction and heat exchanger discharge lines of one loop in Unit 2. Unit 2 and Unit 3 systems are cross-connected in a similar manner. Two normally closed isolation valves are provided in each heat exchanger discharge cross-connection, and four normally closed isolation valves are provided in each suction cross-connection (one at each pump suction), as shown in Figure 4.8-1. RHRS equipment is designed in accordance with Class I seismic criteria (see Appendix C) to resist sufficiently the response motion at the installed location within the supporting building from the Design Basis Earthquake. The system piping and pumps are designed in accordance with the requirements of USAS B31.1.0, 1967 edition, as augmented by GE specifications which were implemented in lieu of the outdated B31 Nuclear Code Cases-N2, N7, N9, and N10. The system is constructed and tested in accordance with TVA construction
BFN-28 4.8-4 specifications. The pumps are also designed and constructed in accordance with the standards of the Hydraulic Institute. The shell side of the heat exchangers is designed in accordance with the ASME Boiler and Pressure Vessel Code, 1965 edition, Section III, Class C vessels, and TEMA Class C; and the tube side is designed in accordance with Section VIII and TEMA Class C. The provisions of the ASME Boiler and Pressure Vessel Code, Section III, Winter Addenda of 1966, paragraph N2113, apply. 4.8.6 Description 4.8.6.1 Shutdown Cooling The shutdown cooling subsystem is an integral part of the RHRS and is placed in operation during a normal shutdown and cooldown. The initial phase of nuclear system cooldown is accomplished by dumping steam from the reactor vessel to the main condenser with the main condenser acting as the heat sink. The RHRS is typically placed in the shutdown cooling mode of operation when reactor vessel pressure has decreased sufficiently to clear the interlocks associated with the shutdown cooling suction valves. The shutdown cooling subsystem alone is capable of completing cooldown to 125F in less than 34 hours and maintaining the nuclear system at 125F so that the reactor can be refueled and serviced. Reactor coolant is pumped by the RHRS pumps from one of the recirculation loops through the RHRS heat exchangers, where cooling takes place by transferring heat to the RHR service water system. Reactor coolant is returned to the reactor vessel via either recirculation loop. During a nuclear system shutdown and cooldown, any one of the four RHR shutdown cooling subsystems can provide the required decay heat removal function and maintain or reduce the reactor coolant temperature as required. The RHRS is normally flushed with water of condensate quality or better in preparation for shutdown cooling operation during the steam dumping phase of plant cooldown. This flush is not required if 1) there is an immediate need for RHR shutdown cooling to control reactor vessel level, temperature, or pressure, or 2) RHR shutdown cooling is removed from and returned to service during an outage and no activities have occurred which could result in water quality degradation below acceptable limits for reactor vessel injection.
BFN-28 4.8-5 4.8.6.2 Containment Cooling The containment cooling subsystem is an integral part of the RHRS and is placed in operation to limit the temperature of the water in the pressure suppression pool so that immediately after the design basis loss-of-coolant accident (guillotine break of a recirculation system suction line) has occurred, the maximum bulk pool temperature does not exceed 179ºF. The maximum permissible bulk pool temperature is limited by the potential for stable and complete condensation of steam discharged from the main steam relief valves as well as the design analyses of the torus attached piping (see Sections 5.2.3.3.2 and 5.2.4.3). With the RHRS in the suppression pool cooling mode of operation, the RHRS pumps are aligned to pump water from the pressure suppression pool through the RHRS heat exchangers where cooling takes place by transferring heat to the RHR service water. For adequate containment cooling, a minimum of two RHR pumps and associated heat exchangers must remain available for several hours after a design basis loss-of-coolant accident. The flow returns to the pressure suppression pool via the flow test line (see Figures 7.4-6a sheets 1, 2, and 3). Pressure suppression pool temperature operational limits are provided in Technical Specification, Section 3.6.2.1. The pressure suppression pool cooling mode of RHRS is initiated to restore pressure suppression pool temperature to within allowable limits during plant operation. IN 87-10 Supplement 1 identifies the potential for the RHRS to be damaged and unable to perform its Low Pressure Coolant Injection function should a LOCA and LOOP occur while RHRS is in the SPC mode of operation. The safety design basis for RHRS requires that only a highly improbable combination of events can result in RHRS being rendered unable to perform its core cooling function (see Section 4.8.3). To meet this requirements, PRA analyses have established a time limit for RHRS operation in the SPC mode and the time RHRS is in SPC mode is tracked to ensure this time limit is not exceeded. The containment spray cooling mode of operation provides additional redundancy to the Core Standby Cooling Systems for post-accident conditions. The water pumped through the RHRS heat exchangers may be diverted to spray headers in the drywell and above the pressure suppression pool. The spray headers in the drywell condense any steam that may exist in the drywell, thereby lowering containment pressure. The spray collects in the bottom of the drywell until the water level rises to the level of the pressure suppression vent lines, where it overflows and drains back to the pressure suppression pool. Approximately 5 percent of this flow may be directed to the pressure suppression chamber spray ring to cool any noncondensable gases collected in the free volume above the pressure suppression pool.
BFN-26 4.8-6 The spray headers of the RHRS cannot be placed in operation unless the core cooling requirements of the low pressure coolant injection subsystem have been satisfied. These requirements may be bypassed by the operator using a keylock switch in the control room (see Section 7.4, "Core Standby Cooling Control and Instrumentation"). 4.8.6.3 Low Pressure Coolant Injection The low pressure coolant injection (LPCI) subsystem is an integral part of the RHRS. It operates to restore and, if necessary, maintain the coolant inventory in the reactor vessel after a loss-of-coolant accident so that the core is sufficiently cooled to preclude fuel clad temperatures in excess of 2200F and subsequent energy release due to a metal-water reaction. A detailed discussion of the requirements and response of the equipment which operates during LPCI for a loss-of-coolant accident may be found in Chapter 6.0, "Core Standby Cooling Systems." A detailed discussion of the requirements and response of the controls and instrumentation of LPCI during a loss-of-coolant accident may be found in Section 7.4, "Core Standby Cooling Control and Instrumentation." In general, LPCI operation involves restoring the water level in the reactor vessel to a sufficient height for adequate cooling after a loss-of-coolant accident. The LPCI subsystem operates in conjunction with the High Pressure Coolant Injection System (HPCIS), the Auto Depressurization System and the Core Spray System to achieve this goal (see Chapter 6.0, "Core Standby Cooling Systems"). This capability satisfies safety design basis 1. The HPCIS is a high-head, low-flow system and pumps water into the reactor vessel when the nuclear system is at high pressure. If the HPCIS fails to maintain the required level of water in the reactor vessel, the automatic depressurization feature of the Nuclear System Pressure Relief System functions to reduce nuclear system pressure so that LPCI operates to inject water into the pressure vessel. LPCI is a low-head, high-flow subsystem and delivers rated flow of 9000 gpm for each pump to the reactor vessel against an indicated pressure of 125 psig. All these operations are carried out automatically. LPCI is designed to reflood the reactor vessel to at least two-thirds core height and to maintain this level. After the core has been flooded to this height, the capacity of one RHR pump is more than sufficient to maintain the level. During LPCI operation, the RHRS pumps take suction from the pressure suppression pool and discharge to the reactor vessel into the core region through both of the recirculation loops. Two pumps discharge to each injection header, assuring flooding of the vessel through at least one loop. Any spillage through a break in the lines within the primary containment returns to the pressure suppression
BFN-26 4.8-7 pool through the pressure suppression vent lines. A bypass line to the pressure suppression pool is provided so that the pumps are not damaged if operating with the discharge valves shut. Added resistance in the pump discharge lines prevents insufficient NPSH in the LPCI mode of operation. It is concluded that safety design basis 3 is satisfied. Service water flow to the RHRS heat exchangers is not required immediately after a loss-of-coolant accident because heat rejection from the containment is not necessary during the time it takes to flood the reactor. Power for the RHRS pumps and the RHR service water pumps comes from the 4-kV AC power shutdown boards. Power for these boards normally comes from the auxiliary supply, but if this source is not available, power is available from the standby (diesel) AC power source. 4.8.6.4 Standby Cooling Standby coolant supply connection and RHR crossties are provided to maintain a long-term reactor core and primary containment cooling capability irrespective of primary containment integrity or operability of the Residual Heat Removal System associated with a given unit. The standby coolant supply connection and RHR crossties provide added long-term redundancy to the other emergency core and containment cooling systems and are designed to accommodate certain situations which, although unlikely to occur, could jeopardize the functioning of these systems. By proper valve alignment (see Figure 4.8-1), the network created by the RHR crossties permits the B (or D) RHR pumps on Unit 1 to circulate Unit 2 pressure suppression pool or reactor vessel water through the B (or D) heat exchangers on Unit 1 in the unlikely event that the Unit 2 RHR pumps are unavailable. The crosstie network is sized for a minimum flow of 5,000 gpm, which will achieve about 91 percent of full flow heat transfer capability of the RHR heat exchangers. In a like fashion, the A (or C) RHR pumps on Unit 2 can be used to circulate Unit 1 pressure suppression pool or reactor vessel water through the A (or C) heat exchangers on Unit 2. The B (or D) RHR pumps on Unit 2 and the A (or C) RHR pumps on Unit 3 can be similarly utilized. Pressure suppression pool water which has been circulated through the RHR heat exchangers on one unit can be used to flood the reactor core, spray the drywell and pressure suppression chamber, or returned to the pressure suppression chamber of the adjacent unit. In this way, decay heat and residual heat can be removed from the reactor core and primary containment of the adjacent unit on a long-term basis. By proper valve alignment (see Figure 4.8-1), the network created by the standby
BFN-26 4.8-8 coolant supply connection and RHR crossties permits the D2 (or D1) RHR service water pump and header to supply raw water directly to the reactor core of Units 1 or 2 as the reactor pressure approaches 50 psig. The service water pump and header can also be valved to supply raw water to the drywell or pressure suppression chamber spray headers or directly to the pressure suppression chamber of either unit. In a similar fashion, the B2 (or B1) RHR service water pump and header can supply raw water to the reactor core of Units 2 or 3 or into the respective drywell/pressure suppression chamber spray headers or directly to the pressure suppression chambers. The Standby Coolant Supply System is sized to supply a minimum raw water flow of 3,250 gpm, against a reactor pressure of 65 psig with a drywell pressure of 15 psig. It is concluded that safety design basis 5 is satisfied. 4.8.6.5 Supplemental Fuel Pool Cooling A description of how the RHRS heat exchangers can be used to assist fuel pool cooling when required is contained in Section 10.5, Fuel Pool Cooling and Cleanup System. 4.8.7 Safety Evaluation Since the LPCI and containment cooling subsystems act with other Core Standby Cooling Systems to satisfy the safety objective, they are properly evaluated in conjunction with the other Core Standby Cooling Systems. This safety evaluation is in Chapter 6.0, "Core Standby Cooling Systems." The safety evaluation of the controls and instrumentation of the LPCI subsystem is in Section 7.4, "Core Standby Cooling Control and Instrumentation." 4.8.8 Inspection and Testing A design flow functional test of the RHRS pumps is performed during normal plant operation by taking suction from the pressure suppression pool and discharging through the test lines back to the pressure suppression pool. The discharge valves to the reactor recirculation loops remain closed during this test and reactor operation is undisturbed. An operational test of these discharge valves is performed by shutting the downstream valve after it has been satisfactorily tested and then operating the upstream valve. The discharge valves to the containment spray headers are checked in a similar manner by operating the upstream and downstream valves individually. All these valves can be actuated from the control room using remote manual switches. Control system design provides automatic return from test to
BFN-28 4.8-9 operating mode if LPCI initiation is required during testing. It is concluded that safety design basis 4 is satisfied. Periodic inspection and maintenance of the RHRS pumps, pump motors, valves and valve motors, and heat exchangers are based on manufacturer's recommendations and sound maintenance practices. A discussion of the availability of engineered safeguards and frequency of testing of equipment is presented in Chapter 6.0, "Core Standby Cooling Systems." 4.8.8.1 RHR Heat Exchanger Performance Monitoring Program Requirements To ensure that the RHR heat exchangers are maintained in a condition that meets or exceeds the minimum performance capability assumed in the containment analyses, which support not taking credit for containment accident pressure in the NPSH analyses, the following program requirements are established. These program requirements are established to satisfy Technical Specification 5.5.1.4, Residual Heat Removal (RHR) Heat Exchanger Performance Monitoring Program.
1. The DBA-LOCA minimum required heat removal rate is 80,136,000 Btu/hour (hr) per heat exchanger with two heat exchangers in service. The EPU fire event minimum required heat removal rate is 124,966,800 Btu/hr with one heat exchanger in service.

2. The thermal performance test acceptance criteria for an RHR heat exchanger is less than or equal to 0.001562 hr-ft2-F/Btu with no more than 77 tubes (4.57% of 1700 tubes) plugged.

3. The nominal (measured) test result (fouling resistance) including the test and measurement uncertainty will be used for comparison to the thermal performance acceptance criteria.

4. The program includes the following requirements:

a. Each RHR heat exchanger is performance tested at a nominal interval of four years, not to exceed five years.

b. The cooling water side of the RHR heat exchangers is inspected and cleaned periodically as determined by the preventative maintenance (PM) program. The maximum interval for performing RHR heat exchanger inspection and cleaning is limited to five years (four years + 25%). Any increase in the inspect and clean interval beyond five years will be evaluated in accordance with PM program procedures, GL 89-13 program implementing procedures and TVA programmatic procedure change
BFN-28 4.8-10 requirements. Inspection results and performance testing results will be used to technically justify extending the inspect and clean interval. The as-found inspection acceptance criteria is less than 77 tubes obstructed (sum of the number of tubes mechanically plugged and the number of tubes obstructed by macrofouling).
5. The following aspects of the RHR Heat Exchanger performance monitoring program meet the guidance provided in EPRI 3002005340, Service Water Heat Exchanger Test Guidelines, May 2015:

a. The Heat Transfer Test Method will be used

b. Temporary surface mounted temperature instrumentation

c. Temporary differential pressure (DP) instrumentation

d. Temporary data acquisition system (DAS) including the associated software

e. Test data analysis - the analysis determines the overall fouling resistance for the heat exchanger and also determines the associated uncertainty in the test result (fouling resistance)

f. The uncertainty analysis methodology

g. Data reduction

6. Temporary instruments are calibrated against standards traceable to the National Institute of Standards and Technology or compared to nationally or internationally recognized consensus standards.

7. Computer programs used in the thermal performance analysis are required to meet the 10 CFR 50 Appendix B, and 10 CFR 21 requirements. PROTO-HX is the computer program used for thermal performance analyses in the RHR Heat Exchanger Performance Monitoring Program.

8. Compensatory measures include entering the condition into the TVA corrective action program, cleaning of the heat exchangers after inspections, determining if more frequent inspections and cleaning are required, and evaluating past operability/functionality when tube plugging and macrofouling acceptance criteria from inspection procedures are not met. The methodology for performing as-found and as-left inspections are provided in TVA Standard Programs and Processes procedures. As-left inspections are procedurally required and ensure that tubes found blocked/obstructed by macrofouling will not be left in the as-found condition.

9. Changes to the program requirements above will be controlled in accordance with 10 CFR 50.59, Changes, tests, and experiments. Change to the program requirements may be made without prior NRC approval provided the changes do not require a change to the Technical Specification requirements and the changes do not require NRC approval pursuant to 10 CFR 50.59.
BFN-28 TABLE 4.8-1 RESIDUAL HEAT REMOVAL SYSTEM EQUIPMENT DESIGN DATA RHRS PUMPS Number Installed per Unit - 4 Design Temperature - 350F Capacity/Pump - 50% (LPCI) Design pressure - 450 psig Shutoff Head - 780 ft Design Conditions/Pump 0 psid* Discharge Flow (gpm) 20,000 (2 in one loop) 10,800 (1 in one loop) Units 1, 2, 3 Rated Pump Capacity 10,000 gpm at 560 ft Total Dynamic Head NPSH Required at 90F (ft) 30 Operating Conditions/Pump Discharge Flow (gpm) 0-12,000 Discharge Head (ft) 780-420 Differential Pressure (psid) 295-0 NPSH Required at 90F (ft) 30-34 RHRS HEAT EXCHANGERS Number Installed per Unit - 4 Shell Side Fluid - Reactor Water or Pressure Suppression Pool Water Tube Side Fluid - RHR Service Water (River Water) Shell and Tube Side Design Pressure - 450 psig and Design Temperature 40-350F Pressure Drop Design Conditions - shell side 10 psi - tube side 6 psi Suppression Pool Cooling Analysis (3952 MWt) - ANS/ANSI 5.1 (with 2 uncertainty) DBA LOCA Analysis: Shell Side Flow (gpm) 6500 Inlet Temperature Shell Side (F) 179.0 Heat Exchanger Duty (Btu/hr) 80,136,000 Tube Side Flow (gpm) 4000 Inlet Temperature Tube Side (F) 95ºF Heat Exchanger K-factor 265 BTU/sec-ºF Limiting NFPA 805 (fire) event Shell Side Flow (gpm) 7500 Inlet Temperature Shell Side (F) 207.7 Heat Exchanger Duty (Btu/hr) 124,966,800 Tube Side Flow (gpm) 4500 Inlet Temperature Tube Side (F) 88.0ºF Heat Exchanger K-factor 290 BTU/sec-ºF
psid - pounds per square inch difference between reactor vessel and drywell.
1 ~o,.cFl>0..'5; LtHTS 1,2 g, 3 BROWNS FERRY NJCLEAR PLANT F !NAL SAFETY Af;ALYS!S REPORT
ES]DU.l.L HEAT REMOVAL SYST

~1 Ul*ilT CPDSS cor~~JECTro*"s AND ST ANOBY COCLANT SUPPLY FIGURE 4. 8-1 AMENDMENT 16 11 I
, I
BFN-26 4.9-1 4.9 REACTOR WATER CLEANUP SYSTEM 4.9.1 Power Generation Objective The Reactor Water Cleanup System maintains high reactor-water purity to limit chemical and corrosive action, thereby limiting fouling and deposition on heat transfer surfaces. The Reactor Water Cleanup System also removes corrosion products to limit impurities available for activation by neutron flux and resultant radiation from deposition of corrosion products. The system also provides a means for removal of reactor water. 4.9.2 Power Generation Design Basis
1.
Provision shall be made for the continuous mechanical and chemical filtration and demineralization of reactor water to quality specifications.
2.
Provision shall be made for discharge of reactor water at reduced activity during startup and shutdown.
3.
Provisions shall be made to limit the heat loss and the fluid loss from the nuclear system. 4.9.3 Description (Figures 4.9-1, 4.9-2, 4.9-3, 4.9-5, 4.9-6, 4.9-7, 4.9-8, 4.9-9, and 4.9-10) The Reactor Water Cleanup System provides continuous purification of a portion of the recirculation flow. The processed fluid is returned to the reactor vessel, to radwaste, or to the main condenser. Regenerative heat exchangers are provided to limit heat loss from the nuclear system. The system can be placed in service at any time during normal reactor operation or shutdown conditions. The major equipment of the Reactor Water Cleanup System is located in the Reactor Building and consists of two pumps, regenerative and nonregenerative heat exchangers, and two filter/demineralizers with supporting equipment. The entire system is connected by associated valves and piping, and controls and instrumentation are provided for proper system operation. Design and construction of pressure-retaining piping and components of the Reactor Water Cleanup System was initially in accordance with the requirements of USAS B31.1.0, 1967 Edition, as supplemented by the requirements of the applicable GE specifications, which were implemented in lieu of the outdated B31 Nuclear Code Cases-N2, N7, N9, and N10. Design data for the major pieces of equipment are presented in Table 4.9-1.
BFN-26 4.9-2 Reactor coolant is continuously removed from the Reactor Coolant Recirculation System, cooled in the regenerative and nonregenerative heat exchangers, filtered and demineralized, and returned to the feedwater system through the shell side of the regenerative heat exchanger. The RWCU System has one return line through reactor feedwater line B. Because the ion exchange resins used in the filter-demineralizer are temperature-limited (Table 4.9-1), the reactor coolant must be cooled prior to processing in the units. The regenerative heat exchanger transfers heat from the influent water to the effluent water. The effluent returns to the feedwater system. The nonregenerative heat exchanger cools the influent water further by transferring heat to the Reactor Building Closed Cooling Water System. During startup and shutdown, excess water in the primary system is sent to the main condenser or to radwaste by diverting part or all of the filter-demineralizer effluent. This reduces the effectiveness of the regenerative heat exchanger. The nonregenerative heat exchanger has the capability of reducing the filter-demineralizer influent temperature to the required level, while maintaining an adequate diversion flow rate. The filter-demineralizer units (Figures 4.9-2, 4.9-3, 4.9-6, 4.9-7, 4.9-9, and 4.9-10) are pressure precoat-type filters which use either finely ground mixed ion exchange resins or a mixture of powdered resins and fibrous material as a precoat medium, they serve as a combination filter-ion exchange medium. Spent resins are not regenerable, but are sluiced from a filter-demineralizer unit to a backwash receiver tank, (from which they are pumped to the cleanup phase separators for dewatering, decay, and disposal). A strainer is installed on the outlet of each filter-demineralizer unit to prevent resins from entering the Reactor Coolant Recirculation System in the event of a resin support failure. Each strainer is provided with an alarm which is energized by high differential pressure (20 psi). A bypass line is provided around the filter-demineralizer units for bypassing the units when necessary. Each unit has a holding pump which starts in the event of low flow, in order to hold the resin in place on the support elements. Relief valves and instrumentation are provided to protect the equipment against overpressurization and the resins against overheating. The system is automatically isolated when signaled by any of the following occurrences.
a.
High temperature downstream of the nonregenerative heat exchanger. To protect the ion exchange resin from damage due to high temperature.
b.
Reactor Vessel Water Level - Low, Level 3. To protect the core in case of a possible break in the Reactor Water Cleanup System piping and equipment (see Subsection 7.3, "Primary Containment and Reactor Vessel Isolation Control System").
BFN-26 4.9-3
c.
Standby Liquid Control System actuation. To prevent removal of the boron by the ion exchange resin.
d.
High temperature indicative of a RWCU pipe break/critical crack in any of the following areas: main steam valve vault, RWCU pipe trench, RWCU pump rooms, or the RWCU heat exchanger room to isolate the system (see subsection 7.3). Sample points are provided upstream and downstream of each filter-demineralizer unit. The sample analysis provides an indication of the effectiveness of the filter-demineralizer units. The influent sample point is also used as the normal source of reactor coolant samples for analysis of coolant system activity required by Technical Specifications, Section 3.4.6 and for coolant chemistry requirements specified in Section 3.4.1 of the Technical Requirements Manual. Reactor Coolant System activity analysis includes a determination of dose equivalent I-131 concentration which includes quantitative measurements for I-131, I-132, I-133, I-134, and I-135. Operation of the Reactor Water Cleanup System is controlled from the Main Control Room. Resin-changing operations, which include backwashing and precoating, are controlled from a local control panel in the Reactor Building. 4.9.4 Inspection and Testing The Reactor Water Cleanup System is normally in service. Satisfactory operation is demonstrated without the need for special testing. Periodic inspection and maintenance are carried out based on manufacturer's recommendations and sound maintenance practices.
BFN-23 TABLE 4.9-1 REACTOR WATER CLEANUP SYSTEM EQUIPMENT DESIGN DATA MAIN CLEANUP RECIRCULATION PUMPS Number Required:.............................. 2 Design Temperature (°F):................... 150 Capacity (each):............................. 50% Design Pressure (psig):..................... 1300 Discharge Flow (gpm/pump):.......... 180 Discharge Head at Rated Flow (ft):..... 500 HEAT EXCHANGERS Regenerative Nonregenerative Reactor Coolant Flow Rate (lb/hr) 133,000/187,530* 133,000/187,530* Shell Side Pressure (psig) 1,450 150 Shell Side Temperature (°F) 575 370 Tube Side Pressure (psig) 1,450 1,450 Tube Side Temperature (°F) 575 575 FILTER-DEMINERALIZERS Number Required:.................................... 2 Design Temperature (°F):.......... 150 Capacity (each):.................................. 50% Design Pressure (psig):........... 1450 Flow Rate/Unit (lb/hr):........ 66,650/93,765* Effluent Conductivity (mmho max):........ 0.1 Effluent pH:.................................. 6.5 to 7.5 Effluent Insolubles (ppb-measured as residue on 0.45 micron filter paper):.................................................. <10
Lower mass flow rate corresponds to the maximum flow of 270 gpm. Higher mass flow rate corresponds to the maximum flow of 340 gpm. Operation of RWCU above 270 gpm depends upon the ability of RBCCW to accommodate the added heat load.
~ ~"'ill.. ' ~... @I II ....::----------------"====-""'~-~ m:-"- 't!l'm:**.. "'"" ::.,:i*~~:rb:ts.:::::*~~~~~~a,,.. f;f"'\\."',"' TO>Of'f,l'"ll'C"'"'M'l ""'l ft"" 6 5 3 ~ f-"-- -----' Hal o-<11110-,""'"""' AMENDME NT 27 P011fERHOUSE UNIT 2 BROWNS FERRY NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT REACTOR WATER CLEANUP SYSTEM FLOW DI AGRAM FIGURE 4.9 - 1 H G D C B A
ozoti L-L£83Lt-Z V'l l9 I DEMINERALIZER TANK ROOM 2A 1 114* I
1-114* I
.1-114* I N (:'. EL 642.0 3" PLUG I i_.--------2_-_11_2_*_* _T_oP_v_E_N_T __ REACTOR CLEANUP DEMINERALIZER TANK 2A [I] rr 1/2" 1/2" r z I DEMINERALIZER TANK ROOM 2B 1 1 /4* I 1-1/4"1 1 114* I N (:'. EL 642.0 3" PLUG i_ REACTOR CLEANUP DEMINERALIZER TANK 2B II] I 1 " 2-1/2" TOP VENT 1/2" 728 1/2" 314" 1THIS DWG,D5 ), PI 52 ~--------------------+-------t-----------------+--------- ~ FCV',--+---IFCV~ 3 3* FCV 25 FCV 26 ~ z w ~ ~ 50 5~ t-z VALVE ROOM ~ I . I 2" 1 i = l"l £',I I I I I Z N N N r-2-1/2" BACKWASH FROM DEMINERALIZER ~ ~ AIR INLET BACKWASH AIR SYSTEM VALVE 2-53-520 I I I I 2" LL.I 2-1/2" L-~11-f--*I----HIH------------1---------- ----------------------11------KJ-.L.--------I---------~ ----1-....;.---,f--L--,'l/'I--Y I I I I FE I -.I 53 1-.1...--'-+,o_,1--'---,.-----< o-+1E846-1, rn oo---1-----ru ~ 2-1 /2" 706 172* 1/2" 1" 1" 1/2" 1/2" 1" 1.395" ~ ....~ 1 " 201A - PI FCV 32 40 [B---{)-~ PDT 3l 202A 1
r DEMIN 2A
~ RESIN TRAP []-i__[~/ + 1" ~ ;~ 203A ~ r, ~ 3/4" 1/2" 658 659 204A © I - 2o8A 7 -f- -f- ~ 1, Fcv -I Fcv, FcvtB--- FCV - -n ~ 74 - f-58 FCV PDT __ !CV _ 1-46 . -\\ 48) 20 __ w ~ ZlOA 1" ~ 45 -- FCV -e-FCV 2a ID s+ 1 1 m ~ N ~ FCV 19 2 !:: ~ rDElv1IN 28 2 47 42 _ 1-5 G:i ~ ((]-----,i.. RESIN TRAP
F4C4V,
--+**--+ *./,-- - : ~ /- - - r 3 ~ FCV .: ~ \\jjJ IT1-£ ~/ ~ --~ ~ .Y,2.. 22, ~: ~"~1" - J3¥*1--2-1'*+-'-~-2-+* 5 -;)m I I I I I 212!, l lZ" f2" I I .+-.----, 3/4" GJ rn-1-aJ ,---~<XJ-...!.L4 f 686 687 L---l---'~-"--*l-.....:=--1-1---------*-1-----r1"------;~<<-T r 4 TANK VENT / 'FI I 49 FE 111' 35 T'L--~'.. 1--'-+'+-
-----1--~3~~4~"-~
FCV 21 FCV 23 FE, fi ~ o 60 ~ L---f'*+-'--i<'+- '----'--*l--'3~/~4-"---1 ~ 205A r RHR SYSTEM FROM RHR SYSTEM VALVE 2-74-702 --(2-47E811-1,H3 FROM DEMINERIALIZER WATER SYSTEM VALVE 2-2-1367 --< 2-47E856-2, E7 PI 69 lJI- ~ N 1 " 740 0 (:'. N N N N lffi ~ ~. 4" "'" L_,:4_".L--r-r--£:,',:J-----L-.,...JI.(/ FE \\ +" - I - I 2""'3 70 ~ 2.580" 3/4" 1 IEISTIIIE TRAP DRW DRAIN r---1i2-47E852-1,G1 FCV 77 I 1-1/2" .4" ' fl"- ,-l<J,-1--f' ,-..i,--,~ 715 112" cl:J 166d ~ 1661 1 4 I IT FCV w (:'. ~ N - l/r cl::J 16881 ~689 1 4 r"c§ ~ ~ /; TANK VENT 213A 4gp ~* ~T~, FCV ~ N 112* r @] l.,\\SOA 2-1/2"
11.
2" FI )------'f--_:::a.::._ _ _;:_.c.:..;; ____ "7""1----l--=,-h--=r:r---r-...l...(:*-rl" f "I-+--'----------------, _ 1-61 --r-66 FCV = TO SAMPLING AND 3/4" 67 35 -~ -~ z ~ r 8

+

714 TO SAMPLING AND 250,, 3/4" 3,,4* WATER QUALITY rc***.'.;;iii'.-j;x:,:J--"~' SYSTEM J" 2-47E610-43-1,H2 1 /2" -- 1 " 3" 24 ~ f ~:f -f-FCV FCV __ @) WATER QUALITY tL2::;5;;;0;,."-tXX:f----l SYSTEM J = ~ 2-4 7 E 61 0-43-1, H 1 6 85 I S'f-.l3:L:! "4-," :;,;--;-;,: '_;,1,..i.;;,-;,--3-£ / OS ~ ----I~ ~ ~ RELIEF SET ,,If '°---t--t-i--- ---~ EL 627'-9" 739 657 EL 625.25' @ 150 PSIG __. '--+---1-=-"'3.ir" =--+----------------1-------+--------+----+----t--------+-----+--------------+----------------l--tl------tl---+- rn HIGH LEVEL ALARM EL 627'-6"(\\ TANK OVERFLOW 0---j MI'XD.6 ~ I 3* DRW DRAIN I 8 2-47E852-1,F4 r 3* 4* I I ~ I 3" 11------===---------:::::-1----:-::--:-;::==========t-::::::;;:;:i~======- ~,~s~ ,2 1/2" PI 1/2" ORW DRAIN (pj\\ 1011 1°'~ t 1020 HOLDING ~ 1 /2" PI 102! RELIEF SET@ 3/4" ~ 41 i--', 2-47E852-1,F4 \\.._Jj) 1 1 = ~ _ 72 5 L--tot--. ,,--, 314.. 1450 PSIG _::__l== [+/- 0 1/ 2"Nlf I') 1/2"_.A 692 N.~ r.2 § ~;~ 112" 3{+"112"_ - 66+ N ~ 1J --g.i L1~/~2!.*~*::1',{,:.'-c'>1'1:.:1:.;2:.*_*1co;.1<1l-..!ii-l"--~-, ~ ~I 3* n 1 S?, L LJ - /.. 1., 'I' RELIEF SET @ ~ 1/ 2" l r1 ~ ~ 'I' 314,~ !z1 1-.1 .=1 ~... 1_112.. fo15 1450 PSIG 3" r I 1.. N r1" ILt '/4"-, j &,"1-1/2 I T"" 1-1/2" 3/4" 1-1/2" I T"" 1-1/2" 1-1/2" '559* r---l~tJ~~~ ')'-t~::'~j~t.:-::~-""-'CO...-t>f-~~-!<l-----'.:..::--*-ll'ra'~tt"""'i 's9/ 3/4 *
557~

0 731 ~32 695 129 730
~ 167~2 676 ~/t°~~J\\G16ol!llfolA -~ ~704 2 703 2 HOLDING PIii' 2B --~ ~ _12_~ 2 4 io. - 10GPM@100' TDH 1 " l CRW DRAIN 2-47E852-2,G3 I - I 673 3/4" - r-- 3/4" ...:....-==--'--==---------------------------* 1-1 1 " 1 " ' -, I 1 " I 1-1 13 4 1672 1" 1* '--------~-----!/ Fif---r----- 1 " 1-1 1011 70"1 FI 1-----------'--.J rn--+---rn I CRW DRAIN L-------t*~"l)2-47E852-2,G3 I ----,_ 3 4 -\\ I . NI---J 4"
n.
BATCH LEVEL ~~~li°"T 1---L..I f, 1'/'rr\\_ 81I
EL 627'-3" r-~
~.250" ~l\\ 1-rn (:'. 1 * *--=, I I 7 , LOW LEVEL ALARM I [+/-I / EL 62+'-5" N 7191" ~ ,o, 1/4" ,,..L-to8;i3,:i1..i...,.,o,r-.'-,J2 ""'\\_,--~.""'"---r=,...--~ .250" PRECOAT 1 l3l----+---f4l TANK OR, 1" 2" d. I .+/-. --1---i-----'----'1-".i.., /-.. **--;_1--....;:*-*4--',' CHEMICAL DRAIN l __L 12-47E852-2.G3 718 ~ 717 FROM CONTROL 3 AIR SYSTEM ~-47E2847-6,B51 FCV 39 I I I ' 674 3" ~ TO CHEM~ ~: I-=~ WASTE TANK ~, ~ ~~~~f8~\\~~E. ~3 ~ 3 1 " 1 " 3" 4" ~ ~~ rn--+---rn RELIEF SET @ 1 1 150 PSIG 1702 1 ~ 3/4"~~ 3/4" I~ 1.. _ ~ 736 ' J3i\\ -738 ~ PRECOAT OUTLET & BACKWASH INLET 3" 4" M n.rf'2:..-.;.1 c;/2:.."!<)-4,-'_._.--{ -PRECOAT PUMP ~3~"-~-J-,-,~r~**~*_1;___-~>-" __ ..,,.,.,,,,,,..,.,-----------:l--1/-2'_'~---\\__-(-NO-T-E-9)-....;._I --.J ~ ~ I I ~ -314 GPM@ 70' TD 3 ~ ,, 6?7 3/4" 4" I ~ (NOTE 9) I 4" I tJ )1111--.::.t...:...., *-;--;------+ '-,-:----------------,,-------1---------;---------I':! I I I I 3/4" 1",:-;_1/2* 314" 314* ~ g 4" PRECOAT PUMP DISCHAR~E,r,,:::-:-;-:--:-~,~3-"*'f--~3~"~----1-f""-st ~ - - ------:------+,,t L I-723 I N 721 (.!J N ~ ~ 4" REGENERATIVE HX TO RWCU SYSTEM <2 47E810 1,C4 n 3" RELIEF SET@ 150 PSIG I ~ 733 7341 735
== I 3~" ~ EL 621.25' ~---~~==~--------------------------------+---------1--------+-------------------------------+--------~~ 3/4" 3/4" r < ~ ~ z ~ " ~ ~ ~ ~
l. ~
~ - r .:a w z tJ w ~ ~ 1" CAPPED \\ OECONTN CONN j 725 r7I 726 o I* ~ - PI 83 1 /2" CRW 4" FROM RELIEF VALVE 2-75-629 DRAIN . \\ OF PSC HEAD TANK ' \\ 2" ~----l---+--+--lf--l----------------------------------------------------------------------------------------t--lf--l--1~2~-~4~7~E=,852-2,G3..... ,.... 1--=---< 2-47E814-1, E2 4' FFI IIFNT CLEANUP SLUDGE Plll'S ROOM 3" -13 1/2" ~i*--------------------------------4-"~*-------'BM:K===='ASH:=.:..::OUT;;:..:;L=E~T~TO:;..:n::...~IIE=CE::.:I~VI=IC:=..~T~Ail(=-------*--4-" _______________________ 1__:_ SYSTEM PRESS.- TEMP DATA ,_ +" INFLUENT NON-REGENERATIVE HX I FROM RWCU SYSTEM 3
4"
~ 3" 3" 2-47E810-1,D3 '>---+,,..._.,..._..._ ___ -f'o.....;f-.&.t,f,----------------------------------------------,--=.,,,.,,.,,,,,,,..,,===----------------------------""' I +" - TO CLEANUP BACKWASH _ -~ 742 RECEIVING TANK (VE2NT1)1,2,, VENT LINE N LINE DESIGN MK NO. PRESSURE (PSIG) 1 2 3 1450 1300 150 DESIGN TEMP( °F) 150 150 150 71 DEMINERALIZER DATA CLEANUP OEMINERALIZER RATE (PER UNIT) DESIGN 340 GPM (MAX).,,..... 1.08 GPM PER SQ FT QUANTITY OF BACKWASH AIR.... 236 SCFM AT 25 PSIG BACKWASH WATER.........,.... NORMAL-157 GPM AT 25 PSIG .............. MAXIMUM-236 GPM AT 25 PSIG PRESSURE.................... DESIGN-1300 PSIG .................... OPERATING-1155 PSIG TEMPERATURE................. DESIGN-150"F ................. OPERATING-130"F
NOTES,

1. FOR DETAILED OPERATING INSTRUCTION, BACKWASH RATES, ETC, SEE MANUFACTURER'S INSTRUCTION MANUAL.

2. ALL VALVES ARE SAME SIZE AS PIPE UNLESS OTHERWISE NOTED.

3. HEAVY LINES SHOW WATER FLOW THROUGH SYSTEM DURING NORMAL DEMINERALIZATION.

4. OPERATIONAL VALVES ARE SHOWN IN THEIR NORMAL OPERATING POSITION.

5. ALL VALVE TAGS SHALL BE PRECEEDED BY THE NUMBER 2-69-, UNLESS OTHERWISE NOTED.

6. [I]ECT, DENOTES DESIGN PRESSURE AND TEMPERATURE AS GIVEN IN TABLE THIS DRAWING.

7. HYDROSTATIC TESTING SHALL BE IN ACCORDANCE WITH THE APPLICABLE COOES.

8. THE DESIGN PRESSURE AND TEMPERATURE OF ALL DRAIN AND VENT LINES THROUGH THE LAST ISOLATION VALVE SHALL BE THE SAME AS THE PROCESS PIPE.

9. THE AIR RELEASE VALVES 69-734 & 737 HAVE A 1* SUPPLY CONNECTION AND 1 /2" VENT CONNECTION.

10. VENT, DRAIN, AND TEST CONNECTIONS 1-1/2* AND BELOW CAN BE PROVIDED WITH PIPE CAPS OR HOSE CONNECTION FITTINGS WHERE REQUIRED BY PLANT PERSONNEL. THIS CONFIGURATION IS SUPPORTED BY ENGINEERING CALCULATION CD-Q0999-923399.
REFERENCE DRAWINGS: MEL..................... VALVE MARKER TAG TABULATION 0-478601-69 SERIES...... INSTRUMENT TABULATION-REACTOR WATER CLEANUP SYSTEM 2-47E610-43-1........... MECHANICAL CONTROL DIAGRAM-SAMPLING & WATER QUALITY SYSTEM 2-47E610-69-1........... MECHANICAL CONTROL DIAGRAM-REACTOR WATER CLEANUP SYSTEM 47EB00-1................ MECHANICAL FLOW DIAGRAM-GENERAL PLANT SYSTEMS 0-47E800-2.............. MECHANICAL SYMBOLS & FLOW DIAGRAM DRAWING INDEX 2-47E810-1.............. MECHANICAL FLOW DIAGRAM-REACTOR WATER CLEANUP SYSTEM 2-47E811-1.............. MECHANICAL FLOW DIAGRAM-RESIDUAL HEAT REMOVAL SYSTEM 2-47E814-1.............. FLOW DIAGRAM-CORE SPRAY SYSTEM 0-47E830-5.............. MECHANICAL FLOW DIAGRAM-RADWASTE 0-47E846-1.............. MECHANICAL FLOW DIAGRAM-DEMINERALIZER BACKWASH AIR 2-47E852-1 & -2......... MECHANICAL FLOW DIAGRAM-CLEAN RADWASTE & DECONTAMINATED DRAINAGE 2-47E856-2.............. MECHANICAL FLOW DIAGRAM-DEMINERALIZED WATER T-12728................. MECHANICAL FLOW DIAGRAM-REACTOR WATER CLEANUP FILTER DEMINERALIZER SYSTEM (GRAVER,.L 2-47E2847-6......,,... --~~~H~~S'f'E~,I/C FLOW DIAGRAM CONTROL AMENDMENT 28 I u, <0-47E830-5,G4 3' OECONTN CONN~ .-~ TO CLEANUP BACKWASH RECEIVING TANK {OUTLET) 5 6 7 0 120 100 40 150 103 100 120 POWERHOUSE UNIT 2 8 ~TO SAMPLING AND WATER QUALITY SYSTEM 2-47E610-43-1,H3 I 7 I -,o-47E830-5, G4 6 I 5 EL 593.0 I 3 BROWNS FERRY FINAL SAFETY NUCLEAR PLANT ANALYSIS REPORT REACTOR WATER CLEANUP DEMINERALIZER FLOW DIAGRAM FIGURE 4.9-2 H G F E D C B A
!,£0~ ~-69-0 ~93Lt-Z V'I L9 FULL OPEN FI 69-24 RCIC SYS .---""2-47E610-71-1,D8 2-47E817-1.D6 THERMAL SLEEVE I 2-45E779-47 TO SHUTDOWN COOLING SUPPLY 2-47E2610-74-1,C1 FROM REACTOR VESSEL BOTTOM HEAD DRAIN 2-47E817-1,C5,,__....,,.., TO FILTER ,----11~ DEMINERALIZER UNIT 2A TO FILTER DEMINERALIZER UNIT 28 2-47E2847-6,F3 HCV 69-22 rev 2-47E2847-6,F3 HCV 9-23 2-47E2847-6,E3 HCV 69-18 2-47E2847-6.E3 25-3 ~ 25-3 69-501~ 2-47E2847-6.88 69-27 -, 2 5-3 {PcV\\ 69-288 25-3 ~ ~: Y-3832-J-2 I FILTER DEMINERALIZER UNIT 2A HCV 69-25 2-47E2847-6,F3 HCV 69-26 25-3 HS FCV 69-26 69-:_ ~~ 25-3 25-3 (,?C~ 69-35A 69-35 ~ 25-3 u u u u u I,-------------------1 mmmmm ll'J U) a:, C7) ~ ~ ~ ~ ~ 9-3 ~ ~ * * * * * ~ 'd!W --------------T-i2-45E620-2 I I PI 9-4 69-78 ULl 25-5-~ 69-13A [J:QQ]_ I.../FI'\\25-2 ~T 69-138 9-19 9-1s I X _..J ~ ~Ti 69 5228 '1'.------' ~9-4 I ~ I ,--~ I 6 9-19.J.....25-2 ~ (FT) 69-13 2-47E2847-6,F4 SEE DET A 2-47E610-43-3 ~~~~~ I ! {TE\\ {TE\\ {TE\\ {TE\\ {TE\\ {TE\\ {TE\\ {TE\\ I ~ 1 ~ ~ ~ 6YA 6YB 6YC SYD 6YE 6YF 6YG 6YH : ~ 6~A 6rJiJ: 5 6~C ~ ~ 6~8 6~~ 5 6~D ~ l ___ L __ j_ __ J __ -r __ L ___ L ___ L ___ L_J--,2-45N2616-51 AS-83 I ,.l...9-85 ,.l...9-84 y A-86 I L...LI.!n..._ I -...lI.!il.-J L...LI.!n.._ I _...lI.!il..J I CLEAN UP SYS. AREA TEMP. 69-834A -rt.= 69-834C 69-8348 ~ 69-834D r - - - - -; ~ T ~ g-83 T~-85 T:: g-.. 44_ 6ill-86(rn ~-22 I ~ 69-839A I 69-839C '1 w 69~98 '1 69~9D 69~D I ~ I g: Y I Y ~~;';c g: 1 r42-73DE927-7.81 I L_ ! 6 du
du 11 6 du 11 I FRDM RHR SYSTE" 11 I q) 69~9A I 69~9C W 69-8398 I
NOTES: ~ I I ~ -- I I l2-47 Ez 51 o 1,Kt>--; 1. THE REACTOR WATER CLEANUP SYSTEM PROVIDES FOR CONTINUOUS I HI "f" I 't I I FROM RCIC SYSTEM I MECHANICAL AND CHEMICAL FILTRATION AND DEMINERALIZATION OF A 2 45E671 25 N ~ L I N ~ 1 PORTION OF THE REACTOR WATER TO QUALITY SPECIFICATION. DIV IIA _ _. ~I I l2-47E610-71-1,HB>-i IMPURITIES ARE REMOVED BY FILTRATION, ABSORPTION, AND ION L I I EXCHANGE. I 12 45E671 31 H I 2 45E671 37 U 12 45E671 43 L1 I lfROM HPCI SYSTEM~ 2. WATER FLOWS fRCM THE REACTOR VESSEL TO THE REGENERATIVE HEAT L I r r 12-47E610-73-1.H8.,-----y EXCHANGER, THE NONREGENERATIVE THROUGH THE RWCU RECIRCULATING L

, I L-------l------:..J...:------J......

FRDM MAIN STEAM SYS I PU.PS TO THE DE*INERALIZERS. THE WATER THEN FLDWS TO THE I REGENERATIVE HEAT EXCHANGER BACK TO THE REACTOR VESSEL. IJZ-..L-r---L--------, '--------INSTRUMENT-LOOP T-69-834----1_.1 2-47 E6l0-l-1.D2 >-' 3. BACKWASH AND PRECOAT OPERATIONS ARE MANUALLY INITIATED TO RWCU ISOLATION AND AUTOMATICALLY CONTINUED. DURATION Of EACH OPERATION WILL REGEN HX 2A CLEAN UP AREA VLV fCV-69_1, _2, _12 BE ADJUSTED IN THE FIELD BASED UPON SYSTEM REQUIREMENTS. R8ccw 4 FCV-69-1, -2 & -12 ARE ISOLATION VALVES THAT MAY BE REGEN HX 2B REGEN HX 2C HIGH FLOOR DRAIN TEMP. OPERATED MANUALLY FROM THE MAIN CONTROL ROOM. THE VALVES .... ---~ 2-47E610-70-1,G2 ARE AUTOMATICALLY CLOSED BY THE FOLLOWING SIGNALS: NON-REGEN HX 2A L------1 NON-REGEN HX 2B L _ _/TE'\\ I 69-68 I PI 25-2 I 69-1D TW"._........ I.._... 6~e I INSTRUMENT LOOP T-69-834 (LOOPS T-69-835,836,837,838 AND 839 SIMILAR, SEE TABLE BELOW) LOOP RTD THEMPERATURE INDICATING SWITCHES T-69-834 TE-69-8J4A,8,C,D TIS-69-BJ4A,8,C,D T-69-835 TE-69-SJSA,8,C,D TJS-69-SJSA,8,C,D T-69-836 TE-69-8J6A,8,C,D TI5-69-8J6A,B.C,D T-69-837 TE-69-8J7A,B,C,D TJS-69-837A.B,C,D T-69-838 TE-69-SJBA,8,C,D TIS-69-SJBA,8,C,D (A) ACTUATION OF STANDBY LIQUID CONTROL SYSTEM.
18)
HIGH TEMPERATURE UPSTREAM Of THE FILTER-DEMINERALIZER. C) HIGH TEMPERATURE IN AREAS CX!CUPIED BY RWCU EQUIPMENT (RWCU HEAT EXCHANGER ROOM OR RWCU PUMP ROOMS 2A & 28), OR HIGH TEMPERATURE IN THE RWCU PIPE TRENCH OR HIGH TEMPERATURE FROM RWCU PIPING IN THE MAIN STEAM VALVE VAULT. ID) REACTOR WATER LEVEL LOW.
5.
THE RECIRCULATION PUMPS ARE PROTECTED FROM OVERHEATING BY TEMPERATURE SWITCHES IN THE COOLING WATER DISCHARGE (TIS-69-4A AND -48) WHICH TRIP THE PUMP AND ALARM ON HIGH COOLING WATER TEMPERATURE. FLOW SWITCHES AT EACH PUMP DISCHARGE (FIS-69-4A AND -48) MONITOR FLOW TO TRIP THE PUMPS UPON LOW FLOW AND ALARM (FA-69-4).
6.
TEMPERATURE ACROSS THE HEAT EXCHANGERS IS MONITORED BY Tl-69-6. T-69-839 TE-69-8J9A.B.C.D TIS-69-8J9A.B.C.D 7. PRESSURE DROP ACROSS THE FILTER DEMINERALIZERS IS L ____ --' RBCCW -47E610-70-1.F2 MONITORED BY (PDT-69-33 AND -58) WHICH ACTUATES AN ALARM (POA-69-33 & -58) AND CLOSES THE DISCHARGE VALVE (fCV-69-35A & -GOA) WHEN THE DIFFERENTIAL PRESSURE EXCEEDS 25 PSIG. A MANUAL SWITCH (HS-69-J5A & -GOA) IS ALSO PROVIDED FOR MANUAL ISOLATION OF THE FILTER DEMINERALIZER VESSEL. FRCM FUPG-32-5105 2-47E2S47-6, ca rev 9-94 t .._...~I >Sc-1 """----.. 9-4 GiS <<-IBA ~5-2 SAh.f'LING SYSTEM

t....-'1---... -----------l(J,l-.... -:-')2-47E61D-43-1.H3 I

2-47E2847-6,F2 83
8.

9.

10.
11 *
12.
LOW FLOW FROM A A 8 DEMINERALIZER (FE-69-35 & -60) WILL START THE HOLDING PUMP BY ACTUATING (FS-69-358 & -608) OPEN VALVE FCV-69-358 & -608 CLOSE (FCV-69-35A & -GOA) AND SOUND ALARM F A-69-JS & -60. ONE CLEANUP PRECOAT SYSTEM SERVES BOTH DEMINERALIZER UNITS. A LEVEL SWITCH (LS-69-82) PROVIDES HIGH AND LOW ALARM AND LOW LEVEL PUMP INTERLOCK PROTECTION. TIS-69-11 WILL ISOLATE THE SYSTEM ON HIGH DISCHARGE TEMPERATURE. THE COMPARTMENT FLOOR DRAIN LEAK DETECTION ELEMENTS (TS-69-30A THROUGH H) CAN BE SET TO ISOLATE THE SYSTEM AT APPROXIMATELY 20" F ABOVE AMBIENT TEMPERATURE. FOR PANEL LOCATIONS OF PILOT LIGHTS FOR CONTROL VALVES, SEE THE PANEL NUMBER FOR THE ASSOCIATED HAND SWITCH. INSTRLIJENTATION ASSOCIATED WITH BACKUP CONTROLS IS LOCATED AS SHOWN ON THE DIAGRAM. IF A LOCATION IS NOT SHOWN. THE DEVICES WILL BE LOCATED ON THE APPROPRIATE 4160-V AC BOARD. 480-V AC BOARD. OR 250-V DC BOARD. (REFERENCE GE DESIGN SPEC 22A1470.)
13. TE-69-98A-1 PUMP 2A INBOARD BEARING T/C, TE-69-98A-2 PUMP 2A OUTBOARD BEARING T/C, TE-69-99A PUMP 2A CASING T/C.
14, TE-69-988-1 PUMP 28 INBOARD BEARING T/C, TE-69-988-2 PUMP 28 OUTBOARD BEARING T/C, TE-69-998 PUMP 28 CASING T/C.
15. CONTROL AIR TO INSTRUMENT SUPPLIED BY CONTROL AIR HEADER IN CABINET 25-3.

16. CONTROL AIR TO INSTRUMENTS SUPPLIED BY CONTROL AIR HEADER ON PANEL 25-JSA & B.

17. ALL COMPONENTS ARE PREFIXED WITH UNIT 2 DESIGNATOR UNLESS OTHERWISE NOTED.
FI 69-49 HCV 69-42 HCV 69-48 HCV 69-51 25-187 (Pr\\
18. CCMPLETE LOGIC FOR REACTOR WATER CLEANUP SYSTEM IS SHOWN ON MECHANICAL LOGIC DIAGRAM LISTED IN REFERENCE DRAWINGS.
PREG-69-35A 2-47E2847-6,F3 2-47E2847-6.B8 2-47E2847-6.F4 HCV 69-46 2-47E2847-6, fS ~ 69-63 rev ta--::-=-+ 69-63 TT 11 25-3 'd 11 FSV HS ~I 69-6.3 69-63 );;;{_ j HS FCV 69-48 69-:_ ~): FILTER OEMINERALIZER UNIT 2B 2-47E2847-6, F5 25-3 25-368 PI 69-56 25-368 PI 69-57 ~25-3 ~25-3 ~25-3 f PX \\_ ( fS ~~ FA) 6~0 '25-368 69-60A 6~0 1----E/~ 25 3 9-4 69-608 FRC 69-60 69-52 2-47[2847-6 . 86 2-47E2847-6, 25-368 CG FM ~ 25-3 2-47E2847-6, 69-60A FS 69-61 FCV 25-3 /H5' 69-77 69-77 y PREG-69-SOA HS 69-65 ~~~~T.-5 -----1 2-47E2847-6 .83 25-3 ~ FSV 69-66 69-66 25-3 TO BACKWASH w ' ~ N w ~ v ' N RECEIVING TANK 2-47E837-1.A5s,------' SAM PL I NG SYS 2-47E610-4J-1,H1 (Pr\\ 69-69 {Fl\\ 69-70 PCV ,_ _ _,69-68 FE.)--L...---[:o<)--,.... t-,-*.6s-10 REFERENCE DRAWINGS: 1,2,3-45E779-5,-8,-11,-12.. SCHEMATIC DIAGRAM 2-45E614-SERIES............ WIRING DIAGRAMS & MISC SCHEMATIC 2-45E2641-1,-7............. CONNECTION DIAGRAM 2-45E2642-1,-4............. CONNECTION DIAGRAM 2-45E2643-1,-5............. CONNECTION DIAGRAM 2-45N2664-1,-2............. CONNECTION DIAGRAM 2-45N2654-1,-2............. CONNECTION DIAGRAM 2-45N2635-16,-16........... CONNECTION DIAGRAM 2-45E2756-9................ CONNECTION DIAGRAM MEL........................ INSTRUMENT TABULATION 2-47E406-1,2 & 3........... REACTOR WATER CLEANUP SYSTEM PIPING 47W552-SERIES.............. MECHANICAL DIAGRAM 47W600-96.................. EQUIPMENT LOCATION 0-47E600-112............... EQUIPMENT LOCATION 2-47E600-58................ EQUIPMENT LOCATION 2-47E610-1-1............... CONTROL DIAGRAM - MAIN STEAM SYSTEM 1.2.3-47E610-3-1........... CONTROL DIAGRAM - REACTOR FEEDWATER 1,2,J-47E610-43-1.......... CONTROL DIAGRAM - SAMPLING AND WATER QUALITY SYSTEM 2-47E610-68-1,..,,,,,.,,,,.CONTROL DIAGRAM - REACTOR RECIRCN SYS 1,2-47E610-70-1............ CONTROL DIAGRAM - REACTOR BLDG CLOSED COOLING WATER SYSTEM 1,2.3-47E610-71-1.......... CONTROL DIAGRAM - RCIC SYSTEM 2-47E610-73-1.............. CONTROL DIAGRAM - HPCI SYSTEM 2-47E2610-74-1, -2......... CONTROL DIAGRAM - RHR SYSTEM 0,1,2,3-47E610-78-1........ CONTROL DIAGRAM - FUEL POOL COOLING & DEMINERALIZER 0-47EB00-1................. FLOW DIAGRAM - GENERAL PLANT SYSTEMS 0-47E800-2................. MECHANICAL - SYMBOLS AND FLOW DIAGRAM DRAWING INDEX 2-47E810-1................. FLOW DIAGRAM - REACTOR WTR CLEANUP SYS 2-47E817-1................. FLOW DIAGRAM - REACTOR WTR RECIRCN SYS 0-47E830-2................. FLOW DIAGRAM - RADWASTE 0-47E830-5................. FLOW DIAGRAM - RADWASTE 2-47E837-1................. FLOW DIAGRAM - REACTOR WATER CLEANUP DEMINERALIZER 0-47E846-1................. FLOW DIAGRAM - DMNRLZR BACKWASH AIR SYS 1,2,3-47E852-2............. FLOW DIAGRAM - CLEAN RADWASTE AND DECON DRAINAGE 2-47E855-1................. FLOW DIAGRAM - FUEL POOL COOLING SYS 0-47E856-1................. FLOW DIAGRAM - DMNRLZR BACKWASH SYS 2-47E856-2................. FLOW DIAGRAM - DEMINERALIZED WATER 2-47E2847 SERIES........... FLOW DIAGRAM - CONTROL AIR SYSTEM 2-47E610-73-1.............. CONTROL DIAGRAM - HPCI SYSTEM 3-47E610-1-1A.............. CONTROL DIAGRAM - MAIN STEAM SYSTEM 2-47E917-1................. FLOW DIAGRAM - REACTOR WTR RECIRCN SYS 2-47E610-70-1.............. CONTROL DIAGRAM - REACTOR BLDG CLOSED COOLING WATER SYSTEM 0-47E610-77-5, -4.......... CONTROL DIAGRAM - RADWASTE SYSTEM 2-47E610-68-1.............. CONTROL DIAGRAM - REACTOR RECIRCN SYS 2-47E2610-74-1............. CONTROL DIAGRAM - RHR SYSTEM 2-47E611-63-SERIES......... MECHANICAL LOGIC DIAGRAM STANDBY LIQUID CONTROL SYSTEM 2-47E611-64-SERIES......... MECHANICAL LOGIC DIAGRAM - PRIMARY CONTAINMENT ISOLATION SYSTEM 2-47E611-69-1,-2........... MECHANICAL LOGIC DIAGRAM - REACTOR WATER CLEANUP SYSTEM GRAVER WATER CONDITIONING CO. T-12728.............. FLOW DIAGRAM GE DRAWINGS: ~5-3 25-3,C:::,.. { p I ) ( PCV)6 - 02 69-28A 11 FCV ffi----'l"'I_. 69-27 '-~,!---;-l~-t2-47E2847-6.B7 /H5' re~*:@: HOLDING PUMP 2A SAMPLING SYSTEM...._... ...._... k 2-47E61D-4.3-1,H2 ~-, 728E900............. REACTOR WATER CLEANUP SYSTEM 730E719..,,,,..,,... FUNCTIONAL CONTROL DIAGRAM REWASH SLOW DRAIN ~ u w ' ~ v ~ N w N 25-3 s 25-J s 25-J FSV FSV 69-28A 69-288 1 i 5-3 I r, PI II 25-~3 I II 69-M3 I HS I I 69-28 1-L ___ :f : 25-3 FSV 69-28C 'Y I I PI _J69-.5 I I I I l_ _L - _:-_:-_:-_:-_:-_:-_:-_:-_:-_:-_:-..: 8 7 69-39 25-3 69-358 25-3 TO RWCU BW r RECEIVING TANK 0-47E830-5,G4 """--!THIS DWG,C2 I

J I

~

PI 69-506 ~: II 2-47E2847-6.F5 FI ).----------------...---.r------------------------' FI r-_. ._ __.,69-3 6~ 6 TO CHEM WASTE TANK INLET HEADER 2-47EB52-2, CJ 5 FCV 'B----,...-i 9-54µ 25-3 "@-- J 1 1 / ' R _ _J 4 HCV 69-7 2-47E2847-6 .85 3 TO PRECOAT TANK DRAIN 2-47E852-2,G3 0-730E163........... PROCESS DIAGRAM 728E905,,,,,,...,,.. CLEANUP FILTER DEMINERALIZER SYSTEM AMENDMENT POWERHOUSE UNIT 2 28 BROWNS FERRY NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT CLEANUP SYSTEM REACTOR WATER MECHANICAL CONTROL DIAGRAM FIGURE 4.9-3 H G F E D C B A
BFN-16 Figure 4.9-4 Deleted by Amendment 9 ~
i BFN-22 Figures 4.9-4a through 4.9-4d (Deleted by Amendment 22)
zgo~ L-0 l83Lt-£ ~ L9 REACTOR PRIMARY CONTAINMENT 3/4" TESTl

ML19298B510

Contents

Text

SUMMARY

REFERENCES:

T=

REFERENCES:

Navigation menu

ML19298B510

Text

SUMMARY

REFERENCES:

T=

REFERENCES:

Navigation menu

Search